Design of a Californium Source-Driven Measurement System for Accountability of Material Recovered from the Molten Salt Reactor Experiment Charcoal Bed

Conference · Sun Jun 07 04:00:00 EDT 1998

OSTI ID:672051

Bentzinger, D. L. ^[1]; Perez, R. B. ^[2]; Mattingly, J. K. ^[2]; Valentine, T. E. ^[2]; Mihalczo, J. T. ^[2]

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1. Univ. of Tennessee, Knoxville, TN (United States)
2. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

The Molten Salt Reactor Experiment Facility (MSRE) operated from 1965 to 1969. The fuel was a molten salt that flowed through the reactor core which consisted of uranium tetrafluoride with molten lithium and beryllium salt used as the coolant. In 1968 the fuel was switched from ²³⁵U to ²³³U. The Molten Salt Reactor Experiment was canceled in 1969 at which time approximately 4800 kg of salt was transferred to the fuel drain tanks. There was about 36.3 kg of uranium, 675 grams of plutonium and various fission products present in the fuel salt. The salt was allowed to solidify in the fuel drain tanks. The salt was heated on a yearly basis to recombine the fluorine gas with the uranium salt mixture. In March 1994, a gas sample was taken from the off gas system that indicated ²³³U had migrated from the fuel drain tank system to the off gas system. It was found that approximately 2.6 kg of uranium had migrated to the Auxiliary Charcoal Bed (ACB). The ACB is located in the concrete-lined charcoal bed cell which is below ground level located outside the MSRE building. Therefore, there was a concern for the potential of a nuclear criticality accident, although water would have to leak into the chamber for a criticality accident to occur. Unstable carbon/fluorine compounds were also formed when the fluorine reacted with the charcoal in the charcoal bed. The purpose of the proposed measurement system was to perform an accountability measurement to determine the fissile mass of ²³³U in the primary vessel. The contents of the primary containment assembly will then be transferred to three smaller containers for long term storage. Calculations were performed using MCNP-DSP to determine the configuration of the measurement system. The information obtained from the time signatures can then be compared to the measurement data to determine the amount of ²³³U present in the primary containment assembly.

• Research Organization: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: AC05-96OR22464
• OSTI ID: 672051
• Report Number(s): ORNL/CP--98159; CONF-980606--; ON: DE98005583
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
22 GENERAL STUDIES OF NUCLEAR REACTORS
ADSORBENTS
Auxiliary Charcoal Bed (ACB)
CALIFORNIUM 252
CONTAINMENT SYSTEMS
DESIGN
HE-3 COUNTERS
MATERIALS RECOVERY
MSRE REACTOR
Molten Salt Reactor Experiment Facility (MSRE)
NEUTRON SOURCES
NUCLEAR MATERIALS MANAGEMENT
Nuclear Criticality Safety Program (NCSP)
OFF-GAS SYSTEMS
REACTOR DECOMMISSIONING
URANIUM 233