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LOFT LOCE fuel module structural response

Description: The purpose of the LOFT fuel structural response evaluation program is to assess the accuracy of the structural response prediction codes and confirm that LOFT fuel bundle structural integrity will be maintained during the LOFT tests. A WHAM/SHOCK/SAP computer code combination has been developed to predict the LOFT fuel module structural response. LOFT test data have been evaluated, and additional computer code (RELAP/SCORE/MOXY) combinations are being developed for predicting guide tube temperatures during the future LOFT tests.
Date: January 1, 1978
Creator: Selcho, H.S.
Partner: UNT Libraries Government Documents Department

Development of a generalized correlation for phase-velocity measurements obtained from impedance-probe pairs in two-phase flow systems. [PWR]

Description: A flag type electrical impedance probe has been developed at the Oak Ridge National Lab (ORNL) to measure liquid- and vapor-phase velocities in steam-water mixtures flowing through rod bundles. Measurements are made by utilizing the probes in pairs, installed in line, parallel to the flow direction, and extending out into the flow channel. The present study addresses performance difficulties by examining from a fundamental point of view the two-phase flow system which the impedance probes typically operate in. Specifically, the governing equations (continuity, momentum, energy) were formulated for both air-water and steam-water systems, and then subjected to a scaling analysis. The scaling analysis yielded the appropriate dimensionless parameters of significance in both kinds of systems. Additionally, with the aid of experimental data obtained at ORNL, those parameters of significant magnitude were established. As a result, a generalized correlation was developed for liquid and vapor phase velocities that makes it possible to employ the impedance probe velocity measurement technique in a wide variety of test configurations and fluid combinations.
Date: January 1, 1983
Creator: Hsu, C.T.; Keshock, E.G. & McGill, R.N.
Partner: UNT Libraries Government Documents Department

Containment analysis capabilities of CONTEMPT4/MOD2. [PWR]

Description: The safety assessment and licensing of nuclear reactor plants by the United States Nuclear Regulatory Commission (USNRC) depend partially on analytical computer programs to predict the response of safeguard systems to accident conditions. CONTEMPT4/MOD2 is a new computer code to predict the long-term thermal hydraulic behavior of water-cooled nuclear reactor containment systems during postulated loss-of-coolant accident (LOCA) conditions. Written in FORTRAN IV, the new code was developed at the Idaho National Engineering Laboratory by EG and G Idaho, Inc., under the sponsorship of the USNRC. The paper describes the features and analytical models available in the code. Comparisons of calculated results with experimental data are also presented which demonstrate the range of containment problems applicable to CONTEMPT4/MOD2.
Date: January 1, 1978
Creator: Metcalfe, L.J.; Hargroves, D.W.; LaChance, J.L. & Wells, R.A.
Partner: UNT Libraries Government Documents Department

Effect of non-heterogeneous wetwell boundaries on pressure suppression system response. [BWR]

Description: The Full-Scale Mark II CRT (Containment Response Test) Program is in progress at the Tokai-Mura Establishment of the Japan Atomic Energy Research Institute (JAERI). The primary objective of the on-going CRT Program is to provide a data base for evaluation of the pressure suppression pool (wetwell) hydrodynamic loads associated with a postulated loss-of-coolant accident (LOCA) in the BWR Mark II containment system. The test facility is 1/18 of full scale in volume and has a wetwell which is a full-scale geometric replica of one 20/sup 0/-sector of a reference 1100MWe Mark II.
Date: August 29, 1980
Creator: McCauley, E.W.; Holman, G.S.; Namatame, K.; Kukita, Y. & Shiba, M.
Partner: UNT Libraries Government Documents Department

Equipment qualification issues research and resolution: Status report

Description: Since its inception in 1975, the Qualification Testing Evaluation (QTE) Program has produced numerous results pertinent to equipment qualification issues. Many have been incorporated into Regulatory Guides, Rules, and industry practices and standards. This report summarizes the numerous reports and findings to date. Thirty separate issues are discussed encompassing three generic areas: accident simulation methods, aging simulation methods, and special topics related to equipment qualification. Each issue-specific section contains (1) a brief description of the issue, (2) a summary of the applicable research effort, and (3) a summary of the findings to date.
Date: November 1, 1986
Creator: Bonzon, L.L.; Wyant, F.J.; Bustard, L.D. & Gillen, K.T.
Partner: UNT Libraries Government Documents Department

Debris dispersal in reactor material experiments on corium-water thermal interactions in ex-vessel geometry

Description: An analysis has been performed of corium sweepout behavior in the ANL/EPRI CWTI-series reactor material experiments involving the gas pressure-driven injection of molten corium into the reactor cavity region of a 1:30 scale mockup of a PWR containment. A computer model was developed to calculate the sweepout versus retention of corium and water from the cavity. The model consists of hydrodynamics and freezing calculations describing the pressure-driven two-phase flow of corium, water, steam and gas out of the cavity, freezing of corium upon structural surfaces, and levitation of corium within the cavity by the vessel blowdown gas jet. The model has had good success predicting the disposition of corium for the available CWTI tests, indicating retention in the cavity of between 40 and 70% of the injected corium masses. For conditions representative of the TMLB' sequence in the reactor system, the model predicts essentially complete sweepout of corium from the full-scale cavity region before the dispersive forces arising from the blowdown of the primary system have decayed. However, this large sweepout does not imply that the swept out material would deliver its energy directly to the containment atmosphere.
Date: January 1, 1984
Creator: Sienicki, J.J.; Spencer, B.W. & Squarer, D.
Partner: UNT Libraries Government Documents Department

Advanced Light Water Reactor Program: Program management and staff review methodology

Description: This report summarizes the NRC/EPRI coordinated effort to develop design requirements for a standardized advanced light water reactor (ALWR) and the procedures for screening and applying new generic safety issues to this program. The end-product will be an NRC-approved ALWR Requirements Document for use by the nuclear industry in generating designs of LWRs to be constructed for operation in the 1990s and beyond.
Date: December 1, 1986
Creator: Moran, D.H.
Partner: UNT Libraries Government Documents Department

Comparison of a TRAC calculation to the data from LSTF run SB-CL-05

Description: Run SB-CL-05 is a 5% break in the side of the cold leg. The test results show that the core was uncovered briefly and that the rods overheated at certain core locations. Liquid holdup on the upflow side of the steam generator tubes was observed. When the loop seal cleared, the core refilled and the rods cooled. The TRAC results are in reasonable agreement with the test data, meaning that TRAC correctly predicted the major trends and phenomena. TRAC predicted the core uncovery, the resulting rod heatup, and the liquid holdup on the upflow side of the steam generator tubes correctly. The clearing of the loop seal allowed core recovery and cooled the overheated rods just as it had in the data, but TRAC predicted its occurrence 20 s late. The experimental and TRAC analysis results of run SB-CL-05 are similar to those for Semiscale Run S-UT-8. In both runs there was core uncovery, rod overheating, and steam generator liquid holdup. These results confirm scaling of these phenomena from Semiscale (1/1650) to LSTF (1/48).
Date: January 1, 1986
Creator: Motley, F. & Schultz, R.
Partner: UNT Libraries Government Documents Department

Multivent effects in a large scale boiling water reactor pressure suppression system

Description: The steam-driven GKSS pressure suppression test facility, which contains 3 full scale vent pipes, has been used for 5 years to investigate the postulated loss-of-coolant accident in a Mark II and Type 69 boiling water reactor. Using the results from several of these tests, wetwell boundary load data (peak pressures and spectral power) during the chugging stage, have been evaluated for sparse pool response (one and two vents in the three vent pool) and for full pool response (one, two, or three vent operation in pools of constant wetwell pool area per vent). The sparse pool results indicate the pool-system, chug event boundary loads are strongly dependent on wetwell pool area per vent, with the load increasing with decreasing area. The full pool results show a substantial increase in the pool-system, chug event boundary loads upon a change from single cell to double cell operation; only minor change occurs in going from double to triple cell operation.
Date: July 6, 1984
Creator: McCauley, E.W.; Aust, E. & Schwan, H.
Partner: UNT Libraries Government Documents Department

SSI sensitivity studies and model improvements for the US NRC Seismic Safety Margins Research Program. Rev. 1

Description: The Seismic Safety Margins Research Program (SSMRP) is a US NRC-funded program conducted by Lawrence Livermore National Laboratory. Its goal is to develop a complete fully coupled analysis procedure for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. In Phase II of the SSMRP, the methodology was applied to the Zion nuclear power plant. Three topics in the SSI analysis of Zion were investigated and reported here - flexible foundation modeling, structure-to-structure interaction, and basemat uplift. The results of these investigations were incorporated in the SSMRP seismic risk analysis. 14 references, 51 figures, 13 tables.
Date: October 1, 1984
Creator: Johnson, J.J.; Maslenikov, O.R. & Benda, B.J.
Partner: UNT Libraries Government Documents Department

HDR (Heissdampfreaktor) Phase 2 vibrational experiments

Description: As part of the second phase of vibrational/earthquake investigations at the HDR (Heissdampfreaktor) Test Facility in Kahl/Main, FRG, high-level shaker tests (SHAG) were performed during June and July 1986. The purpose of these experiments is to investigate full-scale structural response, soil-structure interaction, and piping and equipment response under strong excitation conditions. While global safety considerations imposed load limitations, the HDR soil/structure system was nevertheless tested to its capacity limits. The performance of up to seven different multiple support pipe hanger configurations (ranging from flexible to stiff systems) was evaluated in the tests. Data obtained in the tests serve to validate analysis methods.
Date: October 1, 1986
Creator: Malcher, L. & Kot, C.A.
Partner: UNT Libraries Government Documents Department

Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

Description: Results of reactor-material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address ex-vessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debris characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity.
Date: January 1, 1984
Creator: Spencer, B.W.; McUmber, L.M.; Sienicki, J.J. & Squarer, D.
Partner: UNT Libraries Government Documents Department

Calculation of the neutron source distribution in the venus PWR engineering mock-up and comparison with experimental measurements

Description: The VENUS experiment is sponsored by the USNRC in conjunction with CEN/SCK in Mol, Belgium. The VENUS configuration consists of a central water hole, surrounded by a 2.888 cm thick inner sheet baffle. The inner core zone in the immediate vicinity of the inner baffle contains 752 3.3% /sup 235/U, zircalloy fuel cells, with 48 pyrex rods interspersed among them. The outer core zone contains 1800 4.0% /sup 235/U, steel clad fuel cells. The core itself is surrounded by a 2.858 cm thick outer steel baffle, a water reflector, a 4.972 cm thick steel core barrel, a water gap, a neutron pad, and the pressure vessel. The primary aim of this study is to calculate the VENUS neutron source distribution, as part of the USNRC's overall program goal of benchmarking RPV fluence calculations. Of particular concern is the accuracy of the source calculation near the core-baffle interface, which is the important region for contributing to RPV fluence. The accuracy of the calculations is evaluated by comparison with gamma scan measurements performed by CEN/SCK.
Date: January 1, 1984
Creator: Morakinyo, P.; Williams, M.L. & Kam, F.B.K.
Partner: UNT Libraries Government Documents Department

Adjoint sensitivity analysis of a thermal hydraulic system undergoing phase change due to boiling transition. [BWR]

Description: Sensitivity analysis of practical problems can be performed systematically and very efficiently by using adjoint functions. In areas of interest to nuclear reactors, this efficiency has been amply demonstrated on several widely used codes for neutronics and/or thermal hydraulic calculations. Applications of the adjoint method of sensitivity analysis to models involving phase transitions, where non-differentiability occurs, do not seem to have been reported to date. The purpose of this paper is to report results from a successful adjoint sensitivity analysis of a space- and time-dependent system where phase transition occurs due to boiling. The specific model chosen for this analysis is a simplified but representative model of a BWR pump-trip-type accident. This model is of particular importance to BWR safety, since pump failure is one of the most limiting hypothetical accidents in BWR's. This model simulates an exponential flow decay of initially subcooled FREON-114 flowing through a heated channel and undergoing boiling transition.
Date: June 3, 1984
Creator: Cacuci, D.G.; Wacholder, E.; Kaizerman, S. & Tomerian, N.
Partner: UNT Libraries Government Documents Department

Analysis of core damage frequency from internal events: Peach Bottom, Unit 2

Description: This document contains the internal event initiated accident sequence analyses for Peach Bottom, Unit 2; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 will document the risk of a selected group of nuclear power plants. As part of that work, this report contains the overall core damage frequency estimate for Peach Bottom, Unit 2, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provided additional insights regarding the dominant contributors to the Peach Bottom core damage frequency estimate. The mean core damage frequency at Peach Bottom was calculated to be 8.2E-6. Station blackout type accidents (loss of all ac power) were found to dominate the overall results. Anticipated Transient Without Scram accidents were also found to be non-negligible contributors. The numerical results are largely driven by common mode failure probability estimates and to some extent, human error. Because of significant data and analysis uncertainties in these two areas (important, for instance, to the most dominant scenario in this study), it is recommended that the results of the uncertainty and sensitivity analyses be considered before any actions are taken based on this analysis.
Date: October 1, 1986
Creator: Kolaczkowski, A.M.; Lambright, J.A.; Ferrell, W.L.; Cathey, N.G.; Najafi, B. & Harper, F.T.
Partner: UNT Libraries Government Documents Department

TMI-2 pressure transmitter examination and evaluation of CF-1-PT1, CF-2-LT1, and CF-2-LT2. Volume 2

Description: Pressure transmitters CF-1-PT1, CF-2-LT1, and CF-2-LT2 were removed from the Three Mile Island Unit 2 (TMI-2) Reactor Building and examined during FY-83. The purpose of the examination was to establish the operational characteristics and determine the failure mode of two of the three transmitters.
Date: April 1, 1984
Creator: Yancey, M.E. & Strahm, R.C.
Partner: UNT Libraries Government Documents Department

LOCA verification and data bank. [PWR]

Description: The purpose of this task was to derive local conditions heat transfer parameters and their uncertainties using computer codes and experimentally derived boundary conditions. To accomplish this objective, Semiscale S-02-9 blowdown experiment was used along with the INVERT (an inverse heat conduction code) and RELAP4 (a thermal hydraulic code) codes as the analytical tools. The uncertainties calculated for the local conditions were limited to those introduced by inaccuracies in the experimentally measured boundary conditions. The propagation of the measurement uncertainties through the codes was investigated by varying the code input using statistical methods and a response surface technique.
Date: January 1, 1979
Creator: Varacalle, Jr., D. J.; Cox, N. D.; Atwood, C. L.; Madden, S. C. & Condie, K. G.
Partner: UNT Libraries Government Documents Department

Logic flowgraph model for disturbance analysis of a PWR pressurizer system

Description: The Logic Flowgraph Methodology (LFM) has been developed as a synthetic simulation language for process reliability or disturbance analysis applications. A Disturbance Analysis System (DAS) using the LFM models can store the necessary information concerning a given process in an efficient way, and automatically construct in real time the diagnostic tree(s) showing the root cause(s) of occurring disturbances. A comprehensive LFM model for a PWR pressurizer system is presented and discussed, and the latest version of the LFM tree synthesis routine, optimized to achieve reduction of computer memory usage, is used to show the LFM diagnoses of selected hypothetic disturbances.
Date: January 1, 1984
Creator: Guarro, S. & Okrent, D.
Partner: UNT Libraries Government Documents Department

Design criteria and concepts for vented containment systems

Description: Accident sequences from WASH-1400 were selected and analyzed with the MARCH/CORRAL code to provide an envelope of design conditions. The time-dependent pressures and temperatures in containment were calculated as were the concentrations of steam, noncondensible gases, and airborne fission products in the containment atmosphere. The phenomenon found to be most challenging to containment integrity was a pressure spike resulting from rapid steam generation and/or hydrogen burning. The peak pressures in some sequences exceed the likely failure pressure. Conceptual designs were developed for preserving containment integrity. These include containment pressure relief or depressurization with various venting rates. Anticipatory venting, venting to the atmosphere, venting to a separate building, and venting followed by recirculation back into containment are considered. The effects of these schemes on the important system parameters were identified. The advantages and disadvantages of alternative schemes and their implications for the design of filtration equipment are discussed. For each venting strategy several levels of filtering effectiveness were considered. The simplest option developed is a once-through gravel-filled suppression pool. More sophisticated options involved sand filters, molecular sieves, charcoal adsorbers and HEPA filters. Results of accident consequence calculations using the CRAC code indicate the relatively simple options can provide substantial reductions in consequences of certain accident sequences. 12 figures.
Date: January 1, 1980
Creator: Walling, H.C.; Benjamin, A.S. & Cybulskis, P.
Partner: UNT Libraries Government Documents Department

RELAP5 progress summary: simulation of semiscale isothermal blowdown (Test S-01-4A). [RELAP 5/MOD O code]

Description: The RELAP5/MOD''O'' LOCA analysis code has been applied to Simulation of the Semiscale Isothermal Blowdown Test (S-01-4A) from initiation to 60 seconds. Subcooled ECC injection was simulated from 23 seconds until accumulator emptying. The calculated results are in very good agreement with the experimental data. This is the first full system application of the RELAP5 code and only the pressurizer surge line resistance was modified to achieve the results reported. An analysis of the code execution time using a time-step statistical edit is included.
Date: July 1, 1978
Creator: Kuo, H. H.; Wagner, R. J.; Carlson, K. E.; Kiser, D. M.; Trapp, J. A. & Ransom, V. H.
Partner: UNT Libraries Government Documents Department

Quick look report, Entry 3: Three Mile Island Unit 2, October 16, 1980

Description: This report summarizes tasks performed during the third entry at Three Mile Island Unit 2. During the entry into containment, which was made on October 16, 1980, additional beta, gamma, and neutron surveys were performed to supplement data obtained on previous entries. In addition, several maintenance tasks were completed including testing the operation of both equipment hatch doors, replacing a loose parts monitor system pre-amplifier, and removing a source range monitor. The five-man entry team accomplished these tasks in about 1-1/2 hours.
Date: July 1, 1981
Partner: UNT Libraries Government Documents Department

Quick look report, entry 4: Three Mile Island Unit 2, November 13, 1980

Description: This report summarizes tasks performed during entry 4 at Three Mile Island Unit 2. During the entry into containment, which was made on November 13, 1980, additional beta and gamma surveys were conducted to supplement data acquired on previous entries. A decontamination test was completed on Elevation 305. Power receptables tested on Elevation 305 were deenergized, but receptacles on Elevation 347 were energized. Still photography was acquired of Elevations 305 and 347. During the entry, 86 still photographs were taken. Videotaping (color and black and white) was done on Elevations 305 and 347, but lighting on both elevations was insufficient for high-quality video.
Date: June 1, 1981
Creator: Eidam, G.E.
Partner: UNT Libraries Government Documents Department

RELAP4 stagnation properties option. [PWR]

Description: The stagnation properties option in RELAP4/MOD6 was completely reviewed, from theoretical foundation to code application. The result of this investigation was the identification of a fundamental mismatch between the essentially homogeneous, equilibrium-based, RELAP4 code and the nonhomogeneous and/or nonequilibrium critical flow models imposed on the code. By continuously monitoring fluid Mach numbers and adjusting flow areas such that sonic velocity was never exceeded, the mismatch could be accommodated. This approach was implemented, found to work correctly, and will be incorporated into the MOD7 version of the code.
Date: January 1, 1979
Creator: DeYoung, T. L.
Partner: UNT Libraries Government Documents Department