Experimental benchmark of MCNPX calculations against self-interrogation neutron resonance densitometry (SINRD) fresh fuel measurements

Conference · Fri Jan 01 00:00:00 EST 2010

OSTI ID:1020961

Menlove, Howard O ^[1]; Swinhoe, Martyn T ^[1]; La Fleur, Adrienne M ^[1]; Charlton, William S ^[1]; Lee, S Y ^[1]; Tobin, S J ^[1]

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1. Los Alamos National Laboratory

We have investigated the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the {sup 235}U concentration in a PWR 15 x 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 of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,j) reaction peaks in fission chamber. Due to the low spontaneous fission rate of {sup 238}U (i.e. no curium in the fresh fuel), {sup 252}Cf 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 {sup 235}U 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.

• Research Organization: Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)
• Sponsoring Organization: USDOE
• DOE Contract Number: AC52-06NA25396
• OSTI ID: 1020961
• Report Number(s): LA-UR-10-04144; LA-UR-10-4144; TRN: US201116%%639
• Resource Relation: Conference: INMM 51st Annual Meeting ; July 11, 2010 ; Baltimore, MD
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY
97 MATHEMATICS AND COMPUTING
98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL PROTECTION
ACCURACY
AIR
BENCHMARKS
CALCULATION METHODS
CALIFORNIUM 252
COMPUTERIZED SIMULATION
COMPUTER CODES
COUPLING
CURIUM
FISSILE MATERIALS
FISSION CHAMBERS
FUEL ASSEMBLIES
FUEL PINS
MEASURING METHODS
MEETINGS
MONTE CARLO METHOD
NONDESTRUCTIVE ANALYSIS
RESONANCE
RESONANCE ABSORPTION
SAFEGUARDS
SENSITIVITY
SPENT FUELS
SPONTANEOUS FISSION
TRANSPORT
URANIUM 235