Analysis of Fission Matrix Databases with Nonuniform Fuel Temperature Profiles for Static and Transient Calculations

IF 1.2 3区工程技术 Q3 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Science and Engineering Pub Date : 2024-04-22 DOI:10.1080/00295639.2024.2328944

Maximiliano Dalinger, William Walters

引用次数: 0

Abstract

Monte Carlo codes are the most accurate way to solve the neutronics in a reactor core but can be computationally expensive, especially for when feedback effects are considered or for transient calc...

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用于静态和瞬态计算的非均匀燃料温度曲线裂变矩阵数据库分析

Monte Carlo 代码是求解反应堆堆芯中子学问题的最精确方法，但计算成本很高，特别是在考虑反馈效应或进行瞬态计算时。

本文章由计算机程序翻译，如有差异，请以英文原文为准。

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来源期刊

Nuclear Science and Engineering 工程技术-核科学技术

CiteScore

2.10

自引率

25.00%

发文量

审稿时长

3 months

期刊介绍： Nuclear Science and Engineering, the research journal of the American Nuclear Society, publishes articles on research and development related to peaceful utilization of nuclear energy, radiation, and alternative energy sources.