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DOE OSTI · 1852413

A nodal transport method for coupled fast-thermal reactor analysis

Deng, Puran·Yang, Won Sik
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Abstract

We report that for efficient core analyses of coupled fast-thermal reactors, a new deterministic method has been developed based on the variational nodal transport method of VARIANT. A new multigroup cross section generation procedure was devised by combining Monte Carlo lattice calculations for thermal assemblies and the two-step procedure of the MC 2 -3 code for fast assemblies. To reduce assembly homogenization errors in nodal transport calculations, a new nodal equivalence method was developed based on the partial current discontinuity factor (PCDF) and incorporated in VARIANT. A practical procedure to calculate PCDFs with fixed source supercell calculations was also devised. The performance of the proposed method was investigated using a test problem derived from the versatile coupled test reactor (VCTR) design. The new procedure produced multigroup cross sections accurately for all assemblies. VARIANT transport calculations with PCDFs produced accurate multiplication factor and power distribution compared to reference Serpent-2 Monte Carlo solutions.

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73 NUCLEAR PHYSICS AND RADIATION PHYSICSMonte Carlo lattice calculationassembly homogenizationcoupled fast-thermal reactorpartial current discontinuity factorvariational nodal transport

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Deng, Puran, Yang, Won Sik. 2021-03-16. A nodal transport method for coupled fast-thermal reactor analysis. https://doi.org/10.1016/j.anucene.2021.108229

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