Monte Carlo analysis of neutron slowing-down-time spectrometer for fast reactor spent fuel assay

Conference · Sun Jul 01 04:00:00 EDT 2007

OSTI ID:20979541

Chen, Jianwei; Lineberry, Michael ^[1]

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1. Institute of Nuclear Science and Engineering, Idaho State University, P.O.Box 8150, Idaho Falls, ID 83402 (United States)

Using the neutron slowing-down-time method as a nondestructive assay tool to improve input material accountancy for fast reactor spent fuel reprocessing is under investigation at Idaho State University. Monte Carlo analyses were performed to simulate the neutron slowing down process in different slowing down spectrometers, namely, lead and graphite, and determine their main parameters. {sup 238}U threshold fission chamber response was simulated in the Monte Carlo model to represent the spent fuel assay signals, the signature (fission/time) signals of {sup 235}U, {sup 239}Pu, and {sup 241}Pu were simulated as a convolution of fission cross sections and neutron flux inside the spent fuel. {sup 238}U detector signals were analyzed using linear regression model based on the signatures of fissile materials in the spent fuel to determine weight fractions of fissile materials in the Advanced Burner Test Reactor spent fuel. The preliminary results show even though lead spectrometer showed a better assay performance than graphite, graphite spectrometer could accurately determine weight fractions of {sup 239}Pu and {sup 241}Pu given proper assay energy range were chosen. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 20979541
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ACCOUNTING
ATR REACTOR
FAST REACTORS
FISSILE MATERIALS
GRAPHITE
INVENTORIES
MONTE CARLO METHOD
NEUTRONS
PLUTONIUM 239
PLUTONIUM 241
REPROCESSING
SLOWING-DOWN
SPECTROMETERS
SPENT FUELS
URANIUM 235
URANIUM 238