Determination of IRT-2M fuel burnup by gamma spectrometry

• Measurement of selected fission products in the fuel of a research reactor.
• Evaluation of spent fuel with different burnup history and cooling times.
• Burnup calculations using the MCNPX Monte Carlo code.
• Evaluation of burnup calculations versus experiment data.

A spectrometric system was developed for evaluating spent fuel in the LVR-15 research reactor, which employs highly enriched (36%) IRT-2M-type fuel. Such system allows the measurement of detailed fission product profiles. Within these measurements, nuclides such as 137Cs, 134Cs, 144Ce, 106Ru and 154Eu may be detected in fuel assemblies with different cooling times varying between 1.67 and 7.53 years. Burnup calculations using the MCNPX Monte Carlo code data showed good agreement with measurements, though some discrepancies were observed in certain regions. These discrepancies are attributed to the evaluation of irradiation history, reactor regulation pattern and buildup schemes.