A simple approximation for predicting the concrete erosion rate and depth is derived based on heat balance integral method for conduction with the time dependent boundary conditions. The problem is considered a four-region model including separate, moving heat sinks at the boundaries due to endothermic decomposition reactions. Polynomial temperature profiles are assumed and the results are compared to previous experimental data and other analytical solutions. Since the technique provides an approximate temperature distribution on the average, it does not give the real temperature evaluation but provides a simple prediction of the erosion rates in terms of the parameters that are important during the physical phenomena. Because of its simplicity and reliability, the model might be useful of the larger molten core/concrete interaction models.

The purpose of this paper is to generate a controller to raise the reactor neutronic power from a lower level to a higher level and stabilize it a that level. The dynamic response of research reactors was studied during and after the power transient process. An algorithm for the control process was generated, based on a previously published controller. The inverse-kinetics model-based controller was tested by simulation in order to verify successful completion of the control task, during and after the power transient process. A Kalman filter system was constructed to estimate the state parameters needed by the controller. System dynamic response under different conditions was obtained through simulations. The effects of control performance and system dynamic response of model mismatch between the reactor system and the filter system and of incorrectness of initializations of the filter system were studied. Selection of a filter system with an appropriate Kalman gain matrix was made after conducting several simulation tests. Comparison was made between experimental results obtained from the University of Arizona TRIGA research reactor and simulation results from the computerized simulation systems.

Solutions of fissile materials are often encountered during spent-fuel reprocessing. In order to estimate the hazards from accidental criticalities in these solutions, models have been developed to understand better the dynamics involved. Accurate representation of reactivity feedback mechanisms are a crucial part of such models. Reactivity feedback from uniform volumetric solution expansion is studied. For faster transients, density redistribution may also occur due to a variation of nuclear energy as a function of position in the assembly. Neutronic spectral temperature reactivity effects are studied using two methods: cross section corrections via a Maxwellian flux weighting and creation of temperature dependent cross sections from ENDF/B-VI data. The volumetric and temperature reactivity feedback coefficients are determined for the CRAC, KEWB-5, SILENE, and SHEBA solution assemblies. Spectral temperature coefficients are also calculated for poisoned, unpoisoned, and reflected plutonium solutions. Feedback coefficients are seen to be functions of geometry and isotopic contents of the assemblies. Results for plutonium solutions agree with British calculations reported in 1991, confirming the possibility of autocatalytic excursions in large, dilute solutions.

The disposal of solidified radioactive wastes requires physical, chemical, and radiological characterization to ensure safety. Two quantities important to physical characterization include compressive strength and leach resistance. It is desirable to monitor strength development on-line in processing whenever possible. Therefore, the ability of nondestructive evaluation techniques to predict the long term compressive strength of waste forms during their processing was theoretically and experimentally evaluated. It was determined that the compressive strength of the mixture could be predicted through analysis of both rheological behavior and maturity development. Ultrasonic methods were shown to be less effective. Leach testing requires destructive analysis. The ability of neutron activation analysis to increase the detection limit of leached cobalt from cement waste forms containing EDTA was experimentally examined. A detection limit approaching 53 parts per billion was found. EDTA concentration had a measurable effect on the cobalt release due to cobalt chelation and matrix degradation.

This thesis compares a new dynamic moving boundary thermal hydraulics fuel pin model (FUELPIN) to a known thermal hydraulics code (COBRA) for the following relationships: (1) Minimum Departure from Nucleate Boiling Ratio (MDNBR) versus time for transient conditions in a Pressurized Water Reactor (PWR) channel, and (2) Position of Minimum DNBR versus time for transient conditions. Relationships between MDNBR and position of MDNBR versus time were analyzed by comparing: (1) output of COBRA, a steady state thermal hydraulics model, with inputs from a PWR simulation code (CEPAC-L), (2) output of CEPAC-L linked FUELPIN. CEPAC-L computer model was shown to be a good simulation of a PWR during steady state conditions and a loss of all primary coolant flow casualty by comparison with an accepted PWR simulation code (CEPAC). FUELPIN was shown to predict larger thermal margins than COBRA for loss of all primary coolant flow transients in PWRs.

Collimator and shield configurations for two high-purity germanium detectors were designed for use during a loss of coolant accident at a boiling water reactor. The detectors will return information concerning stack releases to operators within a 15 minute time frame. Operating parameters for the stack monitors are defined by the United States Nuclear Regulatory Commission (USNRC) and a 24 hour source term generated by ORIGEN2. A lead collimator 0.4 cm in diameter, 20 cm in length for the high range detector and 2 cm in diameter, 20 cm in length for the low range detector was shown through a Monte Carlo code, MCNP4 to prevent high range detector saturation and provide enough low range detector response so good statistical data on stack releases result. A lead shield 20 cm thick was shown through MCNP4 to reduce the background radiation interference for both detectors to levels such that the detection of isotopes within the stack effluent is possible as required by the USNRC.

In this paper, the problem of designing an advanced sensor validation system (SVS) which is robust and fault-tolerant under faulty conditions is considered. Associative memories, which provide robust pattern recognition are investigated as an information processing technology that can be applied to sensor validation. Studies of Binary Associative Memories (BAM) and Continuous Associative Memories (CAM) yield many results including (1) the stability condition of exemplars and spurious memories in BAMs, (2) the formula of choosing diagonal weights and bias that eliminates spurious memories most effectively in BAMs, (3) the convergence theory of CAMs that have asymmetric weight matrix with non-zero diagonal elements and non-monotonically increasing activation functions, (4) the energy function that explores the convergence behavior of CAMs, and (5) the hybrid learning algorithm that reduces spurious memories effectively in CAMs. The concept of performability is introduced to the evaluation of SVS. A set of important performability variables is introduced. Stochastic Activity Networks are used as a modeling tool to evaluate the performability of SVS. An illustration example, the evaluation of the pressurizer SVS of a PWR, is provided.

A computer code that utilizes the consistent energy dependent P1 equations and the multigroup diffusion equations was developed to calculate the cold neutron flux that could be generated if a cold neutron source was built at the University of Arizona TRIGA reactor. The cold neutron current at the surface of the cold neutron source and the energy deposition in the cold neutron source due to neutron-proton interactions were also calculated. The computer code models the cold neutron source as a spherically shaped hydrogen medium embedded in the center of a cylindrical shaped homogenous TRIGA core. The outputs of the computer code were compared for various cold neutron source dimensions.

Thesis (Ph.D.)--University of Illinois at Urbana-Champaign, 2009. / Source: Dissertation Abstracts International, Volume: 70-06, Section: B, page: 3760. Adviser: Rizwan Uddin. Includes bibliographical references (leaves 194-204) Available on microfilm from Pro Quest Information and Learning.

Two "collision biasing" schemes have been tested in a major production, Monte Carlo neutron and photon transport code MCNP. The intent is to reduce the variance that is inherent in all Monte Carlo calculations and increase the Monte Carlo efficiency in problems where the phase space is not sampled adequately. The first scheme, called collision biasing, already exists in the French code TRIPOLI. This method samples several post-collision coordinates and randomly selects one in proportion to an input importance function. The second scheme, called track biasing, is original and differs from collision biasing in that several particle tracks (trajectories) are sampled from one distance-to-collision calculation to the next distance-to-collision calculation, and one is chosen randomly in proportion to an input importance function. The effect of track biasing was found to be problem dependent. For transport along a long narrow cylinder the Monte Carlo efficiency increased, however for a deep penetration problem it decreased.