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

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) 1

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 2

1. Concept “Isolation & Containment”(1/9) TRU waste I-129/AgI ＊ Se-79 Zr-93 Tc-99 Pd-107 Sn-126 HLW: vitrified waste TRU waste C-14 Cs-135 ＊ High-level Liquide waste Processes and generation of radioactive waste at reprocessing plant Second progress report on research and development for TRU waste Disposal in Japan – Repository design, safety assessment and means of 3 implementation in the generic phase-, JAEA-Review 2007-010, FEPC TRU-TR2-2007-01, March 2007

1. Concept “Isolation & Containment”(2/9) Process conditions for vitrified waste(Glass) ＊ ＊ 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 4

1. Concept “Isolation & Containment”(3/9) Systems for nuclide migration Radioactivity of vitrified waste vs. Time after disposal* * 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. 5 TN1410 2000-003, Supporting Report 2, Repository Design and Engineering Technology

1. Concept “Isolation & Containment”(4/9) * 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. 6 TN1410 2000-003, Supporting Report 2, Repository Design and Engineering Technology

1. Concept “Isolation & Containment”(5/9) 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 7 - TN1410 2000-004, Supporting Report 3, Safety Assessment of the Geological Disposal System

1. Concept “Isolation & Containment”(7/9) Barrier performance - Containment capability of each barrier component by multi-barrier system Np-237 Cs-135 Np*, low solubility & large distribution coefficient Cs**, highly soluble & less sorptive - 10 7 years, approx. 1 % of Np-237 as of the initial - approx. 95% of the initial inventory of Cs-135 is inventory moves isolated at the area of EBS & geosphere to geosphere from EBS - approx. 1/10 5 of the original Np-237 reaches to the (**T 1/2 =2.30×10 6 years) biosphere (*T 1/2 =2.14×10 6 years) Nuclide inventory in the EBS components and Geosphere 8

1. Concept “Isolation & Containment”(9/9) on the ground HLW Partitioning/ 核種分離 & Reduction of waste volume 高レベル Transmutation/ 核変換 and harmfulness 放射性 減容、有害度低減 高速炉、 ADS 廃棄物 Radionuclide migration 核種移行 f. Human being 人間活動 10 ． Radiation exposure e. Biosphere 生物圏 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 Multi barrier system 多重バリアシステム (a,b,c,d) 7. Dilution 希釈 d. Geological formation/ 1. Isolation host rock 地質 / 岩盤 隔離 6. Advection, Dispersion 移流、分散 ： 透水 ( 量 ) 係数 Natural barrier 5. Sorption ： distribution coefficient 天然バリア 収着： 分配係数 Temporal & spatial uncertainty 4. Diffusion : diffusion coefficient 時間的、空間的な 拡散： 拡散係数 不確実性 3. Precipitation : solubility Engineered Barrier 沈殿生成： 溶解度 人工バリア c. Buffer material/ 緩衝材 2. Dissolution : solubility 2. Containment b. Overpack/ オーバーパック 溶解： 溶解度 閉じ込め a. Vitrified waste/ 1. Contact with groundwater ガラス固化体 地下水との接触 (Bq/kg) 9

Concepts relatin ing to di disposal of ra radioacti ctive ve waste, Important nt !! 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 over all time; this is neither practicable nor necessitated by the hazard associated with waste, which declines with time. 10

2. Post-closure & long-term safety 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. 11

3. Load reduction-1 SNF, Nuclides & Radiotoxicity SNF T 1/2 DCF Contents Glass/HLW Nuclide ( μSv /kBq) (Kg/tonSNF) (year) with P&T Natural-U(9tons, incld. Daughter nuclide) U-235 0.7 Billion 47 10kg U-238 4.5 Billion 45 930kg Pu-238 87.7 230 0.3kg Radiotoxicity/Ingestion Pu-239 24,000 250 6 Pu-240 6,564 250 3 Pu-241 14.3 4.8 1 2.14×10 6 Np-237 110 0.6 Am-241 432 200 0.4 Am-243 7,370 200 0.2 Cm-244 18.1 120 0.06 2.95×10 5 Se-79 2.9 0.006 Sr-90 28.8 28 0.6 Zr-93 1.53×10 6 1.1 1 2.11×10 5 Tc-99 0.64 1 6.50×10 6 Pd-107 0.037 0.3 Time(year) 1×10 5 Sn-126 4.7 0.03 1.57×10 7 I-129 110 0.2 Time vs. Radiological Toxicity/Ingestion/SNF-1tHM (UO2/PWR, 5years, 45,000MWD/t, Cs-135 2.30×10 6 2.0 0.5 U&Pu/Separation/99.5%, MA/Separation/99.5%) Cs-137 30.1 13 1.5 From, Text ”Nuclear Fuel Cycle, 8 - 1 Roles of Partitioning & Transmutation”, K. Tsujimoto, Reprocessing and Recycle Technology Div., 12 Atomic Energy Society of Japan,

3. Load reduction-2 Disposal area ⇒ Waste occupied area ⇒ Mechanical stability of disposal tunnel ⇒ Temperature/buffer material ＜ 100 ℃ Heat generation rate of the vitrified waste Radionuclide content of the waste Remove the heat generating nuclides Reduction of foot-print of the repository Disp sposal osal tunnel el to Pitc tch betwee tween n Wast ste e occupied ed tunnel el dista stance xD [m] waste ste y [m] area ea [m [m 2 ] Verti tical 10 4.44 44.4 emplacemen ement * H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan, Second Progress Report on Research and Development for the 13 Geological Disposal of HLW in Japan, Japan Nuclear Cycle Development Institute (JNC), 2000. TN1410 2000-003, Supporting Report 2, Repository Design and Engineering Technology