1

PHENIX PFC evaluation by tritium Plasma, HEat and Neutron Irradiation eXperiments

2013~2018Japan USA

Representative Yoshio Ueda (Osaka U.) Peter Pappano (DOE)Coordinator Yuji Hatano (U. Toyama) Peter Pappano (DOE)

Task 1Takehiko Yokomine (Kyoto U.)

Yoshio Ueda (Osaka U.)Lance Snead (ORNL)Richard Nygren (SNL)

Task 2Tatsuya Hinoki (Kyoto U.)

Akira Hasegawa (Tohoku U.)Yutai Katoh(ORNL)

Task 3Yasuhisa Oya (Shizuoka U.)

Yuji Hatano (U. Toyama)Brad Merrill (INL)

Dean Buchenauer (SNL)

Japan-US Joint Project Technological Assessment of Plasma

Facing Components for DEMO Reactors

2Summary of the project

The goal of this project is to evaluate the feasibility of He gas-cooleddivertor with neutron irradiated tungsten armor materials for DEMOreactors. Main research subjects are listed below;

1. Heat transfer mechanism and modeling in He-cooled systems,improvement of cooling efficiency and system design.

2. Response of tungsten layered materials and advanced tungstenmaterials to steady state and pulsed heat loads.

3. Thermo-mechanical properties measurement of tungsten basedmaterials after neutron irradiation at elevated temperatures relevantto divertor conditions (500-1200 oC).

4. Effects of high flux plasma exposure on tritium behavior in neutron-irradiated tungsten materials.

5. Evaluation of feasibility (under ~10 MW/m2 heat load with irradiationof plasma and neutrons) and safety (tritium retention andpermeation) of He-cooled PFCs and clarification of critical issues forDEMO divertor design.

3Structure 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

4

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

5Task 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

6Task2 : 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

7Task3 : 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