Retention and surface blistering of helium irradiated tungsten as a first wall material

S. B. Gilliam; S. M. Gidcumb; N. R. Parikh; D. G. Forsythe; B. K. Patnaik; J. D. Hunn; L. L. Snead; G. P. Lamaze

doi:10.1016/j.jnucmat.2005.08.017

Retention and surface blistering of helium irradiated tungsten as a first wall material

S. B. Gilliam, S. M. Gidcumb, N. R. Parikh, D. G. Forsythe, B. K. Patnaik, J. D. Hunn, L. L. Snead, G. P. Lamaze

Particle Fuel Forms

Research output: Contribution to journal › Article › peer-review

96 Scopus citations

Abstract

The first wall of an inertial fusion energy reactor may suffer from surface blistering and exfoliation due to helium ion irradiation and extreme temperatures. Tungsten is a candidate for the first wall material. A study of helium retention and surface blistering with regard to helium dose, temperature, pulsed implantation, and tungsten microstructure was conducted to better understand what may occur at the first wall of the reactor. Single crystal and polycrystalline tungsten samples were implanted with 1.3 MeV ³He in doses ranging from 10¹⁹ m^-2 to 10²² m ^-2. Implanted samples were analyzed by ³He(d,p) ⁴He nuclear reaction analysis and ³He(n,p)T neutron depth profiling techniques. Surface blistering was observed for doses greater than 10²¹ He/m². For He fluences of 5 × 10²⁰ He/m², similar retention levels in both microstructures resulted without blistering. Implantation and flash heating in cycles indicated that helium retention was mitigated with decreasing He dose per cycle.

Original language	English
Pages (from-to)	289-297
Number of pages	9
Journal	Journal of Nuclear Materials
Volume	347
Issue number	3
DOIs	https://doi.org/10.1016/j.jnucmat.2005.08.017
State	Published - Dec 15 2005

Funding

This work was supported under the US Department of Energy High Average Power Laser Program managed by the Naval Reactor Laboratory through subcontract with the Oak Ridge National Laboratory. Also, the authors would like to thank Dr Hugon Karwowski for his helpful discussions with this project.

Funders	Funder number
Naval Reactor Laboratory
U.S. Department of Energy
Oak Ridge National Laboratory

UN SDGs

This output contributes to the following UN Sustainable Development Goals (SDGs)

Access to Document

10.1016/j.jnucmat.2005.08.017

Cite this

@article{6f20196ed03e4764a4f3204f09d68cd8,

title = "Retention and surface blistering of helium irradiated tungsten as a first wall material",

abstract = "The first wall of an inertial fusion energy reactor may suffer from surface blistering and exfoliation due to helium ion irradiation and extreme temperatures. Tungsten is a candidate for the first wall material. A study of helium retention and surface blistering with regard to helium dose, temperature, pulsed implantation, and tungsten microstructure was conducted to better understand what may occur at the first wall of the reactor. Single crystal and polycrystalline tungsten samples were implanted with 1.3 MeV 3He in doses ranging from 1019 m-2 to 1022 m -2. Implanted samples were analyzed by 3He(d,p) 4He nuclear reaction analysis and 3He(n,p)T neutron depth profiling techniques. Surface blistering was observed for doses greater than 1021 He/m2. For He fluences of 5 × 1020 He/m2, similar retention levels in both microstructures resulted without blistering. Implantation and flash heating in cycles indicated that helium retention was mitigated with decreasing He dose per cycle.",

author = "Gilliam, {S. B.} and Gidcumb, {S. M.} and Parikh, {N. R.} and Forsythe, {D. G.} and Patnaik, {B. K.} and Hunn, {J. D.} and Snead, {L. L.} and Lamaze, {G. P.}",

year = "2005",

month = dec,

day = "15",

doi = "10.1016/j.jnucmat.2005.08.017",

language = "English",

volume = "347",

pages = "289--297",

journal = "Journal of Nuclear Materials",

issn = "0022-3115",

publisher = "Elsevier",

number = "3",

}

TY - JOUR

T1 - Retention and surface blistering of helium irradiated tungsten as a first wall material

AU - Gilliam, S. B.

AU - Gidcumb, S. M.

AU - Parikh, N. R.

AU - Forsythe, D. G.

AU - Patnaik, B. K.

AU - Hunn, J. D.

AU - Snead, L. L.

AU - Lamaze, G. P.

PY - 2005/12/15

Y1 - 2005/12/15

N2 - The first wall of an inertial fusion energy reactor may suffer from surface blistering and exfoliation due to helium ion irradiation and extreme temperatures. Tungsten is a candidate for the first wall material. A study of helium retention and surface blistering with regard to helium dose, temperature, pulsed implantation, and tungsten microstructure was conducted to better understand what may occur at the first wall of the reactor. Single crystal and polycrystalline tungsten samples were implanted with 1.3 MeV 3He in doses ranging from 1019 m-2 to 1022 m -2. Implanted samples were analyzed by 3He(d,p) 4He nuclear reaction analysis and 3He(n,p)T neutron depth profiling techniques. Surface blistering was observed for doses greater than 1021 He/m2. For He fluences of 5 × 1020 He/m2, similar retention levels in both microstructures resulted without blistering. Implantation and flash heating in cycles indicated that helium retention was mitigated with decreasing He dose per cycle.

AB - The first wall of an inertial fusion energy reactor may suffer from surface blistering and exfoliation due to helium ion irradiation and extreme temperatures. Tungsten is a candidate for the first wall material. A study of helium retention and surface blistering with regard to helium dose, temperature, pulsed implantation, and tungsten microstructure was conducted to better understand what may occur at the first wall of the reactor. Single crystal and polycrystalline tungsten samples were implanted with 1.3 MeV 3He in doses ranging from 1019 m-2 to 1022 m -2. Implanted samples were analyzed by 3He(d,p) 4He nuclear reaction analysis and 3He(n,p)T neutron depth profiling techniques. Surface blistering was observed for doses greater than 1021 He/m2. For He fluences of 5 × 1020 He/m2, similar retention levels in both microstructures resulted without blistering. Implantation and flash heating in cycles indicated that helium retention was mitigated with decreasing He dose per cycle.

UR - http://www.scopus.com/inward/record.url?scp=28244495550&partnerID=8YFLogxK

U2 - 10.1016/j.jnucmat.2005.08.017

DO - 10.1016/j.jnucmat.2005.08.017

M3 - Article

AN - SCOPUS:28244495550

SN - 0022-3115

VL - 347

SP - 289

EP - 297

JO - Journal of Nuclear Materials

JF - Journal of Nuclear Materials

IS - 3

ER -

Retention and surface blistering of helium irradiated tungsten as a first wall material

Abstract

Funding

UN SDGs

Access to Document

Other files and links

Fingerprint

Cite this