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Title: Using the Tritium Plasma Experiment to evaluate ITER PFC safety

Abstract

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 {times} 10{sup 19} ions/cm{sup 2} {center_dot} s and a plasma temperature of about 15 eV using a plasma that includes tritium. With the closure of the Tritium Research Laboratory at Livermore, the experiment was moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory. An experimental program has been initiated there using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. A considerable lack of data exists in these areas for many of the materials, especially beryllium, being considered for use in ITER. Not only will basic material behavior with respect to safety issues in the divertor environment be examined, but innovative techniques for optimizing performance with respect to tritiummore » safety by material modification and process control will be investigated. Supplementary experiments will be carried out at the Idaho National Engineering Laboratory and Sandia National Laboratory to expand and clarify results obtained on the Tritium Plasma Experiment.« less

Authors:
;  [1];  [2];  [3];  [4]
  1. EG and G Idaho, Inc., Idaho Falls, ID (US)
  2. Los Alamos National Lab., NM (US)
  3. Sandia National Labs., Livermore, CA (US)
  4. MDC Aerospace, St. Louis, MO (US)
Publication Date:
Research Org.:
Los Alamos National Lab., NM (United States); EG and G Idaho, Inc., Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE, Washington, DC (United States)
OSTI Identifier:
10169794
Report Number(s):
LA-UR-93-2148; CONF-9306191-1
ON: DE93016614; TRN: 93:016589
DOE Contract Number:  
W-7405-ENG-36; AC07-76ID01570
Resource Type:
Conference
Resource Relation:
Conference: International Atomic Energy Agency (IAEA) technical committee meeting on developments in fusion safety,Ontario (Canada),7-11 Jun 1993; Other Information: PBD: 1993
Country of Publication:
United States
Language:
English
Subject:
70 PLASMA PHYSICS AND FUSION TECHNOLOGY; ITER TOKAMAK; DIVERTORS; SAFETY ANALYSIS; TRITIUM SYSTEMS TEST ASSEMBLY; SAFETY; EVALUATION; TRITIUM; DIFFUSION; PERMEABILITY; THERMONUCLEAR REACTOR COOLING SYSTEMS; EROSION; THERMONUCLEAR REACTOR MATERIALS; PERFORMANCE; PLASMA; 700420; PLASMA-FACING COMPONENTS

Citation Formats

Longhurst, G R, Anderl, R A, Bartlit, J R, Causey, R A, and Haines, J R. Using the Tritium Plasma Experiment to evaluate ITER PFC safety. United States: N. p., 1993. Web.
Longhurst, G R, Anderl, R A, Bartlit, J R, Causey, R A, & Haines, J R. Using the Tritium Plasma Experiment to evaluate ITER PFC safety. United States.
Longhurst, G R, Anderl, R A, Bartlit, J R, Causey, R A, and Haines, J R. Thu . "Using the Tritium Plasma Experiment to evaluate ITER PFC safety". United States. https://www.osti.gov/servlets/purl/10169794.
@article{osti_10169794,
title = {Using the Tritium Plasma Experiment to evaluate ITER PFC safety},
author = {Longhurst, G R and Anderl, R A and Bartlit, J R and Causey, R A and Haines, J R},
abstractNote = {The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 {times} 10{sup 19} ions/cm{sup 2} {center_dot} s and a plasma temperature of about 15 eV using a plasma that includes tritium. With the closure of the Tritium Research Laboratory at Livermore, the experiment was moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory. An experimental program has been initiated there using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. A considerable lack of data exists in these areas for many of the materials, especially beryllium, being considered for use in ITER. Not only will basic material behavior with respect to safety issues in the divertor environment be examined, but innovative techniques for optimizing performance with respect to tritium safety by material modification and process control will be investigated. Supplementary experiments will be carried out at the Idaho National Engineering Laboratory and Sandia National Laboratory to expand and clarify results obtained on the Tritium Plasma Experiment.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1993},
month = {7}
}

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