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Conference Article: Data For Computing Radiation Induced Damage In HTR Fuel Matrices

A.M. Ougouag , C.A. Wemple, Idaho National Engineering and Environmental Laboratory, Idaho, USA

Abstract

Modifications to the displacement kerma cross section methods employed in the NJOY99 nuclear data processing code are described. Calculations were performed with the modified code for HTR fuel materials using ENDF-6 neutron interaction data. Graphical displays of displacement kerma and gas production cross sections were generated for representative HTR fuel compositions.

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key words: Gas Cooled Reactor, Nuclear Technology
Reference:
Proceedings of the Conference on High Temperature Reactors, Beijing, China, September, 22-24, 2004
International Atomic Energy Agency, Vienna (Austria)
HTR-2004, pp:1-8