On a Continuous Energy Monte Carlo Simulator for Neutron Transport: Optimisation with Fission, Intermediate and Thermal Distributions

DOI10.1007/978-3-319-59387-6_1zbMath1405.82038OpenAlexW2754662539MaRDI QIDQ4562611

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Publication date: 17 December 2018

Published in: Integral Methods in Science and Engineering, Volume 2 (Search for Journal in Brave)

Full work available at URL: https://doi.org/10.1007/978-3-319-59387-6_1

zbMATH Keywords

neuronMonte Carlo simulator

Mathematics Subject Classification ID

Nuclear reactor theory; neutron transport (82D75) Boltzmann equations (35Q20)

Related Items (1)

On a Parametric Representation of the Angular Neutron Flux in the Energy Range from 1 eV to 10 MeV

Uses Software

• MCNP
• OpenMC
• Unnamed Item

Cites Work

• A Novel Method for Simulating Spectral Nuclear Reactor Criticality by a Spatially Dependent Volume Size Control

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