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Thesis (Ph. D.)--Oregon State University, 2010. / Printout. Includes bibliographical references (leaves 183-191). Also available on the World Wide Web.

Thesis (M.S.)--University of Wisconsin--Madison, 1984. / Typescript. eContent provider-neutral record in process. Description based on print version record. Includes bibliographical references (leaf 81).

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Thesis (M.S.)--University of Michigan, 1978.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (p. 263-273). / The economic advantages of longer fuel cycle, improved fuel utilization and reduced spent fuel storage have been driving the nuclear industry to pursue higher discharge burnup of Light Water Reactor (LWR) fuel. A design basis accident, the Reactivity Initiated Accident (RIA), became a concern for further increase of burnup as simulated RIA tests revealed lower enthalpy threshold for fuel failure associated with fuel dispersal, which may compromise the core coolability and/or cause radiological release should this happened in LWRs. Valuable information on the behavior of high burnup fuel during RIA are provided by the simulation tests. However atypical design and operating conditions in simulated tests limited the application of experimental data directly to evaluate the failure potential of LWR fuels. To better interpret the experimental results and improve the capability of the fuel performance codes to predict high burnup fuel behavior, this thesis developed mechanistic models of high burnup fuel during an RIA and implemented models in a transient fuel performance code FRAPTRAN 1.3. Fission gas release (FGR) and swelling were systematically modeled to quantify gaseous loading effects. The grain boundary fission gas inventory is simulated prior to the transient using a diffusion model in FRAPCON 3.3 code. The restructuring of high burnup fuel in rim region is described in terms of porosity, pore size distribution, fission gas concentration, and pore overpressure. The model assumes the fragmentation of fuel upon the separation of grain boundary or when a threshold temperature is exceeded in the rim region. The fission gas in fragmented fuel is assumed to release instantaneously to the free volume when the fuel expansion and swelling creates sufficient pellet-clad gap. / (cont.) The relaxation of rim pore at rapid temperature increase and the thermal expansion of fission gas in fragmented fuel are considered as additional loads on the cladding besides the contact force due to fuel thermal expansion. An analytical approximation is made to calculate the clad radial displacement subjected to fission gas expansion accounting for the constraint of the cladding on the fission gas which would otherwise be neglected in a rigid pellet model FRACAS-I in the FRAPTRAN code. In comparison to the measured FGR from CABRI, NSRR and BIGR test facilities, this mechanistic model can reasonably predict fission gas release fraction for most of the test cases covering a burnup range of 26-64 MWd/kgU and enthalpy deposit of 37-200 cal/g. It reveals the effects of burnup and enthalpy deposit on the fission gas release: burnup is an important parameter affecting fission gas inventory and fuel micro-structure evolution during base irradiation; enthalpy deposit is directly connected to the availability of fission gas release via the grain boundary separation by the intergranular bubble over-pressurization. Analysis of the fission gas radial profile is made with the aid of the neutronic code MCODE to validate the fission gas release from the rim of UO2 fuel. The analysis indicates fission gas release is partly from the rim region and the majority of fission gas release is from grain boundaries for burnup up to 50 MWd/kgU. Fission gas induced hoop strain is predicted to be less than 0.3% in the early phase of RIA with peak fuel enthalpy less than 145 cal/g. Given the fact that the concerned failure mode is the PCMI failure at low energy deposit, the pellet thermal expansion is still considered as effective in analyzing the PCMI failure. However at high level of enthalpy deposit, when clad yield strength is decreased at escalated temperature due to film boiling, the fission gas either released into the plenum or retained in the fuel pellet might strain more the cladding. / (cont.) This is observed in the large deformation of the cladding in some test cases in NSRR and BIGR due to pressure load. A new set of heat transfer correlations were selected and implemented in the FRAPTRAN code to model the cladding-coolant heat transfer of high burnup fuel at room temperature and atmospheric pressure condition. This new set of correlations addressed the effects of subcooling and oxiation on the heat transfer characteristics at pool boiling conditions. They reflect the increase of rewetting temperature and increase of Critical Heat Flux (CHF) due to subcooling. They account for oxidation effects on the transition and film boiling regime and heat conduction through thick oxide as the oxidation is considered as a prominent feature of surface condition change of high burnup fuel. In addition to high burnup fuels tested in NSRR, several fresh fuel tests with different degree of subcooling and a few separate-effects RIA tests are also included to validate the applicabilty of this set of correlations. For fuel enthalpy up to 190 cal/g and oxidation up to 25 micron, the predicted peak cladding temperature (PCT) and duration of DNB achieves generally good agreement with the experimental data. The analysis of high burnup fuel heat transfer reveals that the surface oxidation could cause an early rewetting of high burnup fuel or suppression of DNB. Surface oxidation can delay the heat conducting to the surface while keeping the surface heat transfer in the effective nucleate boiling regime. It also raises the miniumum stable film boiling temperature by lowering the interface temperature during liquid-solid contact resulting from vapor breaking down. Also modeled was Pellet-Cladding Mechanical Interaction (PCMI) failure of irradiated and hydrided cladding. The hydride rim accumulated at outer clad is assumed to cause the crack initiation. The fracture toughness of irradiated and hydrided cladding is obtained by fitting experimental data at different temperature range. / (cont.) The model sets forth a simple criterion for failure associated with crack growth based on the J integral approach. The simplification is that for the thin clad, failure is assumed to occur at the onset of crack tip growth. In comparison to CABRI and NSRR test results and other failure models, the model shows a good capability to separate the failure cases from non-failure cases. These models have been applied to LWR conditions to determine the failure potential of high burnup fuel. It shows that, at high burnup (and therefore high hydride levels in the cladding), the failure enthalpy is smaller than at low burnup. The pulse width is an important parameter in the burnup up to 50 MWd/kg, but starts to become less important for higher burnup with highly corroded cladding. / by Wenfeng Liu. / Ph.D.

This thesis presents the assessment of the Multi Application Small Light Water Reactor (MASLWR) containment design during steady-state and transient conditions. The MASLWR project is a joint effort between Idaho National Environmental and Engineering Laboratory (INEEL), NEXANT Bechtel, and Oregon State University. The project is funded under a Nuclear Energy Research Initiative (NERI) grant from the Department of Energy (DOE). The GOTHIC code was used to simulate the full scale prototype and the Oregon State University MASLWR test facility. Detailed models of the full scale prototype and OSU test facility were generated in GOTHIC. GOTHIC condensation heat transfer models produced heat transfer coefficients that vary by an order of magnitude. This had a significant impact on the pressurization rate and peak pressure achieved within containment. A comparison of the GOTHIC calculation results for the full scale prototype and the test facility model shows reasonable agreement with respect to containment pressure trends and safety system mass flow rates. / Graduation date: 2003

The Oregon State University (OSU) Radiation Center (RC) is the location of a one quarter scale model of the Westinghouse Electric Corporation advanced light-water nuclear reactor design called the AP-600. The full scale AP-600 is a 600 megawatt electric nuclear power plant that incorporates unique passive systems to perform the safety functions currently required of all existing nuclear power plants. Passive safety refers to a system's ability to perform its desired function using natural forces such as gravity and natural circulation. This reduces the reliance on active systems to assure plant safety. The Advanced Plant Experiment (APEX) at the OSU RC is an electrically heated simulation of the AP-600 that includes the Nuclear Steam Supply System (NSSS) and all of the passive safety systems. The APEX facility was funded by the United States Department of Energy and the Westinghouse Electric Corporation. The facility was built to perform the long term cooling tests necessary for design certification of the AP-600. The data taken will be used to benchmark the thermal hydraulic computer codes applied in the design certification process and to better understand the phenomena involved in the full scale AP-600. This paper presents the analysis of the Passive Residual Heat Removal System (PRHR) and in particular the PRHR's "c"-shaped heat exchanger (PRHR Hx). This paper includes analysis and modeling of the PRHR Hx including: hydraulic flow parameters, heat rejection capability, an empirical correlation for determining pressure drop, and an examination of the flow phenomena that occurs in the tank in which the heat exchanger is installed. / Graduation date: 1997

The OSU/APEX thermal hydraulic test facility models the passive safety systems of the Westinghouse AP600 advanced light water reactor design. Numerous experiments have been performed to test these systems, the one of focus here is the station blackout scenario. This experiment simulated the complete loss of AC power to all plant systems. One of the objectives of this experiment was to determine the effectiveness of the Passive Residual Heat Removal (PRHR) system. The PRHR system removes heat by rejecting it into the In-containment Refueling Water Storage Tank (IRWST). The IRWST houses the PRHR and is used as a heat sink for the decay heat. The PRHR is a C-type tube heat exchanger. Heat is removed through two mechanisms: natural convection and nucleate boiling from the surface of the PRHR. As the experiment progressed, a large degree of thermal stratification was observed in the IRWST with no significant thermal mixing. A thermal layer developed in the top of the tank and as the thermal layer approached saturation the rate of heat removal from the sections of the PRHR engulfed by this layer decreased. The effectiveness of these sections of the PRHR continued to decrease until unexpected flow patterns developed at the same time that the thermal layer reached saturation. The IRWST fluid exhibited a bulk azimuthal flow pattern that increased the effectiveness of the PRHR. This increase allowed for more heat to be injected into the IRWST. However, the bulk fluid motion still did not mix the thermal layers. A three-dimensional computational fluid dynamic model using the CFX-4.2 software was developed to study the PRHR/IRWST system. The model uses the RPI method to account for the sub-cooled boiling that is present on the PRHR surface. The model successfully predicted the thermal stratification in the IRWST to within 4 K of experimental data. A counter-current flow was shown to occur along the interface of the thermal layers. This caused an enhancement of the heat transfer and turbulent mixing occurring across the interface of the thermal layers. / Graduation date: 1998