²⁵²Cf-Source-Driven Noise Analysis Measurements for Characterization of Concrete Highly Enriched Uranium (HEU) Storage Vaults

Conference · Sun Nov 14 04:00:00 EST 1993

OSTI ID:10103743

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

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1. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

The ²⁵²Cf-source-driven noise analysis method has been used in measurements for subcritical configurations of fissile systems for a variety of applications. Measurements of 25 fissile systems have been performed with a wide variety of materials and configurations. This method has been applied to measurements for (1) initial fuel loading of reactors, (2) quality assurance of reactor fuel elements, (3) fuel preparation facilities, (4) fuel processing facilities, (5) fuel storage facilities, (6) zero-power testing of reactors, and (7) verification of calculational methods for assemblies with the neutron k < l. These previous measurements, performed with a wide variety of multiplying systems, demonstrated the usefulness of the method. The high sensitivity of noise-measured parameters to small changes in fissile systems has been observed in several measurements. This high sensitivity has been evaluated using Monte Carlo neutron transport methods for an array of light-water reactor spent fuel and aqueous solutions of fissile material. Monte Carlo calculations of these noise-measured parameters for a highly enriched uranium (HEU) storage vault were performed to evaluate whether a subcriticality measurement with existing ²⁵²Cf source and commercially available detectors was feasible and to determine if the measurement could characterize the ability of the concrete to isolate the fissile material.

• Research Organization: Oak Ridge Associated Universities, Inc., TN (United States); Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: AC05-84OR21400; AC05-76OR00033
• OSTI ID: 10103743
• Report Number(s): CONF-931160--29; ON: DE94001468
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220100
663610
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
CALIFORNIUM 252
ENRICHED URANIUM
FISSILE MATERIALS
Fissile Material
Fissile Systems
MEASURING METHODS
MONTE CARLO METHOD
Monte Carlo
NEUTRON PHYSICS
NEUTRON TRANSPORT
NOISE
Nuclear Criticality Safety Program (NCSP)
SUBCRITICALITY
THEORY AND CALCULATION