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Seismic hazard evaluation for the high-flux isotope reactor (HFIR) Oak Ridge National Laboratory, Oak Ridge, Tennessee

Description: This study investigates the probabilistic hazard of earthquake-induced ground shaking at the HFIR facility, Oak Ridge, Tennessee. These results will be used to calculate plant response and potential effects in a Probabilistic Risk Assessment (PRA). For this purpose, several guidelines apply to this work. First, both the frequency of exceedance and the uncertainty in frequency of exceedance of various ground motion levels must be represented. These are required by the PRA so that the frequency a… more
Date: September 1, 1991
Creator: McGuire, R.K. & Toro, G.R. (Risk Engineering, Inc., Golden, CO (United States))
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Detection and diagnosis of abnormal transients in nuclear power plants

Description: This document describes a simulation-based algorithm that combines fuzzy logic with macroscopic conservation equations to diagnose multiple-failure events subject to uncertainties in transient data. Clusters of single-failure data points of similar characteristics are obtained through a pattern recognition algorithm and the cluster centers are combined in the space of macroscopic inventory derivatives to generate multiple-failure cluster centers. A fuzzy membership function is used to represent… more
Date: January 1, 1991
Creator: Lee, J. C.; Rank, P. J.; Hawkes, E.; Wehe, D. K. (Michigan Univ., Ann Arbor, MI (United States). Dept. of Nuclear Engineering) & Reifman, J. (Argonne National Lab., IL (United States))
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Aging assessment of BWR control rod drive systems

Description: This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to… more
Date: January 1, 1991
Creator: Greene, R.H.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests

Description: Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris … more
Date: August 1, 1991
Creator: Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W. (Sandia National Labs., Albuquerque, NM (United States)); Nichols, R.T. (Ktech Corp., Albuquerque, NM (United States)) & Sweet, D.W. (AEA Technology, Winfrith (United Kingdom))
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Transformer failure and common-mode loss of instrument power at Nine Mile Point Unit 2 on August 13, 1991

Description: On August 13, 1991, at Nine Mile Point Unit 2 nuclear power plant, located near Scriba, New York, on Lake Ontario, the main transformer experienced an internal failure that resulted in degraded voltage which caused the simultaneous loss of five uninterruptible power supplies, which in turn caused the loss of several nonsafety systems, including reactor control rod position indication, some reactor power and water indication, control room annunciators, the plant communications system, the plant … more
Date: October 1, 1991
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Identification and assessment of containment and release management strategies for a BWR Mark I containment

Description: This report identifies and assesses accident management strategies which could be important for preventing containment failure and/or mitigating the release of fission products during a severe accident in a BWR plant with a Mark 1 type of containment. Based on information available from probabilistic risk assessments and other existing severe accident research, and using simplified containment and release event trees, the report identifies the challenges a Mark 1 containment could face during t… more
Date: September 1, 1991
Creator: Lin, C. C. & Lehner, J. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Steam generator secondary pH during a steam generator tube rupture

Description: The Nuclear Regulatory Commission requires utilizes to determine the response of a pressurized water reactor to a steam generator tube rupture (SGTR) as part of the safety analysis for the plant. The SGTR analysis includes assumptions regarding the partitioning of iodine between liquid and vapor in steam generator secondary. Experimental studies have determined that the partitioning of iodine in water is very sensitive to the pH. Based on this experimental evidence, the NRC requested the INEL t… more
Date: December 1, 1991
Creator: Adams, J. P. & Peterson, E. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Historical overview of domestic spent fuel shipments: Update

Description: This report presents available historic data on most commercial and research reactor spent fuel shipments in the United States from 1964 through 1989. Data include sources of the spent fuel shipped, types of shipping casks used, number of fuel assemblies shipped, and number of shipments made. This report also addresses the shipment of spent research reactor fuel. These shipments have not been documented as well as commercial power reactor spent fuel shipment activity. Available data indicate th… more
Date: July 1, 1991
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Review of physics methodology of ATR safety analysis

Description: At the request of EG G Idaho, the Pacific Northwest Laboratory (PNL) performed a brief review of the physics methods employed in the safety analyses for the Advanced Test Reactor. PNL determined that the general approach used by EG G was sound. Comparisons were made between the EG G results and a simplified PBL model. These demonstrated good agreement. However, the lack of spacial treatment of the moderator density reactivity coefficient and exclusion of the test loops from the reactivity model… more
Date: September 1, 1991
Creator: Little, W.W. & Heaberlin, S.W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Two-component flow study in large-diameter horizontal pipe

Description: Westinghouse Savannah River Company, Idaho National Engineering Laboratory, and Wyle Laboratory cooperated in a series of single- and two-component calibration tests conducted to obtain sufficient information for calibrating flowmeters, to observe flow patterns, and to estimate void functions. Testing, conducted in large-diameter horizontal pipe, covered total flows of 0.19 to 1.89 m{sup 3}/s (3000 to 30000 gpm) and inlet void fractions up to 40%. A flow regime map, constructed using video imag… more
Date: December 3, 1991
Creator: Eghbali, D. A.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Core cooling under accident conditions at the high flux beam reactor (HFBR)

Description: In certain accident scenarios, e.g. loss of coolant accidents (LOCA) all forced flow cooling is lost. Decay heating causes a temperature increase in the core coolant and the resulting thermal buoyancy causes a reversal of the flow direction to a natural circulation mode. Although there was experimental evidence during the reactor design period (1958--1963) that the heat removal capacity in the fully developed natural circulation cooling mode was relatively high, it was not possible to make a co… more
Date: January 1, 1991
Creator: Tichler, P.; Cheng, L. (Brookhaven National Lab., Upton, NY (USA)) & Fauske, H. (Fauske and Associates, Inc., Burr Ridge, IL (USA))
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Comparison of eigenvalue computations for the Savannah River K Reactor using 5 and 7 digit dimensional and isotopic quantities

Description: A study was undertaken to characterize the reactivity temperature coefficient (RTC) behavior for the Savannah River K-Reactor pursuant to the safety review mandated by the Department of Energy (DOE) in August 1988. During the course of the investigation, it was found that the accuracy levels required in dimensional and isotopic quantities at elevated temperatures were much greater than was initially supposed and are typically used in reactor neutronics calculations. The codes involved do not au… more
Date: January 1, 1991
Creator: Durkee, J.W. Jr.; Mosteller, R.D.; Perry, R.T. & Sapir, J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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World nuclear fuel cycle requirements 1991

Description: The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacit… more
Date: October 10, 1991
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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The IFR modern nuclear fuel cycle

Description: Nuclear power is an essential component of the world's energy supply. The IFR program, by returning to fundamentals, offers a fresh approach to closing the nuclear fuel cycle. This closed fuel cycle represents the ultimate in efficient resource utilization and environmental accountability. 35 refs., 2 tabs.
Date: January 1, 1991
Creator: Hannum, W. H.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Accident Management to Prevent Containment Failure and Reduce Fission Product Release

Description: Brookhaven National Laboratory, under the auspices of the US Nuclear Regulatory Commission, is investigating accident management strategies which could help preserve containment integrity or minimize releases during a severe accident. The strategies considered make use of existing plant systems and equipment in innovative ways to reduce the likelihood of containment failure or to mitigate the release of fission products to the environment if failure cannot be prevented. Many of these strategies… more
Date: January 1, 1991
Creator: Lehner, J.R.; Lin, C.C.; Luckas, W.J. & Pratt, W.T.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Modeling operator actions during a small break loss-of-coolant accident in a Babcock and Wilcox nuclear power plant

Description: A small break loss-of-accident (SBLOCA) in a typical Babcock and Wilcox (B W) nuclear power plant was modeled using RELAP5/MOD3. This work was performed as part of the United States Regulatory Commission's (USNRC) Code, Scaling, Applicability and Uncertainty (CSAU) study. The break was initiated by severing one high pressure injection (HPI) line at the cold leg. Thus, the small break was further aggravated by reduced HPI flow. Comparisons between scoping runs with minimal operator action, and f… more
Date: January 1, 1991
Creator: Ghan, L.S. & Ortiz, M.G.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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An assessment of the risk significance of human errors in selected PSAs and operating events

Description: Sensitivity studies based on Probabilistic Safety Assessments (PSAs) for a pressurized water reactor and a boiling water reactor are described. In each case human errors modeled in the PSAs were categorized according to such factors as error type, location, timing, and plant personnel involved. Sensitivity studies were then conducted by varying the error rates in each category and evaluating the corresponding change in total core damage frequency and accident sequence frequency. Insights obtain… more
Date: January 1, 1991
Creator: Palla, R.L. Jr.; El-Bassioni, A. (Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Reactor Regulation) & Higgins, J. (Brookhaven National Lab., Upton, NY (United States))
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Are limit cycle calculations a stochastic process

Description: Stochasticity is typically associated with processes that produce uncertain results which, in many cases, are due to process nonlinearities and/or extreme sensitivity to initial conditions. By its name, a stochastic process should have a probabilistic or random nature; however, it is well known that many if not all, of the processes that behave stochasticly are indeed deterministic. This is the case with computer calculations to predict the stability of boiling water reactors (BWRs). This paper… more
Date: January 1, 1991
Creator: March-Leuba, J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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The limnology of L Lake: Results of the L-Lake monitoring program, 1986--1989

Description: L Lake was constructed in 1985 on the upper regions of Steel Creek, SRS to mitigate the heated effluents from L Reactor. In addition to the NPDES permit specifications (Outfall L-007) for the L-Reactor outfall, DOE-SR executed an agreement with the South Carolina Department of Health and Environmental Control (SCDHEC), that thermal effluents from L-Reactor will not substantially alter ecosystem components in the approximate lower half of L Lake. This region should be inhabited by Balanced (Indi… more
Date: December 15, 1991
Creator: Bowers, J.A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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On reactor type comparisons for the next generation of reactors

Description: In this paper, we present a broad comparison of studies for a selected set of parameters for different nuclear reactor types including the next generation. This serves as an overview of key parameters which provide a semi-quantitative decision basis for selecting nuclear strategies. Out of a number of advanced reactor designs of the LWR type, gas cooled type, and FBR type, currently on the drawing board, the Advanced Light Water Reactors (ALWR) seem to have some edge over other types of the nex… more
Date: August 22, 1991
Creator: Alesso, H.P. & Majumdar, K.C.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Uncertainties in TRAC plenum pressures for the FI phase of a DEGB LOCA

Description: The TRAC-PF1/MOD1 code (TRAC) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). For this accident, TRAC is used to analyze only the first 5 seconds following the DEGB, which encompasses the Flow Instability (FI) phase of the DBA. The TRAC analysis provides time-dependent … more
Date: May 1, 1991
Creator: Griggs, D. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Transactions of the nineteenth water reactor safety information meeting

Description: This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Res… more
Date: October 1, 1991
Creator: Weiss, A. J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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