http://hdl.handle.net/1721.1/7857 2013-05-19T13:43:37Z http://hdl.handle.net/1721.1/76589 Uncertainty and sensitivity analysis of a fire-induced accident scenario involving binary variables and mechanistic codes Minton, Mark A. (Mark Aaron) In response to the transition by the United States Nuclear Regulatory Commission (NRC) to a risk-informed, performance-based fire protection rulemaking standard, Fire Probabilistic Risk Assessment (PRA) methods have been improved, particularly in the areas of advanced fire modeling and computational methods. As the methods for the quantification of fire risk are improved, the methods for the quantification of the uncertainties must also be improved. In order to gain a more meaningful insight into the methods currently in practice, it was decided that a scenario incorporating the various elements of uncertainty specific to a fire PRA would be analyzed. The NRC has validated and verified five fire models to simulate the effects of fire growth and propagation in nuclear power plants. Although these models cover a wide range of sophistication, epistemic uncertainties resulting from the assumptions and approximations used within the model are always present. The uncertainty of a model prediction is not only dependent on the uncertainties of the model itself, but also on how the uncertainties in input parameters are propagated throughout the model. Inputs to deterministic fire models are often not precise values, but instead follow statistical distributions. The fundamental motivation for assessing model and parameter uncertainties is to combine the results in an effort to calculate a cumulative probability of exceeding a given threshold. This threshold can be for equipment damage, time to alarm, habitability of spaces, etc. Fire growth and propagation is not the only source of uncertainty present in a fire-induced accident scenario. Statistical models are necessary to develop estimates of fire ignition frequency and the probability that a fire will be suppressed. Human Reliability Analysis (HRA) is performed to determine the probability that operators will correctly perform manual actions even with the additional complications of a fire present. Fire induced Main Control Room (MCR) abandonment scenarios are a significant contributor to the total Core Damage Frequency (CDF) estimate of many operating nuclear power plants. Many of the resources spent on fire PRA are devoted to quantification of the probability that a fire will force operators to abandon the MCR and take actions from a remote location. However, many current PRA practitioners feel that effect of MCR fires have been overstated. This report details the simultaneous application of state-of-the-art model and parameter uncertainty techniques to develop a defensible distribution of the probability of a forced MCR abandonment caused by a fire within a MCR benchboard. These results are combined with the other elements of uncertainty present in a fire-induced MCR abandonment scenario to develop a CDF distribution that takes into account the interdependencies between the factors. In addition, the input factors having the strongest influence on the final results are identified so that operators, regulators, and researchers can focus their efforts to mitigate the effects of this class of fire-induced accident scenario. Thesis (Nucl. E. and S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010.; "September 2010." Cataloged from PDF version of thesis.; Includes bibliographical references (p. 72-74). 2010-01-01T05:00:00Z http://hdl.handle.net/1721.1/63242 Feasibility of lateral emplacement in very deep borehole disposal of high level nuclear waste Gibbs, Jonathan Sutton The U.S. Department of Energy recently filed a motion to withdraw the Nuclear Regulatory Commission license application for the High Level Waste Repository at Yucca Mountain in Nevada. As the U.S. has focused exclusively on geologic disposal in shallow mined repositories for the past two decades, an examination of disposal alternatives will be necessary should the Yucca Mountain Project be terminated. This provides an opportunity to study other promising waste disposal technologies. One such technology is the use of very deep boreholes in monolithic granite to permanently segregate high level wastes from the biosphere. While research in this field has focused on vertical emplacement techniques, horizontal emplacement offers the significant advantages of allowing increased emplacement lengths without crushing of the waste package and the use of a single vertical shaft for drilling multiple horizontal shafts. This project examines the application of currently deployed oil and natural gas directional drilling techniques to borehole design. A large trade-space of potential borehole configurations is evaluated and a final design selected using the "V-DeepBoRe" code, a Monte-Carlo simulation based cost model for borehole construction and waste package emplacement. Waste repackaging and reconstitution is evaluated to permit deployment of waste in borehole diameters too small for intact fuel assemblies. A 5 m x 195.26 mm (OD) cylindrical waste package is designed using P-110 drill string steel to meet strength and thermal loading requirements; fuel centerline temperatures are shown to not exceed 190'C by analytical and finite element methods. The total cost of a national borehole repository (including drilling, consolidating and encapsulating the fuel, emplacement, and closure) is shown to fall below $63/kgHM, well within the capacity of the DOE Commercial Spent Nuclear Fuel Waste Fund. Thesis (Nav. E.)--Massachusetts Institute of Technology, Dept. of Mechanical Engineering; and, (Nucl. E.)-- Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010.; Cataloged from PDF version of thesis.; Includes bibliographical references (p. 151-154). 2010-01-01T05:00:00Z http://hdl.handle.net/1721.1/58856 Design of a coordinated plant control system for a marine nuclear propulsion plant Baltra Aedo, Guillermo Thesis (Nucl. E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering; and, (M.S.)--Massachusetts Institute of Technology, Dept. of Ocean Engineering, 1981.; MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE.; Vita.; Includes bibliographical references. 1981-01-01T05:00:00Z http://hdl.handle.net/1721.1/54904 Fuel element performance maps for nuclear reactor operational decisions. Da Silva, Othon Luiz Pinheiro Thesis. 1978. Nucl.E.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering.; MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE.; Includes bibliographical references. 1978-01-01T05:00:00Z