Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Mechanical Engineering, 1982. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND ENGINEERING. / Bibliography: leaf 42. / by Hung Viet Vu. / M.S.

Typescript (photocopy). / Digitized by Kansas Correctional Industries

A literature review of current phase separation publications was conducted. Data sets were collected and compiled into a Two-Phase Flow Separation Database. Examination of this database indicating a need for further investigation into the liquid entertainment phenomena for smaller hot-leg to branch diameters and intermittent flow regimes. A detailed analysis to the prototypic phase separation process is presented and the associated phenomena are identified. Appropriate scaling criteria were employed for the design of a scaled test facility. Geometry and the flow conditions of the test facility were determined accordingly to Wu et. al (1998). A series of phase separation tests conducted at the Air-water Test Loop for Advanced Thermal-hydraulic Studies (ATLATS) and Advanced Plan Experiment (APEX) has been completed. Results show that the criteria developed by Smoglie (1984) used in RELAP5, reasonably predicts the onset of liquid entrainment. However, the steady-state entrainment correlation in RELAP5 significantly underpredicts primary coolant removal rates. This discrepancy is due to the effects of downstream boundary conditions and pool entrainment and carry-over from the reactor vessel. Due to pool entrainment, entrainment through the branch continues when the reactor vessel mixture level drops below the bottom of the hot-leg. This investigation shows that RELAP5 is non-conservative when predicting coolant removal rates due to steady state liquid entrainment in a horizontal mainline with a vertical branch for stratified, stratified-wavy, transition, and stepped hot-leg flow regimes. / Graduation date: 2001

An approach for developing a computer-based aid to assist in monitoring and assessing nuclear power plant status during situations requiring emergency response has been developed. It is based on the representation of regulatory requirements and plant-specific systems and instrumentation in the form of hierarchical rules. Making use of inferencing techniques from the field of artificial intelligence, the rules are combined with dynamic state data to determine appropriate emergency response actions. In a joint project with Portland General Electric Company, a prototype system, called EM-CLASS, was been created to demonstrate the knowledge-based approach for use at the Trojan Nuclear Power Plant. The knowledge domain selected for implementation addresses the emergency classification process chat is used to communicate the severity of the emergency and the extent of response actions required. EM-CLASS was developed using Personal Consultant Plus (PCPlus), a knowledge-based system development shell from Texas Instruments which runs on IBM-PC compatible computers. The knowledge base in EM-CLASS contains over 200 rules. The regulatory basis, as defined in 10 CFR 50, calls for categorization of emergencies into four emergency action level classes: (1) notification of unusual event, (2) alert, (3) site area emergency, and (4) general emergency. Each class is broadly defined by expected frequency and the potential for release of radioactive materials to the environment. In a functional sense, however, each class must be ultimately defined by a complex combination of in- plant conditions, plant instrumentation and sensors, and radiation monitoring information from stations located both on- and off-site. The complexity of this classification process and the importance of accurate and timely classification in emergency response make this particular application amenable to an automated, knowledge-based approach. EM-CLASS has been tested with a simulation of a 1988 Trojan Nuclear Power Plant emergency exercise and was found to produce accurate classification of the emergency using manual entry of the data into the program. / Graduation date: 1997

An overall systems design code was developed to model an advanced in-core thermionic energy conversion based nuclear reactor system for space applications at power levels of 10 to 50 kWe. The purpose of this work was to provide the overall shell for the systems code and to also provide the detailed neutronic analysis section of the code. The design code that was developed is to be used to evaluate a reactor system based upon a single cell thermionic fuel element which uses advanced technology to enhance the performance of single cell thermionic fuel elements. A literature survey provided information concerning how other organizations performed system studies on similar space reactor designs. / Graduation date: 1994

Incore thermionic space reactor design concepts which operate at a nominal power output range of 20 to 50 kWe are described. Details of the neutronic, thermionic, thermal hydraulics and shielding performance are presented. Due to the strong absorption of thermal neutrons by natural tungsten, and the large amount of that material within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced incore thermionic energy conversion based nuclear reactor systems for space applications. The code modules include neutronics and core criticality, a thermionic fuel element performance module with integral thermal hydraulics calculation capability, a radiation shielding module, and a module for the waste heat rejection. The results show that the driverless single cell ATI configuration, which does not have driver rods, proved to be more efficient than the driven core, which has driver rods. It also shows that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one thermionic fuel element (TFE) works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in 40 % decrease of the efficiency of the overall system. / Graduation date: 1993

Thermochemical processes for hydrogen production driven by nuclear energy are promising alternatives to existing technologies for large-scale commercial production of hydrogen, without dependence on fossil fuels. In the Copper-Chlorine (Cu-Cl) cycle, water is decomposed in a sequence of intermediate processes with a net input of water and heat, while hydrogen and oxygen gases are generated as the products. The Super Critical Water-cooled Reactor (SCWR) has been identified as a promising source of heat for these processes. In this thesis, the process analysis and simulation models are developed using the Aspen PlusTM chemical process simulation package, based on experimental work conducted at the Argonne National Laboratory (ANL) and Atomic Energy of Canada Limited (AECL). A successful simulation is performed with an Electrolyte Non Random Two Liquid (ElecNRTL) model of Aspen Plus. The efficiency of the cycle based on three and four step process routes is examined in this thesis. The thermal efficiency of the four step thermochemical process is calculated as 45%, while the three step hybrid thermochemical cycle is 42%, based on the lower heating value (LHV) of hydrogen. Sensitivity analyses are performed to study the effects of various operating parameters on the efficiency, yield, and thermodynamic properties. Possible efficiency improvements are discussed. The results will assist the development of a lab-scale cycle which is currently being conducted at the University of Ontario Institute of Technology (UOIT), in collaboration with its partners. / UOIT

A prototype real-time process monitoring emergency classification expert system, RT/EM-CLASS, has been developed for use at the Trojan Nuclear Power Plant. This knowledge-based system features the integration of electronically sensed plant data with the menu selection data representation of its predecessor, EM-CLASS. This prototype demonstrates the techniques required to acquire plant process data from another computer and use that data in an expert system to determine the proper Emergency Action Level. Several benefits are realized by the RT/EM-CLASS application. These include: The resources required to make a classification are reduced thereby freeing the responsible person to devote time to other important tasks. The classification may be completed more often and with better data than the current system allows. The human user is less likely to make an erroneous Emergency Action Level classification. Prototype implementation required resolution of an efficiency problem of relating plant process data to the expert system data forms. This was achieved through the development of multi-conditional rules that significantly reduce the size of the rule set. The development of RT/EM-CLASS presents a methodology for building knowledge based applications that perform nonintrusive real-time monitoring of dynamic systems. This methodology features Use of existing analytical and Al tools where possible Monitoring of a dynamic system Non-intrusive acquisition of data from the system This technology might be applied to portions of the nuclear engineering design process (control rod programming in Boiling Water Reactors, for example) to emulate the guidance and activities of an expert. A substantial portion of the effort by the expert engineer involves preparation of the code input, running the computer code, analyzing the results, and based on the results, deciding what case to submit next. A suitably designed expert system could act in the place of the engineer in this dynamic design process. This methodology has been tested against the 1988 emergency exercise at the Trojan Nuclear Power Plant. / Graduation date: 1992

Preliminary nuclear design studies have been completed on ten configurations of nuclear reactors for low power (1-30 kWe) space applications utilizing thermionic energy conversion. Additional design studies have been conducted on the TRICE multimegawatt in-core thermionic reactor configuration. In each of the cases, a reactor configuration has been determined which has the potential for operating 7 years with sufficient reactivity margin. Additional safety evaluations have been conducted on these configurations including the determination of sufficient shutdown reactivity, and consideration of water immersion, water flooding, sand burial, and reactor compaction accident scenarios. It has been found, within the analysis conducted using the MCNP Monte Carlo neutron transport code, that there are configurations which are feasible and deserve further analysis. It has also been found that solid core reactors which rely solely on conduction for heat removal as well as pin type cores immersed in a liquid metal bath have merit. The solid cores look attractive when flooding and compaction accident scenarios are considered as there is little chance for water to enter the core and cause significant neutron moderation. A fuel volume fraction effect has also been found in the consideration of the sand burial cases for the SP-100 derived configurations. / Graduation date: 1994

The risk-based life-cycle management of engineering systems in a nuclear power plant is intended to ensure safe and economically efficient operation of energy generation infrastructure over its entire service life. An important element of life-cycle management is to understand, model and forecast the effect of various degradation mechanisms affecting the performance of engineering systems, structures and components. The modeling of degradation in nuclear plant components is confounded by large sampling and temporal uncertainties. The reason is that nuclear systems are not readily accessible for inspections due to high level of radiation and large costs associated with remote data collection methods. The models of degradation used by industry are largely derived from ordinary linear regression methods. The main objective of this thesis is to develop more advanced techniques based on stochastic process theory to model deterioration in engineering components with the purpose of providing more scientific basis to life-cycle management of aging nuclear power plants. This thesis proposes a stochastic gamma process (GP) model for deterioration and develops a suite of statistical techniques for calibrating the model parameters. The gamma process is a versatile and mathematically tractable stochastic model for a wide variety of degradation phenomena, and another desirable property is its nonnegative, monotonically increasing sample paths. In the thesis, the GP model is extended by including additional covariates and also modeling for random effects. The optimization of age-based replacement and condition-based maintenance strategies is also presented. The thesis also investigates improved regression techniques for modeling deterioration. A linear mixed-effects (LME) regression model is presented to resolve an inconsistency of the traditional regression models. The proposed LME model assumes that the randomness in deterioration is decomposed into two parts: the unobserved heterogeneity of individual units and additive measurement errors. Another common way to model deterioration in civil engineering is to treat the rate of deterioration as a random variable. In the context of condition-based maintenance, the thesis shows that the random variable rate (RV) model is inadequate to incorporate temporal variability, because the deterioration along a specific sample path becomes deterministic. This distinction between the RV and GP models has profound implications to the optimization of maintenance strategies. The thesis presents detailed practical applications of the proposed models to feeder pipe systems and fuel channels in CANDU nuclear reactors. In summary, a careful consideration of the nature of uncertainties associated with deterioration is important for credible life-cycle management of engineering systems. If the deterioration process is affected by temporal uncertainty, it is important to model it as a stochastic process.