A high fidelity general purpose 3-D Monte Carlo particle transport program JMCT3.0

Abstract

JMCT is a large-scale, high-fidelity, three-dimensional general neutron–photon–electron–proton transport Monte Carlo software system. It was developed based on the combinatorial geometry parallel infrastructure JCOGIN and the adaptive structured mesh infrastructure JASMIN. JMCT is equipped with CAD modeling and visualizes the image output. It supports the geometry of the body and the structured/unstructured mesh. JMCT has most functions, variance reduction techniques, and tallies of the traditional Monte Carlo particle transport codes. Two energy models, multi-group and continuous, are provided. In recent years, some new functions and algorithms have been developed, such as Doppler broadening on-the-fly (OTF), uniform tally density (UTD), consistent adjoint driven importance sampling (CADIS), fast criticality search of boron concentration (FCSBC) domain decomposition (DD), adaptive control rod moving (ACRM), and random geometry (RG) etc. The JMCT is also coupled with the discrete ordinate SN code JSNT to generate source-biasing factors and weight-window parameters. At present, the number of geometric bodies, materials, tallies, depletion zones, and parallel processors are sufficiently large to simulate extremely complicated device problems. JMCT can be used to simulate reactor physics, criticality safety analysis, radiation shielding, detector response, nuclear well logging, and dosimetry calculations etc. In particular, JMCT can be coupled with depletion and thermal-hydraulics for the simulation of reactor nuclear-hot feedback effects. This paper describes the progress in advanced modeling, high-performance numerical simulation of particle transport, multiphysics coupled calculations, and large-scale parallel computing.

Convergence characteristics and acceleration of the transient fixed source equation solved by Monte Carlo method

“…Among many high-fidelity simulation methods, the Monte Carlo (MC) method is widely regarded as the most accurate method and is usually used for validations of other methods. The MC method has two major advantages: one is the usage of accurate continuous-energy crosssections, and another is the precise 3-dimensional (3D) geometry modeling based on constructive solid geometry (CSG) or computer-aided design (CAD) geometry (Wilson et al, 2010;Davis et al, 2020;Deng et al, 2022). MC codes such as MCNP 10.3389/fenrg.2022.1010482 (Goorley et al, 2012), Serpent (Leppänen et al, 2015), OpenMC (Romano et al, 2015), and RMC (Wang et al, 2015) are widely used in new reactor designs, which have disparate features from normal nuclear reactors in geometry, material, and spectra.…”

High-Fidelity Monte Carlo Analysis of Highly-Localized Fission Power Peaking near a Flux Trap in an MTR Core

“…The "Best Estimate" Approach. In recent years, there has been a growing interest in a promising alternative, the socalled "best estimate" methodology, based on high-fidelity Monte Carlo codes, e.g., Serpent [13], MCNP [14], and OpenMC [15], for detailed neutronic analyses characterized by high accuracy and flexible implementation [16][17][18][19][20]. This class of models, characterized by very fine spatial resolution, continuous energy spectrum, and continuous angular distribution, are referred to as high-fidelity models and allow detailed investigations of localized phenomena.…”