IAE R&D Program Progress ReportDevelopment Project of Supercritical-water Cooled Power Reactors
Overview, Irradiation Testand Mechanical Property Test
Shigeki Kasahara
Hitachi, Ltd.Toshiba Corp.Hokkaido Univ.Univ. of Tokyo
SCPR:Supercritical-water Cooled Power Reactor
Requirement on the Cladding MaterialsReliability, Durability, and Economy
Radiation effectSwelling, Creep, Embrittlement,Radioactivation
Productivity, Cost
CorrosionGeneral
Stress CorrosionCracking
(SCC)
High temp. strengthTensile, Creep(Unirrad.)
Neutron economy(Neutron Absorption)
Design of SCPR fuel assembly
Coolant flow: Outside of claddings, narrow gap (<1 mm)Bundle, Spacer
Specification
ConfigurationRod pitch9.5 mm
Rod dia. φ8 mm
Cladding
Thickness：0.8mm
Coolant(Supercritical
Water)
Fuel assembly : Dimensional accuracy required↓ ↓
Materials : Low swelling, high creep strength,corrosion resistance necessary
SCPR core environment (1)Neutron spectrum and flux(Case of thermal neutron SCPR: Similar to APWR)
→Fluence (Core)2x1025 n/m2/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)
SCPR core environment (2)Temperature: 300°C ~ 550°C Pressure: 25 MPa
Axial NodeKazuaki Kitou : Safety Analysis of a Supercritical Water Cooled Reactor :Ph. D. dissertation, The University of Tokyo (1998)
Material degradation issues
Swelling, Irradiation Creep,Embrittlement
due to precipitatesDatabase from FBR, Fusion Reactor
Over 400°C Under 350°C IASCC,
Radiation hardening,Embrittlement due to
dislocation loopsDatabase from LWR
Max. Temp.(Current design)
Frame work
Technical survey of current status-Technical issues relating to fuel cladding
-Surveillance of promising materials
Planning of R&D program
Preparation of examinationsMaterial Preparation &
Mechanical Property TestsTest loop design and
manufacturing
ExaminationElectron irradiation tests
-Swelling-Embrittlement
Corrosion tests-General corrosion
-SCC*)
Overall Evaluation-Proposal of Candidate Materials
-Clarification of R&D Issues in Next Phase*)SCC: Stress Corrosion Cracking
Schedule
200520022001
Literature Survey
Planning
Preparation of examinations
Corrosion Tests
Irradiation Tests
Overall evaluation
20042000FY
Item
Test loop
Materials, Mechanical properties
General CorrosionSCC
Mechanical properties (High temp.)Literature survey (Creep rupture strength)
Boiler piping materials(Austenitic, Ferritic steel)Comparison of creep strength at 600°C, 105 h
Creep strength: High Ni alloy ≥ Ferritec ≈ Austenitic
General tendency of swellingLiterature Survey
Swelling⇒Ferritic/martensitic steel< Austenitic steel
Swelling (Austenitic steel, Ni base alloy)Literature survey
Optimization of chemical composition for swelling
Swelling resistance(Austenitic steel & Ni Alloys)⇒ Optimize Ni concentration
Candidate austenitic steel for FBR
Improvement of swelling resistance (steel316)=Cold work, P, Ti, Nb, and B addition
Radiation embrittlement in ferritic steelLiterature surveyIrradiated ferritic steel ⇒
Decrease of the upper shelf energyDBTT shift
RadiationEmbrittlement
DBTT: Ductile brittle transient temperature
Candidate ferritic steel for FBROxide dispersion strengthened ferritic steel (ODS)
Cre
ep ru
ptur
e st
ress
(MPa
)
Rupture time (h)
ODS alloy
Corrosion data (Steel, Ni base alloy, Ti alloy)Literature Survey
Test condition: 550°C, 25 MPa
Corrosion Resistance: Ferritic steel Austenitic steel Ni Alloys
(Ti alloys: depend on their chemical compositions)Better->
Selection of test Materials (1) (commercial alloys)
Supercritical Fossil Fired Power Plants(High Temp. Strength, Creep)Supercritical Water Oxidation Plants (Corrosion)= Waste decompositionNuclear 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
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)
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)
The candidate alloys for SCPR fuel claddings will be proposed. Plant design Fuel assembly design
Frame work of screening
Irradiation Test
Nuclear materials
Stainless steel
Fossil fired plants materials
Stainless steelNi base alloy
SCWO plants materials
Ni base alloyTi base alloy
Commercial alloys
Screening of promising alloy＋ Alloy design
Proposal of the candidate materials
(Corrosion Tests)
Irradiation Test
Correlation between design and material properties
Thermal shock resistance
Easy handlingStructural reliability
Optimized configuration(Keeping coolant path)
Optimized hoop stress(PCI, inner gas pressure)
Thermal hydraulic characterizationStructural reliability
Dimensional accuracy
Oxide filmCharacterization
steelceptibility
Swelling
(Radioactive CP reduction/Optimized water chemistry)
SCPR system designGeneral
corrosionThinning
Fuel
ass
embl
y de
signSCC
(IASCC)
Radiationeffect Embrittlement
High Temp.StrengthMechanical
properties Creep Rupture(Literature Survey)
Mechanical properties at 550°CStrain rate：5×10-3 /sec
Yield stress, Tensile stress (MPa) Total elongation (%)
TensileYield
High Strength at 550°C⇒Ferritic steel＜Austenitic steel＜Ni base alloys(Ti base alloy: depends on chemical compositions and heat treatment)
Electron irradiation test (1)Electron irradiation -> Microstructure observation
Test condition：290，450，550°C x 5 dpa (1000keV electron irradiation)
Void formation=>SwellingPrecipitates formation=>Embrittlement
3Φ
0.15(mm)
High voltage electron microscope(Electron irradiation):Hokkaido Univ.
Transmission electron microscope
(Microstructure observation)
Irrad. area
Electron irradiation test (2)TP: steel304 Temp:550°C Damage Rate:2x10-3 dpa/s
100 nm
0.03 dpa 0.5 dpa 1 dpa
5.4 dpa4.4 dpa1.6 dpa
Microstructural observation(1)Austenitic stainless steels irradiated with electrons-1
290℃ 450℃ 550℃
316L
316
100 nm100 nm 100 nm
100 nm100 nm 100 nm
290℃ 450℃ 550℃
310S
304
100 nm
100 nm
100 nm 100 nm
100 nm 100 nm
Microstructural observation(2)Austenitic stainless steels irradiated with electrons-2
Microstructural observation(3)High Ni austenitic alloys irradiated with electrons(1)
290℃ 450℃ 550℃
Alloy 800H
Alloy 825100 nm100 nm100 nm100 nm
100 nm100 nm100 nm
100 nm 100 nm 100 nm
Microstructural observation(4)High Ni austenitic alloys irradiated with electrons(2)
290℃ 450℃ 550℃
HastelloyC276
HastelloyC22
100 nm 100 nm 100 nm
100 nm 100 nm 100 nm
Microstructural observation(5)Ni base alloys irradiated with electrons
290℃ 450℃ 550℃
Alloy 625
Alloy 600
100 nm100 nm 100 nm
100 nm 100 nm 100 nm
Electron irradiation test (3)Microstructure observation => Void formation
100 nm
450°
C55
0°C
HastelloyC276Alloy825Alloy625316L 310S
Higher Ni concentration=Better void swelling resistance
Higher Ni concentration=Better void swelling resistance
Current status of the material database●Comparison of materials (Blue: Advantage, Red: Disadvantage)
Corrosion Radiation damage Mechanical Properties (High temperature) Cost
Austenitic SUS Better (A&B)
Lots of experience Good phase stability
High Swelling
(Modification required)
Improvement required
Low
Ferritic SUS Good (A)
Low Swelling Low radioactivation Good phase stability
DBTT shift(High Temp)
(Precipitates?)
DBTT shift(Low Temp) Improvement required
Low
Ni base alloy (High Ni SUS)
Excellent (B)
Low Swelling
High radioactivation He embrittlement
(Precipitates?)
Good~Excellent*)
Middle~ Expensive
Ti base alloy Good~
Excellent (B)
(Limited DB) Good~Excellent *) Expensive
A: Fossil Fired Plant B: SCWO *)Depend on the chemical compositions and thermal treatment
SummaryLiterature survey
It was almost finished, and the materials for the tests have been selected.
PlanningThe subjects of the development program was clarified.Test matrix of irradiation test and corrosion test were decided.
Preparation of examinationLoop 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.
