Capture cross section measurement analysis in the Californium-252 spectrum with the Monte Carlo method. (July 2015)

• Record Type: Journal Article
• Authors: Manojlovič, Stanko; Trkov, Andrej; Žerovnik, Gašper; Snoj, Luka
• Is Part Of: Applied radiation and isotopes. Volume 101(2015:Jul.)
• Page Start: 101
• Page End: 106
• Publication Date: 2015-07
• Subjects: Cf-252 spontaneous fission spectrum -- Average cross section -- Reaction rate measurements -- Monte Carlo method; Radiology -- Periodicals; Radiation -- Industrial applications -- Periodicals; Nuclear chemistry -- Periodicals; Internet resource; Periodical; 660.298
• DOI: 10.1016/j.apradiso.2015.03.023
• Languages: English
• ISSNs: 0969-8043

Abstract

Abstract: Absolute average capture cross sections of gold, thorium, tantalum, molybdenum, copper and strontium in 252 Cf spontaneous fission neutron spectrum were simulated for two types of experiment setups preformed by Z. Dezso and J. Csikai and by L. Green. The experiments were simulated with MCNP5 using cross section data from the ENDF/B-VII.0 library. The determination of neutron backscattering was calculated with the use of neutron flagging. Correction factors to experimentally measured values were determined to obtain average cross sections in a pure 252 Cf spontaneous fission spectrum. Influence of concrete wall thickness, air moisture and room size on the average cross section was analyzed. Correction factors amounted to about 30%. Corrected values corresponding to average cross sections in a pure 252 Cf spectrum were calculated for 197 Au, 232 Th, 181 Ta, 98 Mo, 65 Cu and 84 Sr. Average cross sections were also calculated with the RR_UNC software using IRDFF-v.1.05 and ENDF/B-VII.0 libraries. The revised average radiative capture cross sections are 75.5±0.1 mb for 197 Au, 87.0±1.6 mb for 232 Th, 98.0±4.5 mb for 181 Ta, 21.2±0.5 mb for 98 Mo, 10.3±0.3 mb for 63 Cu, and 34.9±6.5 mb for 84 Sr. Abstract : Highlights: Average capture cross sections in 252 Cf spontaneous fission spectrum were simulated. Calculations were done using MCNP5 code and ENDF/B-VII.0 library. Correction factors for self-shielding and room return effects were taken into account. The revised average radiative capture cross sections for different materials are published.