A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

T. Makmal, O. Aviv, E. Gilad

Research output: Contribution to journalArticlepeer-review

8 Scopus citations

Abstract

A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections.

Keywords

  • Depletion calculations
  • Fission products
  • Fuel burnup
  • Gamma spectrometry
  • MTR reactor

ASJC Scopus subject areas

  • Nuclear and High Energy Physics
  • Instrumentation

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