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Title: ITER First Wall Module 18 - The US Effort.

Abstract

Abstract not provided.

Authors:
; ; ; ; ; ;
Publication Date:
Research Org.:
Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA)
OSTI Identifier:
1267175
Report Number(s):
SAND2007-3056C
526402
DOE Contract Number:
AC04-94AL85000
Resource Type:
Conference
Resource Relation:
Conference: Proposed for presentation at the 7th International Symposium on Fusion Nuclear Technology held May 22-27, 2005 in Tokyo, JAPAN.
Country of Publication:
United States
Language:
English

Citation Formats

Nygren, Richard E., Ulrickson, Michael A., Martin, Tina T., Youchison, Dennis L, Lutz, Thomas J., Bullock, James H., and Hollis, Kendall J. ITER First Wall Module 18 - The US Effort.. United States: N. p., 2007. Web.
Nygren, Richard E., Ulrickson, Michael A., Martin, Tina T., Youchison, Dennis L, Lutz, Thomas J., Bullock, James H., & Hollis, Kendall J. ITER First Wall Module 18 - The US Effort.. United States.
Nygren, Richard E., Ulrickson, Michael A., Martin, Tina T., Youchison, Dennis L, Lutz, Thomas J., Bullock, James H., and Hollis, Kendall J. Tue . "ITER First Wall Module 18 - The US Effort.". United States. doi:. https://www.osti.gov/servlets/purl/1267175.
@article{osti_1267175,
title = {ITER First Wall Module 18 - The US Effort.},
author = {Nygren, Richard E. and Ulrickson, Michael A. and Martin, Tina T. and Youchison, Dennis L and Lutz, Thomas J. and Bullock, James H. and Hollis, Kendall J.},
abstractNote = {Abstract not provided.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Tue May 01 00:00:00 EDT 2007},
month = {Tue May 01 00:00:00 EDT 2007}
}

Conference:
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  • An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17LI is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Ph-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure.
  • The performance of the ITER first wall and divertor have been analyzed using the Fusion Lifetime Prediction (FLIP) code. The code is a one-dimensional finite difference code which calculates the changes in properties, stress, strain, and temperature overtime for plate structures. The results indicate that the first wall should be able to accommodate up to {approximately}0.6 MW/m{sup 2} heat flux for the reference operating conditions. At much higher levels, fatigue and cracking are predicted to lead to rapid failure. The loss of ductility in irradiated austenitic stainless steel at low temperatures is a concern which may limit operating life. Themore » results of the divertor analysis show that a bare, 2 mm thick plate of Nb-1Zr or TZM can accommodate fluxes of 15--20 MW/m{sup 2} for the ITER conditions. Duplex structures composed of 2 mm of tungsten on 2 mm of Nb-1Zr or TZM are limited to 8--10 MW/m{sup 2}. 3 refs., 8 figs., 3 tabs.« less
  • An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactormore » design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.« less
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  • The recent U.S. effort on the ITER (International Thermonuclear Experimental Reactor) shield has been focused on the limiter module design. This is a multi-disciplinary effort that covers design layout, fabrication, thermal hydraulics, materials evaluation, thermo- mechanical response, and predicted response during off-normal events. The results of design analyses are presented. Conclusions and recommendations are also presented concerning, the capability of the limiter modules to meet performance goals and to be fabricated within design specifications using existing technology.