Bias in Calculated k_eff from Subcritical Measurements by the ²⁵²Cf-source-Driven Noise Analysis Method

Technical Report · Sun Jan 01 04:00:00 EST 1995

DOI:https://doi.org/10.2172/95236· OSTI ID:95236

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

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

The development of MCNP-DSP, which allows direct calculation of the measured time and frequency analysis parameters from subcritical measurements using the ²⁵²Cf-source-driven noise analysis method, permits the validation of calculational methods for criticality safety with in-plant subcritical measurements. In addition, a method of obtaining the bias in the calculations, which is essential to the criticality safety specialist, is illustrated using the results of measurements with 17.771-cm-diam, enriched (93.15), unreflected, and unmoderated uranium metal cylinders. For these uranium metal cylinders the bias obtained using MCNP-DSP and ENDF/B-V cross-section data increased with subcriticality. For a critical experiment [height (h) = 12.629 cm], it was ⁻0.0061 {+-} 0.0003. For a 10.16-cm-high cylinder (k ~ 0.93), it was 0.0060 {+-} 0.0016, and for a subcritical cylinder (h = 8.13 cm, k ~ 0.85), the bias was ⁻0.0137 {+-} 0.0037, more than a factor of 2 larger in magnitude. This method allows the nuclear criticality safety specialist to establish the bias in calculational methods for criticality safety from in-plant subcritical measurements by the ²⁵²Cf-source-driven noise analysis method.

• Research Organization: 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
• OSTI ID: 95236
• Report Number(s): CONF-9509100--6; ON: DE95014587
• Country of Publication: United States
• Language: English

Related Subjects

05 NUCLEAR FUELS
22 GENERAL STUDIES OF NUCLEAR REACTORS
42 ENGINEERING
CALCULATION METHODS
CALIFORNIUM 252
CRITICALITY
EXPERIMENTAL DATA
FISSILE MATERIALS
FUEL ELEMENTS
FUEL FABRICATION PLANTS
FUEL REPROCESSING PLANTS
HIGHLY ENRICHED URANIUM
M CODES
MULTIPLICATION FACTORS
NUCLEAR DATA COLLECTIONS
Nuclear Criticality Safety Program (NCSP)
OR-CEF REACTOR
REACTOR FUELING
REACTOR KINETICS
SAFETY ANALYSIS
SAFETY ENGINEERING
STORAGE FACILITIES
THEORETICAL DATA
USES