Nuclear Remediation

The 13th International Conference on Environmental Remediation and Radioactive Waste Management

October 3-7, 2010 – Tsukuba, Japan, Epochal Tsukuba

http://www.jsme.or.jp/pes/ICEM10/ http://www.asmeconferences.org/ICEM2010/

FINAL PROGRAM

ASME Message

ICEM2010 is the thirteenth in a series of international conferences on environmental remediation and radioactive waste management organized by ASME, in cooperation with various agencies and other technical societies. While the sponsoring leads are the Environmental Engineering and the Nuclear Engineering Divisions within ASME, it provides a focus on environmental restoration and radioactive waste management for the entire ASME. Founded in 1880 as the American Society of Mechanical Engineers, today's ASME promotes the art, science & practice of mechanical & multidisciplinary engineering and allied sciences around the globe –with >120,000 members worldwide.

Having chaired various ICEM conferences, I learned firsthand that success results only through the concerted effort of many individuals and organizations. If I tried to mention all, I would certainly fail, however, I must acknowledge the dedication of the Conference Secretary, Hiroyoshi Ueda, the continued support of the US Department of Energy, the Technical and Track Chairs, numerous volunteers and organizational sponsors, and most of all, the presenters and program participants. Thank you for being here.

I first served as Conference Chair during ICEM ’01, which was held in Bruges, Belgium, during uncertain times for the world – just a few days after a terrorist attack of the United States. Since then the world has changed and the global environment has become much more personal. My personal message for ICEM’01 was of to collect the best of ideas, and “inclusion”. While not intentional, it became the first of a series of thematic, clear, and simple messages.

ICEM returned to Bruges in 2007, and my message had two parts. I believe that they are even more valid today. First, that Education, Energy, Environment, and the Economy are inextricably intertwined, and therefore the logical development of one is strongly dependent on the health of the others. And second, that Society demands to be actively informed --that is, to be aware of the details that influence their reasonable expectation of an adequate return on their investment in Governance. In today's world, Global Environmental Partnerships and cooperative agreements are essential in order to demonstrate good Governance.

The message for ICEM2010 is “A clean environment sustained by nuclear energy”. It is as simple as that.

Please note that ICEM2011 will be held in Reims (France) on September 2011. A call for papers appears at: icemconf.com/2011CallforPapers/tabid/494/Default.aspx. I look forward to an intensive exchange of ideas, and to the exploration of new partnerships. After all that, the only thing left is to Welcome You to the ICEM2010 Conference on behalf of ASME.

Aníbal L. Taboas

ASME Fellow Technical Co-Chair, ICEM2010 October 2010, Tsukuba, Japan

Objectives and Background

The 13th International Conference on Environmental Remediation and Radioactive Waste Management (ICME2010) is the thirteenth in a series of international conferences organized by the ASME and other technical societies, and being held in Japan after an eleven year interval. ICEM was first held in Hong Kong in 1987, followed by Kyoto, Japan in 1989; Seoul, Korea, in 1991; Prague, Czech Republic, in 1993; Berlin, Germany, in 1995; Singapore in 1997; Nagoya, Japan in 1999; Bruges, Belgium in 2001 and 2007; Oxford, UK in 2003; Glasgow, Scotland in 2005; and Liverpool, UK in 2009.

ICEM promotes broad global exchange of information on technologies, operations, management approaches, economics and public policies in the critical areas of environmental remediation and radioactive waste management. ICEM also provides a unique opportunity to foster cooperation among specialists in mature environmental management programs and those involved in emerging programs. Attendees include scientists and engineers, suppliers and vendors, utilities, regulators government, and others seeking solution to environmental problems. Over 30 countries are generally represented in ICEM conferences. Conference Co-Sponsors

ICEM2010 is co-sponsored by:

The Power and Energy System Division of the Japan Society of Mechanical Engineers (JSME), The Division of Nuclear Fuel Cycle and Environment of the Atomic Energy Society of Japan (AESJ), and The Environmental Engineering and the Nuclear Engineering Divisions of the American Society of Mechanical Engineers (ASME). Supporting Organizations

The following national/international organizations support the conference:

The Japan Atomic Energy Agency Tokyo Electric Power Company Nuclear Waste Management Organization of Japan Radioactive Waste Management Funding and

Research Center Tsukuba City The International Atomic Energy Agency The Canadian Nuclear Society The Chinese Nuclear Society The Korean Nuclear Society The Korean Radioactive Waste Society

Conference Formant

The ICEM2010 technical program includes concurrent sessions in six technical tracks:

• Low/Intermediate-Level Radioactive Waste (L/ILW)

• Spent Fuel, Fissile Material, Transuranic and High-Level Radioactive Waste (SF/TRU/HLW)

• Facility Decontamination and Decommissioning (D&D)

• Environmental Remediation (ER) • Environmental Management, Public

Involvement and Crosscutting Issues (EM/PI) • Global Partnership and Multi-National

Programs (GP)

The technical program consists of an opening plenary session and several parallel program tracks with up to six concurrent sessions. The sessions include 25 minute oral presentations, panels, and poster displays which are designed to enhance dialogue between presenters and participants. The program is divided into 36 sessions conducted over three days. A listing of the specific sessions within each of the six technical program tracks can be found in the section for the technical sessions.

The full registration fee includes meeting materials in the conference bag, entrance to all sessions and the exhibition, the welcome reception in the Sunday evening, lunch in the exhibition area and the banquet in the Tuesday evening. Each registrant will receive the conference proceedings on a CD-ROM mailed after the conference. Conference Venue

ICEM2010 is being held at the Tsukuba International Congress Center, Epochal Tsukuba, which is home to many internationally renowned scientists. City of Tsukuba

The City of Tsukuba is about 50 km northeast of Tokyo and 40 km northwest of Narita Airport. It covers an area of 28,000 ha with a population of

General Information

-1-

about 200,000. To the north lies Mount Tsukuba, 877 meters above sea level. Seen from one angle, Tsukuba city, with its streets of traditional stores and houses, has the look of a rural refuge of abundant greenery. Seen from another angle, it is Tsukuba Science City, a superb modern urban landscape seldom seen in Japanese cities. Conference Check-In

Upon your arrival, please check-in, or on-site register, at the Registration Desk and receive your own badge, i.e. registration pass, and other materials. It is strongly recommended that the conference participants check-in on Sunday, October 3, to avoid the rush before the opening session on Monday morning, October 4. Badges are required for the Welcome Reception in the Sunday evening. The Registration Desk is located in the entrance hall on the ground floor, during the following hours:

Sunday, October 3 16:00 to 19:00 Monday, October 4 7:30 to 17:00 Tuesday, October 5 7:30 to 17:00 Wednesday, October 6 7:30 to 13:00.

Welcome Reception

The Welcome Reception will be held from 18:00, Sunday, October 3, at the conference room 101+102 of Epochal Tsukuba, the conference venue. Daily Morning Briefing for Speaker/Session Co-Chair

All speakers including panelists and session co-chairs are invited to the Moring Briefing for Speaker/Session Co-Chair, on the day of their oral presentation/panel discussion/chairing. Coffee and bread will be served at the briefing. This briefing will give them the time for final arrangements before the oral session. The attendance at the briefing will provide an opportunity for the Session Co-Chairs to meet with the speakers, and for all to discuss the topics they will be addressing. It is essential that all speakers and session co-chairs attend the briefing. The briefing will be held at Rooms 401 through 404 during the following time slots:

Monday, October 4 8:30 to 9:00 Tuesday, October 5 8:00 to 8:30 Wednesday, October 6 8:00 to 8:30.

Lunch and Coffee in Exhibition Area

Lunch will be served in the exhibition area, Multi-Purpose Hall on the ground floor. Lunches

served on Monday, Tuesday and Wednesday will mainly consist of a light meal that can be easily consumed while visiting the exhibits. The full registration fee includes the luncheon price for all three days. If you prefer, there are also several local restaurants near the conference venue. During the breaks, coffee/tea will be served in the exhibition area. Banquet

The Banquet will be held from 18:30, Tuesday, October 5, at Banquet Hall "Subaru" on the ground floor of the annex building of Okura Frontier Hotel Tsukuba. Preparatory PC for Speakers

The preparatory PC will be available in Room 301 for speakers to load and review their PowerPoint slides. PC in each oral session room will be also available at the breaks for speakers to check their slides. Paid Business Services and Internet

Paid Business Services are provided on the ground floor (See “Floor Map of Epochal Tsukuba and Paid Business Services” page.) Wireless Internet is available in the public area. Technical Tour

All foreign participants are required to carry their passports to the tour for the security checks. A full day tour to Tokai-Mura, one of the well-known nuclear sites in Japan, will be held on October 7. The tour will integrate visits to ENTRY, JAEA’s research facility for the geological disposal, and the decommissioning site of a JAPC’s gas-cooled reactor, Tokai-1. (The tour is full.) Exhibition

An industrial exhibition will be held in conjunction with the conference sessions during the following hours:

Monday, October 4 9:00 to 18:00 Tuesday, October 5 9:00 to 18:00 Wednesday, October 6 9:00 to 13:30

The Exhibition Area will be Multi-Purpose Hall on the ground floor.

-2-

Insurance and Liability

All participants are encouraged to make their own arrangements for health and travel insurance. Neither JSME, AESJ, ASME nor their agents can be held responsible for any personal injury, loss, damage, accident to private property or additional expenses incurred because of delays or changes in air, rail, sea, road or other services, strikes, sickness, weather or any other cause. Disclaimer

Neither JSME, AESJ nor ASME can accept any liability for death, injury, or any loss, cost or expense suffered or incurred by any person if such loss is caused or results from the act, default or omission of any person other than an employee or agent of JSME, AESJ or ASME. In particular, neither JSME, AESJ nor ASME can accept any liability for losses arising from the provision or non-provision of services provided by hotel companies or transport operators. Nor can JSME, AESJ nor ASME accept liability for losses suffered by reason of war including threat of war, riot and civil strife, terrorist activity, natural disaster, weather, flood, drought, technical, mechanical or electrical breakdown within any premises visited by delegates and/or partners in connection with the conference, industrial disputes, governmental action, regulations or technical problems which may affect the services provided in connection with the conference. Neither JSME, AESJ nor ASME is able to give any warranty that a particular person will appear as a speaker or panelist.

Opening Session

The opening session will start at 9:30 AM, Monday. Opening/welcome addresses from the conference organizers and plenary speeches from national/international representatives will be held. Each speech will be allocated 35 minutes including questions and answers. Concurrent Oral Sessions

Concurrent sessions will be held in the six technical tracks. The sessions will start at 13:30 PM for Monday and at 9:00 AM for Tuesday and Wednesday. Each presentation will be allocated 25 minutes (20 min. presentation and 5 min. discussion). The session co-chairs can extend the discussion time, keeping the time slot of the session. Poster Sessions

The poster sessions will be held at Room 102 during the following time slots:

Monday, October 4 15:15 to 15:40 17:25 to 17:50 Tuesday, October 5 10:20 to 10:45 15:15 to 15:40

The posters should be set up in Monday AM and torn down by 12:00, Wednesday.

Technical Session Schedule

-3-

AECL Atomic Energy of Canada Limited AESJ Atomic Energy Society of Japan AIST National Institute for Advanced Industrial Science and Technology ANL Argonne National Laboratory ANSTO Australian Nuclear Science and Technology Organisation ASME American Society of Mechanical Engineers BGS British Geological Survey CEA Commissariat à l'énergie atomique CRIEPI Central Research Institute of Electric Power Industry EDF Electricité de France EPRI Electric Power Research Institute FEDRAS Far East Department of Russian Academy of Sciences IAE Institute of Applied Energy IAEA International Atomic Energy Agency JAEA Japan Atomic Energy Agency JAEA/NPSTC JAEA/Nuclear Nonproliferation Science and Technology Center JAPC Japan Atomic Power Company JNES Japan Nuclear Energy Safety Organization JNFL Japan Nuclear Fuel Limited JSME Japan Society of Mechanical Engineers KAERI Korea Atomic Energy Research Institute KAIST Korea Advanced Institute of Science and Technology KHNP Korea Hydro and Nuclear Power Company KNF Korea Nuclear Fuel KRMC Korea Radiactive Waste Management Corporation LANL Los Alamos National Laboratory LBNL Lawrence Berkeley National Laboratory MEXT Ministry of Education, Culture, Sports, Science and Technology NDA Nuclear Decommissioning Authority NEL Nuclear Engineering, Ltd. NUMO Nuclear Waste Management Organization of Japan NWMO Nuclear Waste Management Organization OECD/NEA Organization for Economic Cooperation and Development/Nuclear Energy Agency PNNL Pacific Northwest National Laboratory RWMC Radioactive Waste Management Funding and Research Center SCK•CEN Studiecentrum voor Kernenergie•Centre d'Etude de l'Energie Nucléaire (Belgian

Nuclear Research Centre) SRNL Savannah River National Laboratory UCB University of California, Berkeley US DOE U. S. Department of Energy US NRC U. S. Nuclear Regulatory Commission

List of Organization Acronyms

-4-

Map around Conference Venue, Hotels and Tsukuba Station

Okura Frontier Hotel Tsukuba Epochal

Main

Annex

(Epochal Tsukuba)

Okura Frontier Hotel Tsukuba

-5-

Copy, Printer, FAX

Locker

Mailing

Beverage Vending Machine

Information

Floor Map of Epochal Tsukuba and Paid Business Services

-6-

ICEM2010 Technical Sessions at a Glance

9:30

9:45

9:45 - 12:05

12:10 - 13:30

13:30 Room 405 Room 101 Room 406 Room 202A Room 202B Room 303Preparation Preparation Preparation Preparation Preparation Preparation

13:35

14:00

14:25

14:50Break

Preparation

Preparation Preparation Preparation Preparation Preparation15:45

16:10

16:35

17:00

17:25 - 17:50 Poster Poster Poster Poster Poster Poster

Monday, October 4

Break/PosterBreak/Poster Break/Poster15:15 - 15:40 Break/Poster Break/Poster

Main Convention Hall

Opening Session

Lunch

Opening and Welcome Addresses

Plenary Speeches from Representatives

L1: WasteManagement

H1: Nationaland

InternationalPrograms (1)

H3: Panel"Radwaste

HumanResource

Developmentto Support the

NuclearRenaissance"

R1:Environmental

Impacts

D1: National andInternational

D&D Programs

L1: WasteManagement D1: National

andInternational

D&D Programs

R1:Environmental

Impacts

M1:EnvironmentalManagement

M1:EnvironmentalManagement

H2:Transportation,

Storage andWaste Treatment

H2:Transportation,

Storage andWaste Treatment

L/ILWSF/TRU/HLWD&DEREM/PIGP

-7-


9:00 Room 201 Room 101 Room 202 Room 406 Room 405 Room 303Preparation Preparation Preparation Preparation Preparation Preparation

9:05

9:30

9:55

10:20 - 10:45 Break/Poster Break/Poster Break/Poster Break/Poster Break/Poster Break/Poster

Preparation Preparation Preparation Preparation Preparation Preparation10:50

11:15

11:40

12:05

12:30 - 13:30 Lunch Lunch Lunch Lunch Lunch Lunch


14:00

14:25

14:50

15:15 - 15:40 Break/Poster Break/Poster Break/Poster Break/Poster Break/Poster Break/Poster


16:10

16:35

17:00

17:25

Tuesday, October 5

H4: National andInternationalPrograms (2)

G2: IAEATopical forDisusedSealed

RadioactiveSources(DSRS)

H5: Panel"Advances inKnowledge

Managementfor Radioactive

WasteDisposal"

H4: National andInternationalPrograms (2)

L3: NuclideAssay

H5: Panel"Advances inKnowledge

Managementfor Radioactive

WasteDisposal"

G2: IAEATopical forDisusedSealed

RadioactiveSources(DSRS)

L2: Solidificationand Package (1)


D2: Dismantlingand

Decontamination

D2: Dismantlingand

Decontamination

G1: InternationalCollaboration

D3:Planning

H6: CoupledProcess

Modeling andNatural

Analogues

D3:Planning

H6: CoupledProcess

Modeling andNatural

Analogues

L3: NuclideAssay

R2:EnvironmentalRemediation

R2:EnvironmentalRemediation

D4: Panel"ApplyingLessons

Learned fromPast D&DActivities"

D4: Panel"ApplyingLessons

Learned fromPast D&DActivities"


L5: Recyclingand Clearance


-8-


9:00 Room 201 Room 101 Room 406 Room 405 Room 303Preparation Preparation Preparation Preparation Preparation

9:05

9:30

9:55

10:20 - 10:35 Break Break Break Break Break


11:05

11:30

11:55

12:20 - 13:30 Lunch Lunch Lunch Lunch Lunch


14:00

14:25

14:50

15:15 - 15:30 Break Break

Preparation Preparation15:35

16:00

16:25

16:50

17:15

Wednesday, October 6

L6: WasteTreatment

L7: Storage andDisposal Facility

H7: PerformanceAssessmentModeling andParameters

H9: RepositoryEngineering andDemonstration

D6: WasteTreatment andNon-Reactor

L6: WasteTreatment

H7: PerformanceAssessmentModeling andParameters

D5:Measurementand Estimation

H9: RepositoryEngineering andDemonstration

R3: ERTechniques

M2: PublicInvolvement

R3: ERTechniques

H8: SiteCharacterizationand Modeling of

GeologicalEnvironment (1)







D5:Measurementand Estimation


-9-

Technical Sessions

Monday, October 4, 2010

Monday 09:30 Main Convention Hall --------------------------------------------------------------------------------------- Opening and Welcome Addresses:

Satoru Tanaka, Conference General Chair Masanori Aritomi, Technical Program Chair Anibal L. Taboas, Technical Program Co-Chair, ASME

Plenary Speeches:

Tatsujiro Suzuki, Vice Chairman, Japan Atomic Energy Commission (Japan)

“Nuclear Energy Strategy for Sustainable Growth: Aiming at Green Innovation and Life Innovation”

Dae Y. Chung, Principal Deputy Assistant Secretary for Environmental Management, US DOE (USA)

“U.S. Office of Environmental Management - World Leaders in Nuclear Cleanup and Construction”

Ho Taek Yoon, Senior Vice President, KRMC (Korea Rep.) “Recent Progress in Radioactive Waste Management in

Korea”

Irena Mele, Head of Waste Technology Section, Division of Nuclear Fuel Cycle and Waste Technology, IAEA

“Radioactive waste management – achievements, needs and future expectations”

SESSION L1: Waste Management Monday 13:30 Room 405 --------------------------------------------------------------------------------------- Session Co-Chairs: Miklos Garamszeghy, NWMO (Canada) and Kunihiro Nakai, JGC Corporation (Japan)

1. 40081 – Radioactive Waste: Feedback of 40-year Operations in France Michel Dutzer, Gérald Ouzounian, Roberto Miguez, Jean-Louis Tison, Andra (France)

2. 40226 – The Ethics of the Management of Low and Intermediate Radioactive Wastes Generated by Cernavoda NPP, a Challenge for the Romanian Specialists. Gheorghe Barariu, National Authority for Nuclear Activity - Subsidiary of Technology and Engineering for Nuclear Objectives (Romania)

3. 40149 – CEA's radioactive waste and unused fuel inventory - Marcoule site example Jean-Guy Nokhamzon, Marc Butez, Daniel Fulleringer, CEA (France)

——————— Break ——————— 4. 40258 – Management of historical waste legacy at NRG

Petten Renate de Vos, Nuclear Research and Consultancy Group (Netherlands)

5. 40031 – Norwegian Support for Regulations of Radioactive Waste Management from Uranium Mining And Mill Tailings in Central Asia Tamara Zhunussova, Malgorzata Sneve, Astrid Liland, Norwegian Radiation Protection Authority (Norway); Alexander Kim, Kazakhstan Atomic Energy Committee (Kazakhstan); Ulmas Mirsaidov, Tajikistan Nuclear and Radiation Safety Agency (Tajikistan); Baigabyl Tolongutov, Chui Ecological Laboratory of Kyrgiz Republic (Kyrgiz); Per Strand, Norwegian Radiation

Protection Authority, University of Life Sciences (UMB) (Norway)

SESSION H1: : National and International Programs (1) Monday 13:30 Room 101 ---------------------------------------------------------------------------------------

Session Co-Chairs: Stratis Vomvoris, Nagra (Switzerland) and Hiromi Tanabe, RWMC (Japan)

1. 40097 – Overview of NUMO's policy for implementing safe geological disposal and developing supporting technologies Hiroyuki Tsuchi, Kenichi Kaku, Katsuhiko Ishiguro, Akira Deguchi, Yoshiaki Takahashi, NUMO (Japan)

2. 40150 – Stepwise Site Selection in Switzerland - Sectoral Plan: Status and Outlook Thomas Ernst, Markus Fritschi, Stratis Vomvoris, Nagra (Switzerland)

3. 40084– Site Selection for a Geological Disposal in France: an Approach of Convergence Gérald Ouzounian, Roberto Miguez, Jean-Louis Tison, Andra (France)

SESSION H2: Transportation, Storage and Waste Treatment

Monday 13:30 Room 303 --------------------------------------------------------------------------------------- Session Co-Chairs: Hiroshige Kikura, Tokyo Institute of Technology (Japan) and Anibal L. Taboas, ANL (USA)

1. 40155 – Support of the Nuclear Research Institute Rez plc of the Shipment of Spent Nuclear Fuel from Research Reactors to the Russian Federation for Reprocessing in the Frame of the RRRFR Program Josef Podlaha, Karel Svoboda, Nuclear Research Institute Rez plc (Czech Rep.)

2. 40177 – Long Term Storage of Nuclear Spent Fuel as Key Role of Japan's Nuclear Fuel Cycle until 2100: Cost and Benefit Tadahiro Katsuta, Meiji University (Japan)

3. 40285 – TRU Recycling Options for Environmentally Friendly and Proliferation-Resistant Nuclear Fuel Cycle of FBR Sidik Permana, Mitsutoshi Suzuki, JAEA (Japan)

4. 40132 – Transuranic (TRU) Waste Volume Reduction Operation at a Plutonium Facility Michael Cournoyer, Archie E. Nixon, Keith W. Fife, Arnold M. Sandoval, Vincent E. Garcia, Robert L. Dodge, LANL (USA)

——————— Break ——————— 5. 40188 – A Milestone in Vitrification - the Replacement of

a Hot Metallic Crucible with a Cold Crucible Melter in a Hot Cell at the La Hague Plant Sophie Robert, Florence Gassot Guilbert, Benoit Carpentier, SGN (France); Sandrine Naline, AREVA group (France); Frédéric Gouyaud, AREVA NC (France); Christophe Girold, CEA (France)

6. 40265 – Adaptation of CCIM technology for HLW treatment. Results of research and development Vladimir Lebedev, Sergey Stefanovsky, Alexander Kobelev, Fyodor Lifanov, Sergey Dmitriev, SIA Radon (Russia)

OPENING SESSION

-10-

Technical Sessions

SESSION H3: Panel "Radwaste Human Resource Development to Support the Nuclear Renaissance"

Monday 15:15 Room 101 --------------------------------------------------------------------------------------- Session Chair: Ian G. McKinley, McKinley Consulting (Switzerland) The panel focuses on answering three questions that form the basis for establishing an efficient and effective human resource development plan for radioactive waste management:

- what resources do we need and when do we need them? - how can we assure that these resources will be available

(with emphasis on training requirements, especially for generalists and multidisciplinary coordinators)?

- what infrastructure and procedures need to be made available with high priority now?

Panelists: Joonhong Ahn, UCB (USA) Tomio Kawata, NUMO (Japan) Ian G. McKinley, McKinley Consulting (Switzerland) Irena Mele, IAEA Shawn Smith, US NRC (USA)

SESSION D1 : National and International D&D Programs

Monday 13:30 Room 406 --------------------------------------------------------------------------------------- Session Co-Chairs: Sean Bushart, EPRI (USA) and Satoshi Yanagihara, JAEA (Japan)

1. 40003 – Westinghouse PWR and BWR Reactor Vessel Segmentation Experience in Using Mechanical Cutting Process Per Segerud, Stefan Fallström, Westinghouse Electric Sweden (Sweden); Joseph Boucau, Westinghouse Electric Belgium (Belgium); Paul Kreitman, Westinghouse Electric Company (USA)

2. 40130 – EPRI Nuclear Power Plant Decommissioning Technology Program Karen Kim, Sean Bushart, Mike Naughton, Richard McGrath, Electrict Power Research Institute (USA)

3. 40253 – Tokai-1 Decommissioning Project, the First Challenge in Japan Keizaburou Yoshino, JAPC (Japan)

4. 40289 – Activities of the OECD/NEA WPDD in the Field of Decommissioning Claudio Pescatore, Patrick O'Sullivan, OECD/NEA

—————— Break ——————— 5. 40307 – French Decommissioning Feedback Experience

and Lessons Learned Jean-Guy Nokhamzon, CEA (France); Patrick O'Sullivan, OECD/NEA

6. 40032 – Decommissioning Planning for Swedish Operating NPPs Gunnar Hedin, Mathias Edelborg, Niklas Bergh, Westinghouse Electric Sweden (Sweden); Jan Carlsson, Fredrik de la Gardie, SKB (Sweden)

7. 40273 – Considerations for Grout Formulations in Facility Closures Using In Situ Strategies Michael J. Serrato, Christine A. Langton, John B. Gladden, SRNL (USA); John T. Long, John K. Blankenship, Savannah River Nuclear Solutions (USA), Andrew P. Szilagyi, George R. Hannah, Rita B. Stubblefield, US DOE (USA)

SESSION R1: Environmental Impacts

Monday 13:30 Room 202A --------------------------------------------------------------------------------------- Session Co-Chairs: Shinzo Ueta, Mitsubishi Materials Corporation (Japan) and Dawn Wellman, PNNL (USA)

1. 40298 – Main Results of A Remediation of Uranium- and CHC-Contaminated Groundwater Jörg Wörner, Sonja Margraf, Walter Hackel, RD-Hanau (Germany)

2. 40267 – Biogeochemical Gradients, Waste Site Evolution, and Implications for Sustained Metal and Radionuclide Attenuation in Complex Subsurface Environments Karen Skubal, Justin Marble, Kurt D. Gerdes, US DOE (USA); Miles Denham, Karen Vangelas, SRNL (USA)

3. 40260 – Current Mercury Distribution and Bioavailability in Floodplain Soils of Lower East Fork Popular Creek, Oak Ridge, Tennessee, USA Fengxiang X. Han, Yi Su, David L. Monts, Mississippi State University (USA)

4. 40262 – Integrated Strategy to Address Hanford's Deep Vadose Zone Remediation Challenges Mark B. Triplett, Mark D. Freshley, Michael J. Truex, Dawn M. Wellman, PNNL (USA); Kurt D. Gerdes, Briant L. Charboneau, John G. Morse, Robert W. Lober, US DOE (USA); Glen B. Chronister, CH2M Hill Plateau Remediation Company (USA)

—————— Break ——————— 5. 40235 – Advanced Remedial Methods for Metals and

Radionuclides in Vadose Zone Environments Dawn M. Wellman, Shas V. Mattigod, Ann Miracle, Lirong Zhong, Danielle Jansik, PNNL (USA); Susan Hubbard, Yuxin Wu, LBNL (USA); Martin Foote, MSE Technology Applications (USA)

SESSION M1: Environmental Management Monday 13:30 Room 202B --------------------------------------------------------------------------------------- Session Co-Chairs: Motoi Kawanishi, CRIEPI (Japan) and Tadao Tanaka, JAEA (Japan)

1. 40086 – Legacy Management: Turning Liabilities into Assets Joe Legare, Eric Olson, S.M. Stoller Corporation (USA)

2. 40270 – Lessons Learned in Planning the Canadian Nuclear Legacy Liabilities Program Michael E. Stephens, Sheila M. Brooks, Joan M. Miller, Robert A. Mason, AECL (Canada)

3. 40218 – RFID Technology for Environmental Remediation and Radioactive Waste Management Hanchung Tsai, Yung Y. Liu, ANL (USA); James Shuler, US DOE (USA)

4. 40181 – The Radioactivity of 3H in Metals by a High Temperature Furnace and a Liquid Scintillation Counter Hee Reyoung Kim, Geun Sik Choi, Sang Yun Park, Chang Woo Lee, Moon Hee Han, KAERI (Korea Rep.)

—————— Break ——————— 5. 40275 – Next Generation Waste Glass Melters in the U.S.

DOE Waste Processing Program Gary L. Smith, Steven P. Schneider, Kurt D. Gerdes, US DOE (USA)

-11-

Technical Sessions

Tuesday, October 5, 2010

SESSION L2: Solidification and Package (1) Tuesday 09:00 Room 201 --------------------------------------------------------------------------------------- Session Co-Chairs: Miklos Garamszeghy, NWMO (Canada) and Yoshihiko Horikawa, NEL (Japan)

1. 40021 – Commercialization Project of Ulchin Vitrification Hyun-jun Jo, Cheon-Woo Kim, Tae-Won Hwang, KHNP (Korea Rep.)

2. 40023 – Plasma Gasification/Vitrification of Wet ILW Gary Hanus, John Williams, Matt Zirbes, Phoenix Solutions Co. (USA)

3. 40026 – Solidification of Simulated Liquid Waste of Primary Loop Resin Elution Process of PWR Masamichi Obata, Masaaki Kaneko, Michitaka Saso, Nobuhito Ogaki, Taichi Horimoto, Toshiba corporation, Toshikazu Waki, Kansai Electric Power Company (Japan)

—————— Break ——————— 4. 40108 – VUJE Experience with Cementation of Liquid

and Wet Radioactive Waste Kamil Kravárik, .Zuzana Holická, Anton Pekár, Milan Žatkulák, VUJE, Inc. (Slovakia)

5. 40112 – Study of LPOP residue on resin mineralization and solidification Gen-ichi Katagiri, Morio Fujisawa, Fuji Electric Systems Co., Ltd. (Japan); Kazuya Sano, Norikazu Higashiura, JAEA (Japan)

SESSION L3: Nuclide Assay

Tuesday 09:00 Room 303 --------------------------------------------------------------------------------------- Session Co-Chairs: David James, DW James Consulting (USA) and Kunihiro Nakai, JGC Corporation (Japan)

1. 40167 – Feasibility Study on the Nuclide Analysis of the Radwaste Drum Using the Spectrum to Dose Conversion Factor Young-Yong Ji, Dae-Seok Hong, Tae-Kuk Kim, Woo-Seog Ryu, KAERI (Korea Rep.)

2. 40255 – Portable Non-Destructive Assay Methods for Screening and Segregation of Radioactive Waste Alan Simpson, Stephanie Jones, Martin Clapham, Randy Lucero, Pajarito Scientific Corporation (UK)

3. 40093 – Alpha Radioactivity Monitor Using Ionized Air Transportation for Large Size Uranium Waste (1) - Large Measurement Chamber and Evaluation of Detection Performance - Susumu Naito, Shuji Yamamoto, Mikio Izumi, Yosuke Hirata, Yukio Yoshimura, Tatsuyuki Maekawa, Toshiba Corporation (Japan)

—————— Break ——————— 4. 40091 – Alpha Radioactivity Monitor Using Ionized Air

Transport Technology for Large Size Uranium Waste (2) - Simulation model reinforcement for practical apparatus design - Takatoshi Asada, Yosuke Hirata, Susumu Naito, Mikio Izumi, Yukio Yoshimura, Toshiba Corpoation (Japan)

5. 40111 – Preparation of Reference Materials on Radiochemical Analysis for Low-Level Radioactive Waste Generated from Japan Atomic Energy Agency Ken-ichiro Ishimori, Yutaka Kameo, Mikio Nakashima, Kuniaki Takahashi, JAEA (Japan)

SESSION H4: National and International Programs (2)

Tuesday 09:00 Room 101 --------------------------------------------------------------------------------------- Session Co-Chairs: Joonhong Ahn, UCB (USA) and Kenichi Kaku NUMO (Japan)

1. 40213 – U.S. NRC Integrated Spent Fuel Management Plan Catherine Haney, Shawn Smith, US NRC (USA)

2. 40116 – Regulatory Research for Geological Disposal of High-level Radioactive Waste in Japan Shinichi Nakayama, JAEA (Japan); Yoshio Watanabe, AIST (Japan); Masami Kato, JNES (Japan)

3. 40280 – Recent Developments and Trends in Requirements Management Systems Satoru Suzuki, Hiroyoshi Ueda, Kiyoshi Fujisaki, Katsuhiko Ishiguro, Hiroyuki Tsuchi, NUMO (Japan); Stratis Vomvoris, Irina Gaus, Nagra (Switzerland)

——————— Break ——————— 4. 40228 – Development of Requirements Management

System of NUMO and practical experience with development of the database contents Satoru Suzuki, Hiroyoshi Ueda, Kiyoshi Fujisaki, Katsuhiko Ishiguro, Hiroyuki Tsuchi, NUMO (Japan); Kiyoshi Oyamada, JGC Corporation (Japan); Shoko Yashio, Obayashi Corporation (Japan)

5. 40231 – Application of Lifecycle Management to Design of the UK Geological Disposal Facility Henry O'Grady, Malcolm Currie, Parsons Brinckerhoff (UK); Philip Rendell, NDA (UK)

SESSION D2: Dismantling and Decontamination

Tuesday 09:00 Room 202 --------------------------------------------------------------------------------------- Session Co-Chairs: Jean-Guy Nokhamzon, CEA (France) and Motonori Nakagami, Chubu Electric Power Company (Japan)

1. 40036 – Experience in Dismantling and Packaging of Pressure Vessel and Core Internals Peter Pillokat, Jan Hendrik Bruhn, AREVA NP GmbH (Germany)

2. 40102 – Study on evaluation models of management data for decommissioning of Fugen Yuji Shibahara, Masanori Izumi, Takashi Nanko, Mitsuo Tachibana, Tsutomu Ishigami, JAEA (Japan)

3. 40083 – AREVA NP Decontamination Concept for Decommissioning - A Comprehensive Approach Based on Over 30 Years Experience Christoph Stiepani, AREVA NP GmbH (Germany)

——————— Break ——————— 4. 40007 – Chemical Decontamination for

Decommissioning (DFD) and DFDX Ronald Morris, Westinghouse Electric Company (USA)

5. 40127 – Methods for Calculation and Optimisation of Personnel Exposure during Planning of Decommissioning of Nuclear Installation Marek Vaško, Vladimír Daniška, Ivan Rehák, DECOM, a.s. (Slovakia); Vladimír Nečas, Slovak University of Technology in Bratislava (Slovakia)

-12-

Technical Sessions

SESSION D3: Planning

Tuesday 09:00 Room 405 --------------------------------------------------------------------------------------- Session Co-Chairs: Joseph Boucau, Westinghouse Electric Belgium (Belgium) and Toshihiko Higashi, Kansai Electric Power Company (Japan)

1. 40129 – Program Change Management During Nuclear Power Plant Decommissioning Sean Bushart, Karen Kim, Mike Naughton, EPRI (USA)

2. 40245 – Status of the Support Researches for the Regulation of Nuclear Facilities Decommissioning in Japan Yusuke Masuda, Yukihiro Iguchi, Satoru Kawasaki, Masami Kato, JNES (Japan)

3. 40136 – Decommissioning Costing Approach Based on the Standardised List of Costing Items; Lessons Learnt by the OMEGA Computer Code Vladimír Daniška, Ivan Rehák, Marek Vaško, František Ondra, Peter Bezák, Jozef Prítrský, DECOM a.s. (Slovakia); Matej Zachar, Vladimír Nečas, Slovak University of Technology in Bratislava (Slovakia)

—————— Break ——————— 4. 40290 – The Outline of Decommissioning Plan for

Hamaoka Nuclear Power Station Unit-1 and Unit-2 Yoshifusa Fukuoka, Chubu Electric Power Company (Japan)

5. 40015 – Study on Influence of Nuclear Fuel Material Management and Transfer Scenarios on Decommissioning Kazuma Mizukoshi, NEL (Japan)

6. 40100 – Dose Assessment for setting of EPZ in Emergency Plan for Decommissioning of Nuclear Power Plant Hirokazu Minato, Hitachi-GE Nuclear Energy (Japan); Takatoshi Hattori, CRIEPI (Japan); Toshihiko Higashi, Kansai Electric Power Company (Japan); Takehiro Iwata, JAPC (Japan)

SESSION G1 : International Collaboration

Tuesday 09:00 Room 406 --------------------------------------------------------------------------------------- Session Co-Chairs: Hiromi Tanabe, RWMC (Japan) and Robin Heard, IAEA

1. 40118 – Advancing the Use of IAEA Networks in Radioactive Waste Management: Past Successes, Present Challenges and Future Opportunities. Paul Degnan, John Kinker, Irena Mele, Paul J. Dinner, Horst Monken Fernandes, Antonio Morales, Lumir Nachmilner, Shaheed Hossain, IAEA

2. 40287 – The activities of the OECD/NEA RWMC in the Field of HLW and SF disposal Claudio Pescatore, OECD/NEA

3. 40147 – Grimsel Test Site - Phase VI Status and Outlook Ingo Blechschmidt, Sven Peter Teodori, Stratis Vomvoris, Nagra (Switzerland)

SESSION L4: Solidification and Package (2) Tuesday 13:30 Room 201 --------------------------------------------------------------------------------------- Session Co-Chairs: Gérald Ouzounian, Andra (France) and Masamichi Obata, Toshiba Corporation (Japan)

1. 40299 – Treatment of low level radioactive waste by plasma: a proven technology? Jan Deckers, Belgoprocess N.V. (Belgium)

2. 40128 – The Zwilag Plasma Facility - Five Years of Successful Operation Walter Heep, Zwilag Interim Storage Facility (Switzerland)

3. 40293 – Safety Assessment of Disposal Container for Higher Activity Low Level Waste Motonori Nakagami, Seiji Komatsuki, Chubu Electric Power Company (Japan); Kyosuke Fujisawa, Takashi Nishio, Kobe Steel, Ltd. (Japan); Thomas Quercetti, André Musolff, Karsten Müller, Federal Institute for Materials Research and Testing (Germany)

SESSION L5: Recycling and Clearance

Tuesday 15:40 Room 201 --------------------------------------------------------------------------------------- Session Co-Chairs: Kapila Fernando, ANSTO (Australia) and Kunihiro Nakai, JGC Corporation (Japan)

1. 40223 – NPP Bulk Equipment Dismantling Problems and Experience Alexander B. Gelbutovsky, Peter I. Cheremisin, Yuri A. Epikhin, Alexander V. Troshev, Eugeny V. Balushkin, ECOMET-S JSC (Russia)

2. 40073 – Reuse of Conditionally Released Radioactive Materials from NPP Decommissioning Applied in Motorway Bridges Construction Michal Pánik, Tomáš Hrnčíř, Vladimír Nečas, Slovak University of Technology in Bratislava (Slovakia)

3. 40071 – Modelling of Motorway Tunnels Scenario for Utilization of Conditionally Released Radioactive Materials Tomáš Hrnčíř, Michal Pánik, Vladimír Nečas, Slovak University of Technology in Bratislava (Slovakia)

4. 40117 – Estimate of Clearance Levels for Metal Materials Contaminated with Uranium Seiji Takeda, Hideo Kimura, JAEA (Japan)

SESSION H5: Panel "Advances in Knowledge Management for Radioactive Waste Disposal"

Tuesday 13:30 Room 101 --------------------------------------------------------------------------------------- Session Co-Chairs: Hiroyuki Umeki, Hitoshi Makino, JAEA (Japan) and Ian G. McKinley, McKinley Consulting (Switzerland) Part 1: Status and plans of KM activities in national programmes Information exchange with emphasis on identification of potential areas for collaboration. Panelists represent implementing organisations, regulatory authorities and R&D organisations, thus bringing different perspectives on KM. Part 2: Brainstorming on advanced KM tools A common requirement in all geological disposal programmes is efficiently and rigorously managing increasing large and complex fluxes of information. Emphasis will be on suggesting practical applications and further improvements of advanced KM tools that have been developed by JAEA, distinguishing between programme-specific constraints and more generic areas, which could be a focus for future collaborative projects. Panelists:

Johan Andersson, JA Streamflow (Sweden) Kenzi Karasaki, LBNL (USA) Masami Kato, JNES (Japan) Lawrence Kokajko, US NRC (USA) Mark Nutt, ANL (USA) Richard Shaw, BGS (UK)

-13-

Technical Sessions

Hiroyuki Tsuchi, NUMO (Japan) Hiroyuki Umeki, JAEA (Japan)

SESSION H6: Coupled Process Modeling and Natural Analogues

Tuesday 13:30 Room 405 --------------------------------------------------------------------------------------- Session Co-Chairs: Irina Gaus, Nagra (Switzerland) and Gento Kamei, JAEA (Japan)

1. 40306 Keynote – A Discussion of Key Issues in Coupled THM Processes in Clays, Rock Salt and Crystalline Rock with Bentonite Buffer Chin-Fu Tsang, LBNL (USA)

2. 40159 – Environmental Remediation of High-Level Nuclear Waste in Geological Repository: Modified Computer Code Creates Ultimate Benchmark in Natural Systems Geoffrey J. Peter, Oregon Institute of Technology Portland Center (USA)

3. 40196 – Effect of the Residual Heat Release of the Nuclear Waste Stored in an Unsaturated Zone on Radionuclide Release Lubna K. Hamdan, John C. Walton, Arturo Woocay, University of Texas at El Paso (USA)

4. 40072 – Evaluation of behavior of rare earth elements based on determination of chemical state in groundwater in granite Yuhei Yamamoto, Daisuke Aosai, Takashi Mizuno, JAEA (Japan)

—————— Break ——————— 5. 40022 – Natural analogue studies of bentonite reaction

under hyperalkaline conditions: overview of ongoing work at the Zambales Ophiolite, Philippines Naoki Fujii, M. Yamakawa, RWMC (Japan); K. Namiki, Obayashi Corporation (Japan); T. Sato, Hokkaido University (Japan); N. Shikazono, Keio University (Japan); C. A. Arcilla, C. Pascua, University of the Philippines (Philippines); W. Russell Alexander, Bedrock Geosciences (Switzerland)

6. 40063 – Natural Analogues of Cement: Overview of the Unique Systems in Jordan Gento Kamei, JAEA (Japan); W. Russell Alexander, Bedrock Geosciences (Switzerland); Ian D. Clark, University of Ottawa (Canada); Paul Degnan, IAEA; Marcel Elie, Shell (Netherlands); Hani Khoury, Elias Salameh, University of Jordan (Jordan), Antoni E. Milodowski, BGS (UK), Alister F. Pitty, Pitty Consulting (UK); John A.T. Smellie, Conterra (Sweden)

SESSION D4: Panel "Applying Lessons Learned from Past D&D Activities"

Tuesday 13:30 Room 202 --------------------------------------------------------------------------------------- Session Co-Chairs: Koji Okamoto, University of Tokyo (Japan) and Claudio Pescatore, OECD/NEA In this panel, international lessons learned from past D&D activities will be discussed to optimize future decommissioning activities. Each panelist will talk about some of topics below. Discussion will be focused on two or three topics. Later in the panel, a recent TV report in Japan will be introduced to discuss mass media issue, too.

a. Decommissioning Project Management b. Decommissioning Techniques c. Waste Disposal d. Building and Site Remediation

e. Social Issues and Others Panelists:

Koji Okamoto, Keynote, University of Tokyo (Japan) Sean Bushart, EPRI (USA) Satoshi Karigome, JAPC (Japan) Jean-Guy Nokhamzon, CEA (France) Claudio Pescatore, OECD/NEA Andrew P. Szilagyi, US DOE (USA) Satoshi Yanagihara, JAEA (Japan)

SESSION R2: Environmental Remediation

Tuesday 13:30 Room 303 --------------------------------------------------------------------------------------- Session Co-Chairs: Hirofumi Tsukada, Institute for Environmental Sciences (Japan) and Jörg Wörner, RD-Hanau (Germany)

1. 40261 – Reclamation of Three In Situ Uranium Mines - Innovative Techniques Wallace Mays, W M Mining Company (USA)

2. 40005 – Environmental remediation Activities at the Ningyo-toge Uranium Mine, Japan Hiroshi Saito, Tomihiro Taki, JAEA (Japan)

3. 40092 – Radon impact at a remediated uranium mine site in Japan Yuu Ishimori, JAEA (Japan)

4. 40243 – Phosphate based remediation techniques: interaction of phosphate with uranium-bound calcite Chase Bovaird, Dawn Wellman, PNNL (USA)

—————— Break ——————— 5. 40220 – Remediation of Old Environmental Liabilities in

the Nuclear Research Institute Rez plc Karel Svoboda, Josef Podlaha, Nuclear Research Institute Rez plc (Czech Rep.)

SESSION G2: IAEA Topical for Disused Sealed Radioactive Sources (DSRS)

Tuesday 13:30 Room 406 --------------------------------------------------------------------------------------- Session Co-Chairs: Hiromi Tanabe, RWMC (Japan) and Irena Mele, IAEA

1. 40028 – International initiatives addressing the safety and security of Disused Sealed Radioactive Sources (DSRS) Robin Heard, IAEA

2. 40303 – Current Situation and Management Plan of Radioactive Sources in Japan Hirokuni Ito, Tadashi Ishii, Tomokazu Ueta, Takao Nakaya, Kenya Suyama, MEXT (Japan)

3. 40060 – The Deployment of the Mobile Hot Cell to Condition High Activity Disused Sealed Radioactive Sources (DSRS) for Long Term Storage or Removal Gerhardus R. Liebenberg, South African Nuclear Energy Corporation (Necsa) (South Africa)

4. 40266 – Problems with Packaged Sources in Foreign Countries James Matzke, John Zarling, Cristy Abeyta, Joseph A. Tompkins, LANL (USA)

—————— Break ——————— 5. 40085 – The IAEA's approach to the security of

radioactive material Robin Heard, IAEA

6. 40058 – Radioactive Waste Management in Lebanon Munzna Assi, Lebanese Atomic Energy Commission (Lebanon)

7. 40029 – The Ultimate Solution – Disposal of Disused

-14-

Technical Sessions

Sealed Radioactive Sources (DSRS) Robin Heard, IAEA

Wednesday, October 6, 2010

SESSION L6: Waste Treatment Wednesday 09:00 Room 201 --------------------------------------------------------------------------------------- Session Co-Chairs: Gérald Ouzounian, Andra (France) and Hirokazu Tanaka, Mitsubishi Materials Corporation (Japan)

1. 40055 – Drying System For Radioactivated Metal Waste from Nuclear Power Station Nobuhito Ogaki, Yasushi Ooishi, Hironori Takabayashi, Masamichi Obata, Taichi Horimoto, Toshiba Corporation (Japan)

2. 40186 – Macroporous Catalysts for Hydrothermal Oxidation of Metallorganic Complexes at Liquid Radioactive Waste Treatment Valentin Avramenko, Vitaly Mayorov, Dmitry Marinin, Alexander Mironenko, Marina Palamarchuk, Valentin Sergienko, Institute of Chemistry FEDRAS (Russia)

3. 40163 – Impermeable Graphite: a New Development for Embedding Radioactive Waste Johannes Fachinger, Karl-Heinz Grosse, Furnances Nuclear Applications Grenoble (Germany); Richard Seemann, Milan Hrovat, ALD (Germany)

—————— Break ——————— 4. 40165 – THOR® Steam Reforming Technology for the

Treatment of Ion Exchange Resins and More Complex Wastes such as Fuel Reprocessing Wastes J. Brad Mason, Corey Myers, Studsvik, Inc. (USA)

5. 40257 – Phase Behavior and Reverse Micelle Formation in Supercritical CO2 with DTAB and F-pentanol for Decontamination of Radioactive Wastes Kensuke Kurahashi, Osamu Tomioka, Yoshihiro Meguro, JAEA (Japan)

SESSION H7: Performance Assessment Modeling and Parameters

Wednesday 09:00 Room 101 --------------------------------------------------------------------------------------- Session Co-Chairs: Hiroyuki Umeki, JAEA (Japan) and Masaki Tsukamoto, CRIEPI (Japan)

1. 40305 Keynote – Development of a Realistic Repository Performance Assessment Method Joonhong Ahn, UCB (USA)

2. 40204 – Integrated model for the near field of a repository in granite host-rock. Probabilistic approach Lara Duro, Alba Valls, Olga Riba, Jordi Bruno, Amphos XXI Consulting S.L. (Spain); Aurora Martinez-Esparza, ENRESA (Spain)

3. 40017 – Spatial Variability and Parametric Uncertainty in Performance Assessment Models Osvaldo Pensado, James Mancillas, Scott Painter, Southwest Research Institute (USA); Yasuo Tomishima, AIST (Japan)

—————— Break ——————— 4. 40203 – Development of a Radiolytic Model for the

Alteration of Spent Nuclear Fuel. Incorporation of non-oxidative matrix dissolution and hydrogen oxidation inhibition effect Lara Duro, Alba Valls, Olga Riba, Jordi Bruno, Amphos

XXI Consulting S.L.(Spain); Aurora Martinez-Esparza, ENRESA (Spain)

5. 40172 – Evaluated and Estimated Solubility of Some Elements for Performance Assessment of Geological Disposal of High-level Radioactive Waste Using Updated Version of Thermodynamic Database Akira Kitamura, Reisuke Doi, JAEA (Japan); Yasushi Yoshida, Inspection Development Co., Ltd. (Japan)

6. 40049 – Consideration on Soil Origin Carbon Transfer to Leafy Vegetables Using Stable Carbon Isotope Ratios Keiko Tagami, Shigeo Uchida, National Institute of Radiological Sciences (Japan)

7. 40050 – Comparison of Soil-to-plant Transfer Factors for Rice and Wheat Grains Shigeo Uchida, Keiko Tagami, National Institute of Radiological Sciences (Japan)

SESSION H8: Site Characterization and Modeling of Geological Environment (1)

Wednesday 09:00 Room 405 --------------------------------------------------------------------------------------- Session Co-Chairs: Chin-Fu Tsang, LBNL (USA) and Motoi Kawanishi, CRIEPI (Japan)

1. 40121 – Development of Characterization Methodology for Fault Zone Hydrology Kenzi Karasaki, LBNL (USA); Celia Tiemi Onishi, US Geological Survey (USA); Erika Gasperikova, LBNL (USA) Junichi Goto, Tadashi Miwa, Hiroyuki Tsuchi, NUMO (Japan); Keiichi Ueta, Kenzo Kiho, Kimio Miyakawa, CRIEPI (Japan)

2. 40189 – An attempt to evaluate horizontal crustal movement by geodetic and geological approach in the Horonobe area, northern Hokkaido, Japan Tetsuya Tokiwa, Koichi Asamori, Tadafumi Niizato, Tsuyoshi Nohara, JAEA (Japan); Yuki Matsuura, Hitachi Zosen Corporation (Japan); Hideki Kosaka, Kankyo Chishitsu Co., Ltd. (Japan)

3. 40054 – Relationship between hypocentral distribution and geological structure in the Horonobe area, northern Hokkaido, Japan Tetsuya Tokiwa, Koichi Asamori, Naoto Hiraga, Osamu Yamada, Hideharu Yokota, JAEA (Japan); Hirokazu Moriya, Tohoku University (Japan); Hikaru Hotta, Itaru Kitamura, Construction Project Consultants, Inc. (Japan)

—————— Break ——————— 4. 40062 – Technical Know-How for Modeling of

Geological Environment (1) Overview and Groundwater Flow Modeling Hiromitsu Saegusa, Shinji Takeuchi, Keisuke Maekawa, Hideaki Osawa, Takeshi Semba, JAEA (Japan)

5. 40066 – Technical Know-How for Modeling of Geological Environment (2) Geological Modeling Toshiyuki Matsuoka, Kenji Amano, Hideaki Osawa, Takeshi Semba, JAEA (Japan)

6. 40039 – The long-term stability of geological environments in the various rock types in Japan from the perspective of uranium mineralization Eiji Sasao, JAEA (Japan)

SESSION D5: Measurement and Estimation

Wednesday 09:00 Room 406 --------------------------------------------------------------------------------------- Session Co-Chairs: Vladimír Daniška, DECOM, a.s. (Slovakia) and Yukihiko Iguchi, JNES (Japan)

1. 40045 – Improvement of Radioactivity Inventory

-15-

Technical Sessions

Evaluation Procedure In Preparatory Tasks for Decommissioning Ken-ichi Tanaka, Hideaki Ichige, JAPC (Japan); Hidenori Tanabe, JNFL (Japan)

2. 40202 – Verification of Source Term Analysis System for Decommissioning Wastes from a CANDU Reactor Dong-Keun Cho, Gwang-Min Sun, Jongwon Choi, KAERI (Korea Rep.); Donghyeun Hwang, Hak-Soo Kim, Tae-Won Hwang, KHNP (Korea Rep.)

3. 40294 – Evaluation of The Activated Radioactivity of Turbine Equipments in BWR Masato Watanabe, Motonori Nakagami, Chubu Electric Power Company (Japan)

—————— Break ——————— 4. 40014 – Optimization of Quantitative Waste Volume

Determination Technique for Hanford Waste Tank Closure Yi Su, David L. Monts, Ping-Rey Jang, Zhiling Long, Walter P. Okhuysen, Olin P. Norton, Lawrence L. Gresham, Jeffrey S. Lindner, Mississippi State University (USA)

5. 40120 – Implementation of Decommissioning Materials Conditional Clearance Process to the OMEGA Calculation Code Matej Zachar, Vladimír Nečas, Slovak University of Technology in Bratislava (Slovakia); Vladimír Daniška, DECOM, a.s. (Slovakia)

6. 40183 – Quantitative determination of the initial components in the activated pressure tubes of the Wolsong 1st CANDU reactor Gwang-Min Sun, Dong-Keun Cho, KAERI (Korea Rep.)

SESSION R3: ER Techniques

Wednesday 09:00 Room 303 --------------------------------------------------------------------------------------- Session Co-Chairs: Takumi Kubota, Kyoto University Research Reactor Institute (Japan) and Mark B. Triplett, PNNL (USA)

1. 40286 – Sequential Extraction and Determination of Depleted Uranium in the Presence of Natural Uranium in Environmental Soil samples by ICP-MS Mohamed Amr, Alaa E. Negmeldin, Khalid Al-Saad, A. T. Al-Kinani, Qatar University (Qatar); A. I. Helal, Atomic Energy Authority (Qatar)

2. 40096 – Determination of Environmental Uranium Concentration by Utilizing Gamma-Ray Emission from the Progeny Radionuclides Tadao Tanaka, Taro Shimada, Takenori Sukegawa, JAEA (Japan); Takeshi Ito, Japan ATOX Co., Ltd. (Japan)

3. 40034 – Effect of Fertilizer and Soil Amendments on Extraction Yields of Radioiodine and Radiocesium in Soil Hirofumi Tsukada, Akira Takeda, Shunichi Hisamatsu, Institute for Environmental Sciences (Japan)

—————— Break ——————— 4. 40246 – Impact of Mobile-Immobile Water Domains on

the Retention of Technetium (Tc-99) in the Vadose Zone Danielle Jansik, Dawn Wellman, Elsa Cordova, PNNL (USA)

5. 40122 – Remediation of 153Gd-contaminated sand by fulvic and humic materials extracted from fallen cherry leaves Takumi Kubota, Kyoto University Research Reactor Institute (Japan)

SESSION L7: Storage and Disposal Facility

Wednesday 13:30 Room 201 --------------------------------------------------------------------------------------- Session Co-Chairs: Sheila M. Brooks, AECL (Canada) and Mamoru Kumagai, JNFL (Japan)

1. 40284 – Microbial Occurrence in Bentonite-Based Buffer Materials of a Final Disposal Site for Low Level Radioactive Waste in Taiwan Fong-In Chou, Chia-Chin Li, Tzung-Yuang Chen, National Tsing Hua University (Taiwan); Hsiao-Wei Wen, National Chung Hsing University (Taiwan)

2 40153 – Assessing the gas transport mechanisms in the Swiss L/ILW concept using numerical modeling Irina Gaus, Paul Marschall, Joerg Rueedi, Nagra (Switzerland); Rainer Senger, John Ewing, Intera Inc. Swiss Branch (Switzerland)

3. 40283 – The Progress and Results of a Demonstration Test of a Cavern-Type Disposal Facility Yoshihiro Akiyama, Kenji Terada, Nobuaki Oda, Tsutomu Yada, Takahiro Nakajima, RWMC (Japan)

SESSION H9: Repository Engineering and Demonstration

Wednesday 13:30 Room 101 --------------------------------------------------------------------------------------- Session Co-Chairs: Ian G. McKinley, McKinley Consulting (Switzerland) and Hidekazu Asano, RWMC (Japan)

1. 40304 Keynote – Repository engineering and demonstration: special challenges for TRU Ian G. McKinley, McKinley Consulting (Switzerland); Hiroyasu Takase, Quintessa Japan (Japan)

2. 40119 – Half-Scale Test: An important step to demonstrate the feasibility of the Belgian Supercontainer concept for disposal of HLW Lou Areias, SCK•CEN/Euridice (Belgium); Bart Craeye, Artesis Hogeschool Antwerpen (Belgium); Geert De Schutter, Ghent University (Belgium); Hughes Van Humbeeck, William Wacquier, ONDRAF/NIRAS (Belgium); Alain Van Cotthem, Loic Villers, Technum-Tractebel Engineering (Belgium)

3. 40175 – Full-Scale Test on Overpack Closure Techniques for HLW Repository Operation - Welding Methods and UT Systems for Long-Term Structural Integrity of the Weld Joint - Ario Nakamura, Hidekazu Asano, RWMC (Japan); Susumu Kawakami, IHI Corporation (Japan); Takashi Ito, Mitsubishi Heavy Industries, Ltd. (Japan); Takashi Furukawa, Japan Power Engineering and Inspection Corporation (Japan); Kyosuke Fujisawa, Kobe Steel, Ltd. (Japan)

4. 40242 – Design Options for HLW Repository Operation Technology, (I) Demonstration and Evaluation of Remote Handling Technologies Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan); Hideki Kawamura, Obayashi Corporation (Japan)

—————— Break ——————— 5. 40251 – Design Options for HLW Repository Operation

Technology, (II) Bentonite Block Forming and Vertical Emplacement Hajime Takao, Tatsuhiro Takegahara, JGC Corporation (Japan); Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan)

6. 40268 – Design Options for HLW Repository Operation Technology, (III) Transportation and Horizontal Emplacement of Pre-Fabricated EBS Module (PEM)

-16-

Technical Sessions

Susumu Kawakami, IHI Corporation (Japan); Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan)

7. 40236 – Design Options for HLW Repository Operation Technology, (IV) Shotclay Technique for Seamless Construction of EBS Ichizo Kobayashi, Soh Fujisawa, Makoto Nakajima, Masaru Toida, Kajima Corporation (Japan); Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan)

8. 40254 – Design Options for HLW Repository Operation Technology, (V) Preliminary Study and Small Scale Experiments on the Method of Removal of Buffer Material with Salt Solution Satohito Toguri, Jiho Jang, Takashi Ishii, Mitsunobu Okihara, Kengo Iwasa, Shimizu Corporation (Japan); Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan)


Wednesday 13:30 Room 405 --------------------------------------------------------------------------------------- Session Co-Chairs: Kenji Karasaki, LBNL (USA) and Yuji Ijiri, Taisei Corporation (Japan)

1. 40135 – Dry-Run of Site Investigation Planning Using the Manual for Preliminary Investigation in Japan Shigeki Akamura, Tadashi Miwa, NUMO (Japan); Tatsuya Tanaka, Obayashi Corporation (Japan); Hiroshi Shiratsuchi, Tokyo Electric Power Service Co., Ltd. (Japan); Atsushi Horio, DIA Consultants Co., Ltd. (Japan)

2. 40070 – Evaluation of the long-term evolution of the groundwater system in the Mizunami area, Japan Takashi Mizuno, Teruki Iwatsuki, JAEA (Japan); Antoni E. Milodowski, BGS (UK)

3. 40077 – Study on the Estimation Error Caused by Using One-Dimensional Model for the Evaluation of Dipole Tracer Test Yuji Ijiri, Yumi Naemura, Taisei Corporation (Japan); Kenji Amano, Keisuke Maekawa, Kunio Ota, Takanori Kunimaru, Atsushi Sawada, JAEA (Japan)

4. 40056 – Development of Comprehensive Techniques for Coastal Site Characterisation (1) Strategic Overview Kunio Ota, Kenji Amano, Tadafumi Niizato, JAEA (Japan); W. Russell Alexander, Bedrock Geosciences (Switzerland); Yoshiaki Yamanaka, Suncoh Consultants (Japan)

—————— Break ——————— 5. 40052 – Development of Comprehensive Techniques for

Coastal Site Characterisation (3) Conceptualisation of Long-Term Geosphere Evolution Tadafumi Niizato, Kenji Amano, Kunio Ota, Takanori Kunimaru, JAEA (Japan); Bill Lanyon, Nagra (Switzerland); W. Russell Alexander, Bedrock Geosciences (Switzerland)

6. 40048 – Development of Comprehensive Techniques for Coastal Site Characterisation (2) Integrated Palaeohydrogeological Approach for Development of Site Evolution Models Kenji Amano, Tadafumi Niizato, Hideharu Yokota, Kunio Ota, JAEA (Japan); Bill Lanyon, Nagra (Switzerland); W. Russell Alexander, Bedrock Geosciences (Switzerland)

7. 40041 – Development of Methodology of Groundwater Flow and Solute Transport Analysis in the Horonobe Area, Hokkaido, Japan Keisuke Maekawa, Hitoshi Makino, Hiroshi Kurikami, Tadafumi Niizato, Manabu Inagaki, Makoto Kawamura, JAEA (Japan)

SESSION D6: Waste Treatment and Non-Reactor

Wednesday 13:30 Room 406 --------------------------------------------------------------------------------------- Session Co-Chairs: Takeshi Ishikura, IAE (Japan) and Hitoshi Sakai, Toshiba Corporation (Japan)

1. 40105 – Estimation of Radioactivity of Graphite Blocks in Tokai Power Station Using Statistical Method Masaaki Nakano, Fuji Electric Holdings Co., Ltd. (Japan); Hisashi Mikami, Fuji Electric Systems Co., Ltd. (Japan); Hideaki Ichige, Shinich Tsukada, JAPC (Japan)

2. 40115 – The treatment of hexavalent chromium in waste liquid from Fugen Decommissioning Nobuo Ishizuka, Yuji Sato, Wataru Fujiwara, JAEA (Japan); Yuki Yahiro, Seiji Yamamoto, Koji Negishi, Tadashi Fukushima, Hitoshi Sakai, Norimasa Yoshida, Toshiba Corporation (Japan)

3. 40201 – Characterization of Radioactive Waste from Side Structural Components of a CANDU Reactor for Decommissioning Applications in Korea Rizwan Ahmed, Gyunyoung Heo, Kyung Hee University (Korea Rep.); Dong-Keun Cho, Jongwon Choi, KAERI (Korea Rep.)

4. 40068 – Uranium refining and conversion plant decommissioning project Naoki Zaima, Yasuyuki Morimoto, Noritake Sugitsue, Kazumi Kado, JAEA (Japan)

SESSION M2: Public Involvement Wednesday 13:30 Room 303 --------------------------------------------------------------------------------------- Session Chair: Koji Nagano, CRIEPI (Japan)

1. 40288 – Activities of the OECD/NEA in the Field of Stakeholder Confidence for Radwaste Management and Decommissioning Claudio Pescatore, OECD/NEA

2. 40219 – A Comparative Study of Stakeholder Participation in the Cleanup of Radioactive Wastes in the US, Japan and UK William F. Lawless, Fjorentina Angjellari-Dajci, Paine College (USA); Mito Akiyoshi, Senshu University (Japan); John Whitton, Nexia Solutions (UK); Christian Poppeliers, Augusta State University (USA)

3. 40076 – Territorial Integration of the Geological Repository in France Gérald Ouzounian, Sebastien Farin, Roberto Miguez, Jean-Louis Tison, Andra (France)

POSTER SESSIONS

Monday PM, October 4, 2010 Tuesday AM&PM, October 5, 2010

SESSION L8: L/ILW Poster

Room 102 ---------------------------------------------------------------------------------------

1. 40006 – A GoldSim Modeling Approach to Safety Assessment of an LILW Repository System Youn Myoung Lee, Jongtae Jeong, Jongwon Choi, KAERI (Korea Rep.)

2. 40011 – Gas Migration Mechanism of Saturated Highly-Compacted Bentonite and its Modeling Yukihisa Tanaka, Michihiko Hironaga, Koji Kudo, CRIEPI (Japan)

-17-

Technical Sessions

3. 40012 – Development of numerical simulation method for gas migration through highly-compacted bentonite using model of two-phase flow through deformable porous media Yukihisa Tanaka, CRIEPI (Japan)

4. 40020 – Planning of Large-Scale In-Situ Gas Generation Experiment in Korean Radioactive Waste Repository Juyoul Kim, Sukhoon Kim, FNC Technology Co. (Korea Rep.); Jin Beak Park, Sungjoung Lee, KRMC (Korea Rep.)

5. 40024 – Estimation and measurement of porosity change in cement paste Eunyong Lee, Haeryojng Jung, Ki-jung Kwon, KRMC (Korea Rep.); Do-Gyeum Kim, Korea Institute of Construction Technology (Korea Rep.)

6. 40082 – Separation and Recovery of Sodium Nitrate from Low-level Radioactive Liquid Waste by Electrodialysis Yoshihiro Meguro, Atsushi Kato, Yoko Watanabe, Kuniaki Takahashi, JAEA (Japan)

7. 40109 – Study on Mechanical Influence of Gas Generation and Migration on Engineered Barrier System in Radioactive Waste Disposal Facility Mamoru Kumagai, JNFL (Japan); Shuichi Yamamoto, Kunifumi Takeuchi, Obayashi Corporation (Japan); Yukihisa Tanaka, Michihiko Hironaga, CRIEPI (Japan)

8. 40221– Decontamination of Radioactive Concrete Waste by Thermal and Mechanical Processes Byung Youn Min, Wang Kyu Choi, Ki Won Lee, Kune Woo Lee, Un Soo Chung, KAERI (Korea Rep.)

9. 40302 – Latex Particles Functionalized with Transition Metals Ferrocyanides for Cesium Uptake and Decontamination of Solid Bulk Materials Valentin Avramenko, Svetlana Bratskaya, Veniamin Zheleznov, Irina Sheveleva, Dmitry Marinin, Valentin Sergienko, Institute of Chemistry FEDRAS (Russia)

SESSION H11: SF/TRU/HLW Poster

Room 102 ---------------------------------------------------------------------------------------

1. 40001 – Investigation of Colloid-Facilitated Effects on the Radionuclides Migration in the Fractured Rock with a Kinetic Solubility-Limited Dissolution Model Chun-Ping Jen, National Chung Cheng University (Taiwan); Neng-Chuan Tien, Industrial Technology Research Institute (Taiwan)

2. 40013 – Modeling hydraulic conductivity and swelling pressure of several kinds of bentonites affected by salinity of water Yukihisa Tanaka, Takuma Hasegawa, Kunihiko Nakamura, CRIEPI (Japan)

3. 40018 – Current R&D Activities in the Study on Geosphere Stability Takahiro Hanamuro, Ken-ichi Yasue, Yoko Saito-Kokubu, Koichi Asamori, Tsuneari Ishimaru, Koji Umeda, JAEA (Japan)

4. 40019 – In Situ Stress Measurements in Siliceous Mudstones at Horonobe Underground Research Laboratory, Japan Hiroyuki Sanada, Takahiro Nakamura, Yutaka Sugita, JAEA (Japan)

5. 40038 – Low Alkaline Cement Used in the Construction of a Gallery in the Horonobe Underground Research Laboratory Masashi Nakayama, Haruo Sato, Yutaka Sugita, Seiji Ito, JAEA (Japan); Masashi Minamide, Yoshito Kitagawa, Taisei Corporation (Japan)

6. 40040 – Effects of Nitrate on Nuclide Solubility for Co-Location Disposal of TRU Waste and HLW Gento Kamei, Morihiro Mihara, JAEA (Japan); Toshiyuki Nakazawa, Norikazu Yamada, Mitsubishi Materials Corporation (Japan)

7. 40047 – A study on groundwater infiltration in the Horonobe area, northern Hokkaido, Japan Hideharu Yokota, Yoichi Yamamoto, Keisuke Maekawa, Minoru Hara, JAEA (Japan)

8. 40051 – Effective Use of Uranium Resources and Dissolution of Recovered Uranium Storage Accumulation by a Uranium Multi-Recycle System Kenji Kotoh, Yuzo Yamashita, Takeshi Nakamura, Kyushu University (Japan)

9. 40053 – Advanced ORIENT Cycle - Progress on Fission Product Separation and Utilization Isao Yamagishi, Masaki Ozawa, JAEA (Japan); Hitoshi Mimura, Tohoku University (Japan); Shohei Kanamura, Koji Mizuguchi, Toshiba Corporation (Japan)

10. 40064 – Hydrogeological Characterization Based on Long Term Groundwater Pressure Monitoring Shuji Daimaru, Ryuji Takeuchi, Masaki Takeda, Masayuki Ishibashi, JAEA (Japan)

11. 40065 – Development of A Quality Management System (QMS) for Borehole Investigations: (2) Evaluation of Applicability of QMS Methodology for the Hydrochemical Dataset Takanori Kunimaru, Kunio Ota, Kenji Amano, JAEA (Japan); W. Russell Alexander, Bedrock Geosciences (Switzerland)

12. 40067 – An Analytical Model on the Sealing Performance of Space for the Design of Buffer Material and Backfill Material Haruo Sato, JAEA (Japan)

13. 40069 – Current Status of Horonobe URL Project in Construction Phase Hironobu Abe, Koichiro Hatanaka, JAEA (Japan)

14. 40074 – Development of New Ultrafiltration Techniques Maintaining In-Situ Hydrochemical Conditions for Colloidal Study Daisuke Aosai, Yuhei Yamamoto, Takashi Mizuno, JAEA (Japan)

15. 40089 – Sorption Behavior of Iodine on Calcium Silicate Hydrates Formed as a Secondary Mineral Keisuke Shirai, Yuichi Niibori, Akira Kirishima, Hitoshi Mimura, Tohoku University (Japan)

16. 40098 – Development of a quality management system for borehole investigations: (1) Quality assurance and quality control methodology for hydraulic packer testing Shinji Takeuchi, Takanori Kunimaru, Kunio Ota, JAEA (Japan); Bernd Frieg, Nagra (Switzzerland)

17. 40101 – Comparison of Post-Irradiation Experimental Data and Theoretical Calculations for Inventory Estimation of Long-Lived Fission Products in Spent Nuclear Fuel Shiho Asai, Yukiko Hanzawa, Keisuke Okumura, Hideya Suzuki, Masaaki Toshimitsu, Nobuo Shinohara, JAEA (Japan); Satoru Kaneko, Kensuke Suzuki, Tokyo Electric Power Company (Japan)

18. 40103 – Selective Uptake of Palladium from High-Level Liquid Wastes by Hybrid Microcapsules Enclosed with Insoluble Ferrocyanides Hitoshi Mimura, Takashi Sakakibara, Yan Wu, Yuichi Niibori, Tohoku University (Japan); Shin-ichi Koyama, Takashi Ohnishi, JAEA (Japan)

19. 40113 – An Empirical Model to Determine the Modes of Corrosion of Carbon Steel Under Near Field

-18-

Technical Sessions

Environments of Geological Disposal Toshikatsu Maeda, Masatoshi Watanabe, Seiji Takeda, Shinichi Nakayama, JAEA (Japan)

20. 40124 – Trends in Scenario Development Methodologies and Integration in NUMO's Approach Takeshi Ebashi, Katsuhiko Ishiguro, NUMO (Japan); Keiichiro Wakasugi, JAEA (Japan); Hideki Kawamura, Obayashi Corporation (Japan); Irina Gaus, Stratis Vomvoris, Andrew J. Martin, Nagra (Switzerland); Paul Smith, Safety Assessment Management (Switzerland)

21. 40137 – Development of Methodology to Construct a Generic Conceptual Model of River-valley Evolution for Performance Assessment of HLW Geological Disposal Makoto Kawamura, Shin-ichi Tanikawa, Tadafumi Niizato, Ken-ichi Yasue, JAEA (Japan)

22. 40176 – Structural Integrity Evaluation Approach for PWR Spent Nuclear Fuel Yun Seog Nam, Yong Hwan Kim, Kyeong Lak Jeon, Seong Ki Lee, Ki Sung Choi, Chang Sok Cho, KNF (Korea Rep.)

23. 40200 – Development of Program Categories to Assess the Radiological Dosage during Spent Fuel Transportation Suhong Lee, Sangwon Shin, Jaemin Lee, Enesys Co., Ltd. (Korea Rep.); Kiyeoul Seong, Jeonghyoun Yoon, KRMC (Korea Rep.)

24. 40222 – An Analyse of Components and Impact Factors Related to Spent Fuel Transportation Plans SangWon Shin, JaMin Lee, SuHong Lee, Enesys Co., Ltd. (Korea Rep.); ChangYeal Baeg, JeongHyoun Yoon, KRMC (Korea Rep.)

25. 40225 – Exploiting synergies between the UK & Japanese geological disposal programmes Ellie Scourse, Atkins (UK); Hideki Kawamura, Obayashi Corporation (Japan); Ian G. McKinley, McKinley Consulting (Switzerland)

26. 40239 – Realistic Consequence Analysis of River Erosion Scenarios for a HLW Repository Kaname Miyahara, Manabu Inagaki, JAEA (Japan); Makoto Kawamura, MMTEC (Japan); Takanori Ebina, NESI (Japan); Ian G. McKinley, McKinley Consulting (Switzerland); Michael J. Apted, Intera (USA)

27. 40272 – Removal of Fission Products in the Spent Electrolyte Using Iron Phosphate Glass as a Sorbent Ippei Amamoto, Masami Nakada, Yoshihiro Okamoto, JAEA (Japan); Naoki Mitamura, Tatsuya Tsuzuki, Central Glass Co., Ltd. (Japan); Yasushi Takasaki, Atsushi Shibayama, Akita University (Japan); Tetsuji Yano, Tokyo Institute of Technology (Japan)

28. 40295 – Propagation and Interactions of Acoustic Waves in a Waveguide Attached at the Surface of Rock Jin-Seop Kim, Kyung-Soo Lee, S. Kwon, KAERI (Korea Rep.); Gye-Chun Cho, KAIST (Korea Rep.)

SESSION D7: D&D Poster

Room 102 ---------------------------------------------------------------------------------------

1. 40075 – Methods of Selected Input Calculation Data Verification and Their Influence on Decommissioning Cost in the OMEGA Code František Ondra. Vladimír Daniška, Ivan Rehák, DECOM, a.s. (Slovakia); Vladimír Nečas, Slovak University of Technology in Bratislava (Slovakia); Oto Schultz, DECONTA, a.s. (Slovakia)

2. 40126 – Detailed Standardized Decommissioning Parameters Calculation for Larger Technological Aggregates and Relevant Buildings in Nuclear Power

Plants Using the OMEGA Code Peter Bezák, Vladimír Daniška, Ivan Rehák, DECOM, a.s . (Slovakia)

3. 40190 – Dismantling Method of Fuel Cycle Facilities Obtained by Dismantling of the JRTF Fumihiko Kanayama, JAEA (Japan)

4. 40191 – Computer Simulation of Cryogenic Cutting Technology for Dismantling Highly Activated Facilities Sung-Kyun Kim, Kune-Woo Lee, KAERI (Korea Rep.)

5. 40193 – Strippable Core-Shell Polymer Emulsion for Decontamination of Radioactive Surface Contamination Ho-Sang Hwang, Bum-Kyoung Seo, Kune-Woo Lee, KAERI (Korea Rep.)

SESSION R4: ER Poster

Room 102 ---------------------------------------------------------------------------------------

1. 40025 – Improvement of quicklime mixing treatment by carbon dioxide ventilation Yuki Nakagawa, Hisayoshi Hashimoto, Hitachi Construction Machinery Co., Ltd. (Japan); Koichi Suto, Chihiro Inoue, Tohoku University (Japan)

2. 40166 – Procedure and results of decommission of R&D facility of uranium fuel Hirokazu Tanaka, Masao Shimizu, Ryoji Tanimoto, Kazuhiko Maekawa, Shinzo Ueta, Mitsubishi Materials Corporation (Japan); Susumu Tojo, SERNUC Corporation (Japan)

3. 40224 – Hydrogen Production from a PV/PEM Electrolyzer System Using a Neural-Network-Based MPPT Algorithm Abd El-Shafy Nafeh, Electronics Research Institute (Egypt)

4. 40301 – Cement based solidification / stabilization of industrial contaminated soil using various cement additives Grega E. Voglar, RDA (Slovenia); Domen Lestan, University of Ljubljana, (Slovenia)

SESSION M3: EM/PI Poster

Room 102 ---------------------------------------------------------------------------------------

1. 40099 – Removal of Fluorine and Boron from Groundwater Using Radiation-Induced Graft Polymerization Adsorbent at Mizunami Underground Research Laboratory Yosuke Iyatomi, Hiroyuki Hoshina, Noriaki Seko, Noboru Kasai, Yuji Ueki, Masao Tamada, JAEA (Japan)

2. 40184 – The Optimized Risk Management of the Waste from TENORM and Nuclear Industries - How to Harmonize Risk from Various Sources Yoko Fujikawa, Kyoto University Research Reactor Institute (Japan); Michikuni Shimo, Fujita Health University (Japan); Hidenori Yonehara, National Instuitute of Radiological Sciences (Japan); Tadashi Tujimoto, Electron Science Institute (Japan)

3. 40205 – Exposure Dose Evaluation of Worker at Radioactive Waste Incineration Facility on KAERI Sang Kyu Park, Jong Seon Jeon, Youn Hwa Kim, Jae Min Lee, Enesys Co., Ltd. (Korea Rep.); Gi Won Lee, KAERI (Korea Rep.)

4. 40209 – Scenario Development for Safety Assessment of Waste Repository for Feasibility Study on Transmutation of Spent Nuclear Fuel into LILW Using PEACER Sung-yeop Kim, Kun Jai Lee, KAIST (Korea Rep.)

-19-

The following exhibitors are greatly appreciated: (Exhibitors in Alphabetical Order) · Belgoprocess

· Central Research Institute of Electric Power Industry

· Fuji Electric Systems Co., Ltd.

· GEOSCIENCE RESEARCH LABORATORY Co., Ltd.

· JAPAN NUCLEAR FUEL LIMITED

· Kajima Corporation

· Kobe Steel, Ltd. (KOBELCO)

· Mitsubishi Materials Corporation

· NIPPON KOEI Co., Ltd.

· Nuclear Waste Management Organization of Japan (NUMO)

· OBAYASHI CORPORATION

· PIERCAN USA, Inc.

· Raax Co., Ltd. / Earth Scanning Association

· SHIMIZU CORPORATION

· TAISEI CORPORATION

· Toshiba Corporation

· Web I Laboratories, Inc.

· Westinghouse Electric Company

List of Exhibitors

-20-

Fuji s 3 unique technologies which include remote handling, radwastetreatment and high temperature gas-cooled reactor are contributing for ensuring energy resources for longer than a century.

Contact <United States America>: Fuji Electric Corp. of AmericaPhone 201-712-0555 FAX 201-368-8258 URL http//www.fujielectric.com/fecoa /

<Japan> : Fuji Electric Systems Co., Ltd.Phone +81-44-329-2169 FAX +81-44-329-2178URL http://www.fujielectric.com/

Fuel Pellets Manufacturing Facility Internal Equipment within the Glove Box

Exterior View of Fuji Resin Reducer(LPOP)(LPOP : Low Pressure Oxidation Process)

Performance of volume reduction (ion-exchange resin)

(1/10–1/20)Verification test of remote dismantling system

(Tokai-1 Decommissioning)

Gripper

Reactor Vessel

Cutting Torch

Nuclear Power Generation /Nuclear Fuel Cycle Related Facilities

Commercial HighTemperature Gas-cooled Reactor

(HTGR)

Active Core(Fuel Blocks)

Core Bottom Structure

(The HTGR heat utilization plant)

Top view of the core(Outer diameter 4.25m)

Fuel Transfer System

Remote Handling & Fuel Fabrication Technologies1

2 Radwaste Treatment Systems & Decommissioning Technologies

3 Reactor Technology for High Temperature Gas-cooled Reactor

HTTR Core internals

(HTTR : High Temperature Engineering Test Reactor)

pelletinspection

-28-

Abstracts

“Nuclear Energy Strategy for Sustainable Growth: Aiming at Green Innovation and Life Innovation”

Tatsujiro Suzuki, Vice Chairman, Japan Atomic Energy Commission (Japan)

Japan Atomic Energy Commission released a report, entitled “Nuclear Energy Strategy for Growth” on May 25, 2010. The report is intended to contribute the government’s “New Growth Strategy” which was eventually released in June 18, 2010. This JAEC’s report emphasizes the role of nuclear energy technologies for “green innovation” and “life innovation” which are two primary components of New Growth Strategy. The key messages included in the report are “strengthening capability to meet global challenges” and “enhancing citizen’s confidence in nuclear energy technologies”.

The report listed five major recommendations as follows. 1. Improve performance of nuclear power plants aim 2. ing at the world best class and take necessary measures for adding more nuclear plants to achieve

increased share of nuclear power in total power generation for “green innovation” Promote industrial application of radiation technologies, such as medial, agricultural and scientific applications, as “strategic industry” for life innovation

3. Develop social and economic environments for achieving above goals, which include; (1) improve scientific literacy (2) take new initiatives for data/information disclosure (3) enhance economic visualization of CO2 merits of nuclear energy (4) reform safety regulatory structure to improve public confidence (5) develop better approaches for improving local community development and nuclear related facilities (6) develop public/private human network in Asia.

4. Develop appropriate measures to meet increasing global needs for nuclear power and radiation technologies

5. Develop long term social platform for sustainable growth for nuclear energy technologies.

“U.S. Office of Environmental Management - World Leaders in Nuclear Cleanup and Construction”

Dae Y. Chung, Principal Deputy Assistant Secretary for Environmental Management, US DOE (USA)

U.S. Department of Energy Principal Deputy Assistant Secretary for Environmental Management, Dae Chung, will present an overview of the progress, challenges, and future opportunities associated with the world's largest nuclear cleanup program. The Office of Environmental Management (EM) is responsible for the safe cleanup of over two million acres of land located in 34 states. EM employs more than 30,000 people – primarily scientists, engineers, and hazardous waste technicians, and has an annual budget of approximately $6 billion.

“Recent Progress in Radioactive Waste Management in Korea”

Ho Taek Yoon, Senior Vice President, KRMC (Korea Rep.)

Republic of Korea is set to become a major nuclear power generation country. Nuclear energy initiative is a strategic priority for R.O.K, and of which nuclear share in electricity generation is planned to become 56% by 2030, with nuclear installed capacity of 35 GWe. Currently, 20 reactors with installed capacity of 17.7 GWe provide almost 40% of domestic electricity demand. However, the very active nuclear power program inevitably causes increase in the build-up of the radioactive wastes including Low- and Intermediate-Level radioactive Waste (LILW) and spent nuclear fuel. Therefore, reliable and effective management of the radioactive waste and spent nuclear fuel has become a key agenda for the continuous growth of the nuclear power program.

LILW generated by the nuclear power plants is being tentatively stored at each reactor site. The accumulated amount of LILW in Korea’s nuclear power plants as of Dec. 2009 is approximately 87,000 drums (200 liter drum). And together with those of research institutes, nuclear fuel manufacturer, general industry and medical sectors reached approximately 112,000 drums (200 liter drum). The construction of a final repository is urgently needed to manage the radioactive waste effectively and safely.

Efforts to secure site for the radioactive waste disposal facility in Korea began in 1986. In 18 years from 1986 to 2004, nine attempts of site selection had been executed unfruitfully. In the 10th attempt which was made in 2005, Korean government requested that the public acceptance shall be confirmed through resident voting at the each local government volunteering for the repository hosting and decided that the spent nuclear fuel site will be separated from LILW disposal site, and the community support packages will be guaranteed by the law. Under the above mentioned basic principles, the resident vote was implemented in the four volunteered local governments including Gunsan-kun, Youngdok-kun, Pohang city, and Gyeongju city. Among these local governments, Gyeongju city recorded the highest consent vote rate (89.5%)

OPENING SESSION

-36-

Abstracts

and became the hosting city of LILW repository site. After the vote, on January 2, 2006, the Ministry of Knowledge and Economy (MKE) designated Bonggil-ri,

Yangbuk-myeon, Gyeongju-city, North Gyeongsang Province (approximately 2,100,000 m2) as the LILW repository site. The first stage Wolsong disposal facility construction will be completed by December 2012 as underground silos type with a disposal capacity of 100,000 drums.

Spent nuclear fuels generated from nuclear power plants are stored at reactor sites pending construction of an interim storage facility or final repository. The cumulative amount of spent fuel is about 10,761 tons as of Dec. 2009, and is expected to increase up to 20,000 tons by 2020. The existing storage capacity as of Dec. 2009 is 13,532 tons.

National policy for spent nuclear fuel management shall be decided in a timely manner through public participation process taking into consideration of current international trends and technology development. KRMC launched the spent fuel management alternative study and establishment of road map in Dec. 2009 to promote expert group’s consensus and for the public involvement in the future which is approaching imminently.

“Radioactive waste management – achievements, needs and future expectations”

Irena Mele, Head of Waste Technology Section, Division of Nuclear Fuel Cycle and Waste Technology, IAEA

International nuclear community is facing a period of dynamic changes. The growing interest for nuclear energy brings new optimism to the community but also new challenges and increased responsibilities. Radioactive waste management remains on the agenda even in most developed nuclear countries. Although significant progress has been achieved in this area and many disposal solutions for different types of waste have been successfully implemented worldwide, unresolved disposal of HLW and spent nuclear fuel remains the major concern of the public related to the use of nuclear energy and may become an obstacle for the planned expansion of their nuclear programmes. There are also countries that are still missing necessary radioactive waste management infrastructure and, thus, capacity and capability to safely manage their low level and intermediate level radioactive waste. In spite of well developed and available waste management technologies and good practices many our Members States are still meeting great deficiency in radioactive waste management infrastructure and need strong support and assistance to develop adequate level for safe and efficient management of their radioactive waste. In particular countries with only institutional waste experience great difficulties in providing waste management facilities due to the lack of capacity and expertise.

Many new countries have also expressed their interest and intention to embark on nuclear energy. When planning and preparing for new nuclear build, major attention is given to the construction and operation of nuclear power plant and infrastructure building. Much less attention of those countries and also nuclear industry is given to the fact that in parallel with the NPP the radioactive waste management infrastructure needs to be developed as well. A big challenge is to increase awareness among those countries and also among nuclear industry on burning need to include also preparations for spent fuel and radioactive waste management in early planning process.

Decommissioning of nuclear facilities has become a mature technology and planning for decommissioning and implementing a decommissioning plan is commonly accepted practice in most of nuclear countries. However, additional encouragement and support is still needed in countries with less experience and smaller programmes. Another important issue is cleaning-up of sites with legacy waste emanating from past research, medical or military activities. Important legacies from earlier uranium mining are known in many parts of the world. Remediation of land contaminated by radioactive material residues is needed in many countries, like in the Central Asian republics, where the remedial challenges will require a consolidated international support approach.

The paper will provide an overview of current situation and future prospects in all these areas of radioactive waste management, decommissioning and environmental remediation and highlight the role and efforts of the IAEA to ensure that adequate safety standards in this area are applied worldwide, that experiences and good practices gained in radioactive waste management area are shared and communicated between the countries and potential users, and that proper assistance is provided to countries considering nuclear power.

SESSION L1: Waste Management

1) 40081 – Radioactive Waste: Feedback of 40-year Operations in France

Michel Dutzer, Gérald Ouzounian, Roberto Miguez, Jean-Louis Tison, Andra (France)

France's experience in the management of radioactive waste is supported by forty years of operational activities in the field of surface disposal. This feedback is related to three disposal facilities:

- Centre de la Manche disposal, not far away Cherbourg, from design to post-closure facility, - Centre at Soulaines-Dhuys from site selection to design to operation during nearly 20 years, - Centre at Morvilliers from site selection to operation for seven years now.

During the operational period of Centre de la Manche disposal facility (1969-1994), the safety concept for low-and intermediate level short lived waste (LIL-SLW) was developed and progressively incorporated in the procedures of the facility. The facility entered its institutional control period and the experience of this facility has been useful for the operating facilities.

-37-

Abstracts

Centre de l’Aube that took over Centre de la Manche, and Morvilliers for very low level wastes. Both facilities currently accommodate the major part of the volume of radioactive wastes that are generated in France. However disposal facilities have to be considered as rare resources. Then new waste management options are being investigated as the disposal of large components or recycling metallic wastes within the nuclear industry.

2) 40226 – The Ethics of the Management of Low and Intermediate Radioactive Wastes Generated by Cernavoda NPP, a Challenge for the Romanian Specialists

Gheorghe Barariu, Subsidiary of Technology and Engineering for Nuclear Objectives (Romania)

In Romania there was an extensive interest on treatment of radioactive wastes generated by nuclear technology used in industrial and research field. Still the existence of NPP Cernavoda and the provision for a new nuclear power plant impose the existence of the disposal capacities for future radioactive wastes that will be further generated by operation and decommissioning of new units. The key aspects of the radioactive wastes management at the Romanian Cernavoda NPP equipped with CANDU 600 reactors result from missing of a Radioactive Waste Treatment Plant. The Strategy for radioactive waste management was elaborated by the former National Agency for Radioactive Waste (ANDRAD), the jurisdictional authority for final disposal and the coordination of nuclear spent fuel and radioactive waste management (Order 844/2004), with attributions established by Governmental Decision (GO) 31/2006. The Strategy specifies the commissioning of the Saligny L/IL Radwaste Repository near Cernavoda NPP, in 2014. The new Agency AN&DR, Nuclear Agency and for Radioactive Waste which was appointed in 2010, probable will follows the same Strategy. A lot of constraints, including limited available surface and bearing capacity on site of the Saligny Repository near Cernavoda NPP limits the possibilities for selecting of the technologies for radioactive wastes treatment. For the new NPP not sised and located yet wil be provided a separate strategy. Based on the input data the main constraints for the new repository design were identified: Cernavoda NPP specific waste characterization is not finalized, the wastes are not properly conditioned for the final disposal, environmental constraints due to characteristics of the Saligny Repository, regulatory constraints, public acceptance constraints, constraints related Utility concept for treatment of radioactive waste, uncertenties related the New Governamental Agency AN&DR. For every constraint mentioned above, suitable measures will be taken to reduce the uncertainties. The most important technological dilema is related to selection betwen technogies that implies impact on present generation ( incineration, radwaste transfer from ss drums to cs drums and ss drums supercompaction ), and technologies that confined the tritium and C-14 in the Repository with impact for next generations. The paper will present the implicative of this selection on the future Radioactive Waste Treatment Plant.

3) 40149 – CEA's radioactive waste and unused fuel inventory - Marcoule site example

Jean-Guy Nokhamzon, Marc Butez, Ddaniel Fulleringer, CEA (France)

In the field of clean up and decommissioning of nuclear facilities the knowledge of the radioactive waste and unused fuel inventories is crucial. For CEA (French Commission in charge of Atomic Energy and Alternative Energies), this knowledge is a key issue : in the short term for the flows stream lining as well as in the middle and long terms for investment strategy in nuclear support facilities.

These needs appear in a larger extent at the national scale in France, with the enforcement of two laws promulgated in 2006 dealing with radioactive waste and with transparency in Nuclear Safety.

More generally, a precise and reliable knowledge of inventory and flow production forecast is indispensable to give pertinent answers to the questions related to waste valuation. Thus, all these subjects are also worked out with our authorities in charge of financial aspects because there are essential for the cost control of clean-up and decommissioning projects CEA has to manage. In effect, in these projects, waste management represents a great part of cost and risk.

For all these reasons CEA has carried out an important work to collect all his waste and unsued fuel inventories. For this reason we have developed a new "INFLUVAL" data base which will be combined with a softare featured with coherency checking and analysis modules. Indeed, we must guarantee uniqueness and relevance of data coming from decommissioning projects and facility managers.

As a consequence, in 2009 CEA Clean-up and Decommissioning Division has decided to perform, , a comprehensive inventory of waste and unused fuel with the following objectives :

- Consolidation of stock and expected flows, - Evaluation of reference costs - Help to dimensioning and investment decision for facilities programs ( storage, waste conditioning facility,…) - Validation of waste cost ratio.

In this paper we will take Marcoule as a more precise example. On this CEA center, heavy operations of clean-up and decommissioning are carried out on the first French reprocessing plants –UP1-, and the related workshops.

Theses installations have been shut down in 1996. Since this date the main operations were focussed on definitive shutdown and dismantling of equipments which were to be realized in priority in the goal of reducing radiological activity and operating costs.

Thus, the consolidation of the inventory data is essential to valid a scenario and then to design facilities for the following stages of the UP1 project.

-38-

Abstracts

4) 40258 – Management of historical waste legacy at NRG Petten

Renate de Vos, Nuclear Research and consultancy Group (Netherlands)

Since the 1960s, the Waste Storage Facility (WSF) in Petten has been used to store radioactive waste. In 1993, the Dutch government decided that all nuclear waste should be stored in the central storage facility COVRA (Dutch Central Organisation for Radioactive Waste). This means that the drums with historical waste must all be opened, their contents sorted according to intermediate- and low-level radioactivity and subsequently packed for transport to the COVRA facility.

The intermediate-level radioactive waste processing Unit (HAVA-VU) has been designed specifically for this purpose.

In this paper an overview will be given of the process of this HAVA-VU project in the last decade. Subsequently the current status and path forward will be described. Also the chosen solutions for waste processing and characterization will be discussed.

5) 40031 – Norwegian Support For Regulations Of Radioactive Waste Management From Uranium Mining And

Mill Tailings In Central Asia

Tamara Zhunussova, Malgorzata Sneve, Astrid Liland, Norwegian Radiation Protection Authority (Norway); Alexander Kim, Kazakhstan Atomic Energy Committee (Kazakhstan);

Ulmas Mirsaidov, Tajikistan Nuclear and Radiation Safety Agency (Tajikistan); Baigabyl Tolongutov, Chui Ecological Laboratory of Kyrgiz Republic (Kyrgiz);

Per Strand, Norwegian Radiation Protection Authority, University of Life Sciences (UMB) (Norway)

In Central Asia the radioactive waste comes mainly from uranium mining and milling, nuclear weapon testing and nuclear power development and other ionizing sources. This waste was produced, to a greater extent, by the military-industrial complex and the uranium and non-uranium industry, and, to a lesser extent, by the nuclear industry and in the process of use of isotope products. Exploitation and mining of uranium and thorium deposits produce a large amount of solid and liquid radioactive waste, as well volatile contaminants which need a proper management. In Central Asia the wastes are mainly stored at the surface in large piles and represent a long-term potential health and environmental hazard. The process of remediating legacy sites of the past and reducing the threats is now getting under way, with the design and implementation of remediation activities, partly with international support. However, there is a significant lack in the regulatory basis for carrying out such remediation work, including a lack of relevant radiation and environmental safety norms and standards, licensing procedures and requirements for monitoring etc., as well as expertise to transform such a basis into practice. Accordingly, the objective of the proposed project is to assist the relevant regulatory authorities in Kazakhstan, Kirgizstan and Tajikistan to develop national robust and adequate regulations and procedures, taking into account the international guidance and Norwegian experience in regulatory support projects in Russia. Specific expected results in the project period include: a threat assessment report identifying priority areas for regulatory development, based on the status of current regulatory documents and the hazard presented by the different sites and facilities; development of national radioactive waste management strategies in each country; development of an enhanced regulatory framework.

SESSION H1: : National and International Programs (1)

1) 40097 – Overview of NUMO's policy for implementing safe geological disposal and developing supporting technologies

Hiroyuki Tsuchi, Kenichi Kaku, Katsuhiko Ishiguro, Akira Deguchi, Yoshiaki Takahashi, NUMO (Japan)

Based on the Act on Final Disposal of Specified Radioactive Waste (Final Disposal Act), the Nuclear Waste Management Organization of Japan (NUMO) was established in October 2000 as the implementing organization for geological disposal of high-level radioactive waste. In December 2002, NUMO issued a nationwide call for volunteers to initiate the siting process. With the revision of the Final Disposal Act in 2007, some types of TRU waste were included as waste destined for geological disposal and thus fall within the remit of NUMO.

NUMO has been developing the technologies required for implementation of the project and has carried out a range of siting and performance assessment (PA) activities. However, no application has been received as yet from a volunteer municipality. Together with the national government, NUMO is now making increased efforts to obtain public acceptance for initiating literature surveys (the first step in the siting process). In this connection, NUMO decided to prepare a 2010 technical report as a tool for improving the understanding of the geological disposal project by the stakeholders. The report will present the safety concept that describes how NUMO aims to achieve safe geological disposal through ca.100-year project and will document the progress made with developing the technologies that support the safety concept.

Three policies for ensuring safety are followed by NUMO, namely a) staged project implementation and decision making based on iterative confirmation of safety, b) project implementation based on reliable technologies (or Best Available Technologies - BAT) and c) technical activities for building confidence in NUMO’s safety concept. One of the highlights of the report is the introduction of a “roadmap” which includes all project milestones, goals in each stage and main activities for achieving these goals. The roadmap is

-39-

Abstracts

also being used to identify the technologies required for implementing the geological disposal project, thus providing a guideline for evaluating currently available technologies and identifying needs for further technology development. The technologies developed to date and the plan for further technical development are explained in this paper.

2) 40150 – Stepwise Site Selection in Switzerland - Sectoral Plan: Status and Outlook

Thomas Ernst, Markus Fritschi, Stratis Vomvoris, Nagra (Switzerland)

The starting point is the nomination of siting regions that fulfil the criteria for long-term safety and engineering feasibility defined in the Sectoral Plan. In October 2008, Nagra proposed a total of 6 siting regions for the LLW repository and three for the HLW repository; the latter would also be suitable for shared use with the surface facilities and part of the access tunnels for the two types of repositories.

The review of Nagra’s proposals by the safety authority (ENSI – the Swiss Federal Nuclear Safety Inspectorate) and its supporting commission was completed in February 2010. The decision-making process foresees evaluation of the review and recommendations of ENSI by the various agencies at the governmental, cantonal and local level, an open public consultation and finally, a resolution of the comments received by the Swiss Federal Office of Energy, a recommendation to the Federal Government and a decision by the Federal Government.

At the same time, the procedures and guidelines for the selection of the (at least) two potential sites for each of the repositories are being finalised by the safety authorities. Although the safety and engineering criteria remain the same, Stage 2 foresees a provisional safety analysis for each potential site as one of the criteria to be used in the site selection process.

This paper will highlight the main findings of the review by the authorities and Nagra’s response to issues raised. The next steps and preparatory activities for the initiation of Stage 2 will also be described, as well as how the criteria and guidelines specified by ENSI will be applied by Nagra in order to meet the requirements for a successful completion of Stage 2.

3) 40084 – Site Selection for a Geological Disposal in France: an Approach of Convergence

Gérald Ouzounian, Roberto Miguez, Jean-Louis Tison, Andra (France)

On December 1991, the French National Assembly passed the French Waste Management Research Act, authorizing a 15 year research program of three options for HLW: separation and/or transmutation, long-term storage, and geologic disposal. On June 2006, the “Planning Act on the sustainable management of radioactive materials and waste" sets a new framework and new aims to the above mentioned options.

This paper deals only with the geologic disposal research program. In a step by step approach, this program has been broken down into three phases having intermediate goals : site selection for an Underground research Laboratory (URL), potential disposal feasibility, potential reversible disposal design.

The first step of the research program aimed at URL site selection. From 1994 to 1996, Andra carried out geological-characterization work in four departments. This enabled to make the Request for Licensing and Operation of the laboratory facility on three sites. During this phase, wells, 2D seismic campaigns and land studies of geologic outcrops were the essential activities. The result was the selection of the most suitable site for the implementation of an underground laboratory. Main results on Bure URL will be presented in the paper.

In the second phase the research program targeted the safety and technical feasibility of a potential reversible disposal somewhere in Meuse and Haute Marne D departments site, chosen by the government in 1998. Andra conducted geologic survey during the URL shaft construction and experiments in drifts at depths of 445 and 490 m. This program allowed consolidating the knowledge already acquired: geological environment, stability of the rock, containment properties and it confirms that the rock will maintain its qualities. The 2005 Progress Report presents the results of this phase. The main conclusion is that a potential disposal facility may be safely constructed over a zone around the URL, called transposition zone (about250 km2). The paper will present the most important results in this phase.

From 2006, the third phase of the program, the activities were oriented, inside the transposition zone, to determine a smaller zone in which à potential disposal facility could be designed. In 2009, Andra reported to the French authorities a proposal describing such a zone. In this paper the main results of this phase will be presented. Finally, next steps towards a final implementation will be given.

SESSION H2: Transportation, Storage and Waste Treatment

1) 40155 – Support of the Nuclear Research Institute Rez plc of the Shipment of Spent Nuclear Fuel from Research Reactors to the Russian Federation for Reprocessing in the Frame of the RRRFR Program

Josef Podlaha, Karel Svoboda, Nuclear research institute plc. (Czech Republic)

In 2007, spent nuclear fuel (SNF) from the Nuclear Research Institute Rez plc (NRI) was shipped to the Russian Federation for reprocessing. A large amount of SNF of Russian origin has been accumulated after 50 years of research reactor operation.

The shipment was realized in the frame of the Russian Research Reactor Fuel Return (RRRFR) program under the

-40-

Abstracts

US-Russian Global Threat Reduction Initiative (GTRI). SNF shipment from NRI to the Russian Federation (RF) represented a very complex and complicated technical, legal and contractual scope of work.

The SNF shipment has been realized under specific conditions: 1. High capacity SKODA VPVR/M casks were used for transportation for the first time. 2. For the first time, high enriched uranium SNF from a research reactor has been sent to the RF from a European Union country under the appropriate intergovernmental agreements, legal regulations and conditions.

NRI also participates in shipments of SNF from other countries within the framework of the RRRFR program. NRI`s participation consists of: - SKODA VPVR/M casks leasing, including service and maintenance inspections of the casks, - transportation of the casks to the research reactor site, -providing cask documentation in support of development of licenses, certificates, etc. for authorization of the shipment, - training of personnel in cask use and SNF loading, - technical oversight and expertise during the cask handling, fuel loading and cask closing and sealing, - the drying and helium leak testing of casks, -the return transportation of the empty casks from the Russian/Ukrainian border to NRI.

NRI participated in shipments of SNF from Bulgaria and Hungary in 2008, from Poland in 2009 and 2010 adn from Ukraine in 2010. Shipments from Belarus and Serbia are planned in 2010. The second shipment of the residue of high enriched SNF from NRI after changeover of the reactor operation to low enriched fuel will be implemented in 2013.

The experiences gained during the SNF transportation are described in the paper together with the present and future NRI activities in support of the SNF shipment from other countries.

2) 40177 – Long Term Storage of Nuclear Spent Fuel as Key Role of Japan's Nuclear Fuel Cycle until 2100: Cost and Benefit

Tadahiro Katsuta, Meiji University (Japan)

Political and technical advantages to introduce spent nuclear fuel interim storage into Japan's nuclear fuel cycle are examined. Once Rokkasho reprocessing plant starts to be operated, 80,000 tHM of spent Low Enriched Uranium (LEU) fuel must be stored in an Away From Reactor (AFR) interim storage site until 2100. If a following reprocessing plant starts the operation, the Spent LEU reaches its peak of 30,000 tHM before 2050, and then decreases until the end of the following reprocessing plant operation. Throughput of the following reprocessing plant is assumed as twice of Rokassho reprocessing plant case, namely 1,600tHM/year. On the other hand, three times of final disposal site of High Level Nuclear Waste (HLW) will be necessary with this condition. Besides, large amount of plutonium surplus will occur, even if First Breeder Reactors (FBR)s consume the plutonium. The maximum of plutonium surplus will reach almost 500 tons of plutonium. These results indicate that current nuclear policy does not solve the spent fuel problems but even complicate them. Thus, reprocessing policy could put off the problems in spent fuel interim storage capacity and other fears could appear such as difficulties in large amount of HLW final disposal management or separated plutonium management. If there is no reprocessing or MOX use, the amount of spent fuel will reach over 115,000 tones at the year of 2100. However, the spent fuel management could be simplified and also the cost and the security would be improved by using an interim storage primarily.

3) 40285 – TRU Recycling Options for increasing Protected Plutonium Production of FBR

Sidik Permana, Mitsutoshi Suzuki, JAEA (Japan)

The embodied challenges for introducing closed fuel cycle are utilizing advanced fuel reprocessing and fabrication facilities as well as nuclear nonproliferation aspect. Optimization target of advanced reactor design should be maintained properly to obtain high performance of safety, fuel breeding and reducing some long-lived and high level radioactivity of spent fuel by closed fuel cycle options. In this paper, the contribution of loading trans-uranium to the core performance, fuel production, and reduction of minor actinide in high level waste (HLW) have been investigated during reactor operation of large fast breeder reactor (FBR). Excess reactivity can be reduced by loading some minor actinide in the core which affect to the increase of fuel breeding capability, however, some small reduction values of breeding capability are obtained when minor actinides are loaded in the blanket regions. As a total composition, MA compositions are reduced by increasing operation time. Relatively smaller reduction value was obtained at end of operation by blanket regions (9 %) than core regions (15 %). In addition, adopting closed cycle of MA obtains better intrinsic aspect of nuclear nonproliferation based on the increase of even mass plutonium in the isotopic plutonium composition.

4) 40132 – Transuranic (TRU) Waste Volume Reduction Operation at a Plutonium Facility

Michael Cournoyer, Archie E. Nixon, Keith W. Fife, Arnold M. Sandoval, Vincent E. Garcia, Robert L. Dodge, LANL (USA)

Programmatic operations at the Los Alamos National Laboratory Plutonium Facility (TA-55) involve working with various amounts of plutonium and other highly toxic, alpha-emitting materials. The spread of radiological contamination on surfaces, airborne contamination, and excursions of contaminants into the operator’s breathing zone are prevented through use of a variety of gloveboxes (the glovebox, coupled with an adequate negative pressure gradient, provides primary confinement). Size-reduction operations on glovebox equipment are a common activity when a process has been discontinued and the room is being modified to support a new customer. The Actinide Processing Group at TA-55 uses one-meter or longer glass columns to process plutonium. Disposal of used columns is a challenge, since they must be

-41-

Abstracts

size-reduced to get them out of the glovebox. The task is a high-risk operation because the glass shards that are generated can puncture the bag-out bags, leather protectors, glovebox gloves, and the worker’s skin when completing the task. One of the Lessons Learned from these operations is that Laboratory management should critically evaluate each hazard and provide more effective measures to prevent personnel injury. A bag made of puncture-resistant material was one of these enhanced controls. We have investigated the effectiveness of these bags and have found that they safely and effectively permit glass objects to be reduced to small pieces with a plastic or rubber mallet; the waste can then be easily poured into a container for removal from the glovebox as non-compactable transuranic (TRU) waste. This size-reduction operation reduces solid TRU waste volume generation by almost 2½ times. Replacing one-time-use bag-out bags with multiple-use glass crushing bags also contributes to reducing generated waste. In addition, significant costs from contamination, cleanup, and preparation of incident documentation are avoided. This effort contributes to the Los Alamos National Laboratory Continuous Improvement Program by improving the efficiency, cost-effectiveness, and formality of glovebox operations. In this report, the technical issues, associated with implementing this process improvement are addressed, the results discussed, effectiveness of Lessons Learned evaluated, and waste savings presented.

5) 40188 – A Milestone in Vitrification - the Replacement of a Hot Metallic Crucible with a Cold Crucible Melter in a Hot Cell at the La Hague Plant

Sphiee Robert, Benoit Carpentier,SGN (France); Florence Gassot Guilbert, SGN Service procédé (France); Sandrine Naline, AREVA (France); Frédéric Gouyaud, AREVA NC (France); Christophe Girold, CEA (France)

Vitrification of high-level waste is the internationally recognized way to optimize the conditioned High Level Waste to be disposed of.

AREVA’s successful operation of AVM at the Marcoule plant, R7 and T7 at the La Hague Plants demonstrates the capabilities and experience of the Group to deploy innovative high level waste (HLW) processing technologies in industrial facilities, in partnership with R&D (CEA) and AREVA engineering (SGN). CEA and AREVA engineering team have continuously improved the hot metallic crucible melter vitrification technology through operational feedback aswell as ongoing research and developement. They have led the way in the development and implementation of Cold Crucible Induction Melter technology (CCIM).

The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass are molten by direct high frequency induction. This technology can handle with highly corrosive solutions and high operating temperature, which offers, among others advantages: o a great flexibility in matrix compositions, o the increase of the industrial glass production capacity, ° increase of melting temperature, ° allow higher waste loading, ° extend to a wide and varied waste types

In order to take advantage of CCIM, a new project, called “Vitrification 2010”, was launched in 2005 with the objective to put in commercial operation a CCIM in one of the existing and active R7 vitrification line in 2010. The main challenge for the project team is to replace the hot metallic crucible melter with the CCIM: o in an existing hot cell which is in operation for 20 years with high level waste solutions, o with respect to the current hot cell’s layout (calciner, through-wall for utility for example…).

The paper will illustrate why this event is a major milestone in vitrification and how the “Vitrification 2010” project succeeds in record time and without any impact on the La Hague Plant production.

6) 40265 – Adaptation of CCIM technology for HLW treatment. Results of research and development

Vladimir Lebedev, Alexander Kobelev, Fyodor Lifanov, Sergey Dmitriev, SIA Radon (Russia)

Results of research of possibility of Cold Crucible Inductive Melter (CCIM) technology application for HLW treatment (on examples of Savannah River (USA) HLW and PA “Mayak” (Russia) HLW), carried out in Moscow SIA RADON, and also results of development of new perspective facility for appliance in lab scale and then for real HLW treatment, are shown in this report. From 2003 till 2010 the research of possibility of different HLW treatment with CCIM technology are carried out. The principal possibility of vitrification of wastes with different compositions in the cold crucible was shown. The features of appliance of various materials, natural and industrially produced, as glass forming and modifying additions were explored. Charge compositions, made with using each researched sample of HLW imitator, were formulated and maximal HLW containing in charge was determined. Then the samples of glass obtained were explored. The main technological features of melting process were determined and principals of organization and control of the process were formulated. The pattern of cold crucible for HLW imitators treatment being included in experimental stand in PA “Mayak” was created. Also the System of Automatic Control for crucible was created. The tests and research of equipment work were carried out. On the base of analysis of data gained earlier and newly obtained the main requirements to designing of inductive melters and additional equipment, intended for real HLW treatment, were worked out. These requirements seem like a part of Task for designing of experimental-industrial HLW treatment plant, that is contemplated to built in PA “Mayak” in nearest time.

SESSION H3: Panel "Radwaste Human Resource Development to Support the Nuclear Renaissance"

Abstract Not Required

-42-

Abstracts

SESSION D1 : National and International D&D Programs

1) 40003 – Westinghouse PWR and BWR Reactor Vessel Segmentation Experience in Using Mechanical Cutting Process

Joseph Boucau, Stefan Fallström, Per Segerud, Paul Kreitman, Westinghouse Electric Belgium (Belgium)

Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both PWR’s and BWR’s and its process has been continuously improved over the years. This paper will describe the sequential steps required to segment, separate, and package each individual component, based on this mechanical cutting method. Detailed planning is essential to a successful project, and typically a “Segmentation and Packaging Plan” is prepared to document the effort. The usual method is to start at the end of the process, by evaluating the waste disposal requirements imposed by the waste disposal agency, what type and size of containers are available for the different disposal options, and working backwards to select the best cutting tools and finally the cut geometry required. These plans are made utilizing advanced 3-D CAD software to model the process. Another area where the modeling has proven invaluable is in determining the logistics of component placement and movement in the reactor cavity, which is typically very congested when all the internals are out of the reactor vessel in various stages of segmentation. The main objective of the segmentation and packaging plan is to determine the strategy for separating the highly activated components from the less activated material, so that they can be disposed of in the most cost effective manner. Usually, highly activated components cannot be shipped off-site, so they must be packaged such that they can be dry stored with the spent fuel in an Independent Spent Fuel Storage Installation (ISFSI). Less activated components can be shipped to an off-site disposal site depending on space availability. Several of the plants dismantled to date in the US have repackaged the less activated waste back into the reactor vessel and shipped the entire assembly to the disposal site. Decisions like these can be driven by many factors such as disposal costs, transportation logistics, licensing fees, etc., but will have a significant impact on the segmentation and packaging plan so they must be considered early in the planning phase. All segmentation tools are remotely controlled since the mechanical segmentation projects that Westinghouse has executed, so far, have been performed under water due to the high radiation levels. ALARA and personal safety is the number one priority during the site work. The complexity of the work requires well designed and reliable tools. Westinghouse has optimized the technologies from its experiences accumulated over the years. Its main focus has always been to improve tool handling and cutting speed, water cleanliness, fail-safe and safety aspects. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Mechanical cutting has a number of advantages compared to other cutting techniques. ? The technique produces almost no secondary waste. ? The visibility during cutting is very good because the cutting produces only a negligible amount of micro particles. ? Chips from the cutting process falls down to the bottom of the cutting pool and are easy to collect. ? No gases are produced that can cause airborne contamination. ? The technique is safe and reliable. ? All reactor internal sizes, materials and thicknesses can be cut.

Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. Westinghouse continues to develop new methods and products in order to further reduce the waste volume.

In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

2) 40130 – EPRI Nuclear Power Plant Decommissioning Technology Program

Karen Kim, Sean Bushart, Mike Naughton, Richard McGrath, Electrict Power Research Institute (USA)

The Electric Power Research Institute (EPRI) is a non-profit research organization that supports the energy industry. The Nuclear Power Plant Decommissioning Technology Program conducts research and develops technology for the safe and efficient decommissioning of nuclear power plants.

Several USA nuclear power plants shut down and entered active decommissioning in the 1990s. At that time the decommissioning of a commercial nuclear power plant was a first of a kind evolution; nuclear power plant decommissioning is a combination of industrial decommissioning with the complexities of radioactive materials and waste management. The first plants to undergo this process encountered various challenges related to decommissioning planning and project management, removal of large radioactive components, radiation protection, site remediation, final status survey, and license termination. The USA regulatory framework for nuclear power plant decommissioning developed and evolved along with these first decommissioning projects. As of 2009 these nuclear power plants have successfully completed decommissioning activities and have achieved license termination. Two major contributors to the successful completion of these projects were a) the development of technology specifically tailored to tasks unique to nuclear plant decommissioning and b) learning from each other’s experiences.

The EPRI Nuclear Power Plant Decommissioning Technology Program conducts research and develops technology that addresses the range of decommissioning tasks from pre-planning, site characterization, removal of large components,

-43-

Abstracts

site remediation, to license termination. A key feature of the EPRI Program is the collection and transfer of lessons learned and experiences from the major decommissioning projects conducted to date. The EPRI Program conducts research and development based on the needs of its utility members; the EPRI Program membership consists of utilities from the USA, France, Spain, Sweden, Italy, Taiwan, and Japan. EPRI is the mechanism through which these international utilities can conduct collaborative research and development and learn from each other’s experiences and lessons learned. In addition to collaborative research and development, the EPRI Program Team of decommissioning experts provides direct and site-specific support to its member utilities to facilitate knowledge transfer and decommissioning project optimization.

This presentation will provide information about the EPRI Nuclear Power Plant Decommissioning Technology Program and discuss USA and International Decommissioning experiences to date.

3) 40253 – Tokai-1 Decommissioning Project

Keizaburou Yoshino , JAPC (Japan)

Tokai-1 is the oldest and historical commercial Magnox reactor in Japan, which had started commercial operation in 1966. The unit had helped introduction and establishment of the construction and operation technologies regarding nuclear power plant at early stage in Japan by its construction and operating experiences. However, The Japan Atomic Power Company (JAPC), the operator and owner of Tokai-1, decided to cease its operation permanently because of a fulfillment of its mission and economical reason. The unit was finally ceased in March 1998 after about 32 year operation. It took about three years for removal of all spent fuels from the site, and then decommissioning started in 2001. JAPC, always on the forefront of the nation’s nuclear power generation, is now grappling Japan’s first decommissioning of a commercial nuclear power plant, striving to establish effective, advanced decommissioning. The decommissioning for Tokai-1 was scheduled as 20 years project. At the beginning, the reactor was started to be in a static condition for ten years (“safe storage period”). While the reactor had been safely stored, the phased decommissioning works started from non-radioactive or low radioactive equipments toward high radioactive equipments. First five years of the project, JAPC concentrated to drain and clean spent fuel cartridge cooling pond and to remove conventional equipments such as turbine, feed water pump and fuel charge machine as planed and budgeted.

From 2006, the project came into new phase. JAPC has been trying to remove four Steam Raising Units (SRUs). The SRUs are huge component (7ton, φ6.3m, H24.7m) of the Gas Cooling Reactor (GCR) and inside of the SRUs are radioactively contaminated. It is concerned that workers are required safety and minimizing contamination areas during SRU removal. Therefore, JAPC is developing and introducing Jack-down method and remote control multi functional removal system. This method is the method by which to remove the SRUs in turn from the bottom by lifting the SRU by a large jack and cutting for removing SRU is done remotely with this system. The system enables cutting and holding not only SRU body but also internals. This technology and experiences would be useful for the reactor removal in the near future.

4) 40289 – Activities of the OECD/NEA WPDD in the Field of Decommissioning

Claudio Pescatore, Patrick O'Sullivan, OECD/NEA

The OECD/NEA seeks to assist its member countries in developing strategies for the management of all types of radioactive material, including waste, that are safe and sustainable and that meet the broad needs of society - with particular emphasis on the management of long-lived waste and spent fuel and on decommissioning of disused nuclear facilities. The programme of work in the area of decommissioning is supervised by the Radioactive Waste Management Committee (RWMC) and carried out by the Working Party on Decommissioning and Dismantling (WPDD). The latter is made of senior representatives from regulatory authorities, decommissioning organisations, policy making bodies, and researchand-development institutions from the NEA countries. It includes also representatives of the IAEA and of the the European Commission. The WPDD supports the RWMC by keeping under review the policy, strategic, and regulatory aspects of decommissioning of phased-out nuclear installations. Its scope of work includes decommissioning and dismantling of shutdown facilities up to and including the release of the site, but excluding fuel removal, removal of nuclear processing fluids, post-operational clean out of fuel residues and removal of operational wastes. The work programme comprises activities in the following key areas: Policy, regulation and strategy; Funding and costs; Techniques; Decommissioning materials management and site release; and Human and organisational factors. Recent trends in the above areas, as noted in a recent analysis by the WPDD, include: • Policy, regulations and strategy – most countries have adopted policies for decommissioning and for funding provisioning but a range of options were being applied to the decommissioning of nuclear power plants (early dismantling, safe store and entombment), • Funding and costs – cost calculations for decommissioning are very sensitive to the assumed end-state and to the levels of contaminations on the sites and may be greatly influenced by stakeholder requirements, • Techniques – the focus of current R&D is on the development of innovative technology for segmentation, dismantling and concrete decontamination, together with better instrumentation for material management and control • Materials management and site release – there is a significant acceleration in the number of decommissioning projects, including provision of infrastructure for storage of materials and for undertaking clearance and recycling of disused materials; • Human and organisational factors – efforts are being made to structure decommissioning contracts to optimise the supply chain relationships, with greater transparency and better communication amongst all concerned parties; and • Stakeholder participation and knowledge management – there is a range of approaches to stakeholder involvement in developing decommissioning plans and to rule making. In general, this issue is gaining momentum. A number of specific activities,

-44-

Abstracts

projects and reports address the above areas. They will be reviewed for the ICEM-2010 audience

5) 40307 – French Decommissioning Feedback Experience and Lessons Learned

Jean-Guy Nokhamzon, CEA (France); Patrick O'Sullivan, OECD/NEA

The French (CEA, EDF and AREVA) experience in decommissioning of nuclear facilities goes back many decades and relates to many installations of very different types.

Some twenty facilities were dealt with by 2009, corresponding to around half of all the nuclear facilities permanently closed up to date, beginning in March 2001 with the decommissioning of AT1 facility at La Hague, the pilot plant used by the CEA in the seventies for the reprocessing of spent fuel from fast neutron reactors (former IAEA stage 3, excluding the civil engineering structures), as well as the blasting of G1 stack at Marcoule, on 19 July 2003 and ending with the total demolition of the research reactors Triton and Néréide in Fontenay aux Roses in 2004 and the final steps for the total decommissioning of Siloë, Siloette and Mélusine in Grenoble research Centre by 2012.

All the work done demonstrates the following: 1) Decommissioning can and has been done in a safe, cost-effective and environmentally friendly manner. 2) Current technologies have demonstrated their effectiveness and robust performance in numerous

decommissioning activities. Feedback experience on design, construction and operation is a considerable help for reliable planning, cost evaluation and successful realization of a decommissioning project.

3) During decommissioning radiological risks are very small in comparison to non-radiological risks. 4) The dissemination of best practices and sharing of information in international workshops, conferences and

especially within working groups as the OECD/NEA/CPD and IAEA/IDN has proven to be a good basis for an effective cooperation and support to master new challenges on decommissioning projects.

5) Future challenges will require further international cooperation to establish sustainable regulations and guidance to achieve objectives without being burdensome or overly conservative.

6) A consistent, internationally accepted rationale is necessary for the elaboration of concepts and for the derivation of numerical values on clearance, exemption and authorized releases.

7) With decommissioning moving towards being a fully mature industrial process, increased dialogue among regulators, implementers and international standards organisations is necessary.

6) 40032 – Decommissioning Planning for Swedish Operating NPPs

Gunnar Hedin, Mathias Edelborg, Niklas Bergh, Westinghouse Electric Sweden (Sweden); Jan Carlsson, Fredrik de la Gardie, SKB (Sweden)

Decommissioning studies have been carried out for the three BWR units of Oskarshamn and the three BWR units of Forsmark nuclear power plants. The final closure of these units is far ahead but anyhow there has been a need for developing a more general decommissioning planning basis. The main objectives of the studies have been to establish an estimate of the waste amounts arising from these units during decommissioning and dismantling as well as providing a firm basis for funding of the decommissioning phase for these units. The waste amounts will be used when designing a separate repository for decommissioning waste of the same type as the existing facility for final disposal of short-lived low- and intermediate level waste, the SFR, at Forsmark, Sweden. The broader studies are also used for verification that the existing national decommissioning fund is of an adequate size.

The studies have been performed by Westinghouse in cooperation with the utilities of Forsmark (FKA) and Oskarshamn (OKG), on behalf of SKB, the utility-owned Swedish waste management organization, responsible for coordination of the national waste fund as well as for designing, building and operation of waste management facilities.

Each of the studies contains a general description of the plant with the purpose to characterize them to facilitate decommissioning project comparative studies of different plants. Also, a detailed inventory of all materials in the plant have been put together and modeling of radioactivity data have been carried out to establish typical contamination levels for the main parts of the plant. The procedures and technologies foreseen to be used for the future dismantling and waste management work are briefly described based on standard, presently easily available techniques.

The combination of the materials inventory data and the radiological data is used to calculate the number of different waste containers that will be produced during dismantling of the plant. Standard 20 feet ISO containers will be used for the low-level waste while special steel containers that fit the transport overpack system will be used for intermediate level waste. The long-lived waste will be loaded into thick-walled steel containers.

The inventory data will also be used by experience models that calculate work hours for taking care of all the different types of plant components. The working time estimates are then combined, together with general duration data for different activities during plant decommissioning, into a time schedule for the complete program, from initial planning and preparatory activities to non-radioactive building demolition and site restoration.

Costs are then estimated for each of the work breakdown structure (WBS) elements of the time schedule. Summarized it provides a budgetary estimate for the complete decommissioning program. By rearranging the elements of the cost estimate according to the IAEA/EC/OECD-NEA common matrix (the “Yellow Book”), cost comparison with other plants is facilitated.

The paper will provide detailed data of waste characterization and amounts, as well as of time schedules and cost estimates for the main part of the Swedish BWR fleet and show how decommissioning will fit with the existing

-45-

Abstracts

well-developed waste management system of the country.

7) 40273 – Considerations for Grout Formulations in Facility Closures using In Situ Strategies

John Gladen, Mike Serrato, Chris Langton, Savannah River National Laboratory (USA); Andy Szilagyi, US DOE (USA)

The U.S. Department of Energy is conducting In Situ closures (entombment) at a large number of facilities throughout the complex. Among the largest closure actions currently underway is the closure of the P and R Reactors at the Savannah River Site. Below grade open spaces in these facilities are being stabilized with grout to ensure the long term structural integrity of the facility which ensures the permanent immobilization of residual contamination.

The large size and structural complexity of these facilities present a wide variety of challenges for the identification and selection of appropriate fill materials. Considerations for grout formulations must account for flowability, long term stability, set times, heat generation and interactions with materials within the structure. The large size and configuration of the facility necessitates that grout must be pumped from the exterior to the spaces to be filled, which requires that the material must retain a high degree of flowability to move through piping without clogging while achieving the required leveling properties at the pour site. Set times and curing properties must be controlled to meet operations schedules, while not generating sufficient heat to compromise the properties of the fill material.

The properties of residual materials can result in additional requirements for grout formulations. If significant quantities of aluminum are present in the facility, common formulations of highly alkaline grouts may not be appropriate because of the potential for hydrogen generation with the resultant risks. The SRS is developing specialized inorganic grout formulations that are designed to address this potential problem. One circum-neutral chemical grout formulation identified for initial consideration did not possess the proper chemical characteristics having exceptionally short set times and high heat of hydration. Research efforts are directed toward developing formulations that can meet operational requirements for chemical compatibility, extended set times and reduced heat generation.

SESSION R1: Environmental Impacts

1) 40298 – Main Results of A Remediation of Uranium- and CHC-Contaminated Groundwater

Jörg Wörner, Sonja Margraf, Walter Hackel, RD-Hanau (Germany)

The ground-water of a former research- and production site for various nuclear fuel-elements has been remediated for the last 8 years taking place after an intensive soilremediation by large-scale drilling and additional soil-exchange within the saturated zone. The remediation-activities were improved successfully by several measures of which one was the increase of the plant’s throughput and another the setting of additional remediation-wells at selected places. This procedure is mainly based on the regular observation of the remaining contaminants’ concentrations in the ground-water and the aquifer’s specific lithography. The cleaning process was adopted for a second installation which was taken in operation last November. Both installations’ experience will be presented by which about 450.000 m³ ground-water will have been extracted by midyear. The area affected by the contaminants has been downsized by the remediation which is manifested by the results of the monthly monitoring. The conventionally set target-values have not yet been reached, but the progress will offer the chance to release the remediation out of the rules of the German Atomic Law by the year 2013. Details in the ground-water’s Uranium-nuclide composition allow the attribution of the ground-water to two main plumes associated to different former production-activities. Blending of the two plumes with respect to their specific Uranium-nuclide composition caused by the chosen pump-regime has been observed and is meanwhile a helpful means for reaching the set aims on a planned time-scale. Figures for the totally extracted amounts of Uranium and CHC (chlorinated hydrocarbons) will be presented which are increasing actually for both contaminants as result of setting additional remediation-wells close to the still- observable contamination-centres. Completely remediated areas are observed on the other hand where intensive pumping was executed from the very beginning of the remediation-activities. These observations give data to estimate the residual amounts of ground-water which have to be treated in the future. The decrease of the Uranium-contamination is observed as a function of the extracted amount of ground-water whereas the CHC-contamination has decreased irregularly due to the fact of its dependence of the pump-regime and the remediation-wells’ different distances to the former handling locations. The CHC-compound-spectrum shows natural attenuation effects in the close down-stream area which can be helpful after achievement of the remediation’s target-value for the Uranium and possible finalizing of a continuous pumping procedure. As detailed results were published in two former ICEM-papers [1,2] this paper will highlight main results and discuss taken measures to finalize the remediation-activities within the next 3 to 4 years. [1] ICEM07-7270 Remediation of a Uranium-contamination in groundwater, [2] ICEM09-16244 Remediation of Uranium- and CHC-contaminated groundwater on a former nuclear fuel-element production site.

-46-

Abstracts

2) 40267 – Biogeochemical Gradients, Waste Site Evolution, and Implications for Sustained Metal and Radionuclide Attenuation in Complex Subsurface Environments

Karen Skulbal, Justin Marble, Kurt D. Gerdes, US DOE (USA); Miles Denham, Karen Vangelas, Savannah River National Laboratory (USA)

The U.S. Department of Energy (DOE) Office of Technology Innovation and Development sponsors applied research to improve the understanding of metal and radionuclide behavior in soil, sediment, and groundwater. DOE-supported researchers are developing advanced site characterization and simulation capabilities, as well as novel remediation tools and technologies to control contaminants over long time frames in highly complex subsurface environments. Natural and enhanced attenuation strategies are being studied in the context of subsurface biogeochemical gradients, such as variations in pH and redox potential. These gradients are spatially and temporally dynamic and can strongly influence metal and radionuclide migration, speciation, and reactivity. Knowledge of gradient evolution is essential for site evaluation, initial remedy selection, and technology transitioning to sustainable, low-energy remediation approaches such as monitored natural attenuation. Research activities in this area will be discussed using case studies from DOE sites. One case study focuses on a former disposal area at the Savannah River Site (SRS) in Aiken, South Carolina. Unlined seepage basins at the site received approximately seven billion liters of acidic, aqueous, low-level radioactive waste over more than three decades. The resulting groundwater plume contains multiple contaminants, including nitrate, iodine-129, strontium-90, technetium-99, tritium, and uranium isotopes. Since 1991, groundwater near the source area has been analyzed for evidence that plume acidity is naturally attenuating. Other biogeochemical gradients are also being evaluated to determine natural waste site evolution and impacts from active remediation systems. These systems include a neutralization barrier to mitigate plume acidity and a biostimulation zone to enhance in situ reductive metal precipitation. The influence of reactive facies on plume dynamics is also being examined using geophysical tools to detect these subsurface zones of unique mineralogy, hydrology, and microbiology. Site characterization data, plume dynamics, and contaminant behavior are incorporated into an evolving site conceptual model, which serves as the foundation for reactive transport modeling. This unique approach for site assessment is expected to improve the selection of remedial strategies and decision making for long-term environmental stewardship at SRS and other sites. The work is complementary to the U.S. Environmental Protection Agency’s recent guidance on monitored natural attenuation of inorganics.

3) 40260 – Current Mercury Distribution and Bioavailability in Floodplain Soils of Lower East Fork Popular Creek, Oak Ridge, Tennessee, USA

Fengxiang X. Han, Yi Su, David L. Monts, Mississippi State University (USA)

The objectives of this study were to investigate the current status of mercury distribution, speciation and bioavailability in the floodplain soils of Lower East Fork Poplar Creek (LEFPC) after decades of US Department of Energy’s remediation. Historically as part of its national security mission, the U.S. Department of Energy’s Y-12 National Security Facility in Oak Ridge, TN, USA acquired a significant fraction of the world’s supply of elemental mercury. During the 1950s and 1960s, a large amount of elemental mercury escaped confinement and is still present in the watershed surrounding the Y-12 facility. A series of remediation efforts have been deployed in the watersheds around the Oak Ridge site during the following years. The sampling fields were located in a floodplain of LEFPC of Oak Ridge, TN, USA. A series of surface soils (10-20 cm) were sampled from both wooded areas and wetland/grass land. Two 8x8 m fields were selected in the woodland. Five profiles each consisting of three layers were randomly taken from each field. The three layers were the surface layer at 0-10cm, subsurface layer at 50-60 cm, and bottom layer at 100-110 cm. Soil in both wood and wetland areas was well developed with a clear B horizon. The present study clearly shows that the total mercury in floodplain soils of LEFPC significantly decreased after the series of remediation. This study confirmed the long-term effectiveness of these remediation actions, especially after excavation of highly contaminated floodplain soils. However, the average total mercury level of all soil samples collected are in the range of 50-80 mg/kg, still significantly above toxic level (> 5mg/kg). Furthermore, contrary to conventional believing, the major mercury form in current soils of this particular area of floodplain of LEFPC is mainly in non-cinnabar mercury bound in clay minerals (after decades of remediation). The floodplains can act both as a medium-term sink and as long-term sources. Native North American earthworms (Diplocardia spp.) and adjacent soils were taken from each spot in each field. Our results show strong linear relationships between mercury concentrations in earthworms (both mature and immature groups) and non-cinnabar mercury form, while cinnabar mercury is less bioavailable to native earthworms. Earthworms may be used as a potential mercury ecological bioindicator (bio-marker) for demonstrating mercury bioavailability and ecotoxicity in the ecosystem. The long-term stability, mobility and bioavailability of mercury contaminants in these floodplains still needs to be monitored continuously and closely.

4) 40262 – Integrated Strategy to Address Hanford's Deep Vadose Zone Remediation Challenges

Mark B. Triplett, Mark D. Freshley, Michael J. Truex, Dawn Wellman, PNNL (USA); Kurt D. Gerdes, Briant L. Charboneau, John G. Morse, Robert W. Lober, US DOE (USA);

Glen B. Chronister, CH2M Hill Plateau Remediation Company (USA)

A vast majority of Hanford’s remaining in-ground contaminants reside in the vadose zone of the 200 Area Central

-47-

Abstracts

Plateau, where reprocessing operations occurred. The vadose zone at this location is comprised of about 75 meters of water-unsaturated, unconsolidated, stratified sediments above groundwater that discharges to the Columbia River. Contaminants in this zone originated from intentional discharges to cribs, retention basins, and trenches, and from unintended tank waste releases in the tank farms. The “deep vadose zone” is defined as the region below the practical depth of surface remedy influence (e.g., excavation or surface barrier). At the Hanford Site, this region poses unique challenges for characterization and remediation.

In 2008, the Department of Energy initiated a deep vadose zone treatability test to seek remedies for technetium-99 and uranium contamination. The treatability approach includes laboratory, modeling, and field tests building on previous work at the Hanford Site. Initial laboratory and field tests for technetium-99 have focused on use of desiccation which could be used in combination with an infiltration barrier to slow the transport of technetium-99 in the subsurface. For uranium contamination, reactive gas technologies are being tested as one component of the overall treatability test approach.

More recently, in recognition of the need for a broader strategy, the Department of Energy initiated an integrated study of the deep vadose zone to establish a technical basis for addressing risk-driving contaminants and supporting selection of remediation approaches for the deep vadose zone that are protective of groundwater. This is a cooperative effort that combines the resources and contributions of research scientists, technology developers, and remediation contractors.

The objectives of the deep vadose zone strategy are to: • Develop sufficient understanding of the nature and extent of deep vadose zone contamination and processes that

affect fate and transport; • Develop predictive capabilities for describing contaminant fate and transport as well as flux from the vadose

zone to the groundwater; • Develop, test and deploy effective methods for remediating contaminated areas; • Develop and deploy effective monitoring methods for assessing the performance of remedies and for

determining the long-term threat of contaminants to the groundwater.

5) 40235 – Advanced Remedial Methods for Metals and Radionuclides in Deep Vadose Zone Environments

Dawn Wellman, Shas Mattigod, Ann Miracle, Lirong Zhong, Danielle Jansik, PNNL (USA); Susan Hubbard, Yuxin Wu, LBNL (USA); Martin Foote, MSE (USA)

Deep vadose zone environments can be a primary source and pathway for contaminant migration to groundwater. These environments present unique characterization and remediation challenges that necessitate scrutiny and research. The thickness, depth, and intricacies of the deep vadose zone, combined with a lack of understanding of the key subsurface processes (e.g., biogeochemical and hydrologic) affecting contaminant migration, make it difficult to create validated conceptual and predictive models of subsurface flow dynamics and contaminant behavior across multiple scales. These factors also make it difficult to design and deploy sustainable remedial approaches and monitor long-term contaminant behavior after remedial actions.

Functionally, the methods for addressing contamination must remove and/or reduce transport of contaminants. This problem is particularly challenging in the arid western USA where the vadose zone is hundreds of feet thick, rendering transitional excavation methods exceedingly costly and ineffective. Delivery of remedial amendments is one of the most challenging and critical aspects for all remedy-based approaches. The conventional approach for delivery is through injection of aqueous remedial solutions. However, heterogeneous deep vadose zone environments present hydrologic and geochemical challenges which limit the effectiveness. Because the flow of solution infiltration is dominantly controlled by gravity and suction, injected liquid preferentially percolates through highly permeable pathways, by-passing low-permeability zones which frequently contain the majority of contamination. Moreover, the wetting front can readily mobilize and enhance contaminant transport to the underlying aquifer prior to stabilization. Development of innovative, in-situ technologies may be the only way to meet remedial action objectives and long-term stewardship goals.

Shear-thinning fluids (i.e., surfactants) can be used to lower the liquid surface tension and create stabile foams, which readily penetrate low permeability zones. Although surfactant foams have been utilized for subsurface mobilization efforts in the oil and gas industry, so far, the concept of using foams as a delivery mechanism for transporting remedial amendments into deep vadose zone environments to stabilize metal and long-lived radionuclide contaminants has not been explored. Foam flow can be directed by pressure gradients, rather than being dominated by gravity; and, foam delivery mechanisms limit the volume of water (< 5% vol.) required for remedy delivery and emplacement, thus mitigating contaminant mobilization. We will present the results of a numerical modeling and integrated laboratory- / intermediate-scale investigation to simulate, develop, demonstrate, and monitor (i.e. advanced geophysical techniques and advanced predictive biomarkers) foam-based delivery of remedial amendments to remediate metals and radionuclides in vadose zone environments.

-48-

Abstracts

SESSION M1: Environmental Management

1) 40086 – Legacy Management: Turning Liabilities into Assets

Joe Legare, Eric Olson, S.M. Stoller Corporation (USA)

The U.S. Department of Energy (DOE) Office of Legacy Management (LM) is responsible for stewardship of over 85 sites across the USA dating back to the Manhattan Project and associated with the nuclear weapons production mission. This responsibility for the long-term environmental management of these sites includes the key requirements to ensure protection of human health and protection of the environment from residual contamination, and to manage over 100,000 ft3 of associated records.

The Legacy Management program, through its prime contractor, provides program management, field and technical support, data collection and analysis and project planning and implementation in support of the DOE to achieve its mission. With a defined program baseline of 75 years, and a theoretical baseline that may extend beyond this time period, effective programmatic integration of these myriad sites, and appropriate consideration of life-cycle implications to risk and cost of stewardship decisions is essential to the success of the program.

This presentation will provide an overview of the more than 85 sites in the Legacy Management program, including a discussion of prior mission and associated environmental contamination issues; ramifications of the different regulatory and cleanup approaches taken during active remediation; and the challenges of the current long-term surveillance and maintenance mission. Additionally, the presentation will address lessons learned during the first eight years of the Legacy Management program in terms of risk management, public and stakeholder involvement, institutional controls, regulatory agreements and the issues associated with transition from active remediation to long-term surveillance and maintenance. Some of the case histories of how environmental management of these varied sites has evolved over time will be presented.

2) 40270 – Lessons Learned in Planning the Canadian Nuclear Legacy Liabilities Program

Michael Stephens, Sheila M. Brooks, Joan Miller, Robert Mason, AECL (Canada)

In 2006, AECL and Natural Resources Canada (NRCan) began implementing a Nuclear Legacy Liabilities Program (NLLP) to manage the liabilities at AECL’s nuclear sites in Canada. These liabilities include shutdown research and prototype power reactors, fuel-handling facilities, radiochemical laboratories, support buildings, radioactive waste storage facilities, and contaminated lands. The delivery of the Program is managed by a Liability Management Unit (LMU) within AECL.

The NLLP comprises a large program of interlinked decommissioning, waste management and environmental restoration activities being executed at different sites, and by various technical groups as suppliers to the LMU. This paper describes the lessons learned in planning the “start-up” phase, which will conclude 2011 March, and the planning of the second phase, a 5-year program, which is currently being finalized.

The initial “start-up” phase was planned by a small group of experts. Although effort was made to communicate the goals and overall strategy of the Program to the groups that would carry out the work, progress was slower than anticipated because AECL was ramping up from a minimal maintenance mode and the required increase in staff and technical resources was not sufficiently understood. Many of the projects being addressed within the Program are one-of-a-kind, and the essential base information on which to prepare detailed execution plans was not available to accurately plan the work.

Internal reviews of the Program examined progress and identified several improvements to planning. These included strengthening communications, conducting more advance planning of the interlinked activities, and building flexibility into the commitments made about activities that had yet to reach major decision points.

The priorities for executing the required activities in the Program were set using criteria based on risks the liabilities presented to health, safety, the environment and to AECL’s business. In future, the decision criteria will also include the value gained for funds expended, and greater consideration will be given to mitigating risks to the execution of the Program. It was also determined that licensing strategies and processes need to be well-defined, and waste characterization methods and disposition pathways must be in place to deal with the wastes the Program will generate.

Case studies will be presented in the paper to illustrate these lessons learned. The NLLP is funded by the Government of Canada through NRCan.

3) 40218 – RFID Technology for Environmental Remediation and Radioactive Waste Management

Hanchung Tsai, Yung Liu, ANL (USA); James Shuler, US DOE (USA)

We have developed an advanced Radio Frequency IDentification (RFID) system capable of tracking and monitoring a wide range of materials and components for the nuclear industry – from fissionable stocks to radioactive wastes. The RFID technology has a number of advantages, such as enhanced safety and security, reduced personnel exposure to radiations, and improved inventory control. The sensors in the RFID tags monitor the state of health (e.g., temperature, shock, seal, dose, and dose rate) of the item and send out an alarm instantly when the sensor threshold is violated. Nonvolatile memories in the tag store data from sensors and event records, as well as manifest information if so desired. In irradiation tests, the tag electronics was confirmed to possess significant radiation

-49-

Abstracts

resistance and, therefore, would yield a satisfactory service life. Long-life batteries and smart management circuitries permit the RFID tags to operate for up to 10 years without a battery replacement. The form factor of the tags can be modified to suit different container types. The read range can be up to 100 m, and no line-of-sight between the tagged items and the interrogator (reader) is necessary. With careful implementation, even a large-size processing or storage facility with a complex configuration can be monitored with a handful of readers in a network. In transportation, by using Global Positioning System (GPS) and satellite/cellular communication protocols, the locations and the conditions of the tagged containers can be continuously tracked. The RFID system also integrates Geographic Information Systems (GIS) technology, which uses information in preexisting geodatabases to generate and issue reports instantly to first responders for situation recovery in case of incidents. In stand-alone applications, the monitoring and tracking data are contained within the local control computer; with a secure Internet connection, multiple users can share the data in real time within the complex or beyond. As with the deployment of any new technology, overcoming the cultural resistance is part of the developmental process. With a strong institutional support and multiple convincing live demonstrations, the cultural resistance has been largely overcome. As a result, implementation of the RFID technology is taking place at several of U.S. Department of Energy installations for processing, storage, and transportation applications.

4) 40181 – The Radioactivity of 3H in Metals by a High Temperature Furnace and a Liquid Scintillation Counter

Hee Reyoung Kim, Geun Sik Choi, Sang Yun Park, Chang Woo Lee, Moon Hee Han, KAERI (Korea Rep.)

The radioactivity of 3H of the metal samples from the nuclear sites was analyzed by using a commercialized high temperature furnace and a Liquid Scintillation Counter (LSC). The 3H activity of the sample was measured according to the duration of the high temperature combustion and the oxidation temperature. Basically, the recovery from the furnace was 90% for 3H and the LSC had a quenching efficiency of approximately 30 %. HNO3 was used as a trapping solution for 3H and the solution was cocktailed with a scintillator. The activity extracted from the sample was increased till the combustion time elapsed 60 minutes and the increasing rate was reduced continuously thereafter at 600 degree in centigrade whereas 80 % of radioactivity was extracted during the first 15 minutes at 900 degree in centigrade. Also, the pretreatment for the metal sample, which included a high temperature combustion and trapping, had the time required of at least four hours at 900 degree in centigrade. Finally, it was suggested that this high temperature combustion method could be applied to analyze the activity of the radioactive metal waste from the nuclear power plants.

5) 40275 – Next Generation Waste Glass Melters in the U.S. DOE Waste Processing Program

Steven P. Schneider, Gary Smith, US DOE (USA)

The USA Department of Energy (U.S. DOE) Office of Environmental Management (EM) is evaluating alternative options for waste glass melting technologies. Specifically, DOEEM is assessing advanced melter technologies and developing a comprehensive research plan for next generation waste glass melter design and demonstration. Resolution of the USA nuclear waste legacy requires the design, construction and operation of large and technically complex one-of-a-kind processing facilities coupled to equally complex waste treatment and vitrification facilities. The loading of nuclear waste into glass and the glass production rates at U.S. vitrification facilities are limited by the current melter technology. Significant reductions in glass volumes for disposal and mission life are only possible with advancements in melter technology and glass formulations. Melters with higher throughput rate may shorten cleanup mission, in addition melters that allow for higher waste loading in glass may significantly reduce lifecycle costs.

To help focus the next generation waste glass melter program, DOE-EM convened an international workshop to assess nuclear waste melter technologies and have used the melter workshop to help develop a comprehensive research plan for melter design and demonstration. The workshop included both oral presentations and discussion sessions from waste glass melter experts from around the world to assess the "state of the art" in melter technology and to lay the groundwork for a program plan that includes evolutionary changes to existing Joule-heated ceramic-lined liquid fed melters, as well as transformational melter technologies such as induction and hybrid-heated systems. At that workshop, representatives from many nations and international organizations (IAEA, China, France, Germany, India, Japan, Korea, Russia, UK, and the USA), universities (Catholic University of America, Missouri University of Science and Technology), and private companies (EnergySolutions, Kurion, URS) met to assess advanced melter technologies which helped the U.S. develop a comprehensive research plan for advanced waste glass melter design and demonstration with the goal of improved performance and reduced cost. The U.S. DOE Next Generation Waste Glass Melter Program is discussed in this paper.

SESSION L2: Solidification and Package (1)

1) 40021 – Commercialization Project of Ulchin Vitrification

Hyun-jun Jo, Cheon-Woo Kim, KHNP (Korea Rep.); Tae-Won Hwang, Nuclear Engineering & Technology Institure (Korea Rep.)

The Ulchin Vitrification Facility (UVF), to be used for the vitirification of low-and intermediate-level radioactive

-50-

Abstracts

waste (LILW) generated by nuclear power plants (NPPs), is the world’s first commercial facility using Cold Crucible Induction Melter (CCIM) technology. The construction of the facility was begun in 2005 and was completed in 2007. From December 2007 to September 2009, all key performance tests, such as the system functional test, the cold test, the hot test, and the real waste test, were successfully carried out. The UVF commenced commercial operation in October 2009 for the vitrification of radioactive waste.

2) 40023 – Plasma Gasification/Vitrification of Wet ILW

Gary Hanus, John Williams, Matt Zirbes, Phoenix Solutions Co. (USA)

Magnox South Ltd has authorized a variety of decommissioning programmes to evaluate various technologies for timely and cost-effective remediation of a spectrum of waste streams resulting from the operation of their reactors. Of particular current interest is wet, intermediate level waste (ILW) in the form of solids, sludges and liquids. Hinkley Point A has over 137,000 litres of ILW organic cation resin containing significant quantities of radio-cesium (Cs). In 2009 Phoenix Solutions Co was awarded a contract by Magnox South Ltd to demonstrate the effectiveness of thermal plasma treatment of this wet ILW resin waste stream.

The objective of the wet ILW plasma treatment project was to feed organic resin sludge (containing a cesium surrogate) and a borosilicate glass frit simultaneously into a thermal plasma reactor to demonstrate the gasification of the organic content of the resin while capturing the cesium within a molten glass bath. A total of approximately 200 litres of glass product were produced. Approximately 5 metric tonnes of organic resin simulant were provided to Phoenix Solutions Co consisting of a mixture of 68% water / NaOH solution, 30% organic cation resin material with bound Cs and 2% other, including a small proportion of cesium in solution. The off-gas species from the plasma reactor exhaust were sampled by an independent, qualified sampling organization focusing on VOCs, HCl, NOx, SOx, CO, CO2, dioxins and furans as well as particulate levels. Glass samples were obtained and analyzed for crystalline inclusions, elemental identification, and viscosity characteristics at temperatures near the process melt conditions. The process temperature was maintained near 1000 ºC while the molten glass bath was held between 1100 and 1200 ºC. The processing produced 200 litres of glass, 150 litres of which were successfully tapped from the plasma reactor into a standard Sellafield vitrified HLW disposal canister.

This paper will describe the basic waste processing approach, the process hardware utilized, the process control features and the test results. Several trial tests were conducted, the longest of which processed over 2 metric tonnes of dewatered resin waste together with 400 kg of glass. Further testing with improved process controls demonstrated an increase in cesium retention in the glass product.

3) 40026 – Solidification Of Simulated Liquid Waste Of Primary Loop Resin Elution Process Of PWR

Masamichi Obata, Michitaka Saso, Masaaki Kaneko, Nobuhito Ogaki, Taichi Horimoto, Toshiba corporation (Japan);Toshikazu Waki, The Kansai Electirc Power Co., Inc. (Japan)

Primary loop resin waste is eluted by sulfuric acid in The Kansai Electic Company Mihama,Takahama and Oi nuclear power station. Waste solution from this elution process is planned to be solidified by cement. This study bring out a range of chemical composition and crud concentration of waste solution from this elution process, and examine the properties of alumina cement solidification process and solidified material. Test for sulfate ion, borate, lithium, ammonium ion was carried out. Volume reduction ratio of over 0.5 was archieved for 5 to 25wt% of sulfate ion and <5,000ppm of borate.Lithium ion restrained the solidification delay by borate. Also, ammonium ion shows no significant effect. Based on this study, we concluded that the aluminum cement is applicable to all range of composition of waste solution from the resin elution process. We are planning the real waste solidification test and alumina cement solidification system/equipment design. This study is a part of commited work of The Kansai electric company.

4) 40108 – VUJE Experience with Cementation of Liquid and Wet Radioactive Waste

Kamil Kravarik, Zuzana Holická, Anton Pekár, Milan Žatkulák, VUJE, Inc. (Slovakia)

Liquid and wet LLW generated during operation as well as decommissioning of NPPs is treated with different methods and fixed in a suitable fixation matrix so that a final product meets required criteria for its disposal in a final repository. Cementation is an important process used for fixation of liquid and wet radioactive waste such as concentrate, spent resins and sludge. Active cement grout is also used for fixation of low level solid radioactive waste loaded in final packing containers. VUJE Inc. has been engaged in research of cementation for long. The laboratory for analyzing radioactive waste properties, prescription of cementation formulation and estimation of final cement product properties has been established. Experimental, semi-production cementation plant has been built to optimize operation parameters of cementation. VUJE experience with cementation of liquid and wet LLW is described in the presented paper. VUJE has assisted in commissioning of Jaslovské Bohunice Treatment Centre. Cement formulations for treatment of concentrate, spent resins and sludge have been developed. Research studies on the stability of a final concrete packaging container for disposal in repository have been performed. Gained experience has been further utilized for design and manufacture of several cementation plants for treatment of various liquid and wet LLW. Their main technological and technical parameters as well as characterization of treated waste are described in the paper. Applications include the Mochovce Final Treatment Centre, Movable Cementation Facility utilizing in-drum mixing for treatment of sludge, Cementation

-51-

Abstracts

Facility for treatment of tritiated water in Latvia and Cementation Facility for fixation of liquid and solid institutional radioactive waste in Bulgaria, which utilizes lost stirrer mixer. Key words: Cementation process, cementation formulation, concentrate, spent resins, sludge, cementation plant, in-drum mixing, lost stirrer mixer.

5) 40112 – Study of LPOP residue on spent-resin mineralization and solidification

Gen-ichi Katagiri, Morio Fujisawa, Fuji Electric Systems Co., Ltd. (Japan); Kazuya Sano, Norikazu Higashiura, JAEA (Japan)

The used ion exchange resin generated as radioactive waste in water purifying system at nuclear power stations or related facilities is currently stored in the sites, but its volume is increasing year by year. The used resin is planned to be buried in the future, the establishment of a disposal technique, which enables volume-reduction and stabilization for final disposal, is required. Fuji Electric had developed Low pressure oxygen plasma (LPOP) technology for mild decomposition and mineralization of an organic material such as ion exchange resin. LPOP method is suitable for radioactive used resin volume reduction and stabilization for final disposal. On LPOP process, the ion-exchange resins are vaporized and decomposed into gas-phase with pyrolysis, and then, they are decomposed and oxidized with low-pressure plasma activity based on oxygen. And this process is achieved under moderate condition for radio active waste. ?-incinerate temperature: 400 – 700deg C ?-low-pressure (low-temperature) plasma condition: 10-50 Pa From the result of LPOP process by the inductively coupled plasma, we have confirmed that the process is applicable for organic fireproof waste including ion-exchange resin, and found that the used resin treatment performance is the same as cold test (using imitate spent resin) as mentioned below. (1)LPOP attained weight reduction over 95% for the used resins from heavy water system and condensate demineralizer, and over 90% for (metal-rich) chemical decontamination resins. (2)Co-60 carry-over to the waste gas system is less than 10-4 g/g. (3)C, H, N, and S elements were completely removed from the resins by the LPOP treatment. In this paper, the outline of the LPOP technology, and two research results on the possibility of solidification with cement of LPOP residue for final disposes are reported. (1)Study of the residue chemical form after LPOP process - On sample of resin exchanged with transition metals ion, more than 90wt% element of transition metals in the residue is transformed to oxide. -On sample of resin exchanged with alkali metals ion, 70-95wt% element of alkali metals in the residue is transformed to sulfate. (2)Study of the solidification character with cement -On sample of resin exchanged with much cobalt and mixed with Fe2O3 as crud material, LPOP residuum is easily mixed with cement, and solidity with high strength can be obtained.

SESSION L3: Nuclide Assay

1) 40167 – Feasibility Study on the Nuclide Analysis of the Radwaste Drum Using the Spectrum to Dose Conversion Factor

Young-Yong Ji, Dae-Seok Hong, Tae-kuk Kim, Woo-Seog Ryu, KAERI (Korea Rep.)

There are several methods for non-destructive assay of a radwaste drum that are based on the gamma ray scanning and the in-situ objet counting system. Although these methods can be processed using high resolution, shielded detector and specific correction techniques to quantitatively analyze the nuclides in a drum, time and cost constraints compared with their accuracy dictate the use of simpler method to apply. Dose to curie (DTC) conversion method can simply and easily provide a reasonable estimate of the nuclide inventory in a radwaste drum. The measured dose rate as well as the relative abundance of gamma nuclides in a drum is a very important factor to be appreciated at the DTC conversion method because of the direct linearity between the measured dose rate and the gamma emitters in a drum. The dose rate is directly measured with the field detector. However, the relative abundance of gamma nuclides in a drum to be assayed is determined form an indirect measurement using the material balance by the waste stream. The uncertainty of the nuclide inventory of the assayed drum from the DTC conversion method could be increased because of the different detection mechanism between the dose rate and the relative abundance of gamma emitters in a drum. Unfortunately, that expands the limitation of using the DTC method. It is, therefore, necessary to find out a suitable measurement method of which two variables could be obtained from the drum to be assayed at once. This method could be realized by using the dose conversion factor, which has been widely using in the field of the environmental radiation measurement. The dose rate from a drum could be directly calculated from the measured gamma ray spectra by using the dose conversion factor, and also, the relative abundance of gamma nuclides could be easily obtained from the net count peaks in the spectra. In this study, the dose conversion factor for 3”?x3” NaI(Tl) scintillation detector around lead shield with the thickness of about 3 cm was calculated by a MCNP code. The experimental verification for using this dose conversion factor was performed by using the simulated drum that has several holes for locating a standard source.

2) 40255 – Portable Non-Destructive Assay Methods for Screening and Segregation of Radioactive Waste

Alan Simpson, Martin Clapham, Stephanie Jones, Randy Lucero, Pajarito Scientific Corporation (UK)

Significant cost-savings and operational efficiency may be realised by performing rapid nondestructive classification of radioactive waste at or near its point of retrieval or generation. There is often a need to quickly

-52-

Abstracts

categorize and segregate bulk containers (drums, crates etc.) into waste streams defined at various boundary levels (based on its radioactive hazard) in order to meet disposal regulations and consignor waste acceptance criteria.

Recent improvements in gamma spectroscopy technologies have provided the capability to perform rapid in-situ analysis using portable and hand-held devices such as battery-operated medium and high resolution detectors including lanthanum halide and high purity germanium (HPGe). Instruments and technologies that were previously the domain of complex lab systems are now widely available as touch-screen “off-the-shelf” units. Despite such advances, the task of waste stream screening and segregation remains a complex exercise requiring a detailed understanding of programmatic requirements and, in particular, the capability to ensure data quality when operating in the field. This is particularly so when surveying historical waste drums and crates containing heterogeneous debris of unknown composition. The most widely used portable assay method is based upon far-field High Resolution Gamma Spectroscopy (HRGS) assay using HPGe detectors together with a well engineered deployment cart (such as the PSC TechniCART™ technology). Hand-held Sodium Iodide (NaI) detectors are often also deployed and may also be used to supplement the HPGe measurements in locating hot spots. Portable neutron slab monitors may also be utilised in cases where gamma measurements alone are not suitable.

Several case histories are discussed at various sites where this equipment has been used for in-situ characterization of debris waste, sludge, soil, high activity waste, depleted and enriched uranium, heat source and weapons grade plutonium, fission products, activation products, americium, curium and other more exotic nuclides. The process of acquiring and analyzing data together with integration of historical knowledge to resolve and delineate waste streams (for example between low-level waste and transuranic waste) is described.

3) 40093 – Alpha radioactivity monitor using ionized air transportation for large size uranium waste (1) - Large measurement chamber and evaluation of detection performance –

Susumu Naito, Shuji Yamamoto, Miikio Izumi, Yosuke Hirata, Yukio Yoshimura, Tatsuyuki Maekawa,Toshiba Corporation (Japan)

Massive amounts of waste contaminated with alpha-radioactive uranium have accumulated in back-end facilities of the nuclear fuel cycle. In order to dispose of it adequately, it is necessary to classify such waste according to its so-called clearance level. However, when measuring alpha radioactivity using a conventional survey meter, one must hold it as close as possible to the surface, due to the relatively short flight range of alpha particles in air. As a result, measuring a large amount of diverse waste takes a long time and involves high labor costs. Therefore, a new efficient alpha measurement technique is strongly desired. To satisfy this demand, we have been developing an alpha radioactivity monitor based on the principle of alpha radioactivity measurement using ionized air transportation. The measurement principle is as follows. Air is ionized by alpha particles emitted from the waste. The produced ions are transported to an ion sensor with the air flow produced by a blower, where the number of ions is measured as an electric current (ion current). The alpha radioactivity is then evaluated from the ion current value. This indirect measurement method is very efficient, because the ions produced near the entire surface of the waste can be detected both all at once and at a distant position from the waste. In previous work, we developed a prototype monitor with an about 1000 mm cubic measurement chamber to measure the cut waste. However, in a survey of target waste, we found that it is desired to measure not only the cut waste but also the lengthy waste such as uncut cylinders. Therefore, we developed an alpha radioactivity monitor with a long and large measurement chamber (effective sizes: 500 mm x 900 mm x 3200 mm) for the long and large cylindrically-shaped waste (maximum size: 300 mm in diameter and 3000 mm in length, weight: 10 to 200 kg). We aimed <1000 Bq as the target value of alpha radioactivity detection limit (ADL), which is one-tenth of the clearance level (1 Bq/g) for 10 kg waste. The issue to size up the measurement chamber was to suppress the rise of ADL. To achieve this, we optimized the prototype design. The prototype showed 60 to 120 Bq of ADL for simulated waste with several cylindrical shapes and they sufficiently satisfied the target ADL. In conclusion, this monitor has sufficient performance to perform the clearance level inspection for large uranium waste.

4) 40091 – Alpha Radioactivity Monitor using Ionized Air Transport Technology for Large Size Uranium Waste (2) - Simulation model reinforcement for practical apparatus design –

Takatoshi Asada, Yosuke Hirata, Susumu Naito, Miikio Izumi, Yukio Yoshimura, Toshiba Corpoation (Japan)

In alpha radioactivity measurement using ionized air transportation (AMAT), conversion from ion currents to radioactivity with sufficient accuracy is required. However, the ion transport from the measurement chamber to the ion sensor is subject to change depending on the location or the shape of the waste since the ion loss is sensitive to the flow condition near the contamination. Measured ion currents can be different even for the same target shape, depending on the contaminated location. For operational convenience the conversion factor should be determined in advance. Direct measurement using simulated radioactive wastes is straightforward and convincing. Practically, however, the contaminated location is usually unknown, and the waste shape is not always the same as pre-measured samples. Measurement, from this viewpoint, is costly if comprehensive measurement is inevitable. An accurate ion transport simulation provides ways of complementarily determining conversion factors. We have developed an ion transport simulation model based on CFD including ion diffusion and loss. Simulation results were compared with experiments with relatively high air speed, namely 1 – 4 m/s, at the alpha source, achieving good agreement. In a practical AMAT apparatus capable of measuring a multiple wastes at once for operation efficiency, the flow condition can be more complex. The air flow at the alpha source may be slower than 1 m/s or even stagnant, and ion loss is likely to be large.

-53-

Abstracts

Reinforcement of the ion transport model to cover the lower air speed region is, therefore, very effective. Ions are generated by an alpha particle in a very thin column with a radius of a few micro-meters and a height of about 0.05 m. Since the ion density at this temporal stage is very high, the recombination loss, proportional to the square of ion density, is exclusively dominant within a few milli-seconds after the ion generation. The spatial and temporal scales of this columnar recombination are too small for CFD simulation. We, therefore, solve an ion transport equation during the period of columnar recombination with diffusion and recombination terms and incorporated the relation between ion loss and turbulent parameters of the air flow into CFD. Using this new CFD model, simulations have been done for various air speeds and targets. Dependence of obtained ion currents on the air speed shows improved accuracy at low air speed. Those for various shapes and numbers also agree with experiments, showing improvement of simulation accuracy.

5) 40111 – Preparation of Reference Materials on Radiochemical Analysis for Low-Level Radioactive Waste Generated from Japan Atomic Energy Agency

Ken-ichiro Ishimori, Mikio Nakashima, Kuniaki Takahashi, Yutaka Kameo, JAEA (Japan)

We have advanced the development of simple and rapid determination method for important radionuclides on safety assessment for disposal of low-level radioactive wastes generated from Japan Atomic Energy Agency (JAEA). In the radiochemical analyses of the radioactive wastes, it is necessary to manage accuracy and precision of determined values of radioactivity concentration using reference materials. However, since appropriate reference materials of radioactive waste are hardly available at the present state, the developments of laboratory-scale preparation methods are required to supply reference materials. In this work, we investigated preparation methods for the reference materials containing important nuclides and confirmed the validity of the prepared materials. Additionally a reference material for cemented liquid waste was also prepared. --Solidified product containing alpha-ray and gamma-ray emitting nuclides--In waste management in JAEA, non-metallic low-level radioactive solid wastes will be treated by plasma melting at the Advanced Volume Reduction Facilities (AVRF). In order to clarify optimum melting conditions of solidified products using a laboratory-scale electric furnace instead of plasma heating device, we conducted melting tests of a miscellaneous simulated solid waste in the presence of stable isotope tracers. Over 90% of Cs remained in the solidified product by keeping the basicity (CaO[wt%] / SiO2[wt%]) to be 0.05. Under the optimum melting conditions, we prepared reference materials containing alpha-ray (237Np, 241Am, and 244Cm) and gamma-ray (60Co, 137Cs, and 152Eu) emitting nuclides. The characteristics observed in SEM-EDX measurement and chemical durability against acids suggested that glass structure of the reference materials was almost same as that of solidified products produced by plasma melting. --Solidified product containing 14C or 36Cl-- Since 14C and 36Cl easily vaporize at a high temperature, it is difficult to remain the nuclides in solidified product on melting treatment with an electric furnace. First, melting conditions for solidified glasses containing N or Cl were optimized. We attempted a preparation method which produces 14C or 36Cl in the solidified glass using nuclear reaction 14N(n, p)14C or 35Cl(n, gamma)36Cl by thermal neutron irradiation. Reference materials containing 14C or 36Cl were successfully prepared by the proposed method. The radioactivity concentrations of the reference materials were evaluated from the developed simple and rapid determination method. From the results, it was confirmed that reference materials, which was useful for routine radiochemical analysis, could be successfully prepared on the present preparation methods.

SESSION H4: National and International Programs (2)

1) 40213 – U.S. NRC Integrated Spent Fuel Management Plan

Catherine Haney, Shawn Smith, US NRC (USA)

The U.S. Nuclear Regulatory Commission (NRC) is developing an integrated plan for regulating the interrelated activities that are involved in the management of spent fuel and high-level waste. Treating the system as a whole is essential for several reasons: (1) decisions made about one component or activity of the waste management system could significantly affect other components (e.g., a decision made in isolation could inadvertently impact alternatives to the system as a whole); (2) treating waste management as a system is a more efficient and effective way to determine priorities, logically complete activities or to appropriately deal with unexpected situations that may arise with first-of-a-kind programs like reprocessing or alternative waste disposal options; and (3) viewing the system as a whole avoids gaps and unnecessary duplication in regulations leading to more effective and efficient development and application of regulatory oversight.

Near-term flexibility is a key consideration because the national policy in the USA appears to be changing and will likely remain in flux for some time. For example, the Secretary of Energy has convened a commission to conduct a comprehensive review of the policies for managing the back end of the nuclear fuel cycle. NRC activities will be informed by this and other relevant developments as it develops its plans. However, integration is essential regardless of the direction of national policy, as the NRC needs to remain flexible and agile under a range of policy outcomes. This presentation will describe key aspects of the plan and how the plan will be implemented.

-54-

Abstracts

2) 40116 – Regulatory Research for Geological Disposal of High-level Radioactive Waste in Japan

Shinichi Nakayama, JAEA (Japan); Yoshio Watanabe, AIST (Japan); Masami Kato, JNES (Japan)

The Nuclear and Industrial Safety Agency of the Ministry of Economy, Trade and Industry (NISA) has renewed its regulatory role and supporting research needs, taking into consideration recent circum-stances of radioactive waste management in Japan. In October 2009, a technical support organization of the Japan Nuclear Energy Safety Organization (JNES) released its independent five-year research plan, “Regulatory Support Research Plan on Radioactive Waste Management 2010-2014”, in cooperation with the research institutes of Japan Atomic Energy Agency (JAEA) and the National Institute of Advanced In-dustrial Science and Technology (AIST). The plan covers low- through high-level radioactive waste management. The geological disposal research plan and the future research activities are outlined in this paper. Japan’s nuclear regulation law, “Law for Regulation of Nuclear Source Materials, Nuclear Fuel Ma-terials and Nuclear Reactors”, was amended in 2007 to address the safety of geologic disposal of high-level radioactive waste. NISA announced its involvement or supervision not only for the licensing application but during the site selection process. Two major research areas required by NISA have been identified: studies to review the validity of preliminary survey results during the site selection, and studies to review the safety assessment for a licensing application. These research areas are aimed at 1) developing “safety indicators” to judge the adequacy of site investigation results presented by the operator, 2) compiling basic requirements of safety design and safety assessment needed to make a technical evaluation of the license application, as well as developing safety indicators for objective evaluation, and 3) developing an inde-pendent safety assessment methodology. In addition, NISA plans to periodically issue the report, “The Regulatory Research Report on Geological Disposal” consistent with the operator’s technical report sched-ule. This report would be intended to confirm the regulatory status of the program as well as strengthen-ing the competence of NISA as a regulatory body. JNES launched safety studies on geological disposal in its establishment year in 2003. JAEA and AIST joined as regulatory support research institutes in 2005. In October 2007, the three parties signed an agreement of cooperative study on geological disposal, which enhanced joint studies, as well as exchanges of staff, data, and results. One of the ongoing joint studies has been aimed at investigating regional-scale hydrogeological modeling using JAEA’s Horonobe Underground Research Center. The three parties have begun to discuss expanding the joint studies and the agreement areas in response to the new five-year re-search plan.

3) 40280 – Recent Developments and Trends on Requirements Management Systems

Satoru Suzuki, Hiroyoshi Ueda, Kiyoshi Fujisaki, Katsuhiko Ishiguro, Hiroyuki Tsuchi, NUMO (Japan); Stratis Vomvoris,Irina Gaus, Nagra (Switzerland)

Although the management of requirements for the development of geologic repository systems has been practiced in all radioactive waste disposal programs from the beginning, the systematic management through IT-based requirement management systems (RMS) has a younger history. Traditionally requirements have been gathered in various documents and quite often the rationale for design choices is difficult to trace. Thanks to the development of RMS in other industries (software development, aerospace), many radioactive waste disposal programs have recently been able to set-up and tailor RMS to their needs quickly. In a recent international meeting, five radioactive waste disposal organizations (NUMO/Japan; NAGRA/Switzerland; ONDRAF/NIRAS/Belgium; POSIVA/Finland and SKB/Sweden) have discussed the status and developments of RMS in their respective programs. The majority have already implemented an IT-based system, or, are testing and developing such systems. The level of detail of requirements depends on the stage of the program. Those approaching the license application have integrated all components of the repository concept, including the processes for the operational phase. It was recognized that the earlier you implement a systematic tracking of requirements and the decisions taken to satisfy these requirements, the easier it would be to implement an RMS over the whole duration of the geologic disposal program – in the order of 100 years. The documented transition from the general requirements (highest level) to the lower level ones – especially those related to the repository system concept or its components – can be then used to enable the assessment of potential changes at future stages of the system. Even if the host rock or, the repository concept has not been selected, a ‘dry’ run with assumed conditions can be very elucidating of the most useful set-up of the RMS. Requirements management is closely associated with the quality management system. Combining requirement and decision-tracking has been expressed as an explicitly goal for some programs. Caution was expressed regarding the expectations for the RMS being developed. There is a risk that such systems are perceived as expert systems that can derive decisions, which then will be unquestionably accepted. It is nevertheless recognized that they can be of great help in communicating with the various stakeholders and with relative ease demonstrating how their requirements have been considered and satisfied with the proposed repository systems. Further efforts need to be undertaken to integrate the requirement management systems, and the processes that they represent, in the day-to-day operations of the organizations. Different organizational schemes are being considered, for example, the composition of the team that defines the requirements at the various levels and the function of its members within the technical and scientific program of the waste disposal organizations. First positive experiences of the latter were reported.

4) 40228 – Development of Requirements Management System of NUMO and practical experience with development of the database contents

Satoru Suzuki, Hiroyoshi Ueda, Katsuhiko Ishiguro, Hiroyuki Tsuchi, Kiyoshi Fujisaki, NUMO (Japan); Kiyoshi Oyamada, JGC Corporation (Japan); Shoko Yashio, Obayashi Corporation (Japan)

-55-

Abstracts

Decision-making and work activities in the geological disposal program need to be implemented in such a way as to fulfill various requirements such as safety, practicality, quality and socio-economic aspects. Since a stepwise approach is applied for implementing the program, the number, weighting and specific nature of the requirements will change depending on the premises and constraints in each stage of implementation. Requirements management with a long-term perspective is therefore required for consistent implementation of the program. NUMO has developed a requirements management (RM) methodology that is suitable for the long-term, stepwise disposal program in Japan, as well as a supporting requirements management system (RMS) tool. The basic concept of the RMS was already presented at the last ICEM 2009. In this presentation, we will focus on practical experience with development of the database content for the RMS. The RM methodology was first applied in the HLW repository design work. Requirements for repository design were considered primarily from the viewpoint of post-closure safety and engineering feasibility. The repository design requirements are structured hierarchically to those assigned to the ‘post-closure safety concept’, the ‘required system functions’ and the ‘design requirements’. The post-closure safety concept, which is based on isolation by a multibarrier system, is further subdivided into fundamental safety functions and operational functions under the heading ‘required system functions’. The fundamental safety functions are ‘post-closure containment’, ‘retardation/reduction of radionuclide migration’ and ‘length of migration pathway’ and they need to be satisfied in order to fulfill the safety concept. The aim of the operational functions is to realize these safety functions by providing a feasible and reliable operating system (e.g. welding facility, transport and emplacement). Finally, the lowest level – the design requirements - are defined for each system component, so that each component can fulfill the upper ranking ‘required system functions’. The system components are the engineered barrier system (metal overpack and the bentonite buffer) and the natural barrier (the host formation and its surroundings). In the NUMO RMS tool, four elements -‘Decision/Work (D/W)’, ‘Requirements (R)’, ‘Conditions (C)’ and ‘Arguments (A)’ are used for describing the RM information. The design requirements discussed above are recorded in the ‘R’ element and attributed to the design work of each component of the ‘D/W’ elements. The ‘C’ elements record information such as the site environment, which in turn affects the D/W and R elements. ‘A’ elements record the synthesis of evidence that fulfills the requirements. The database content for these four elements will be defined in advance before initiating the design work in each stage of the disposal program.

5) 40231 – Application of Lifecycle Management to Design of the UK Geological Disposal Facility

Henry O'Grady, Malcolm Currie, Parsons Brinckerhoff (UK); Philip Rendell, Radioactive Waste Management Directorate (UK)

The Radioactive Waste Management Directorate (RWMD) of the UK’s (UK) Nuclear Decommissioning Authority (NDA) has been charged with the responsibility for design and delivery of a Geological Disposal Facility (GDF) for High Level Waste / Intermediate Level Waste in accordance with government policy. Over the next few years, the design of the GDF will be developed and the organisation built up to provide the necessary design and delivery capability. In the short term, this capability must also be demonstrated to UK government regulators in order to gain approval for nuclear operations. As part of this process, RWMD have developed a staged approach to engineering design, which addresses the overall lifecycle of the GDF in terms of seven phases, from initial concept development through to operation and, finally, closure. Each phase finishes with a formally defined milestone (a “gate”) comprising a technical review and a specific set of engineering deliverables. The phases and milestones have been built up from a number standard approaches described in open literature, which have then been crafted to address the specific needs of the GDF. Roles and responsibilities are also considered, along with the interface issues between various functional groups both within and outside of the engineering sphere. The process also incorporates a number of good practices based on the authors’ experiences, such as requirements management, progressive assurance and the use of architectural frameworks. As the lifecycle of the GDF will extend over decades, the process uses a modular approach which will permit it to evolve to meet the changing needs of the project, the organisation and the regulatory process. The process is intended to help the engineering effort provide timely support to, the higher-level needs of the overall project and the community engagement programmes, as well as with the activities of non-engineering functions. The process also provides for the management of risk by ensuring that the requirements, designs, risk registers and detailed procedures are accepted before further funding is released. This paper describes the background to the UK GDF programme, the organisational issues associated with the RWMD’s evolving role, the relationship between the top-level UK Government Managing Radioactive Waste Safely programme and the RWMD engineering lifecycle, the formal reviews, the milestones and the overall contribution this makes to RWMD organisational development and UK regulatory approval.

SESSION D2: Dismantling and Decontamination

1) 40036 – AREVA NP: Experience in Dismantling and Packaging of Pressure Vessel and Core Internals

Peter Pillokat, Jan Hendrik Bruhn, AREVA NP GmbH (Germany)

AREVA NP GmbH, German Regional Sector of French Nuclear Company AREVA is proud to look back on versatile experience in successfully dismantling nuclear components. After performing several minor dismantling projects and studies for nuclear power plants, AREVA NP completed the order to dismantle all remaining reactor pressure vessel internals at German boiling water reactor Wuergassen NPP in October ´08. During the onsite activities

-56-

Abstracts

about 121 tons of steel where successfully cut and packed under water into 200l- drums, as the dismantling was performed partly in situ and partly in an underwater working tank. AREVA NP deployed a variety of different cutting techniques such as band sawing, milling, nibbling, compass sawing and water jet cutting throughout this project. After successfully finishing this task, AREVA NP is currently dismantling the cylindrical part of the Wuergassen Pressure Vessel. During this project approxi-mately 320 tons of steel are cut and packaged for final disposal, as dismantling is mainly performed by on air use of water jet cutting with vacuum suction of abrasive and kerfs material. The main clue during this assignment is the logistic challenge to handle and convey cut pieces from the pressure vessel to the packing area. For this an elevator is installed to transport cut segments into the turbine hall, where a special housing is built for final storage conditioning. At the beginning of ‘07 another complex dismantling project of great importance was acquired by AREVA NP. The contract included dismantling and conditioning for final storage of all the RPV Internals at German pressurized water reactor Stade NPP. Very similar cutting techniques turned out to be the proper policy to cope this task. Onsite activities took place in up to 5 separate working areas including areas for post segmentation and packaging to perform optimized parallel activities. All together about 85 tons of core internals where successfully dismantled at Stade NPP until September ´09. To accomplish the best possible on-site performance and to achieve a minimization of the applied collective dose rate, each onsite activity was previously planned in detail and personnel exercised each task at original size mockups under most realistic onsite conditions. Planning was especially focused on an optimized size minimization and packaging concept to reduce the number of filled waste packages. The segmentation of components strictly followed a sophisticated cutting and packaging concept developed under consideration of possible cutting techniques, the resulting geometry and logistical conditions. Therefore segments were post processed by hydraulical press and band saw in order to minimize their volume, were applicable.

2) 40102 – Study on evaluation models of management data for decommissioning of Fugen

Yuji Shibahara, Masanori Izumi, Takashi Nanko, Mitsuo Tachibana, Tsutomu Ishigami, JAEA (Japan)

In the Fugen nuclear power plant (FUGEN), the dismantling of equipments in the turbine building has started in 2008, and the dismantling of equipments around the reactor is scheduled around in 2015. To evaluate the management data on this dismantling of equipments around reactor appropriately, it is very important to study whether the conventional evaluation models have the applicability for FUGEN or not. Thus, the management data on the dismantling of equipments in 3rd/4th feedwater heater room conducted in 2008 was calculated with the conventional evaluation models. The conventional evaluation models were made by data obtained from Japan Power Demonstration Reactor (JPDR) decommissioning program. It was found that there were large differences between the calculated values and the actual data. For finding the cause in the difference between them, the dismantling of equipments in 3rd/4th feedwater heater room was divided into three processes: i) the preparation process, ii) the dismantling process, and iii) the clean-up process. In the both process of preparation and clean-up, the calculated values were smaller than the actual data. These were mostly caused by the plant scale difference between JPDR and FUGEN, because the conventional evaluation models were built by analyzing the actual data on the decommissioning of JPDR which is smaller than FUGEN. The primary expression depending on the area of working space was built as new evaluation models for the preparation and clean-up processes. In the dismantling process, on the other hand, it was found that there were characteristic differences in the dismantling of feedwater heater as follows: 1) the calculated values were significantly larger than the actual data, 2) the actual data for the dismantling of 3rd feedwater heater was larger than that of 4th one, though these equipments were almost same weight. It was found that these were brought by the difference in the descriptions of dismantling of feedwater heaters, and the new evaluation models reflecting the descriptions of dismantling were built for the appropriate evaluation of the management data. The calculated values with the new evaluation models for each process showed the good agreement with the actual data. In this report, study on evaluation models of management data for dismantling of equipments in the feedwater heater room will be described.

3) 40083 – CORD Decontamination Technologies for Decommissioning - A Comprehensive Approach Based on Over 30 Years Experience

Christoph Stiepani, AREVA NP GmbH (Germany)

Decontamination prior to Decommissioning and Dismantlement is imperative. Not only does it provide for minimization of personnel dose exposure but also maximization of the material volume available for free release. Since easier dismantling techniques in lower dose areas can be applied, the licensing process is facilitated and the scheduling and budgeting effort is more reliable. The most internationally accepted approach for Decontamination prior to Decommissioning projects is the Full System Decontamination (FSD). FSD is defined as the chemical decontamination of the primary cooling circuit, in conjunction with the main auxiliary systems.

AREVA NP has long-term experience with Full System Decontamination for return to service of operating nuclear power plants as well as for decommissioning after shutdown. Since 1976, AREVA NP has performed over 500 decontamination applications and, from 1986, AREVA NP has performed Decontaminations prior to Decommissioning projects which comprise virtually all NPP designs and plant conditions:

- NPP designs: HPWR, PWR, and BWR by AREVA, Westinghouse, ABB and GE - Decontaminations performed shortly after final shutdown or several years later, and even after re-opening Safe

Enclosure

-57-

Abstracts

- Gamma / Alpha inventory - Main Coolant chemistry (e.g., with and without Zn injection during operation)

Fifteen Decontamination prior to Decommissioning Projects have been performed successfully to date. The sixteenth FSD for Decommissioning at the French NPP Chooz A is now in the detailed engineering phase with onsite application scheduled for Fall 2010.

This paper will describe the AREVA NP Decontamination Concept for Decommissioning (DCD) and present highlights of previous FSDs performed prior to decommissioning using the CORD / AMDA technology.

4) 40007 – Chemical Decontamination for Decommissioning (DFD) and DFDX

Ronald Morris, Westinghouse Electric Company (USA)

DFD is an acronym for the “Decontamination for Decommissioning” process developed in 1996 by the Electric Power Research Institute (EPRI). The process was designed to remove radioactivity from the surfaces of metallic components to allow these components to be recycled or free-released for disposal as non-radioactive. DFD is a cyclic process consisting of fluroboric acid, potassium permanganate and oxalic acid. The process continues to uniformly remove base metal once oxide dissolution is complete.

The DFD process has been applied on numerous components, sub-systems and systems including the reactor systems at Big Rock Point and Maine Yankee in the USA, and the Jose Cabrera (Zorita) NPP in Spain. The Big Rock Point site has been returned to Greenfield and at Maine Yankee the land under the license was reduced for an Independent Spent Fuel Storage Installation (ISFSI). In the upcoming months, the Zorita NPP in Spain will initiate dismantlement and decommissioning activities to return the site to a non-nuclear facility.

The development of the EPRI DFD process has been an on-going evolution and much has been learned from its use in the past. It is effective in attaining very high decontamination factors; however, DFD also produces secondary waste in the form of ion exchange resins. This secondary waste generation adds to the decommissioning quota but can be improved upon at a time when radioactive waste storage at nuclear facilities and waste disposal sites is limited.

To reduce the amount of secondary waste, EPRI has developed the DFDX process. This new process is an enhancement to the DFD process and produces a smaller amount of metallic waste rather than resin waste; this reduction in volume being a factor of ten or greater. Electrochemical ion exchange cells are the heart of the DFDX system and contain electrodes and cation ion exchange resin. It has been used very successfully in small system applications and the next evolution is to design, build and implement a system for the chemical decontamination for decommissioning of larger reactor systems and full system decontamination (FSD).

The purpose of this paper will be to provide a reference for those planning future chemical decontaminations for plant decommissioning. It will be based on actual experience from the work that has been performed to date and the planned development of the DFDX process.

5) 40127 – Methods for Calculation and Optimisation of Personnel Exposure during Planning of Decommissioning of Nuclear Installation

Marek Vasko, Ivan Rehak, DECOM, a.s. (Slovakia); Vladimir Daniska, DECONTA, a.s. (Slovakia); Vladimir Necas, Slovak University of Technology in Bratislava (Slovakia)

Assessment of personnel exposure is one of safety issues being evaluated within decommissioning plans. It is required to show an impact of planned decommissioning activities to personnel and demonstrate even at the stage of decommissioning planning that the personnel exposure will be minimized in line with ALARA principle. The paper presents a methodology for evaluation of personnel exposure developed within standardized decommissioning costing code OMEGA. It deals with a methodology for calculation of external and internalpersonnel exposure based on calculated individual manpower components for each profession within working group and relevant dose rates, the dose rate from equipment, average dose rate in the rooms and background dose rate, as well as air volume radioactivities of radionuclides as they are recorded in facility inventory database. It also deals with the methodology for evaluation and optimization of the individual effective dose for individual members of working group based on their profession. Developed methodology enables optimization of deployment for individuals assigned to given professions within deployed working group in order to keep the individual effective dose within stipulated limits.

SESSION D3: Planning

1) 40129 – Program Change Management During Nuclear Power Plant Decommissioning

Mike Naughton, Sean Bushart, Karen Kim, EPRI (USA)

Decommissioning a nuclear power plant is a complex project. The project involves the coordination of several different departments and the management of changing plant conditions, programs, and regulations. As certain project Milestones are met, the evolution of such plant programs and regulations can help optimize project execution and cost.

-58-

Abstracts

This paper will provide information about these Milestones and the plant departments and programs that change throughout a decommissioning project. The initial challenge in the decommissioning of a nuclear plant is the development of a definitive plan for such a complex project. EPRI has published several reports related to decommissioning planning. These earlier reports provided general guidance in formulating a Decommissioning Plan. This Change Management paper will draw from the experience gained in the last decade in decommissioning of nuclear plants. The paper discusses decommissioning in terms of a sequence of major Milestones. The plant programs, associated plans and actions, and staffing are discussed based upon experiences from the following power reactor facilities: Maine Yankee Atomic Power Plant, Yankee Nuclear Power Station, and the Haddam Neck Plant. Significant lessons learned from other sites are also discussed as appropriate. Planning is a crucial ingredient of successful decommissioning projects. The development of a definitive Decommissioning Plan can result in considerable project savings. The decommissioning plants in the U.S. have planned and executed their projects using different strategies based on their unique plant circumstances. However, experience has shown that similar project milestones and actions applied through all of these projects. This allows each plant to learn from the experiences of the preceding projects. As the plant transitions from an operating plant through decommissioning, the reduction and termination of defunct programs and regulations can help optimize all facets of decommissioning. This information, learned through trial in previous plants, can be incorporated into the decommissioning plan of future projects so that the benefits of optimization can be realized from the beginning of the projects. This process of the collection of information and lessons learned from plant experiences is an important function of the EPRI Decommissioning Program.

2) 40245 – Status of the Support Researches for the Regulation of Nuclear Facilities Decommissioning in Japan

Yusuke Masuda, Yukihiro Iguchi, Satoru Kawasaki, Masami Kato, JNES (Japan)

In Japan, 4 nuclear power stations are under decommissioning and some nuclear fuel cycle facilities are expected to be decommissioned in the future. On the other hand, the safety regulation of decommissioning of nuclear facilities was changed by amending law in 2005. An approval system after review process of decommissioning plan was adopted and applied to the power stations above. In this situation, based on the experiences of the new regulatory system, the system should be well established and moreover, it should be improved and enhanced in the future. Nuclear Industry and Safety Agency (NISA) is in charge of regulation of commercial nuclear facilities in Japan and decommissioning of them is included. Japan Nuclear Energy Safety Organization (JNES) is in charge of technical supports for NISA as a TSO (Technical Support Organization) also in this field. As for decommissioning, based on the needs in terms of regulation, JNES has been continuing research activities from October 2003, when JNES has been established. Considering the “Prioritized Nuclear Safety Research Plan (August 2009)” of the Nuclear Safety Commission of Japan and the situation of operators facilities, “Regulatory Support Research Plan between FY 2010-2014” was established in November 2009, which shows the present regulatory needs and a research program. This program consists of researches for 1. review process of decommissioning plan of power reactors, 2. review process of decommissioning plan of nuclear fuel cycle facilities, 3. termination of license at the end of decommissioning and 4. management of decommissioning waste. For the item 1, JNES studied safety assessment methods of dismantling, e.g. obtaining data and analysis of behavior of dust diffusion and risk assessment during decommissioning, which are useful findings for the review process. For the item 2, safety requirements for the decommissioning of nuclear fuel cycle facilities was compiled, which will be used in the future review. For the item 3, measuring method, release procedure and analysis code for the site release were studied for the establishment of the license termination process in the future. From FY 2010, based on the new plan, we have started the researches for the standardization of review process of decommissioning plan for power reactors and nuclear fuel cycle facilities, establishing the process and criteria of license termination and appropriate method of management of decommissioning waste based on the waste form confirmation process.

3) 40136 – Decommissioning Costing Approach Based on the Standardised List of Costing Items; Lessons Learnt

Vladimir Daniska, Frantisek Ondra, Peter Bezak, DECONTA, a.s. (Slovakia); Ivan Rehak, Marek Vasko, Jozef Pritrsky, DECOM a.s. (Slovakia) ;

Matej Zachar, Vladimir Necas, Slovak University of Technology in Bratislava (Slovakia)

The document “A Proposed Standardised List of Items for Costing Purposes” was issues in 1999 by OECD/NEA, IAEA and European Commission for promoting the harmonisation in decommissioning costing. It is a systematic list of typical decommissioning activities classified into chapters 01 to 11 with three hierarchical numbered levels. Cost groups are defined in chapter 12 for presenting cost for each activity. In this way the document is the standardised matrix of decommissioning activities and cost groups with unambiguous meaning of items. Knowing what is behind the items of the standardised cost structure makes the comparison of cost for various decommissioning projects transparent. There are two principal approaches for use of the standardised cost structure. First approach converts the cost items from specific cost structures into the standardised cost structure for the purpose of cost presentation. Second approach uses the extended standardised cost structure as the calculation structure; the cost data have the standardised format directly. Several advantages can be identified in this approach. The paper presents general aspects of the costing methodology based on the standardised cost structure and lessons learnt from last ten years of implementation of the standardised cost structure as the cost calculation structure in the computer code OMEGA. The paper presents following aspects:

• general principles of decommissioning costing in current costing methodologies based on information on costing

-59-

Abstracts

methodologies available, • standardised cost structure, methods of its use in costing, cost calculation structures based on the standardised

cost structure, • principles of cost calculation methodology based on the standardised cost structure, input data for the

methodology, cost calculation process, management of calculated data • interactions of main cost structures involved in standardised costing: standardised cost calculation structure,

work breakdown structure of a decommissioning project, relations to cost accounting structures • main features of the computer code OMEGA: implementation of the standardised cost structure as the cost

calculation structure, management of the material and radioactivity flow in the decommissioning process, generation of Gantt charts for managing of the decommissioning projects, data feedback from real processes, links to cost accounting systems

• international activities for promoting of the standardised cost structure and for its upgrading

4) 40290 – The Outline of Decommissioning Plan for Hamaoka Nuclear Power Station Unit-1 and Unit-2

Yoshifusa Fukuoka, Chubu Electric Power Co., Inc. (Japan)

Hamaoka Nuclear Power Station's Unit-1 and Unit-2 ended their operation on January 30, 2009. The Unit-1 and 2 will be dismantled and removed, and doing this requires establishing a nuclear reactor facility decommissioning plan, and getting the approval of the national government. Their decommissioning procedure commenced after the submission of the "Application for the approval of the decommissioning plan for Hamaoka Nuclear Power Station Unit-1 and 2" to the Minister of Economy, Trade and Industry on June 1, 2009 , and subsequent approval for the application on November 18, 2009. ?The application includes an overall plan for dismantling reactor facilities safely and surely, a description of tasks to be performed during the period Chubu Electric is preparing to dismantle the facilities in the coming years and safety assurance measures, among other information. ?According to the plan, the decommissioning of Unit-1 and Unit-2 is to be completed by the end of FY2036. The 28-year schedule is divided into four phases. ?Phase 1?Preparation Stage, FY2009 ? FY2014 Transporting and transferring nuclear fuel out of the plants, investigating the status of contamination, conducting system decontamination, and dismantling / removing facilities and equipment outside RCA ?Phase 2?Dismantling and Removal Stage for Reactor Zone Peripheral Facilities, FY2015 ? FY2022 Dismantling / removing reactor zone peripheral facilities, safely storing dismantlement debris, and installing facilities for processing dismantlement debris ? Phase 3?Dismantling and Removal Stage for Reactor Zones, FY2023 ? FY2029 Dismantling and removing the reactor zone (covering the reactor vessel, core support structures, and radiation shields surrounding the reactor vessel) ?Phase 4?Dismantling and Removal Stage for Building Structures, FY2030 ? FY2036 Removing radioactive materials inside plant buildings, and dismantling / removing the buildings.

Of waste to be generated in decommissioning work at Unit-1 and Unit-2, low-level radioactive waste accounts for approx. 17,000 tons . Low-level radioactive waste is sorted according to the types of radioactive substances contained or the level of radiation based on laws and regulations, and appropriately put to underground disposal based on the classifications. Specific methods for disposal, including the disposal site, will be decided before the start of dismantlement work on reactor zone's peripheral facilities, and reflected to the decommissioning plan for approval.

The decommissioning plan for Hamaoka Nuclear Power Station Unit-1 and Unit-2 represents Japan's first decommissioning of a commercial light-water nuclear power plant. The priority is given to safety to steadily implement the decommissioning plan with transparency on Hamaoka NPS, and to acquisition of the trust from everyone concerned.

5) 40015 – Study on Influence of Nuclear Fuel Material Management and Transfer Scenarios on Decommissioning

Kazuma Mizukoshi, Nuclear Engineering, Ltd. (Japan)

1. Summary The Japanese electric utilities are required to prepare plans to transfer its nuclear fuel material by the end of the decommissioning period. There can be several scenarios regarding the management and transfer of nuclear fuel material. It is necessary to fully understand the characteristics of individual scenarios so that the most suitable method can be selected according to the conditions specific to each plant. We have examined how the nuclear fuel material management and transfer scenarios cause influences on the decommissioning and evaluated the characteristics (advantages and disadvantages) of each scenario. We expect that the result of this study will be useful for a nuclear power station which plans decommissioning to choose a suitable and effective scenario for the nuclear fuel material management and transfer.

2. Method We collected information about methods of nuclear fuel material management and transfer and extracted several scenarios as shown below: Scenario A? Spent fuel (SF) is transported to a reprocessing plant after storing it in the same SFP as it was stored during the in-service period. Scenario B? SF is transported to an another unit of the same power station to transfer the management of the SF to the unit. ?The transportation between units.? Scenario C? SF is transported to an interim storage facility to transfer the management of SF to the interim storage facility. Scenario D: SF is transported to a reprocessing plant after storing it in a SFP which is isolated by replacing existing system components with temporarily installed components. Scenario E?SF is transported to an interim storage facility after storing it in a SFP which is isolated by replacing existing system components with temporarily installed components, and its management is transferred to the interim storage facility.

-60-

Abstracts

Regarding the scenarios mentioned above, we examined the influence on decommissioning conditions (dismantling / storage areas, and so on), decommissioning process, and costs.

3. Result A study on the influence of the scenarios of nuclear fuel material management and transfer on decommissioning has clarified the characteristics (advantages and disadvantages) of individual scenarios and which scenario is suitable to specific conditions.

4. Conclusion We could fully understand the characteristics of the scenarios of nuclear fuel material management and transfer. We expect the result obtained from this study will be useful for a nuclear power plant to choose a suitable scenario of nuclear fuel material management and transfer.

6) 40100 – Dose Assessment for setting of EPZ in Emergency Plan for Decommissioning of Nuclear Power Plant

Hirokazu Minato, Hitachi-GE Nuclear Energy (Japan); Takatoshi Hattori, CRIEPI (Japan); Toshihiko Higashi,The Kansai Electric Power Co., Inc. (Japan); Takehiro Iwata, JAPC (Japan)

In emergency plan for a nuclear power plant, taking enough measures in EPZ boundary (Emergency Planning Zone) is one of priority matters, to have protection against the release of radioactive materials in accident efficiently and quickly to minimize environmental impacts. EPZ is set as the zone which emergency plans should be mainly prepared on emergency conditions. The criteria of EPZ is that dose value in the area between plant and EPZ boundary have to be less than 10mSv, even if very conservative release mechanisms and path are supposed. The released amount of radioactive material from a nuclear power plant is calculated to the accident scenario were supposed with the each phase of decommissioning. Moreover, the dose value is calculated as the evaluation of environmental impacts, using atmospheric diffusion parameters are determined by the plume concentration gaussian type distribution model at steady state, and annual meteorological data of the reference plants. Both of 'the spent fuel storage phase' and 'the safe maintenance and dismantling phase' on each of the expected accident scenario, the dose value in EPZ boundary is much less than safety criteria (10mSv), and there is no need to plan offsite emergency plan, such as the Sheltering and Escape for a reference plant. This result is agreeing with the opinion of Waste and decommission working group 2006 of Western European Nuclear Regulator's Association (WENRA).

SESSION G1 : International Collaboration

1) 40118 – Advancing the Use of IAEA Networks in Radioactive Waste Management: Past Successes, Present Challenges and Future Opportunities

Paul Degnan, John Kinker, Irena Mele, Paul J. Dinner, Horst Monken Fernandes, Antonio Morales, Lumir Nachmilner, Shaheed Hossain, IAEA

Since 2001 the International Atomic Energy Agency has championed the concept and use of Networks to advance radioactive waste management across the globe. At the present time there are four Networks managed on behalf of Member States by the IAEA and a fifth one is currently being implemented. The scopes of interest covered by the Networks encompass near-surface and deep geological disposal, the decommissioning of nuclear facilities, the environmental remediation of sites contaminated with radioactive materials and the characterisation of low- and intermediate-level radioactive wastes. To date over 100 organisations from more than 40 Member States are involved in the Networks. Many of these Network participants generously donate resources, time and effort to support Network activities, while others with nascent or otherwise less well developed programmes are still in the process of acquiring experience, capabilities and know-how. Regardless of the stage of development, all Network participants share in the mutual benefits that arise from improved communications with sister organisations and the sharing of experience and knowledge. The universal Goal of the Networks is the promotion of methods and technologies that will enhance the safety and sustainability of radioactive waste management practices and facilities. This Goal is being achieved through continuous improvements in communication and knowledge sharing between Network participants and the provision of enhanced opportunities for training, involvement in demonstration projects and the development of novel technologies and methodologies. We recognise that interdisciplinary understanding and the coordination of efforts at key interfaces at the back-end of the fuel cycle are critical aspects for achieving the Network Goal efficiently and effectively. Consequently, the IAEA Networks that will be operational by the end of 2010 are themselves are being molded into an organic “Network of Networks” where the use of new electronic media and the possibilities presented by enhanced communication channels will be exploited. Here we provide an overview of the IAEA Networks in radioactive waste management and present a new tool that is under development, an internet-based portal for enhanced communications and the provision of improved training opportunities.

2) 40287 – The activities of the OECD/NEA RWMC in the Field of HLW and SF disposal

Claudio Pescatore, OECD/NEA

The OEDCD/NEA seeks to assist its member countries in developing strategies for the management of all types of radioactive material, including waste, that are safe and sustainable and that meet the broad needs of society - with

-61-

Abstracts

particular emphasis on the management of long-lived waste and spent fuel and on decommissioning of disused nuclear facilities. The programme of work in the area of radioactive waste management is supervised by the Radioactive Waste Management Committee (RWMC) made of senior representatives from regulatory authorities, radioactive waste management and decommissioning organisations, policy making bodies, and research-and-development institutions from the NEA countries. The IAEA participates in the work of the committee, and the European Commission (EC) is a full member. Strong ties are maintained with national high-level advisory bodies to governments and with international bodies such as the International Committee on Radiation Protection. RWMC helps the national programmes through a broad programme of work that: fosters a shared and broad-based understanding of the state of the art and emerging issues; facilitates the elaboration of waste management strategies that respect societal requirements; helps to provide common bases to the national regulatory frameworks; enables the management of radioactive waste and materials to benefit from progress of scientific and technical knowledge; contributes to knowledge consolidation and transfer, e.g., through the organization of specialist meetings and publication of technical reports, consensus statements and short flyers; and helps advance the state of the art, e.g., by providing a framework for the conduct of international peer reviews. The latest collective statement of the RWMC dates to 2008 on “moving forward with geological disposal of radioactive waste”; the latest peer review was of the French Dossier 2005. A new peer review of the safety report for a spent fuel repository in Sweden is being organised. The RWMC holds multi-stakeholder, national workshops. The latest (2009) was in France, in the siting region for the national HLW repository. The RWMC manages a database of country information, in the form of a country reports and country profiles updated yearly; a summary of the regulatory infrastructure in NEA countries is also maintained. Current work areas include: promoting greater understanding of radioactive waste management and decommissioning disciplines; an international project on the topic of “retrievability and reversibility”; assisting the organisation of an “International Conference on Geological Repositories” in Japan in 2011; initiating dialogue with ICRP with a view to updating the ICRP guidance in the field of geological disposal; starting a project in the field of information and memory keeping. Specific technical areas also include optimization, treatment of the very long time scales, assessing the state of the art in safety assessment methods, and the operation phase of repositories. An overview is provided of these activities and the relevant issues.

3) 40147 – Grimsel Test Site - Phase VI Status and Outlook

Ingo Blechschmidt, Sven Peter Teodori, Stratis Vomvoris, Nagra (Switzerland)

The Grimsel Test Site (GTS) (www.grimsel.com) is a generic underground research laboratory located in the crystalline rocks of the Aare Massif of the Swiss Alps and is owned and operated by the National Cooperative for the Disposal of Radioactive Waste (Nagra). The GTS is unique in that it includes a class B/C control zone in which repository relevant radionuclides can be used as tracers of rock processes. 2009 marked the 25th year of GTS’s operation and the current running Phase VI is planned until 2013. Experiments have evolved from those focused on characterising the structural geology, groundwater geochemistry and hydrogeology of the test site during the 80s; to radionuclide migration experiments in the 90s; and then more recently to assessing perturbation effects of repository implementation and demonstrating engineering and operational aspects of the repository system for the last 10 years.

Currently over 25 international organisations participate in various projects at the GTS. On-going international partner projects are as follows: evaluation of full-scale engineered systems under simulated heat production and long-term natural saturation (FEBEXe); emplacement of a shotcrete low-pH plug (EC Project ESDRED); testing and evaluation of standard monitoring techniques (TEM/MoDeRn) including both wired and non-wired techniques; long-term cement studies (LCS) which aims at improving the understanding of low-pH cement interaction effects in water conducting features; the Colloid Formation and Migration Project (CFM) which focuses on colloid generation and migration from a bentonite source doped with radionuclides; and the Long-Term Diffusion (LTD) project which aims at in-situ verification of long term diffusion concepts for radionuclides.

Additional experiments include techniques for determination of fracture network hydraulic and migration parameters and behaviour of grouting materials. New large-scale experiments to test the emplacement techniques and behaviour seals/plugs at 1:1 scale, under hydraulic pressure differential of up to 50 bars with parallel migration of repository-generated gas are in the planning stage.

The status of the on-going experiments as well as the future plans and possibilities for new partners are summarised in this poster presentation.

SESSION L4: Solidification and Package (2)

1) 40299 – Treatment of low level radioactive waste by plasma: a proven technology?

Jan Deckers, Belgoprocess, NV (Belgium)

Introduction Large amounts of actual and historical low level radioactive waste, with varying characteristics, are stored and generated from the operation and maintenance of nuclear power plants, the nuclear fuel cycle, research laboratories, pharmaceutical and medical facilities. Virtual all of these waste streams can be treated by the plasma technology resulting in a final product free from organics, liquids and moisture, and meets without a doubt the acceptance criteria for safe storage and disposal.

-62-

Abstracts

Review The plasma is a highly desirable heat source. Its high temperature of up to 10,000 °C can treat the radioactive waste as is. The inorganic materials are melted into a glassy slag, containing most of the radioactive isotopes while the organic material are vaporised into a syngas and afterwards oxidised in an afterburner. This technology is very suitable for historical waste containing mixtures of inorganic, organic, liquids, sludge, etc, with almost no preparation of the waste and minimum risks for radioactive contamination and exposure.

Discussion In the interest of the waste producers and future generations, a high volume reduction factor (VRF) of the waste is desired in order to minimise the volume and overall costs of storage and waste disposal. Not only the VRF is of importance but also the growing requirements for improved quality of the final waste form. Therefore the plasma technology can also be used to recondition previous conditioned waste packages that no longer meet the current acceptance criteria for final disposal.

Conclusion This paper describes in detail the principles of plasma, the different waste feed systems, off gas treatment, operational experience and future plasma plants.

2) 40128 – The Zwilag Plasma Facility - Five Years of Successful Operation

Walter Heep, Zwilag Interim Storage Facility (Switzerland)

This paper is about a treatment facility of low level radioactive wastes that operates with plasma technology.The first processing of low level radioctive wastes from Swiss nuclear power plants marked the successful completion of the commissioning of this facility in March 2004. The commercial operation of the plasma plant owned by Zwilag Zwischenlager Würenlingen AG (Zwilag) has also enabled this technology to be used successfully for the first time in the nuclear field in a way that addresses the issue of radiation protection. The plasma facility has been operating now for five years and was granted an unrestricted operating license in September 2009.

3) 40293 – Safety Assessment of Disposal Container for Higher Activity Low Level Waste

Motonori Nakagami, Seiji Komatsuki, Chubu Electric Power Co., Inc. (Japan); Kyosuke Fujisawa, Takashi Nishio, KOBE STEEL, LTD. (Japan);

Thomas Quercetti, André Musolff, Karsten Müller, Federal Institute for Materials Research and Testing (Germany)

As one of the studies on "yoyushindo disposal" whose concept is similar to an intermediate disposal, the development of a disposal container has been conducted by the Federation of Electric Power Companies of Japan. To assess a drop event of a waste package in which stored the radioactive wastes from nuclear power plants, the toughness of the disposal container was evaluated by drop tests using three specimens which have actual dimensions, drop analysis, fracture mechanics assessment and macroscopic tests. The three specimens for drop tests were manufactured in consideration of the design specifications and the manufacture operations in nuclear power plants. The lid plates of the specimens were welded to the body plates without pre- and post-weld heat treatment by using a remote automated welding machine. The drop tests showed that no penetration cracks or splash of its content occurred in the disposal container under conservative conditions such as the maximum weight and height in the handling. Drop analysis and the fracture mechanics assessment indicate that the strain induced by the drop impact did not exceed the fracture strain and an unstable fracture did not occur. And macroscopic tests showed that penetration cracks did not occur at 8m drop events. These tests and evaluations confirmed that the disposal container had sufficient toughness.

SESSION L5: Recycling and Clearance

1) 40223 – NPP Bulk Equipment Dismantling Problems and Experience

Alexander B.Gelbutovsky, Eugeny V. Balushkin, Jury A. Epikhin, Alexander V. Troshev, Peter I. Cheremisin, ECOMET-S JSC (Russia)

NPP bulk equipment dismantling problems and experience are summarized. “ECOMET-S” JSC is shown as one of the companies which are able to make NPPs industrial sites free from stored bulk equipment with its further utilization. “ECOMET-S” JSC is the Russian Federation sole specialized metallic LLW (MLLW) treatment and utilization facility. Company/s main objectives are waste predisposal volume reduction and treatment for the unrestricted release as a scrap. Leningrad NPP decommissioned main pumps and moisture separators / steam super heaters dismantling results are presented. Prospective fragmentation technologies (diamond and electro-erosive cutting) testing results are described. The electro-erosive cutting machine designed by “ECOMET-S” JSC is presented. The fragmentation technologies implementation plans for nuclear industry are presented too. Key words: cutting, fragmentation, NPP bulk equipment, recycling, utilization, main pump casing, gas cutting, melting, metal ingots, unrestricted use, separator, steam super heater, air plasma cutting machine, diamond cutting, electro erosive cutting, electro-erosive cutting machine.

-63-

Abstracts

2) 40073 – Reuse Of Conditionally Released Radioactive Materials From NPP Decommissioning Applied In Motorway Bridges Construction

Michal Panik, Tomas Hrncir, Vladimir Necas, Slovak University of Technology in Bratislava (Slovakia)

During the operation and especially during decommissioning of nuclear installation is produced considerable amount of solid materials (metals, non-metals, building structures) that can fix radioactivity in forms of contamination or activation. The materials present radioactive waste, part of radioactive waste may just slightly exceed limits for unconditional release of materials into the environment. On the other side, there is possible, after proving of defined safety limits fulfillment, to conditionally release radioactive waste for special purpose. In opposite case it would be inevitable to dispose radioactive waste in radioactive waste repository. Approaches of different countries to release of materials vary and the extent of this issue processing is related to each approach. Requirements set down in Slovak Republic legislation are given in the paper. Before the conditional release of materials there must be done consistent analysis of the materials impact on the inhabitants and the environment in short and long time period. The analysis comprises the evaluation of considered scenarios of specific utilization of conditionally released materials. This analysis necessarily precedes the realization of utilization. Scenarios describing utilization of radioactive waste carbon steel in the motorway bridge building process is stated in the paper. Radioactive steel can be utilized in many parts of the bridge. In the paper it is described its use as reinforcement of piles. Short time period external irradiation of workers and inhabitants is taken into account. Critical group (i.e. the group that gets the highest accumulated dose) of workers or inhabitants is chosen. Specific mass activity of released radioactive waste carbon steel is related to individual effective dose taken by critical group. Following legislation rules, annual effective dose taken by critical group must not overstep the limit of 10 µSv/year. The determination of value of the specific mass activity is the target of scenarios evaluating. Evaluation of model scenarios can be realized with the appropriate calculation tool. In the paper VISIPLAN 3D ALARA planning tool was chosen.

3) 40071 – Modelling of Motorway Tunnels Scenario for Utilization of Conditionally Released Radioactive Materials

Tomas Hrncir, Michal Panik, Vladimir Necas, Slovak University of Technology in Bratislava (Slovakia)

Considerable amount of solid radioactive waste is generated during the decommissioning of nuclear installations. Some of the materials can be released into the environment either in a direct way, if they meet the releasing limits, or after the application of some procedures leading to meeting the limits. If materials exceed these limits, they are treated, conditioned and disposed of at appropriate repository of radioactive waste. Another possible releasing way is the conditional release of materials, which is discussed in this paper. The basic principles of conditional release as well as possibilities of reusing of the conditionally released materials are described. One of these possibilities of the reusing was chosen and application proposal of conditional release of metal waste - steel reinforcement in the concrete, which could be used for construction of motorway tunnels, was created. The computer code Visiplan 4.0 3D ALARA planning tool software was used for the calculation of effective individual dose for personnel constructing the tunnel and for critical group related to scenario. Particular models for individual scenarios of conditional release have been developed within the scope of this software code. The aim of the paper is to determine a level of the radioactivity of conditional released materials to avoid over exceeding the value of annual individual effective dose 10microSv/year established by international recommendations.

4) 40117 – Estimate of Clearance Levels for Metal Materials Contaminated with Uranium

Seiji Takeda, Hideo Kimura, JAEA (Japan)

The Nuclear Safety Commission of Japan (NSC) published the draft report on the derivation of clearance levels for the solid materials contaminated with uranium (uranium-bearing wastes) in October 2009. The authors provide NSC with the estimated results of the clearance levels of major radionuclides, U-234, U-235 and U-238, for metal reuse scenario. The metal reuse scenario is categorized as two sub-scenarios on a series of recycling process of scrap metals and on reuse of items made from recycled metal and slag. By applying an effective dose criterion of 10?Sv/y, the activity concentrations are estimated from the deterministic dose calculation. The activity concentrations for U-234, U-235 and U-238 are calculated to be 1.5Bq/g, 1.4Bq/g and 1.8Bq/g respectively. In order to confirm the validity of the calculated concentrations, we estimate the uncertainties on scenario description after metal recycling and on parameter values used in the deterministic calculation. It is difficult to rule out the possibility that a small amount of residue of slag generated from recycling process is disposed of as an industrial waste. Accordingly, the authors estimate the dose for a worker involved with landfill disposal of residue of slag and moreover the total uranium concentration derived from the slag in the disposal site. The calculated dose of the worker for U-234, U-235 and U238 is about 0.05 times as low as that for the metal reuse scenario respectively. This result, therefore, leads to the predominance of estimating the dose on the basis of the metal reuse scenario. The total uranium concentration derived from the slag in the disposal site is estimated to be lower than the mean value of measured uranium concentration data in natural environment of Japan. This result indicates that the landfill disposal of residue of slag hardly brings to increase uranium concentration of natural origin. Monte Carlo-based uncertainty analysis was carried out in order to estimate the influence of parameter uncertainties to the result of deterministic calculation. The results of the uncertainty analysis, which are corresponding to the 97.5th percentile of

Session L5 - H5 - H6

-64-

Abstracts

minimum activity concentration estimated by applying an effective dose criterion of 100?Sv/y, are higher than those of deterministic calculation. The calculated activity concentrations by deterministic calculation for U-234, U235 and U-238 were confirmed from the results of both the analysis for an additional scenario on the landfill disposal of residue of slag and the Monte Carlo-based uncertainty analysis.

SESSION H5: Panel "Advances in Knowledge Management for Radioactive Waste Disposal"


SESSION H6: Coupled Process Modeling and Natural Analogues

1) 40306 Keynote – A Discussion of Key Issues in Coupled THM Processes in Clays, Rock Salt and Crystalline Rock with Bentonite Buffer

Chin-Fu Tsang, LBNL (USA)

Abstract Not Available

2) 40159 – Environmental Remediation of High-Level Nuclear Waste in Geological Repository: modified Computer Code Creates Ultimate Benchmark in Natural systems

Geoffrey Peter, Oregon Institute of Technology Portland Center (USA)

Isolation of high-level nuclear waste in permanent geological repositories has been a major concern for over 30 years due to the migration of dissolved radio nuclides reaching the water table (10,000-year compliance period) as water moves through the repository and the surrounding area. Repositories based on mathematical models allow for long-term geological phenomena and involve many approximations; however, experimental verification of long-term processes is impossible. Countries must determine if geological disposal is adequate for permanent storage. Many countries have extensively studied different aspects of safely confining the highly radioactive waste in an underground repository based on the unique geological composition at their selected repository location. This paper discusses two computer codes developed by various countries to study the coupled thermal, mechanical, and chemical process in these environments, and the migration of radionuclides. Further, this paper presents the results of a case study of the Magma-hydrothermal (MH) computer code, modified by the author, applied to nuclear waste repository analysis. The MH code verified by simulating natural systems thus, creating the ultimate benchmark. This approach based on processes similar to those expected near waste repositories currently occurring in natural systems. Keywords: High-Level Nuclear Waste, Repository, Environmental Remediation, Computer Codes, Compliance Period, Benchmark.

3) 40196 – Effect of the Residual Heat Release of the Nuclear Waste Stored in an Unsaturated Zone on Radionuclide Release

Lubna K. Hamdan, John C. Walton, Arturo Woocay, University of Texas at El paso (USA)

Over time, nuclear waste packages disposed in an unsaturated zone geological repository, such as the proposed repository at Yucca Mountain, are expected to fail gradually due to localized and general corrosion. As a result, water will have access to the nuclear waste and radionuclides will be transported to the accessible environment by ground water. In this paper we consider a serious failure case in which penetrations at the top and bottom of the waste package will allow water to flow through it (flow-through model). We introduce a new conceptual model that examines the effect of the residual heat release of the nuclear waste stored in an unsaturated environment on radionuclide release. This model predicts that the evaporation of water at the hotter sheltered areas (from condensate and seepage) inside the failed waste package will create a capillary pressure gradient that drives water to wick with its dissolved and suspended contents toward these relict areas, effectively preventing radionuclides release. We drive a dimensionless group to estimate the minimum length of the sheltered areas required to sequester radionuclides and prevent their release. The implications of this model on the performance of the proposed repository at Yucca Mountain or unsaturated zone geological repositories in general are explored.

4) 40072 – Evaluation of behavior of rare earth elements based on determination of chemical state in groundwater in granite

Yuhei Yamamoto, Daisuke Aosai, Takashi Mizuno, JAEA (Japan)

Rare earth elements (REEs) in natural environment have been studied as a useful geochemical tracer and analogues for trivalent

-65-

Abstracts

actinides such as Am(III) which is contained in High-Level radioactive Waste (HLW). Since Am does not occur in natural environment, behavior of REEs in deep groundwater gives us valuable information for research and development relating to the safety assessment of geological disposal of HLW. In groundwater, REEs often form complexes with aqueous and colloidal ligands according to their large ionic potential (ionic charge/ionic radius). Therefore, behavior of REEs strongly depends on physical and chemical properties of these complexes. However, it is difficult to speciate complexes of REEs in groundwater mainly due to difficulties in direct measurement of REEs complexes with their low concentrations and alteration of groundwater condition during collection. The aim of this study is determination of chemical states of REEs in groundwater by combination of ultrafiltration techniques maintaining in-situ pressure and anaerobic conditions, speciation calculation taking care about contribution of natural organic matters, and finger-printing method using REE pattern of stability constants for probable complexes of REEs in groundwater. Groundwater samples were collected from a borehole located in the 200 m substage of the Mizunami Underground Research Laboratory (MIU), Gifu, Japan. Our results suggest that colloidal ligands play an important roll on the behavior of REEs in groundwater.

5) 40022 – Natural analogue studies of bentonite reaction under hyperalkaline conditions: overview of ongoing work at the Zambales Ophiolite, Philippines

Naoki Fujii, M. Yamakawa, RWMC (Japan); K. Namiki, Obayashi Corporation (Japan); T. Sato, Hokkaido University (Japan); N. Shikazono, Keio University (Japan);

C. A. Arcilla, C. Pascua, University of the Philippines (Philippines); W Russell Alexander, Bedrock Geosciences (Switzerland)

Bentonite is one of the most safety-critical components of the engineered barrier system for the disposal concepts developed for many types of radioactive waste. However, bentonite – especially the swelling clay component that contributes to its essential barrier functions – is unstable at high pH. To date, results from laboratory tests on bentonite degradation have been ambiguous as the reaction rates are so slow as to be difficult to observe in the laboratory. As such, a key goal in this project is to examine the reaction of natural bentonites in contact with natural hyperalkaline groundwaters to determine if any long-term alteration of the bentonite occurs.

Ophiolites have been identified as sources of hyperalkaline groundwaters that can be considered natural analogues of the leachates produced by cementitious materials in repositories for radioactive waste. At the Zambales ophiolite in the Philippines, widespread active serpentinisation results in hyperalkaline groundwaters with measured pH values of up to 11.1, falling into the range typical of low-alkali cement porewaters. These cements are presently being developed worldwide to minimise the geochemical perturbations which are expected to result from the use of OPC-based concretes (see Kamei et al., this conference, for details). In particular, it is hoped that the lower pH of the low-alkali cement leachates will reduce, or even avoid entirely, the potential degradation of the bentonite buffer which is expected at the higher pH levels (12.5 and above) common to OPC-based concretes. During recent field campaigns at two sites in the Zambales ophiolite (Mangatarem and Bigbiga), samples of bentonite and the associated hyperalkaline groundwaters have been collected by drilling and trenching. At Mangatarem, qualitative data from a ‘fossil’ (i.e. no groundwater is currently present) reaction zone indicates some alteration of the bentonite to zeolite, serpentine and CSH phases. Preliminary reaction path modelling suggests that the zeolites could have been produced as a product of smectite reaction in the hyperalkaline groundwaters. Although not included in this calculation to date, the CSH phases identified are completely consistent with reaction of clays with hyperalkaline groundwaters, as seen at other sites worldwide.

At the Bigbiga site, an active hyperalkaline groundwater/bentonite reaction zone (at the base of the bentonite deposit) has recently been drilled and samples are currently undergoing analysis. A comparison of the data sets from both sites will be included in the presentation. The paper will also outline the potential future programme of research at these sites.

6) 40063 – Natural Analogues of Cement: Overview of the Unique Systems in Jordan

Gento Kamei, JAEA (Japan); W Russell Alexander, Bedrock Geosciences (Switzerland); Ian D. Clark, University of Ottawa (Canada);

Paul Degnan, IAEA; Marcel Elie, Shell (UK); Hani Khoury, Elias Salameh, University of Jordan (Jordan); Antoni E. Milodowski, British Geological Survey (UK); Alister F. Pitty, Pitty Consulting (UK);

John A.T. Smellie, Conterra (Sweden)

In L/ILW repository designs, cement-based materials are expected to dominate – for example, the proposed Swiss repository will contain over 1.5 million tonnes of cementitious material. Models of cement evolution predict that leaching of the cement in the repository by groundwater will produce an initial stage of hyperalkaline (pH~13.5) leachates, dominated by alkali hydroxides, followed by a longer period of portlandite and CSH buffered (pH~12.5) leachates. It has also been predicted that the hyperalkaline porewater leached out of the near-field will interact with the repository host rock (and, where applicable, bentonite buffer and backfill). This could possibly lead to deterioration of those features for which the host rock formation and bentonite were originally chosen.

Here, the safety assessment implications of the novel data from the Jordan Natural Analogue Study, which looked into interaction of natural cementitious hyperalkaline leachates on repository host rocks and clays, are presented. Several sites across Jordan have been studied, but the focus here will be on two particularly contrasting sites:

-66-

Abstracts

• Maqarin in northern Jordan – this represents repository host rocks with advective groundwater systems. Hydrogeological, hydrochemical and structural data collected on the fractured rock at the site have been used to assess the likely implications of hyperalkaine leachate interaction on the long-term flow conditions in similar repository host rocks (e.g. granites, fractured sediments) and this will be discussed in detail

• Khushaym Matruk in south-central Jordan – this represents repository host rocks with diffusive groundwater systems. Here, hyperalkaline leachates have diffused through an impermeable, clay-rich sediment, so providing information on the likely controls on leachate interaction in tight repository host rocks (e.g. claystones)

In addition, data are provided on the nature of the secondary phases produced following interaction of the leachates with clays present at the sites. These clays include mixed-layer illite/smectite and so are particularly good analogues of reaction of cement leachates on the bentonite buffer which is an integral part of the EBS in some L/ILW repository designs. This work will be contrasted with that presented by Fujii et al (this conference) which is focussed on bentonite reaction in leachates from low alkali cements. These cements are under consideration for use in repositories where bentonite and concrete will be placed together as they produce lower pH leachates (pH 10 to 11) which are believed to be less aggressive to bentonite than the higher pH leachates from OPC which are discussed here.

SESSION D4: Panel "Applying Lessons Learned from Past D&D Activities"


SESSION R2: Environmental Remediation

1) 40261 – Reclamation of Three In Situ Uranium Mines - Innovative Techniques

Wallace Mays, W M Mining Company (USA)

RECLAMATION OF THREE IN SITU URANIUM MINES IN TEXAS-INNOVATIVE APPROACHES By: Wallace Mays, President of IEC, Chairman and COO of Powertech Uranium, President W M Mining Company From 1990 through 2010, Wallace Mays has been restoring the ground water and reclaiming the surface of three In Situ Leach Uranium Mines in south Texas; the Lamprecht, Zamzow and Pawnee In Situ Leach Mines located in Bee and Live Oak Counties in south Texas. These mines were operated by Westinghouse Corporation subsidiary Wyoming Minerals and by Intercontinental Energy Corporation (IEC). These were among the first In Situ Mines operated and utilized high levels of ammonia carbonate, which complicated the ground water reclamation project. IEC was acquired by Oren Benton, who declared bankruptcy in 1995 in the middle of the ground water restoration process. Westinghouse had set up an annuity to fund their reclamation obligations. W. Mays formed a company, Cima Energy to contract to manage the reclamation and to finance the funding of the working capital to be reimbursed. These ISL mines posed very difficult ground water restoration and surface reclamation problems as they were developed when the industry was developing the techniques that later were successful. In addition, the reclamation bond was limited and required innovative techniques to restore and reclaim more efficiently. The paper will describe the Restoration Funding Methods and speak to issues related to determining bonding and how these could be improved. The paper will present a significantly enhanced ground water restoration procedures, describing the theory and presenting the data from both Reverse Osmosis and Ground Water Sweep restoration methods. Three complete ISL process facilities were dismantled, decommissioned and transported to licensed low level radioactive waste disposal. Innovative decontamination procedures were developed and described in the paper. This paper describes the restoration of pre mining ground water quality in three mines of more than 3.2 kilometers of ground water by circulation more than 3 billion gallons of ground water through reverse osmosis, decontaminating, dismantling and disposal of three complete ISL Process Plants, removing, decontaminating and disposing of more than 50 miles of 4 inch pvc pipe from the well fields, plugging and reclaiming more than 6,600 ISL wells, and transporting to licensed low level radioactive waste disposal sites more than 650 trucks of low level waste. Surface reclamation procedures and problems will be presented.

2) 40005 – Environmental remediation Activities at the Ningyo-toge Uranium Mine, Japan

Hiroshi Saito, Tomihiro Taki, JAEA (Japan)

Ningyo-toge Uranium Mine is located at and around the boundary of Okayama and Tottori Prefectures, western part of Japan. In the Ningyo-toge Mine, exploration activities had been carried out in 1950’s and 60’s after the outcrop was discovered in 1955. Mining activities using galleries and an open-pit had been carried out for sending uranium ores to the conversion and the enrichment plants to 1987 when the mining activities were terminated to begin the environmental remediation activities.

There are many facilities subject to the environmental remediation, including a mill tailings pond, a former open-pit mine and waste rock yards. The main purposes of the environmental remediation common to these facilities, are to take

-67-

Abstracts

measures to reduce the radiation exposure from the exposure pathways to humans in future, and to prevent the occurrence of relevant environmental contamination.

So far, a great number of data have been acquired and technical methods have been examined for the future remediation. And using the above-mentioned data, JAEA has been conducting the remediation activities at the related facilities. Among them, the mill tailings pond, operated since 1965 with the approved volume about 40,000m3, has deposited mining waste and impounded mine water as a buffer reservoir before it is transferred to the water treatment facility. It is located at the upstream of the water-source river, and therefore, its presence is a cause for worry to the local residents. Also, social impact is thought to be extensive in case of an outflow incident of mill tailings, like dam failure by the earthquake. Thus the highest priority has been put to the pond.

JAEA has planned to conduct the remediation and close the pond in coming couple of years. Some activities have already begun, and the results have been produced steadily. According to the current plan, the pond will be covered by the multi-layered capping following dewatering and reshaping of mill tailings. The capping is composed of “radon barrier” for lowering radon-gas dissipation and dose rate, and “low-permeable protective layer” for protecting the radon barrier and reducing the amount of permeated rainwater. Natural material, including bentonite and sand, is planned for use to alleviate the future maintenance. Currently, designing is underway for the upstream half of the pond. Data will be accumulated after capping to verify its effectiveness, and if proved effective, it will be utilized for the capping of the downstream half of the pond.

3) 40092 – Radon impact at a remediated uranium mine site in Japan

Yuu Ishimori, JAEA (Japan)

This paper mainly illustrates the radon impact of the closed uranium mine site remediated in 2007. The site remediated is the waste rock site located on the steep slope of a hill about1.5 km upstream from a residential area along a main ravine. Major remedial action was to cover these waste rock yards with weathering granite soil. The radon flux density after remediation was intended to be 0.1 Bqm-2s-1 in consideration with the natural background level around Ningyo-toge because there is no value of radon flux density regulated in Japan. Our action decreased the radon concentration in the site to natural background level, approximately from 10 to 40 Bqm-3, although relatively high concentration in excess of 100 Bqm-3 was observed before remediation. On the other hand, our action did not decrease the radon concentrations around the site in general. This fact proved that the limited source such as waste rocks affected the radon concentrations at neighboring area only. The similar tendencies were also observed in other environmental data such as radon progeny concentrations. In conclusion, these findings proved that our remedial action was successful against radon. This fact will lead to more reasonable action plans for other closed mine sites.

4) 40243 – Phosphate based remediation techniques: interaction of phosphate with uranium-bound calcite

Chase Bovaird, Dawn Wellman, PNNL (USA)

Despite several decades of studies, effective uranium cleanup strategies remain elusive for contamination in deep subsurface settings that prevail in a number of Department of Energy sites in the western USA. Numerous strategies have been proposed, including iron barriers, soluble reductive agents, and microbial stabilization via reduction and precipitation, but have limited applicability for deep subsurface remediation in an oxidative environment. In-situ phosphate based remediation techniques can potentially delay the precipitation of phosphate phases for controlled in situ precipitation of stabile phosphate phases to control the long-term fate of uranium. The basic principles underlying phosphate stabilization is that aqueous phosphate (PO43-), whether injected as an aqueous solution or solubilized from a source reacts with heavy metals to form insoluble metal-phosphate minerals. The sorption of uranyl species onto minerals is dependent on the nature and availability of binding sites, solution composition and pH, and aqueous complexation. Uranium can be sequestered either through ion exchange or surface complexation, and the rates of release of uranium are dependent on the sorption mechanism. The Hanford Site in southeastern Washington State is a former nuclear defense production facility. Uranium has been identified as a contaminant of concern for groundwater and the deep vadose zone. The vadose zone is comprised of highly alkaline, calcareous sediment. EXAFS analyses at shallow depths suggest that uranium-rich calcite is one of the major controlling phases. Calcite can also serve as a source of Ca2+ and CO32- ions to form mobile, aqueous, uranyl-carbonate species [Ca2UO2(CO3)3] under circumneutral to alkaline conditions. Detailed understanding of the rate and mechanism of the interaction between phosphate and uranium-rich calcite will allow a more effective design of aqueous phosphate-based infiltration strategies to minimize the mobilization of uranium during remediation. The objective of this investigation was to evaluate the interaction of phosphate species with uranium-rich calcite to determine the effects of geochemical conditions on the partitioning of phosphate and its degradation products with uranium-rich calcite, quantify the release of uranium from uranium-rich calcite based on the identity and concentration of aqueous phosphate species, and quantify the rate and mechanism of uranium immobilization based on the identity and concentration of aqueous phosphate species. The information obtained from this line of inquiry is essential to effectively develop phosphate-based remediation strategies for uranium in calcareous environments.

5) 40220 – Remediation of Old Environmental Liabilities in The Nuclear Research Institute Rez plc

Karel Svoboda, Josef Podlaha, Nuclear Research Institute Rez plc (Czech Republic)

-68-

Abstracts

The Nuclear Research Institute Rez plc (NRI) after 55 years of activities in the nuclear field produced some environmental liabilities that shall be remedied. There are three areas of remediation: (1) decommissioning of old obsolete facilities (e.g. decay tanks, RAW treatment technology, special sewage system), (2) processing of RAW from operation and dismantling of nuclear facilities, and (3) elimination of spent fuel from research nuclear reactors operated by the NRI. The goal is to remedy the environmental liabilities and eliminate the potential negative impact on the environment. Remediation of the environmental liabilities started in 2003 and will be finished in 2014. The character of the environmental liabilities is very specific and requires special remediation procedures. Special technologies are being developed with assistance of external subcontractors. The NRI has gained many experiences in the field of RAW management and decommissioning of nuclear facilities and will use its facilities, experienced staff and all relevant data needed for the successful realization of the remediation. The most significant items of environmental liabilities are described in the paper together with information about the history, the current state, the progress, and the future activities in the field of remediation of environmental liabilities in the NRI.

SESSION G2: IAEA Topical for Disused Sealed Radioactive Sources (DSRS)

1) 40028 – International initiatives addressing the safety and security of Disused Sealed Radioactive Sources (DSRS)

Robin Heard, IAEA

High activity radioactive sources provide great benefit to humanity through their utilization in agriculture, industry, medicine, research and education, and the vast majority is used in well-controlled environments. None-the-less, control has been lost over a small fraction of those sources resulting in accidents of which some had serious – even fatal – consequences. Indeed, accidents and incidents involving radioactive sources indicate that the existing regime for the control of sources needs improvement. Additionally, today’s global security environment requires more determined efforts to properly control radioactive sources. Consequently, the current regimes must be strengthened in order to ensure control over sources that are outside of regulatory control (orphan sources), as well as for sources that are vulnerable to loss, misuse, theft, or malicious use. Besides improving the existing situation, appropriate norms and standards at the national and international levels must continue to be developed to ensure the long-term sustainability of control over radioactive sources. In order to improve the existing situation, concerted national and international efforts are needed and, to some degree, are being implemented to strengthen the safety and security of sources in use, as well as to improve the control of disused sources located at numerous facilities throughout the world. More efforts must also be made to identify, recover, and bring into control orphan sources. The IAEA works closely with Member States to improve the safety and security of radioactive sources worldwide. Besides the IAEA Technical Assistance Programme and Technical Cooperation Fund, donor States provide significant financial contributions to the Nuclear Security Fund and/or direct technical support to other States to recover, condition and transfer disused sources into safe and secure storage facilities and to upgrade the physical protection of sources that are in use. Under the USA-Russian Federation-IAEA (“Tripartite”) Initiative, for example, disused sources of a total activity of 2120 TBq (57251 Ci) were recovered and transported into safe and secure storage facilities in six countries of the former Soviet Union. Additionally, physical protection upgrades were performed in thirteen former Soviet Union republics at facilities using or storing high activity radioactive sources. Canada has also provided funding for projects related to the safety and security of radioactive sources in the same region. Additionally, the EU and other countries are making regular and significant contributions to the IAEA for projects aimed at upgrading the safety and security of radioactive sources in South-Eastern Europe, the Middle East, Asia and Africa. Depending on the status of the radioactive source (in use, disused, or orphan) and the actual technical, safety and security situation, several options exist to ensure the source is properly brought or maintained under control. This paper will describe those options and the systematic approach followed by the IAEA in deciding on the most appropriate actions to take for the high activity sources that need to be recovered or removed from the countries under that request assistance.

2) 40303 – Current situation and Management Plan of Radioactive Sources in Japan

Hirokuni Ito, Tadashi Ishii, Tomokazu Ueta, Takao Nakaya, Kenya Suyama, MEXT (Japan)

1, Status of Radioactive Material Distribution in Japan Almost all radioactive products in Japan are directly imported or produced by using imported radioactive materials. Approximately 99% (in terms of radioactivity; as of 2009) of sealed radioactive sources distributed in Japan are imported and supplied through Japan Radio Isotope Association (JRIA).

2, Disused Sealed Radioactive Sources (DSRS) and Orphan sources Substantially, in Japan the users of the sealed radioactive sources will return to manufacturers or distributors. JRIA collects the DSRS based on the sales contract between JRIA and the purchasers of the sealed radioactive sources. The collected DSRS are tested in JRIA’s facility to check the surface contamination and the radioactivity for the transport safety. JRIA also receives orphan radioactive sources in Japan in accordance with MEXT’s administrative instruction if they are found.

JRIA returns collected DSRS, except for the products using very short half life nuclides, to the radioactive source manufacturers in accordance with international guidelines and the import agreement between JRIA and the source manufacturers.

-69-

Abstracts

JRIA keeps the collected DSRS in their own storage facility in cases where they can not return them to the radioactive source manufacturers. However, generally speaking, such DSRS kept temporary in the storage facility of JRIA have been returned successively to the manufacturer, through the discussion among relevant parties if necessary. In Japan, by the system that JRIA receives disused sealed sources based on the sales contract mentioned above, orphan sources are scarcely found. Nevertheless, in such case, they will be picked up and kept safely in storage facility of JRIA.

3, Long term Management Plan of DSRS in Japan There is no disposal facility for DSRS in Japan. But as we explained above, it is not the serious issue which should be resolved soon. This is because i) almost all of DSRS are returned to source manufacturers, ii) JRIA has still enough storage capacity, iii) and the number of DSRS stored in JRIA is settled.

We understand the potential necessity of the disposal facility of DSRS depending on the number of stored DSRS. But considering current situation, it is not required to construct the additional storage facility or the disposal facility for DSRS in Japan. Of course, this view depends on the increase rate of stored DSRS and the capacity of the storage facility in future. We would cooperate with the international society and keep attention to the status of DSRS in Japan for the safety regulation of the radioactive source considering security aspects.

3) 40060 – The Deployment of the Mobile Hot Cell to Condition High Activity Disused Sealed Radioactive Sources (DSRS) for Long Term Storage or Removal

Gerhardus R. Liebenberg, South African Nuclear Energy Corporation (Necsa) (South Africa)

The International Atomic Energy Agency (IAEA) Waste Technology Section with additional support from the U.S. National Nuclear Security Agency (NNSA) through the IAEA Nuclear Security Fund has funded the design, fabrication, evaluation, and testing of a mobile hot cell intended to address the problem of high activity disused sealed radioactive sources (DSRS) in obsolete irradiation devices such as teletherapy heads and dry irradiators.

Operations to condition high activity DSRS using the mobile hot cell has successfully been undertaken in various countries since April 2009. The project was initially targeting the African continent but is now also expanding to other parts of the world such as Latin America and Asia. The mobile hot cell allows for source removal, characterization, consolidation, repackaging in stainless steel capsules as special form, and secure storage of high risk DSRS in modern long term storage shields at single sites in each IAEA Member State.

The mobile hot cell and related equipment is transported in two shipping containers to a specific country where the following process takes place: - Assembly of hot cell - Removal of high activity DSRS from working shields, encapsulation into a stainless steel capsule to obtain special form status and placement into a long term storage shield - Conditioning of any other spent sources the country may require. - Dismantling of the hot cell - Shipping equipment out of country.

This presentation will discuss the design of the mobile hot cell as well as the deployment of the unit for manipulation of high activity DSRS in various countries worldwide. As a result of this project, excess high activity DSRS could be managed safely and securely and possibly be more easily repatriated to their country of origin for appropriate final disposition.

4) 40266 – Problems with Packaged Sources in Foreign Countries

James Matzke, John Zarling, Cristy Abeyta, Joseph A. Tompkins, LANL (USA)

The Global Threat Reduction Initiative’s (GTRI) Off-Site Source Recovery Project (OSRP), which is administered by the Los Alamos National Laboratory (LANL), removes excess, unwanted, abandoned, or orphan radioactive sealed sources that pose a potential risk to health, safety, and national security. In total, GTRI/OSRP has been able to recover more than 19,000 excess and unwanted sealed sources from over 750 sites. In addition to transuranic sources, the GTRI/OSRP mission now includes recovery of beta/gamma emitting sources, which are of concern to both the U.S. government and the International Atomic Energy Agency (IAEA). This paper provides a synopsis of cooperative efforts in foreign countries to remove sealed sources by discussing three topical areas: 1) The Regional Partnership with the International Atomic Energy Agency; 2) Challenges in repatriating sealed sources; and 3) Options for repatriating sealed sources.

5) 40085 – The IAEA's approach to the security of radioactive material

Robin Heard, IAEA

Over the past decade, the threat has increased of terrorism and other malevolent acts by terrorist groups and other malicious non-State actors, involving the potential use of radioactive materials. This has led to an international effort to build a nuclear security framework and regime, both for prevention and consequence management. Legally binding and non-binding international instruments have been established that form the international framework for an effective nuclear security regime. Adherence to and implementation of these instruments is vital for effective nuclear security. IAEA implements a comprehensive programme to assist States in strengthening their nuclear security. The third Nuclear Security Plan covers the period 2010–2013. Through the implementation of these plans, IAEA conducts advisory services and provides technical advice, support and training. It also addresses the longer-term effort of development of nuclear security guidance and it facilitates outreach and information exchange through databases, conferences,

-70-

Abstracts

workshops and fellowships. Nuclear security issues relating to the prevention and detection of, and response to, theft, sabotage, unauthorized access and illegal transfer or other malicious acts involving nuclear material and other radioactive substances and their associated facilities are addressed in the IAEA Nuclear Security Series of publications. These publications are consistent with, and complement, international nuclear security instruments, such as the Code of Conduct on the Safety and Security of Radioactive Sources. Nuclear security missions, evaluations and technical visits continue to be the Agency’s main tool for helping States to assess their nuclear security needs, and provide a basis for formulating plans of action for improving nuclear security. The needs identified by such missions can be subsequently addressed by the State alone, in conjunction with Agency support, or with the assistance of a bilateral partner. Annually, the IAEA conducts more than 60 training events for Member States as well as for Non-Member States. These are based on findings and insights resulting from the various advisory missions and organized in response to the requests formulated by the States themselves. Providing urgently needed technical upgrades and equipment has been a foundation for IAEA assistance to States in enhancing the security of radioactive material since the establishment of the Nuclear Security Programme. The equipment needed was provided to States as follow-up to assessment missions and the training needed to operate the equipment was arranged in separate events.

6) 40058 – Radioactive Waste Management in Lebanon

Munzna Assi, Lebanese Atomic Energy Commission (Lebanon)

The disused sealed radioactive sources including orphan sources in Lebanon, along with the growing industry of sealed radioactive sources in medical, industrial and research fields have posed a serious problem for authorities as well as users due to the lack of a national store for disused radioactive sources. Assistance form International Atomic Energy Agency (IAEA) was requested to condition and store disused radium needles and tubes present at two facilities. The mission took place on July 25, 2001 and was organized by the (IAEA) in cooperation with the Lebanese Atomic Energy Commission (LAEC). Other disused radioactive sources were kept in the facilities till a safer and securer solution is provided; however orphan sources, found mainly during export control, were brought and stored temporarily in LAEC. The necessity of a safe and secure store became a must. Prior to October 2005, there was no clear legal basis for establishing such store for disused radioactive sources, until the ministerial decree no 15512 dated October 19, 2005 (related to the implementation of decree-law no 105/83) was issued which clearly stated that “The LAEC (Lebanese Atomic Energy Commission) shall, in cooperation with the Ministry of Public Health, establish a practical mechanism for safe disposal of radioactive waste”. Following this, the work on inventory of disused sealed sources along with collecting orphan sources and placing them temporarily in LAEC was legally supported. Moreover, several missions were planned to repatriate category I and II sources, one of which was completed specifically in August 2009; other missions are being worked on. In 2008, a national technical cooperation project with the International Atomic Energy Agency, IAEA, was launched. Under the reference TC number LEB3002, the project was entitled “"Assistance in the establishment of a safe temporary national storage at the Lebanese Atomic Energy Commission for orphan sources and radioactive waste" which cycle is 2009-2011. Under this project, a national store for radioactive sources in the third basement of LAEC is being established. The area is being reconstructed currently and will be equipped when ready under LEB3002 project. Along with this, a system for sealed disused sources management has been prepared, part of which is applied now and the rest will be applied upon the establishment of the store. This paper will cover the inventory collection process, the study for the establishment of this store, the present and prospective waste management system, and the the waste acceptance criteria.

7) 40029 – The ultimate solution – disposal of Disused Sealed Radioactive Sources (DSRS)

Robin Heard, IAEA

For countries with no access to existing or planned geological disposal facilities for radioactive wastes, the only options for managing high activity or long-lived disused radioactive sources are to store them indefinitely, return them to the supplier or find an alternative method of disposal. Disused sealed radioactive sources (DSRS) pose an unacceptable radiological and security risk if not properly managed. Out of control sources have already led to many high-profile incidents or accidents. Countries without solutions in place need to consider the future management of DSRSs urgently. In the frame of a regional Technical Cooperation (TC) project, a number of countries in the African region have come together under the auspices of the IAEA to explore the option of a Borehole Disposal Concept (BDC) for their small inventories of DSRSs.

Disposal in boreholes is intended to be simple and effective, meeting the same high standards of long-term radiological safety as any other type of radioactive waste disposal. It is believed that the BDC can be readily deployed with simple, cost-effective technologies. These are appropriate both to the relatively small amounts and activities of the wastes and the resources that can realistically be found in African countries. The South African Nuclear Energy Corporation Ltd (Necsa) has carried out project development and demonstration activities since 1996. The project looked into the technical feasibility, safety and economic viability of BDC under the social, economic, environmental and infrastructural conditions currently prevalent in Africa. Conceptually, the disposal concept comprises a borehole (diameter ranging from 150 to 260 mm) drilled down to a depth ranging from 30 to 100 metres. The depth will be dependent on the site-specific safety assessment. The borehole will have a casing with a plug at the bottom. Grouting will then be applied to seal the annulus and all fractures and crevices outside of the casing. The spacing of waste packages is ~1 metre. The space between packages is backfilled with a suitable material such as cement or concrete grout. In the

-71-

Abstracts

generic design of the Borehole Disposal Concept, a 100-meter deep borehole is to be filled with 50 packages up to the depth of 50 metres. The rest will be backfilled with concrete to act as a borehole plug.

The project was completed in 2004 and subject to an IAEA peer review by an independent group of international experts in April 2005. One of the main outcomes of the international peer review was the positive statement by the expert team that the BDC developed by Necsa had been demonstrated to be a safe, economic, practical and permanent means of disposing of disused radioactive sealed sources.

Implementation is near at hand with work being done in Ghana with support from the IAEA. Here the site selection is complete and studies are being carried out to test the site parameters for inclusion into the safety assessment.

SESSION L6: Waste Treatment

1) 40055 – Drying System For Radioactivated Metal Waste From Nuclear Power Station

Nobuhito Ogaki, Yasushi Ooishi, Hironori Takabayashi, Masamichi Obata, Taichi Horimoto, Toshiba Corporation (Japan)

High dose rate metal waste from core internals, BWR channel box or control rod is stored in fuel pool or site bunker pool. Waste form for final disposal of these high dose rate metal waste should elinimate water to prevent hydrogen gas which is caused by radialysis of water. Toshiba's newly developed drying system enables short drying time and easy maintenance. Toshiba will provide the total system to fabricate the waste form for high dose rate metal waste.

2) 40186 – Macroporous Catalysts for Hydrothermal Oxidation of Metallorganic Complexes at Liquid Radioactive Waste Treatment

Valentin Avramenko, Dmitry Marinin, Vitaly Mayorov, Alexander Mironenko, Marina Palamarchuk, Valentin Sergienko, Institute of Chemistry FEDRAS (Russia)

One of the main problems of liquid radioactive waste (LRW) management is concerned with treatment of decontamination waters containing organic ligands. The organic ligands like oxalic, citric and ethylenediaminetetraacetic acids form stable complexes with radionuclides which puts restrictions on application of many technologies of LRW management. One of the ways of destruction of metallorganic complexes consists in using the catalytic oxidation. However, the heterophase catalytic oxidation is rather problematic due to formation of metal oxides on the catalyst surface and calmatation of meso- and micropores. A possible solution of the above problem can be found in synthesis of macroporous catalysts for oxidation having a regular macroporous structure. The present paper describes the template synthesis of macroporous metalloxide catalysts performed with using siloxane-acrylate microemulsions as templates. The method for impregnation of precious metals (PM) particles into the template, which enables one to produce PM nanoparticles of a specific size and immobilize them in the porous structure of the synthesized metalloxide catalysts, is presented. A possible mechanism of the synthesis of macroporous catalysts is suggested and the comparison of the electronic and photon-correlation spectroscopy results obtained at different stages of catalysts synthesis was conducted.

3) 40163 – Impermeable graphite: A new development for the waste management of irradiated graphite

Johannes Fachinger, Karl-Heinz Grosse, Furnances Nuclear Applications Grenoble (Germany); Richard Seemann, Milan Hrovat ALD (Germany)

Graphite is a geological stable material proven by its natural occurrence. However its porous structure anticipates the use of graphite as long term stable waste matrix for final disposal. A slow corrosion in aquatic phases can be induced by high irradiation doses in the range of 10E-5 to 10E-7 g/m²d. The porous structure is related to large surface areas and therefore the radiation induced corrosions process cannot be neglected for final disposal. Furthermore aqueous phases will penetrate into the pore system and radionuclides adsorbed on the surface will be dissolved. All this problems can be solved with a graphite material with a closed pore system. A graphite composite material with an inorganic binder has been developed with a density > 99,9 % of theoretical density and therefore a negligible porosity. A first draft calculation predicts that the life time of such a material will be at least 2 orders of magnitude higher than porous graphite. This material represents a long term stable leaching resistant matrix for the embedding of i-graphite. Granulated i-graphite will be mixed with natural graphite and inorganic binder and pressed as block. Other radioactive wastes can be embedded in this matrix additionally, e.g. coated particles of spent fuel from HTR reactors. First investigations proofed the expected pore free structure and good mechanical strength properties. The paper further will present a selection of potential additonal applications of the new material as embedding material for of different nuclear waste streams like noble metals from reprocessing or iodine.

-72-

Abstracts

4) 40165 – THOR® Steam Reforming Technology for the Treatment of Ion Exchange Resins and More Complex Wastes such as Fuel Reprocessing Wastes

J. Brad Mason, Corey Myers, Studsvik, Inc. (USA)

The THOR fluid bed steam reforming process has been successfully operated for more than 10 years in the USA for the treatment of low- and intermediate-level radioactive wastes generated by commercial nuclear power plants. The principle waste stream that has been treated is primarily ion exchange resins but various liquids, sludges and solid organic wastes have also been treated. The primary advantages of the THOR process include: (a) maximum volume reduction on the order of 5:1 to 10:1 depending on the waste type and waste characteristics; (b) environmentally compliant off-gas emissions, (c) reliable conversion of wastes into mineralized products that are durable and leach-resistant, and (d) no liquid effluents.

Over the past five years, the THOR process has been adapted for the treatment of more complex wastes including historic defense wastes, reprocessing wastes, and other wastes associated with the fuel cycle. As part of U. S. Department of Energy (DOE) environmental remediation activities, the THOR dual bed steam reforming process was successfully demonstrated to process: Idaho National Laboratory (INL) Sodium-Bearing Waste (SBW); Savannah River Tank 48 High Level Waste (HLW); and Hanford Low Activity Waste (LAW) and Hanford Waste Treatment Plant Secondary Waste (WTP SW) liquid slurry simulants. The THOR process has been demonstrated in pilot plant operations to successfully process various simulated liquid, radioactive, nitrate-containing wastes into environmentally safe, leach-resistant solid mineral products. The solid products incorporate normally soluble ions, such as sodium, potassium, cesium, technetium, and sulfate, chloride and fluoride salts into an alkali alumino-silicate mineral matrix that inhibits the leaching of those ions into the environment. The solid mineral products produced by the THOR process exhibit durability and leach resistance characteristics that are superior to borosilicate waste glasses. As a result of this work, a full-scale THOR process facility is currently under construction at the DOE’s Idaho site for the treatment of SBW and a full-scale facility is in the final design stage for the DOE’s Savannah River Site for the treatment of Tank 48 Waste.

Recent work has focused on the development of new monolithic waste formulations, the extension of the THOR process to new waste streams, and the development of modular THOR processes for niche waste treatment applications. This paper will provide an overview of current THOR projects and summarize the processes and outcomes of the regulatory and safety reviews that have been necessary for the THOR process to gain acceptance in the USA.

5) 40257 – Phase Behavior and Reverse Micelle Formation in Supercritical CO2 with DTAB and F-pentanol for Decontamination of Radioactive Wastes

Kensuke Kurahashi, Osamu Tomioka, Yoshihiro Meguro, JAEA (Japan)

Decontamination of radioactive wastes is useful for the volume reduction and re-categorization of them. However, a large amount of secondary wastes are often generated from a wet decontamination using traditional solvents. Therefore, the development of decontamination method decreasing the secondary wastes is required. Supercritical CO2 (scCO2), which is gaseous matter at ordinary conditions, has the potential to minimize the amount of solvent wastes. However, neat scCO2 is not suitable for the separation of polar materials, such as metal compounds, because non-polar scCO2 cannot dissolve them. A reverse micelle is a stable aggregate of amphiphilic surfactants surrounding a water pool in a non-polar solvent. Polar materials can be dissolved in the water pool and dispersed in the nonpolar solvent. Therefore, the formation of reverse micelle in scCO2 could increase the availability of scCO2 as a solvent for separation of radionuclides from radioactive wastes. In the present study, we investigated the reverse micelle formation in scCO2 with dodeceytrimethylammonium bromide (DTAB) and 2,2,3,3,4,4,5,5-octafluoro-1-pentanol (Fpentanol). After putting a mixture of DTAB, F-pentanol and water in a high pressure view cell of 54 cm3 at 45 °C, CO2 was introduced into the cell slowly. When the mixture was dissolved completely in the scCO2, the pressure was measured as the cloud point pressure (CPP). When 1.08 mmol DTAB were employed, 13.5 mmol H2O could be dissolved in scCO2 with 6.3 vol% F-pentanol at 20.9 MPa. Formation of reverse micelles was confirmed by dissolving an aqueous solution of methyl orange into the scCO2 with DTAB and water. The CPP values were determined for various volume of water. The CPP values increased slowly with an increase of water concentration in the cell less than 0.35 mol/dm3. The increase of water concentration led the growth of reverse micelle size, and then the CPP value would rise. On the other hand, the CPP did not simply increase with the water concentration more than 0.35 mol/dm3 and a sharp and concave peak appeared in the relationship between the CPP value and the water concentration. This suggests that the state of water in the scCO2 changed before and after the water concentration of 0.35 mol/dm3. The water concentration giving the CPP peak depended on the F-pentanol concentration but was independent of the DTAB concentration. From this fact, it is expected that the interaction between F-pentanol and water affects the state of water in the scCO2.

-73-

Abstracts

SESSION H7: Performance Assessment Modeling and Parameters

1) 40305 Keynote – Development of a Realistic Repository Performance Assessment Method

Joonhong Ahn, UCB (USA)

Abstract Not Available

2) 40204 – Integrated model for the near field of a repository in granite host-rock. Probabilistic approach

Lara Duro, Alba Valls, Olga Riba, Jordi Bruno, Amphos XXI Consulting S.L.(Spain); Aurora Martinez-Esparza, ENRESA (Spain)

The application of probabilistic approaches to the performance of underground repositories for long-lived radioactive waste has received special attention in the last years. Numerous exercises have been developed in order to elicit the Probability Distribution Functions (PDFs) of the several parameters needed for these developments. Several integrated models allow the implementation of PDFs in the long-term simulations needed for Performance Assessment. In this work we present how the deterministic compartmental model for a repository of high level nuclear waste (HLNW) located in a crystalline host-rock has been modified to include PDFs for some of the parameters. The implementation of probabilistic approaches gives also information on the most influencing parameter on the migration of the different radionuclides from a deep repository concept.

3) 40017 – Spatial Variability and Parametric Uncertainty in Performance Assessment Models

Osvaldo Pensado, James Mancillas, Scott Painter, Southwest Research Institute (USA); Yasuo Tomishima, AIST (Japan)

The problem of defining an appropriate treatment of distribution functions (which could represent spatial variability or parametric uncertainty) is examined based on a generic performance assessment model for a high-level waste repository. The generic model incorporated source term models available in GoldSim®, the TDRW code for contaminant transport in sparse fracture networks with a complex fracture-matrix interaction process, and a biosphere dose model known as BDOSE(TM). Using the GoldSim framework, several Monte Carlo sampling approaches and transport conceptualizations were evaluated to explore the effect of various treatments of spatial variability and parametric uncertainty on dose estimates. Results from a model employing a representative source and ensemble-averaged pathway properties were compared to results from a model allowing for stochastic variation of transport properties along streamline segments (i.e., explicit representation of spatial variability within a Monte Carlo realization). We concluded that the sampling approach and the definition of an ensemble representative do influence consequence estimates. In the examples analyzed in this paper, approaches considering limited variability of a transport resistance parameter along a streamline increased the frequency of fast pathways resulting in relatively high dose estimates, while those allowing for broad variability along streamlines increased the frequency of "bottlenecks" reducing dose estimates. On this basis, simplified approaches with limited consideration of variability may suffice for intended uses of the performance assessment model, such as evaluation of site safety.

4) 40203 – Development of a Radiolytic Model for the Alteration of Spent Nuclear Fuel. Incorporation of non-oxidative matrix dissolution and hydrogen oxidation inhibition effect

Lara Duro, Alba Valls, Olga Riba, Jordi Bruno, Amphos XXI Consulting S.L.(Spain); Aurora Martinez-Esparza, ENRESA (Spain)

In the last years, there have been numerous efforts from national waste management agencies to develop models able to predict the dissolution behaviour of spent nuclear fuel under interim and/or long-term storage conditions. One of the most evolved models is the so called Matrix Alteration Model (MAM), which is based on the radiolytic oxidative dissolution of UO2 (Martínez-Esparza et al., 2004) and which has been applied to different experimental results with certain level of agreement. The calibration of the MAM model in front of new experimental data has resulted in the identification of some important drawbacks that may result in limited applicability of the model as a predictive tool. In this work we present the modifications made to the MAM in order to improve it and expand its range of applicability: - Incorporation of the non oxidative alteration of the matrix. Ignoring the incorporation of the non-oxidative alteration of the matrix has proved to be non-conservative over long-term experiments, resulting in an underestimation of the actual concentrations of uranium and radionuclides dissolving congruently with the matrix (Bruno et al., 2009). The incorporation of this mechanism in the MAM has been done by considering the different rates of irradiated and unirradiated UO2+x determined under reducing conditions and published in the open scientific literature. The modified MAM is able to reproduce experimental data gathered under a diverse range of experimental conditions. Incorporation of the catalytic effect of the surface on hydrogen activation. Hydrogen can be generated by radiolysis of water in the vicinity of the spent nuclear fuel as well as by anaerobic corrosion of metallic components forming on the container of the fuel

-74-

Abstracts

under storage conditions. The ability of molecular hydrogen to decrease the rate of dissolution of the spent fuel matrix has been attributed to the presence of metallic surfaces in the fuel that can act as catalyst for the activation of molecular to atomic hydrogen (Carbol et al., 2009). The MAM model integrates the H2 inhibition effect on UO2 matrix oxidation by including reactions between H2 and the UO2 oxidising species, (i.e. consumption of oxidants species by H2) (Duro et al., 2009) but until now the activation of hydrogen on the surface of the solids present in the system had not been implemented. In this work we present the implementation of the heterogeneous activation of hydrogen in order to reproduce experimental conditions and couple the predictions of the long-term rates of fuel dissolution of the matrix.

5) 40172 – Evaluated and Estimated Solubility of Some Elements for Performance Assessment of Geological Disposal of High-level Radioactive Waste Using Updated Version of Thermodynamic Database

Akira Kitamura, Reisuke Doi, JAEA (Japan); Yasushi Yoshida, Inspection Development Co., Ltd. (Japan)

A thermodynamic database was established for performance assessment of geological disposal of high-level radioactive waste (HLW) and TRU waste (equated to long-lived intermediate level waste) based on the thermodynamic database (JNC-TDB) developed by the Japan Nuclear Cycle Development Institute in 1999. Twenty-five elements (actinides, fission products, activated nuclides and their progenies) which were important for the performance assessment of geological disposal were selected. The fundamental plan was in principle based on the guidelines established by the Nuclear Energy Agency (NEA) in the Organisation for Economic Co-operation and Development (OECD). Additional unique guidelines were established due to a requirement from the performance assessment to select tentative thermodynamic data obtained from chemical analogues and/or models for elements and species with insufficient thermodynamic data. Thermodynamic data for nickel, selenium, zirconium, technetium, thorium, uranium, neptunium, plutonium and americium, which were critically reviewed by the NEA, were taken to our thermodynamic database and modified. Thermodynamic data for cobalt and molybdenum, which were important for the performance assessment of TRU waste, were newly reviewed, selected and compiled. Some of thermodynamic data for other elements were updated or modified. All thermodynamic data were extrapolated zero ionic strength using the model called “specific ion interaction theory (SIT)”. We tried to assign uncertainty of the selected data as many as possible. Selected thermodynamic data were compiled as the JAEA’s thermodynamic database (JAEA-TDB), which were available for some geochemical calculation programs, e.g., PHREEQC. We evaluated and estimated solubility of the 25 elements in the simulated pore waters established in the second progress report (H12) for safety assessment of geological disposal of HLW in Japan using the JAEA-TDB and compared with the solubility values evaluated and estimated using the JNC-TDB. Furthermore, we tried to establish a technique to determine the solubility limiting solid for all the elements of interest. It was found that most of the evaluated and estimated solubility values were not changed drastically, but the solubility values and dominant aqueous species for some elements were changed using the JAEA-TDB, e.g., due to introducing the formation constant of polynuclear hydrolysis species of zirconium and replacing the formation constant of mixed carbonatohydoxo complexes of thorium. Detail of the comparison and discussion about the evaluated and estimated solubility values between the JAEA- and the JNC- TDBs will be presented.

6) 40049 – Consideration on Soil Origin Carbon Transfer to Leafy Vegetables Using Stable Carbon Isotope Ratios

Keiko Tagami, Shigeo Uchida, National Institute of Radiological Sciences (Japan)

From the viewpoint of radiation dose assessment for humans from transuranic waste disposal sites, carbon-14 (half-life: 5730 y) in organic forms is thought to be one of the most important radionuclides. Thus, understanding the C-14 fate in soil-to-crop systems is important since C-14 might transfer from waste disposal sites to crops through the soil environment. It is well known that carbon dioxide in the air is the major source for carbon assimilation by plants, thus, carbon in soil is less considered for plant uptake. Recent results by radiotracer experiments, however, indicated some contribution of soil origin carbon to plant carbon. In this study, we used stable carbon isotope ratios (C-13/C-12) in crop and soil samples. It is well known that C-13 and C-12 are fractionated in the photosynthesis process at a certain ratio, depending on plant types, i.e. C3, C4 and CAM plants. If all the carbon in crops originated only from the air, the stable carbon isotope ratio would take a constant value; however, if there were any contributions from soil carbon, the ratio would be changed. This approach has been used in our previous study for rice (Tagami and Uchida, 2009) and we tried to use the method for leafy vegetables. Leafy vegetables were cabbage, Chinese cabbage, lettuce, and leeks, and associated soil samples were also collected in Japan. Stable carbon isotope ratios and total C concentrations using an elemental analyzer connected to an isotope ratio mass spectrometer (Thermo Fisher Scientific, Flash EA and Delta V Advantage). Finally we found that the percentage of soil origin carbon to the total plant carbon was 2.9%, and the soil-to-plant transfer factors ranged 0.12-1.1 with an average of 0.5, which was within the previously reported values for radish (Sheppard et al., 1991).

7) 40050 – Comparison of Soil-to-plant Transfer Factors for Rice and Wheat Grains

Shigeo Uchida, Keiko Tagami, National Institute of Radiological Sciences (Japan)

It is important to understand the behaviors of radionuclides in the agricultural systems since radionuclides enter humans through ingestion of foods. Critical staple foods are rice and wheat; the former is the dominant staple food crop

-75-

Abstracts

in humid tropical and sub-tropical countries across the globe, and the latter is cultivated worldwide. Both are members of the Poaceae Family; however, they are different in cultivation methods. For example, rice is produced under flooded conditions but wheat is grown under upland field conditions. The difference in cultivation methods may affect plant uptake of radionuclides from soil. Thus we measured soil-to-grain transfer factors (TFs) of naturally existing elements as analogues of radionuclides for rice and wheat cultivated in Japan. For this work, we collected from 63 sites for rice and 7 sites for wheat; to the results of wheat, we added 2 barley samples since their cultivation methods were the same. The elements we measured for TFs were Li, N, Na, Mg, Al, Si, P, K, Ca, Sc, Ti, V, C, Mn, Fe, Co, Ni, Cu, Zn, Ga, As, Se, Br, Rb, Sr, Y, Zr, Nb, Mo, Ag, Cd, Sn, Sb, I, Cs, Ba, La, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Hf, W, Tl, Pb, Ra, Th and U. Geometric means of TFs of these elements were compared between rice and wheat grains; the tendency was almost the same and the correlation factor was more than 0.98 by Student’s t-test (p<0.001). However, there were elements showing TF differences by a factor of 5 or more; that is, As, Ba and Ra. The obtained TF values were also compared with recently published data for cereals (IAEA TRS-472, 2010). Details will be discussed at the presentation.


1) 40121 – Development of Characterization Methodology for Fault Zone Hydrology

Kenzi Karasaki, LBNL (USA); Tiemi Onishi, US Geological Survey (USA); Erika Gasperikova, LBNL (USA); Junichi Goto, Tadashi Miwa, Hiroyuki Tsuchi, NUMO (Japan);

Keiichi Ueta, Kenzo Kiho, Kimio Miyakawa, CRIEPI (Japan)

Several deep trenches were cut, and a number of geophysical surveys were conducted across the Wildcat Fault in the hills east of Berkeley, California. The Wildcat Fault is believed to be a strike-slip fault and a member of the Hayward Fault System, with over 10 km of displacement. So far, three boreholes of ~ 150m deep have been core-drilled and borehole geophysical logs were conducted. The rocks are extensively sheared and fractured; gouges were observed at several depths and a thick cataclasitic zone was also observed. While confirming some earlier, published conclusions from shallow observations about Wildcat, some unexpected findings were encountered. Preliminary analysis indicates that Wildcat near the field site consists of multiple faults. The hydraulic test data suggest the dual properties of the hydrologic structure of the fault zone. A fourth borehole is planned to penetrate the main fault believed to lie in-between the holes. The main philosophy behind our approach for the hydrologic characterization of such a complex fractured system is to let the system take its own average and monitor a long term behavior instead of collecting a multitude of data at small length and time scales, or at a discrete fracture scale and to “up-scale,” which is extremely tenuous.

2) 40189 – An attempt to evaluate horizontal crustal movement by geodetic and geological approach in the Horonobe area, northern Hokkaido, Japan

Tetsuya Tokiwa, Koichi Asamori, Tadafumi Niizato, Tsuyoshi Nohara, JAEA (Japan); Yuki Matsuura, Hitachi Zosen Co., Ltd. (Japan); Hideki Kosaka, Kankyo Chishitsu Co., Ltd. (Japan)

In this study, we present the preliminary results for the estimation of a horizontal crustal movement by using geodetic and geological approach in the Horonobe area, northern Hokkaido, Japan. The estimations have been carried out by using a GPS data and a geological cross section obtained by applying balanced-section method. As results of this study, both of the shortening rates estimated by GPS data and balanced-section method indicate several millimeters per year. Namely, there is no contradiction between geodetic and geological data, and it is considered that Horonobe area is still situated similar tendency and magnitude of a crustal movement. It is seemingly considered that geodetic data is unhelpful for estimating the long-term crustal movement, because period of geodetic observations is a very short. However, the result of this study indicates that geodetic data provide valuable information for estimating the long-term crustal movement in the area, and it is considered that geodetic approach play an important role in improvement of the credibility of evaluation for prediction of long-term stability.

3) 40054 – Relationship between hypocentral distribution and geological structure in the Horonobe area, northern Hokkaido, Japan

Tetsuya Tokiwa, Koichi Asamori, Naoto Hiraga, Osamu Yamada, Hideharu Yokota, JAEA (Japan); Hirokazu Moriya, Tohoku University (Japan);

Hikaru Hotta, Itaru Kitamura, Construction Project Consultants, Inc. (Japan)

In this paper, we discuss the relationship between the accurate hypocentral distribution and three-dimensional (3-D) geological structure in and around the Horonobe area in northern Hokkaido, Japan. The multiplet-clustering analysis was applied to the 421 micro-earthquakes which occurred from 1 September, 2003 to 30 September, 2007. The 3-D geological structure model was mainly constructed from previous seismic reflection profiles and borehole data. As a result of this analysis, although with slight differences in depth between them, the hypocenters were found to be distributed in the NNW-SSE direction and become deeper from the west toward the east. The distributed pattern of the

-76-

Abstracts

hypocenters is similar to that of the geological structure. These results indicate that the hypocentral distribution may represent existence of active zone related to the geological structure, and provide effective information which can contribute to establishing methods for estimating the future evolution of the geological environment.

4) 40062 – Technical know-how for modeling of geological environment (1) Overview and groundwater flow modeling

Hiromitsu Saegusa, Shinji Takeuchi, Keisuke Maekawa, Hideaki Osawa, Takeshi Semba, JAEA (Japan)

It is important for site characterization projects to manage the decision-making process with transparency and traceability and to transfer the technical know-how accumulated during the research and development to the implementing phase and to future generations. The modeling for a geological environment is to be used to synthesize investigation results. Evaluation of the impact of uncertainties in the model is important to identify and prioritize key issues for further investigations. Therefore, a plan for site characterization should be made based on the results of the modeling. The aim of this study is to support for the planning of initial surface-based site characterization based on the technical know-how accumulated from the Mizunami Underground Research Laboratory Project and the Horonobe Underground Research Laboratory Project. These projects are broad scientific studies of the deep geological environment that are a basis for research and development for the geological disposal of high-level radioactive wastes. In this study, the work-flow of the groundwater flow modeling, which is one of the geological environment models, and is to be used for setting the area for the geological environment modeling and for groundwater flow characterization, and the related decision-making process using literature data have been summarized.

5) 40066 – Technical Know-how for Modeling ff Geological Environment (2) Geological Modeling

Toshiyuki Matsuoka, Kenji Amano, Hideaki Osawa, Takeshi Semba, JAEA (Japan)

It is important for site characterization projects to manage the decision-making process with transparency and traceability and to transfer the technical know-how developed and accumulated during the research and development to the implementing phase as well as to future generations. The modeling of a geological environment supports efforts to clarify the degree of understanding regarding that geological environment, including uncertainty. Evaluation of the impact of uncertainties in a geological environment model is important to identify and prioritize key issues for further investigations. Therefore, a plan for site characterization should be made based on the results of the modeling. The aim of this study is to support the planning of initial surface-based site characterization based on the technical know-how accumulated from the Mizunami Underground Research Laboratory Project and the Horonobe Underground Research Laboratory Project. These projects are broad scientific studies of the deep geological environment that are a basis for research and development for the geological disposal of high-level radioactive wastes. In this study, the work-flow followed in developing the geological model, one of the geological environment models, and the related technical know-how acquired from literature data have been summarized.

6) 40039 – The long-term stability of geological environments in the various rock types in Japan from the perspective of uranium mineralization

Eiji Sasao, JAEA (Japan)

Long-term stability of the geological environment is one of the important keys for deep geological disposal of high-level radioactive waste in the Japanese Islands due to their location in a tectonically active island-arc. Uranium occurrences in Japan have been subjected to many geological processes inherent to the island-arc setting. Geological environments associated with uranium mineralization are considered favorable for HLW disposal, because uranium mineralization is considered a natural analogue of the radionuclides in HLW. Studies on the long-term stability of the uranium mineralization in Japan can be instructive as these could provide useful information on the long-term stability of the geological environment. Information on host rock and mode of occurrence of uranium mineralization was compiled from published data. The mineralization occurs in these types of deposits, i.e., sedimentary formations, association with metallic ore mineralization of magmatic origin and stratiform manganese mineralization, pegmatite, and alluvial placer deposit. The mineralization occurs in various geological settings in Japan. This fact suggests that geological environments suitable for geological isolation are widely distributed in the Japanese Islands, despite their location in a geologically active area. This study will support building confidence in HLW disposal in the Japanese Islands.

SESSION D5: Measurement and Estimation

1) 40045 – Improvement of Radioactivity Inventory Evaluation Procedure In Preparatory Tasks for Decommissioning

Ken-ichi Tanaka, Hidenori Tanabe, Hideaki Ichige, JAPC (Japan)

-77-

Abstracts

Preparatory tasks for decommissioning of nuclear power plant start with radiological characterization. Residual radioactivity inventory evaluation is a main part of the characterization. Reliable information on the inventory is important for specification for decommissioning plan. Japan Atomic Power Company (JAPC) has already started these tasks for Tsuruga Nuclear Power Plant Unit 1 (TS-1). We can optimize decommissioning plan using the information. To obtain the reliable information, we improved an evaluation procedure. The procedure is divided into two main steps. First step is neutron flux distribution calculation and second one is radioactivity distribution calculation. Radioactivity distribution is calculated using neutron flux distribution. In this work, we improved the evaluation procedure to obtain the reliable information on the inventory Because of the limitation of computer resource, two-dimension (2D) approximation model was applied to radioactivity distribution around Reactor Pressure Vessel (RPV). We can calculate reliable 2D neutron flux distribution by having better understanding of neutron transport phenomena. Neutron flux was measured at 30 locations in TS-1 Primary Containment Vessel (PCV) using activation foils. And in order to understand the neutron transport phenomenon inside the PCV, we also calculated neutron flux distribution with the three-dimensional (3D) discrete ordinates method calculation (Sn) code. By consideration about the result of the measurement and 3D calculation, we could understand the characteristics of the neutron flux distribution inside the PCV. To simulate the neutron flux distribution well with 2D Sn code, neutron flux behaviors inside the PCV had been investigated with referencing the measurement values and with observing calculated 3D neutron flux distribution. 2D calculation model had been modified repeatedly until reliable calculation result was provided. After several model modifications, the reliable 2D calculation was accomplished and important neutron transport phenomena that are necessary to simulate the neutron flux distribution well was understood. Network-parallel-computing technique was applied to radioactivity distribution calculation. Using this technique, we could calculate radioactivity at all space mesh points that were used with 2D Sn code and we obtained the radioactivity distribution. By using this distribution, we can estimate a quantity of radioactivity around RPV more accurately and optimize dismantling designs.

2) 40202 – Verification of Source Term Analysis System for Decommissioning Wastes from CANDU Reactor

Dong-Keun Cho, Gwang-Min SUN, Jongwon Choi, KAERI (Korea Rep.); Donghyeun Hwang, Hak-Soo Kim, Tae-Won Hwang, KHNP (Korea Rep.)

There are now twenty commercial nuclear power reactors operating as of May 2010 in South Korea. As nuclear capacity becomes higher and installations age, the Korean government and industry have launched R&D to estimate appropriate decommissioning costs of power reactors. In this paper, MCNP/ORIGEN2 code system which is being developed as a source term evaluation tool was verified by comparing the estimated nuclide inventory from MCNP/ORIGEN2 simulation with the measured nuclide inventory from chemical assay in an irradiated pressure tube discharged from Wolsong Unit 1 in 1994. Equilibrium core model of Wolsoung unit 1 was used as a neutron source to activate in-core and ex-core structural components. As a result, the estimated values from the analysis system agreed with measured data within 20% difference. Therefore, it can be concluded that MCNP/ORIGEN system could be a reliable tool to estimate source terms of decommissioning wastes from CANDU reactor, although this system assumes constant flux irradiation and snapshot equilibrium core model as a reference core.

3) 40294 – Evaluation of The Activated Radioactivity of Turbine Equipments in BWR

Masato Watanabe, Motonori Nakagami, Chubu Electric Power Co., Inc. (Japan)

The equipments in the main steam system under the plant operation were irradiated by the neutron emitted from N-17 which is produced by (n,p) reaction of the coolant in Boiling Water Reactor (BWR) plants. However, only few measures, regarding the radioactive concentration of N-17 in the main steam system of BWR, were indicated up to now.

In this study, the radioactive concentration of N-17 in the main steam system of the advanced boiling water reactor (ABWR) plant, Hamaoka unit-5, is evaluated by the neutron flux measurements and calculations.

It was found that the concentration of N-17 at the reactor pressure vessel outlet nozzle of the main steam pipe is approximately 3 Bq/cm3. It turns out that the concentration has small dependency on nuclear thermal power and that it is relatively common among the BWR plants.

Then, the activated activities of turbine equipments and pipes in the main steam system are evaluated by using the measured radioactive concentration of N-17. The activities of the equipments in the main steam system are relatively very low level.

4) 40014 – Optimization of Quantitative Waste Volume Determination Technique for Hanford Waste Tank Closure

Yi Su, David L. Monts, Ping-Rey Jang, Zhiling Long, Walter P. Okhuysen, Olin P. Norton, Lawrence L. Gresham, Jeffrey S. Lindner, Mississippi State University (USA)

The Hanford Site is currently in the process of an extensive effort to empty and close its radioactive single-shell and double-shell waste storage tanks. Before this can be accomplished, it is necessary to know how much residual material is left in a given waste tank and the uncertainty with which that volume is known.

The Institute for Clean Energy Technology (ICET) at Mississippi State University is currently developing a quantitative in-tank imaging system based on Fourier Transform Profilometry, FTP. FTP is a non-contact, 3-D shape

-78-

Abstracts

measurement technique. By projecting a fringe pattern onto a target surface and observing its deformation due to surface irregularities from a different view angle, FTP is capable of determining the height (depth) distribution (and hence volume distribution) of the target surface, thus reproducing the profile of the target accurately under a wide variety of conditions. Hence FTP has the potential to be utilized for quantitative determination of residual wastes within Hanford waste tanks. In this paper, efforts to characterize the accuracy and precision of quantitative volume determination using FTP and the use of these results to optimize the FTP system for deployment within Hanford waste tanks are described.

5) 40120 – Implementation of Decommissioning Materials Conditional Clearance Process to the OMEGA Calculation Code

Matej Zachar, Vladimir Necas, Slovak University of Technology (Slovakia); Vladimir Daniska, DECONTA a.s. (Slovakia)

The activities performed during nuclear installation decommissioning process inevitably lead to the production of large amount of radioactive material to be managed. Significant part of materials has such low radioactivity level that allows them to be released to the environment without any restriction for further use. On the other hand, for materials with radioactivity slightly above the defined unconditional clearance level, there is a possibility to release them conditionally for a specific purpose in accordance with developed scenario assuring that radiation exposure limits for population not to be exceeded. The procedure of managing such decommissioning materials, mentioned above, could lead to recycling and reuse of more solid materials and to save the radioactive waste repository volume. In the paper an implementation of the process of conditional release to the OMEGA Code is analyzed in details; the Code is used for calculation of decommissioning parameters. The analytical approach in the material parameters assessment, firstly, assumes a definition of radiological limit conditions, based on the evaluation of possible scenarios for conditionally released materials, and their application to appropriate sorter type in existing material and radioactivity flow system. Other calculation procedures with relevant technological or economical parameters, mathematically describing e.g. final radiation monitoring or transport outside the locality, are applied to the OMEGA Code in the next step. Together with limits, new procedures creating independent material stream allow evaluation of conditional material release process during decommissioning. Model calculations evaluating various scenarios with different input parameters and considering conditional release of materials to the environment are performed to verify the implemented methodology. Output parameters and results of the model assessment are presented, discussed and concluded in the final part of the paper.

6) 40183 – Quantitative determination of the initial components in the activated pressure tubes of the Wolsong

1st CANDU reactor

Gwang-Min Sun, Dong-Keun Cho, KAERI (Korea Rep.)

When the end of the original lifespan is coming for the Wolsong 1st CANDU reactor in Korea, decommissioning has been one of the largest issues requiring large expense, which is facing the nuclear industry and the government. Korean government intends to provide public funds, which have been accumulated using a portion of the electrical charge, to dismantle reactors when the time comes. The radioactive sources in the construction stuffs and components of reactor such as pressure tube, steam generator and so on must be evaluated for the estimation of appropriate expense for the decommissioning. The overall objective of this study is to confirm the methods for the elemental or isotopic analysis of the pressure tube samples from nuclear reactor in order to make possible the reliable application of the methods to the case of the Wolsong 1st CANDU reactor. To achieve this objective we need to categorize the methods into two groups taking into account the applicability of the methods to the target we have interests in. The radiochemical analytical methods such as liquid scintillation counting, X-ray spectroscopy and normal gamma-ray spectroscopy and so on to measure the trace isotopic beings such as 60Co, 94Nb, 54Mn, 55Fe, 59Ni, 63Ni and so on. Elemental analysis methods such as instrumental neutron activation analysis (INAA) and prompt gamma activation analysis (PGAA) to measure the initial compositions of the samples. In present work, we will deal with INAA and PGAA to determine an elemental composition of the pressure tube. The prepared samples are irradiated in the NAA #1, NAA #2 and SNU-KAERI PGAA facility of the HANARO research reactor in Korea Atomic Energy Research Institute. Because the samples are from pressure tube of reactor, the most abundant element must be iron and other elements can be determined relative to the iron content. From this idea, we determined the initial elemental composition of the samples through two-step procedure. For the absolute determination of iron content in the sample, the count rate must be calibrated according to the iron elemental content as shown in figure 8 where the slope of the line is a specific count rate or analytical sensitivity given in cps/?g. The relative mass of other elements in the samples were determined by using a k0-standardization method which has been usually used in the INAA and also in the PGAA.

-79-

Abstracts

SESSION R3: ER Techniques

1) 40286 – Sequential Extraction and Determination of Depleted Uranium in the Presence of Natural Uranium in Environmental Soil samples by ICP-MS

Mohamed Amr, Alaa E. Negmeldin, Khalid Al-Saad, A. T. Al-Kinani, Qatar University (Qatar); A. I. Helal, Atomic Energy Authority (Qatar)

Determination of depleted uranium (DU) in the presence of natural uranium (NU) by inductively coupled plasma mass spectrometry (ICP-MS) was applied on environmental soil samples collected from Qatar. The soils were artificially spiked by soaking them in a mixture of DU and NU. The detection limit of 235U and 238U isotopes were 141 ppt and 1.28 ppt, respectively. It was observed from sequential experiment that uranium was brought into solution mainly appeared at steps for dilute acid-soluble, carbonate-bound and organic matter-bound species. Little redistribution was observed at steps for exchangeable and Fe-Mn oxide-bound species.

2) 40096 – Determination of Environmental Uranium Concentration by Utilizing Gamma-ray Emission from the Progeny Radionuclides

Tadao Tanaka, Taro Shimada, Takenori Sukegawa, JAEA (Japan); Takeshi Ito, Japan ATOX Co., Ltd. (Japan)

Nuclear facility sites such as enrichment plant and fabrication plant are allowed to be released from nuclear safety regulations after the plants are decommissioned. The sites are necessary to confirm to be decontaminated, prior to be released.

Gamma-ray emission from the progeny radionuclides of uranium such as Th-234, Pa234m, Ra-226 has been utilized for the determination of uranium concentration in soils. Gamma-ray emission radionuclides occurring in spacious areas of land was often measured by the in-situ method with portable germanium semiconductor detector (portable Ge detector), to confirm that there is no significant gamma-ray emission radionuclides distributed in the vast land areas. In the present study, we proposed a determination method for U-238 concentration of background level in environment and for probate of vast site areas, in which the gamma-ray radioactivities from Th-234, Pa-234m, Ra-226 are measured with the portable Ge detector.

Validity of the estimation method of U-238 concentration from the progeny radionuclides was examined by the comparison between the U-238 concentration estimated by the in-situ method with portable Ge detector and that directly measured by ICP-MS. The U-238 concentration by the in-situ method was estimated from peak counting rate at 63 keV of the gamma-ray emission corresponding to Th-234 and from that at 186 keV corresponding to Ra226. The estimated U-238 concentration was in the order of 0.01 Bq/g in radioactive concentration, and was in comparable level with the concentrations decided by the ICP-MS.

The determination method of U-238 concentration from the progeny radionuclides was applied to the site contaminated by low-level uranium resourced from an uranium handling facility. The U-238 concentration could be determined from the peak counting rates corresponding to Th-234 and Ra-226. The proposed method utilizing gamma-ray radioactivities from the progeny radionuclides may be available for the U-238 concentration determination in vast land areas.

3) 40034 – Effect of Fertilizer and Soil Amendments on Extraction Yields of Radioiodine and Radiocesium in Soil

Hirofumi Tsukada, Akira Takeda, Shunichi Hisamatsu, Institute for Environmental Sciences (Japan)

Mobile fraction of radionuclides in a soil, which is water-extractable and exchangeable, is important information for remediation in the terrestrial environment. Radioiodine and radiocesium derived from atmospheric nuclear weapons tests and released from nuclear facilities are major radionuclides for the assessment of radiation exposure to the public. In the present study, the effects of fertilizer and soil amendments on the extraction yields of radioiodine and radiocesium in a soil were investigated. A surface soil (Andosol) was collected from grassland in Aomori, Japan. The soil was mixed with the following amendments: Chemical fertilizer (NH4, PO4, K): 0.24 mg N, 0.16 mg P2O5, or 0.16 mg K2O per g soil, Compost (rice straw mixed with cattle feces): 50 mg per g soil, and Clay mineral (illite, kaolinite, montmorillonite, sericite, vermiculite, zeolite): 20 mg per g soil.

They were mixed and then 10 kBq of 125I or 137Cs was added to one g of the soil sample. They were stored in an artificial climate chamber, and a wetting-and-drying treatment was repeated by adding one ml of deionized water every 2 weeks. The extraction yields of 125I with water, and 137Cs with water or 1 M ammonium acetate solution were determined at 10, 30 and 120 d after addition. The extraction yield of 125I in the soil without amendment (the control sample) decreased to 4% of the added amount at 120 d after addition. The extraction yield of 125I in the soil mixed with the compost at 10 and 30 d after addition was higher than that in each control soil, and the difference was accompanied by an increase of the dissolved organic carbon content in the water extract of the compost application soil. The extraction yield of 125I in the soil mixed with the montmorillonite was higher than that in the control soil through the aging period. The extraction yield of 137Cs in the control soil with water and ammonium acetate decreased to 0.12% and 32% at 120 d after addition, respectively. The extraction of 137Cs with water was enhanced by the application of the nitrogen fertilizer and the compost, whereas that was depressed by the application of the clay minerals. The extraction of 137Cs with

-80-

Abstracts

ammonium acetate was also depressed by the clay minerals, and this may be caused by the large cesium interception sites in the clay minerals. It was clear that the mobile fractions of radioiodine and radiocesium in the soil were changed by several treatments. This work was conducted under contract with the Aomori Prefectural Government, Japan.

4) 40246 – Impact of Mobile-Immobile Water Domains on the Retention of Technetium (Tc-99) in the Vadose Zone

Danielle Jansik, Dawn Wellman, Elsa Cordova, PNNL (USA)

The transport of technetium (Tc-99), like many other radionuclides, is of interest due to the potential for human exposure and impact on ecosystems. Technetium has been released to the environment through nuclear power production and nuclear fuel processing; as a result, further spreading of Tc-99 is a concern at DOE sites across the US. Specifically, technetium is a contaminant of concern at Hanford, Savannah River, Idaho, and Oak Ridge National Laboratory. The current body of work conducted on Tc-99 has provided a wealth of information regarding the redox relationships, sorption, solubility, and stability of the mineral phases (Artinger et al., 2003; Beals and Hayes, 1995; Cui and Eriksen, 1996b; Gu and Schulz, 1991; Jaisi et al., 2009; Keith-Roach et al., 2003; Kumar et al., 2007), however little work has been conducted on the physical transport of the highly soluble pertechnetate oxyanion, in the subsurface. Current conceptual models do not explain the persistence and presence of the anion in deep vadose zone environments such as the Hanford site. In an oxic reducing environment with low organic content the residence time of technetium is the soil would be expected to be near low, due to its low sorption. Surprisingly, nearly 50 years following the release of contamination into the site, much of the element has persisted in the subsurface, in its most mobile form. Using an Unsaturated Flow Apparatus (UFA) we have conducted a series of experiments to examine the impact of mobile-immobile domains on the transport of Tc-99. By varying sand/silt ratios and saturations we examined how changes in pore geometry and moisture content impact the transport of Tc-99 within our experimental system. Results demonstrating the impact of sediment texture pore morphology, and soil moisture content on physical impediments to Tc-99 transport will be presented.

5) 40122 – Remediation of 153Gd-contaminated sand by fulvic and humic materials extracted from fallen cherry leaves

Takumi Kubota, Kyoto University Research Reactor Institute (Japan)

Natural organic substances can provide the remediation of metal-contaminated soil with their complex capability and are environment-friendly due to their biodegradability in remediation where those, used as eluent, may remain in the environment. It is furthermore desirable to use eluent obtained from plant wastes in the point of recycling.

In this study organic substances were extracted from fallen cherry leaves with dilute NaOH, separated into fulvic and humic substances by acid treatment, and then added to contaminated sand samples which were artificially loaded with 153Gd for laboratory experiment. Gadolinium, one of lanthanides, was used for soil contamination and remediation in consideration that lanthanides are trivalent in aqueous solution and strong to adsorb onto soil rather than monovalent and divalent and that lanthanides are chemically similar to trivalent actinides which are released through use of nuclear power.

In our elution examination a two gram of aliquot of the Toyoura standard sand (Yamaguchi, Japan) was loaded with 153Gd in 0.01M HCl and was washed with four ml fulvic or humic eluent whose pH was regulated with HCl and/or NaOH. The distribution of 153Gd between eluant and soil was determined from gamma activity.

The elution ratio of 153Gd increased with the concentration of total organic carbon (TOC) in eluent. The effect of decontamination with organic substances was large at pH of 3 to 7. At lower pH the decontamination was regulated by acid solution (e.g. HCl) rather than organic substances because of competition of hydrogen and metal ion to them. Fulvic substances of 4000 ppm TOC at pH ranging from 3 to 5 yielded the elution ratio of 50% and humic substances of 700 ppm TOC at pH ranging from 6 to 7 yielded that of 40%. The complexation capability of fulvic substance was smaller than EDTA. The comparison between both solutions of the same TOC of 15ppm showed the elution ratio of EDTA and fulvic substances were 60% and negligible, respectively. Gadolinium humate, recovered as eluant, was added with 6M HCl to be separated into acidic gadolinium solution and gadolinium-free (153Gd-free) humic precipitation.

Soil remediation was simulated with fulvic and humic substances, which indicated the effective decontamination capability of high TOC organic substances, easily extracted from fallen cherry leaves, and the feasibility of repeated use of humic substances purified by acid treatment for environmental remediation.

SESSION L7: Storage and Disposal Facility

1) 40284 – Microbial Occurrence in Bentonite-based Buffer Materials of a Final Disposal Site for Low Level Radioactive Waste in Taiwan

Fong-In Chou, Chia-Chin Li, Tzung-Yuang Chen, National Tsing Hua University (Taiwan); Hsiao-Wei Wen, National Chung Hsing University (Taiwan)

-81-

Abstracts

This research addresses the potential of microbial implications in bentonite for use as a buffer and backfill material in final disposal site for low-level radioactive waste (LLRW) in Taiwan, where has a special island-type climate. Microbe activities naturally present in this site were analyzed, and buffer materials (BM) consisted of 100%, 70% or 50% bentonite were prepared for laboratory studies. A total of 39 microbial strains were isolated, and the predominant strains included four bacterial, one yeast and four fungal strains. Growth inhibition was not detected in any tested strain cultured in a radiation field with a dose rate of 0.2 Gy/h. Most of the isolated strains grew under a dose rate of 1.4 Gy/h. The D10 values of the tested strains ranged from 0.16 to 2.05 kGy. The mycelia of tested fungal strains could spread over 5 cm during six months of inoculation in BM. The spreading activity of the tested bacteria was less than that of the fungi. Moreover, biofilms were observed on the surfaces of the BM. Since a large and diverse population of microbes is present in Taiwan, microbes may contribute to the mobilization of radionuclides in the disposal site.

2) 40153 – Assessing the gas transport mechanisms in the Swiss L/ILW concept using numerical modeling

Irina Gaus, Paul Marschall, Joerg Rueedi, Nagra (Switzerland); Rainer Senger, John Ewing, Intera Inc. Swiss Branch (Switzerland)

In low/intermediate-level waste (L/ILW) repositories, anaerobic corrosion of metals and degradation of organic materials produce hydrogen, methane, and carbon dioxide. Gas accumulation and gas transport in a L/ILW repository is an important component in the safety assessment of proposed deep repositories in low-permeability formations. The dominant gas transport mechanisms are dependent on the gas overpressures as with increasing overpressure the gas transport capacity of the system increases. The dominant gas transport mechanisms occurring with increasing gas pressure within the anticipated pressure ranges are: diffusion of gas dissolved in pore water (1), two phase flow in the hostrock and the excavation damaged zone (EDZ) whereby no deformation of the pore space occurs (2), gas migration within parts of the repository (if repository materials are appropriately chosen) (3) and pathway dilation (4). Under no circumstances the gas is expected to induce permanent fractures in the hostrock. This paper focuses on the gas migration in parts of the repository whereby materials are chosen aimed at increasing the gas transport capacity of the backfilled underground structures without compromising the radionuclide retention capacity of the engineered barrier system (EBS). These materials with enhanced gas permeability and low water permeability can supplement the gas flow that is expected to occur through the EDZ and the host rock. The impact of the use of adapted backfill and sealing materials on the gas pressure build-up and the major gas paths were assessed using numerical two-phase flow models on the repository scale. Furthermore, both the gas and water fluxes as a function of time and gas generation rate can be evaluated by varying the physical properties of the materials and hence their transport capacity. Results showed that by introducing seals with higher gas permeability, the modeled gas flow is largely limited to the access tunnels and the excavation disturbed zone for the case of a very low permeability host rock. The bulk of the gas flows through the repository seal and the adjacent EDZ into the tunnel system. In addition to the demonstration of the gas flow in the seal and access tunnel system by numerical models, laboratory results confirm the high gas transport capacity of the sand/bentonite mixtures. In a next step a multi year demonstration scale experiment (GAST) at the Grimsel Test Site is envisioned.

3) 40283 – The progress and results of Demonstration Test of Cavern-Type Disposal Facility

Yoshihiro Akiyama, Kenji Terada, Nobuaki Oda, Tsutomu Yada, Takahiro Nakajima, RWMC (Japan)

The cavern-type disposal facilities for low-level waste (LLW) with relatively high radioactivity levels mainly generated from power reactor decommissioning and for part of transuranic (TRU) waste mainly from spent fuel reprocessing are designed to be constructed in a cavern 50 to 100 meters below ground, and to employ an engineered barrier system (EBS) of a combination of bentonite and cement materials in Japan. In order to advance the feasibility study for these disposal, a government-commissioned research project named Demonstration Test of Cavern-Type Disposal Facility started in fiscal 2005, and since fiscal 2007 a full-scale mock-up test facility has been constructed under actual subsurface environment. The main objective of the test is to establish construction methodology and procedures which ensure the required quality of the EBS on-site. By fiscal 2009 some parts of the facility have been constructed, and the test has demonstrated both practicability of the construction and achievement of the quality. They are respectively taken as low-permeability of less than 5x10-13 m/s and low-diffusivity of less than 1x10-12 m2/s at the time of completion of construction. This paper covers the project outline and the test results obtained by the construction of some parts of a bentonite and cement materials.

SESSION H9: Repository Engineering and Demonstration

1) 40304 Keynote – Repository engineering and demonstration: special challenges for TRU

Ian G. McKinley, McKinley Consulting (Switzerland)

The diverse range of long-lived radioactive wastes without significant heat output specified for deep geological disposal (here termed TRU) pose challenges that are potentially more serious than those from vitrified high-level waste

-82-

Abstracts

and spent fuel. Despite this, the latter tend to be the focus of R&D in national programmes. Such challenges are particularly severe for the case for countries that are not considering evaporite host rocks or have a volunteering approach to siting and those with inventories of TRU resulting from reprocessing of spent fuel. While there is little doubt that safe disposal of TRU is feasible, it is tricky to develop a convincing safety case for a site during early stages of characterisation as, compared to HLW/SF, less credit can be taken for robust, long-term performance of current designs of the engineered barrier systems. In order to improve this situation and increase flexibility with respect to host rock properties, two different options are available – improving the conditioning of particular waste streams or improving the overall repository safety concept. Although the former has been a focus for work in some countries (particularly Japan), much less effort has been invested in the latter and hence this will be illustrated by some examples. In the paper these options will be compared in terms of their pros and cons with respect to practicality of implementation, environmental impact and cost. Additionally, the ease with which the resulting safety case can be supported by demonstrations of key arguments will be discussed, which may indicate the likely degree of acceptance by stakeholders.

2) 40119 – Half-Scale Test: An important step to demonstrate the feasibility of the Belgian Supercontainer concept for disposal of HLW

Lou Areias, SCK•CEN and VUB (Belgium); Bart Craeye, Ghent University/Artesis Hogeschool Antwerpen (Belgium);

Geert De Schutter, Ghent University (Belgium); Hughes Van Humbeeck, William Wacquier, ONDRAF/NIRAS (Belgium);

Alain Van Cotthem, Loic Villers Technum-Tractebel (Belgium)

This paper presents results of a half scale test performed by ESV EURIDICE, an Economic Interest Grouping between the Belgian Nuclear Research Centre (SCK•CEN) and the Belgian Agency for Radioactive Waste and Enriched Fissile Materials (ONDRAF/NIRAS). The primary objective of the test was to assess the feasibility of constructing the Supercontainer and to provide experimental data to validate modelling calculations obtained using the finite element program HEAT/MLS. The test focused on the early age behaviour of the concrete matrix materials and the practical aspects of construction. Generally, the results obtained from the half scale test confirm that it is feasible to construct the Supercontainer with currently available techniques. The results also validate scoping calculations obtained earlier with the finite element model. These findings contribute an important step to demonstrate the feasibility to construct the Supercontainer and to validate the Belgian Supercontainer concept proposed by ONDRAF/NIRAS for disposal of high level waste (HLW) in Belgium.

Keywords: Belgian supercontainer concept, high-level radioactive waste, half-scale test, concrete buffer, overpack, construction feasibility

3) 40175 – Full-Scale Test on Overpack Closure Techniques for HLW Repository Operation - Welding Methods and UT Systems For Long-Term Structural Integrity of the Weld Joint -

Ario Nakamura, Hidekazu Asano, RWMC (Japan); Takashi Furukawa, JAPAN POWER ENGINEERING AND INSPECTION CORPORATION (Japan);

Kyosuke Fujisawa, KOBE STEEL, LTD. (Japan); Susumu Kawakami, IHI Corporation (Japan); Takashi Ito, Mitsubishi Heavy Industries, Ltd. (Japan).

Overpack, a high-level radioactive waste package for Japan’s geological disposal program, is required for preventing sealed vitrified waste from contact with groundwater for at least 1,000 years. The weld joint between the body and lid must also meet this requirement. Certain welding methods were examined for applicability through full-scale welding tests using various welding depths up to 190 mm and two different lid structures. Results show that generation of welding flaws must be considered unavoidable. Therefore, ultrasonic testing (UT) must be conducted on the assumption that weld flaws will be present. Such UT systems must be designed for natural defects. Several types of UT must be evaluated for detection and size estimation capability at depths ranging from the surface to the bottom of the weld joint. Certain UT methods were examined for their ability to detect natural defects that were created on the surface of and inside a 190-mm thick carbon steel specimen. Probability of detection (POD) of each UT method was calculated by comparing the results of UT and destructive examination. In consideration of the preferred range of scanning depth for each UT method, a concept that combines UT methods was proposed as a practical UT system for the overpack weld joint.

KEYWORDS: radioactive waste management, geological disposal, waste package, overpack, final closure weld, structural integrity, natural defects, UT, POD

4) 40242 – Design Options for HLW Repository Operation Technology, (I) Demonstration and Evaluation of Remote Handling Technologies

Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan); Hideki Kawamura, Obayashi Corporation (Japan)

The long-term safety of a geological repository for high-level radioactive waste(HLW) in Japan is achieved using a multi-barrier system, with emphasis on an engineered barrier system (EBS) consisting of bentonite-based buffer material and a steel overpack containing vitrified wastes. Providing a wide variety of technical options for repository operation,

-83-

Abstracts

which can allow EBS construction under various repository conditions, is valuable for the volunteering approach to siting a HLW repository in Japan. In order to confirm technological applicability of candidate techniques and to evaluate compatibility in an operation system, full-scale tests of various remote-handling and-emplacement techniques for the EBS were conducted. The test results are summarized in the technical menu which is hierarchical structured database. Since the long-term performance of EBS is significantly influenced by remote handling and emplacement techniques for EBS, consistency of the operation technology must be evaluated from the view point of system standing. In this program, evaluation methodology was studied and several indexes, representing requirements for the repository operation and management, were identified. Indices include current technical availability, long-term safety of the resultant EBS, operational safety, engineering reliability, retrievability, etc., with expert assessment complementing a more quantitative evaluation.

5) 40251 – Design Options for HLW Repository Operation Technology, (II) Bentonite Block Forming and Vertical Emplacement/JGC

Hajime Takao, Tatsuhiro Takegahara, JGC Corporation (Japan); Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan)

RWMC and JGC have been running an all-round R&D program for the period of 2000-2010 to develop the concept of Vertical Emplacement for disposal of vitrified waste. The conceptual design of its basic equipment was worked out in 2000, followed by forming the large-scale bentonite block in 2001-2004. Study has also been conducted on a mechanism to convey and position the large-scale block using a vacuum suction device. Subsequent to these developments, various technologies necessary for designing the Vertical Emplacement equipment have been reviewed, which would enhance engineering feasibility and reliability.

Full-scale demonstration program under a joint research program with JAEA (Japan Atomic Energy Agency) started in 2008 with the twin objectives i) supporting of public relations and ii) technical verification. The large-scale bentonite block and part of the full-scale Vertical Emplacement equipment are now on view at the Full-scale demonstration facility in Horonobe, Hokkaido, Japan.

6) 40268 – Design Options for HLW Repository Operation Technology, (III) Transportation and Horizontal Emplacement of Pre-Fabricated EBS Module (PEM)

Susumu Kawakami, IHI Corporation (Japan); Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan) As one of the repository operation technologies for high-level radioactive waste?(HLW), the pre-fabricated

engineered barrier system (EBS) module (PEM) was carried out the examination of handling and emplacement technique for EBS. The PEM technology was examined to confirm technological applicability. The PEM is concept of the integration of EBS as the module in the surface facilities, and transporting the module underground facilities. This concept is the one of the candidate concepts of horizontal emplacement techniques for EBS in Japan. Therefore, PEM is the same level large size and heavy weight as EBS, and it is necessary to examine the applicability of handling and emplacement techniques. Full-scale level tests were performed to confirm the applicability of these techniques with the air bearing/air jack devices. In the tests, we prepared the testing devices of full-scale level size/weight and confirmed the applicability of these technologies as an elemental technology on the condition of considering the environment of an underground tunnel. The air bearing test that produced the surface-roughness of the tunnel environment was carried out the evaluation concerning the transportation performance of the air bearing. And, the air jack test was carried out the holding and emplacement of PEM. The repository operation technology with the air bearing/jack device was confirmed to execute the examination, and to apply to handling and emplacement technique for PEM.

7) 40236 – Design Options for HLW Repository Operation Technology, (IV) Shotclay Technique for Seamless Construction of EBS

Ichizo Kobayashi, Soh Fujisawa, Makoto Nakajima, Masaru Toida, Kajima Corporation (Japan); Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan)

In Japan, the construction method of the buffer material has been investigated focusing on the block emplacement and in-situ compaction methods. Under the current concept of geological disposal of radioactive waste, since it has been important that the barrier satisfies the dry density requirement as a mass, no enough attention has been paid to the distribution of density and the gaps between blocks which were caused by the conventional methods. This is based on the assumption that bentonite swells from the permeation of groundwater and that the density eventually becomes homogeneous. However, it is not clear whether the high density bentonite does in fact swell until the density is homogeneous. The effects of density distribution and the gaps in long-term high impermeability of a bentonite-engineered barrier may be not small. Therefore, the option of the construction method for a bentonite-engineered barrier that does not have such uncertainty concerning the implementation of a disposal concept should be prepared. In order to remove such uncertainty, the high-density spray method for bentonite, termed the shotclay method, was developed as a method for constructing a uniform bentonite-engineered barrier without gaps even in a narrow space. The shotclay method is used to construct compacted soil at high density by spraying material preconditioned with water. Using this method, the dry density of 1.6 Mg /m3, which was considered impossible with the

-84-

Abstracts

spraying method, is achieved. In this study, the applicability of the shotclay method to HLW bentonite-engineered barriers was confirmed experimentally. In the tests, an actual scale vertical-type HLW bentonite engineered barrier was constructed. This was a bentonite-engineered barrier with a diameter of 2.22 m and a height of 3.13 m. The material used was bentonite with 30% silica sand, and water content was adjusted by mixing chilled bentonite with powdered ice before thawing. Work progress was 11.2 m3 and the weight was 21.7 Mg. The dry density of the entire buffer was 1.62 Mg/m3, and construction time was approximately 8 hours per unit. After the formworks were removed, the core and block of the actual scale HLW bentonite-engineered barrier were sampled to confirm homogeneity. As a result, homogeneity was confirmed, and no gaps were observed between the formwork and the buffer material and between the simulated waste and the buffer material. The applicability to HLW of the shotclay method has been confirmed through this examination.

8) 40254 – Design Options for HLW Repository Operation Technology, (V) Preliminary Study and Small Scale Experiments on the Method of Removal of Buffer Material with Salt Solution

Satohito Toguri, Jiho Jang, Takashi Ishii, Mitsunobu Okihara, Kengo Iwasa, SHIMIZU CORPORATION (Japan); Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan)

During the construction of geological disposal facilities for high-level radioactive waste, it may be decided to free and retrieve the emplaced overpack for some reasons. Thus we have been paying attention to a method of slurrying bentonite buffer around the overpack with fluid (salt solution) for freeing it. A few laboratory tests were performed to research the feasibility and verify the applicability of the method. The test piece of the buffer material consists of 70wt% bentonite and 30wt% sand, and its dry density is 1.6 g/mL, and the volume of the cylindrical test piece is 100mL. First, the time was measured for test pieces to be dissolved in various strength NaCl solutions. In case of 47% saturated test pieces, they could be easily dissolved in NaCl 3 or 4wt% solution, and the dissolved material was deposited soon in the slurry. Second, the results from the flushing experiments applied to a small specimen of buffer material suggest that flushing 4wt% NaCl solution is effective for speedy stripping of buffer material. Third, the times were measured for various percentage saturated test pieces to be dissolved in 4wt% NaCl solution. The time for 47% saturated one was shorter than 1 Hr, but the time for 67% was longer than 5 Hrs. In case of 88%, it had not been dissolved yet in 7Hrs, but after drying the test piece it could be easily dissolved in 5 minutes. The results from the laboratory tests indicate that the unsaturated buffer material can be easily dissolved in NaCl 4wt% solution, and the wet buffer material can be dissolved by immersing in NaCl 4wt% solution after process of dry-hot-air blowing. Finally, in order to confirm the effect of removal processes, simulated experiments were executed using small scale (1/14 scale) specimens. The 47% saturated buffer material specimen was easily removed by the process of immersing in NaCl solution, flushing NaCl solution, removing slurry by vacuum device, and re-using NaCl solution after deposition. In the nearly saturated case, it was removed in cyclic process of dry-hot-air blowing and immersing in NaCl 4wt% solution.


1) 40135 – Dry-run of Site Investigation Planning using the Manual for Preliminary Investigation in Japan

Shigeki Akamura, Tadashi Miwa, NUMO (Japan); Tatsuya Tanaka, Obayashi Corporation (Japan); Hiroshi Shiratsuchi, Tokyo Electric Power Service Co.,Ltd. (Japan);

Atsushi Horio, DIA CONSULTANTS CO., Ltd. (Japan)

A stepwise site selection process has been adopted for geological disposal of HLW in Japan. Literature surveys, followed by preliminary investigations (PI) and, finally, detailed investigations in underground facilities will be carried out in the successive selection stages. In the PI stage, surface-based investigations such as borehole surveys and geophysical prospecting will be implemented.

NUMO recognizes that sustained improvement of internal expertise is very important to ensure that the PI will be implemented rigorously and efficiently. Therefore existing knowledge and experience in the planning and management of site investigations were compiled in the form of two manuals: the Preliminary Investigation Planning Manual (PIPM) and the Preliminary Investigation Management Manual (PIMM). The manuals were based on the experience in overseas site investigation programs, and has further been refined by taking experience from site investigations in Japan (e.g. from generic URLs) into account. NUMO has applied them in its own R&D programs, such as the dry-run studies for various geological environments (e.g. coastal and island).

This paper outlines the process and the results of a dry-run study which applied the revised PIPM to the Yokosuka area where the demonstration and validation project for PI technologies has been conducted. The planning of the PI was performed according to the eight steps in the PIPM. The GET (Geosphere Evaluation Team) consisting of staffs from site characterization, performance assessment and repository design groups in NUMO, was established for the PI planning. This task force team has responsibility not only for the planning but also for making key decisions regarding the performance of the program. The SDMT (Site Descriptive Modeling Team) was formed for site modeling activities. Initial geoscientific conceptual models were established based on the interpretation of existing site information around the area. Various conceptual models were constructed due to the limited amount of existing information. Site descriptive

-85-

Abstracts

models were derived taking such variations in the conceptual models into account. The main targets for the PI planning were specified considering the uncertainties in the models, requests from the repository concept as well as legal requirement. The integrated PI program set out the sequence of investigations and characterization activities to achieve the main targets. Lessons learned in the PI planning and the applicability of the revised PIPM were addressed as the conclusion of the dry-run.

2) 40070 – Evaluation of the long-term evolution of the groundwater system in the Mizunami area, Japan

Takashi Mizuno, Teruki Iwatsuki, JAEA (Japan); Antoni E. Milodowski, British Geological Survey (UK)

This study aimed to develop a methodology for assessing the evolution of the long-term groundwater system, using fracture-filling calcite. Fracture-filling calcite mineralization, closely associated with groundwater flow paths, in deep (to ca. 1000 m) granitic rocks in and around Mizunami area, Japan, was studied by optical and cathode-luminescence microscopy, SEM, laser ablation microprobe-ICP MS, stable isotope geochemistry, and fluid inclusion analysis. As the results, four generations (I to IV) of calcite precipitation can be differentiated based on their morphological and isotopic characteristics. Calcite I - calcite with indistinct morphology including the wall rock fragments (d13CPDB -31.4 ‰ ~ -7.5 ‰, d18OPDB -32.7 ‰ ~ -8.0 ‰); Calcite II -calcite with euhedral rhombohedral and hexagonal crystal forms (d13CPDB 22.5 ‰ ~ -1.8 ‰, d18OPDB -15.5 ‰ ~ -6.5 ‰); Calcite III - calcite forming c-axis elongated rhombohedral overgrowths seeded on top of Calcite II (d13CPDB -18.0 ‰ ~ 5.8 ‰, d18OPDB -11.1 ‰ ~ -2.3 ‰); Calcite IV - calcite forming small rhombohedral crystals nucleated on the surface of Calcite III (d13CPDB -12.5 ‰ ~ -1.7 ‰, d18OPDB -12.7 ‰~7.8 ‰). Carbon and oxygen isotopic ratios suggests that the Calcite I is of hydrothermal origin. On the other hand, Calcite II, IV and III were precipitated from freshwater and marine water, respectively. From the geohistorical point of view, depositional setting in Mizunami area was changed during Tertiary and Quaternary. Mizunami Group overlying above Toki granite is changed from lacustrine strata to marine strata. In addition, upper Seto Group is a lacustrine stratum. The change of depositional setting corresponds to groundwater system assumed by mineralogical study. As there is no other evidence linked to penetration of high-salinity water into deep environment in this area, it is suggested that both of deposition of marine strata (upper part of Mizunami group) and forming Calcite III precipitation were possibly caused by same transgression event. After Calcite III precipitation, calcite IV was precipitated from fresh water during flushing of the marine water. As summary, integrated morphological, mineralogical, microchemical and isotopic analysis of multilayered calcite fracture mineralization associated with groundwater flow paths provides valuable information to evaluate long-term evolution of groundwater system.

3) 40077 – Study on the Estimation Error Caused by Using One-dimensional Model for the Evaluation of Dipole Tracer Test

Yuji Ijiri, Yumi Naemura, Taisei Corporation (Japan); Kenji Amano, Keisuke Maekawa, Kunio Ota, Takanori Kunimaru, Atsushi Sawada, JAEA (Japan)

In-situ tracer tests are of value in obtaining parameters for repository performance assessment. A one-dimensional model is simple and has thus been commonly employed to identify the radionuclide transport parameters by fitting the model to breakthrough curves obtained from the tracer tests. It can, however, be considered that the one-dimensional model could increase uncertainty in the identified parameters. In particular, such errors are not negligible when the parameters are evaluated for the test conducted in a dipole (twodimensional) flow field between injection and pumping wells. In this study, the effects of various experimental conditions including pumping rate, dipole ratio, heterogeneity of fracture transmissivity and background groundwater flow on the identified parameters are investigated using computer simulations for the case of tracer tests in a fracture plane. Longitudinal dispersivity tends to be overestimated by using the one-dimensional model and to become larger when the pumping rate becomes smaller, the dipole ratio larger, the heterogeneity of the fracture stronger and orthogonal oriented background groundwater flow greater. Such information will be of much help for planning tracer tests in an appropriate manner at underground research laboratories both at Mizunami in central Japan and Horonobe in northern Japan. Definition of appropriate experimental conditions will contribute to decreasing the uncertainty in the results of the tracer tests.

4) 40056 – Development of Comprehensive Techniques for Coastal Site Characterisation (1) Strategic Overview

Kunio Ota, Kenji Amano, Tadafumi Niizato, JAEA (Japan); W Russell Alexander, Bedrock Geosciences (Switzerland);

Yoshiaki Yamanaka, Suncoh Consultants (Japan)

The assurance of the long-term stability of the geological environment is sine qua non for deep geological disposal. Any assessment of repository safety will thus require development of a set of analyses and arguments to demonstrate the persistence of the key safety functions of the geological environment up to several hundred thousand years into the future, taking into account the likely future evolution of the repository host rock.

Concern presently focusses on a sea-level rise caused by anthropogenic global warming but, within a period of several tens of thousands of years, a return to glacial-period conditions is to be expected. Based on previous glaciations, global sea-level could drop by up to 150 m and, in the case of Japan, such a decrease in sea-level would result in an extremely dramatic change in the location of coastal lines with a subsequent significant change to hydraulic and

-86-

Abstracts

hydrochemical conditions at coastal sites. It is thus of great importance in the Japanese disposal programme to establish comprehensive techniques for characterising the overall evolution of coastal sites over geological time with focus very much on the persistence of the key safety functions throughout episodes of uplift/subsidence and climatic and sea-level changes.

To this end, based on practical experience from the ongoing underground research laboratory projects of JAEA, a transparent and traceable roadmap for planning and implementing a sequence of field investigations at any coastal site has been formulated. Known as a “Geosynthesis Data Flow Diagram (GDFD)”, this system illustrates linkages between a range of parameters, including investigations of key aspects to be addressed, interpretation of data acquired, synthesis of the results of different studies and analyses and final clarification of the key properties and processes of the geological environment. In particular, the GDFD defines a geosynthesis methodology for describing temporal and spatial changes of various characteristics and processes, with particular focus on the site palaeohydrogeology.

Such a geosynthesis methodology has been introduced in an ongoing collaborative programme for characterising the coastal geological environment around Horonobe in northern Hokkaido, Japan. A basic strategy for stepwise investigations has been proposed, which incorporates the geosynthesis methodology in an effective manner in each step from initial survey/review of existing information, through aerial, terrestrial and marine exploration, to the final borehole programme. This technique has now been tested and optimised based on technical findings and experience that have been accumulated with the progress of the investigations.

5) 40052 – Development of comprehensive techniques for coastal site characterisation: (3) Conceptualisation of long-term geosphere evolution

Tadafumi Niizato, Kenji Amano, Kunio Ota, Takanori Kunimaru, JAEA (Japan); Lanyon Bill, Nagra (Switzerland);

W Russell Alexander, Bedrock Geosciences (Switzerland)

A critical issue for building confidence in the long-term safety of geological disposal is to demonstrate the stability of the geosphere, taking into account its likely future evolution. This stability is broadly defined as the persistence of Thermal-Hydrological-Mechanical-Chemical conditions considered favourable for the long-term safety of a geological repository. The geosphere is slowly but constantly evolving, and then the stability, in this case, does not imply that steady-state conditions exist. What is important is that the evolution of the geosphere can be understood. In general, an understanding of the evolution is gained by studying the palaeohydrogeological evolution of a site, defining temporal and spatial changes of various characteristics, events, and processes over geological time. The site palaeohydrogeology refers to natural events and processes that have occurred in the past and contributed to the present state of the geosphere, which include sub-surface processes (e.g. crustal movement, diagenesis, etc.) and earth-surface processes (e.g. climatic and sea-level changes, geomorphological processes, etc.). An understanding of the palaeohydrogeological evolution of the site provides the firm foundation to describe the likely future evolution of the site. An ongoing collaborative programme aims to establish comprehensive techniques for characterising the overall evolution of coastal sites through studying the palaeohydrogeological evolution in the coastal system around the Horonobe area, Hokkaido, northern Japan. In this study, the current status for the conceptualisation of the long-term geosphere evolution in the coastal area, is based on data from the JAEA’s underground research laboratory project. Information on surface and sub-surface processes has been integrated into a chronological conceptual model which indicates space-time sequences of the events and processes in the area over geological time. Spatial scale for the conceptualisation is ca. 100 km in the East-West direction through the locations of the underground research laboratory and of the borehole investigations on the coast in the Horonobe area. Temporal conceptualisations over the last few million years are focused on the spatial and temporal changes of the geosphere caused by sub-surface processes and over the last several hundred thousand years focused on the changes caused by earth-surface processes. The methodology for conceptualisation of the geosphere evolution will be applied to other analogous coastal areas on Japan’s western seaboard to produce comprehensive techniques to support understanding the geosphere evolution of potential coastal sites for deep geological repositories.

6) 40048 – Development of Comprehensive Techniques for Coastal Site Characterisation (2) Integrated Palaeohydrogeological Approach for Development of Site Evolution Models

Kenji Amano, Tadafumi Niizato, Hideharu Yokota, Kunio Ota, JAEA (Japan); Bill Lanyon, Nagra (Switzerland);

W Russell Alexander, Bedrock Geosciences (Switzerland)

Radioactive waste repository designs consist of multiple safety barriers which include the waste form, the canister, the engineered barriers and the geosphere. In many waste programmes, it is considered that the three most important safety features provided by the geosphere are mechanical stability, favourable geochemical conditions and low groundwater flux. To guarantee that a repository site will provide such conditions for timescales of relevance to the safety assessment, any repository site characterisation has to not only define whether these features will function appropriately today, but also to assess if they will remain adequate up to several thousand to hundreds of thousand years into the future, depending on the repository type.

In general, this is done by studying the palaeoKHNPgeological evolution of a site, defining temporal and spatial changes of various characteristics and processes. These may include hydrogeology, geology, groundwater flow

-87-

Abstracts

characteristics, groundwater chemistry and site tectonics, including uplift and erosion processes. These key aspects are studied to build up a conceptual model for the overall site evolution over geological time, up to the present and this is used to define the likely future evolution of the site and to assess if the main safety features will continue to function adequately.

The collaborative programme described here is focussed on the palaeohydrogeology of the coastal area around Horonobe in northern Hokkaido, Japan. Data from JAEA’s ongoing underground research laboratory project is being synthesised in a Site Descriptive Model (SDM) with new information from the collaborating research institutes to develop a Site Evolution Model (SEM), with the focus very much on changes in the Sea of Japan seaboard over the last few million years. This new conceptual model will then be used to assess the palaeohydrological evolution of other analogous sites on Japan’s western seaboard, with the final aim of producing a set of comprehensive techniques to understand the palaeohydrogeological evolution of the deep geosphere of all coastal sites on the Sea of Japan.

7) 40041 – Development of Methodology of Groundwater Flow and Solute Transport Analysis in the Horonobe Area, Hokkaido, Japan

Keisuke Maekawa, Hitoshi Makino, Hiroshi Kurikami, Tadafumi Niizato, Manabu Inagaki, Makoto Kawamura, JAEA (Japan)

It is important for establishment of safety assessment techniques of geological disposal to understand groundwater flow and solute transport accurately. Therefore, we are positioning to confirm an applicability of the techniques in realistic environment as a crucial issue in R&D. We have attempted and planed some relevant studies as below: -A methodology to integrate activities from site investigations to evaluation of solute transport was examined. We have carried out groundwater flow analysis on a regional scale using geological and hydrological information from surface-based investigations at the Horonobe area, and also solute transport analysis based on the information of the trajectory analysis. -We have carried out a preliminary simulation of groundwater flow and salinity concentration distribution using information on climatic and sea-level changes, and evolution of geological structures considering the impacts of natural events and processes. Consequently, we could outline the impacts of natural events and processes on geological environment including hydrogeology, hydrochemistry and their evolutions. -We have been planning to develop and apply a methodology of groundwater flow and solute transport analysis to the shallow part, the Horonobe coastal area and around the URL. These techniques would become a basis for future site specific safety assessment in Japan.

SESSION D6: Waste Treatment and Non-Reactor

1) 40105 – Estimation of Radioactivity of Graphite Blocks in Tokai Power Station using Statistical Method

Masaaki Nakano, Fuji Electric Holdings Co., Ltd. (Japan); Hisashi Mikami, Fuji Electric Systems Co., Ltd. (Japan);

Hideaki Ichige, Shinich Tsukada, JAPC (Japan)

Tokai Power Station (graphite moderated, gas-cooled reactor, GCR) stopped its commercial operation in March 1998 and is decommissioning now. Since graphite blocks in Tokai reactor core are major low level wastes (LLWs), the realistic and reasonable method to estimate radioactivity of graphite blocks is required for final disposal and its licensing procedure. In general, LLWs, which were installed in or around a reactor core, have large radioactivity, theoretical calculations can be applied to the estimation of the radioactivity. This paper describes the concept of the method using statistical approach to determine the radioactivity of the graphite blocks in the reactor core. This method directly considers the variations of input calculation conditions, for example, compositions of impurity elements, irradiation neutron flux and irradiation period. In this paper, the variations of the compositions of impurity elements were statistically considered with the mean value and the standard deviation that were determined with elemental analyses. Many activation calculations were performed with the compositions that were determined with pseudorandom numbers, the mean value and the standard deviation. The calculated radioactivities distribute also statistically and a mean value and a standard deviation of radioactivity can be determined. The distribution of calculated radioactivities shows consistency to radiochemical analyses of graphite blocks from the reactor core and this shows that the method is applicable to the estimation of the graphite block radioactivity. Furthermore, this method can be considered to reduce over-excess estimation margin and can obtain reasonable radioactivity rather than using maximum or conservative values of all input conditions. This method is now being developed and approved as one of basic procedure for determining the radioactivity of wastes by Standards Committee of the Atomic Energy Society of Japan.

2) 40115 – The treatment of hexavalent chromium in waste liquid from Fugen Decommissioning

Yuki Yahiro, Seiji Yamamoto, Koji Negishi, Hitoshi Sakai, Tadashi Fukushima, Norimasa Yoshida, Toshiba Corporation (Japan); Nobuo Ishizuka, Yuji Sato, Wataru Fujiwara, JAEA (Japan)

-88-

Abstracts

The prototype Advanced Thermal Reactor (ATR) Fugen Nuclear Power Station of JAEA terminated its operation in 2003 and is under decommissioning. In cooling water (90m3) for iron-water shield concrete surrounding the core region, hexavalent chromium (K2CrO4) was dissolved as corrosion inhibitor. The hexavalent chromium has to be reduced to trivalent chromium, for the safety. Authors had developed a new treatment technique for the waste liquid containing hexavalent chromium with small amount of secondary waste. We manufactured and installed test equipment, and we treated the waste liquid in Fugen. We added formic acid as pH adjuster in the water and hydrogen peroxide to reduce hexavalent chromium to trivalent chromium. The residual hydrogen peroxide was decomposed to oxygen and water with iron catalysis and UV lamp. Then cationic species, such as trivalent chromium and potassium, in waste liquid were removed by cation exchange resin. The hydrogen peroxide and iron catalysis were added again to decompose the residual formic acid to carbon dioxide gas and water. Next, the residual hydrogen peroxide was decomposed by catalase. Finally, all the ionic species in the waste liquid were removed by mixed resin. It was confirmed that the concentration of hexavalent chromium could be reduced to below 0.5ppm and total chromium concentration to below 2.0ppm. This value is below the water pollution prevention law. It took 5 days for the whole treatment of 5m3 waste liquid containing the hexavalent chromium 340ppm.

3) 40201 – Characterization of Radioactive Waste from Side Structural Components of a CANDU Reactor for Decommissioning Applications in Korea

Rizwan Ahmed, Gyunyoung Heo, Kyung Hee University (Korea Rep.); Dong-Keun Cho, Jongwon Choi, KAERI (Korea Rep.)

Reactor core components and structural materials of nuclear power plants to be decommissioned have been irradiated by neutrons of various intensities and spectrum. This long term irradiation results in the production of large number of radioactive isotopes that serve as a source of radioactivity for thousands of years for future. Decommissioning of a nuclear reactor is a costly program comprising of dismantling, demolishing of structures and waste classification for disposal applications. The estimate of radio-nuclides and radiation levels forms the essential part of the whole decommissioning program. It can help establishing guidelines for the waste classification, dismantling and demolishing activities. ORIGEN2 code has long been in use for computing radionuclide concentrations in reactor cores and near core materials for various burn-up-decay cycles, using one-group collapsed cross sections. Since ORIGEN2 assumes a constant flux and nuclide capture cross-sections in all regions of the core, uncertainty in its results could increase as region of interest goes away from the core. This uncertainty can be removed by using a Monte Carlo Code, like MCNP, for the correct calculations of flux and capture cross-sections inside the reactor core and in far core regions. MCNP has greater capability to model the reactor problems in much realistic way that is to incorporate geometrical, compositional and spectrum information. In this paper the classification of radioactive waste from the side structural components of a CANDU reactor is presented. MCNP model of full core was established because of asymmetric structure of the reactor. Side structural components of total length 240 cm and radius 16.122 cm were modeled as twelve (12) homogenized cells of 20 cm length each along the axial direction. The neutron flux and one-group collapsed cross-sections were calculated by MCNP simulation for each cell, and then those results were applied to ORIGEN2 simulation to estimate nuclide inventory in the wastes. After retrieving the radiation level of side structural components of in- and ex-core, the radioactive wastes were classified according to the international standards of waste classification. The wastes from first and second cell of the side structural components were found to exhibit characteristics of class C and Class B wastes respectively. However, the rest of the waste was found to have activity levels as that of Class A radio-active waste. The waste is therefore suitable for land disposal in accordance with the international standards of waste classification and disposal.

4) 40068 – Uranium refining and conversion plant decommissioning project

Naoki Zaima, Yasuyuki Morimoto, Noritake Sugitsue, Kazumi Kado, JAEA (Japan)

The uranium refining and conversion plant at Ningyo-toge (URCP) was constructed in 1981 for the purpose of demonstration on refining and conversion processes from yellow cakes to UF6 via UF4, and then as modified to develop the conversion of reprocessed uranium production of natural UF6 and purification of reprocessed UF6. Through 20 years, 385 tons of natural uranium of UF6 and 336 tons of reprocessed uranium of UF6 had been conducted. There are two refining processes in the URCP facilities. One is the wet-type process for the natural uranium and the other is the dry-type process for the reprocessed uranium. It was found the large amount of uranium residuals such as wet slurry and dried powder inside the vessels and pipings. Careful consideration had always been required against the diffusion of contamination. The basic policies concerning plant decommissioning are the optimization of the labor costs and the minimization of the radioactive wastes. The procedures are followings; i)measuring doserate by high sensitivity surveymeters and identificating nuclide by gamma ray spectrometry, ii)estimating uranium mass inventory, iii) planning workers distributions including of radiation control staffs, iv)deciding dismantling methods and decontaminating schematically if required, v)measuring and classifying doserate and contamination level, vi) managing for radioactive waste container, vii)control for personal exposures. Through two years and half, almost all equipment had been dismantled except building decontamination. Several hundreds tons of dismantled wastes had accumulated in 200 litter drums approximately. In addition, the secondary wastes had also been generated. Several thousands day of working time had spent totally. The radiation monitoring of working places had been performed during dismantling, the results were generally less than 20?Sv/h under the doserate limitation. However, followed by the trace of the reprocessed uranium,

-89-

Abstracts

U-232 progenies nuclides such as Th-228 and Tl-208 were observed. The expected exposures are apprehensive by accumulation of the high energy gamma emission nuclides. For example, the fluidization media storage tank in which Th-228 progenies originated from U-232 accumulated. There arises the case of remote controlled tools for decreasing personal exposure if required. As for the waste disposal, the determination of uranium content must be necessary. We are now developing for measuring systems with better accuracy. The further tasks imposed us summarized the followings; i)dismantling method for high doserate area, ii)reduction of radioactive wastes volume, iii)decontamination for the buildings, iv)waste disposal.

SESSION M2: Public Involvement

1) 40288 – Activities of the OECD/NEA in the Field of Stakeholder Confidence for Radwaste Management and Decommissioning

Claudio Pescatore, OECD/NEA

The OECD/NEA seeks to assist its member countries in developing strategies for the management of all types of radioactive material, including waste, that are safe and sustainable and that meet the broad needs of society - with particular emphasis on the management of long-lived waste and spent fuel and on decommissioning of disused nuclear facilities. The programme of work in the area of stakeholder involvement is supervised by the Radioactive Waste Management Committee (RWMC) and carried out by its Forum on Stakeholder Confidence (FSC). The latter is made of senior representatives from regulatory authorities, decommissioning organisations, policy making bodies, and research-anddevelopment institutions from the NEA countries. First and foremost, the FSC is a “learning organisation”. Through the FSC, members seek to improve themselves as responsive actors in the governance of radioactive waste management and decommissioning. Delegates attend in order to benefit from in-depth pragmatic exchanges with both peers and stakeholders beyond the membership. They then consolidate their learning: the FSC takes as a responsibility to mature its lessons in discussion and cooperation with those concerned, and then to validate its conclusions with the help of academic researchers. The third step in the cycle is to make the learning available to others. The FSC has held workshops in six countries and local communities therein according to its own well tested approach. In each place, the host defines the principal themes for discussion in radioactive waste management. FSC members learn from national presenters about the history and context of each case study and hear a broad range of stakeholder voices describe their position, actions and concerns. Sitting together in small groups for roundtable discussions, FSC members can ask questions of hosting stakeholders, understand better their point of view, and share experience from their own institutions and countries. Each table then gives feedback to the entire audience, and the main observations are published alongside the texts of stakeholder presentations. The FSC then elaborates further on the lessons to be learnt. Over the 10 years of its existence, the FSC has established itself as a key international player and adviser in the field of stakeholder confidence. Its numerous studies and its advice are available on-line in the form, respectively, of academic reports and policy brochures and flyers. Explored topics include: stepwise decision-making, principles for stakeholders involvement, the partnering approach for siting and developing fuel cycle facilities, lessons learned in decommissioning, providing value-added in view of building a durable relationship between a facility and its host community, etc. Current study areas include: the interests and roles of regional authorities, how to increase the knowledge base of journalists, and providing added value (beyond economic benefits) to communities hosting waste management facilities. The activities and lessons to be learnt will be reviewed for the ICEM-2010 audience.

2) 40219 – A Comparative Study of Stakeholder Participation in the Cleanup of Radioactive Wastes in the US, Japan and UK

Mito Akiyoshi, Senshu University (Japan); William Lawless, Fjorentina Angjellari-Dajci, Paine College (USA); Christian Poppeliers, Augusta State University (USA); John Whitton, National Nuclear Laboratory (UK)

We review case studies of stakeholder participation in the environmental cleanup of radioactive wastes in the USA, Japan and UK (e.g., [21,26,27,66,78]). Citizen participation programs in these three countries are at different stages: mature in the US, starting in Japan, and becoming operational in the UK. The US issue at the US Department of Energy’s (DOE) Savannah River Site (SRS) in South Carolina (SC) had been focused on citizens encouraging Federal (DOE; US Environmental Protection Agency, or EPA; and the US Nuclear Regulatory Commission, or NRC) and State (SC's Department of Health and Environmental Compliance, or DHEC) agencies to pursue "Plug-in-RODs" at SRS to simplify the regulations to accelerate closing seepage basins at SRS. In Japan, the Reprocessing of spent fuel and deep geological disposal of vitrified high-level waste have been among Japan's priorities. A reprocessing plant in Rokkasho, Aomori Prefecture is expected to commence operations in October 2010. The search of a site for a deep geological disposal facility has been ongoing since 2002. But the direct engagement of stakeholders has not occurred in Japan. Indirectly, stakeholders attempt to exert influence on decision-making with social movements, local elections, and litigation. In the UK, the issue is gaining effective citizen participation with the UK's Nuclear Decommissioning Authority (NDA). We hope that the case studies from these countries may improve citizen participation.

-90-

Abstracts

3) 40076 – Territorial Integration of the Geological Repository in France

Gérald Ouzounian, Sebastien Farin, Roberto Miguez, Jean-Louis Tison, Andra (France)

In France, a framework has been drawn up by the National Assembly and implemented by the government, in order to get the best relationship between Andra, among others, and the stakeholders and the inhabitants of the towns and countries where disposal facilities or projects are or could be established. The main threads of the two Acts passed* in 2006, being relevant to the relationship with inhabitants are the information exchange and the local economic development. Dealing with the information exchange and diffusion: - The Local Information Committee (CLI), for each nuclear facility, has been reinforced and a specific, Local Information and Oversight Committee (for the Underground Laboratory in Meuse-Haute-Marne) has been renewed. The CLI was in charge of a general assignment to inform and consult on nuclear safety, radioprotection and environmental topics. Now, since 2006, the nuclear facility’s CLI and the CLIS are able to order study reports, measures and analyses to experts freely selected. - Creation of the High Committee for Transparency and Information on Nuclear Safety (HCTISN). This new authority aims to inform, consult and debate about the risks relevant to nuclear activities and their impacts on people’s health, environment and nuclear safety. Andra contributes to the functioning of CLI in disposal facilities at Manche and Aube Departments, and CLIS of the underground Laboratory at Meuse and Haute-Marne departments. This paper will present these contributions and how Andra’s action helps to reach the goals of information and exchange with the people around its facilities. Concerning the local economic development, there are specific organizations or schemes, depending on the facility.: - Local taxes contributions based on the disposal facilities activities as is usual in France. - A High Level Committee (CHN) and two public interest groups (GIP) in Meuse and Haute-Marne departments have been set up since 1991 and 2005. Andra is represented in these three institutions, but they are not funded at all by Andra. This paper will show the Andra’s involvement in the local economic and territorial developments. Within this general framework Andra has developed information and exchanges actions with the stakeholders and the inhabitants around its facilities. Examples of these actions will be presented also. (*) -Planning Act No. 2006-739 of 28 June 2006 concerning the sustainable Management of Radioactive materials an waste. - Act No. 2006-686 of 13 June 2006 on Transparency and Security in the Nuclear Field.

SESSION L8: L/ILW Poster

1) 40006 – A GoldSim Modeling Approach to Safety Assessment of an LLW Repository System

Youn Myoung Lee, Jongtae Jeong, Jongwon Choi, KAERI (Korea Rep.)

An assessment program for the safety assessment and performance evaluation of a low- and intermediate level waste (LILW) repository system has been developed by utilizing GoldSim, by which nuclide transports in the near- and far-field of a repository as well as a transport through a biosphere under various natural and manmade disruptive events affecting a nuclide release could be modeled and evaluated. To demonstrate its usability, some illustrative cases under the selected scenarios including the influence of degradation of man-made barriers and the possible disruptive events caused by an accidental human intrusion or an earthquake have been investigated and illustrated for a hypothetical LILW repository. Even though all the parameter values applied to a hypothetical repository are assumed without any real base, the illustrative cases are very informative especially when seeing the result of the probabilistic calculation with scenarios which are possibly happen for nuclide release and further transport in and around the repository system.

2) 40011 – Gas Migration Mechanism of Saturated Highly-compacted Bentonite and its Modeling

Yukihisa Tanaka, Michihiko Hironaga, Koji Kudo, CRIEPI (Japan)

In the current concept of repository for radioactive waste disposal, compacted bentonite will be used as an engineered barrier mainly for inhibiting migration of radioactive nuclides. Hydrogen gas can be generated inside the engineered barrier by anaerobic corrosion of metals used for containers, etc. If the gas generation rate exceeds the diffusion rate of gas molecules inside of the engineered barrier, gas will accumulate in the void space inside of the engineered barrier until its pressure becomes large enough for it to enter the bentonite as a discrete gaseous phase. It is expected to be not easy for gas to entering into the bentonite as a discrete gaseous phase because the pore of compacted bentonite is so minute. Therefore the gas migration tests are conducted in this study to investigate the mechanism of gas migration. On the basis of the experimental facts obtained through the gas migration tests, possible gas migration mechanism is proposed. A simplified method for calculating gas pressure at large breakthrough, which is defined as a sudden and sharp increase in gas flow rate out of the specimen is also proposed.

3) 40012 – Development of numerical simulation method for gas migration through highly-compacted bentonite using model of two-phase flow through deformable porous media

Yukihisa Tanaka, CRIEPI (Japan)

-91-

Abstracts

In the current concept of repository for radioactive waste disposal, compacted bentonite will be used as an engineered barrier mainly for inhibiting migration of radioactive nuclides. Hydrogen gas can be generated inside of the engineered barrier by anaerobic corrosion of metals used for containers, etc. It is expected to be not easy for gas to entering into the bentonite as a discrete gaseous phase because the pore of compacted bentonite is so minute. Therefore it is necessary to investigate the effect of gas pressure generation and gas migration on the engineered barrier, peripheral facilities and ground. In this study, a method for simulating gas migration through the compacted bentonite is proposed. The proposed method can analyze coupled hydrological-mechanical processes using the model of two-phase flow through deformable porous media. Validity of the proposed analytical method is examined by comparing gas migration test results with the calculated results, which revealed that the proposed method can simulate gas migration behavior through compacted bentonite with accuracy.

4) 40020 – Planning of Large-scale In-situ Gas Generation Experiment in Korean Radioactive Waste Repository

Juyoul Kim, Sukhoon Kim, FNC Technology Co.(Korea Rep.); Jinbeak Park, Sungjoung Lee, KRMC (Korea Rep.)

In the Korean LILW (Low- and Intermediate-Level radioactive Waste) repository at the Gyeongju site, the degradation of organic wastes and the corrosion of metallic wastes and steel containers would be important processes that a?ect repository geochemistry, speciation and transport of radionuclides during the lifetime of a radioactive waste disposal facility.

Gas is generated in association with these processes and has the potential to pressurize the repository, which can promote the transport of groundwater and gas, and consequently radionuclide transport. Microbial activity plays an important role in organic degradation, corrosion and gas generation through the mediation of reduction–oxidation reactions.

The Korean research project on gas generation will be performed by Korea Radioactive Waste Management Corporation (KRMC). A large-scale in-situ experiment will form a central part of the project, where gas generation in real radioactive low-level maintenance waste from nuclear power plants will be studied.

In order to examine gas generation from an LILW repository which is being constructed at present by 2012, two large-scale facilities for the gas generation experiment (GGE) will be established, each equipped with a concrete container for 16 or 9 drums of LILW from Korean nuclear power plants. Each container will be enclosed within a gas-tight and acid-proof steel tank. The experiment facility will be filled with ground water that provides representative chemical conditions and microbial inoculation in the repository near field.

In the experiment, the design includes long-term monitoring for the rate and composition of gas generated, and aqueous geochemistry and microbe populations present at various locations through on-line analyzers and manual sampling.

A main schedule for establishing the experiment facility is as follows: Completion of the detailed design until the second quarter of the year 2010; Completion of the manufacture and on-site installation until the second quarter of the year 2011; Start of the operation and monitoring from the third quarter of the year 2011.

5) 40024 – Estimation and measurement of porosity change in cement paste

Eunyong Lee, Haeryojng Jung, Ki-jung Kwon, KRMC (Korea Rep.); Do-Gyeum Kim, Korea Institute of Construction Technology (Korea Rep.)

Laboratory-scale experiments were performed to understand the porosity change of cement pastes. The cement pastes were prepared using commercially available Type-I ordinary Portland cement (OPC). As the cement pastes were exposed in water, the porosity of the cement pastes sharply increased; however, the slow decrease of porosity was observed as the dissolution period was extended more than 50 days. As expected, the dissolution reaction was significantly influenced by w/c raito and the ionic strength of solution. A thermodynamic model was applied to simulate the porosity change of the cement pastes. It was highly influenced by the depth of the cement pastes. There was porosity increase on the surface of the cement pastes due to dissolution of hydration products, such as portlandite, ettringite, and CSH. However, the decrease of porosity was estimated inside the cement pastes due to the precipitation of cement minerals.

6) 40082 – Separation and Recovery of Sodium Nitrate from Low-level Radioactive Liquid Waste by Electrodialysis

Yoshihiro Meguro, A Kato, Y Watanabe, Kuniaki Takahashi, JAEA (Japan)

Low-level radioactive liquid wastes including TRU elements have been generated from several treatments in reprocessing facilities. These wastes usually include highly concentrated sodium nitrate. In Japan the liquid wastes are planned to dispose in geological layer with HLW following solidification of them and in this plan it is considered that leached nitrate from the solidified materials and its decomposition products would change sorption coefficients of radionuclides and then their migration in the long-term would be influenced. Therefore a method to remove the nitrate ions prior to the solidification is required. The authors have been developed a separation and recovery method of sodium nitrate from the low-level radioactive liquid waste by using an electrodialysis method not only to remove the nitrate but

-92-

Abstracts

also to reduce waste volume and to reuse resources. Here, nitrate ion is recovered as nitric acid and sodium ion as sodium hydroxide through ion-exchange membranes. Especially, a ceramic cation-exchange membrane, in which sodium ion was selectively transported, was employed to prevent contamination of radioactive nuclides in the recovered sodium hydroxide. And also an anion-exchange membrane, in which monovalent anions were selectively transported anions, was employed to improve the nitric acid purity. In the present paper, we showed our recent results of investigation for transport efficiency of the target ions through the membranes and their selectivity. Aqueous solutions of 5 mol/L sodium nitrate containing a small amount of potassium, cesium, and strontium ions or containing nitrite, carbonate, sulfate, and phosphate ions were prepared as simulated waste solutions. As a sodium-selective cation-exchange membrane, NaSICON membrane produced by Ceramatec Inc. was employed and NEOSEPTA ACS membrane produced by ASTOM Co. as a mono-anion permselective membrane. Sodium ions were transferred through the NaSICON membrane as much as the quantity of electricity used in the electrodialysis, but only hundredth to 10thousandth amount of potassium, cesium, and strontium were transferred through the membrane. The sodium ions could be effectively and selectively removed from the highly concentrated solution using the NaSICON membrane. Through the ACS membrane monovalent anions were efficiently transferred, but less than 10% of multivalent anions were moved at which 90% nitrate was. It was confirmed from these results that the electrodialysis method with the ion-selective membranes was useful procedure to remove and recover sodium nitrate from the liquid waste. This work was funded by ANRE: Agency for Natural Resources and Energy, of METI: Ministry of Economy, Trade and Industry, of Japan.

7) 40109 – Study on Mechanical Influence of Gas Generation and Migration on Engineered Barrier System in adioactive Waste Disposal Facility

Mamoru Kumagai, JNFL (Japan); Shuichi Yamamoto, Kunifumi Takeuchi, Obayashi Corporation (Japan); Yukihisa Tanaka, Michihiko Hironaga, CRIEPI (Japan)

In Japan, some radioactive waste with a relatively higher radioactivity concentration from nuclear facilities is to be packaged in rectangle steel containers and disposed of in sub-surface disposal facilities, where normal human intrusion rarely occurs. After the closure of a facility, its pore is saturated with groundwater. If the dissolved oxygen of the pore water is consumed by steel corrosion, hydrogen gas will be generated from the metallic waste, steel containers, and reinforcing bars of concrete mainly by anaerobic corrosion. If the generated gas accumulates and the gas pressure increases excessively in the facility, the facility’s barrier performance might be degraded by mechanical influences such as crack formation in cementitious material or deformation of bentonite material.

Firstly, in this study, we assessed the time evolution of the gas pressure and the water saturation in a sub-surface disposal facility by using a multi-phase flow numerical analysis code, GETFLOWS, in which a pathway dilation model is introduced and modified in order to reproduce the gas migration mechanism through the highly compacted bentonite.

Next, we calculated the stress applied to the engineered barriers of the facility from the results of the time evolution of the pressure and the saturation. Then, we conducted a mechanical stability analysis of the engineered barriers by using a nonlinear finite element code, ABAQUS, in order to evaluate their performances after the closure of the facility.

8) 40221 – Decontamination of Radioactive Concrete Wastes by Thermal and Mechanical Processes

B.Y. Min, Ki Won Lee, Un Soo Chung, W.K Choi, Kune-Woo Lee, KAERI (Korea Rep.)

The purpose of this paper is to provide the results of a series of volume reduction tests consucted in thermal and mechanical processes. Korea Atomic Energy Research Institites (KAERI) has developed volume reduction technology applocable to an activated heavy concrete waste generated by dismantled Korea research reactor 2 (KRR-2) and a uranium conversion plant (UCP). the volume reductuin could be achieved above 70% by lab scale test by thermal and mechanical separation processes. Pilot test were performed with radioactively contaninated dismantled concrete waste. The developed processes were quite effective for volume reduction of radioactively contaminated dismantled concrete waste.

9) 40302 – Latex Particles Functionalized with Transition Metals Ferrocyanides for Cesium Uptake and Decontamination of Solid Bulk Materials

Dmitry Marinin, Valentin Avramenko, Dmitry Marinin, Valentin Sergienko, Institute of Chemistry FEDRAS (Russia);

Veniamin Zheleznov, Irina Sheveleva, Russian Academy of Sciences (Russia)

Decontamination of spent ion-exchange resins, corrosion-unstable metal structures, soil, ground, and construction materials contaminated by fission, corrosion and transuranic radionuclides remains one of the most urgent and complicated ecological problems. Among the existing methods having different efficiencies in regard to such materials decontamination, application of selective sorbents put into a humid medium to be decontaminated (ground, bulk materials) appears to be rather extensive. However, the efficiency of such an approach is significantly limited by difficulties concerned with uniform sorbent distribution in porous media and completeness of spent sorbents removal for final disposal. In this paper we suggest a principally new approach to preparation of colloid-stable selective sorbents for cesium uptake using immobilization of transition metals (cobalt, nickel, and copper) ferrocyanides in nanosized carboxylic latex emulsions. The effects of ferrocyanide composition, pH, and media salinity on the sorption properties of

-93-

Abstracts

the colloid-stable sorbents toward cesium ions were studied in solutions containing up to 200 g/l sodium nitrate or potassium chloride. The sorption capacities of the colloid sorbents based on mixed potassium/transition metals ferrocyanides were in the range 1.45-1.86 mol Cs/mol ferrocyanide with the highest value found for the copper ferrocyanide. It was shown that the obtained colloid-stable sorbents were capable to penetrate through bulk materials without filtration that makes them applicable for decontamination of solids, e.g. soils, zeolites, spent ion-exchange resins contaminated with cesium radionuclides. After decontamination of liquid or solid radioactive wastes the colloid-stable sorbents can be easily separated from solutions by precipitation with cationic flocculants providing localization of radionuclides in a small volume of the precipitates formed. Besides, functionalized latex particles can be used for preparation of carbon fiber/ferrocyanide composite materials for cesium uptake using electrodeposition method. Application of the carbon fibers as an inert support for ferrocyanides, in general, significantly improves the sorption kinetics, but washing out of ferrocyanide fines from the fiber surface limits the potential of such materials. When ferrocyanides are deposited in a form of nanocrystals stabilized by latexes which undergo electropolymerization on the fiber surface, the thin polymeric film formed substantially improves the stability of the composite and prevents loss of ferrocyanide during sorbent application. The effect of electrodeposition conditions on composite morphology, ferrocyanide loading and cesium distribution coefficient in media with different salinity has been discussed.

SESSION H11: SF/TRU/HLW Poster

1) 40001 – Investigation of Colloid-facilitated Effects on the Radionuclides Migration in the Fractured Rock with a Kinetic Solubility-Limited Dissolution Model

Chun-Ping Jen,Nation Chung Cheng University (Taiwan); Neng-Chuan Tien, Energy and Environment Research Laboratories,

Industrial Technology Research Institute (Taiwan)

Nuclides can move with the groundwater either as solutes or colloids, where the latter mechanism generally results in much shorter traveling time as they interact strongly with solid phases, such as actinides. It is therefore essential to assess the relative importance of these two transport mechanisms for different nuclides. The relative importance of colloids depends on the nature and concentration of colloids in groundwater. Plutonium (Pu), neptunium (Np), uranium (U) and americium (Am) are four nuclides of concern for long-term emplacement of nuclear wastes at potential repository sites. If attached to iron oxide, clay or silica colloids in groundwater. Strong sorption of the actinides by colloids in groundwater may facilitate transport of these nuclides along potential flow paths. Solubility-limited dissolution model models can be used to determine the release of the safety assessment for nuclear waste in geological disposal sites. The present study investigates the effect of colloid on the transport of solubility limited nuclide under the kinetic solubility limited dissolution (KSLD) boundary condition in fractured media. The release rate of nuclide would proportional to the difference between the saturation concentration and the inlet aqueous concentration of nuclide. The presence of colloids could decrease the aqueous concentration of nuclide and thus could increase the release flux of nuclide from the waste form.

2) 40013 – Modeling hydraulic conductivity and swelling pressure of several kinds of bentonites affected by salinity of water

Yukihisa Tanaka, Takuma Hasegawa, Kunihiko Nakamura, CRIEPI (Japan)

In case of construction of repository for radioactive waste near the coastal area, the effect of salinity of water on hydraulic conductivity as well as swelling pressure of bentonite as an engineered barrier should be considered because it is known that the hydraulic conductivity of bentonite increases and swelling pressure decreases with increasing in salinity of water. Though the effect of salinity of water on hydraulic conductivity and swelling pressure of bentonite has been investigated experimentally, it is necessary to elucidate and to model the mechanism of the phenomenon because various kinds of bentonites may possibly be placed in various salinity of ground water. Thus, in this study, a model for evaluating hydraulic conductivity as well as swelling pressure of compacted bentonite is proposed considering the effect of salinity of water as follows :

a) Change in number of flakes of a stack of montmorillonite because of cohesion b) Change in viscosity of water in interlayer between flakes of montmorillonite.

Quantitative evaluation method for hydraulic conductivity and swelling characteristics of several kinds of bentonite under saline water is proposed based on the model mentioned above.

3) 40018 – Current R & D activities of the study on long-term stability of geological environments

Takahiro Hanamuro, Kenichi Yasue, Yoko Saito-Kokubu,Koichi Asamori, Tsuneari Ishimaru, Koji Umeda, JAEA (Japan)

Japanese islands are located in a tectonically active zone, where earthquakes and volcanic eruptions frequently

-94-

Abstracts

occur. Therefore the understanding of the long-term stability of geological environment is required for assessing the long-term behaviour of the geological disposal system of high level radioactive waste (HLW) in Japan. The Japan Atomic Energy Agency (JAEA) is promoting the establishment of investigation and assessment methods of the long-term stability of geological environment necessary for site selection and safety assessment of HLW geological disposal.

The Nuclear Waste Management Organization of Japan (NUMO) is the implementation body of HLW disposal in Japan. The preliminary investigation areas for HLW disposal site, selected by NUMO in the future, are supposed to be selected excluding the zones affected by already-known active faults identified by nationwide and site-specific literature surveys. They are also supposed to be selected with excluding the area within 15 km radius of the center of a Quaternary volcano and the zones affected by thermal and hydrothermal activities identified by site-specific literature surveys. For uplift, denudation and climatic/sea-level changes, it is necessary that the change of geological environment caused by uplift/denudation and climatic/sea-level changes is assessed for HLW geological disposal system.

For seismicity and faulting, some detection techniques for active faults with no surface expression, by using helium isotope ratio of hot spring gas or detection of hydrogen gas, and studying on the assessment of fault activities are developed. For volcanism and geothermal activity, heat source of geothermal anomaly area in the non-volcanic region are considered and some detection techniques for high-temperature fluid and magma at deep underground, by using seismic tomography, magnetotelluric method and helium isotope ratio of hot spring gas, are constructed. For uplift, denudation and climatic/sea-level changes, a methodology to understand future topographic change with time is developed. Also, for dating techniques as an essential part to proceed on these studies, C-14 dating by using AMS and (U-Th)/He dating by using QMS and ICP-MS have developed, and Be-10 dating by using AMS has been being developed.

We are planning the establishment of assessment methods for long-term stability of geological environment; assessment of activities of faults encountered by underground excavation, development of long-term estimation methods of volcanisms and hydrothermal activities, and hydrogeological analyses considering topographic change in the future.

4) 40019 – In Situ Stress Measurements in Siliceous Mudstones at Horonobe Underground Research Laboratory, Japan

Hiroyuki Sanada, Takahiro Nakamura, Yutaka Sugita, JAEA (Japan)

As part of the research and development program on the HLW geological disposal, JAEA has been implementing the Horonobe Underground Research Laboratory (URL) Project investigating sedimentary rock formations distributed in Horonobe area, Hokkaido, Japan. To optimize the design and construction of underground excavations for any HLW repository, a thorough evaluation of initial stress condition will be required. Studies of initial stress conditions in sedimentary rock aren't enough, since important underground structures such as those for underground power plants and caverns for storage of liquefied petroleum or natural gas in Japan have mainly been constructed in hard rock. Additionally, initial stress measurements in sedimentary rocks are inherently more difficult compared to hard rock. The authors have been implementing the research and development program at Horonobe to clarify the in situ stress conditions in siliceous mudstones at the URL. The objective of this work is to establish a strategy for an in situ stress measurement program for geological disposal and to develop an understanding of the in situ stress conditions in the deep underground formed by the sedimentary rocks. The application of several stress measurement methods to the Horonobe siliceous mudstones carried out during the surface-based investigations and the investigations during construction of the underground facilities, as well as information on the initial stress state around the Horonobe URL are described in this paper. During the surface-based investigations, determination of deep in situ stress was done using hydraulic fracturing (HF), borehole breakout information in deep boreholes and core-based methods such as AE and DSCA. Subsurface investigations during construction of the underground facilities utilized, the Compact Conical Ended Borehole Overcoring (CCBO) method and hydraulic testing of pre-existing fractures (HTPF) were conducted in order to validate results from initial stress measurements in the surface-based investigations. HF results indicate that horizontal maximum and minimum principal stresses increase linearly with depth. The minimum principal stress is almost equivalent to overburden pressure. The maximum principal stress estimated from the HF and borehole breakout data is almost E-W. This is similar to the tectonic movement direction in the vicinity of the Horonobe URL. Due to tectonic movement, horizontal maximum stress is almost 1.5 times larger than the horizontal minimum stress. Vertical stress determined from HTPF during construction of the underground facilities is almost equal to the overburden pressure.

5) 40038 – Low alkaline Cement Used in the Construction of a Gallery in the Horonobe Underground Research Laboratory

Masashi Nakayama, Haruo Sato, Yutaka Sugita, Seiji ITO, JAEA (Japan); Masashi Minamide, Yoshito Kitagawa, Taisei Cooperation (Japan)

In Japan, any high level radioactive waste repository is to be constructed at over 300m depth below surface. Tunnel support is used for safety during the construction and operation, and shotcrete and concrete lining are used as the tunnel support. Concrete is a composite material comprised of aggregate, cement and various additives. Low alkaline cement has been developed for the long term stability of the barrier systems whose performance could be negatively affected by highly alkaline conditions arising due to cement used in a repository. Japan Atomic Energy Agency (JAEA) has developed a low alkaline cement, named as HFSC (High fly-ash silicafume cement), containing over 60wt% of silica-fume (SF) and coal ash (FA). JAEA are presently constructing an underground research laboratory (URL) at Horonobe for research and development in the geosciences and repository engineering technology. HFSC was used

-95-

Abstracts

experimentally as the shotcrete material in construction of part of the 140m deep gallery in Horonobe URL. The objective of this experiment was to assess the performance of HFSC shotcrete in terms of mechanics, workability, durability, and so on. HFSC used in this experiment is composed of 40wt% OPC (Ordinary Portland Cement), 20wt% SF, and 40wt% FA. This composition was determined based on mechanical testing of various mixes of the above components. Because of the low OPC content, the strength of HFSC tends to be lower than that of OPC in normal concrete. The total length of tunnel using HFSC shotcrete is about 73m and about 500m3 of HFSC was used. This experimental construction confirmed the workability of HFSC shotcrete. Although several in-situ experiments using low alkaline cement as shotcrete have been performed at a small scale, this application of HFSC at the Horonobe URL is the first full scale application of low alkaline cement in the construction of a URL in the world. In the paper, we present detailed results of the in-situ construction test and the future works.

6) 40040 – Effects of Nitrate on Nuclide Solubility for Co-location Disposal of TRU Waste and HLW

Gento Kamei, Morihiro Mihara, JAEA (Japan); Toshiyuki Nakazawa, Norikazu Yamada, Mitsubishi Materials Corp. (Japan)

TRU wastes are generated by reprocessing spent fuel from nuclear power plants and by fabricating MOX fuel in Japan. Some of the TRU wastes are expected to be disposed of deep underground to isolate it from the biosphere in the long-term. To optimize the disposal of TRU waste, a co-location disposal with high level waste, HLW, is being considered. A part of TRU waste includes a large amount of nitrate salt, the effects of which have to be evaluated in a safety assessment of co-location disposal.

Solubility is one of the important parameters for the safety assessment of HLW disposal. Large concentrations of nitrate ions from TRU waste might affect the oxidation state and consequently the solubility of different radionuclides in the HLW waste. In addition, it is necessary to consider complex formation of nitrate ions with radionuclides, as well as the formation of ammonia by microbes and/or by reactions with reducing materials in the disposal facility. Consequently, complex formation of ammonia with radionuclides must also be evaluated.

In the current study, the effects of nitrate salt on radionuclide solubility were investigated experimentally with consideration given to the above perturbations. Solubility experiments of important and redox sensitive radionuclides, Tc(IV), Np(IV) and Se(0), were performed using various concentrations of sodium nitrate (NaNO3) and of Np(V) in NaNO3 solutions to investigate complex formation with NO3- ions. Solubility experiments of Pd(II), Sn(IV) and Nb(V) using ammonium chloride (NH4Cl) solution were also undertaken to investigate complex formation with NH3/NH4+ ions. A chemical equilibrium model was applied to assist the interpretation of the experimental results. No significant solubility enhancement was observed for Np and Se due to oxidation by nitrate ions. Tc solubility was, however, increased by one order of magnitude under high nitrate concentrations. Solubility enhancement by complex formation of nitrate ions with Np(V) was not observed. Solubility enhancement by complex formation of Sn and Nb were not also observed, only Pd solubility was increased by complex formation with NH3/NH4+ ions. Tendencies of the enhancement of Pd solubility were explained by the chemical equilibrium model.

This work was funded by ANRE: Agency for Natural Resources and Energy, of METI: Ministry of Economy, Trade and Industry, of Japan.

7) 40047 – A study on groundwater infiltration in the Horonobe area, northern Hokkaido, Japan

Hideharu Yokota, Yamamoto Yoichi, Keisuke Maekawa, Minoru Hara, JAEA (Japan)

It is important for assessing the safety of geological disposal of high-level radioactive waste to understand groundwater flow as the driving force of mass transport. In the groundwater-flow simulation, hydraulic boundary conditions are required, including groundwater-recharge rates. In the Horonobe area of northern Hokkaido, the Japan Atomic Energy Agency (JAEA) has been carrying out the Horonobe Underground Research Laboratory (URL) Project to understand characteristics of the geological environment. To obtain various hydrological data to estimate the recharge rate by water balance, meteorological observation and observation of river flux, etc. in the Horonobe area (snowy cold region) have been carried out. However, infiltration of water from the surface is difficult to clarify in detail because water near ground surface is sensitive to external influence such as climatic variations. It is important for precise evaluation of groundwater flow to understand shallow groundwater-flow systems (ground surface to tens of meters at depth) as a part of hydraulic boundary conditions. In the Horonobe area, subsurface temperature and soil moisture content have been observed at the URL (GL-0.7m to GL-2.3m) site since 2005 and at Hokushin Meteorological Station (HMS, GL-0.1m to GL-1.1m) since 2008. As results of these observations, it is clear that similar processes operate at both sites. Subsurface temperatures become lower with depth in the summer and higher with depth in the winter. The lowest subsurface temperatures at the shallowest and deepest at depth are observed in the middle of April and early May respectively. Concurrently, soil moisture content increases rapidly. In addition, the observed data also show that the subsurface temperature is higher than 0°C throughout the year, and keeps decreasing until early May (snow-melting season). From these results, it is suggested that, regardless of the air temperature, water at 0°C is supplied from the bottom of snow-cover to ground surface by bottom snow melting due to the insulating effect of snow. Therefore, the subsurface temperature firstly becomes the lowest at the shallowest depth. Subsequently, subsurface temperatures at greater depths decrease as the cold water infiltrates to depth with time. For the estimation of boundary conditions in groundwater-flow simulation, in this study, the shallow groundwater-flow system has been examined qualitatively on the basis of the seasonal variation of the groundwater infiltration. Results have revealed the groundwater recharge occurring in a

-96-

Abstracts

snow-covered region. In the future, it is planed that quantitative assessments will be made by the observed data of the weighing lysimeter.

8) 40051 – Effective Use of Uranium Resources and Dissolution of Recovery Uranium Storage Accumulation by a Uranium Multi-recycle System

Yuzo Yamashita, Yuzo Yamashita, Takeshi Nakamura, Kyushu University (Japan)

The uranium recovered from LWR spent fuels, containing an amount of U-235 comparable to that in the natural uranium, can be recycled as uranium fuels for LWRs by re-enrichment using a conventional centrifuge cascade. However, the remade fuel is inferior to the original fuel produced from the natural uranium on the burn-up performance, because the former includes U-236 as neutron absorber which is yielded in the spent fuel and then enhanced in re-enrichment process. Therefore, a few times recycle of recovered uranium may be available but its successive multi-recycle is not recommendable because of successive decline in burn-up resulting from U-236 accumulating in remade fuels. In this study, an idea of uranium fuel recycling is proposed which is of the feed-back of recovered uranium to a natural uranium enriching cascade. Since this cascade processes a mixture of natural and recovered uranium, the product contains U-236 enriched but diluted with the U-236-free bulk of natural uranium. In this process, the concentration of U-236 in the fuels reaches a balance with increase in recycle times. The multi-recycling fuels perform the burn-up degrees in PWRs declining but comparing favorably with convetional uranium ones. The multi-recycling system not only makes effective use of nuclear fuel resources, but also dissolves the problem of accumulating inventory in the storage of recovered uranium taking the major volume of spent fuels.

9) 40053 – Advanced ORIENT Cycle - Progress on Fission Product Separation and Utilization

Isao Yamagishi, Masaki Ozawa, JAEA (Japan); Hitoshi Mimura, Tohoku Univ.(Japan); Shohei Kanamura, Koji Mizuguchi, Toshiba Corporation (Japan)

Fission reaction of U-235 generates more than 40 elements and 400 nuclides in the spent fuel. Among them, 31 elements are categorized as rare metals. Typical yields of Pd, Ru, Rh (PGM) and Tc as rare metals will reach around 11kg, 13kg, 4kg and 3kg, respectively per metric ton of the reference FBR spent fuel (150GWd/t, cooled 5 years). Based on a ground swell on enhancements of minimization of radio-ecological impacts and economical expenditures, nuclear fuel cycle concept itself is required to be changed. Adv.-ORIENT (Advanced Optimization by Recycling Instructive Elements) Cycle strategy was hence drawn up for the minimization of radioactive waste and utilization of elements/nuclides in the wastes simultaneously., and has been developed at Japan Atomic Energy Agency. The present paper deals with the separation process of fission products in the Adv.-ORIENT Cycle, which consists of the chromatographic separation of heat-generating Cs and Sr, and the electrodeposition of Pd, Ru, Rh and Tc. Highly functional inorganic adsorbent (silica gel loaded with ammonium molybdophosphate, AMP-SG) and organic microcapsule (alginate (ALG) gel polymer enclosed with crown ether D18C6) were investigated for separation of Cs and Sr, respectively, from high-level liquid waste (HLLW). The AMP-SG adosorbed more than 99% of Cs from the simulated HLLW selectively. The ALG microcapsule adsorbed 0.0249 mmol/g of Sr and exhibited the order of its selectivity: Ba > Sr > Pd >> Ru > Rb > Ag. Separated Cs and Sr will be immobilized in ceramic forms using zeolites for the utilization as heat and/or radiation source. The electrodeposition is advantageous for both recovery and utilization of PGM and Tc because these elements are recovered as metal (Ru, Rh, Pd, (Tc)) and/or oxide (Tc) form on a Pt electrode. In the presence of Pd or Rh the reduction of Ru and Tc was accelerated in hydrochloric acid media. This co-deposition effect, however, was less in nitric acid. In the simulated HLLW, the redox reaction of Fe(III)/Fe(II) disturbed deposition of elements except for Pd. The deposits on Pt electrode showed higher catalytic reactivity on electrolytic hydrogen production than the original Pt electrode.

Keywords; Nuclear waste, Separation, Utilization, Adsorption, electrolysis, Cesium, Strontium, Ruthenium, Rhodium, Palladium, Technetium, PGM

10) 40064 – Hydrogeological Characterization based on Long Term Groundwater Pressure Monitoring

Shuji Daimaru, Ryuji Takeuchi, Masaki Takeda, Masayuki Ishibashi, JAEA (Japan)

The Mizunami Underground Research Laboratory (MIU) is now under construction by the Japan Atomic Energy Agency in the Tono area of central Japan. The MIU project is being implemented in three overlapping Phases: Surface-based Investigation (Phase I), Construction (Phase II) and Operation (Phase III). The changes of groundwater pressure due to shaft excavation can be considered analogous to a large-scale pumping test. Therefore, there is the possibility that the site scale groundwater field (several km square) can be approximated by the long-term groundwater pressure monitoring data from Phase II. Based on the monitoring observations, hydrogeological characteristics were estimated using the s–log(t/r^2) plot based on the Cooper-Jacob straight line method. Results of the s-log(t/r^2) plots are as follows. The groundwater flow field around the MIU construction site is separated into domains by an impermeable fault. In other words, the fault is a hydraulic barrier. Hydraulic conductivity calculated from s-log(t/r^2) plots are in the order of 1.0E-7(m/s). The above results from the long term monitoring during PhaseII are a verification of the hydrogeological characteristics determined in the Phase I investigations. Keywords: Large scale pumping test, long term pressure monitoring, shaft excavation, s-log(t/r^2) plot, hydrogeological characteristics.

-97-

Abstracts

11) 40065 – Development Of The Quality Management System For Borehole Investigations: (2) Quality Management Systems For Hydrochemical Investigations,

Takanori Kunimaru, Kunio Ota, Kenji Amano, JAEA (Japan); W Russell Alexander, Bedrock Geosciences (Switzerland)

An appropriate Quality Management System (QMS), which is among the first tools required for repository site characterisation, will save on effort by reducing errors and the requirement to resample and reanalyse – but this can only be guaranteed by continuously assessing if the system is truly fit-for-purpose and amending it as necessary based on the practical experience of the end-users on-site. As part of the national research and development programme for deep geological disposal of radioactive waste, the Japan Atomic Energy Agency has established two Underground Research Laboratories (URL) based around Horonobe, northern Japan, and Mizunami, central Japan. The main aim of the URL programme is to define comprehensive techniques for future repository site characterisation. At the Horonobe URL, investigation of the geological environment within sedimentary host formations is currently ongoing and one facet of this is one of the few examples worldwide (and the first use in Japan) of the study of rock matrix porewater hydrochemistry, in conjunction with groundwater hydrochemistry. The data and interpretation are described in detail elsewhere but, before detailed data interpretation, the first priority should always be an analysis of the data quality and this is addressed here. The quality of the porewater data have been categorised based on an existing system of ranking groundwater data quality, which has been developed for the fractured crystalline rocks of the Fennoscandian Shield. The ranks range from Category 1 (highest quality) to Category 5 (lowest quality) and of particular importance in the categorisation criteria are (i) the degree of porewater contamination with drilling fluid, (ii) indications of sample perturbations such as oxidation or CO2 reaction and (iii) the completeness of major ion and isotope analytical data. With the integration of all available information on hydrochemistry, hydrogeology and borehole history, more than 150 porewater and 24 groundwater datasets have been examined to date. The results of the QA audit by multi-operators have been compared in order to assess the objectivity of the categorisation, methodology and this clearly shows operator-dependent differences in categorisation of the same data set. As such, the initial guidelines for assigning the QA categories have been improved so as to reduce dependence on expert judgement and this approach will be discussed in detail here. This approach will increase transparency in the data handling and so increase stakeholders’ confidence in both the data set itself and in the results obtained using such data for a repository site characterisation.

12) 40067 – An Analytical Model on the Sealing Performance of Space for the Design of Buffer Material and Backfill Material

Haruo Sato, JAEA (Japan)

The self-sealing function of spaces between buffer material and overpack, tunnel wall and disposal pit by swelling occurred with water penetration is expected for bentonite which will be used as buffer material and part of the backfill material in the geological of high-level radioactive waste in Japan. The clearance filling properties of Na-bentonite for buffer material and backfill material specifications have been studied for distilled water and saline water conditions, for example, it is reported that Na-bentonite seals clearance even under saline water conditions in a range of effective bentonite densities which are higher than 1.3 kg/dm3, for a bentonite dry density of 1.8 kg/dm3 and a clearance ratio (volumetric ratio of clearance volume to total volume including pore and the clearance) of 10 % (1.11 in volumetric swelling ratio: Rvs) in experiments for Kunigel-V1 (Na-montmorillonite content ca. 50 wt.%). Although such information is useful for judging whether clearance is filled, the filling properties of bentonite depend on groundwater condition, silica sand content, montmorillonite content in the bentonite and the bentonite dry density, even though at the same effective bentonite density. In the present study, the author constructed an analytical model on the clearance filling performance for the design of buffer material and backfill material, based on the swelling properties of Na-montmorillonite which is the clay mineral constituent of bentonite. In the modelling, experimental data on the Rvs of bentonite (Kunigel-V1) reported so far were analyzed. Consequently, it was found that dry density of the bentonite when the free swelling reached equilibrium state was approximately in the range of 0.204-0.241 kg/dm3 for distilled water condition. These dry densities are equivalent to the range of 0.106-0.126 kg/dm3 in montmorillonite partial density. On the other hand, the Rvs of bentonite decreased for saline water condition. Based on these experimental data for Rvs, the author derived an analytical expression to be able to calculate Rvs values against various dry densities and bentonites (different montmorillonite and silica sand contents), and showed the manner for judging whether clearance is sealed for an arbitrary clearance ratio. In the paper, the author shows the calculated results of Rvs versus dry density under various conditions.

13) 40069 – Current Status of Horonobe URL Project in Construction Phase

Hironobu Abe, Koichiro Hatanaka, JAEA (Japan)

Horonobe URL project has been pursued by JAEA (Japan Atomic Energy Agency) to establish and demonstrate site characterization methodologies, engineering technologies, and safety assessment methodologies for HLW geological disposal in relevant geological environment with sedimentary rock and saline groundwater distributing in Horonobe area, Hokkaido, Japan. In the Horonobe URL project, surface-based investigation phase (Phase I) has already completed in the year 2005, and then construction phase (Phase II) has initiated in the same year. Currently, construction of the

-98-

Abstracts

underground facilities such as shafts/drifts which were designed in Phase I, investigations of the geological environment in the excavated shafts/drifts and confirmation of applicability of engineering technologies has been alternately carried out as Phase II activities of the project. At the end of January, 2010, ventilation shaft has reached at the depth of GL-250m and east access shaft has reached at the depth of GL210m. Concerning horizontal drifts, construction of GL-140m gallery with total length of 184m has completed and GL-250m gallery has partially constructed with the length of 42m. During the construction so far, monitoring for the construction safety such as convergence measurements, tunnel wall observation, sampling of groundwater and rock, investigations for evaluating excavation damaged zone along shaft/drift were carried out. In addition, shotcrete construction test and grout injection test by using low alkaline cement material were carried in the horizontal drifts. In this paper, status of the URL construction and research activities mentioned above are outlined as the current achievement of the Horonobe URL project.

14) 40074 – Development of New Ultrafiltration Techniques Maintaining In-Situ Hydrochemical Conditions for Colloidal Study

Daisuke Aosai, Yuhei Yamamoto, Takashi Mizuno, JAEA (Japan)

Chemical state of elements in groundwater is one of the most important information for understanding behavior of elements in underground environment. Chemical state of elements controlled mainly by groundwater physico-chemical parameters. Because the change of physico-chemical parameters of groundwater, due to pressure release and oxidation during sampling, causes changes in chemical state of elements, systematic methodologies for understanding in situ chemical state is required. In this study, in order to understand chemical state of elements in groundwater, an ultrafiltration instrument for maintaining in-situ pressure and anaerobic conditions was developed. The instrument developed in this study for ultrafiltration made of passivated Stainless Used Steel (SUS) materials, was designed to keep groundwater samples maintaining in-situ pressure/anaerobic conditions. Ultrafiltration of groundwater was conducted at a borehole drilled from the 200 mbGL (meters below ground level) Sub-stage at a depth of 200 m at the Mizunami Underground Research Laboratory. Chemical analyses of groundwater were also conducted using samples filtered under both pressurized/anaerobic and atmospheric conditions and passivated SUS materials with different elapsed times after passivation. The results indicate that our ultrafiltration method is suitable for collection of filtered groundwater and passivation is an essential treatment before ultrafiltration. Keywords: groundwater, ultrafiltration.

15) 40089 – Sorption Behavior of Iodine on Calcium Silicate Hydrates Formed as a Secondary Mineral

Keisuke Shirai, Yuichi Niibori, Akira Kirishima, Hitoshi Mimura, Tohoku Univ. (Japan)

This study examined the sorption behaviors of iodine into CSH gel without dried processes, considering the repository system saturated with groundwater after the backfilling. In glove box saturated with N2 gas, each sample of CSH gel was synthesized with CaO, SiO2, and distilled water with liquid/solid ratio 20. Then, 1 mM iodine solution is added into the aqueous solution including the CSH gel with various Ca/Si molar ratios under the isothermal condition (298 K).

In the results, even if the Ca/Si ratio is relatively small (<1.0), the distribution coefficients (Kd) of iodine on CSH gel without dried processes were two-order of magnitude larger than those with dried processes. For example, the value of Kd (ml/g)?was about 380 in the case of 0.5 in Ca/Si molar ratio. Furthermore, it was confirmed the sorption behavior attain equilibrium in 24 hours. These suggest secondary mineral of CSH gel would retard the migration of anionic nuclides under the condition saturated with groundwater. In comparison, this study prepared also the co-precipitated samples of CSH gel and iodine, where the 1 mM iodine solution is added before curing the CSH gel. These distribution coefficients and the kinetics almost agreed with those mentioned above.

16) 40098 – Development of the Quality Management System for Borehole Investigations: (1) Quality Management System for Hydraulic Packer Tests

Shinji Takeuchi, Takanori Kunimaru, Kunio Ota, JAEA (Japan); Bernd Frieg, Nagra (Switzzerland)

The Japan Atomic Energy Agency has been conducting two underground research laboratory projects at Mizunami in central Japan and Horonobe in northern Japan, with the main aim of establishing the relevant techniques for future repository site characterization. Surface-based investigations have been completed and one of the remaining issues is the development of an appropriate quality management system (QMS) which is applicable to all aspects of the site characterization process. A number of hydraulic packer tests (HT) have been carried out during the borehole investigations and the hydraulic properties of the host rock, which ensure reliability of hydrogeological conceptualization and groundwater flow simulations, such as transmissivity and a flow model, have been determined. It is thus of importance to ensure the quality of HT data for increasing confidence in such modeling and simulation results. A QMS has been established for the entire HT process, from data acquisition through to data analysis, and it has been tested in the borehole investigation program. This involves; Test interval selection: The test interval is selected based on the results of detailed fluid logging, core description, and drilling fluid and hydraulic pressure monitoring. A tally list for ensuring the depth of the test interval has been prepared. Test tool inspection: Pressure transducers, packers and shut-in valves of the test equipment are checked to ensure these components work correctly before starting the test. Data quality audit: The data quality is checked by assessing the pressure changes and their derivative for each test phase. The whole test events

-99-

Abstracts

and decisions made during HT are recorded in an appropriate format to ensure the traceability of the test. Quick analysis: The most representative test phase is selected based on the data quality and behavior, from a sequence of different phases, and the hydraulic parameters are estimated under the simple flow model. The results of this analysis are described with a compilation of the whole data and graphs on a quick look report. Detailed analysis: Detailed analysis with type curve matching or numerical analysis is adopted. Transmissivity normalized plot is of use to derive transmissivity estimations and a plausible flow model. This paper presents the preliminary development of the QMS for HT data acquisition and analysis during the borehole investigations and explores aspects which can be improved following on-site tests.

17) 40101 – Comparison of Post-Irradiation Experimental Data and Theoretical Calculations for Inventory Estimation of Long-Lived fission Products in Spent Nuclear Fuel

Shiho Asai ,Yukiko Hanzawa, Keisuke Okumura, Hideya Suzuki, Masaaki Toshimitsu, Nobuo Shinohara, JAEA (Japan);

Kensuke Suzuki, Satoru Kaneko, Tokyo Electric Power Company (Japan)

An inventory estimation of long-lived radionuclides in high-level radioactive wastes (HLW) has been a major concern for a long-term safety assessment of HLW disposal. Among such nuclides, Se-79, Cs-135, and Sn-126 attract much attention due to their potential migration ability to the environment through multi-barrier components. However, their yields for U-235 fission have not yet been sufficiently confirmed by post irradiation experiments because of the difficulties in the measurements. In order to evaluate the reliability of the inventory estimation with theoretical calculations, the amounts of Se-79, Cs-135, and Sn-126 generated during an irradiation of UO2 fuel are necessary to be determined experimentally. In this study, mass spectrometry was applied to the determination of Se-79, Cs-135, and Sn-126. The results were compared with those obtained by ORIGEN2 calculations with JENDL-3.3 based nuclear data library. About 5 g of UO2 fuel irradiated in Japanese commercial PWR with an average burn-up of 44.9 GWd/t was sampled as a specimen and dissolved with 50 mL of 4 M nitric acid in a hot cell. The resultant solution was filtered to remove the insoluble residue. The portion of the supernatant was taken and diluted in 1 M nitric acid to adjust the amount of uranium to 0.5 mg. After Se, Cs, and Sn were chemically separated, each fraction was analyzed to determine the concentrations of Se-79, Cs-135, and Sn-126. The results for Se-79 and Cs-135 obtained in this study showed a good agreement with those obtained by ORIGEN2 calculations. This indicates that ORIGEN2 calculation is applicable to the estimation of the amounts of Se-79 and Cs-135 generated in a spent nuclear fuel. In contrast, the experimentally determined concentration of Sn-126 was equivalent to 60% of that obtained by ORIGEN2 calculation, which suggests that a part of Sn initially existed in the sample UO2 pellet precipitated as insoluble residue during the dissolution process.

18) 40103 – Selective Uptake of Palladium from High-Level Liquid Wastes by Hybrid Microcapsules

Hitoshi Mimura, Takashi Sakakibara, Yan Wu, Yuichi Niibori, Tohoku Univ. (Japan); Shin-ichi Koyama, Takashi Ohnishi, JAEA (Japan)

Fine crystalline powders of KCuFC were immobilized with alginate gel polymers by sol-gel methods. The uptake properties of KCuFC-microcapsules (KCuFC-MC) were examined by batch and column methods. The size of KCuFC-MC particle was estimated to be about 1 mm in diameter, and KCuFC powders were uniformly dispersed in KCuFC-MC particles. The uptake rate of Pd2+ for KCuFC-MC was attained within 3 d, and the uptake of Pd2+ was found to be independent of the temperature and coexisting HNO3 concentration. As for the breakthrough properties of Pd2+ through a column packed with KCuFC-MC, a breakpoint of 5% breakthrough was enhanced with lowering of flow rate and independent of coexisting HNO3 concentration. The Pd2+ ions were selectively adsorbed in the KCuFC crystal phase, while other metal ions such as Ru(NO)3+ and ZrO2+ were absorbed in the alginate phase. High uptake percentage of 98.6% was obtained by using the dissolved solutions of spent fuel from FBR-JOYO (119 GWd/t, JAEA). The alginate film enclosing KZnFC was further prepared by using the support of cellulose filter paper, where the Pd2+ ions were selectively adsorbed on the KZnFC-MC film. The alginate film enclosing insoluble ferrocyanides are predicted for the selective separation of Pd2+ as an ion-exchange filter. Thus, the microcapsules enclosing insoluble ferrocyanides are effective for the selective separation of Pd2+ from high-level liquid waste (HLLW).

19) 40113 – An Empirical Model To Determine The Modes Of Corrosion Of Carbon Steel

Toshikatsu Maeda, Masatoshi Watanabe, Seiji Takeda, Shinichi Nakayama, JAEA (Japan)

Carbon steel is a candidate material for overpack of high-level radioactive waste disposal in Japan. One of its expected functions is to avoid groundwater from contacting with vitrified waste form. The corrosion rate is a determining factor for the overpack lifetime, and it is highly dependent on the mode of corrosion. Carbon steel is an alloy that could be attacked by localized corrosion in the form of pitting corrosion or crevice corrosion under certain water chemistries. For example, it is known that carbon steel is passivated in solutions above a limiting pH; pH that is referred to as the general corrosion/passivation transition pH and is denoted as pHd. Carbon steel corrosion rates depend on the mode of corrosion (general or local corrosion), which is determined by the pHd value. Predicting the pHd for carbon steel in near field environments is essential to evaluate the expected lifetime of overpacks, especially considering the possibility of highly alkaline environments induced by cementitious materials in the disposal facility. In this study, an empirical model was developed to determine whether near field environments fall in the passivation or non-passivation domain for carbon

-100-

Abstracts

steel. Using the experimental data obtained by previous studies, the pHd was defined as a function of four factors shown in equation (1), where the activity of proton ion ([H+]) for pHd is assumed to be a linear combination of the logarithms of the total carbonate concentration (C total), the chloride ion concentration (Cl-), and the limiting current density of dissolved oxygen diffusion (iO2), and the inverse of absolute temperature of contacting solution (T). The derived equation fitted well experimental data from previous studies.

20) 40124 – Trends in Scenario Development Methodologies and Integration in NUMO's Approach

Takeshi Ebashi, Katsuhiko Ishiguro,NUMO (Japan); Keiichiro Wakasugi, JAEA (Japan); Hideki Kawamura, Obayashi Corporation (Japan);

Irina Gaus, Stratis Vomvoris, Andrew Martin, Nagra (Switzerland); Paul Smith, Safety Assessment Management (UK)

The development of scenarios for quantitative or qualitative analysis is a key element of the assessment of the safety of geological disposal systems. As an outcome of an international workshop attended by European and the Japanese implementers, a number of features common to current methodologies could be identified, as well as trends in their evolution over time. In the late nineties, scenario development was often described as a bottom-up process, whereby scenarios were said to be developed in essence from FEP databases. Nowadays, it is recognised that, in practice, the approaches actually adopted are better described as top-down or "hybrid", taking as their starting point an integrated (top-down) understanding of the system under consideration including uncertainties in initial state, sometimes assisted by the development of "storyboards". A bottom-up element remains (hence the term "hybrid") to the extent that FEP databases or FEP catalogues (including interactions) are still used, but the focus is generally on completeness checking, which occurs parallel to the main assessment process. Recent advances focus on the consistent treatment of uncertainties throughout the safety assessment and on the integration of operational safety and long term safety.

21) 40137 – Development of Methodology to Construct a Generic Conceptual Model of River-valley Evolution for Performance Assessment of HLW Geological Disposal

Makoto Kawamura, Kenichi Yasue, Tadafumi Niizato, Shin-ichi Tanikawa, JAEA (Japan)

In order to assess the long-term safety of a geological disposal system for high-level radioactive waste (HLW), it is important to consider the impact of uplift and erosion, which cannot be precluded on a timescale in the order of several hundred thousand years for many locations in Japan. Geomorphic evolution, caused by uplift and erosion and coupled to climatic and sea-level changes, will impact the geological disposal system due to resulting spatial and temporal changes in the disposal environment. Degradation of HLW barrier performance will be particularly significant when the remnant repository structures near, and are eventually exposed at, the ground surface. In previous studies, river erosion was identified as the key concern in most settings in Japan. Here, therefore, we present a methodology for development of a generic conceptual model for performance assessment based on best current understanding of river erosion in Japan.

Critical considerations that have to be taken into account when interpreting the geological record of past river-valley evolutions, as preserved in ancient fluvial deposits, include: 1) the spatial variation in the relative significance of erosion and sedimentation at any time between upper- and lower-reaches of rivers originating in mountainous terrain 2) the temporal variation in the extent of erosion / sedimentation at any specific location during glacial / interglacial cycles 3) the balance between uplift and vertical erosion as a result of the hardness of the riverbed rock 4) the balance between vertical and lateral erosion – ranging from formation of narrow gorges to wide meandering flood plains 5) the varying duration and intensity (as assessed by sea level change) of past glacial / interglacial cycles.

Interpretation of the impact of such phenomena at relevant locations in Japan has led to development of a generic conceptual model which contains the features typical of mid-reach rivers. This paper presents the methodology to develop the conceptual model, identifies the simplifications and uncertainties involved and assesses their consequences in the context of repository performance. Details of resultant analyses using this conceptual model will be discussed in another paper presented in ICEM’10 by Miyahara et al.

22) 40176 – Structural Integrity Evaluation Approach for PWR Spent Nuclear Fuel

Yun Seog Nam, Seong Ki Lee, Yong Hwan Kim, Jeon Kyeong Lak, Choi Ki Sung, Chang Sok Cho, KNF (Korea Rep.)

PWR fuel assembly experiences many changes from the time it is manufactured, loaded in the reactor and repositioned in the core several times until finally removed from the reactor for the interim storage, reprocessing or final disposal etc. Under a severe radiation and a thermo-mechanical condition, this mechanism can alter fuel assembly characteristics such as its mechanical properties, geometrical shape, material characteristics etc. Any of these alterations which impact spent nuclear fuel (SNF) integrity should be considered to design a cask/canister for its transportation or intermediate dry storage. Regarding the cask/canister design, there could be a freedom to design a system that mitigates the forces transmitted to SNF and fuel rods. If the storage cask/canister or transport package design prevents or mitigates forces transmitted to its contents such that structural integrity is not significantly compromised, the detailed SNF properties are necessary to make a decision of the elaborated design parameters, assuming other factors (temperature,

-101-

Abstracts

inert atmosphere, etc.) have been adequately addressed. An approach to those work formations is to analyze mechanical characteristics of structural components which are the mechanical properties of grid spring reflecting its irradiation effect, SNF strength and SNF structural deformation such as its bow, twist etc. Those informations are also used to evaluate hypothetical accident like a drop accident of SNF cask/canister, to select limiting fuel assembly for cask/canister to accommodate various kinds of SNFs and to design transportation or storage system for SNFs. Especially, the fuel assembly structural properties are a sort of essential data. Thus, in this paper, some approaches to evaluate SNF mechanical characteristics are suggested through the existing technical information review, some test data and the analysis methodology, and also closely study the mechanical characteristics of a representative SNF for its general comprehension.

23) 40200 – Development of Program Categories to Assess the Radiological Dosage during Spent Fuel Transportation

Suhong Lee , Sangwon Shin, Enesys, Jaemin Lee, Enesys.Co.,Ltd. (Korea Rep.); Kiyroul Seong, Jeonghyoun Yoon, KRMC (Korea Rep.)

Korea is seeking the plan to put spent fuels in interim storage facility as disposal business of low and intermediate-Level radioactive waste started. First, There is a strong probability that transport of the spent fuels will be arranged in marine transportation. Accordingly, building of transportation risk assessment system in the marine transportation comes to the force as a real problem. Therefore, in this study, I organized the basic categories of program to develop transportation risk assessment program about specialized transport in marine transportation of spent fuels. The possibility of radiation exposure in marine transportation consists of the normal condition and the accident condition. The direct radiation exposure that the workers in loading and crews of the vessels will receive is considered in the normal condition. I decide the leakage rate dividing the accident condition that the radioactive materials are spilled into air and marine and doing the effect by emission by fire and physical damage. Also, as for the radioactive materials floated in the air, calculated applying the atmospheric dispersion factors to them and evaluate the direct radiation exposure and internal radiation exposure. As for marine contamination, the radioactive materials move along ocean currents. Generally, there can be the direct radiation exposure of ocean activists due to the moving radioactive materials according to the flow of the ocean currents in contaminated marine and the direct radiation exposure and the indirect radiation exposure by breathing of the subjects near the shores due to the smoke and fog form that occurs by seawater. Also the internal radiation exposure by intake contains one of marine life. I set the categories to build the overall programs to build this with the comprehensive program that have effects on radiation while the spent fuels are transported combining it with the transport conditions or the characteristics of the subjects of radiation exposure and the environmental factors. And building of the program will be completed through it. The completed program can evaluate various conditions of marine transportation after it will be specialized in marine transportation.

24) 40225 – Exploiting synergies between the UK & Japanese geological disposal programmes

Ellie Scourse, Atkins (UK); Hideki Kawamura, Obayashi Corporation (Japan); Ian G. McKinley, McKinley Consulting (Switzerland)

The early ‘80s UK programme for deep geological disposal of high-level radioactive waste was advanced and at the stage of characterising potential sites. When this project was put on hold in the mid ‘80s, much expertise in this field was lost.

In Japan R&D in the ‘80s resulted in major generic safety assessments to demonstrate feasibility in the ‘90s. This led to the establishment of NUMO (Nuclear Waste Management Organization of Japan) and the initiation of siting based on volunteerism. This novel approach required more flexible methodology and tools for site characterisation, repository design and safety assessment. NUMO and supporting R&D organisations in Japan have invested much time and effort preparing for volunteers but, unfortunately, no discussions with potential host communities have yet developed to the point where technical work is initiated.

Presently, the UK is moving forward; with the NDA RWMD (Nuclear Decommissioning Agency Radioactive Waste Management Directorate) adopting a NUMO-style volunteering approach and a flexible design catalogue. Communities have already shown interest in volunteering. The situation is thus ideal for collaboration.

The paper will expand on the opportunities for the UK and Japan to benefit from an active collaboration and discuss how this can be most efficiently implemented.

Key Words: UK, Japan, collaboration

25) 40239 – Realistic Consequence Analysis of River Erosion Scenarios for a HLW epository

Kaname Miyahara , Makoto Kawamura, Manabu Inagaki, JAEA (Japan); Ian G. McKinley, McKinley Consulting (Switzerland);

Michael J. Apted, Monitor Scientific LLC (USA)

Uplift and erosion cannot be precluded in most sites in Japan. As no assessment cut-off times have yet been defined, erosive radionuclide release scenarios must be developed and analyzed, even if these occur far in the future. Obviously, uplift and erosion will cause major disruption of the engineered and natural barriers when the repository nears, and is

-102-

Abstracts

eventually exposed at, the ground surface. In a previous study, a simple linear uplift process was combined with a more detailed assessment of river erosion, which was identified as the key erosion process in Japan. This indicated the robustness of the reference HLW disposal system: the consequences of erosion of the repository being small when compared to the yardstick provided by natural radionuclide fluxes.

The original model was rather simple, but highlighted the importance of cyclic erosion in response to the changing environment in glacial and inter-glacial periods. Therefore, the geological record of such cycles preserved in Japan as river terraces has been studied further, with the aim of constructing a more realistic river erosion model. It is evident that, although riverbed deepening occurs during glacial cycles, significant sedimentation also occurs and the timing of these phases during the cycle differs in the upper and lower reaches of rivers. Unfortunately the situation is particularly complicated in mid-reach settings – and these may be most relevant for repositories that avoid mountain locations and cannot be established in the highly populated coastal plains. Here interpretation of complex river terrace structures must take into account both variations in the length of past glacial/inter-glacial periods and in the resistance of different riverbeds to erosion. Further, when the width of river channel is significantly less than that of the valley, meanders develop giving sequences of river terraces that are observed only on one side (asymmetric) or on both sides (symmetric) of river. Unlike in the upper reaches, in such environments it cannot be assumed that valley profiles are effectively constant in time and the gradual progress from eroding mountains towards formation of a peneplain has to be considered.

This paper describes a conceptual model based on generalization of these observations and resultant consequence analyses, again using comparisons with natural radionuclide fluxes. Geological evidence supporting such erosion models will be discussed in another paper presented in ICEM’10 by Kawamura et al.

26) 40272 – Removal of Fission Products in the Spent Electrolyte Using Iron Phosphate Glass as a Sorbent

Ippei Amamoto, Masami Nakada, Yoshihiro Okamoto, JAEA (Japan); Naoki Mitamura, Tatsuya Tsuzuki, Central Glass Co. Ltd. (Japan);

Yasushi Takasaki, Atsushi Shibayama, Akita University (Japan); Tetsuji Yano, Tokyo Institute of Technology (Japan)

The [3LiCl(59.5mol%)-2KCl(40.5mol%)] eutectic medium used in the pyroreprocessing by the electrorefining method will be contaminated by the accumulation of various actinoid elements (An) and fission products (FP) due to prolonged electrolytic operation. Though the An can be removed at the pyrocontactor step by extraction using cadmium melt, the FP remain in the eutectic medium, which is regarded as a spent electrolyte at certain stage, can cause the rising of the melting point of the electrolytic bath and /or lowering the current efficiency. Some measures e.g., its replacement by a virgin medium, etc. should be taken to maintain its stable condition. The constant replacement of the electrolyte, however, could lead to the generation of enormous volume of high-level radioactive waste (HLW). In terms of environmental load reduction and economical improvement, it is desirable to have the spent electrolyte purified for recycling by removing its FP. Some technical developments on spent electrolyte treatments have been carried out in several countries. One of them is the zeolite sorption of FP following An removal which is being developed in the USA, Japan, etc.. Meanwhile, the contaminants precipitation methods by converting contaminants to insoluble compounds are undertaken in Russia and a few other countries and are anticipated to reduce the total volume of HLW. In the case of the Russian process, its purpose is to remove contaminants such as the minor actinoids (MA) and rare earth elements (REE) from the medium before its disposal. In our case, the main objective is to recycle the purified medium, delaying its disposal for as long as possible. It is with this objective in mind that this study was undertaken. We have introduced the simple filtration method to remove REE particles which were formed due to the conversion of REE chlorides to phosphates. Here, the iron phosphate glass is used as a filtration medium for the removal of FP particles. However, some soluble FP such as compounds of alkali-metals, alkaline-earth metals, etc. still remain in the eutectic medium. This time around, on an experimental basis, the iron phosphate glass has been used as a sorbent instead, to remove the soluble FP. We have obtained some positive results and have intention to incorporate it into the spent electrolyte recycle process as a part of the FP separation and immobilization system.

27) 40295 – Propagation and Interactions of Acoustic Waves in a Waveguide Attached at the Surface of Rock

Jin-Seop Kim, Kyung-Soo Lee, S. Kwon, KAERI (Korea Rep.); Gye-Chun Cho, KAIST (Korea Rep.)

The dynamic response of a rock mass is inevitable in the systematic long-term monitoring and management in the radioactive waste disposal repository. With this point of view, AE (acoustic emission) detection is considered to be a promising technique for monitoring the in-situ performance of near-field rock mass. In this study, propagation and interactions of guided acoustic waves in a waveguide, which is required to in-situ application of AE monitoring system were investigated. The changes in acoustic wave amplitude, time delay, frequency variation, and system transfer function were measured between the waveguide and a rock sample. Subsequently the waveguide coupling conditions filled with epoxy were compared with mechanical type of coupling for the validity of field application. Three type of coupling methods were used to compare each other. One is a type of direct contact with granite specimen(Ch.2) which is same kind of rock mass with KURT(Korea Underground Research Tunnel) by using a vacuum grease as a reference data. Another is a waveguide-aided connection method without any coupling fluid(Ch.1) and the other is a same type with Ch.1 except for the use of epoxy resin for a coupling fluid between the gap of a specimen and a waveguide. The time delay between CH.1 and CH.2 was 0.025 msec and in case of CH.3 and CH.2 0.012 msec. Energy decay ratio of first

-103-

Abstracts

arrival signal for CH.1/CH.2 was 0.68 and 0.81 for CH.3/CH.2. Normalized amplitude attenuation model with a distance was successfully obtained. The similarity of two signals were quantified using a mathematical tool called mean magnitude squared coherence(MSC). The mean MSC between CH.1-CH.6, CH.2-CH.6, and CH.3-CH.6 are 0.8727, 0.9262 and 0.9040 respectively. The signal directly attached to the surface of a specimen presents the most similar pattern with the applied source signal. While the signals from the epoxy filled waveguide show closer relationship with the source than that without filling couplant. Additionally by using of system transfer function, the effect of waveguide is apparently shown at the frequency of 118 kHz. The results derived from this study can be valuable information for the quantitative analysis of signal processing in AE source localization and degree of crack damage in a radioactive waste repository.

SESSION D7: D&D Poster

1) 40075 – Methods of Selected Input Calculation Data Verification And Their Influence On Decommissioning Cost In the OMEGA Code

Frantisek Ondra. Vladimir Daniska, Ivan Rehak, Oto Schultz, DECONTA, a.s. (Slovakia); Vladimir Necas, Slovak University of Technology in Bratislava (Slovakia)

The aim of this contribution is development of methodology for verification of selected input calculation data (performance unit parameters, work group structure, and duration of time-dependent activities) of the OMEGA Code in the individual PSL (Proposed Standardised List) structure parts. The OMEGA Code, developed by DECOM, a.s., is used for calculation of nuclear power plant decommissioning parameters and decommissioning planning. The code represents a complex tool modelling a real flow of materials and radioactivity in the whole process of decommissioning, beginning from pre-dismantling decontamination and terminating with either final disposal of radioactive waste or release of non-contaminated materials to the environment. The analytical methodology for evaluation of input data inaccuracy impacting on calculation of cost and other output decommissioning parameters was developed. This methodology is based on application of coefficients representing calculated cost relative change for contingency adjustment purpose. The decommissioning process includes significant amount of various technical as well s administrative activities. Decommissioning parameters calculation is performed by calculation procedures modelling these decommissioning activities. The procedures use for calculation a lot of input and unit data. Decommissioning calculation parameters projects having been performed using the OMEGA Code showed the necessity of development of a new verification module. This module is able to display selected input data used for calculation for either specific inventory database item or specific decommissioning activity in the PSL structure. The new verification module allows compare selected input parameters to reference values including graphic display of differences. There is also a possibility to perform recalculation of the some decommissioning option using reference values of input data and consequently to compare calculated decommissioning data. These data (e.g. exposure, labor, and cost) and their differences can be compared both in tables and graphs, for either specific inventory database item or specific PSL activity. The verification module within the OMEGA Code brings a new view on impact of selected input data change on calculated decommissioning data. It is auxiliary tool for contingency calculation using analytical methods developed last year.

2) 40126 – Detailed Standardized Decommissioning Parameters Calculation for Larger echnological Aggregates and Relevant Buildings in Nuclear Power Plants using the OMEGA Code

Peter Bezak, Vladimir Daniska, DECONTA, a.s. (Slovakia); Ivan Rehak, DECOM, a.s. (Slovakia)

Computer code OMEGA, developed by DECOM a.s., is used for evaluation of nuclear facility decommissioning activities. The Code implements in full extent the standardised cost structure PSL. Decommissioning activities of nuclear facility are involved in compact calculation structure. Calculation models a real material and radioactivity flow and reflects a radioactivity decay during decommissioning process. Calculation processes material items, which are linked to decommissioning procedures (dismantling, demolition and decontamination procedures) in calculation structure. Inventory database contains approx 90 standard material items of technological equipment (pipes, valves, tanks etc.), approx 50 specific items (pieces, planked components etc.). The inventory database also contains 14 building categories (masonry, concrete, steel construction etc.). A new task is costing the larger technological aggregates decommissioning in nuclear power plants. The paper introduces development of larger technological aggregates inventory database. These aggregates include: 1. Reactor and its internals 2. Steam generator 3. Pressurizer 4. Refueling machine, etc.. Larger technological aggregates decommissioning activities need to be implemented into decommissioning planning and costing Code OMEGA. So decommissioning procedures representing these activities have to be developed for technological aggregates, and dismantling unit factors need to be set up as well. A proper definition of dismantling techniques and workgroups performing the techniques is also important, taking into account presence of activated and contaminated materials to be dismantled. Where a dose rate is higher than a limit for manual dismantling, remote dismantling techniques are applied. Paper introduces manual and remote techniques available for larger technological aggregates dismantling. Dismantling procedures of larger technological aggregates are based on reversed sequence of their commissioning. The paper also deals with specific dismantling procedures, when equipment is dismantled as a whole and moved to a fragmentation facility. There it will be fragmented to smaller parts, put into containers for disposal in

-104-

Abstracts

radioactive waste repository. Decommissioning of larger technological aggregates relates to buildings where aggregates are housed in and the paper also deals with their demolition.

3) 40190 – Dismantling Method of Fuel Cycle Facilities Obtained by Dismantling of the JRTF

Fumihiko Kanayama, JAEA (Japan)

The Japan Atomic Energy Research Institute Reprocessing Test Facility (JRTF) was the first reprocessing facility which was constructed by applying only Japanese technology to establish basic technology on wet reprocessing. JRTF had been operated since 1968 to 1969 using spent fuels (uranium metal / aluminum clad, about 600kg as uranium metal and 600MWD/T) from the Japan Research Reactor No.3 (JRR-3). Reprocessing testings on PUREX process were implemented at 3 runs, so that, 200g of plutonium dioxide were extracted. After JRTF was shut down at 1970, it had been used for research and development of reprocessing since 1971. The more mature research and development of nuclear are, the more opportunity of dismantling of old nuclear facilities would be. JAEA has an experience of full scale of dismantling through decommissioning of JPDR. On the other hand, we didn’t have that of fuel cycle facility. Moreover, it is considered that dismantling methods of nuclear reactor and fuel cycle facility are different for following reason, components contaminated TRU nuclide including Pu, contamination form being many kinds, and components installed inside narrow cells. Dismantling methods are important factor to decide manpower and time for dismantling. So, it is indispensable to optimize dismantling method in order to minimize manpower and time for dismantling. Considering the background mentioned above, the decommissioning project of JRTF was started in 1990. The decommissioning project of JRTF is carrying out phase by phase. Phase 1; Investigation for dismantling of the JRTF. Phase 2; R&D of decommissioning technologies for dismantling of the JRTF. Phase 3; Actual dismantling of the JRTF. There were several components used for reprocessing and a system for liquid radwaste storage, and those were installed inside of each of several thick concrete cells. The inner surfaces of each cell were contaminated by TRU nuclides including Pu. In phase 3, components used in reprocessing and a system for liquid radwaste storage were dismantled. Moreover, concrete walls (including ceiling) were opened to make entrances in this work. Effective practices for dismantling fuel cycle facilities were obtained through these works. On this report, I introduce effective dismantle method obtained by actual dismantling activities in JRTF.

4) 40191 – Computer Simulation of Cryogenic Jet Cutting for Dismantling Highly Activated Facilities

Sung-Kyun Kim, Kune-Woo Lee, KAERI (Korea Rep.)

Cryogenic cutting technology is one of the most suitable technologies for dismantling nuclear facilities due to the fact that a secondary waste is not generated during the cutting process. In this paper the feasibility of cryogenic cutting technology has been investigated by using a computer simulation. In the computer simulation, a hybrid method combined with the SPH (smoothed particle hydrodynamics) method and with the FE (finite element) method was used. And also, a penetration depth equation, for the design of the cryogenic cutting system, was used and the design variables and operation conditions to cut a 10 mm thickness for steel were determined. Finally the main components of the cryogenic cutting system were developed on the basis of the obtained design values.

5) 40193 – Strippable Core-shell Polymer Emulsion for Decontamination of Radioactive Surface Contamination

Bum-Kyoung Seo, Bum-Kyoung Seo, Kune-Woo Lee, KAERI (Korea Rep.)

Strippable coatings are innovative technologies for decontamination that effectively reduce loose contamination. These coatings are polymer mixtures, such as water-based organic polymers that are applied to a surface by paintbrush, roller or spray applicator. In this study, the core-shell composite polymer for decontamination from the surface contamination was synthesized by the method of emulsion polymerization and blends of polymers. The strippable polymer emulsion is composed of the poly(styrene-ethyl acrylate) [poly(St-EA)] composite polymer, poly(vinyl alcohol) (PVA) and polyvinylpyrrolidone (PVP). The morphology of the composite emulsion particle was core-shell structure, with polystyrene (PS) as the core and poly(ethyl acrylate) (PEA) as the shell. Core-shell polymers of styrene (St)/ethyl acrylate (EA) pair were prepared by sequential emulsion polymerization in the presence of sodium dodecyl sulfate (SDS) as an emulsifier using ammonium persulfate (APS) as an initiator. Related tests and analysis confirmed the success in synthesis of composite polymer. The products are characterized by FT-IR spectroscopy, TGA that were used, respectively, to show the structure, the thermal stability of the prepared polymer. Two-phase particles with a core-shell structure were obtained in experiments where the estimated glass transition temperature and the morphologies of emulsion particles. Decontamination factors (DF) of the strippable polymeric emulsion were evaluated with the polymer blend contents. The decontamination factors obtained for Sr-90 on the disk plate studies were observed the DF values of 8.9 to 12.8 at all the polymer composition.

-105-

Abstracts

SESSION R4: ER Poster

1) 40025 – Improvement of quicklime mixing treatment by carbon dioxide ventilation

Yuki Nakagawa, Hisayoshi Hashimoto, Hitachi Construction Machinery Co., Ltd. (Japan); Koichi Suto, Chihiro Inoue, Tohoku University ( Japan)

This report describes fundamental examination about a quicklime mixing treatment combined with carbon dioxide ventilation for the remediation process of soils polluted with volatile organic compounds (VOCs). The quicklime mixing treatment is widely applied to remove volatile pollutants in soils using 65.17 kJ/mol of the heat from the following reaction; CaO+H2O?Ca(OH)2?? To keep higher temperature and to ensure most of VOCs are volatilized, 10 % of calcium oxide is usually mixed with soils in this treatment. However, much amount of addition of calcium oxide results in higher alkaline soil pH and gives serious damage to the soil ecosystems. To solve this problem, a simultaneous ventilation of carbon dioxide during calcium oxide mixing to polluted soil was conducted. The formation of calcium carbonate according to following reaction produces 92.84 kJ/mol of the heat; Ca(OH)2+H2CO3? CaCO3+2H2O?? It is expected that the heat from the above reaction can be used for the treatment and the amount of calcium oxide addition for the treatment can be reduced. Laboratory experiments showed that more than half of calcium oxide changed to calcium carbonate when carbon dioxide ventilated to the mixed soil sample after 5% of calcium oxide and 5% of water mixed with the soil. Maximum soil temperature for this treatment increased same as that for the treatment with 10% calcium oxide. Pilot level and operational level experiments confirmed the effectiveness of the simultaneous ventilation of carbon dioxide during the quicklime mixing process.

2) 40166 – Procedure and result of decommission of R&D facility of Uranium fuel

Hirokazu Tanaka, Masao Shimizu, Ryoji Tanimoto, Kazuhiko Maekawa, Shinzo Ueta, Mitsubishi Materials Corporation (Japan); Susumu Tojo, SERNUC Corporation (Japan)

Mitsubishi Materials Corporation (MMC) had carried out nuclear engineering researches and developments such as Uranium fuel development since 1954 in MMC’s research laboratory in Omiya city in Japan. The facilities of the laboratory had been decommissioned since 1998 to 2005. The research activities were transferred to the new research laboratories established in different location. At the time of the decommission work, there was no suitable law restriction to distinguish the decontaminated waste as radioactive or not. MMC discussed and adopted the pragmatic procedure under its own responsibility with consultation of regulatory authority and technical authority outside the company. The decommissioned material could be divided into two categories. One was research apparatus and building. Another was shallow rand soil contacted to the facility. Metal wastes were mainly rose from the apparatus. Metal, concrete, and wood wastes were rose from building. At first, these wastes were decontaminated for example by blasting. Then the surface radioactivity was measured mainly by GM-survey meter and temporally by alpha-survey meter to compare the radioactivity criteria. In this case, the radioactivity criteria was decided as the natural radioactivity measured outside the facility without any influence of artificial radioactivity. Natural radioactivity varies within a certain range, therefore it was measured and treated statistically. Radioactivity of the soils contacted to the facility were measured because there was possibility that small amount contaminated liquid had flew out from the facility through the crack of pipes and fracture in basement concrete. It was very difficult to decide whether the soil was contaminated or not, because the soil naturally contains radioactive nuclides as U-238, Th-232, K-40, etc. Soil from ten areas in the city far away from the laboratory were collected and analyzed its radioactivity by the automatic alpha-beta counting system consisted of a ZnS(Ag) scintillation detector and a plastic scintillation detector. The analyzed data were treated statistically to decide the appropriate value of natural radioactivity. The law on decommission work was established later, however there is still no suitable quantitative criteria to distinguish uranium contaminated material or not. MMC’s criteria was considered as very low, which was close to natural radioactivity. In other words, the facility was decommissioned under very safely manner. The wastes generated from the decommission work are being held in storehouse build in the same area.

3) 40224 – Hydrogen Production from a PV/PEM Electrolyzer System Using a Neural-Network-Based MPPT Algorithm

Abd El-Shafy Nafeh, Electronics Research Institute (Egypt)

The electrolysis of water using a polymer electrolyte membrane (PEM) electrolyzer is a very vital and efficient method of producing hydrogen (H2). The performance of this method can be significantly improved if a photovoltaic (PV) array, with maximum-power-point (MPP) tracker, is utilized as an energy source for the electrolyzer. This paper suggests a stand-alone PV/PEM electrolyzer system to produce pure hydrogen. The paper also develops the different mathematical models for each constituent subsystem. Moreover, the paper develops the suitable maximum-power-point tracking algorithm that is based upon utilizing the neural network. This algorithm is utilized together with the action of the PI controller to improve the performance of the suggested stand-alone PV/PEM electrolyzer system through maximizing the hydrogen production rate for every instant. Finally, the suggested hydrogen production system is simulated using the Matlab/Simulink and neural network toolbox. The simulation results of the system indicate the improved relative performance of the suggested hydrogen production system compared with the traditional case of direct

-106-

Abstracts

connection between the PV array and the PEM electrolyzer.

4) 40301 – Cement based solidification / stabilization of industrial contaminated soil using various cement additives

Grega E. Voglar, RDA (Slovenia); D. Lestan, Agronomy Department, Biotechnical Faculty, University of Ljubljana, (Slovenia)

A large number of industrial activities produce wastes and contaminants that reach the soil through direct disposal, emissions, spills, leaks and other pathways. An increasing number of abandoned industrial sites (brownfields) have emerged as a result of weak environmental regulation over decades. Soil clean-up operations (remediations) of brownfields, followed by redevelopment is essential to lower the urbanization pressure on arable and other farmland (greenfields). The town of Celje in central Slovenia has a long tradition of metallurgical and chemical industries, which started in 1874 with a zinc smelter and was subsequently expanded to the production of Zn and Pb oxides and Ba salts. In 1912, the synthesis of H2SO4 and in 1970 the production of TiO2 started. Obsolete manufacturing plants have gradually been closed and replaced with new ones in a new location. The whole industrial site of “old Cinkarna” was finally demolished in 2003, leaving a brownfield area of some 170,000 m2 very close to the city centre and with highly contaminated soils, primarily with potentially toxic metals and metaloides (PTMs) and to some extend also with organic pollutants. In a laboratory study, 15% (w/w) of ordinary portland cement (OPC), black portland cement (BPC) and puculanic cement (PC) combined with various cement additives were used for solidification / stabilization (S/S) of Cd, Pb, Zn, Cu, Ni and As contaminated soils from the former industrial site. Soils formed solid monoliths with all cements. S/S effectiveness was assessed by measuring the mechanical strength of the monoliths, concentrations of metals in deionised water and TCLP (Toxicity Characteristic Leaching Procedure) soil extracts, and mass transfer of metals. Concentrations of Cd, Pb, Zn and Ni in water extracts from S/S soils generally decreased, while concentrations of As and Cu increased. Concentrations of Cd, Pb, Zn, Cu and Ni in the TCLP extracts from S/S soils were lower than from original soils, while the extractability of As from S/S soils increased. Overall, the concentration of metals in deionised water and TCLP solution, obtained after extraction of the S/S soils, was below the regulatory limits. S/S greatly reduced the mass transfer of Cd (up to 300-times), Pb (up to 53.7-times) and Zn (up to 3843-times). Mass transfer of Ni was generally also reduced, while that of Cu and As increased in some S/S soils. Based on the findings of mass-transfer mechanism analysis the predominant mechanism of release was surface wash-off of metals otherwise physically encapsulated within the cementous soil matrix.

SESSION M3: EM/PI Poster

1) 40099 – Removal of Fluorine and Boron from Groundwater Using Radiation-induced Graft Polymerization Adsorbent at Mizunami Underground Research Laboratory

Yosuke Iyatomi, Hiroyuki Hoshina, Noriaki Seko, Noboru Kasai, Yuji Ueki, Masao Tamada, JAEA (Japan)

High fluorine and boron contents in groundwater are commonly reduced using coagulation and ion-exchange treatments. As an alternative, we tested the efficiency of fluorine and boron removal from groundwater using radiationinduced graft polymerization adsorbent. The durability of the adsorbent was also determined by varying groundwater flowthrough rates and repetitive use of the adsorbent. The results indicated that it was possible for the adsorbent to remove more than 95% of boron and fluorine from the groundwater, and that the performance of the adsorbent for boron removal was better than commonly used ion-exchange resin. The adsorbent used several times was able to remove boron, indicating that the adsorbent can be used for efficient boron removal.

2) 40184 – The Optimized Risk Management of the Waste from NORM and Nuclear Industries - How to Harmonize Risk from Various Sources

Yoko Fujikawa, Kyoto University Research Reactor Institute (Japan); Michikuni Shimo, Fujita Health University (Japan);

Hidenori Yonehara, National Instuitute of Radiological Sciences (Japan); Tadashi Tujimoto, Electron Science Institute (Japan)

We compared the existing regulation on management of radioactive and non-radioactive wastes with the ideal legislation procedures for protection of environment. The comparison revealed the necessity of risk-based regulation, consideration for ethics and cost-effectiveness of the regulation, and optimal usage of regulation resources. In order to assess the cost-effectiveness of several different waste disposal options, the concept of disposal cost per unit radiotoxicity (mSv or m3) in waste (CPR hereafter) was introduced and calculated. The results revealed that current disposal option of high level radioactive waste (underground burial) was more cost-effective than that of TENORM and asbestos containing waste.

-107-

Abstracts

3) 40205 – Exposure Dose Evaluation of Worker at Radioactive Waste Incineration Facility on KEARI

Sang Kyu Park, Jong Seon Jeon, Younhwa Kim, Jaemin Lee, NESYS.CO., LTD. (Korea Rep.); Ki Won Lee, KAERI, (Korea Rep.)

In this study, we evaluate the exposure dose of worker by operating at radioactive waste incineration facility that is in KAERI(Korea Atomic Energy Research Institute) for safety analysis. The incineration facility that is areas are 570 ?, annual treatment quantities are 40,000 kg is separated with radiation zone and non-radiation area to installed extra equipment for incineration process. The incineration facility consist of preparing system, incineration processing system, exhaust gas treatment system, ash treatment system and controlled with measurement. Radioactive waste incineration facility during normal operation, worker is exposured radiation by the exposure route of external and internal. They are exposured by external radiation that was effected by treated waste material and characteristic. In case of internal exposure, it is primary fact that is inhalation of contaminated air. The evaluation assumption is the next: One is that internal exposure is concentration of radioactive material release basis and work is 200 days per year(8 hr per day) because of the incineration facility is operated and managed in critical regulation under maximum acceptance air contamination. Result of evaluation, maximum exposure dose is 3.07 mSy/y and internal exposure dose is 2.5 mSy/yr according as selecting the treated radioacitve waste. The exposure dose of assessment result is lower the basis of domestic nuclear law and regulation limit in IAEA.

4) 40209 – Scenario Development for Safety Assessment of Waste Repository for Feasibility Study on Transmutation of Spent Nuclear Fuel into LILW using PEACER

Sung-yeop Kim, Kun Jai Lee, KAIST (Korea Rep.)

Safety assessment of radioactive waste repository is necessary for feasibility study on transmutation of spent nuclear fuel into LILW(Low and Intermediate Level Waste) using transmutation reactor PEACER(Proliferation-resistant, Environmental-friendly, Accident-tolerant, Continuable-energy and Economical Reactor). PEACER is a conceptual liquid metal fast reactor using Pb-Bi as a coolant. Scenario development is important to the safety assessment for several reasons. Scenarios provide the context in which safety assessments are performed. Scenarios influence model development and data collection efforts. They have become a very important aspect of confidence building for the post-closure safety assessment. In this study, condition of medium depth disposal about 100-200m in granite is considered. Waste from PWR and PEACER considering DF(Decontamination Factor) is disposed in this condition. Scenario for these concepts are established by screening FEP(Feature, Event, Process) database and benchmarking reference safety assessment report.

-108-

Kapila FERNANDO Australian Nuclear Science and Technology Organisation Australia Lynn TAN Australian Nuclear Science and Technology Organisation Australia Robin George HEARD IAEA Austria Irena MELE IAEA Austria Roman BEYERKNECHT Nuclear Engineering Seibersdorf GmbH Austria Wolfgang STUDECKER Nuclear Engineering Seibersdorf GmbH Austria Jan DECKERS Belgoprocess Belgium Henri VANBRABANT Belgoprocess Belgium Paul LUYCX Castor Consulting Belgium Lou AREIAS SCK.CEN Belgium Alain VAN COTTHEM Technum-Tractebel Engineering Nv Belgium Joseph, Ghislain BOUCAU Westinghouse Belgium Sheila M. BROOKS Atomic Energy of Canada Limited Canada Miklos GARAMSZEGHY Nuclear Waste Management Organization Canada Peter BATTEN PBC ASIA China Josef PODLAHA Nuclear Research Institute Rez plc Czech Rep. Frantisek SVITAK Nuclear Research Institute Rez plc Czech Rep. Karel SVOBODA Nuclear Research Institute Rez plc Czech Rep. Eduard Josef HANSLÍK T. G. Masaryk Water Research Institute Czech Rep. Abd El-Shafy NAFEH Electronics Research Institute Egypt Gérald OUZOUNIAN ANDRA France Jean-Guy NOKHAMZON Commissariat à l'énergie atomique France Christine BRUN-YABA Institute de Radioprotection et Sûreté Nucléaire France Claudio PESCATORE OECD/NEA France Florence GASSOT-GUILBERT SGN AREVA GROUP France Peter PILLOKAT AREVA NP GmbH Germany Christoph Michael STIEPANI AREVA NP GmbH Germany Johannes FACHINGER Furnaces Nuclear Applications Grenoble Germany Klaus BÜTTNER NUKEM Technologies GmbH Germany Markus Adam HARTUNG NUKEM Technologies GmbH Germany Jörg WÖRNER RD-Hanau Germany Katharina AYMANNS Research Centre Juelich Germany Natalia GIRKE Research Centre Juelich Germany Hans-Juergen STEINMETZ Research Centre Juelich Germany Yasuo TOMISHIMA AIST Japan Shizue FURUKAWA Central Research Institute of Electric Power Industry Japan Motoi KAWANISHI Central Research Institute of Electric Power Industry Japan Koji NAGANO Central Research Institute of Electric Power Industry Japan Yukihisa TANAKA Central Research Institute of Electric Power Industry Japan Masaki TSUKAMOTO Central Research Institute of Electric Power Industry Japan Yoshifusa FUKUOKA Chubu Electric Power Company Japan Motonori NAKAGAMI Chubu Electric Power Company Japan Masato WATANABE Chubu Electric Power Company Japan Yoshio KIMURA Chuden CTI Co.,LTD. Japan Shinichi HOSOYA DIA Consultants Co., Ltd. Japan Miyoshi YOSHIMURA DIA Consultants Co., Ltd. Japan Haruyoshi SHIMAZOE Ebara Industrial Cleaning Co., Ltd. Japan Chikao Chuck MIYAMOTO EPRI International Japan Masaaki NAKANO Fuji Electric Holdings Co., Ltd. Japan Gen-Ichi KATAGIRI Fuji Electric Systems Co., Ltd. Japan Kiyoshi AMEMIYA Hazama Corporation Japan Yuki NAKAGAWA Hitachi Construction Machinery Co., Ltd. Japan Motonari HARAGUCHI Hitachi, Ltd. Power Systems Company Japan Hirokazu MINATO Hitachi-GE Nuclear Energy Ltd. Japan Susumu KAWAKAMI IHI Corporation Japan Hirofumi TSUKADA Institute for Environmental Sciences Japan Takeshi ISHIKURA Institute of Applied Energy Japan

List of Registrants

-109-

Hironobu ABE Japan Atomic Energy Agency Japan Ippei AMAMOTO Japan Atomic Energy Agency Japan Kenji AMANO Japan Atomic Energy Agency Japan Daisuke AOSAI Japan Atomic Energy Agency Japan Shiho ASAI Japan Atomic Energy Agency Japan Shuji DAIMARU Japan Atomic Energy Agency Japan Takahiro HANAMURO Japan Atomic Energy Agency Japan Kazumasa HIOKI Japan Atomic Energy Agency Japan Sohei IKEGAMI Japan Atomic Energy Agency Japan Ken-Ichiro ISHIMORI Japan Atomic Energy Agency Japan Yuu ISHIMORI Japan Atomic Energy Agency Japan Yosuke IYATOMI Japan Atomic Energy Agency Japan Gento KAMEI Japan Atomic Energy Agency Japan Fumihiko KANAYAMA Japan Atomic Energy Agency Japan Akira KITAMURA Japan Atomic Energy Agency Japan Takanori KUNIMARU Japan Atomic Energy Agency Japan Kensuke KURAHASHI Japan Atomic Energy Agency Japan Toshikatsu MAEDA Japan Atomic Energy Agency Japan Keisuke MAEKAWA Japan Atomic Energy Agency Japan Hitoshi MAKINO Japan Atomic Energy Agency Japan Toshiyuki MATSUOKA Japan Atomic Energy Agency Japan Yoshihiro MEGURO Japan Atomic Energy Agency Japan Kaname MIYAHARA Japan Atomic Energy Agency Japan Takashi MIZUNO Japan Atomic Energy Agency Japan Shinichi NAKAYAMA Japan Atomic Energy Agency Japan Masashi NAKAYAMA Japan Atomic Energy Agency Japan Tadafumi NIIZATO Japan Atomic Energy Agency Japan Kunio OTA Japan Atomic Energy Agency Japan Sidik PERMANA Japan Atomic Energy Agency Japan Hiromitsu SAEGUSA Japan Atomic Energy Agency Japan Hiroshi SAITO Japan Atomic Energy Agency Japan Hiroyuki SANADA Japan Atomic Energy Agency Japan Eiji SASAO Japan Atomic Energy Agency Japan Haruo SATO Japan Atomic Energy Agency Japan Yuji SHIBAHARA Japan Atomic Energy Agency Japan Seiji TAKEDA Japan Atomic Energy Agency Japan Shinji TAKEUCHI Japan Atomic Energy Agency Japan Tadao TANAKA Japan Atomic Energy Agency Japan Tetsuya TOKIWA Japan Atomic Energy Agency Japan Hiroyuki UMEKI Japan Atomic Energy Agency Japan Isao YAMAGISHI Japan Atomic Energy Agency Japan Yuhei YAMAMOTO Japan Atomic Energy Agency Japan Satoshi YANAGIHARA Japan Atomic Energy Agency Japan Hideharu YOKOTA Japan Atomic Energy Agency Japan Naoki ZAIMA Japan Atomic Energy Agency Japan Tomohisa ZAITSU Japan Atomic Energy Agency Japan Tatsujiro SUZUKI Japan Atomic Energy Commission Japan Satoshi KARIGOME Japan Atomic Power Company Japan Ken-Ichi TANAKA Japan Atomic Power Company Japan Kazuhisa YAMAGUCHI Japan Atomic Power Company Japan Kazuhiko YAMAMOTO Japan Atomic Power Company Japan Keizaburou YOSHINO Japan Atomic Power Company Japan Hiroomi AOKI Japan Nuclear Energy Safety Organization Japan Yukihiro IGUCHI Japan Nuclear Energy Safety Organization Japan Masami KATO Japan Nuclear Energy Safety Organization Japan Yusuke MASUDA Japan Nuclear Energy Safety Organization Japan Shigeyuki SAITO Japan Nuclear Energy Safety Organization Japan Mamoru KUMAGAI Japan Nuclear Fuel Limited Japan Masataka KAMITSUMA Japan NUS Co. Ltd. Japan Keiji KUSAMA Japan Radioisotope Association Japan

-110-

Takao IKEDA JGC Corporation Japan Atsushi MUKUNOKI JGC Corporation Japan Kunihiro NAKAI JGC Corporation Japan Kiyoshi OYAMADA JGC Corporation Japan Hajime TAKAO JGC Corporation Japan Ichizo KOBAYASHI Kajima Corporation Japan Toshihiko HIGASHI Kansai Electric Power Company Japan Takashi NISHIO Kobe Steel, Ltd. Japan Ryutaro WADA Kobe Steel, Ltd. Japan Yoko FUJIKAWA Kyoto University Japan Takumi KUBOTA Kyoto University Japan Kenji KOTOH Kyushu University Japan Takeshi NAKAMURA Kyushu University Japan Yuzo YAMASHITA Kyushu University Japan Keita AKEHASHI Marubeni Utility Services, Ltd. Japan Tadahiro KATSUTA Meiji University Japan Hirokuni ITO Ministry of Education, Culture, Sports, Science and Technology Japan Akira SAKASHITA Mitsubishi Heavy Industries, INC Japan Toshiyuki NAKAZAWA Mitsubishi Materials Corporation Japan Hirokazu TANAKA Mitsubishi Materials Corporation Japan Shinzo UETA Mitsubishi Materials Corporation Japan Makoto KAWAMURA Mitsubishi Materials Techno Corporation Japan Keiko TAGAMI National Institute of Radiological Sciences Japan Shigeo UCHIDA National Institute of Radiological Sciences Japan Yoji KUSAKA NGK Insulators, Ltd. Japan Hitoshi OHATA NGK Insulators, Ltd. Japan Katsutoshi TORITA NGK Insulators, Ltd. Japan Kojuro YAMAMOTO NGK Insulators, Ltd. Japan Yoshihiko HORIKAWA Nuclear Engineering, Ltd. Japan Kazuma MIZUKOSHI Nuclear Engineering, Ltd. Japan Satoru TAKEDA Nuclear Services Company Japan Shigeki AKAMURA Nuclear Waste Management Organization of Japan Japan Takeshi EBASHI Nuclear Waste Management Organization of Japan Japan Kiyoshi FUJISAKI Nuclear Waste Management Organization of Japan Japan Takahiro GOTO Nuclear Waste Management Organization of Japan Japan Keisuke ISHIDA Nuclear Waste Management Organization of Japan Japan Kenichi KAKU Nuclear Waste Management Organization of Japan Japan Tomio KAWATA Nuclear Waste Management Organization of Japan Japan Kazumi KITAYAMA Nuclear Waste Management Organization of Japan Japan Satoru SUZUKI Nuclear Waste Management Organization of Japan Japan Hiroyuki TSUCHI Nuclear Waste Management Organization of Japan Japan Hiroyoshi UEDA Nuclear Waste Management Organization of Japan Japan Hideki KAWAMURA Obayashi Corporation Japan Tatsuya TANAKA Obayashi Corporation Japan Shuichi YAMAMOTO Obayashi Corporation Japan Hidekazu ASANO Radioactive Waste Management Funding and Research Center Japan Naoki FUJII Radioactive Waste Management Funding and Research Center Japan Takahiro NAKAJIMA Radioactive Waste Management Funding and Research Center Japan Ario NAKAMURA Radioactive Waste Management Funding and Research Center Japan Hitoshi NAKASHIMA Radioactive Waste Management Funding and Research Center Japan Hiromi TANABE Radioactive Waste Management Funding and Research Center Japan Takahiro YOSHIDA Radioactive Waste Management Funding and Research Center Japan Mito AKIYOSHI Senshu University Japan Satohito TOGURI Shimizu Corporation Japan Yuji IJIRI Taisei Corporation Japan Hitoshi MIMURA Tohoku University Japan Keisuke SHIRAI Tohoku University Japan Yu AOKI Tokyo Electric Power Company Japan Satoru KANEKO Tokyo Electric Power Company Japan Kazuhiro TAKEI Tokyo Electric Power Company Japan

-111-

Hiroshi SHIRATSUCHI Tokyo Electric Power Services Co., Ltd. Japan Masanori ARITOMI Tokyo Institute of Technology Japan Hiroshige KIKURA Tokyo Institute of Technology Japan Takatoshi ASADA Toshiba Corporation Japan Susumu NAITO Toshiba Corporation Japan Masamichi OBATA Toshiba Corporation Japan Yuki YAHIRO Toshiba Corporation Japan Hitoshi SAKAI Toshiba Corporation Power Systems Company Japan Yasuhiro MOROMOTO Transnuclear Tokyo Japan Koji OKAMATO University of Tokyo Japan Satoru TANAKA University of Tokyo Japan Jongseon JEON Enesys. Co., Korea Rep. Juyoul KIM FNC Technology Co., Ltd. Korea Rep. Sukhoon KIM FNC Technology Co., Ltd. Korea Rep. Sung-Yeop KIM Korea Advanced Institute of Science and Technology Korea Rep. Dong-Keun CHO Korea Atomic Energy Research Institute Korea Rep. Young-Yong JI Korea Atomic Energy Research Institute Korea Rep. Youn Myoung KEE Korea Atomic Energy Research Institute Korea Rep. Hee Reyoung KIM Korea Atomic Energy Research Institute Korea Rep. Jin-Seop KIM Korea Atomic Energy Research Institute Korea Rep. Bum-Kyoung SEO Korea Atomic Energy Research Institute Korea Rep. Gwangmin SUN Korea Atomic Energy Research Institute Korea Rep. Donghyeun HWANG Korea Hydro and Nuclear Power Company Korea Rep. Hyun Jun JO Korea Hydro and Nuclear Power Company Korea Rep. Yun Seog NAM Korea Nuclear Fuel Korea Rep. Sun-Joung LEE Korea Radioactive Waste Management Corporation Korea Rep. Jin-Beak PARK Korea Radioactive Waste Management Corporation Korea Rep. Ho-Taek YOON Korea Radioactive Waste Management Corporation Korea Rep. Rizwan AHMED Kyung Hee University Korea Rep. Muzna ASSI Lebanese Atomic Energy Commission (LAEC) Lebanon Renate DE VOS Nuclear Research and consultancy Group Netherlands Tamara ZHUNUSSOVA Norwegian Radiation Protection Authority Norway Mohamed AMR Qatar University Qatar Gheorghe BARARIU R.A.A.N. - S.I.T.O.N. Romania Mikhail BOGOD ECOMET-S Russia Alexander GELBUTOVSKIY ECOMET-S Russia Alexander KOLPAKOV ECOMET-S Russia Alexey VOTYAKOV FSUE RosRAO Russia Sevastyanov STANISLAV Gazprombank Russia Dmitry V. MARININ Institute of Chemistry FEDRAS Russia Alexander KOBELEV Scientific and Industrial Association RADON Russia Vladimír DANIŠKA DECOM, a.s. Slovakia Peter BEZÁK DECONTA, a.s. Slovakia František ONDRA DECONTA, a.s. Slovakia Marek VAŠKO DECONTA, a.s. Slovakia Helena DANISKOVA Grammar School of St. Michael the Archangel in Piestany Slovakia Tomáš HRNČÍŘ Slovak University of Technology in Bratislava Slovakia Matej ZACHAR Slovak University of Technology in Bratislava Slovakia Vladimír NEČAS Slovak University of Technology in Bratislava Slovakia Michal PÁNIK Slovak University of Technology in Bratislava Slovakia Kamil KRAVÁRIK VUJE, Inc. Slovakia Grega E. VOGLAR Regional Development Agency Celje /University of Ljubljana Slovenia Gerhardus R LIEBENBERG South African Nuclear Energy Corporation (Necsa) South Africa Lara DURO Amphos 21 Spain Johan ANDERSSON JA Streamflow AB Sweden Gunnar HEDIN Westinghouse Sweden Ian Gerard MCKINLEY McKinley Consulting Switzerland Irina GAUS Nagra Switzerland Sven-Peter TEODORI Nagra Switzerland Stratis VOMVORIS Nagra Switzerland

-112-

Dorothea SCHUMANN Paul Scherrer Institute Switzerland Walter M. HEEP ZWILAG Interim Storage Switzerland Chih Tien LIU Atomic Energy Council Taiwan Chun-Ping JEN National Chung Cheng University Taiwan Fong-In CHOU National Tsing Hua University Taiwan Chia Chin LI National Tsing Hua University Taiwan Oleksandr NOVIKOV Chernobyl NPP Ukraine Ellie SCOURSE Atkins UK Ricahrd Peter SHAW British Geological Survey UK Alan SIMPSON Pajarito Scientific Corporation UK Henry O'GRADY Parsons Brinckerhoff UK W Mark NUTT Argonne National Laboratory USA Anibal L. TABOAS Argonne National Laboratory USA Hanchung TSAI Argonne National Laboratory USA David Walter JAMES DW James Consulting, LLC USA Sean Paul BUSHART Electric Power Research Institute USA Karen KIM Electric Power Research Institute USA Kenzi KARASAKI Lawrence Berkeley National Laboratory USA Chin-Fu TSANG Lawrence Berkeley National Laboratory USA Frank COCINA Los Alamos National Laboratory USA Michael Edward COURNOYER Los Alamos National Laboratory USA John ZARLING Los Alamos National Laboratory USA Yi SU Mississippi State University USA Geoffrey John PETER Oregon Institute of Technology - Portland Center USA Chase Collins BOVAIRD Pacific Northwest National Laboratory USA Danielle JANSIK Pacific Northwest National Laboratory USA Mark B TRIPLETT Pacific Northwest National Laboratory USA Dawn M. WELLMAN Pacific Northwest National Laboratory USA William LAWLESS Paine College USA Gary Joseph HANUS Phoenix Solutions Co USA Joseph LEGARE S.M. Stoller Corporation USA Eric OLSON S.M. Stoller Corporation USA Michael SERRATO Savannah River National Laboratory USA Corey Adam MYERS Studsvik, Inc. USA Dae Y. CHUNG U.S. Department of Energy USA Kurt GERDES U.S. Department of Energy USA Gary L. SMITH U.S. Department of Energy USA Andrew SZILAGYI U.S. Department of Energy USA Catherine HANEY U.S. Nuclear Regulatory Commission USA Lawrence Edward KOKAJKO U.S. Nuclear Regulatory Commission USA Shawn Rochelle SMITH U.S. Nuclear Regulatory Commission USA Joonhong AHN University of California, Berkeley USA John Calvin WALTON University of Texas at El Paso USA Ronald MORRIS Westinghouse Electric Company USA Wallace M. MAYS WM Mining Company LLC USA John Phil HAYFIELD WorleyParsons Polestar USA

-113-

Conference General Chair Satoru Tanaka, University of Tokyo

Technical Program Chair Masanori Aritomi, Tokyo Institute of Technology

Technical Program Co-Chair Anibal L. Taboas, Argonne National Laboratory

Technical Program Vice-Chair Kaname Miyahara, Japan Atomic Energy Agency

Administrative Chair Kazuhiro Takei, Tokyo Electric Power Company

Conference Secretary Hiroyoshi Ueda, Nuclear Waste Management Organization of Japan

Exhibition Secretary Kiyoshi Fujisaki, Nuclear Waste Management Organization of Japan

ASME Project Directors John Bendo, American Society of Mechanical Engineers

Vince Dilworth, American Society of Mechanical Engineers

Track Chairs/Co-Chairs

Low/Intermediate-Level Radioactive Waste Management:

Track Chair Kunihiro Nakai, JGC Corporation

Track Co-Chair Miklos Garamszeghy, Nuclear Waste Management Organization

Spent Fuel, Fissile material, Transuranic and High-Level Radioactive Waste Management:

Track Chair Kaname Miyahara, Japan Atomic Energy Agency

Track Co-Chair Stratis Vomvoris, Nagra

Facility Decontamination and Decommissioning:

Track Chair Takeshi Ishikura, Institute of Applied Energy

Track Co-Chair Toshihiko Higashi, Kansai Electric Power Company

Environmental Remediation:

Track Chair Tomohisa Zaitsu, Japan Atomic Energy Agency

Track Co-Chair Yuu Ishimori, Japan Atomic Energy Agency

Environmental Management / Public Involvement / Crosscutting Issues:

Track Chair Masaki Tsukamoto, Central Research Institute of Electric Power Industry

Global Partnership and Multi-National Programs:

Track Chair Hiromi Tanabe, Radioactive Waste Management Funding and Research Center

ICEM2010 Conference Organizers

-114-

Nuclear Remediation

Documents

Transcript of Nuclear Remediation