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,"
Proceedings of the 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.
Associated Project(s):Development of Self-Interrogation Neutron Resonance Densitometry (SINRD) to Measure the Fissile Content in Nuclear Fuel