Development of High-fidelity Multi-physics System for Light Water Reactor Analysis

Open Access
Author:
Magedanz, Jeffrey William
Graduate Program:
Nuclear Engineering
Degree:
Doctor of Philosophy
Document Type:
Dissertation
Date of Defense:
October 10, 2013
Committee Members:
  • Maria Nikolova Avramova, Dissertation Advisor
  • Maria Nikolova Avramova, Committee Chair
  • Kostadin Nikolov Ivanov, Committee Chair
  • Armin Seubert, Special Member
  • Fan Bill B Cheung, Committee Member
  • Suzanne Michelle Shontz, Committee Member
Keywords:
  • coupled codes
  • light water reactor simulation
  • high fidelity
  • fuel performance
  • discrete ordinates
Abstract:
Nuclear simulation codes are vital for optimizing the performance of reactors while ensuring that they remain within regulatory safety margins. These codes tend to specialize in one part of the physical phenomena of a reactor core, such as Reactor Physics, Thermal Hydraulics, or Fuel Performance. However, the overall behavior of the system depends on the synergistic interaction of all of these phenomena – they cannot be entirely separated from each other. For example, the reactivity and thus rate of change of the power generation is affected by the fuel temperature and coolant density, and the power in turn deposits heat inside the fuel rods. Thus, in a code specializing in one field, parameters from the other fields have traditionally been determined by simplified models, or by boundary conditions. Given that such simplification is not always acceptable, especially for transient and accident analysis involving reactivity, coupled codes have become a common way of representing the feedback between fields – the codes are combined in such a way that they exchange information. It is especially common to develop core simulators that consist of coupled neutron-kinetics (time-dependent Reactor Physics) and thermal-hydraulics codes. The neutron-kinetics code calculates the neutron flux and the resulting power distribution in the core, which it passes to the thermal-hydraulics code, which in turn solves the heat transfer in the fuel rods, and the energy, momentum, and mass equations for each phase of the coolant, passing back to the neutron-kinetics code the feedback from the fuel and coolant conditions. The spatial exchange of feedback parameters currently is performed in a coarse-mesh framework based on assembly-wise neutronics model and channel-wise model in thermal-hydraulics. While current methods are adequate for many applications, there has been a tendency in recent years toward greater heterogeneity in reactor cores, due to the use of mixed-oxide (MOX) fuel, burnable absorbers, and longer cycles with consequently higher fuel burnup. The resulting asymmetry of the neutron flux and energy spectrum between regions with different compositions causes a need to account for the directional dependence of the neutron flux, instead of the traditional diffusion approximation. Furthermore, the presence of both MOX and high-burnup fuel in the core increases the complexity of the iv heat conduction. The heat transfer properties of the fuel pellet change with irradiation, and the thermal and mechanical expansion of the pellet and cladding strongly affect the size of the gap between them, and its consequent thermal resistance. These operational tendencies require higher fidelity multi-physics modeling capabilities, and this need is addressed by the developments performed within this PhD research. The dissertation describes the development of a High-Fidelity Multi-Physics System for Light Water Reactor Analysis. It consists of three coupled codes – CTF for Thermal Hydraulics, TORT-TD for Neutron Kinetics, and FRAPTRAN for Fuel Performance. It is meant to address these modeling challenges in three ways: (1) by resolving the state of the system at the level of each fuel pin, rather than homogenizing entire fuel assemblies, (2) by using the multi-group Discrete Ordinates method to account for the directional dependence of the neutron flux, and (3) by using a fuel-performance code, rather than a Thermal Hydraulics code's simplified fuel model, to account for the material behavior of the fuel and its feedback to the hydraulic and neutronic behavior of the system. While the first two are improvements, the third, the use of a fuel-performance code for feedback, constitutes an innovation in this PhD project. Also important to this work is the manner in which such coupling is written. While coupling involves combining codes into a single executable, they are usually still developed and maintained separately. It should thus be a design objective to minimize the changes to those codes, and keep the changes to each code free of dependence on the details of the other codes. This will ease the incorporation of new versions of the code into the coupling, as well as re-use of parts of the coupling to couple with different codes. In order to fulfill this objective, an interface for each code was created in the form of an object-oriented abstract data type. Object-oriented programming is an effective method for enforcing a separation between different parts of a program, and clarifying the communication between them. The interfaces enable the main program to control the codes in terms of high-level functionality. This differs from the established practice of a master/slave relationship, in which the slave code is incorporated into the master code as a set of subroutines. v While this PhD research continues previous work with a coupling between CTF and TORT-TD, it makes two major original contributions: (1) using a fuel-performance code, instead of a thermal-hydraulics code's simplified built-in models, to model the feedback from the fuel rods, and (2) the design of an object-oriented interface as an innovative method to interact with a coupled code in a high-level, easily-understandable manner. The resulting code system will serve as a tool to study the question of under what conditions, and to what extent, these higher-fidelity methods will provide benefits to reactor core analysis.