WHAT IS TRIMON?
TRIMON is a software package that runs Monte Carlo simulation of neutron random walk movements inside a TRIGA reactor.
TRIMON helps physicists and students to analyze and visualize neutronic behaviour such as the effect of reactor power feedback on the core reactivity
and the effect of nuclear fuel burnup on core reactivity. TRIMON could also be used to analyze the change in core reactivity after modifying the fuel configuration, including the fuel arrangements and the fuel type being used.
TRIGA REACTOR
TRIGA is a commercial research reactor and it has been installed in 24 different countries. The reactor has been used for many diverse applications such as radioisotopes production, non-destructive testing, research on the properties of matter and for education and training. The reactor is a pool-typed water reactor and the reactor core is loaded with hydride fuel-moderator elements, specifically the U-ZrH alloy. The fuel meat is a solid, homogeneous alloy of U-ZrH with the uranium enriched to 20% U-235. Also, the fuel meat is cladded by a 0.051cm thick aluminium or stainless steel (SUS304) can.
GETTING TRIMON
The current standalone version of TRIMON is free to use. It runs on all Windows machines with .NET framework 2.0.
RESEARCH PAPER
Click the link below to read the research paper about TRIMON. In this paper, express details of the theories and laws implemented in TRIMON are presented.
TRIMON MANUAL
TRIMON users can refer to the user manual before using TRIMON. Also, the input format is easy-to-learn and not time consuming.
THEORIES AND LAWS
ALAS, IT IS BEST TO UNDERSTAND THE VARIOUS THEORETICAL CONCEPTS AND PHYSICAL LAWS IMPLEMENTED IN TRIMON BEFORE USING IT. WITH ALL OF THESE THEORETICAL CONCEPTS IN MIND, USERS CAN GAIN MORE KNOWLEDGE IN NUCLEAR MONTE CARLO METHOD AND GET DEEP INSIGHT INTO THE NEUTRONIC BEHAVIOUR IN A TRIGA REACTOR CORE.
Chapter I ― An Overview of Nuclear Monte Carlo Method
In this chapter, a brief introduction to nuclear Monte Carlo method is presented to the reader without any mathematical jargons related to it.
Chapter II ― Neutron Transport Theory
In this chapter, an exact integro-differential equation which describes the neutron transport phenomena will be introduced. Such an equation is recognized as the neutron transport equation and the key objective of this study is to solve the equation. The readers will also be introduced with the basic concepts of the neutron transport theory before jumping into the battle of solving the equation.
Chapter III ― Criticality Calculation Basics
In this chapter, the multigroup method will be introduced to simplify and reduce the general transport equation into the multigroup equations. Then, readers will be presented with the most vital calculation in reactor physics, that is, the criticality calculation. At this point, the criticality calculation of a nuclear system will allow us to evaluate the stability of the fission chain reaction.
Chapter IV ― Theories and Laws Implemented in TRIMON
This chapter outlines the details of the development of TRIMON (TRIGA Monte Carlo Code), a reactor code that integrates homogenized group cross sections into the Monte Carlo method for TRIGA reactors. The details of the theories and laws implemented in TRIMON are extensively discussed in this chapter.
TRIMON SUPPORTS THREE DIFFERENT TYPES OF U-ZRH (URANIUM ZIRCONIUM HYDRIDE ALLOY) FUELS INCLUDING THE 8.5%WT (FE08), 12%WT (FE12) AND THE 20%WT (FE20) WITH EACH HAVING U-235 ENRICHED UP TO 19.9%. OTHER TYPES OF FUELS ARE NOT SUPPORTED.
It is also important to borne in mind that the core height correction needs to be done prior to the use of TRIMON. To understand why, it is customary to define a homogenized region as a simplified sub-region within the reactor core. For instance, the fuel meat, plus the cladding and the surrounding coolant water can be regarded as a sub-region. Such a sub-region can be homogenized and represented using a set of homogenized group cross sections. However, preserving the net neutron leakage across the boundary of the sub-region is often cumbersome. It is also impossible to avoid the fact that neutrons migrate from one sub-region to another which mean the group cross sections of the homogenized sub-region is dependent on its surroundings.
Figure 1: A cross-sectional schematic diagram of a TRIGA UZrH fuel.
The current state-of-the-art technique for generating homogenized neutron cross section imposes a limitation such that the strategies of preserving the leakage at the homogenized cell boundary remain inconsistent. Since the theoretical solution to the current homogenization issue remains unknown or yet to be discovered, it is convenient to introduce a free parameter to suppress the leakage effect on the boundary of the sub-region. Here, the free parameter is a factor which scales the core height (also the buckling-squared) such that the net axial leakage of the reactor core matches with the actual operating reactor core. Therefore, the homogenized neutron cross section data provided by TRIMON need to get calibrated for your TRIGA reactor.
Figure 2: An illustration of homogenization of a sub-region within a reactor core.
THE FOLLOWING STEPS NEED TO BE DONE DURING THE FIRST USE OF TRIMON
Step I ― The actual critical core configuration (fuels arrangement and composition that yields core criticality) is specified to the TRIMON input. By default, the core height is set to 38.1cm which is the actual active height of U-ZrH fuels designed for TRIGA.
Step II ― Run the TRIMON's criticality calculation. Identify the final effective multiplication factor yields after the calculation.
Step III ― Your aim is to find the simulated core height for such an actual critical configuration that yields criticality in TRIMON calculation. To do this, you may adjust the simulated core height in TRIMON input file by modifying the core height input value of #DIMENSIONS. The input manual of TRIMON can be found here.
Step IV ― Steps II-III are repeated until the effective multiplication factor obtained from TRIMON calculation is equal to 1.
AUTHORS
The development of TRIMON is contributed by M. R. Omar (as a PhD student) and his supervisors:
M. R. Omar and Dr. Yoon Tiem Leong
School of Physics,
Universiti Sains Malaysia,
11800, Minden,
Penang, Malaysia.
Dr. Julia Abdul Karim
Reactor Technology Centre,
Technical Support Division,
Malaysian Nuclear Agency,
48000, Bangi,
Selangor, Malaysia.