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Technical Assistance in Review of Source Term-Related Issues of Advanced Reactors

Description: The distribution of iodine in containment during an AP-600 design-basis accident was evaluated using models in the "TRENDS" code. The AP-6003BE accident sequmce calculations showed that a pH >7 was maintained for at least 30 days. Because the pH was maintained at this level, most of the iodine was in the form of iodid~ only 3 x 10-3% was present as aqueous 12, and only 1 x 10< `/0 was present as J in the vapor phase.
Date: October 1, 1998
Creator: Beahm, E.C.; Dillow, T.A. & Weber, C.F.
Partner: UNT Libraries Government Documents Department


Description: A transient, two-dimensional code has been developed to provide a detailed description of fuel-clad conditions during a TOP accident. Emphasis has been directed toward development of a framework within which fuel motion and ejection can be viewed following pin failure. All code modules have been rigorously verified. Illustrative application of the code, with the exercise of its many and varied features, have been included.
Date: August 1978
Creator: Meek, C. C.
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

Study on severe accident fuel dispersion behavior in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

Description: Core flow blockage events are a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel could result in core heatup and melting under full coolant flow condition. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Heat transfer between melt particle and coolant, which affects particle breakup, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to relative motion of particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. Results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also core debris tends to move together upon melting and entrainment.
Date: December 31, 1995
Creator: Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V. & Xiang, J.Y.
Partner: UNT Libraries Government Documents Department

Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

Description: The Solution High-Energy Burst Assembly (SHEBA) was originally constructed during 1980 and was designed to be a clean free-field geometry, right-circular, cylindrically symmetric critical assembly employing U(5%)O{sub 2}F{sub 2} solution as fuel. A second version of SHEBA, employing the same fuel but equipped with a fuel pump and shielding pit, was commissioned in 1993. This report includes data and operating experience for the 1993 SHEBA only. Solution-fueled benchmark work focused on the development of experimental measurements of the characterization of SHEBA; a summary of the results are given. A description of the system and the experimental results are given in some detail in the report. Experiments were designed to: (1) study the behavior of nuclear excursions in a low-enrichment solution, (2) evaluate accidental criticality alarm detectors for fuel-processing facilities, (3) provide radiation spectra and dose measurements to benchmark radiation transport calculations on a low-enrichment solution system similar to centrifuge enrichment plants, and (4) provide radiation fields to calibrate personnel dosimetry. 15 refs., 37 figs., 10 tabs.
Date: October 1, 1997
Creator: Cappiello, C.C.; Butterfield, K.B. & Sanchez, R.G.
Partner: UNT Libraries Government Documents Department

DOSFAC2 user`s guide

Description: This document describes the DOSFAC2 code, which is used for generating dose-to-source conversion factors for the MACCS2 code. DOSFAC2 is a revised and updated version of the DOSFAC code that was distributed with version 1.5.11 of the MACCS code. included are (1) an overview and background of DOSFAC2, (2) a summary of two new functional capabilities, and (3) a user`s guide. 20 refs., 5 tabs.
Date: December 1, 1997
Creator: Young, M.L. & Chanin, D.
Partner: UNT Libraries Government Documents Department

Developing safety culture-rocket science or common sense?

Description: Despite evidence of significant management contributions to the causes of major accidents, recent events at Millstone Nuclear Power Station in the US and Ontario Hydro in Canada might lead one to conclude that the significance of safety culture, and the role of management in developing and maintaining an appropriate safety culture, is either not being understood or not being taken serious as integral to the safe operation of some complex, high-reliability operations. It is the purpose of this paper to address four aspects of management that are particularly important to safety culture, and to illustrate how development of an appropriate safety culture is more a matter of common sense than rocket science.
Date: August 1, 1998
Creator: Mahn, J.A.
Partner: UNT Libraries Government Documents Department

Dissolution of Stainless Steel by Molten Aluminum and Aluminum Alloys - Final Report

Description: The purpose of this task was to investigate on a laboratory-scale the interactions of molten aluminum with stainless steel under hypothetical severe reactor accident conditions. This experimental effort provided data necessary to assess the susceptibility of the reactor vessel to breaching (general through-wall failure of vessel) in accident scenarios where contact of molten aluminum and stainless steel may occur. This report summarizes the results of the extensive experimental program.
Date: July 11, 2001
Creator: Marra, J.C.
Partner: UNT Libraries Government Documents Department

Precursors to potential severe core damage accidents: 1992, a status report; Volume 18: Appendices B, C, D, E, F, and G

Description: This document is part of a report which documents 1992 operational events selected as accident sequence precursors. This report describes the 27 precursors identified from the 1992 licensee event reports. It also describe containment-related events; {open_quote}interesting{close_quote} events; potentially significant events that were considered impractical to analyze; copies of the licensee event reports which were cited in the cases above; and comments from the licensee and NRC in response to the preliminary reports.
Date: December 1, 1993
Partner: UNT Libraries Government Documents Department

Characterization of jet breakup mechanisms observed from simulant of molten fuel penetrating coolant. Technical progress report, 1989--1990

Description: The objective of the proposed experiments is to replicate approximately, by injecting low melting point metal alloys into Freon-11 and liquid nitrogen, the dispersal of corium streams in water. To first gain a better understanding of the corium dispersal process to be simulated, experimental data from the CCM experiments, in which the injection of streams of molten corium into water was studied, was interpreted in cooperation with Argonne National Laboratory (ANL) staff. The results of these experiments are discussed briefly below. This is followed by a description of the preparations made to date for the present simulant experiments.
Date: December 31, 1990
Creator: Jones, B.G.
Partner: UNT Libraries Government Documents Department

Final Report of Fuel Dynamics Test E7

Description: Test data from an in-pile failure experiment of high-power LMFBR-type fuel pins in a simulated $3/s transient-overpower (TOP) accident are reported and analyzed. Major conclusions are that (1) a series of cladding ruptures during the 100-ms period preceding fuel release injected small bursts of fission gas into the flow stream; (2) gas release influenced subsequent cladding melting and fuel release (there were no measurable FCI's (fuel-coolant interactions), and all fuel motion observed by the hodoscope was very slow); (3) the predominant post-failure fuel motion appears to be radial swelling that left a spongy fuel crust on the holder wall; (4) less than 4 to 6 percent of the fuel moved axially out of the original fuel zone, and most of this froze within a 10-cm region above the original top of the fuel zone to form the outlet blockage. An inlet blockage approximately 1 cm long was formed and consisted of large interconnected void regions. Both blockages began just beyond the ends of the fuel pellets.
Date: April 1977
Creator: Doerner, R. C.; Murphy, W. F.; Stanford, G. S. & Froehle, P. H.
Partner: UNT Libraries Government Documents Department

Combined Motion of Fuel and Coolant Due to Fuel-Coolant Interactions under High Ramp Rate Reactivity Insertion

Description: An analysis has been made of the combined motion of fuel and coolant due to fuel-coolant interactions following a massive fuel failure in a high-ramp overpower transient. The motion of fuel and coolant was described using a two-fluid model formulation in which the mixture of sodium liquid and vapor and of fission gas, on the one hand, and the fuel particles, on the other, were treated as two superimposed continua. The method of solution employed a numerical procedure called the ACE method, a modified version of the IMF technique.
Date: July 1978
Creator: Chang, K. I. & Cho, D. H.
Partner: UNT Libraries Government Documents Department

Stationary low power reactor No. 1 (SL-1) accident site decontamination & dismantlement project

Description: The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the reactor pressure vessel surrounded by gravel shielding. Above the pressure vessel, in the center portion of the building, was a turbine generator and plant support equipment. The upper section of the building contained an air cooled condenser and its circulation fan. The major support facilities included a 2,500 ft{sup 2} two story, cinder block administrative building; two 4,000 ft{sup 2} single story, steel frame office buildings; a 850 ft{sup 2} steel framed, metal sided PL condenser building, and a 550 ft{sup 2} steel framed decontamination and laydown building.
Date: November 1, 1995
Creator: Perry, E.F.
Partner: UNT Libraries Government Documents Department

Systems analysis programs for hands-on integrated reliability evaluations (SAPHIRE), Version 5.0

Description: The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs), primarily for nuclear power plants. The Graphical Evaluation Module (GEM) is a special application tool designed for evaluation of operational occurrences using the Accident Sequence Precursor (ASP) program methods. GEM provides the capability for an analyst to quickly and easily perform conditional core damage probability (CCDP) calculations. The analyst can then use the CCDP calculations to determine if the occurrence of an initiating event or a condition adversely impacts safety. It uses models and data developed in the SAPHIRE specially for the ASP program. GEM requires more data than that normally provided in SAPHIRE and will not perform properly with other models or data bases. This is the first release of GEM and the developers of GEM welcome user comments and feedback that will generate ideas for improvements to future versions. GEM is designated as version 5.0 to track GEM codes along with the other SAPHIRE codes as the GEM relies on the same, shared database structure.
Date: October 1, 1995
Creator: Russell, K. D.; Kvarfordt, K. J. & Hoffman, C. L.
Partner: UNT Libraries Government Documents Department

Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants, Appendix 1

Description: From introduction: In conventional safety analyses, a suitable design basis, including redundancy, is specified to assure a minimum level of operability of ESFs, and the likelihood or consequences of total failure of ESFs are not considered further. In this study all failures are considered possible, but appropriate probabilities are assigned to them. Thus, many potential accident sequences are described in the following discussions as if they will surely occur, with no reservations expressed as to their likelihood or significance. However, most of these sequences have such low probability that they do not contribute to the overall risk from reactor accidents. In fact, in order to make an overall risk assessment, a major task of this study was to identify the sequences that are the dominant contributors to risk. In this study the initial failures or initiating events that could lead to significant consequences were examined to varying degrees. Those that seemed to contribute significantly to potential risks were analyzed in considerable detail; those that did not, received less detailed consideration. This is discussed more fully in section 3 of this appendix.
Date: October 1975
Creator: U.S. Nuclear Regulatory Commission
Partner: UNT Libraries Government Documents Department

Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants, Appendices 3 and 4

Description: From section 1: In the quantitative system probability estimates performed in this study, component behavior data in the form of failure rates and repair times are required as inputs to the system models. Since the goal of this study is risk assessment, as opposed to reliability analysis, larger errors (e.g. order of magnitude type accuracy) can be tolerated in the quantified results. This has important implications on the treatment of available data. In standard reliability analysis, point values (i.e., "best-estimates") are generally used for both data and results in quantifying the system model. In risk assessment, since results accurate to about an order of magnitude are sufficient, data and results using random variable and probabilistic approaches, can be usefully employed. The base of applicable failure rate data is thus significantly broadened since data with large error spreads and uncertainties can now be utilized. The data and associated material that were assembled for use in this study and that are presented here are to be used in the random variable framework (which will be described). The data and the accompanying framework are deemed sufficient for the study's needs. Care must be taken, however, since this data may not be sufficiently detailed, or accurate enough for use in general quantitative reliability models.
Date: October 1975
Creator: U.S. Nuclear Regulatory Commission
Partner: UNT Libraries Government Documents Department

Proceedings of the Third Post-Accident Heat Removal Information Exchange November 2-4, 1977, Argonne National Laboratory, Argonne, Illinois

Description: Papers presented at the third Post-Accident Heat Removal Information Exchange concerning heat distribution and criticality considerations, particulate-bed phenomena, pool heat transfer and melt-front phenomena, behavior of heated concrete and sodium-concrete interactions, design-related studies, gas bubbling and boiling effects, and materials interactions at high temperatures and experimental methods.
Date: 1978?
Creator: Baker, Louis, Jr. & Bingle, James D.
Partner: UNT Libraries Government Documents Department