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Multilevel transport solution of LWR reactor cores

Description: This work presents a multilevel approach for the solution of the transport equation in typical LWR assemblies and core configurations. It is based on the second-order, even-parity formulation of the transport equation, which is solved within the framework provided by the finite element-spherical harmonics code EVENT. The performance of the new solver has been compared with that of the standard conjugate gradient solver for diffusion and transport problems on structured and unstruc-tured grids. Numerical results demonstrate the potential of the multilevel scheme for realistic reactor calculations.
Date: September 1, 2008
Creator: Damian, Jose Ignacio Marquez; Oliveira, Cassiano R.E. de & Park, HyeonKae
Partner: UNT Libraries Government Documents Department

Uncertainties in the analysis of plutonium fueled light water moderated assemblies

Description: A theoretical analysis of UO/sub 2/-- PuO/sub 2/ fueled, light-water- moderated lattice experiments has been performed to aid in establishing technical bases and design criteria for the utilization of plutonium bearing fuel in thermal power reactors. Results for UO/sub 2/ and Al-- Pu lattices are included in order to understand the effects due to uranium and plutonium separately. The problems involved in calculating high leakage critical experiments are discussed. Estimates of the effects of various approximations inherent in the theoretical methods and/or analysis procedures are included along with the consequence on the results of the correlation. Uncertainties in the measurements and the neutron crosssection data are related to uncertainties in the calculated values K/sub eff/ .The results of other studies which bear on evaluating the calculational methods are summarized. Areas which should be investigated in future analyses are also identified. (111 references) (auth)
Date: May 1, 1973
Creator: Liikala, R.C.; Uotinen, V.O. & Jenquin, U.P.
Partner: UNT Libraries Government Documents Department

Results in the application of pattern recognition methods to nuclear reactor core component surveillance

Description: From nuclear science symposium; San Francisco, California, USA (14 Nov 1973). Pattern recognition methods were applied to analyze and interpret neutron noise data from the High Flux Isotope Reactor (HFIR) at ORNL. The results show that it is feasible to detect some core component failures by means of machine- discernible differences in the time-dependent noise power spectra. These neutron spectra (signatures) were analyzed by using a clusterseeking algorithm to derive a set of templates for automatic computer evaluation of the reactor's mechanical integrity and soundness. (auth)
Date: January 1, 1973
Creator: Gonzalez, R.C.; Fry, D.N. & Kryter, R.C.
Partner: UNT Libraries Government Documents Department

An Innovative Three-Dimensional Heterogeneous Coarse-Mesh Transport Method for Advanced and Generation IV Reactor Core Analysis and Design

Description: This project has resulted in a highly efficient method that has been shown to provide accurate solutions to a variety of 2D and 3D reactor problems. The goal of this project was to develop (1) an accurate and efficient three-dimensional whole-core neutronics method with the following features: based sollely on transport theory, does not require the use of cross-section homogenization, contains a highly accurate and self-consistent global flux reconstruction procedure, and is applicable to large, heterogeneous reactor models, and to (2) create new numerical benchmark problems for code cross-comparison.
Date: November 12, 2009
Creator: Rahnema, Farzad
Partner: UNT Libraries Government Documents Department

Development of An On-Line, Core Power Distribution Monitoring System

Description: The objective of the proposed work was to develop a software package that can construct in three-dimensional core power distributions using the signals from constant temperature power sensors distributed in the reactor core. The software developed uses a mode-based state/parameter estmation technique that is particularly attractive when there are model uncertainties and/or large signal noise. The software yields the expected value of local power at the detector locations and points in between, as well as the probability distribution of the local power density
Date: October 2, 2007
Creator: ALdemir, Tunc; Miller, Don & Wang, Peng
Partner: UNT Libraries Government Documents Department

Localized Deformation as a Primary Cause of Irradiation Assisted Stress Corrosion Cracking

Description: The objective of this project is to determine whether deformation mode is a primary factor in the mechanism of irradiation assisted intergranular stress corrosion cracking of austenitic alloys in light watert reactor core components. Deformation mode will be controlled by both the stacking fault energy of the alloy and the degree of irradiation. In order to establish that localized deformation is a major factor in IASCC, the stacking fault energies of the alloys selected for study must be measured. Second, it is completely unknown how dose and SFE trade-off in terms of promoting localized deformation. Finally, it must be established that it is the localized deformation, and not some other factor that drives IASCC.
Date: March 31, 2009
Creator: Was, Gary S.
Partner: UNT Libraries Government Documents Department

High-Order Homogenization Method in Diffusion Theory for Reactor Core Simulation and Design Calculation

Description: Most modern nodal methods in use by the reactor vendors and utilities are based on the generalized equivalence theory (GET) that uses homogenized cross sections and flux discontinuity factors. These homogenized parameters, referred to as infinite medium parameters, are precomputed by performing single bundle fine-mesh calculations with zero current boundary conditions. It is known that for configurations in which the node-to-node leakage (e.g., surface current-to-flux ratio) is large the use of the infinite medium parameters could lead to large errors in the nodal solution. This would be the case for highly heterogeneous core configurations, typical of modern reactor core designs.
Date: September 30, 2003
Creator: Rahnema, Farzad
Partner: UNT Libraries Government Documents Department

A summary of generation IV non-classical power reactor concepts.

Description: The summary of this report are: (1) Despite many of the technology gaps and data uncertainties, there is no lack of innovation and revolutionary ideas in Non-Classical reactor concepts. (2) Several concepts such as gas/vapor core reactors meet or exceed Gen IV goals for sustainability, safety, and economy, and have potential for making significant inroads toward achieving the optimum utilization of nuclear energy. (3) Gas/vapor core reactors set the upper performance potential in sustainability and safety with no insurmountable technology challenge. (4) Evaluation of modular deployable concepts are underway. (5) Direct energy conversion and non-convective cooled nuclear reactor systems are eliminated from further evaluation process.
Date: August 1, 2001
Creator: Lewis, D.
Partner: UNT Libraries Government Documents Department

LER screening algorithm for identification of potential accident sequence precursor events

Description: A computer algorithm has been developed and implemented to search the Sequence Coding and Search System Licensee Event (LER) database for failures or conditions common to Accident Sequence Precursor (ASP) events. Use of the algorithm has greatly improved the efficiency and timeliness in identifying potential ASP events and, by focusing attention on the most likely precursor events, has reduced the likelihood that these events will be overlooked in manual screening.
Date: September 1996
Creator: Poore, W. P., III
Partner: UNT Libraries Government Documents Department

FFTF (Fast Flux Test Facility) Reactor Characterization Program: Absolute Fission-rate Measurements

Description: Absolute fission rate measurements using modified National Bureau of Standards fission chambers were performed in the Fast Flux Test Facility at two core locations for isotopic deposits of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu. Monitor chamber results at a third location were analyzed to support other experiments involving passive dosimeter fission rate determinations.
Date: May 1, 1981
Creator: Fuller, J.L.; Gilliam, D.M.; Grundl, J.A.; Rawlins, J.A. & Daughtry, J.W.
Partner: UNT Libraries Government Documents Department

Development of the BWR Dry Core Initial and Boundary Conditions for the SNL XR2 Experiments

Description: The objectives of the Boiling Water Reactor Experimental Analysis and Model Development for Severe Accidents (BEAMD) Program at the Oak Ridge National Laboratory (ORNL) are: (1) the development of a sound quantitative understanding of boiling water reactor (BWR) core melt progression; this includes control blade and channel box effects, metallic melt relocation and possible blockage formation under severe accident conditions, and (2) provision of BWR melt progression modeling capabilities in SCDAP/RELAP5 (consistent with the BWR experimental data base). This requires the assessment of current modeling of BWR core melt progression against the expanding BWR data base. Emphasis is placed upon data from the BWR tests in the German CORA test facility and from the ex-reactor experiments [Sandia National Laboratories (SNL)] on metallic melt relocation and blockage formation in BWRs, as well as upon in-reactor data from the Annular Core Research Reactor (ACRR) DF-4 BWR test (conducted in 1986 at SNL). The BEAMD Program is a derivative of the BWR Severe Accident Technology Programs at ORNL. The ORNL BWR programs have studied postulated severe accidents in BWRs and have developed a set of models specific to boiling water reactor response under severe accident conditions. These models, in an experiment-specific format, have been successfully applied to both pretest and posttest analyses of the DF-4 experiment, and the BWR severe fuel damage (SFD) experiments performed in the CORA facility at the Kernforschungszentrum Karlsruhe (KfK) in Germany, resulting in excellent agreement between model prediction and experiment. The ORNL BWR models have provided for more precise predictions of the conditions in the BWR experiments than were previously available. This has provided a basis for more accurate interpretation of the phenomena for which the experiments are performed. The experiment-specific models, as used in the ORNL DF-4 and CORA BWR experimental analyses, also provide a basis for ...
Date: January 1, 1994
Creator: Ott, L. J.
Partner: UNT Libraries Government Documents Department