A.M. LaFleur, W.S. Charlton, H.O. Menlove, M.T. Swinhoe, S.Y. Lee, S.J. Tobin, “Experimental Benchmark of MCNPX Calculations Against Self-Interrogation Neutron Resonance Densitometry (SINRD) Fresh Fuel Measurements”, 51st Annual Meeting for the Institute of Nuclear Materials Management, Baltimore, Maryland, July 11-15, 2010.
We have investigated the use Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the 235U concentration in a PWR 15×15 fresh LEU fuel assembly in air. Different measurement configurations were simulated in Monte Carlo N-Particle eXtended transport code (MCNPX) and benchmarked against experimental results. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect ofresonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in fission chamber. Due to the low spontaneous fission rate of 238U (i.e. no curium in the fresh fuel), 252Cf sources were used to self-interrogate the fresh fuel pins. The resonance absorption of these neutrons in the fresh fuel pins can be measured using 235U fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the number of unknowns we are trying measure because the neutron source strength and detector-fuel assembly coupling cancel in the ratios. The agreement between MCNPX results and experimental measurements confirms the accuracy of the MCNPX models used. The development of SINRD to measure the fissile content in spent fuel is important to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in LWR spent fuel in water.