OpenMOC simulation for core-133 and core-134 of the TRIGA Mark II Research Reactor

OpenMOC is a relatively new reactor physics code that can solve the neutron transport equation using the method of characteristics (MOC), based on the multi-group and a steady-state form of the neutron transport equation. In this study, calculations and simulations were carried out for a TRIGA Mark...

Full description

Saved in:
Bibliographic Details
Main Authors: Kuswanto, Casimiro Rino, Aprianti, Nur Asiah, Irwanto, Dwi
Format: Conference Proceeding
Language:English
Subjects:
Mathematical analysis
Method of characteristics
Multiplication
Nuclear reactors
Reactor design
Reactor physics
Simulation
Transport equations
Online Access:Get full text
Tags: Add Tag
No Tags, Be the first to tag this record!
Description
Summary:OpenMOC is a relatively new reactor physics code that can solve the neutron transport equation using the method of characteristics (MOC), based on the multi-group and a steady-state form of the neutron transport equation. In this study, calculations and simulations were carried out for a TRIGA Mark II Research Reactor using ENDF/B-VII.1 as the nuclear data library to study and test the accuracy of this code and its method. Calculation of the important neutronic parameters in a nuclear reactor design, namely the effective multiplication factor, was carried out for the configuration of core-133 and core-134 in the TRIGA Mark II Research Reactor. The simulation results were compared with experimental results. The results obtained in this study showed a good agreement between the simulation results, using the method of characteristics through the OpenMOC code, and the experimental results with a difference of 0.07% for core-133 and 0.18% for core-134.
ISSN:0094-243X
1551-7616
DOI:10.1063/5.0127161