MCNP-DSP calculations of measurements with uranyl nitrate solution system

Journal Article · Tue Sep 01 04:00:00 EDT 1998 · Transactions of the American Nuclear Society

OSTI ID:644269

Valentine, T E ^[1]

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1. Oak Ridge National Lab., TN (United States)

The {sup 252}Cf-source-driven noise analysis method has been used to determine the subcriticality of various configurations of fissile materials. In the past, the application of this method was limited because point-kinetics models had to be used to interpret the data; however, with the development of the Monte Carlo code MCNP-DSP, the measurements can be analyzed using the more general Monte Carlo models. The results of the Monte carlo calculations will be dependent on the ability to model the experiment accurately and on the nuclear data used to perform the calculations. This paper presents a comparison of the measured and calculated ratio of spectral densities for a subset of measurements performed with a uranyl nitrate solution tank filled to various heights. The results presented are for calculations that were performed with both ENDF/B-IV and ENDF/B-V cross-section data sets.

• OSTI ID: 644269
• Report Number(s): CONF-980606--
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 78; ISSN 0003-018X; ISSN TANSAO
• Country of Publication: United States
• Language: English

Related Subjects

42 ENGINEERING
99 GENERAL AND MISCELLANEOUS
CALIFORNIUM 252
COMPUTER CALCULATIONS
CRITICALITY
EVALUATION
M CODES
MONTE CARLO METHOD
NUCLEAR DATA COLLECTIONS
SPECTRAL DENSITY
URANYL NITRATES