Supercritical-water Cooled Power Reactor Development Project

Transcription

1 Supercritical-water Cooled Power Reactor Development Project 1. IAE* Fund Program K. Kataoka / Material & Water chemistry N. Saito Long Term Scope 2. IAE R & D Program Progress Report S. Kasahara - Overview - Irradiation Test & Mechanical Property 3. IAE R & D Program Progress Report N. Saito - Corrosion, SCC, Water Chemistry 4. R&D Collaboration (Discussion) * 1 The Institute of Applied Energy, founded under the auspices of leading industries and the Ministry of Economy, Trade and Industry (METI former MITI).

2 Supercritical-water Cooled Power Reactor Development Project Long Term Scope and Milestone for Material & Water Chemistry Toshiba Corp. Hitachi Ltd. Hokkaido Univ. Tokyo Univ.

3 SCPR Research Groups - Materials & Water Chemistry - USA Japan INEEL U. Michigan MIT Hokkaido U. U. Wisconsin ANL U. Tokyo Hitachi Ltd. Toshiba Corp.

4 Schedule of Generation IV Program Screening of Poential R & D Projects Test item Tech. Working Gr TW1 (Water-cooled Reactors) Supercritical Water-cooled Power Reactor, CANDU-NG SCWR(T, F) Technical Working Group TW2 (Gas-cooled Reactors) TW3 (Liquid Metal Reactors) Gas Cooled Thermal Reactor, Prismatic Fuel Modular Reactor, Very High Temperature Reactor, Gas Cooled Fast Reactor (Na, Pb, Pb-Bi) X (MOX, U-TRU,-Zr metal, Th-U-TRU-Zr metal, nitride), total 33 concepts GFR, VHTR Na, Pb-Bi 6 Reactor Types, Gen IV International R & D Program? TW4 (Non-classical Reactors) Transition Cross-cutting Group Vepor Core Reactor, Advanced High Temperature Reactor, Molten Salt Reactor Fuel Cycle Crosscut Fuels & Materials Risk & Safty Economics MSR PMR CANDU-NG Energy Products Budget NERI $16M I-NERI $8.3M Gen IV $8M SCPR supporter (US): INEER, U. Michigan, Wisconsin U., MIT, ANL (J): U. Tokyo, Toshiba, Hitachi, Kyushu U., Hokkaido U. (EU): Framatome ANP, Karlsruhe(FZK), PSI, etc.

5 Cost Estimation for Material Tests in SCPR R & D (Gen IV Report) Test item Facilities Corrosion & SCC Water chemistry Materials stability (Swellig, Phase) Mechanical property (Strength, Embr'mnt, Creep) Cold test: 4 loops Hot cell test: 2 loops (pre-irrad. sample) In-pile test * : 1-2loops (thermal) 1-2loops (fast) Out-pile test loops In-pile test loops Accelerator In-pile test loops Cold tests Hot cell test: (pre-irrad. sample) DBTT 3 to 5 years ($4M/y) Research period (years) 3 to 5 years ($6M/y) Subtotal: $250 to 330 million Thermal ($8M/y/loop, $2-4M/y for PSI) Fast neutron condition ($8M/y/loop, $2-4M/y for PSI) 8 years ($15M/y) Subtotal: $120 million 7-8 years ($3M/y) 15 years ($3M/y) Subtotal: $45 million 10 years ($5M/y) 10 years ($3M/y) Subtotal: $100 million Total : $ million/25-30years

6 Innovative & Viable Nuclear Energy Technology Development Project Sponsor IAE* / METI** *The Institute of Applied Energy **Ministry of Economy, Trade and Industry MEXT*** ***Ministry of Education, Culture, Sports, Science and Technology SCPR-T (Materials, General & Water design chemistry, study, Thermohydraulics, General design study, Material Thermohydraulics) & Water chemistry) Fund: 100 M yen/year(1 M$) M$) Cost-reduced Low-moderation Spectorum BWR Passive Safety Reducedmoderation LWR Integrated Modular Reactor Internal CRD BWR Water chemistry (from LWR to SCPR) Fund: M yen/year?? Cladding Material Viability On-going Future plan Water Chemistry Control Viability National, Utility Joint Project SCPR Phase II Fund: 1000 M yen/year SCPR Phase III Fund: 1000 M yen/year Material Optimization Manufacturing Process

7 Materials & Water chemistry R & D (Phase II & III) Test Item Facilities Corrosion / SCC / Uniform corrosion / SCC / Hydrogen embrittlement Irradiation property / Swelling / Embrittlement / Phase Stability / PCI Mechanical Property / Tensile strength / Thermal creep / Corrosion loop / SSRT loop / Hydrogen absorption test / electron irradiation / ion irradiation / neutron irradiation (I) (specimen) / neutron irradiation (II) (cladding) / High temperature tensile test / Thermal creep test Corrosion tests SCC tests Hydrogen embrittlement electron irradiation test ion irradiation test Tensile strength neutron irradiation test (I) Creep test neutron irradiation test (II) Manufacturing Process / Tube, Plate etc. / Welding Water chemistry / Dissolution Rate / Deposition / Water chem. Control / Processing test / Welding test / g-irrad. loop / Hot-cell loop g-irradiation loop test Hot-cell loop test Manufacturing Process Phase II Material Optimization Phase III Manufacturing Process

8 SCPR Target Condition IAE Program (J) Gen IV R & D Report (US) Reactor Type Thermal Once through Thermal Fast Temperature Pressure n-spectrum Max. dose (dpa) Lifetime (years) o C(Water) Comparable to PWR(Vessel) 25 MPa 5 (Cladding) 60 (Plant) * (Cladding) to 500 o C(Cladding) LWR+30 to 50 o C(Vessel) 10-20(Cladding) (Cladding) 60 (Plant) 6-10 (Cladding) 60 (Plant) (Cladding)

9 Materials & Water chemistry R & D Collaboration 1. Total Plan 2. Facilities 3. Data Share 4. Cross Check of Evaluation Method 5. Cooperative Experiment Plan

10 Supercritical-water Cooled Power Reactor Development Project 1. IAE Fund Program (J-NERI) K. Kataoka / Material & Water chemistry N. Saito Long Term Scope and Milestone 2. IAE R & D Program Progress Report S. Kasahara - Overview - Irradiation Test & Mechanical Property 3. IAE R & D Program Progress Report N. Saito - Corrosion, SCC, Water Chemistry 4. R&D Collaboration (Discussion)

11 IAE/METI R & D Program Progress Report 1.Corrosion & SCC 2.Water Chemistry Toshiba Corp. Hitachi Ltd. Hokkaido Univ. Tokyo Univ.

12 Goal in IAE Project (Phase I) - Materials & Water chemistry - Test Item Facilities (Condition) Corrosion / SCC / Uniform corrosion / SCC Irradiation property / Swelling Mechanical Property / Tensile strength Water chemistry / Radiolysis / Dissolution rate / E-pH diagram / E, ph monitoring / Dose rate simulation / Corrosion loop (Cold, 550 o C, 25MPa) / SSRT loop (Cold, 550 o C, 25MPa) / Electron beam 1000kV, (290, 450, 550 o C 5 dpa) / Tensile test (RT, 550 o C) / g-irrad. loop (Cold, 600 o C, 50MPa) / SCW loop (Cold, 600 o C, 50MPa) Materials screening / Commercial alloys (5 Austenitic alloys, 7 Ni-base alloys, 2 Ferritic alloys, 6 Ti alloys) / New alloys (Austenitic alloys Ferritic alloys Ni-base alloys) <40-50 M yen / year> On-going Future plan / Candidate alloys (2 Alloy types X 2-3 alloys) / Alloy development plan (Preferable additives, microstructure, etc.) / Requirements & Criteria (irradiation, corrosion, SCC, Mechanical / G value / Corrosion environment / Dissolution rate / Water chemistry control basic plan < M yen / year> Cladding Material Viability Water Chem. Control Viability SCPR Materials R & D Team (J) : Toshiba, Hitachi, Hokkaido U., U. Tokyo SCPR Water chemistry Team (J) : U. Tokyo, Toshiba, Hitachi, JAERI, CRIEPI

13 R & D Plan of SCPR Core Component Materials (Phase I) R & D Flow of SCPR Core Component Materials Supercritical Thermal Power Stainless Steels, Ni-Alloys (High-Temp. strength, Creep) Screening commercial alloys Electron Beam Tests for Void Swelling SCWO Ni Alloys, Ti-Alloys (Corrosion resistance) Viability of Existing Materials Promising Alloys Selection + Alloy Design for Improvement Electron Beam Irrad. Tests for Void Swelling Optimization of Chemical Composition & Microstructure for Candidate Alloys (Plant Design) FBR Stainless Steel (Irradiation resistance) Mechanical Properties Assessment High-temp. Tensile Tests Corrosion Properties Assessment Uniform Corrosion Tests SCC Tests Zircaloy is not applicable for SCPR cladding Viability assessment of existing alloys in terms of CorrosionSCC properties Irradiation properties High-temp. strength Improvement of existing materials Optimization for SCPR core component materials

14 Candidate materials (Interim) Existing materials (commercial, prototype) Property, achievements, Data base etc. Material Corrosion Irradiation Mechanical resistance resistance properties Stainless steel Austenitic Ti alloy Ferritic Ni alloy (High-Ni alloy) Good (TP, WP) Poor (TP) Good (WP) Excellent (WP) Poor (FBR) Good (FBR) Poor FBR) Few data available Good high temp. strength Exist improved material high temp. strength Good high temp. strength Insufficient thermal creep resistance Cost low middle high Subject Improve irradiation resistance Needs corrosion data Improve irradiation resistance Needs irradiation property data Typical material Modified 12Cr- 1Mo Alloy625 Alloy690 Alloy800 Ti-6Al-4V Ti-13V-11Cr-3Al New material (Additives, Microstructure) Trace element addition Fine grain Single crystal A few date available Improve irradiation resistance Equivalent to 316L Better high temp. strength Better high temp. strength Needs cost evaluation Needs corrosion data SS316L SS316L SS310 SS316L TP: Supercritical pressure thermal power plant materials WP: Waste processing plants materials FBR: Fast Breeder Reactor

15 Environmental Condition Materials Mechanical Condition Evaluation (Analysis) Test Specimen Austenitic Ferritic Ni-base alloy Ti Alloy New alloy Experimental Plan Corrosion & SCC- Uniform corrosion Weight change Oxide film analysis Coupon specimen: 10X20X2 mm 304, 304H, 316, 316L,310S 12C-1Mo-1WVNb, Mod.9Cr-1Mo 600, 625, 825, 800H, 690, 718, C276, C22 Ti-6Al-4V, Ti-3Al-2.5V, Ti-15V-3Sn-3Al-3Cr, Ti-15Mo-5Zr-3Al ODS, Fine grain SS, Single crystal - e: Constant Double U-bend Temp. : 290, 450, 550 o C Pressure: 25MPa O 2 : 8 ppm Crack initiation Coupon specimen: 10X60X2 mm, 8 mmir Sens. 304, 304H, 316, 316L, 310S 12C-1Mo-1WVNb, Mod.9Cr-1Mo SCC 600, 625, 825, 800H, 690, 718, C276, C22 Ti-6Al-4V, Ti-3Al-2.5V, Ti-15V-3Sn-3Al-3Cr, Ti-15Mo-5Zr-3Al -. e: 4X10-7 /s SSRT Fracture surface analysis (SCC ratio) Maximum stress Cylindrical: 4 mmf, 20 mmg.l. Sens. 304*, 316L, Selected SS Selected alloys *Finished

16 SCW Loop for Uniform Corrosion and U-bend Test Ion exchange resin Test vessel N 2 N 2 +O 2 Control Tank P DO ms Cooler Sampling line Heat exchanger Test vessel Heater SCW SCW P Water chemistry control section Supercritical water Specification Max. Temp. : 600 o C Max. Press. : 25MPa Flow Rate : 50 ml/min 20 mm Coupon specimen Uniform corrosion 20 mm U-bend specimen SCC Susceptibility

17 SCW Loop for Slow Strain Rate Test Ion exchange resin N 2 N 2 +O 2 Control Tank P DO ms Cooler Sampling line SSRT Apparatus Test Piece Example 550 o C, 25MPa, Sens. Type 304SS P Heat exchanger Heater Water chemistry control section Supercritical water Specification Max. Temp. : 600 o C Max. Press. : 25MPa Flow Rate : 50 ml/min Web-monitoring system Data and external view can be monitored with PC from outside of building

18 Requirements & Criteria for Corrosion & SCC Tests Plant Condition (Assumption) Test Condition Water temp. Oxidizing species Flow rate (m/s) Requirement Temp. ( o C) Oxidizing species Flow rate (m/s) Criteria CP release * Refer to BWR Oxide film (Deposition) Thinning ( o C) O 2 (?ppm) H 2 O 2 (?ppm) Radicals (?mmp) 2-5 * * O 2 (8ppm) 10-4 Refer to BWR Refer to BWR SCC Not Susceptible Not Susceptible * Needs Discussion Cladding Material Viability Assessment by

19 Supercritical-water Cooled Power Reactor Development Project 1. Corrosion & SCC 2. Water Chemistry Toshiba Corp. Hitachi Ltd. Tokyo Univ. CRIEPI JAERI

20 Water Chemistry R & D Plan Research Items Radiolysis & Kinetics H 2 O 2, O 2, H 2 Concentration Activation of Water Reactor Water Clean up High-temp. Filter Water Chem. Modification Co-free Material Model Calculation Engineering -Reactor Water, MS- Radiolysis of SCW H 2 O 2, H 2 Evolution Rate N-16 Generation Radiation Buildup Model Corrosion Environment Corrosion Tests Chemical Form of N-compound FP,TRU Generation Fundamentals Degas system Radionuclide Release Rate Requirements for Oxygen conc. in Condensate Metal Release into Coolant Chemical form of FP,TRU Key Factor Condensate Demin. Design Electrochem. Monitoring Co-free Material Chemical Form & Dissolution Rate of Metal Oxides Engineering -Off Gas, Condensate, FW- Hightemp. Filter Amount of Corrosion Products Condensate, Feed Water, SCW Oxide Deposition on Cladding Surface Activation Rate Calc. Engineering Requirements for CP Reduction in FW Thermodynamics, Electrochemistry Main Subjects 1.Corrosion Environment Radiolysis Electrochemical Monitoring 2.Radionuclide Transportation Dissolution Rate Chemical Form Filtration / Deposition 3.Fundamentals of SCW Thermodynamics Kinetics Radiation Buildup Model Radionuclide Removal System

21 Thermodynamics Calculation of SCW = + Revised HKF Model + = = = ( ) = ( ) ( ) = + + ( ) ( ) = : : = + ( ) = = ( ) ( ) Calc. E-pH Diagram of SCW ( ) ( ) ( ) ( + ) ( ) ( ) = ( ) = = ( ) ( ) = = = ( ) ( ) ( ) = + + ( ) Conventional (Criss & Cobble) ( ) + ( ) + ( ) + ( ) + ( ) ( ) ( ) + ( )

22 Materials & Water chemistry R & D Collaboration 1. Total Plan 2. Facilities 3. Data Share 4. Cross Check of Evaluation Method 5. Cooperative Experiment Plan Facilities Available for SCPR R & D SCW Test Loop Irradiation Facilities Japan 1. SSRT Loop (Toshiba, 600 o C, 25 MPa) 2. Corrosion Loop (Hitachi 600 o C, 25MPa) 3. Multi-purpose Monitoring Loop* (Toshiba, 600 o C, 50 MPa) 4. g-irrad. Loop* (Hitachi, 600 o C, 50MPa) 1. Electron beam accelerator (Hokkaido U., 1000kV) US Needs Information Needs Information * Future Plan