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Safety rod latch inspection

Description: During an attempt to raise control rods from the 100 K reactor in December, one rod could not be withdrawn. Subsequent investigation revealed that a small button'' in the latch mechanism had broken off of the lock plunger'' and was wedged in a position that prevented rod withdrawal. Concern that this failure may have resulted from corrosion or some other metallurgical problem resulted in a request that SRL examine six typical latch mechanisms from the 100 L reactor by use of radiography and met… more
Date: February 1, 1992
Creator: Leader, D.R.
Partner: UNT Libraries Government Documents Department
open access

Rapid quenching of molten lithium-aluminum jets in water under loss-of-control-rod-cooling conditions

Description: A series of fifteen tests were performed to investigate the thermal interactions between molten LiAl control rod material and water under conditions prototypic of the loss-of-control-rod-cooling (LCRC) accident scenario. The experimental parameters such as melt mass, stream diameter, melt temperature and flowrate, water depth and water temperature were controlled or varied to agree with analytically determined conditions, thus insuring prototypicality of the experiments and applicability of the… more
Date: January 1, 1992
Creator: Greene, G. A.; Finfrock, C. C.; Schwarz, C. E.; Allison, D. K. & Hyder, M. L.
Partner: UNT Libraries Government Documents Department
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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))
Partner: UNT Libraries Government Documents Department
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Validation issues in aging risk evaluations

Description: Validation issues in aging risk evaluations are examined by considering how variabilities and uncertainties due to sparse component aging data, modeling assumptions, and risk quantification approaches may affect aging risk evaluation results and inferences. Sensitivity studies using a NUREG-1150 PWR evaluated the effect of component aging data uncertainties and variations in test and maintenance frequencies on aging prioritizations. Preliminary results indicate that while individual component r… more
Date: January 1, 1992
Creator: Hassan, M.; Samanta, P. (Brookhaven National Lab., Upton, NY (United States)) & Vesely, W. (Science Applications International Corp., San Diego, CA (United States))
Partner: UNT Libraries Government Documents Department
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Calculation of Savannah River K Reactor Mark-22 assembly LOCA/ECS power limits

Description: This paper summarizes the results of TRAC-PF1/MOD3 calculations of Mark-22 fuel assembly of loss-of-coolant accident/emergency cooling system (LOCA/ECS) power limits for the Savannah River Site (SRS) K Reactor. This effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to perform confirmatory power limits calculations for the SRS K Reactor. A method using a detailed three-dimensional (3D) TRAC model of the Mark-22 fuel assembly was d… more
Date: January 1, 1992
Creator: Fischer, S. R.; Farman, R. F. & Birdsell, S. A.
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))
Partner: UNT Libraries Government Documents Department
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Natural convection heat transfer analysis of ATR fuel elements

Description: Natural convection air cooling of the Advanced Test Reactor (ATR) fuel assemblies is analyzed to determine the level of decay heat that can be removed without exceeding the melting temperature of the fuel. The study was conducted to assist in the level 2 PRA analysis of a hypothetical ATR water canal draining accident. The heat transfer process is characterized by a very low Rayleigh number (Ra {approx} 10{sup {minus}5}) and a high temperature ratio. Since neither data nor analytical models wer… more
Date: May 1, 1992
Creator: Langerman, M.A.
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.
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))
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
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.
Partner: UNT Libraries Government Documents Department
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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.
Partner: UNT Libraries Government Documents Department
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Fatigue and environmentally assisted cracking in light water reactors

Description: Fatigue and stress corrosion cracking (SCC) for low-alloy steel used in piping and in steam generator and reactor pressure vessels have been investigated. Fatigue data were obtained on medium-sulfur-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor water, and in air. Analytical studies focused on the behavior of carbon steels in boiling water reactor (BWR) environments. Crack-growth rates of composite fracture-mechanics specime… more
Date: March 1, 1992
Creator: Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y. et al.
Partner: UNT Libraries Government Documents Department
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The development of code inservice inspection (ISI) requirements for Low Temperature Heavy Water Reactors (LTHWR)

Description: DOE Savannah River Field office requested that the American Society of Mechanical Engineers (ASME) develop rules for inservice inspection (ISI) of Savannah River Site (SRS) Low Temperature Heavy Water Reactors (LTHWR's) in January 1990. The request is part of the SRS Reactor Safety Improvement Program (RSIP). RSIP will implement an ASME B PV Code Section XI based ISI program after restart of K Reactor. The establishment of a Code based ISI program at SRS will affect a transition from a standing… more
Date: January 1, 1992
Creator: Cowfer, C.D.
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
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.
Partner: UNT Libraries Government Documents Department
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Savannah River Laboratory monthly report, February 1992

Description: This report is a progress report for the Savannah River Laboratory for the month of February 1992. The progress and activities in six categories were described in the report. The categories are reactor, tritium, separations, environmental, waste management, and general. Each category described numerous and varied activities. Some examples of these activities described are such things as radiation monitoring, maintenance, modifications, and remedial action.
Date: February 1, 1992
Creator: Ferrell, J.M. (comp.) & Ice, L.W. (ed.)
Partner: UNT Libraries Government Documents Department
open access

Rcs Pressure Under Reduced Inventory Conditions Following a Loss of Residual Heat Removal

Description: The thermal-hydraulic response of a closed-reactor coolant system to loss of residual heat removal (RHR) cooling is investigated. The processes examined include: core coolant boiling and steam generator reflux condensation, pressure increase on the primary side, heat transfer mechanisms on the steam generator primary and secondary sides, and effects of noncondensible gas on heat transfer processes.
Date: January 1, 1992
Creator: Palmrose, D. E.; Hughes, E. D. & Johnsen, G. W.
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.
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))
Partner: UNT Libraries Government Documents Department
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Correlation of analysis with high level vibration test results for primary coolant piping

Description: Dynamic tests on a modified 1/2.5-scale model of pressurized water reactor (PWR) primary coolant piping were performed using a large shaking table at Tadotsu, Japan. The High Level Vibration Test (HLVT) program was part of a cooperative study between the United States (Nuclear Regulatory Commission/Brookhaven National Laboratory, NRC/BNL) and Japan (Ministry of International Trade and Industry/Nuclear Power Engineering Center). During the test program, the excitation level of each test run was … more
Date: January 1, 1992
Creator: Park, Y.J.; Hofmayer, C.H. (Brookhaven National Lab., Upton, NY (United States)) & Costello, J.F. (Nuclear Regulatory Commission, Washington, DC (United States))
Partner: UNT Libraries Government Documents Department
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WSRC Reactor Tank Inspection Program (RTIP) status report

Description: Westinghouse Savannah River Company (WSRC) recently completed the initial phase of nondestructive inspections of the Savannah River Site's (SRS) reactor tanks. This program required almost three years to be conceptualized, fabricated, and tested. An additional 20 months were required to complete the NDE inspection of the P, K and L reactor tanks. The overall cost of the program to date is approximately $25 MM. This status report will address: (1) A brief review of the RTIP program and the const… more
Date: January 1, 1992
Creator: Loibl, M. W.
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.
Partner: UNT Libraries Government Documents Department
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