A single-step framework to generate spatially self-shielded multi-group cross sections from Monte Carlo transport simulations-Reference-Cited by-同舟云学术

A single-step framework to generate spatially self-shielded multi-group cross sections from Monte Carlo transport simulations

Published:2019-03 Issue: Volume:125 Page:261-271
ISSN:0306-4549
Container-title:Annals of Nuclear Energy
language:en
Short-container-title:Annals of Nuclear Energy

Author:

Boyd William,Forget Benoit^ORCID,Smith Kord

Funder

National Science Foundation Graduate Research

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering

Reference23 articles.

1. OpenCG: a combinatorial geometry modeling tool for data processing and code verification;Boyd,2015

2. An analysis of condensation errors in multi-group cross-section generation for fine-mesh neutron transport calculations;Boyd;Ann. Nucl. Eng.,2018

3. Multi-group cross section generation with the OpenMC Monte Carlo particle transport code;Boyd;Nucl. Technol.,2018

4. The OpenMOC method of characteristics neutral particle transport code;Boyd;Ann. Nucl. Energy,2014

5. Condensation and Homogenization of Cross Sections for the Deterministic Transport Codes with Monte Carlo Method: Application to the GEN IV Fast Neutron Reactors;Cai,2014

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1. A Moose-Based Neutron Diffusion Code with Application to a LMFR Benchmark;Springer Proceedings in Physics;2023

2. Generation method and verification of pebble type VHTR multigroup cross sections based on OpenMC;Annals of Nuclear Energy;2021-09

3. Research on the fabrication method of macroscopic cross section base on the Monte Carlo simulation;Annals of Nuclear Energy;2021-04

4. Generalized Equivalence Theory Used with Spatially Linear Sources in the Method of Characteristics for Neutron Transport;Nuclear Science and Engineering;2020-05-28