Perspective on Long-Term Safety Research

for New Reactors and New Technologies

Masashi HIRANONuclear Safety Research Center

Japan Atomic Energy Agency (JAEA)

Presented atU.S. NRC Regulatory Information Meeting,

March, 2010

NSRR ROSA/LSTF NUCEF JMTR

1Outline

Nuclear Energy Policy in JapanAdvanced LWRs: “Next Generation LWRs”

– Development by Joint Project of METI, Utilities and Three Major Vendors

– R&D for Seismic Isolation System

FBR– FR Cycle Technology Development in Japan– Innovative Technologies for Safety Enhancement– R&D for Prevention of Re-criticality during CDA

HTGR– Development toward “Hydrogen Society”– Technical Issues and R&D Plan

Challenges for Safety Research

2Nuclear Energy Policy in Japan

Framework for Nuclear Energy Policy (AEC, 2005)Maintain nuclear power’s share 30-40% or more even beyond 2030.Prepare advanced LWRs for replacement of existing NPPs at around 2030.Large LWRs are a prime candidate from scale merit.Strive for commercial use of FBRs from around 2050 based on “FS on Commercial FBR Cycle” and operation of “Prototype FBR, Monju.”

1960 1970 1980 1990 2000 2010 2020 2030 2040 2050 2060 2070 2080 2090 2100

70

60

50

40

30

20

10

0Elec

tric

ity G

ener

atio

n Ca

paci

ty (

GW

e)

“Next Generation LWRs”

(1,700 – 1,800 MWe)

Existing LWRs(40 year-life)

Existing LWRs(Extend to 60 year-life)

Commercial FBRsDemo - JSFR (500-750MWe)

Experimental FBRMonju(280MWe)

Prototype FBRJoyo(140MWt)

LWRs to be replaced(60 year-life)

as of end of 2008PWR BWR APWR ABWR Total

23 26 0 4 53

1 1 5 8 12

In OperationUnder construction or preparation 153 10

3

In April 2008, METI with industries launched development of “Next Generation LWRs” aiming at coping with the need to replace existing NPPs in Japan after 2030. They are also expected to be a global standard design. The Institute of Applied Energy (IAE) undertakes overallmanagement of this development program.

Six Core Concepts:1) Fuel with uranium enrichment above 5%

⇒ High capacity factor, high burnup2) Seismic isolation technologies3) Plant life of 80 years and significant reduction

of occupational dose4) Significant shortening of construction period5) Optimized combination of passive and

active systems6) Innovative digital technologies

One PWR and one BWR1,700 – 1,800 MWe class

Roadmap

Basic design

2016 - 20302006 - 2007

Feasibility study Establish plant

design concept

Site specific designSafety review

Construction

2008 - 2010 - 2015

Conceptual design

Start full-scale developments

Determine basic specifications

Complete basic design

Start commercial operation

Planning of R&DLong-term testingR&D to realize core concepts

“Next Generation LWRs”Development by Joint Project of METI, Utilities and Three Major Vendors

4“Next Generation LWRs”R&D for Seismic Isolation System

BackgroundSeismic risk is a dominant contributor in Japan.Regulatory Guide for Reviewing Seismic Design of NPPs (amended in Sep. 2006) requires evaluation of “residual risk”. Plant specific seismic PSA is on-going.

GoalStandardized structural design independent from site specific conditions enabled by application of seismic isolation system

R&D PlanDevelopment of large-scale seismic isolation system

Full-scale / Large-scale model tests in Phase II (FY 2011-2015)Characterization/Demonstration of seismic isolation systemPiping integrity at the interface between seismically isolated building and non-isolated ones, etc.

Development of evaluation methods and standards

Bolt HolesLead plug

damper Elastomer

Inner steel plates

Elastomer cover

Flange

Seismic isolation system

5

20052015

2015

2025

～2050

Experimental FR “Joyo”

Commercial Introduction of FR Cycle Technologies

Detailed Design & Construction

Start of Operation of Demonstration FR & its Fuel Cycle Facility

2010 (JFY)

Fast Reactor Cycle Technology Development Project (FaCTFaCT）

R&D of Innovative Technologies

Preliminary Design of Commercial & Demonstration FR Cycle Facilities

Commercialized Commercialized FR CycleFR Cycle

2015

R&D at “Monju”

Demonstrating its Reliability as a Nuclear Power PlantEstablish Sodium Handling Tech.

Conceptual Design of Commercial and Demonstration FR Cycle Facilities, with Providing Necessary R&D Programs

Feasibility Study

Prototype FBR “Monju “

C&R

Rev

iew

& B

asic

Pol

icy

by M

EXT

&A

EC

◆International Cooperation（GNEP, GEN-IV, INPRO etc.）◆Cooperation with Various Organizations

Identify Most Promising Concept

Decision on Innovative Tech. Decision on Innovative Tech. (2010)(2010) Approved Confirmation (2015)Approved Confirmation (2015)

(JFY 1999-2005) Validation of Economy & Reliability

Toward Commercialization of FBRFR Cycle Technology Development in Japan

6Toward Commercialization of FBRInnovative Technologies for Safety Enhancement

Passive reactor shutdown and decay heat removal systems (DHRS)

Passive DHRS by natural circulation

Design measures to avoidre-criticality and achieve stable IVR during CDA

In-pile test using IGR in of Kazakhstan

RV

IHX/PUMP

PUMP

DRACS

SG withdouble-walledtubes

PRACSx2DRACS: Direct Reactor Aux. Cooling SystemPRACS: Primary Reactor Aux. Cooling System

Drop

Retention

Ret

enti

on F

orce

(N

)

1600

700

600 640 680Temperature (℃)

Use of change in magnetic characteristics at Curie point

Retention Drop

Temperature-sensitive alloy(Ni-Co-Fe)

MagnetCoil

Iron-core

Coupling face

Self Actuated Shutdown System (SASS)

7Toward Commercialization of FBRR&D for Prevention of Re-criticality during CDA

Modified FAIDUS:Fuel Assembly with Inner Duct Structure

Internal duct

Calculated results of ULOF with modified FAIDUS

-5 0 5 10 15Time after power burst (s)

1000

100

10

1

0.1

0.01

10

0

-10

-20

-30

-40

1.0

0.8

0.6

0.4

0.2

0.0

Rel

ativ

e po

wer

(-) R

eactivity ($)

Fuel m

ass fraction (-)

Power

Reactivity

Fuel mass0.96$148P0

SAS4A SIMMER-III

EAGLE in-pile test with IGR

Pressurehousing

Test fuel pin bundle

Neutron detectors for fuel relocation

Steel duct

Relocated fuelreceiver

Fuel: 8 kg (enriched UO2)Na : App. 15 kg

Stable IVR with avoiding re-criticality

Relocation of molten fuel

Basic conceptDischarge of

molten fuel by design measure

8HTGR Development toward “Hydrogen Society”

Roadmap for Innovative R&D of Nuclear Energy against Global Warming (AEC, 2008)

One of the candidates is:HTGR technology and innovative hydrogen production technology by thermochemical water splittingIt is expected to propose a prototype commercial system around 2020.

“Safety Demonstration Tests” at HTTR planned to be conducted as OECD/NEA project will demonstrate its intrinsic safety characteristics during total loss-of-forced cooling under ATWS condition.

Hydrogen production

by IS process

Reactor technology using HTTR

Pilot-scale test

HTGR technology

HTTRHTTR

2010 203020202015 2040

Commercial system around 2030

2050

HTGR-IS

Commercialization Commercialization 

Basic R&D

Engineeringscale R&D

R&D for demonstrationJAEA’s R&D

IS process: Thermochemical water splitting by Iodine-Sulfur cycle

9HTGR Development toward “Hydrogen Society”Technical Issues and R&D Plan

Irradiation tests with HTTR, etc.RIA tests with NSRR

Fuel behavior under transient / accident conditions

Fuel (Coated fuel particles / Prismatic block type)

Development of turbo-machinery and high temperature isolation technologies (high temperature valve, hot gas duct, etc.)Development of safety design guide for coupling

Coupling technologies between hydrogen production system and HTGR

Process heat application

Pilot plant test of the thermochemical IS processImprovement of process efficiency

Hydrogen production

Verification through “Safety Demonstration Tests”(OECD/NEA HTTR project being proposed)Development of PSA methods and application

Upgrade of methods and models

Safety assessment

Deployment of carbon/carbon composite, etc.Development of structural integrity evaluation methodIrradiation tests with HTTR, JOYO, etc.

Deployment of new materials with longer life

Hightemperature materials

HTTR operation data analysisFP transport

JAEA’s R&D PlanKey Issues

0.92

mm

10

Under the environment of decreasing priority / budget for R&D on existing LWRs worldwide, major challenge is “declining infrastructure (facilities and expertise)” :

We need to maintain infrastructure while resolving current safety issues on such as:

Life extension, burnup extension, power uprate,Use of risk information, PSA for external events, andRadioactive waste disposal.

Lincensing and operation of new reactors would benefit from the availability of such infrastructure:

ROSA for system integral tests, RIA tests with NSRR for new fuel, NUCEF for criticality, JMTR for irradiation of new fuel / materials, andExperience and knowledge on safety review and analysis.

Challenges for Safety Research- Short Term and Mid Term -

ROSA : Rig of Safety AssessmentNSRR : Nuclear Safety Research ReactorNUCEF : Nuclear Fuel Cycle Safety Engineering Research FacilityJMTR : Japan Material Testing Reactor

11

Public expectations for new reactors would be directed toward “lower risks”.

PSA would play a more prominent and fundamental role in the licensing process to deal with diverse design measures to cope with beyond-DBA scenarios.Understanding of phenomena gained from R&D would play a vital role for high level of confidence on PSA.

Seismic isolation system,Measures to avoid re-criticality during CDA, Coupling between hydrogen production system and HTGR, etc.

For efficient use of available resources of both regulatory and industry sides, “regulatory – industry cooperation in R&D” with due consideration of regulatory independence will be a key. International role- and cost-sharing with industry participationwould become vital.

Challenges for Safety Research- Long Term -

12HTTR:High Temperature Engineering Test Reactor:

Major specificationMajor specificationThermal power 30 MWFuel Coated fuel particle /

Prismatic block typeCore material GraphiteCoolant HeliumInlet temperature 395 °COutlet temperature 950 °C (Max.)Pressure 4 MPa

Achievement of reactor outlet coolant temp. of 950℃: April, 2004

Containment vessel

Reactor pressure vessel

Intermediate heat

Exchanger(IHX)

Hot-gas duct

HTTR Graphite-moderated and helium-cooled HTGR

First criticality : 1998Full power operation : 2001

Coated fuel particle

(0.92mm,Dia.)

Low density PyC

Fuel kernelHigh density

PyC

Fuel compact

SiC

8mm

26mm

39mm

Appendix

13Toward Commercialization of FBRR&D needs

○ Core safety

○ Seismic reliability

⑬ Seismic reliability in core assemblies

⑫ Re-criticality free core

⑪ Passive shutdown and decay heat removal

Economic CompetitivenessEconomic Competitiveness Enhanced reliabilityEnhanced reliability

Enhanced safetyEnhanced safety

○ Sodium technology

⑨ Higher reliable SG with double-walled tubes

⑩ Higher inspection ability inside of sodium boundary

⑧ Sodium leak tightness withdouble-walled piping

○ Reduction of Mass & Volume

○ Long operation by high burn-up fuel

① Shortened piping with high chromium steel

④ Compact reactor vessel

⑦ Advanced fuel material

② 2 loop cooling system

⑤ Simplified fuel handling system

⑥ CV with steel plates andreinforced concrete building

③ Integrated pump-IHX component

Secondary pump

Integrated pump‐IHX

Reactor vessel

⑭ Plant design study(Demo-FR/Commercial-FR)

⑮ Large-scale sodium tests

SG

Appendix