Benchmarking Calculation of a Soluble-Boron-Free SMR Lattice Model using Deterministic, Hybrid Monte Carlo and Monte Carlo Codes


Numerical benchmark calculations for a soluble-boron-free (SBF) 2D SMR assembly have been performed using the deterministic transport code WIMS, Monte Carlo (MC) code Serpent and hybrid MC code MONK. Each code has the capability to perform neutronic transport calculations using different methods to determine important neutronic parameters. Two sets of fuels: 18% U-235 enriched micro-heterogeneous ThO2-UO2 duplex fuel and 15% U-235 enriched homogeneously mixed all-UO2 fuel have been examined in this study. A comparison between the three codes has been carried out with a 2D fuel assembly model. The eigenvalue (k) and 2D assembly pin power distribution at different burnup states in the assembly depletion are compared using several nuclear data files, such as ENDF/B-VII, JEF2.2 and JEF3.1. A good agreement in k values was observed among the codes for both the candidate fuels. The differences in k∞ between the codes are 200 pcm when cross-sections based on the same nuclear data file are used. A higher difference (up to 450 pcm) in the k∞ values is observed among the codes using cross-sections based on different data files. Finally, it can be concluded from this study that the good agreement in the results between the codes found provides enhanced confidence that the deterministic reactor physics code WIMS can be reliably used in modeling SMR propulsion core systems, offering the advantage of less expensive computation compared to the MC code Serpent and hybrid MC code MONK.

Meeting Name

2018 Pacific Basin Nuclear Conference, PBNC 2018 (2018: Sep. 30-Oct. 5, San Francisco, CA)


Nuclear Engineering and Radiation Science

Keywords and Phrases

Boron; Eigenvalues and eigenfunctions; Fuels; Monte Carlo methods; Small nuclear reactors; Thoria; Uranium dioxide; Deterministic transport; Hybrid Monte Carlo; Lattice modeling; Micro-heterogeneous; Monte Carlo codes; Numerical benchmark; Pin power distribution; Transport calculation; Codes (symbols)

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Article - Conference proceedings

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© 2018 PBNC 2018 - Pacific Basin Nuclear Conference, All rights reserved.

Publication Date

01 Sep 2018