Hidekazu ASANO Radioactive Waste Management Funding and Research - - PowerPoint PPT Presentation

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Symposium on Nuclear Data TokyoTech, 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)

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Key words

Geological Disposal Long-term Safety Radionuclide Migration Radiation Dose vs. Radiological Toxicity Environmental Impact Viewpoints Data needs from new fields

Topics/GD

• 1. Concept “Isolation & Containment”
• 2. Post closure & long-term safety
• 3. Load reduction
• 4. Integrated approach
• 5. Data required
• 6. Summary & Needs

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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-79 Zr-93 Tc-99 Pd-107 Sn-126 Cs-135 I-129/AgI C-14 TRU waste TRU waste HLW: vitrified waste

＊ ＊

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

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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) ＊ ＊

• 1. Concept “Isolation & Containment”(2/9)

Radioactivity of vitrified waste vs. Time after disposal*

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* 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

• 1. Concept “Isolation & Containment”(3/9)

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* 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

• 1. Concept “Isolation & Containment”(4/9)

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HLW(Vitrified waste) Result : safety assessment/long-term & post-closure safety

Radiation 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
• 1. Concept “Isolation & Containment”(5/9)

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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)

• 1. Concept “Isolation & Containment”(7/9)

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• 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

• n the ground

Radionuclide migration

核種移行

Temporal & spatial uncertainty 時間的、空間的な 不確実性

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

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SSR-5 Specific Safety Requirement/Disposal of Radioactive waste, IAEA 2011 1.10. 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

• ver all time; this is neither practicable nor necessitated by the hazard associated with waste,

which declines with time.

Concepts relatin ing to di disposal of ra radioacti ctive ve waste, Important nt !!

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SSR-5 Specific Safety Requirement/Disposal of Radioactive waste, IAEA 2011

RADIATION PROTECTION IN THE POST-CLOSURE PERIOD 2.15 Safety objective The 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

Nuclide

T1/2

(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

Radiotoxicity/Ingestion Time(year) SNF, Nuclides & Radiotoxicity

SNF Glass/HLW with P&T Natural-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

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

Disp sposal

• sal tunnel

el to tunnel el dista stance xD [m] Pitc tch betwee tween n waste ste y [m] Wast ste e occupied ed area ea [m [m2] Verti tical emplacemen ement

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

• 3. Load reduction-2

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Reactor actor Fuel uel Spen ent t fuel el Rep eproc roces essi sing Vitr trification ation Vitr trified ed waste ste Geol

• logi
• gica

cal dispo sposa sal － UO UO2/ MOX Burn urn- up up Cool

• oling

peri eriod

• d

Separ aration ation proces rocess Separ aration ation ratio tio Nuc uclides Glas ass s Matr trix Mel elter ter

• perati

eration

• n

Waste ste load ading Storage rage peri eriod

• d

Waste ste

• ccu

cupied ed area ea LWR UO UO2 45 45 GWD /THM 4 Purex urex 99.5 U, Pu

• Approx

rox. 20wt% t% 50 50 years ars 44m 44m2/glass ass LWR, FR, etc., UO UO2 MOX/ X/ Pu Pu thermal ermal MOX/ X/ ful ull Low ～ High gh ＞4 4 years ars Nuc uclides s and nd thei eir r sep eparat aration

• n

ratio tio Requ quest est from

• m

geol

• log
• gical

cal dispo sposa sal MA: Np, Am, Cm Cm Measu Measure － High gher er waste ste load ading Heat t gene nerati ration

• n

Rep eposi

• sitor

tory y area ea Waste ste

• ccu

cupied ed area ea Waste ste emplac ace- ment nt method hod (V,H) Cs/Sr － － Heat t gene nerati ration

• n

Mo Mo Measu Measure e Yel ellow

• w

pha hase se － PGM: M: Ru, Rh, h, Pd Pd Measu Measure Sedimen men- tation ion －

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 4 Nuclide 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, 35 Value of CAERA index* 8 Waste occupied are m2/glass 44 ～ 300

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

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Case SNF Cooling period [year] Cs/Sr separation [wt%] MA separation [wt%] Mo/PGM separation [wt%] Vitrified waste, waste loading [wt%] CAERA* [kg/m2] Reduction of waste occupied area [%] 1 4 90 70 35 2.25 43 2 15 70 70 25 1.35 72 3 20 70 70 25 1.15 84 4 30 21 0.97 100 5 40 21 0.97 100 6 50 90 70 35 2.25 43 7 100 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

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(1) Calculation code & nuclear data To 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)

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(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-up Pu- enrichment Nuclide composition (wt%) GWd/THM wt% P u - 2 3 8 Pu-239 Pu-240 Pu-241 Pu-242 Am-241 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 92.834 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 87.834 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 82.834

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

• 5. Data required(2/6)

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(5) Heat generation /nuclides/SNF/UO2

70GWd/THM

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”

• 5. Data required(3/6)

SNF cooling period after discharge(year) Heat generation (kW/THM) SNF cooling period after discharge(year)

45GWd/THM

SNF cooling period after discharge(year) SNF cooling period after discharge(year) Contribution of nuclides Contribution of nuclides

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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, example Crystalline rock, -1,000m Vertical emplacement

Time vs. Heat generation rate of vitrified waste

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 Separation Vitrified waste

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

Time after Disposal, year

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 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 坑道離間距離：3D

Time vs. Temperature, at Overpack/Buffer material interface

SNF cooling period: 4 years Distance : 3D SNF cooling period: 50 years Distance : 3D Time after Disposal, year Temperature, ℃ Temperature, ℃ Temperature, ℃ Time after Disposal, year

Heat generation rate of vitrified waste, W/glass unit

Results from RWMC’s own research program entitled “ Study on the effects of advanced nuclear fuel cycle technology to the geological disposal concept, Fy.2016”

• 5. Data required(4/6)

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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 冷却期間：50年

No MA Separation : Pitch = 6.4 m 50% MA Separation : Pitch = 3.2 m 50% reducing the repository area

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 冷却期間：50年 坑道離間距離：3D 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%

Heat generation rate of vitrified waste, W/glass unit Time after Disposal, year 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 years Burn-up : 45 GWd/t Cooling period: 50 years Distance : 3D Pitch : 3.2m

Time after Disposal, year Temperature, ℃

Heat generation rate/Vitrified waste/with MA Separation

Results from RWMC’s own research program entitled “ Study on the effects of advanced nuclear fuel cycle technology to the geological disposal concept, Fy.2016”

• 5. Data required(5/6)

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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.2017 Calculation by G. Chiba, Hokkaido Univ.

• 5. Data required(6/6)

100% :Separation/recycle

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• 6. Summar
• 6. Summary

y & & Needs 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)

Needs

• 1. 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 operation
• 5. FR : Fuel cycle and discharged wastes
• 6. EI : Environmental load index

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Thank you for your attention.