Fusion

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Fusion

Fusion

Plasma Facing Materials in Nuclear Fusion

 

  Plasma facing materials (PFMs) have been studied especially in the interests of International Thermonuclear Experimental Reactor (ITER) application. As a member of ITER organization, Republic of Korea has contributed many years to the intensive R&D programs of tungsten as PFM against high plasma thermal and irradiation loadings in nuclear fusion reactors. The most updated research is ongoing in the interests of high ductility W alloy development with a close cooperation with National Fusion Research Institute (NFRI) and Seoul National University (SNU).

 

 

Research Key Words and Materials

 

Physical vapor deposition (PVD) coating

Plasma thermal loading

Deuterium irradiation

W alloy development

Thermo-mechanical treatment

(TMT) process

High T mechanical property analysis

Tungsten (W)

Graphite

 

 

 

Related Project

(I) Investigation on basic technologies for high toughness tungsten alloys for ITER Divertor application(NFRI)

(II)Tungsten coating as plasma facing material (NFRI)

 

*Related Project I

Project Title:

Investigation on basic technologies for high toughness tungsten alloys for ITER Divertor application

 

Project Period:

2016.01.01 - 2018.12.31

 

Project Purpose:

(1) Alloy development of high toughness W alloys for ITER divertor application

(2) Development of thermo-mechanical treatment for toughness improvement of ITER W alloys

(3) Development of life-time evaluating program

 

Brief Description:

  Bulk tungsten (W) has been selected as ITER divertor structural material, yet improvement of high temperature toughness is currently in demand. Development of high toughness W alloys is important in which implementation of post thermo-mechanical treatments (TMTs) is also being considered to improve mechanical properties of developed alloys.

   In this laboratory, development of W alloys has been investigated to improve high temperature toughness for ITER divertor application. Sintering alloying and addition of minor elements including Ti, Re, Y, K are particularly interested in the design process. In addition to the W alloys design, TMTs have been applied on various grade W alloys including ITER grade to improve microstructure as well as high temperature mechanical properties.

 

*Related Project II

Project Title:

Technology Follow-up and Development for ITER Plasma Facing Material

 

Project Period:

2013.01.01 - 2015.12.31

 

Project Purpose:

(1) Recent technology follow-up for ITER plasma facing material

(2) Development of nano-structured tungsten technique by PF-PVD

(3) Development of life-time evaluating program

 

Brief Description:

  First wall and divertor are the plasma facing components (PFC) of the nuclear fusion reactors like ITER. These components will be exposed to plasma particles (to high temperature and energy) during the operation period. As a result, surface of the PFC would be damaged and loss of thickness would be happened. Therefore, the technology of PFC material is essential to determine the life time and the capacity of nuclear fusion reactors. Although beryllium (Be), carbon fiber composite (CFC), and tungsten (W) are used in present nuclear fusion experimental reactors, Be and CFC are thought to be replaced to W in ITER and DEMO reactor due to high sputter resistance of W than that of low atomic number materials such as Be and CFC. Nevertheless, brittle characteristic of W is a main problem to be worked out. Small grain and carbide distribution in W are helpful to solve this problem.

   In this laboratory, technology follow-up for ITER plasma facing material has been investigated since 2012. As sequence of this, development and evaluation of tungsten coating technique with nano-structured W and carbide distributed W has also conducted to improve brittle resistance of W.

 

Multilayered PVD/PS W coating behavior under plasma thermal loading (left) and deuterium irradiation (right)

 



RESEARCH

LWR
Gen-IV
Fusion
 




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