Development and coupling of the 1-D neutronics code PoKiMON with ANSYS CFX for reactor shutdown simulations
Introduction
A nuclear system operates at constant power, as long as the neutrons produced via fission reactions balance the neutrons lost due to absorption or leakage.
Often, the reactor is set off balance by changes made by the reactor operators to adjust the power level. Other times, unforeseen events like a Loss of Cooling Accident (LOCA), Loss of Force Cooling (LOFC), a pump malfunction, or main steam line break accident can also alter the balance of the system.
As a matter of fact, it is of the uttermost importance to be able to predict the dynamic variations of the neutron population in order to assess and mitigate the possible consequences deriving from an accident scenario.
At the moment, nuclear reactor safety simulations of real nuclear plants are performed using approximated system codes which rely on one-dimensional lumped-parameters or subchannel methods (Moorthi et al., 2018).
Of course, one-dimensional codes are not able to fully simulate intrinsic multi-dimensional effects that occur within a nuclear reactor (Smith, 2010, Chen et al., 2017, Fanning et al., 2009).
On the other hand, CFD codes, thanks to the recent advances in computer technology, are now capable of tackling complex multi-dimensional fluid-dynamics problems, for a large variety of nuclear and non-nuclear applications (D’Auria, 2017).
Therefore, in order to correctly simulate the overall behaviour of a nuclear system, it is important to capture all the neutronics / fluid dynamics interactions that occur in such systems. This is done by coupling the CFD code to a neutronics one.
In recent years, other researchers have coupled kinetics codes to CFD ones to simulate short transients, using code to code comparisons to validate them (Pérez Mañes et al., 2014, Grahn et al., 2014, Chen et al., 2014).
The aim of this work is to extend the previous methodologies in order to simulate reactor shutdowns with decay heat power and to verify the accuracy of the coupling with code to code comparisons as well as experimental data.
While applying CFD code to reactor simulation, the complexity of the reactor core structure is often a challenge, particularly for transient calculation during safety analysis. In cases where the reactor core detailed distributions are not the focus of the investigation, the porous media approximation is a good approach to simplify the problem. It can offer a reasonable level of accuracy while simplifying the problem and saving computational time.
In our coupling scheme, the neutronics code, in this case PoKiMON, provides the nuclear power distribution to the CFD one, while the latter provides updated data on the material densities and temperatures.
Although this article focuses on capturing the reactor behaviour that follows a boron ingress shut down, it is evident how the scope of this work can be extended towards more generic applications.
It is worth noting that the boron ingress shutdowns and the so-called “density lock” mechanism have been investigated in the 80s and 90s in the context of safety analyses of the Process Inherent Ultimate Safety (PIUS) reactor concept (Pedersen, 1993, Atom Atom, 1989), and they were also part of an experimental campaign (Asak et al., 1992, Tasaka et al., 1994).
However, this was done making use of methodologies and codes nowadays been considered outdated (Harmony et al., 1994, Lime et al., 1993).
In the following section of this paper, a description of the underlying theory is presented. Section three shortly describes the technical details of the coupling implementation.
The fourth section contains a description of the benchmarks that are used to verify and validate that coupled system, like the IFA-507 experiment performed at the Halden reactor facility (Kolstad, 1984).
Section snippets
Theory
In this section, a short description of the theoretical background, that was needed to implement the code, is given.
Coupling
To perform dynamic analysis of reactor core behaviour simulations, PoKiMON code and ANSYS CFX are coupled. The coupling is performed via the ANSYS CFX User Defined Fortran (UDF) interface and its Memory Management System (MMS) (ANSYS Inc, 2016).
In the coupling scheme, as a preprocessing step,
the neutron cross sections and the temperature functions g(T), from eq (15), are calculated via Serpent and read by PoKiMON code as input parameters.
The coupling is an explicit scheme. At the end of each
Results
Several benchmarks were performed to test and verify PoKiMON behaviour and its coupling to ANSYS CFX. In this section, some of these calculations are discussed to show the code performance and accuracy.
Conclusions
In this work, a 1-D code transport reactor dynamic code was developed and coupled with ANSYS CFX code.
The code solves the point-kinetics equation for time-dependent reactor conditions while the axial power distribution is calculated by solving the 1-D neutron transport equation. The reactivity from the boron ingress is accounted for changing the coolant neutron cross sections while the feedback from fuel and coolant temperatures are taken into consideration in the neutron kinetics model via
CRediT authorship contribution statement
F. Tantillo: Conceptualization, Methodology, Software, Investigation. T. Zhang: Conceptualization, Methodology, Investigation, Validation. H.-J. Allelein: Supervision.
Declaration of Competing Interest
The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper.
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2022, Annals of Nuclear EnergyCitation Excerpt :They suggested that the model could be utilized in real-time simulators and the design of control systems. Tantillo et al. (Tantillo et al., 2021) investigated the dynamic behavior of a reactor at full scale by combining a point kinetics model with the CFX commercial CFD solver. They used neutronic feedback to calculate energy sources to deliver to the CFX.