structure of this project 1...structure of this project 1 task 3 (tpe, idaho nl) permeation devices...
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1Structure of this Project
Task 3(TPE, Idaho NL)
Permeation devices (INL, SNL)
Tritium Behaviorin n damaged W
Task 2(HFIR, Oak Ridge NL)Neutron-irrad. Effects Physical PropertiesThermo mechanical
Task 1(PAL facility, ORNL)
(He loop,GIT)Heat Load Tests
Heat TransferSystem Evaluation
MaterialProperties(n-irradiated sample )
TritiumBehavior
Neutron-irradiated samples
2
Incident plasma particles(T, D, He, Electrons)Neutron + Radiation
Armor
Melting layer(liquid
surface)
Eroded layer
Solid particles
He inflow He outflowHe outflow
CartridgeCap
Liquid droplets
熱機械特性に及ぼす
Research Subjects for Task1, Task2, Task3
Property Improvement of W-alloyAssessment of Complex
Heat Load Effect Assessment of Tritium Retention in Armor Material
Neutron Irradiation Effect on
Thermomechanical Property
Development of Bonding between
Armor and Structural Materials
Assessment of Soundness of Sealed Structure with High
Pressure/High Temperature Helium
Understanding of Heat Transfer Mechanism
Optimization of Cooling Configuration
Assessment of Tritium Permeation to
Coolant
Task2Task3
Task1Task1
Task1
Task2
Task2 Task3
3Task 1:Investigation of Overall Heat Flow Response in PFC
PFM
Structural Material
Coolant
PlasmaHeat flux
Thermal conductivity
Heat transfer coefficient
Heat Flow Analysis
Safety (Tritium)
Thermal-mechanical analysis
He gas300~600oC~10MPa
SS thermal load
5~10MW/m2
Modeling
He gas impingingJET
Heat load test ~10 MW/m2
∇T -> Thermal stressThermal fatigue
W based materials,W layered materials
(n- irradiated)
Integrate results from Task1, Task2, and Task 3
Thermal resistance
material property
4Task2 : Material Performance - Evaluation & Irradiation-
Post Irradiation Experiment
Post irradiation experiment• LAMDA, IMET: thermal conductivity, toughness, strength, microstructure• PAL Facility: Thermal cycle (Task 1)• Oarai, Japan: supporting PIE
Heat load
Material selection and modification• W based material, advanced W
material including alloy and composites
• Layered material• Information exchange and
feedback for development
Neutron Irradiation (HFIR)• Temperature: 500℃~1200℃, Dose : ~1.5 dpa• Rabbit : no shield, low dose• RB* : Thermal neutron shield to be installed• High dose(~15 dpa): Utilizing ion irradiation
5Task3 : Tritium behavior in tungsten at high temperatures
Retention and permeation of hydrogen including tritium in tungstenare studied in high temperature conditions after neutron irradiation.
Neutron irrad.
Ion irrad.
Surface
Defe
ctdi
strib
utio
nDepth
Neutron irradiation at HFIR and sample transport to INL
Tritium diffusion coefficient Trap density(Tritium
retention) Actvation energy for
desorption Recombination coefficient
T T
TT
T
T
TT
TT
T
TT
T
T
T
TT
TT
T
T
T TT
TT
TTPE
Diffusion behavior
Permeation behavior(TPE)
High Temp. Neutron irradiated sample
Surface Bulk
Microstructure observation(TEM)
Retention behavior (TDS)
Permeation holder designed by US side for in-situ permeation study
Tritium retention and permeation rate evaluation by modeling and simulation (TMAP etc)
Safety analysis
TDS apparatus