Bias in calculated k{sub eff} from subcritical measurements by the {sup 252}Cf-source-driven noise analysis method

Technical Report · Sat Jul 01 00:00:00 EDT 1995

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

Mihalczo, J T; Valentine, T E

The development of MCNP-DSP, which allows direct calculation of the measured time and frequency analysis parameters from subcritical measurements using the {sup 252}Cf-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 {minus}0.0061 {+-} 0.0003. For a 10.16-cm-high cylinder (k {approx} 0.93), it was 0.0060 {+-} 0.0016, and for a subcritical cylinder (h = 8.13 cm, k {approx} 0.85), the bias was {minus}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 {sup 252}Cf-source-driven noise analysis method.

• Research Organization: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 95236
• Report Number(s): CONF-9509100-6; ON: DE95014587; TRN: 95:018503
• Resource Relation: Conference: 5. international conference on nuclear criticality safety, Albuquerque, NM (United States), 17-22 Sep 1995; Other Information: PBD: [1995]
• Country of Publication: United States
• Language: English

Related Subjects

42 ENGINEERING NOT INCLUDED IN OTHER CATEGORIES
05 NUCLEAR FUELS
22 NUCLEAR REACTOR TECHNOLOGY
CRITICALITY
CALCULATION METHODS
HIGHLY ENRICHED URANIUM
FISSILE MATERIALS
MULTIPLICATION FACTORS
M CODES
CALIFORNIUM 252
SAFETY ANALYSIS
NUCLEAR DATA COLLECTIONS
REACTOR FUELING
FUEL FABRICATION PLANTS
FUEL REPROCESSING PLANTS
STORAGE FACILITIES
SAFETY ENGINEERING
USES
EXPERIMENTAL DATA
THEORETICAL DATA
OR-CEF REACTOR
FUEL ELEMENTS
REACTOR KINETICS