Efficiency of thermal outgassing for tritium retention measurement and removal in ITER

G. De Temmerman, M.J. Baldwin, D. Anthoine, K. Heinola, A. Jan, I. Jepu, J. Likonen, C.P. Lungu, C. Porosnicu, R.A. Pitts

Research output: Contribution to journalArticleScientificpeer-review

18 Citations (Scopus)

Abstract

As a licensed nuclear facility, ITER must limit the in-vessel tritium (T) retention to reduce the risks of potential release during accidents, the inventory limit being set at 1. kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be) eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID) system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513. K for the FW and 623. K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure) co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.
Original languageEnglish
Pages (from-to)267-272
Number of pages6
JournalNuclear Materials and Energy
Volume12
DOIs
Publication statusPublished - 1 Aug 2017
MoE publication typeA1 Journal article-refereed

Fingerprint

Tritium
outgassing
Degassing
tritium
Deposits
deposits
baking
desorption
Desorption
Beryllium
Facings
Thermal desorption
baffles
accidents
beryllium
Pulsed lasers
Extrapolation
vessels
monitors
extrapolation

Keywords

  • Tritium retention
  • ITER
  • divertor
  • outgassing

Cite this

De Temmerman, G., Baldwin, M. J., Anthoine, D., Heinola, K., Jan, A., Jepu, I., ... Pitts, R. A. (2017). Efficiency of thermal outgassing for tritium retention measurement and removal in ITER. Nuclear Materials and Energy, 12, 267-272. https://doi.org/10.1016/j.nme.2016.10.016
De Temmerman, G. ; Baldwin, M.J. ; Anthoine, D. ; Heinola, K. ; Jan, A. ; Jepu, I. ; Likonen, J. ; Lungu, C.P. ; Porosnicu, C. ; Pitts, R.A. / Efficiency of thermal outgassing for tritium retention measurement and removal in ITER. In: Nuclear Materials and Energy. 2017 ; Vol. 12. pp. 267-272.
@article{4c388ad0a21644ea8a34c2e4d3b8a1c0,
title = "Efficiency of thermal outgassing for tritium retention measurement and removal in ITER",
abstract = "As a licensed nuclear facility, ITER must limit the in-vessel tritium (T) retention to reduce the risks of potential release during accidents, the inventory limit being set at 1. kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be) eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID) system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513. K for the FW and 623. K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure) co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.",
keywords = "Tritium retention, ITER, divertor, outgassing",
author = "{De Temmerman}, G. and M.J. Baldwin and D. Anthoine and K. Heinola and A. Jan and I. Jepu and J. Likonen and C.P. Lungu and C. Porosnicu and R.A. Pitts",
year = "2017",
month = "8",
day = "1",
doi = "10.1016/j.nme.2016.10.016",
language = "English",
volume = "12",
pages = "267--272",
journal = "Nuclear Materials and Energy",
issn = "2352-1791",
publisher = "Elsevier",

}

De Temmerman, G, Baldwin, MJ, Anthoine, D, Heinola, K, Jan, A, Jepu, I, Likonen, J, Lungu, CP, Porosnicu, C & Pitts, RA 2017, 'Efficiency of thermal outgassing for tritium retention measurement and removal in ITER', Nuclear Materials and Energy, vol. 12, pp. 267-272. https://doi.org/10.1016/j.nme.2016.10.016

Efficiency of thermal outgassing for tritium retention measurement and removal in ITER. / De Temmerman, G.; Baldwin, M.J.; Anthoine, D.; Heinola, K.; Jan, A.; Jepu, I.; Likonen, J.; Lungu, C.P.; Porosnicu, C.; Pitts, R.A.

In: Nuclear Materials and Energy, Vol. 12, 01.08.2017, p. 267-272.

Research output: Contribution to journalArticleScientificpeer-review

TY - JOUR

T1 - Efficiency of thermal outgassing for tritium retention measurement and removal in ITER

AU - De Temmerman, G.

AU - Baldwin, M.J.

AU - Anthoine, D.

AU - Heinola, K.

AU - Jan, A.

AU - Jepu, I.

AU - Likonen, J.

AU - Lungu, C.P.

AU - Porosnicu, C.

AU - Pitts, R.A.

PY - 2017/8/1

Y1 - 2017/8/1

N2 - As a licensed nuclear facility, ITER must limit the in-vessel tritium (T) retention to reduce the risks of potential release during accidents, the inventory limit being set at 1. kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be) eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID) system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513. K for the FW and 623. K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure) co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.

AB - As a licensed nuclear facility, ITER must limit the in-vessel tritium (T) retention to reduce the risks of potential release during accidents, the inventory limit being set at 1. kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be) eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID) system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513. K for the FW and 623. K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure) co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.

KW - Tritium retention

KW - ITER

KW - divertor

KW - outgassing

UR - http://www.scopus.com/inward/record.url?scp=85006804232&partnerID=8YFLogxK

U2 - 10.1016/j.nme.2016.10.016

DO - 10.1016/j.nme.2016.10.016

M3 - Article

VL - 12

SP - 267

EP - 272

JO - Nuclear Materials and Energy

JF - Nuclear Materials and Energy

SN - 2352-1791

ER -