Since the 1990s, researchers around the world have been creating antineutrino detectors for monitoring power reactors. These detectors have been deployed at light water reactors and are able to determine power levels and burn up throughout a fuel cycle. This technology could allow the IAEA to monitor LWRs remotely and unobtrusively to determine if they are operating using normal parameters. Very soon, the next generation of detector will be deployed at a CANDU reactor for a trial operation. While physical observation of these detectors is necessaryl in determining their usefulness, reactor physics simulations have proven to be very accurate in their prediction of detector performance. Since there are many designs still in development, reactor physics simulations are the only way to determine the efficacy of the detector technology. In addition to this, reactor simulations are the best way to evaluate the detector technology to ascertain its usefulness during diversion scenarios. In this research, antineutrino source terms were calculated for a High Temperature Gas Cooled Reactor core. These source terms were a function of power level and initial enrichment. SCALE6.1, developed by Oak Ridge National Laboratory, was used to calculate the isotopic inventory in the core as a function of depletion. These fertile and fissile isotopics, along with the fission cross-section and number of antineutrinos emitted per fission, were used to predict the antineutrino source rate for the core. It was found that changing the power yields a linear response from the antineutrino source term. By increasing the power by five percent, the source term also increased by five percent. Substantial changes in the initial enrichment also lead to a detectable change in the antineutrino source term. / Graduation date: 2012

Two topics from nuclear engineering field are included in this dissertation. One study is the air-ingress phenomenon during a loss of coolant accident (LOCA) scenario, and the other is a 5-by-5 bundle assembly problem under a design of PWRs. The objectives are to investigate the Kelvin-Helmholtz instability of the gravity-driven stratified flows inside a coaxial pipe and the effects caused by two types of spacers at the downstream of the rod bundle problem. Richardson extrapolation is used for the grid independent study. Simulation results give good agreements with the experiments. Wavelet analysis and Proper Orthogonal Decomposition (POD) are used to study the flow behaviors and flow patterns. For the air-ingress phenomenon, Brunt-Vaisala frequency, or buoyancy frequency, predicts a frequency of 2.34 Hz, which is confirmed by the dominant frequency of 2.4 Hz obtained from the wavelet analysis between times 1.2 s and 1.85 s. For the rod bundle study, the dominant frequency at the center of the subchannel is given as 2.4 Hz with a secondary dominant frequency of 4 Hz and a much minor frequency of 6 Hz. Generally, wavelet analysis has much better performance than POD in the air-ingress phenomenon that is a strongly transient scenario; they both appropriate for the rod bundle study. Based on this study, when the fluid pair in a real condition is used, the time which air intrudes into the reactor is predictable.

High Temperature Gas-cooled Reactors (HTGRs) can provide clean electricity,as well as process heat that can be used to produce hydrogen for transportation and other sectors. A prototypic HTGR, the Next Generation Nuclear Plant (NGNP),will be built at Idaho National Laboratory.The need for HTGR analysis tools and methods has led to the addition of gas-cooled reactor (GCR) capabilities to the light water reactor code MELCOR. MELCOR will be used by the Nuclear Regulatory Commission licensing of the NGNP and other HTGRs. In the present study, new input techniques have been developed for MELCOR HTGR analysis. These new techniques include methods for modeling radiation heat transfer between solid surfaces in an HTGR, calculating fuel and cladding geometric parameters for pebble bed and prismatic block-type HTGRs, and selecting appropriate input parameters for the reflector component in MELCOR. The above methods have been applied to input decks for a water-cooled reactor cavity cooling system (RCCS); the 400 MW Pebble Bed Modular Reactor (PBMR), the input for which is based on a code-to-code benchmark activity; and the High Temperature Test Facility (HTTF), which is currently in the design phase at Oregon State University. RCCS results show that MELCOR accurately predicts radiation heat transfer rates from the vessel but may overpredict convective heat transfer rates and RCCS coolant flow rates. PBMR results show that thermal striping from hot jets in the lower plenum during steady-state operations, and in the upper plenum during a pressurized loss of forced cooling accident, may be a major design concern. Hot jets could potentially melt control rod drive mechanisms or cause thermal stresses in plenum structures. For the HTTF, results will provide data to validate MELCOR for HTGR analyses. Validation will be accomplished by comparing results from the MELCOR representation of the HTTF to experimental results from the facility. The validation process can be automated using a modular code written in Python, which is described here.

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No description available.

With the coming of Generation IV systems, there is a need for thermal-hydraulic codes to model such advanced reactors. Codes for High Temperature Gas-cooled Reactors (HTGRs) already exist, but often suffer from insufficient validation and little user experience. Therefore, some existing codes created for Light Water Reactors are updated with HTGR related features. In this study, the feasibility of providing the TRACE thermal-hydraulics code with those features is analysed. Two models were used, one of a pebble bed core, one of a prismatic block reactor. For this purpose the effective conductivity test of the High Temperature Test Unit was taken as a benchmark for the pebble bed core. For the prismatic block reactor a model of the GEMINI+ reactor was created. This would allow to simulate not only steady state, but also Depressurised Loss of Forced Cooling scenarios. For both models the effective conductivity is known to play a major role and, consequently, a model to incorporate such feature was developed and implemented within TRACE's control system module. Results show that TRACE has a good potential for HTGR simulation, but currently available models still provide unstable solutions. It is concluded that TRACE needs additional adjustments in order to be employed for HTGR safety analyses in the future. / Med fjärde generationens system på ingång, finns det ett behov av termisk-hydrauliska koder för att modellera sådana avancerade reaktorer. Koder för gaskylda högtemperaturreaktorer (HTGR) finns redan, men lider ofta av otillräcklig validering och liten användarupplevelse. Därför uppdateras vissa befintliga koder som skapats för lättvattenreaktorer med HTGR-relaterade funktioner. I denna studie analyseras möjligheten att tillhandahålla TRACE termisk-hydraulisk kod med dessa funktioner. Två modeller användes, den ena av en pebble-bed reaktor, den andra av en prismatisk blockreaktor. För detta ändamål togs det effektiva konduktivitetstestet för högtemperaturtestenheten som ett riktmärke för pebble-bedens härd. För den prismatiska blockreaktorn skapades en modell av GEMINI+-reaktorn. Detta skulle göra det möjligt att simulera inte bara steady state, utan även scenarier med trycklös förlust av forcerad kylning. För båda modellerna är den effektiva konduktiviteten känd för att spela en stor roll och följaktligen utvecklades och implementerades en modell för att införliva en sådan funktion inom TRACEs kontrollsystemmodul.. Resultaten visar att TRACE har en god potential för HTGR-simulering, men för närvarande tillgängliga modeller ger fortfarande instabila lösningar. Slutsatsen är att TRACE behöver ytterligare justeringar för att kunna användas för HTGR-säkerhetsanalyser i framtiden.

This research project will investigate the potential business case and technical feasibility of using nuclear generated heat in a crude oil refinery located some distance away. The key design element is an energy transportation mechanism that doesn’t compromise the safety, licensing or operability of the nuclear plant. In a crude oil refinery processing heat is generated by combusting fuels that are generally sellable products. The inherent safety features and high output temperature of a HTGR make it an appropriate replacement heat source for such a processing plant. An opportunity thus exists to replace the refinery hydrocarbon fuel usage with nuclear energy thereby improving refinery profitability. Three alternate proposed were generated. Alt 1: Generation of steam at HTGR, piped to the refinery to replace current supply. Alt 2: Closed loop reversible methanation reaction delivering potential chemical energy to the refinery which is released to the process in heat exchangers. Alt 3: Hydrogen production from water splitting at the HTGR, piped to the refinery and combusted in boilers or used for hydrotreating diesel. Utilizing data from refinery plant historian and journals, a basic engineering study assessed technical feasibility thereof. An economic model for the 2 most promising alternates was set up using quotations and factored data and evaluated against the existing refinery situation. A consistently increasing crude price was assumed. Alternates 1, 2 and 3 proved technically feasible and delivered 86 MW, 59 MW and 48MW to the refinery respectively. Generating steam at the HTGR (Alt 1) demonstrated an attractive business case, strengthened by co-locating the nuclear plant at the refinery. It is therefore concluded that using a HTGR for process heat in a petrochemical plant such as a refinery is techno-economically practical and demands further consideration. If future carbon emission legislation is promulgated this proposal will be key component of the solution. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2014

This research project will investigate the potential business case and technical feasibility of using nuclear generated heat in a crude oil refinery located some distance away. The key design element is an energy transportation mechanism that doesn’t compromise the safety, licensing or operability of the nuclear plant. In a crude oil refinery processing heat is generated by combusting fuels that are generally sellable products. The inherent safety features and high output temperature of a HTGR make it an appropriate replacement heat source for such a processing plant. An opportunity thus exists to replace the refinery hydrocarbon fuel usage with nuclear energy thereby improving refinery profitability. Three alternate proposed were generated. Alt 1: Generation of steam at HTGR, piped to the refinery to replace current supply. Alt 2: Closed loop reversible methanation reaction delivering potential chemical energy to the refinery which is released to the process in heat exchangers. Alt 3: Hydrogen production from water splitting at the HTGR, piped to the refinery and combusted in boilers or used for hydrotreating diesel. Utilizing data from refinery plant historian and journals, a basic engineering study assessed technical feasibility thereof. An economic model for the 2 most promising alternates was set up using quotations and factored data and evaluated against the existing refinery situation. A consistently increasing crude price was assumed. Alternates 1, 2 and 3 proved technically feasible and delivered 86 MW, 59 MW and 48MW to the refinery respectively. Generating steam at the HTGR (Alt 1) demonstrated an attractive business case, strengthened by co-locating the nuclear plant at the refinery. It is therefore concluded that using a HTGR for process heat in a petrochemical plant such as a refinery is techno-economically practical and demands further consideration. If future carbon emission legislation is promulgated this proposal will be key component of the solution. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2014

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