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NRC plan for cleanup operations at Three Mile Island Unit 2

Description: This NRC Plan, which defines NRC's functional role in cleanup operations at Three Mile Island Unit 2 and outlines NRC's regulatory responsibilities in fulfilling this role, is the first revision to the initial plan issued in July 1980 (NUREG-0698). Since 1980, a number of policy developments have occurred which will have an impact on the course of cleanup operations. This revision reflects these developments in the area of NRC's review and approval process with regard to cleanup … more
Date: February 1, 1982
Creator: Lo, R. & Snyder, B.
Partner: UNT Libraries Government Documents Department
open access

Characterization of solids in the Three Mile Island Unit 2 Reactor defueling water: Addendum

Description: Shortly after ORNL/TM-10362 was issued, it was discovered that a series of 31 figures had been inadvertently omitted. These figures, which consist of scanning electron microscope (SEM) photographs and energy-dispersive X-ray fluorescence scans, provide significant information about the results obtained in the tests performed with water sample W3. This Addendum includes these figures. Details of and comments on the SEM photographs may be found in ORNL/TM-10362.
Date: March 1, 1988
Creator: Campbell, D.O.
Partner: UNT Libraries Government Documents Department
open access

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR]

Description: The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-… more
Date: February 1, 1978
Creator: Benedetti, R. L.; Lords, L. V. & Kiser, D. M.
Partner: UNT Libraries Government Documents Department
open access

LOFT system structural response during subcooled blowdown

Description: The Loss-of-Fluid Test (LOFT) facility is a highly instrumented, pressurized water reactor test system designed to be representative of large pressurized water reactors (LPWRs) for the simulation of loss-of-coolant accidents (LOCAs). Detailed structural analysis and appropriate instrumentation (accelerometers and strain gages) on the LOFT system provided information for evaluation of the structural response of the LOFT facility for loss-of-coolant experiment (LOCE) induced loads. In general, th… more
Date: January 1, 1978
Creator: Martinell, J. S.
Partner: UNT Libraries Government Documents Department
open access

Nuclear plant-aging research on reactor protection systems

Description: This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show th… more
Date: January 1, 1988
Creator: Meyer, L.C.
Partner: UNT Libraries Government Documents Department
open access

PNL technical review of pressurized thermal-shock issues. [PWR]

Description: Pacific Northwest Laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS). Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be required to prevent an unsafe operating condition. PNL reviewed these responses and the r… more
Date: July 1, 1982
Creator: Pedersen, L. T.; Apley, W. J.; Bian, S. H.; Defferding, L. J.; Morgenstern, M. H.; Pelto, P. J. et al.
Partner: UNT Libraries Government Documents Department
open access

Quarterly Progress Report on Fission Product Behavior in LWRs for the Period October-December 1977

Description: Analysis of release data obtained during High Burnup Fuel Test 10 (HBU-10) has been completed. In this test the fuel rod segment was ruptured by internal pressurization at 900/sup 0/C, at which time the temperature was rapidly increased to 1200/sup 0/C and maintained at this temperature for 10 min. Approximately 0.061% of the total cesium inventory in the rod segment was released; this was accompanied by the release of about 1.69% of the total /sup 85/Kr inventory. Moreover, about 0.022% of the… more
Date: February 1, 1978
Creator: Malinauskas, A. P.; Lorenz, R. A.; Collins, J. L.; Osborne, M. F.; Whatley, S. K. & Towns, R. L.
Partner: UNT Libraries Government Documents Department
open access

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
open access

TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

Description: The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydr… more
Date: February 1, 1984
Creator: Liles, D.R. & Mahaffy, J.H.
Partner: UNT Libraries Government Documents Department
open access

RLPORINC: the RELAP4/MOD5 interface to ORINC

Description: The Oak Ridge Inverse code, ORINC, calculates the temperature distribution in an electric core pin from internal temperatures and the conditions in the surrounding environment. While the internal temperatures can be obtained from thermocouples in the Thermal Hydraulic Test Facility, the needed environmental conditions cannot be determined as directly from experimental data. However, this information can be determined from a modified version of the Reactor Linearized Analysis Program, RELAP. In … more
Date: January 10, 1978
Creator: Cliff, S.B.
Partner: UNT Libraries Government Documents Department
open access

Hydraulic pressure pulses with elastic and plastic structural flexibility: test and analysis (LWBR Development Program)

Description: Pressure pulse tests were conducted with a flexible test section in a test vessel filled with room temperature water. The pressure pulses were generated with a drop hammer and piston pulse generator and were of a sufficient magnitude to cause plastic deformation of the test section. Because of the strong pressure relief effect of the deforming test section, pressure pulse magnitudes were below 265 psig in magnitude and had durations of 50 to 55 msecs. Calculations performed with the FLASH-35 bi… more
Date: March 1978
Creator: Schwirian, R. E.
Partner: UNT Libraries Government Documents Department
open access

PWR Blowdown Heat Transfer Separate-Effects Program. Thermal-Hydraulic Test Facility experimental data report for test 103

Description: Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 103, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system.
Date: March 7, 1978
Creator: Clemons, V. D.; White, M. D.; Moore, P. A. & Hedrick, R. A.
Partner: UNT Libraries Government Documents Department
open access

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
open access

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
open access

Severe accident testing of electrical penetration assemblies

Description: This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. … more
Date: November 1, 1989
Creator: Clauss, D.B. (Sandia National Labs., Albuquerque, NM (USA))
Partner: UNT Libraries Government Documents Department
open access

Simmer analysis of prompt burst energetics experiments

Description: The Prompt Burst Energetics experiments are designed to measure the pressure behavior of fuel and coolant as working fluids during a hypothetical prompt burst disassembly in an LMFBR. The work presented in this report consists of a parametric study of PBE-5S, a fresh oxide fuel experiment, using SIMMER-II. The various pressure sources in the experiment are examined, and the dominant source identified as incondensable contaminant gasses in the fuel. The important modeling uncertainties and limit… more
Date: March 1, 1982
Creator: Hitchcock, J.T.
Partner: UNT Libraries Government Documents Department
open access

Experimental investigations of uncovered-bundle heat transfer and two-phase mixture-level swell under high-pressure low heat-flux conditions. [PWR]

Description: Results are reported from a series of uncovered-bundle heat transfer and mixture-level swell tests. Experimental testing was performed at Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF). The THTF is an electrically heated bundle test loop configured to produce conditions similar to those in a small-break loss-of-coolant accident. The objective of heat transfer testing was to acquire heat transfer coefficients and fluid conditions in a partially uncovered bundle. Test… more
Date: March 1, 1982
Creator: Anklam, T. M.; Miller, R. J. & White, M. D.
Partner: UNT Libraries Government Documents Department
open access

Seismic Safety Margins Research Program: Phase II program plan (FY 83-FY 84)

Description: The Seismic Safety Margins Research Program (SSMRP) is an NRC-funded, multiyear program conducted by Lawrence Livermore National Laboratory (LLNL). Its goal is to develop a complete, fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-caused radioactive release from a commercial nuclear power plant. The analysis procedure is based upon a state-of-the-art evaluation of the current seismic analysis and design process and explicitly incl… more
Date: August 2, 1982
Creator: Bohn, M. P.; Bernreuter, D. L.; Cover, L. E.; Johnson, J. J.; Shieh, L. C.; Shukla, S. N. et al.
Partner: UNT Libraries Government Documents Department
open access

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
open access

Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 9. PRAISE computer code user's manual. Load Combination Program Project I final report

Description: The PRAISE (Piping Reliability Analysis Including Seismic Events) computer code estimates the influence of earthquakes on the probability of failure at a weld joint in the primary coolant system of a pressurized water reactor. Failure, either a through-wall defect (leak) or a complete pipe severance (a large-LOCA), is assumed to be caused by fatigue crack growth of an as-fabricated interior surface circumferential defect. These defects are assumed to be two-dimensional and semi-elliptical in sh… more
Date: June 1, 1981
Creator: Lim, E.Y.
Partner: UNT Libraries Government Documents Department
open access

Posttest REALP4 analysis of LOFT experiment L1-3A

Description: This report presents selected results of posttest RELAP4 modeling of LOFT loss-of-coolant experiment L1-3A, a double-ended isothermal cold leg break with lower plenum emergency core coolant injection. Comparisons are presented between the pretest prediction, the posttest analysis, and the experimental data. It is concluded that pressurizer modeling is important for accurately predicting system behavior during the initial portion of saturated blowdown. Using measured initial conditions rather th… more
Date: October 1, 1977
Creator: White, J.R. & Holmstrom, H.L.O.
Partner: UNT Libraries Government Documents Department
open access

PWR blowdown heat transfer separate-effects program data evaluation report: system response for thermal-hydraulic test facility test series 100

Description: Selected reduced instrument responses and analyses of the indicated phenomena are presented for Thermal-Hydraulic Test Facility (THTF) test series 100, which is part of the PWR Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomena that govern the energy transfer and transport processes occurring during a postulated loss-of-coolant accident in a pressurized-water reactor system. Comparisons are made between the trends indi… more
Date: October 21, 1977
Creator: Hedrick, R. A.; Craddick, W. G.; Turnage, K. G. & Hyman, C. R.
Partner: UNT Libraries Government Documents Department
open access

BWR refill-reflood program: core spray distribution experimental task plan

Description: An experimental task plan for the BWR/4 core spray task of the Refill-Reflood Test Program is presented. The test program will provide core spray distribution data for a 30 degree sector of the BWR/4 and 5-218 design. This design uses different nozzle types and different sparger elevations than the BWR/6-218 design which was tested previously. Test parameter ranges are specified; individual tests are defined; and measurement and data utilization plans are defined.
Date: February 1, 1981
Creator: Eckert, T.
Partner: UNT Libraries Government Documents Department
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