This paper introduces a quick method for improving the accuracy of Monte Carlo simulations by generating one-and twodimensional cross sections at a user-defined temperature before performing transport calculations. A finite difference method is used to Doppler-broaden cross sections to the desired temperature, and unit-base interpolation is done to generate the probability distributions for double differential two-dimensional thermal moderator cross sections at any arbitrarily user-defined temperature. The accuracy of these methods is tested using a variety of contrived problems. In addition, various benchmarks at elevated temperatures are modeled, and results are compared with benchmark results. The problem-dependent cross sections are observed to produce eigenvalue estimates that are closer to the benchmark results than those without the problem-dependent cross sections.
No abstract
Advanced reactor concepts being developed throughout the industry are significantly different from light-water reactor (LWR) designs with respect to geometry, materials, and operating conditions, and consequently, with respect to their reactor physics behavior. Given the limited operating experience with non-LWRs, the accurate simulation of reactor physics and the quantification of associated uncertainties are important for ensuring that the nuclear design for advanced reactor concepts include appropriate margins. Nuclear data are a major source of input uncertainties in reactor physics analysis. As part of a project sponsored by the US Nuclear Regulatory Commission at Oak Ridge National Laboratory (ORNL), key nuclear data relevant to reactor safety analysis in selected advanced reactor technologies 1 were identified, and their impacts on important key figures of merit were assessed based on (1) a review of available advanced reactor specifications, (2) analysis of previous studies performed at ORNL and other research institutions, and (3) sensitivity and uncertainty analyses performed for six selected benchmarks-three experimental and three computational-to quantify the impacts of the identified key nuclear data on several key metrics. This report summarizes the key nuclear data-nominal data and nuclear data uncertaintiesconsidering the most important nuclear reactions in the fuel and in various materials for the moderator, coolant, and structure of the considered advanced reactors.The major nominal missing data that were identified consist of thermal scattering data and 135m Xe cross section data for molten salt reactor (MSR) analysis. The identified major gaps with respect to nuclear data uncertainties are (1) the missing uncertainties in the thermal scattering data for high-temperature gas-cooled reactors and moderated MSR systems, and (2) the incomplete uncertainties on angular distributions, particularly for fast spectrum systems such as sodium-cooled fast reactors, fast molten salt reactors, and heat pipe reactors.Large uncertainties of reactions that are not commonly considered to be relevant in LWR studies were found to be significant for several advanced reactor systems. The large uncertainty of 238 U inelastic scattering in the fast energy range contributes significantly to large output uncertainties in all fast spectrum systems. The large uncertainty of 235 U (n,) in the fast energy range causes significant reactivity uncertainties in fast neutron spectrum systems that use 235 Uenriched fuel. A large uncertainty of 7 Li (n,) causes a large fraction of uncertainty in the output quantities investigated for MSR systems in which lithium is part of the salt. Special attention should be paid to differences in cross section and uncertainties of different evaluated nuclear data library releases. Significant differences were found in nuclear data that can lead to major differences in reactivity calculations, even for well-known nuclides. In particular, differences in 235 U, 238 U, and 239 Pu nominal and unce...
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