Innovative Fuel Cycle with Minimizing Environmental Burden and Future Direction

Transcription

1 Innovative Fuel Cycle with Minimizing Environmental Burden and Future Direction Tadashi Inoue Research Advisor to CRIEPI, Emeritus Corresponding to; Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

2 Conventional Fuel Cycle U Enrichiment Fuel fabrication LWR Spent fuel Reprocessing FP MA Disposal U, Pu MOX LWR FBR MOX fabrication U, Pu recycle in LWR and/or FR -Pure Pu is conventionally extracted by PUREX process. (U is mixed with Pu not to make a weapon-usable material in Japan) - Minor actinides with long half-live and high toxicity are disposed as HLW into geologic depository Flow of Reprocessing Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

3 Paradigm Change to be Required for the Next Generation Conventional fuel cycle has been developed for effective use of uranium resources. High level wastes with minor actinides and fission products will be disposed of into geologic formation Current Issue to be Overcome Radioactive waste issue is critical for further use of nuclear energy Waste issue should be more crucial for fact reactor era as well as current era Paradigm change is emerging issue for further use of nuclear energy The most prioritized exploration is the research for innovation on radioactive waste management Fuel cycle innovation should be explored for the next generation in use of nuclear energy Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

4 Long-lived Nuclides in HLLW (LWR Spent UO2; 43Gwd/t) TRU FP Nuclide Half-life (y) Weight in 1tHM (g) Heat Generation (W/tHM) Np m 582 <0.1 Pu Pu , <0.1 Pu <0.1 Pu <0.1 Am Am Cm Se-79 65, <0.1 Sr Zr m 926 <0.1 Tc , <0.1 Pd m 282 <0.1 Sn , <0.1 I m 2.33 <0.1 Cs m 487 <0.1 Cs Sm <0.1 1) UO 2 -LWR 43GWd/t 5 years after discharge. After 99.5% separation of Pu, I 1% in HLLW Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

5 What is required in order to minimize the Environmental Burden at Waste Treatment and Disposal On spent fuel management Minor actinide management Long-lived fission product management Innovation of waste treatment/disposal from technical, societal and economical view points On fuel cycle processing Simply-designed reprocessing stream with minimizing a number of process Avoiding a complex associated-process for cleaning or scrubbing solvent Reducing a waste volume and to minimize secondary waste Simply-designed fuel fabrication (due to mixing high heat-generated nuclides) Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

6 Progress of Partitioning and Transmutation in Japan History in Japan Basic research at former JAERI (~1980 s) Four group separation; TRUs, Cs-Sr, PGM, Others (aqueous process) Long-term research plan on Partitioning and Transmutation (Oct. 1988, JAEC) Started OMEGA Project (Institutions; JAERI, JNC, CRIEPI) Current status and future direction of P&T of long-lived nuclides (as Phase 1) (March, 2000, JAEC) Current status and future direction of P&T of long-lived nuclides (as Phase 1) (April, 2009, JAEC) Feasibility Study of FBR Cycle;, Phase-1( ); Determination of applicable technology Phase-2( ); Selection of main concept and proposal of R/D plan Fast Reactor Cycle Technology Development Project (FaCT Project) - Started 2006, No further large program after the Fukushima accident History in CRIEPI Concept and technological feasibility research of Transmutation (~1983) Pyro-partitioning of TRUs and burning in metal fuel FBR (1986 ~ ) Cooperation for researches with TRU and irradiated fuel (1988 ~ ) Engineering installation, and experiment with U and Pu Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

7 Intitating OMEGA Project in Japan (1980 s) Concept of Partitioning Concept and Transmutation in OMEGA Project Transmutation TRUs Pu-239, Np-237, Am-241, Cm-244, etc. Actinide Burner FBR Disposal ADS Conversion/Fabrication Enrichment Conversion U Pu Generation (LWR) Pu Partitioning Sr, Cs Group Sr-90, Cs-137, etc. Utilization ADS Disposal U U HLLW Utilization U, (Pu) Reprocessing Tc-PGM Group Tc-99, Rd, Ru, Pd, etc. Utilization Conversion/Fabrication Generation (FBR) Waste from reprocessing Others Mo, Zr, La, etc. Undissolved residue Utilization Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

8 Direction on Partitioning and Transmutation (Decision by JAEC on April 2000) Partitioning and transmutation of long-lived nuclides has a potential to reduce a burden on disposal and to utilize valuable elements from the viewpoint of reduction of toxic waste associated industrial activity The technology is a basic stage and is, hereafter, required a steady progress of research and development The technology makes possible long-lived nuclides to short-lived or stable ones by transmutation, and contains issues to be solved by implementing innovative technology, which requires creative ideas It is important/effective to provide a research field of this technology for young scientists and engineers, which leads to activate a nuclear research Translated by T. Inoue Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

9 2 Types of Concept for P&T 旧 JNC: 酸化物燃料高速炉電中研 : 金属燃料高速炉 Homogeneous cycle FBR for both electricity and transmutation Fuel fabrication Pu Double strata Fresh fuel without MA FBR (or LWR) for electricity 旧原研 Spent fuel Fresh fuel with MA Pu,MA FP FP Partitioning (Reprocessing) MA Spent fuel Fuel fabrication Partitioning (Reprocessing) Disposa l Fuel fabrication Fresh MA fuel ADS for transmutation Spent MA fuel Transmutation using electricity generating plant MA is recycled as Pu in the next-generation reprocessing. The fresh fuel for the FBR contains small amount of MA up to 5 wt%. Disposa l Reprocessing Transmutation in the dedicated (second) fuel cycle with the Accelerator-Driven System (ADS) MA is confined in the compact transmutation-cycle. The fresh fuel for the ADS mainly consists of MA. Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

10 Transmutation of I-129 and Cs-137 Neutron capture reaction Iodine-129 Neutron Half-life; 1.57x10 7 hrs. Photonuclear reaction Cesium-137 γ-ray Half-life; 30 hrs. Neutron capture Neutron Half-life; 12.4 hrs. Half-life; 12.6 d. β-ray β-ray Stable Stable Figure. Transmutation of fission products with use of neutron capture and photonuclear reaction Ref; Current status and future direction on P&T research, JAEC, March, 2003 ImPACT Seminar July 3,

11 Contribution to the Area Reduction Required for Disposal of HLW (Unit m 2 /TWh) UO2-LWR(43GWd/t), CT= 5 years UO2-LWR(43GWd/t), CT=20 years Without MA transmutation With MA transmutation MOX-LWR(43GWd/t), CT= 5 years MOX-LWR(43GWd/t), CT=20 years FBR(79GWd/t), CT= 5 years FBR(79GWd/t), CT=20 years CT: Cooling time between fuel discharge and reprocessing Emplacement area required for HLW disposal (m 2 / TWh) Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

12 Contribution to the Area Reduction Required for Disposal of HLW (Unit m 2 /TWh) UO2-LWR, CT= 5 年 UO2-LWR, CT=20 年 MOX-LWR, CT= 5 年 MOX-LWR, CT=20 年 従来型 PUREX 法 MA 回収 リサイクル FP 群分離 MA FP の分離変換 FBR(79GWd/t), CT= 5 年 FBR(79GWd/t), CT=20 年 FBR(147GWd/t), CT= 5 年 FBR(147GWd/t), CT=20 年 MA リサイクル高速炉 ( 平衡状態 ) のため MA 回収が前提条件 CT: 再処理前の冷却期間 単位発電量あたりに生じる高レベル廃棄物の定置に要する面積 (m 2 / TWh) Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

13 Characteristic of Pyroprocessing with MA Recycle Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

14 Characterization Pyroprocess Fuel cycle Separation of U and Pu Pyro-processing (Metal electrorefining) FR fuel cycle Dedicated cycle for MA burning Electrorefining (Pu separated together with minor actinide) Separation of minor actinide No additional process is required Separation of long-lived fission products Extraction solvent No process has not been explored Molten salt and liquid metal No degraded by radiation Product U, (U+)Pu+Np+Am+Cm Purity of product Low decontamination Operating temperature More than 500 C, some process needs over 1000 C Type of waste Salt Solid waste; scrap of used equipment, such as crucible, mold Waste form Zeolite, Sodalite (with glass mixed) Accessibility to product (proliferation resistance) Applicability on fuel cycle Handled only in hot cell for every product Metal fuel cycle Oxide fuel cycle Treatment of HLLW Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

15 Simply Designed Process with Minimizing a Number of Process Spent metal fuel Decladding & Chopping (U-Pu-Zr) +MA, FP NM U Zr Pu,U,MA RE Pu,U,MA AM,AEM RE, I Electrorefining U,Pu,MA U,Pu,MA -Cd U metal with salt Spent salt (LiCl-KCl + U,Pu,MA,FP) U/Pu/MA Distillation Distillation of salt and Cd Evaporated and collected Cd and salt U,Pu,MA U metal Fresh metal fuel Fuel fabrication (Injection casting) Fresh U Zr Spent salt Cd-U,Pu,MA Recycle Salt with FP Salt purified Cd-Li Recycle d salt Zeolite column (FP decontamination) Zeolite contained FP Multistage counter current extraction Zeolite absorbed alkali and alkaline earth FP in salt Spent salt(licl-kcl + FP) Zeolite+FP + LiCl-KCl U,Pu+MA recovery into Cd from salt Salt waste solidification Boric acid, Al2O3 Quartz mold used Sodalite Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

16 Electrorefining Technology Developments Liquid Cd cathode Anode/solid cathode Test with simulants Test with U Cd ingot with U+Pu (Cd:120g U+Pu:14.7g) 3.55 kg-u Cd 50mm (U,Pu)Cd 11 (U,Pu)Cd 6 500μm Zr/U : (in weight) salt content : 33.0 wt % Collected product after 1 st scraper operation (most U-rich product) Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

17 Counter Current Extraction to Extract Actinides CRIEPI Report; L11906 (2012) Remark; Picture without heat-insulating cover Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

18 Salt Waste Recycle Electric furnace ( C) Level gauge, Thermocouple, etc. Element with different valence is expected to be captured depending on a type of zeolite Zeolite column CRIEPI Report; L11906 (2012) Picture without heat-insulating cover Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

19 Application of Pyro-process to LWR Fuel Cycle and PUREX Liquid waste LWRs LWR fuel cycle FBRs U-Pu-MA-Zr FBR metal fuel cycle Spent fuel MOX Reduction to metal U, Pu, MA U, Pu Pyro-processing Electrorefining Reductive extraction Injection casting Purex HLLW Pyro-partitioning MA Denitration Chlorination Reductive extraction Waste (salt, metal) Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

20 TC Li-Pb ref. UO 2 cathode Application on Oxide Fuel Pt anode MgO shroud Experimental conditions Electrolyte: LiCl-1wt%Li 2 O UO 2 : Run 1 = sinter-a 87 g Run 2 = sinter-b 104 g Current: 15 A 1 A Time: Run 1 = 11.5 h Run 2 = 9.3 h Cathode basket with UO 2 sinters UO 2 LiCl-Li 2 O (650 o C) The cell for electrolytic reduction Run 1 Run 2 Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

21 (a) Electrolytic Reduction (b) (d) Inside of cathode basket (c) SEM image (e) 1 mm External appearance of cathode basket Polished cross section of reduced UO 2 sinter 50 m Fig. Reduction Product of Run 2: all of UO 2 sinters were completely reduced. Current efficiency: Run 1= 64%, Run 2= 62% One halfway-reduced UO 2 sinter was found in Run 1. Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

22 Application on HLLW (Pyro-partitioning) Separation factors and other basic data are obtained in the TRU group-partitioning process development program. The points are to recover TRU more than 99:5% with the enough separation factor from rare earths existing 10 times more than TRU in HLLW. In case of commercial FBR without MA recycling, MA should be recovered from HLLW for burning by dedicated reactors, such as ABR or ADS system. alkali elements Tc, Mo, Se Fe, Zr, Mo noble metals Electro-winning U HLLW denitration chlorination reductiveextraction multistageextraction Electro-winning >99% of TRU TRU / RE=1 / 1 TRU / RE=1 / 10 (in weight) H 2 O, NO X Cl 2, LiCl-KCl salt waste treatment RE, Sr, Ba Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

23 Distribution coefficient of element = Molar fraction in Cd/molar fraction in salt Separation of Transuranium elements in HLLW Oxide denitrated Concentrated HLL Drying( ) 脱硝物全量 (7.3g) を回収揮発物は微量の Ru のみ ( %) Chloride after chlorination Distribution coefficient of U-238 = Molar fraction in Cd/molar fraction in salt ImPACT Seminar July 3,

24 Entire Process in an Unit Cell Spent fuel storage Reprocessing Plane Fuel Fabrication Plant Fresh Fuel Storage PUREX MOX Spent Fuel Assemblies Chopping, Dissolution Separation by Extraction Product: pure U, Pu Blending of U and Pu Pellet fabrication Pin filling & assembling Storage of fresh fuel assemblies Pyroreprocessing Metal Spent Fuel Assemblies Chopping, Electrorefining Cathode processing (Product; U, TRUs(f.p.) Blending of U and TRUs Centrifugal casting Pin filling & assembling Storage of fresh fuel assemblies Biological shielding α-containment Different to PUREX/MOX plant The process can be operated in an unit cell till making fresh fuel Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

25 Metallography of CR10 alloy: 44U-18Pu-10Zr-9Np-5Am-3Ce-10Nd (wt%) U-Pu-Zr-Np phase Miscibility of MA and RE into U-Pu-Zr U-Pu-Zr-MA-RE alloys of different compositions were mixed by arc-melting. Pu-Am-RE phase Metallography of CR101/3 alloy: 39U-22Pu-12Zr-15Np-10Am-0.6Ce-1.8Nd Pu-Am-RE precipitates U-Pu-Zr-Np phase In the alloys of high RE content, Matrix segregates into upper and lower parts. 100µm In the alloys of low RE content ( 5%), RE-rich precipitates ( 30µm) were uniformly dispersed. RE content should be reduced to 5%. Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

26 Metal fuel Pin No.1 Metal fuel Pin No.2 Metal fuel Pin No.3 Oxide fuel driver pin 285 1, METAPHIX Experiment for MA Transmutation in Metal Fuel Irradiation experiment was carried out in PHÉNIX with support of CEA 3 metal fuel pins & 16 oxide fuel pins were fabricated and arranged in an capsule. Pin No.1 : U-19Pu-10Zr Pin No.2 : U-19Pu-10Zr-2MA-2RE Pin No.3 : U-19Pu-10Zr-5MA / -5MA-5RE MA~60Np-30Am-10Cm, RE~10Y-10Ce-70Nd-10Gd. Burnup goals ~2.5at.% (METAPHIX-1), ~ 7at.% (METAPHIX-2), ~10at.% (METAPHIX-3). Initial bond sodium level U-Pu-Zr U-Pu-Zr U-Pu-Zr U-Pu-Zr -2MA-2RE U-Pu-Zr U-Pu-Zr -5MA-5RE U-Pu-Zr -5MA Irradiation Capsule U-Pu-Zr U-Pu-Zr U-Pu-Zr No.1 No.2 No.3 Oxide Fuel Pin (Driver) Metal Fuel Pin 6.55mm Fuel pin arrangement in irradiation capsule. Schematic views of irradiated fuel pins. Unit [mm] Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

27 Irradiation Rig and Fuel Pin after Irradiation Irradiation rig after irradiation Fuel pin in irradiation rig (during disassembling) Spot with color change to dark Appearance of fuel pin Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

28 Metallography of Irradiated U-19Pu-10Zr-5MA- 5RE Central Zone 100 m Porous phase γ-phase Unirradiated U-Pu-Zr-5MA-5RE 1mm 100 m Intermediate Zone Dense phase (γ+ζ)-phases Cross-Sectional Overview 100 m Peripheral Zone The burnup is ca. 2.5 at%. Matrix morphology is similar to that of U-Pu-Zr fuel (Sample #1). Large precipitations (MA and RE inclusions) appear in γ phase zone. Some narrow layered phases (MA-RE inclusions) spread along grain boundaries in γ+ζ zone. In low-temperature regions, small dark spots (MA and RE inclusions) are observed. Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

29 Challenge to Minor Actinides Management Management of minor actinides accompanied extremely high radiation and heat is essential to sustain nuclear energy utilization in 21 st century. Partitioning and transmutation of minor actinide within a homogeneous or a heterogeneous fuel cycle can provide a potential solution Over 99% separation of TRUs from HLLW makes the radiotoxicity of the waste possible to the level of 7.5 ton of natural uranium after several hundred years of disposal, and reduces remarkably an emplacement area of HLW at the geological disposal. Pyroprocessing and transmutation by metal fuel gives a potential solution of the goal by achieving the strong proliferation resistance. Currently R&D of pyroprocessing and transmutation by metal fuel has been advanced While, large barriers should be solved to make the technology maturity, and to accept social/economic issues. Fuel fabrication, handling and storage should be remarkably a big hurdle by implementation of partitioning and transmutation. International collaboration is a potential approach to achieve the goal of partitioning and transmutation Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

30 Issues of Pyro-processing form point of Waste Minimization From point of long-lived nuclides management -Minor actinides are recovered together with U and Pu due to the intrinsic electrochemical property -Most of long-lived fission products are dissolved into electrolyte or remained in anode Nuclides with half-life more than 100,000 years ; Se-79, Zr-93, Tc-99, Pd-107, Sn-126, I-129, Cs-135 From point of process -Minimizing secondary waste arising from used process equipment and salt scrubbing -Reducing a volume of waste matric, such as zeolite, sodalite, Recovery ratio - Determining the recovery ratio in order to achieve a simplified waste treatment/disposal Balance of cost and benefit with Societal view-point are another critical issue. Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

31 Summary Radioactive waste issue is the most critical and is to be highly prioritized for the next generation nuclear era. Management of long-lived nuclides -Minor actinides; R&D in progress -Long-lived fission products (I, Tc); R&D done and stopped due to no large efficiency for transmutation and technological difficulty -Long-lived fission products (Others); fundamental research started, ex. ImPACT Required innovative research and technology -If we could achieve a waste system without HLW/SF disposal, Paradigm of nuclear world must be changed. Balance of cost and benefit with Societal view-point are another critical issue Waste issue in Fukushima decommissioning program; not exceptional Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

32 Codfish with 9.7 kg Consultancy Meeting on Advanced Fuel Cycle for Waste Minimization, IAEA, Vienna, June 21 24,

33 Technology Readiness Level based on NASA Standard TRL Partitioning Transmutation Concept developed Literature review Flowsheet developed Option evaluated Bench-scale hot test Criteria presented Industrial scale cold unit test Scale-up Industrial scale hot unit test Process/equipment designed Full scale hot test Design feasibility confirmed 7 Commercial plant cold test Concept developed Literature review Concept materialization Option evaluated Basic study activated Criteria presented Labo-scale inidividual tests Simulation study activated Individual technology developed Research reactor conctructed Industrial scale tests Research reactor operated Proto-type reactor operated Commercial plant designed Current stage 8 Commercial plant hot test Commercial plant constructed 9 Commercial plant operated Commercial plant operated Regarding MA-technology, maturity of major R&D programs is mostly in TRL-3 and some below TRL-5. There is a deep technology valley between TRL-4 and TRL-5 due to difficulty on MA-treatment. This could cause stagnation of various R&D programs. ImPACT Seminar July 3,

34 適用した還元反応 NpO 2 + 2Ca 2CaO + Np (in CaCl 2 ) PuO 2 + Th ThO 2 + Pu Am 2 O 3 + 2La La 2 O 3 +2Am CmO 2 + Th ThO 2 + Cm 還元セル Am 金属 ヒーター 電極 Cm 金属 MA 酸化物の金属への還元装置 ImPACT Seminar July 3,