The Generation IV International Forum has established several goals for the next generation of nuclear energy systems, which are to be substantial improvements over contemporary designs. In Canada Generation IV research efforts have focused on developing the Pressure Tube type SuperCritical Water-cooled Reactor (PT-SCWR), an evolution of CANada Deuterium Uranium (CANDU) technology. An integral part of the PT-SCWR is the High Efficiency Re-entrant Channel (HERC), wherein coolant first travels downward through a centre flow tube and then upward around the fuel. The large density variation of supercritical fluids, combined with the negative Coolant Void Reactivity (CVR), make the concept similar to a Boiling Water Reactor (BWR). The objective of this study was thus to apply the state-of-the-art in BWR analysis to the PT-SCWR. Models were created using the DRAGON (neutron transport), DONJON (neutron diffusion/spatial kinetics), and CATHENA (channel thermalhydraulics) computer codes. A procedure for DONJON-CATHENA coupling was developed to enable simulation of coupled transients. The specifications of the HERC necessitated multiple coolant reactivity feedbacks be included in the model, in turn requiring extensions to the DONJON source code. The model created for this work is thus among the first to incorporate multiple coolant feedbacks in core-level coupled spatial kinetics and thermalhydraulics transient analysis, and is uniquely capable of simulating such transients in the PT-SCWR. This work found that while the total CVR was negative as required, the reactivity effect of coolant void solely around the fuel was positive. As a consequence additional heat delivered from fuel to coolant, which decreases the coolant density, has a positive reactivity effect making BWR-like coupled instabilities impossible. On the other hand, in some postulated transients, such as Loss-OfCoolant Accidents (LOCAs) or Loss-Of-Flow Accidents (LOFAs), this positive reactivity results in temporary power excursions. A fast-acting shutdown system is potentially necessary to limit damage to the fuel in such transients. / Dissertation / Doctor of Philosophy (PhD)

Recent work by Yasseri and Rahnema has introduced a consistent spatial homogenization (CSH) method completely in transport theory. The CSH method can very accurately reproduce the heterogeneous flux shape and eigenvalue of a reactor, but at high computational cost. Other recent works for homogenization in diffusion or quasi-diffusion theory are accurate for problems with low heterogeneity, such as PWRs, but are not proven for more heterogeneous reactors such as BWRs or GCRs. To address these issues, a consistent hybrid diffusion-transport spatial homogenization (CHSH) method is developed as an extension of the CSH method that uses conventional flux weighted homogenized cross sections to calculate the heterogeneous solution. The whole-core homogenized transport calculation step of the CSH method has been replaced with a whole- core homogenized diffusion calculation. A whole-core diffusion calculation is a reasonable replacement for transport because the homogenization procedure tends to smear out transport effects at the core level. The CHSH solution procedure is to solve a core-level homogenized diffusion equation with the auxiliary source term and then to apply an on-the-fly transport-based re-homogenization at the assembly level to correct the homogenized and auxiliary cross sections. The method has been derived in general geometry with continuous energy, and it is implemented and tested in fine group, 1-D slab geometry on controlled and uncontrolled BWR and HTTR benchmark problems. The method converges to within 2% mean relative error for all four configurations tested and has computational efficiency 2 to 4 times faster than the reference calculation.

The South African government’s integrated resource plan for electricity IRP2010 states that the country plans to have an additional 9.6 GW of nuclear power on the national electricity grid by 2030. In support of this, the NRF-funded SARChI Research Chair in Nuclear Engineering within the School of Mechanical and Nuclear Engineering at the North-West University recently initiated research studies focused on Light Water Reactor (LWR) systems. These studies inter alia involve coupled neutronic and thermal hydraulic analyses of selected LWR systems. This study focuses on the steady state neutronic analysis of the European Pressurised Reactor (EPR) using Monte-Carlo N-Particle (MCNP5). The neutronic model will in due course be coupled to a thermal hydraulic model forming part of a broader study of the system. The Monte Carlo neutron transport code MCNP5 has been widely used since the 1950s for analysis of existing and future reactor systems due to its ability to simulate complex fuel assemblies without making any significant approximations. The primary aim of the study was to develop an input model for a representative fresh fuel assembly of the US EPR reactor core from which the fluxes and fission power of the reactor can be obtained. There after a 3D model of full EPR core developed by the school of mechanical and nuclear engineering based on findings of this work is also tested. The results are compared to those in the US EPR Final Safety Analysis Report. Agreement in major core operational parameters including the keff eigenvalue, axial and radial power profiles and control rod worth are evaluated, from which consistency of the model and results will be confirmed. Further convergence of the model within a reasonable time is assessed. / MSc (Engineering Sciences in Nuclear Engineering), North-West University, Potchefstroom Campus, 2014

The South African government’s integrated resource plan for electricity IRP2010 states that the country plans to have an additional 9.6 GW of nuclear power on the national electricity grid by 2030. In support of this, the NRF-funded SARChI Research Chair in Nuclear Engineering within the School of Mechanical and Nuclear Engineering at the North-West University recently initiated research studies focused on Light Water Reactor (LWR) systems. These studies inter alia involve coupled neutronic and thermal hydraulic analyses of selected LWR systems. This study focuses on the steady state neutronic analysis of the European Pressurised Reactor (EPR) using Monte-Carlo N-Particle (MCNP5). The neutronic model will in due course be coupled to a thermal hydraulic model forming part of a broader study of the system. The Monte Carlo neutron transport code MCNP5 has been widely used since the 1950s for analysis of existing and future reactor systems due to its ability to simulate complex fuel assemblies without making any significant approximations. The primary aim of the study was to develop an input model for a representative fresh fuel assembly of the US EPR reactor core from which the fluxes and fission power of the reactor can be obtained. There after a 3D model of full EPR core developed by the school of mechanical and nuclear engineering based on findings of this work is also tested. The results are compared to those in the US EPR Final Safety Analysis Report. Agreement in major core operational parameters including the keff eigenvalue, axial and radial power profiles and control rod worth are evaluated, from which consistency of the model and results will be confirmed. Further convergence of the model within a reasonable time is assessed. / MSc (Engineering Sciences in Nuclear Engineering), North-West University, Potchefstroom Campus, 2014

No description available.

The National Bureau of Standards Reactor (NBSR) located on the National Institute of Standards and Technology (NIST) Gaithersburg campus, is currently underway of fuel conversion from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. One particular challenging part of the conversion of the NBSR is the high average flux level (2.5×1014 n/cm2-s) required to maintain experimental testing capabilities of the reactor, without significant changes to the external structures of the reactor. Recently the General Atomics (GA) Training Research Isotopes General Atomics (TRIGA) fuel has shown some promising features as a LEU candidate for the high performance research reactors such as the NBSR. The GA fuel has a long history of success in conversion of research reactors since it was developed in 1980s. The UZrH compound in the GA fuel has seen success in long term TRIGA reactors, and is a proven safe LEU alternative. This study performs a neutronics evaluation of the TRIGA fuel under the schema of the NBSR’s heavy conversion requirements in order to examine whether the TRIGA fuel is a viable option for conversion of the NBSR. To determine the most optimal path of conversion, we performed a feasibility study with particular regard to the fuel dimensions, fuel rod configurations, cladding, as well as fuel structure selection. Based on the outcome of the feasibility study, an equilibrium core is then generated following the NBSR’s current fuel management schema. Key neutronics performance characteristics including flux distribution, power distribution, control rod (i.e., shim arms) worth, as well as kinetics parameters of the equilibrium core are calculated and evaluated. MCNP6, a Monte Carlo based computational modeling software was intensively used to aid in these calculations. The results of this study will provide important insight on the effectiveness of conversion, as well as determine the viability of the conversion from HEU to LEU using the GA fuel.

Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2008.

This work presents a whole-core benchmark problem based on a 2-loop pressurized water reactor with both UO₂and MOX fuel assemblies. The specification includes heterogeneity at both the assembly and core level. The geometry and material compositions are fully described and multi-group cross section libraries are provided in 2, 4, and 8 group formats. Simplifications made to the benchmark specification include a Cartesian boundary, to facilitate the use of transport codes that may have trouble with cylindrical boundaries, and control rod homogenization, to reduce the geometric complexity of the problem. These modifications were carefully chosen to preserve the physics of the problem and a justification of these modifications is given. Detailed Monte Carlo reference solutions including core eigenvalue, assembly averaged fission densities and selected fuel pin fission densities are presented for benchmarking diffusion and transport methods. Three different core configurations are presented in the paper namely all-rods-out, all-rods-in, and some-rods-in.

The high and ever growing demand for electricity coupled with environmental concerns and a worldwide desire to shed petroleum dependence, all point to a shift to utilization of renewable sources of energy. The under developed nature of truly renewable energy sources such as, wind and solar, along with their limitations on the areas of applicability and the energy output calls for a renaissance in nuclear energy. In this second nuclear era, deliberately small reactors are poised to play a major role with a number of Small Modular Reactors (SMRs) currently under development in the U.S. In this work, an SMR model of the Integral Pressurized Water Reactor (IPWR) type is created, analyzed and optimized to meet the publically available performance criteria of the mPower SMR from B&W. The Monte Carlo codes MCNP5/MCNPX are used to model the core. Fuel enrichment, core inventory, core size are all variables optimized to meet the set goals of core lifetime and fuel utilization (burnup). Vital core behavior characteristics such as delayed neutron fraction and reactivity coefficients are calculated and shown to be typical of larger PWR systems, which is necessary to ensure the inherent safety and to achieve rapid deployment of the reactor by leveraging the vast body of operational experience amassed with the larger commercial PWRs. Inherent safety of the model is analyzed with the results of an analytical single channel analysis showing promising behavior in terms of axial and radial fuel element temperature distributions, the critical heat flux, and the departure from nucleate boiling ratio. The new fleet of proposed SMRs is intended to have increased proliferation resistance (PR) compared to the existing fleet of operating commercial PWRs. To quantify this PR gain, a PR analysis is performed using the Proliferation Resistance Analysis and Evaluation Tool for Observed Risk (PRAETOR) code developed by the Nuclear Science and Security Policy Institute at Texas A&M University. The PRAETOR code uses multi-attribute utility analysis to combine 63 factors affecting the PR value of a facility into a single metric which is easily comparable. The analysis compared hypothetical spent fuel storage facilities for the SMR model spent fuel assembly and one for spent fuel from a Westinghouse AP1000. The results showed that from a fuel material standpoint, the SMR and AP1000 had effectively the same PR value. Unable to analyze security systems and methods employed at specific nuclear power plant sites, it is premature to conclude that the SMR plants will not indeed show increased PR as intended.

Traditionally coupled field reactor analysis has been carried out using several loosely coupled solvers, each having been developed independently from the others. In the field of multi-physics, the current generation of object-oriented toolkits provides robust close coupling of multiple fields on a single framework. This research investigates the suitability of such frameworks, in particular the Open-source Field Operation and Manipulation (OpenFOAM) framework, for the solution of spatial reactor dynamics problems. For this a subset of the theory of the Time-dependent Neutronics and Temperatures (TINTE) code, a time-dependent two-group diffusion solver, was implemented in the OpenFOAM framework. This newly created code, called diffusionFOAM, was tested for a number of steady-state and transient cases. The solver was found to perform satisfactorily, despite a number of numerical issues. The object-oriented structure of the framework allowed for rapid and efficient development of the solver. Further investigations suggest that more advanced transport methods and higher order spatial discretization schemes can potentially be implemented using such a framework as well. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2008.