Overview, Irradiation Test and Mechanical Property Test

Transcription

1 IAE R&D Program Progress Report Development Project of Supercritical-water Cooled Power Reactors Overview, Irradiation Test and Mechanical Property Test Shigeki Kasahara Hitachi, Ltd. Toshiba Corp. Hokkaido Univ. Univ. of Tokyo SCPR:Supercritical-water Cooled Power Reactor

2 Requirement on the Cladding Materials Reliability, Durability, and Economy Neutron economy (Neutron Absorption) Productivity, Cost Corrosion General Stress Corrosion Cracking (SCC) High temp. strength Tensile, Creep(Unirrad.) Radiation effect Swelling, Creep, Embrittlement, Radioactivation

3 Design of SCPR fuel assembly Configuration Cladding Coolant (Supercritical Water) Rod pitch 9.5 mm Thickness 0.8mm Rod dia. φ8 mm Specification Coolant flow: Outside of claddings, narrow gap (<1 mm) Bundle, Spacer Fuel assembly : Dimensional accuracy required Materials : Low swelling, high creep strength, corrosion resistance necessary

4 SCPR core environment (1) Neutron spectrum and flux (Case of thermal neutron SCPR: Similar to APWR) Fluence (Core) 2x10 25 n/m 2 /cycle (LWR) 2~3 dpa/cycle (Fe base, Ni base alloy) Max. damage = 15 dpa The material database obtained under similar neutron irradiation condition can be applied to the evaluation of candidate materials for the cladding. Yasteelhi Okano : Conceptual Design of a Supercritical Pressure Light Water Reactor using Double Tube Water Rods : Ph. D. dissertation, The University of Tokyo (1997)

5 SCPR core environment (2) Temperature: 300 C ~ 550 C Pressure: 25 MPa Max. Temp. (Current design) Axial Node Kazuaki Kitou : Safety Analysis of a Supercritical Water Cooled Reactor : Ph. D. dissertation, The University of Tokyo (1998) Over 400 C Under 350 C Material degradation issues Swelling, Irradiation Creep, Embrittlement due to precipitates Database from FBR, Fusion Reactor IASCC, Radiation hardening, Embrittlement due to dislocation loops Database from LWR

6 Frame work Technical survey of current status -Technical issues relating to fuel cladding -Surveillance of promising materials Planning of R&D program Preparation of examinations Test loop design and manufacturing Electron irradiation tests -Swelling -Embrittlement Examination Material Preparation & Mechanical Property Tests Corrosion tests -General corrosion -SCC *) Overall Evaluation -Proposal of Candidate Materials -Clarification of R&D Issues in Next Phase *)SCC: Stress Corrosion Cracking

7 Schedule Item FY Literature Survey Planning Preparation of examinations Test loop Materials, Mechanical properties Corrosion Tests General Corrosion SCC Irradiation Tests Overall evaluation

8 Mechanical properties (High temp.) Literature survey (Creep rupture strength) Boiler piping materials(austenitic, Ferritic steel) Comparison of creep strength at 600 C, 10 5 h Creep strength: High Ni alloy Ferritec Austenitic

9 General tendency of swelling Literature Survey Swelling Ferritic/martensitic steel< Austenitic steel

10 Swelling (Austenitic steel, Ni base alloy) Literature survey Optimization of chemical composition for swelling Swelling resistance(austenitic steel & Ni Alloys) Optimize Ni concentration

11 Candidate austenitic steel for FBR Improvement of swelling resistance (steel316) =Cold work, P, Ti, Nb, and B addition

12 Radiation embrittlement in ferritic steel Literature survey Irradiated ferritic steel Decrease of the upper shelf energy DBTT shift Radiation Embrittlement DBTT: Ductile brittle transient temperature

13 Candidate ferritic steel for FBR Oxide dispersion strengthened ferritic steel (ODS) Creep rupture stress (MPa) ODS alloy Rupture time (h)

14 Corrosion data (Steel, Ni base alloy, Ti alloy) Literature Survey Test condition: 550 C, 25 MPa Corrosion Resistance: Ferritic steel Austenitic steel Ni Alloys Better-> (Ti alloys: depend on their chemical compositions)

15 Selection of test Materials (1) (commercial alloys) Supercritical Fossil Fired Power Plants(High Temp. Strength, Creep) Supercritical Water Oxidation Plants (Corrosion)= Waste decomposition Nuclear Power Plants(Radiation Damage) Austenitic Steel(7) steel316l(commercial, Modified), steel316, steel304, steel310s(commercial, Modified), steel304h Ferittic Steel(2) 12Cr-1Mo-1WVNb(HCM12), Mod. 9Cr-1Mo Nickel Base Alloy(8) Alloy 600, Alloy 625, Hastelloy C276, Alloy825, Hastelloy C22, Alloy 800H, Alloy 690, Alloy 718 Titanium Base Alloy(4) Ti-6Al-4V, Ti-3Al-2.5V, Ti-15V-3Al-3Sn-3Cr, Ti-15Mo-5Zr-3Al

16 Selection of Test Materials (2) (developed alloys) Some developed materials (mainly nuclear fields) will be examined to evaluate their viability as candidate materials for SCPR fuel claddings. Austenitic stainless steel Modified 316L (Zr added) Ultra fine grained steel PCA (Primary candidate alloys)(if available) Ferritic stainless steel F82H(Low radioactivation steel: JAERI) ODS ferritic/martensitic steel (JNC) JFMS (Japan Ferritic Maltensitic Steel) for fusion reactors (DEMO reactor) (If available)

17 Screening of the candidate materials Screening of test materials after irradiation and corrosion test Alloy design (modification) on the screened materials (Improve the properties to meet the requirements of the cladding design) Fossil fired plants materials Stainless steel Ni base alloy Commercial alloys Frame work of screening Irradiation Test SCWO plants materials Ni base alloy Ti base alloy Nuclear materials Stainless steel Screening of promising alloy Alloy design Irradiation Test (Corrosion Tests) The candidate alloys for SCPR fuel claddings will be proposed. Proposal of the candidate materials Plant design Fuel assembly design

18 General corrosion Correlation between design and material properties Oxide film Characterization Thinning SCPR system design (Radioactive CP reduction/ Optimized water chemistry) Thermal hydraulic characterization Structural reliability SCC (IASCC) Radiation effect Mechanical properties steelceptibility Swelling Embrittlement High Temp. Strength Creep Rupture (Literature Survey) Optimized hoop stress (PCI, inner gas pressure) Dimensional accuracy Thermal shock resistance Optimized configuration (Keeping coolant path) Easy handling Structural reliability Fuel assembly design

19 Mechanical properties at 550 C Strain rate /sec Yield Tensile Yield stress, Tensile stress (MPa) Total elongation (%) High Strength at 550 C Ferritic steel Austenitic steel Ni base alloys (Ti base alloy: depends on chemical compositions and heat treatment)

20 Electron irradiation test (1) Electron irradiation -> Microstructure observation Test condition C x 5 dpa (1000keV electron irradiation) Void formation=>swelling Precipitates formation=>embrittlement (mm) Irrad. area High voltage electron microscope (Electron irradiation):hokkaido Univ. Transmission electron microscope (Microstructure observation)

21 Electron irradiation test (2) TP: steel304 Temp:550 C Damage Rate:2x10-3 dpa/s 0.03 dpa 0.5 dpa 1 dpa 1.6 dpa 4.4 dpa 5.4 dpa 100 nm

22 Microstructural observation(1) Austenitic stainless steels irradiated with electrons nm 100 nm 100 nm 316L 100 nm 100 nm 100 nm

23 Microstructural observation(2) Austenitic stainless steels irradiated with electrons nm 100 nm 100 nm 310S 100 nm 100 nm 100 nm

24 Microstructural observation(3) High Ni austenitic alloys irradiated with electrons(1) Alloy nm 100 nm 100 nm 100 nm Alloy 800H 100 nm 100 nm 100 nm

25 Microstructural observation(4) High Ni austenitic alloys irradiated with electrons(2) Hastelloy C nm 100 nm 100 nm Hastelloy C nm 100 nm 100 nm

26 Microstructural observation(5) Ni base alloys irradiated with electrons Alloy nm 100 nm 100 nm Alloy nm 100 nm 100 nm

27 Electron irradiation test (3) Microstructure observation => Void formation 316L 310S Alloy625 Alloy825 HastelloyC C 450 C Higher Ni Ni concentration =Better void swelling resistance 100 nm

28 Current status of the material database Comparison of materials (Blue: Advantage, Red: Disadvantage) Austenitic SUS Ferritic SUS Ni base alloy (High Ni SUS) Ti base alloy Corrosion Better (A&B) Good (A) Excellent (B) Good~ Excellent (B) Radiation damage Lots of experience Good phase stability High Swelling (Modification required) Low Swelling Low radioactivation Good phase stability DBTT shift(high Temp) (Precipitates?) Low Swelling High radioactivation He embrittlement (Precipitates?) Mechanical Properties (High temperature) Improvement required DBTT shift(low Temp) Improvement required Good~Excellent*) Cost Low Low Middle~ Expensive (Limited DB) Good~Excellent *) Expensive A: Fossil Fired Plant B: SCWO *)Depend on the chemical compositions and thermal treatment

29 Summary Literature survey It was almost finished, and the materials for the tests have been selected. Planning The subjects of the development program was clarified. Test matrix of irradiation test and corrosion test were decided. Preparation of examination Loop facility for corrosion test was designed and manufactured. The materials for the tests have been purchased. Irradiation Test (Simulation by electron irradiation) The tests have been started. The materials containing higher Ni tend to suppress the void formation.