Geological Disposal of High-Level Radioactive …chiba/NDS2018/Presentation/S702...Reactor Fuel...

1

Symposium on Nuclear DataTokyoTech, Nov. 29-30, 2018

Geological Disposal of High-Level Radioactive Waste:

Long-term Safety and Reduction of Environmental Impact

Hidekazu ASANO

Radioactive Waste Management Funding and Research Center（RWMC)

2

Key words

Geological Disposal

Long-term Safety

Radionuclide Migration

Radiation Dose vs. Radiological Toxicity

Environmental Impact

Viewpoints

Data needs from new fields

Topics/GD1. Concept “Isolation & Containment”2. Post closure & long-term safety 3. Load reduction4. Integrated approach5. Data required 6. Summary & Needs

3

High-level

Liquide waste

Second progress report on research and development for TRU waste Disposal in Japan – Repository design, safety assessment and means of

implementation in the generic phase-, JAEA-Review 2007-010, FEPC TRU-TR2-2007-01, March 2007

Processes and generation of radioactive waste at reprocessing plant

Se-79Zr-93Tc-99Pd-107Sn-126Cs-135

I-129/AgI

C-14

TRU waste

TRU waste HLW: vitrified waste

＊

＊

1. Concept “Isolation & Containment”(1/9)

4

H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan, Second Progress Report on Research and Development for the

Geological Disposal of HLW in Japan,

Japan Nuclear Cycle Development Institute (JNC), 2000.

- TN1410 2000-004, Supporting Report 3, Safety Assessment of the Geological Disposal System

Process conditions for vitrified waste(Glass)

＊

＊


Radioactivity of vitrified waste vs. Time after disposal*

5

* H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan, Second

Progress Report on Research and Development for the Geological Disposal of HLW in Japan, Japan

Nuclear Cycle Development Institute (JNC), 2000.

TN1410 2000-003, Supporting Report 2, Repository Design and Engineering Technology

Systems for nuclide migration


6

* H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan, Second Progress Report on Research

and Development for the Geological Disposal of HLW in Japan, Japan Nuclear Cycle Development Institute (JNC), 2000.



7

HLW(Vitrified waste)Result : safety assessment/long-term & post-closure safetyRadiation dose caused by groundwater usage by a person living on the surface,Estimation modeled as 40,000 - vitrified wastes disposed at 1,000m below the surface, and contact with groundwater after 1,000 years.

Figure from NUMO-TR-04-01, 2004

Original Data from H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan, Second Progress Report on Research

and Development for the Geological Disposal of HLW in Japan, Japan Nuclear Cycle Development Institute (JNC), 2000.

- TN1410 2000-001, Project Overview Report

- TN1410 2000-004, Supporting Report 3, Safety Assessment of the Geological Disposal System


8

Barrier performance- Containment capability of each barrier component by multi-barrier system

Np-237 Cs-135

Nuclide inventory in the EBS components and Geosphere

Np*, low solubility & large distribution coefficient- 107 years, approx. 1 % of Np-237 as of the initial

inventory moves to geosphere from EBS

- approx. 1/105 of the original Np-237 reaches to the biosphere

(*T1/2 =2.14×106 years)

Cs**, highly soluble & less sorptive- approx. 95% of the initial inventory of Cs-135 is

isolated at the area of EBS & geosphere

(**T1/2 =2.30×106 years)


9

1. Isolation

隔離

2. Containment

閉じ込め

c. Buffer material/緩衝材b. Overpack/オーバーパックa. Vitrified waste/ガラス固化体(Bq/kg)

d. Geological formation/host rock 地質/岩盤

e. Biosphere 生物圏f. Human being 人間活動

Engineered Barrier人工バリア

Multi barrier system多重バリアシステム(a,b,c,d)

Natural barrier天然バリア

10．Radiation exposure被ばく線量評価 : Sv/y

9. Dose conversion factor:/DCF線量換算係数： Sv/Bq

8. Water utilization : Ingestion河川水利用：経口摂取

7. Dilution 希釈

6. Advection, Dispersion移流、分散：透水(量)係数

5. Sorption：distribution coefficient収着：分配係数

4. Diffusion : diffusion coefficient拡散：拡散係数

3. Precipitation : solubility沈殿生成：溶解度

2. Dissolution : solubility溶解：溶解度

1. Contact with groundwater地下水との接触

HLW高レベル放射性廃棄物

Partitioning/核種分離 &

Transmutation/核変換高速炉、ADS

Reduction of waste volume

and harmfulness減容、有害度低減

on the ground

Radionuclide migration核種移行

Temporal & spatial

uncertainty

時間的、空間的な不確実性


10

SSR-5 Specific Safety Requirement/Disposal of Radioactive waste, IAEA 20111.10.The specific aims of disposal are:(a) To contain the waste;(b) To isolate the waste from the accessible biosphere and to reduce substantially the

likelihood of, and all possible consequences of, inadvertent human intrusion into the waste;(c) To inhibit, reduce and delay the migration of radionuclides at any time from the waste to

the accessible biosphere;(d) To ensure that the amounts of radionuclides reaching the accessible biosphere due to any

migration from the disposal facility are such that possible radiological consequences are acceptably low at all times.

1.12. Disposal facilities are not expected to provide complete containment and isolation of waste over all time; this is neither practicable nor necessitated by the hazard associated with waste, which declines with time.

Concepts relating to disposal of radioactive waste, Important !!

11

SSR-5 Specific Safety Requirement/Disposal of Radioactive waste, IAEA 2011RADIATION PROTECTION IN THE POST-CLOSURE PERIOD2.15Safety objectiveThe safety objective is to site, design, construct, operate and close a disposal facility so that protection after its closure is optimized, social and economic factors being taken into account. A reasonable assurance also has to be provided that doses and risks to members of the public in the long term will not exceed the dose constraints or risk constraints that were used as design criteria.

Criteria(a) The dose limit for members of the public for doses from all planned exposure situations

is an effective dose of 1 mSv in a year [3].(b) To comply with this dose limit, a disposal facility (considered as a single source) is so

designed that the calculated dose or risk to the representative person who might be exposed in the future as a result of possible natural processes affecting the disposal facility does not exceed a dose constraint of 0.3 mSv in a year or a risk constraint of the order of 10–5 per year.

(c) …..inadvertent human intrusion after closure,….. less than 1 mSv….. are not warranted.(d)….. human intrusion….. annual dose of more than 20 mSv….. options for waste disposal

are to be considered.(e)…..annual doses in the range 1–20 mSv….. reasonable efforts are warranted.(f) Similar considerations apply….. for deterministic effects in organs may be exceeded.

2. Post-closure & long-term safety

NuclideT1/2(year)

DCF(μSv/kBq)

Contents(Kg/tonSNF)

U-235 0.7 Billion 47 10kg

U-238 4.5 Billion 45 930kg

Pu-238 87.7 230 0.3kg

Pu-239 24,000 250 6

Pu-240 6,564 250 3

Pu-241 14.3 4.8 1

Np-237 2.14×106 110 0.6

Am-241 432 200 0.4

Am-243 7,370 200 0.2

Cm-244 18.1 120 0.06

Se-79 2.95×105 2.9 0.006

Sr-90 28.8 28 0.6

Zr-93 1.53×106 1.1 1

Tc-99 2.11×105 0.64 1

Pd-107 6.50×106 0.037 0.3

Sn-126 1×105 4.7 0.03

I-129 1.57×107 110 0.2

Cs-135 2.30×106 2.0 0.5

Cs-137 30.1 13 1.5

Rad

ioto

xici

ty/I

nge

stio

n

Time(year)

SNF, Nuclides & RadiotoxicitySNFGlass/HLWwith P&TNatural-U(9tons, incld. Daughter nuclide)

Time vs. Radiological Toxicity/Ingestion/SNF-1tHM(UO2/PWR, 5years, 45,000MWD/t, U&Pu/Separation/99.5%, MA/Separation/99.5%)

From, Text ”Nuclear Fuel Cycle, 8-1 Roles of Partitioning & Transmutation”, K. Tsujimoto, Reprocessing and Recycle Technology Div.,

Atomic Energy Society of Japan, 12

3. Load reduction-1

http://www.aesj.or.jp/~recycle/nfctxt/nfctxt.html

13

Disposal area⇒ Waste occupied area ⇒ Mechanical stability of disposal tunnel⇒ Temperature/buffer material＜100℃

Disposal tunnel to

tunnel distance xD [m]

Pitch between

waste y [m]

Waste occupied

area [m2]

Vertical

emplacement10 4.44 44.4

Remove the heat generating nuclides

Heat generation rate of the vitrified waste

Radionuclide content of the waste

Reduction of foot-print of the repository

* H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan, Second Progress Report on Research and Development for the

Geological Disposal of HLW in Japan, Japan Nuclear Cycle Development Institute (JNC), 2000.


3. Load reduction-2

14

Reactor FuelSpent

fuelReprocessing Vitrification

Vitrified

waste

Geological

disposal

－UO2/

MOX

Burn-

up

Cooling

period

Separation

process

Separation

ratioNuclides

Glass

Matrix

Melter

operation

Waste

loading

Storage

period

Waste

occupied

area

LWR UO2

45

GWD

/THM

4 Purex 99.5 U, Pu - -Approx.

20wt%

50

years44m2/glass

LWR,

FR,

etc.,

UO2

MOX/

Pu

thermal

MOX/

full

Low

～High

＞4

years

Nuclides

and their

separation

ratio

Request

from

geological

disposal

MA:

Np,

Am,

Cm

Measure －

Higher

waste

loading

Heat

generation

Repository

area

Waste

occupied

area

Waste

emplace-

ment

method

(V,H)

Cs/Sr －－Heat

generation

Mo Measure Yellow

phase－

PGM:

Ru, Rh,

Pd

MeasureSedimen-

tation－

Case Item Unit Parameters(example) Notation

1 Fuel type － UO2, MOX separate UO2 &MOX

2 Burn-up GWd/THM 28, 33, 45, 55, 70 influence by BU difference is negligible

3 Cooling period/spent fuel Year 4, 10, 20, 30 classification by color coding

4Nuclide separation ratio

Cs,Sr % 0, 70, 90 classification by symbol

5 MA % 0, 70, 90 same as above

6 Mo,PGM % 0, 70 same as above

7 Waste loading ratio wt%/glass 20, 30, 35Value of CAERA index*

8 Waste occupied are m2/glass 44 ～ 300

4. Integrated approach(1/3) : Reference & Variation

15

Case

SNFCooling period[year]

Cs/Srseparation

[wt%]

MAseparation

[wt%]

Mo/PGMseparation

[wt%]

Vitrified waste, waste loading

[wt%]

CAERA*[kg/m2]

Reduction of waste occupied

area [%]

1 4 90 0 70 35 2.25 43

2 15 70 0 70 25 1.35 72

3 20 70 0 70 25 1.15 84

4 30 0 0 0 21 0.97 100

5 40 0 0 0 21 0.97 100

6 50 0 90 70 35 2.25 43

7 100 0 70 70 35 2.25 43

Result : reduction of waste occupied area/hard rock, -1,000m, vertical emplacement

UO2

CAERA(kg/m2) = Waste loading ratio(wt%/glass) ー Na2O content ratio(wt%/glass)

Waste occupied area(m2/glass)× Glass weight(kg ) ×

1

100

Maximum temperature of buffer material

*CAERA : Comprehensive Analysis of Effects on Reduction of disposal Area(包括的な検討による廃棄体専有面積削減効果)

Results from RWMC’s own research program entitled “ Study on the effects of advanced nuclear fuel cycle technology to the geological disposal concept, Fy.2017”

4. Integrated approach(2/3)

5.1 General Properties of Spent Nuclear Fuel

16

(1) Calculation code & nuclear dataTo estimate isotope composition of spent nuclear fuel(SNF) for evaluation of;

- amount of nuclear fuel materials- management of SNF- fuel cycle properties incl. various chemical/physical processes - safety assessment for radioactive waste generated from reprocessing of SNF

(2) Code & Data/example- ORIGEN-2.0, -2.1, -2.2, -ALP : ORNL Isotope Generation and Depletion Code - ORLIBJ40 : cross section data library- ENSDF : nuclide decay library

(3) Reactor core, fuel type and fuel burn-up/UO2*

Burn-up GWd/THM 28 45 70

Power ratio MW/THM 38

Operation Period

Day 736.84 1184.21 1842.11

Enrichment/U-235

Wt% 2.6 4.5 6.5

Cross section library

-PWR34J40(3.4wt%)

PWR41J40(4.1wr%)

PWR47J40(4.7wt%)

burn-up and cross section data/UO2 fuel

U-232 0.0001 μg/gU

U-234 10×103 μg/gU-235

U-236 250 μg/gU

Tc-99 0.01 μg/gU

Impurities of U-isotopes in UO2 fuel

5. Data required(1/6)

17

(4) Reactor core, fuel type and fuel burn-up/MOX*

Burn-up GWd/THM 28 45 70

Power ratio MW/THM 38

Operation Period

Day 736.84 1184.21 1842.11

Enrichment/Pu

Wt% 7 12 17

Cross section library

-PWRM0210J40(10wt%)

PWRM0213J40(13wr%)

PWRM0213J40(13wt%)

burn-up and cross section data/MOX fuel

Burn-upPu-enrichment

Nuclide composition (wt%)

GWd/THM wt% P u - 2 3 8 Pu- 2 3 9 P u - 2 4 0 P u - 2 4 1 Pu- 2 4 2 A m - 2 4 1 U - 2 3 5 U - 2 3 8

28 7 0 . 1 4 7 3 . 8 1 5 1 . 7 5 0 . 6 5 1 0 . 4 4 8 0 . 1 8 9 0 . 1 6 6 9 2 . 8 3 4

45 12 0 . 2 5 2 6 . 5 4 3 . 0 0 1 . 1 1 6 0 . 7 6 8 0 . 3 2 4 0 . 1 6 6 8 7 . 8 3 4

70 17 0 . 3 5 7 9 . 2 6 5 4 . 2 5 1 . 5 8 1 1 . 0 8 8 0 . 4 5 9 0 . 1 6 6 8 2 . 8 3 4

Chemical composition/MOX fuel

*UO2 & MOX fuel data/settled on TokyoTech/RWMC joint research program for nuclear fuel cycle and geological disposal, Fy.2017


18

(5) Heat generation /nuclides/SNF/UO2

70GWd/THM



SNF cooling period after discharge(year)

Hea

t ge

ne

rati

on

(kW

/TH

M)


45GWd/THMSNF cooling period after discharge(year)


Co

ntr

ibu

tio

n o

f n

ucl

ides

C

on

trib

uti

on

of

nu

clid

es

19

Effects of Nuclide Separation on Repository Layout Removal of heat generation elements Increase waste loading of vitrified waste Layout : Pitch, Distance and Area

Repository layout, exampleCrystalline rock, -1,000m

Vertical emplacement

Time vs. Heat generation rate of vitrified waste

0

50

100

150

200

250

300

350

0.001 0.01 0.1 1 10 100 1000

ガラス固化体の発熱量

, W

/本

処分後の時間, 年

4年

20年

50年

100年

燃焼度：45GWd/t

Heat generation rate/Vitrified waste/without MA SeparationVitrified waste

- PWR- UO2- Burn-up : 45GWd/t- SNF cooling period : 4, 20, 50 & 100 years

Time after Disposal, year

0

20

40

60

80

100

120

140

160

0.001 0.01 0.1 1 10 100 1000

温度, 度

時間, 年

P313 P400

P500 P600

燃焼度：45GWd/t

冷却期間：4年

坑道離間距離：3D

0

20

40

60

80

100

120

140

160

0.001 0.01 0.1 1 10 100 1000

温度

, 度

時間, 年

P313 P400

P500 P600

P640

燃焼度：45GWd/t

冷却期間：50年


Time vs. Temperature, at Overpack/Buffer material interface

SNF cooling period: 4 years

Distance : 3D

SNF cooling period: 50 years

Distance : 3D


Tem

per

ature

, ℃

Tem

per

ature

, ℃

Tem

per

ature

, ℃


Hea

t gen

erat

ion

rat

e o

f v

itri

fied

was

te,

W/g

lass

un

it



20

0

50

100

150

200

250

300

350

0.001 0.01 0.1 1 10 100 1000

ガラス固化体の発熱量

, W

/本

処分後の時間, 年

0%

50%

70%

90%

燃焼度：45GWd/t


No MA Separation : Pitch = 6.4 m50% MA Separation : Pitch = 3.2 m

50% reducing the repository area

0

20

40

60

80

100

120

140

160

0.001 0.01 0.1 1 10 100 1000

温度

, 度

時間, 年

0%

50%

70%

90%

燃焼度：45GWd/t



廃棄体ピッチ：3.2m

Vitrified waste- PWR- UO2- Burn-up : 45GWd/t- SNF cooling period : 50 years- Nuclide separation : Np, Am & Cm - Separation ratio : 0, 50, 70 & 90%

Hea

t g

ener

atio

n r

ate

of

vit

rifi

ed w

aste

,

W/g

lass

un

it


Repository- Crystalline rock- 1,000m below the surface- Horizontal emplacement

Time vs. Temperature at Overpack/Buffer material interface

Burn-up : 45 GWd/t

Cooling period: 50 yearsBurn-up : 45 GWd/t

Cooling period: 50 years

Distance : 3D

Pitch : 3.2m


Tem

per

atu

re, ℃

Heat generation rate/Vitrified waste/with MA Separation



21

MA recycle(100 & 99.9%) by large-scale FR/MOX,/equilibrium phase/discharged fuel composition- Case 1 : Recycle/Np, Am, Cm - Case 2 : Recycle/Np, Am- Case 3 : Recycle/Np- Case 4 : no recycle

*Results from Hokkaido Univ./RWMC joint research program for MA transmutation by fast reactor, Fy.2017Calculation by G. Chiba, Hokkaido Univ.


100% :Separation/recycle

22

6. Summary & Needs

General & Current status/Geological Disposal - Implementers published their own safety case report, incl. Japan- Long-term safety/concept of GD/Isolation & containment is proven- Further improvement

- Site specific- Uncertainty, time & space

View points/Nuclear energy use - GD : Reference vs. Variation- Radionuclide : Migration vs. Existing - Long-term safety : Dose vs. Toxicity- Environmental Impact : Definition (Radiation, amount/RW, area/Repository)

Needs1. General property of SNF : BU & Fuel/chemical composition 2. Separation/Partitioning : MA, Am, FPs 3. Recycle : MA, Am, etc.,4. Vitrification : Quality of glass & Glass-melter operation5. FR : Fuel cycle and discharged wastes6. EI : Environmental load index

23

Thank you for your attention.

Geological Disposal of High-Level Radioactive …chiba/NDS2018/Presentation/S702...Reactor Fuel...

Documents

Transcript of Geological Disposal of High-Level Radioactive …chiba/NDS2018/Presentation/S702...Reactor Fuel...