Tritium Decay Helium-3 Effects in Tungsten
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
A critical challenge for long-term operation of ITER and beyond to a Demonstration reactor (DEMO) and future fusion reactor will be the development of plasma-facing components (PFCs) that demonstrate erosion resistance to steady-state/transient heat fluxes and intense neutral/ion particle fluxes under the extreme fusion nuclear environment, while at the same time minimizing in-vessel tritium inventories and permeation fluxes into the PFC’s coolant. Tritium will diffuse in bulk tungsten at elevated temperatures, and can be trapped in radiation-induced trap site (up to 1 at. % T/W) in tungsten [1,2]. Tritium decay into helium-3 may also play a major role in microstructural evolution (e.g. helium embrittlement) in tungsten due to relatively low helium-4 production (e.g. He/dpa ratio of 0.4-0.7 appm [3]) in tungsten. Tritium-decay helium-3 effect on tungsten is hardly understood, and its database is very limited. Two tungsten samples (99.99 at. % purity from A.L.M.T. Co., Japan) were exposed to high flux (ion flux of 1.0x1022 m-2s-1 and ion fluence of 1.0x1026 m-2) 0.5%T2/D2 plasma at two different temperatures (200, and 500°C) in Tritium Plasma Experiment (TPE) at Idaho National Laboratory. Tritium implanted samples were stored at ambient temperature in air for more than 3 years to investigate tritium decay helium-3 effect in tungsten. The tritium distributions on plasma-exposed was monitored by a tritium imaging plate technique during storage period [4]. Thermal desorption spectroscopy was performed with a ramp rate of 10°C/min up to 900°C to outgas residual deuterium and tritium but keep helium-3 in tungsten. These helium-3 implanted samples were exposed to deuterium plasma in TPE to investigate helium-3 effect on deuterium behavior in tungsten. The results show that tritium surface concentration in 200°C sample decreased to 30 %, but tritium surface concentration in 500°C sample did not alter over the 3 years storage period, indicating possible tritium retention in helium-3 bubble. This paper reports the initial experimental observation of tritium-decay helium-3 in tungsten exposed to deuterium/tritium plasma along with electron microscope analysis and also discusses a Tritium Migration Analysis Program (TMAP) analysis of tritium-decay helium-3 effects on tritium retention in tungsten for DEMO and future fusion reactor. [1] Y. Hatano, et.al., Nucl. Fusion 53 (2013) 073006 [2] M. Shimada, et.al., Nucl. Fusion 55 (2015) 013008 [3] M. Sawan, Fus. Sci. Technol. 66 (2014) 272 [4] T. Otsuka, Fus. Sci. Technol. 60 (2011) 1539 This work was prepared for the U.S. Department of Energy, Office of Fusion Energy Sciences, under the DOE Idaho Field Office contract number DE-AC07-05ID14517.
- Research Organization:
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- USDOE Office of Science (SC), Fusion Energy Sciences (FES)
- DOE Contract Number:
- AC07-05ID14517
- OSTI ID:
- 1358199
- Report Number(s):
- INL/CON-15-37549
- Resource Relation:
- Conference: 22nd International Conference on Plasma Surface Interactions in Controlled Fusion Devices, Rome, Italy, May 30–June 3, 2016
- Country of Publication:
- United States
- Language:
- English
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