Methodology and Determination of Field of View of Neutron and Gamma Detectors in the Atucha Spent Fuel Storage Pool
- Lawrence Livermore National Laboratory (LLNL), Livermore, CA (United States)
In this paper we seek to create a model by determining the field of view (FOV) of a detector (i.e. which assemblies contribute to the detector response) in the Atucha-I spent fuel pool. The FOV is determined by solving the adjoint transport equation using the 3-D, parallel PENTRAN (Parallel Environment Neutral-particle TRANsport) Sn code, with the detector cross section as the adjoint source. If this adjoint function is coupled with the source spectrum, then the contribution to the detector from each assembly can be determined. First, the reactor criticality was modeled using the MCNP5 (Monte Carlo N-Particle) Monte Carlo code in order to determine the power distribution in each assembly. Using the power distribution data, the assemblies were divided and homogenized into 8 axial and 3 radial zones for burnup analysis. Depletion calculations were performed for each zone using the ORIGEN-ARP (Automatic Rapid Processing) utility from the SCALE 5.1 (Standardized Computer Analyses for Licensing Evaluation) code package. Spent fuel pool and detector were modeled in 2-D in PENTRAN as the detector plus 3 fuel assemblies along both x and y axes. Using the resulting adjoint function combined with the source spectrum, they have determined the FOVs of the fission chamber neutron detector that was used at Atucha, and concluded that 2 assemblies along x and y axes are needed for both cases (i.e. the 4 adjacent assemblies plus the next surrounding 12). For the neutron detector, 88% of the response comes from the nearest 4 assemblies, with 99% from the nearest 16. Results for a uniformly sensitive gamma detector indicate that 2 assemblies in both directions are also needed, with 89% of the response coming from the adjacent assemblies. A Monte Carlo calculation using MCNP was performed to benchmark the neutron result, giving a similar result (87% MCNP vs. 88% PENTRAN). Based on these studies, we have developed a database of FOVs as a function of burnup and decay conditions for different detector types, and a methodology/algorithm which uses this database to analyze the response of a detector placed in a spent fuel pool with the aim of detecting gross defects.
- Research Organization:
- Lawrence Livermore National Laboratory (LLNL), Livermore, CA (United States)
- Sponsoring Organization:
- USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
- DOE Contract Number:
- AC52-07NA27344
- OSTI ID:
- 964510
- Report Number(s):
- LLNL-CONF-414296; TRN: US0903862
- Resource Relation:
- Conference: Methodology and Determination of Field of View of Neutron and Gamma Detectors in the Atucha Spent Fuel Storage Pool, Tucson, AZ (United States), 12-16 Jul 2009
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
SAFEGUARDS AND PHYSICAL PROTECTION
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY
BENCHMARKS
BURNUP
COMPUTERS
CRITICALITY
CROSS SECTIONS
DECAY
DEFECTS
FISSION CHAMBERS
FUEL ASSEMBLIES
LICENSING
NEUTRAL-PARTICLE TRANSPORT
NEUTRON DETECTORS
NEUTRONS
POWER DISTRIBUTION
SPENT FUEL STORAGE
SPENT FUELS
Nuclear Criticality Safety Program (NCSP)