Dry Spent Fuel Cask Monitoring by ²⁵²Cf-Source-Driven Frequency Analysis Measurements

Conference · Sun Jul 28 00:00:00 EDT 1996

OSTI ID:395574

Valentine, T. E. ^[1]; Mattingly, J. K. ^[2]; Mihalczo, J. T. ^[1]

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1. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States). Instrumentation and Controls Division, Measurements Science Section
2. Oak Ridge Inst. for Science and Education (ORISE), Oak Ridge, TN (United States)

If developed, a nondestructive method would be useful for verifying canister contents without requiring the canister to be opened. This paper addresses the application of the ²⁵²Cf-source-driven frequency analysis measurements for verification of the fissile material content of sealed spent fuel canisters. The cross-power spectral density (CPSD) between the ²⁵²Cf source in an ionization chamber and external neutron detectors depends only on the induced fission rate in the fissile system and is independent of inherent sources. Thus, the source-to-detector CPSD is ideal for determination of fissile material content of the spent fuel. This paper evaluates the application of this method to a 125-ton spent fuel canister that contained 21 pressurized-water reactor fuel elements. The results demonstrate that the fissile materials content of a sealed spent fuel canister could be obtained using the ²⁵²Cf frequency analysis method if calibration standards were available. The results also indicate that a measurement could be performed in less than a day for burnups up to 36 GWd/MTU and in less time for lower burnups.

• Research Organization: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States). Instrumentation and Controls Division, Measurements Science Section; Oak Ridge Inst. for Science and Education (ORISE), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP); USDOE Laboratory Cooperative Postgraduate Research Training Program
• DOE Contract Number: AC05-96OR22464
• OSTI ID: 395574
• Report Number(s): CONF-960767-69; ON: DE96015229; TRN: 97:000017
• Resource Relation: Conference: 37. Annual Meeting of the Institute of Nuclear Materials Management, Naples, FL (United States), 28 Jul - 1 Aug 1996; Other Information: PBD: [1996]
• Country of Publication: United States
• Language: English

Related Subjects

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
SPENT FUEL ELEMENTS
NONDESTRUCTIVE ANALYSIS
ARMS CONTROL
VERIFICATION
CALIFORNIUM 252
INSPECTION
PWR TYPE REACTORS
FUEL SCANNING
BURNUP
NEUTRON TRANSPORT
MONTE CARLO METHOD
FISSILE MATERIALS
SAFEGUARDS
Nuclear Criticality Safety Program (NCSP)
Pressurized Water Reactor (PWR)
Reactor Fuel Elements
Spent Fuel Canister
Fissile Material
Cross-Power Spectral Density (CPSD)
Induced Fission Rate
Monte Carlo Code