In order to generate cross sections for fast reactor calculation, a code named TXMAT based on object–oriented programming and allocate memory technology has been developed. It has the capability to generate macroscopic cross sections for transport or diffusion calculation and microscopic cross sections for burnup calculation, and can also deal with the ultrafine group cross sections (more than 2000 groups and P5 Legendre order) which TRANSX 2.15 can’t do. It works together with a generalized cross section data library called MATXS to give the transport code users easier access to appropriate nuclear data and capabilities which are difficult or impossible to get with any other systems. The TXMAT can handle the shielding effects of many isotopes through background cross section iteration. Several critical benchmarks are calculated. It is shown that the total cross section, absorption cross section, fission neutron spectrum and zero Legendre scattering matrix have been verified using TRANSX 2.15, and the maximum relative difference for the main groups is less than 0.2%. After the critical benchmark calculation, the RBEC-M benchmark is used for the whole core calculation. It is shown that the effective multiplication factor of the calculation is consistent with that of other codes, and the power distribution is also in good agreement with that of other codes except for the blankets. The maximum relative difference of the power distribution among core-1, core-2 and core-3 regions is less than 2.3%. But in the blankets the relative error is about 33%, which may be caused by the difference of the weight function between IWT = 8 and real model. Further analysis will be performed in the future.
- Nuclear Engineering Division
Development and Verification of Multi-Group Cross Section Process Code TXMAT for Fast Reactor RBEC-M Analysis
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Qiu, R, Ma, X, Xu, Q, Liu, J, & Chen, Y. "Development and Verification of Multi-Group Cross Section Process Code TXMAT for Fast Reactor RBEC-M Analysis." Proceedings of the 2017 25th International Conference on Nuclear Engineering. Volume 3: Nuclear Fuel and Material, Reactor Physics and Transport Theory; Innovative Nuclear Power Plant Design and New Technology Application. Shanghai, China. July 2–6, 2017. V003T02A015. ASME. https://doi.org/10.1115/ICONE25-66550
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