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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

Experiment data report for semiscale Mod-1 test S-28-4 (steam generator tube rupture test)

Description: Recorded test data are presented for Test S-28-4 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-4 was conducted from initial conditions of 15 646 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Thirty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg.
Date: October 1, 1977
Creator: Esparza, V. & Sackett, K.E.
Partner: UNT Libraries Government Documents Department

Experiment data report for semiscale Mod-1 test S-06-4 (LOFT counterpart test)

Description: Recorded test data are presented for Test S-06-4 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-4 was conducted from initial conditions of 15,653 kPa and 564 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 100 percent of the maximum peak power density (52.5 kW/m).
Date: December 1, 1977
Creator: Gillins, R. L.; Sackett, K. E. & Coppin, C. E.
Partner: UNT Libraries Government Documents Department

Assessment of soil-structure interaction effects based on simple modes. [PWR; BWR]

Description: Soil-structure interaction effects are investigated using a simple mathematical model which employs three degrees-of-freedom. The foundation is approximated by a homogeneous, isotropic, elastic half-space. Harmonic functions and a recorded earthquake are used to represent the free-field input motion. Variations of the response characteristics due to structural and interaction parameters are demonstrated. Response spectra are presented that display the magnitude of the maximum structural response for a range of fixed-base structural frequencies, interaction frequencies and damping. Conclusions are obtained regarding the behavior of the response of the soil-structure system. The findings reported herein can be used for the interpretation of the results of soil-structure interaction analyses of nuclear plant structures that are performed with available computer codes.
Date: January 1, 1983
Creator: Philippacopoulos, A.J. & Miller, C.A.
Partner: UNT Libraries Government Documents Department

Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor building decontamination. Summary status report. Volume 2

Description: This document summarizes information relating to decontamination of the Three Mile Island Unit 2 (TMI-2) reactor building. The report covers activities for the period of June 1, 1979 through March 29, 1985. The data collected from activity reports, reactor containment entry records, and other sources were entered into a computerized data system which permits extraction/manipulation of specific information which can be used in planning for recovery from an accident similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during decontamination of the reactor building. Support activities conducted outside of radiation areas are excluded from the scope of this report. Computerized reports included in this document are: a chronological summary listing work performed relating to reactor building decontamination for the period specified; and summary reports for each major task during the period. Each task summary is listed in chronological order for zone entry and subtotaled for the number of personnel entries, exposures, and man-hours. Manually-assembled table summaries are included for: labor and exposures by department and labor and exposures by major activity.
Date: May 1, 1986
Creator: Doerge, D.H.; Miller, R.L. & Scotti, K.S.
Partner: UNT Libraries Government Documents Department

Improving the safety of LWR power plants. Final report

Description: This report documents the results of the Study to identify current, potential research issues and efforts for improving the safety of Light Water Reactor (LWR) power plants. This final report describes the work accomplished, the results obtained, the problem areas, and the recommended solutions. Specifically, for each of the issues identified in this report for improving the safety of LWR power plants, a description is provided in detail of the safety significance, the current status (including information sources, status of technical knowledge, problem solution and current activities), and the suggestions for further research and development. Further, the issues are ranked for action into high, medium, and low priority with respect to primarily (a) improved safety (e.g. potential reduction in public risk and occupational exposure), and secondly (b) reduction in safety-related costs (improving or maintaining level of safety with simpler systems or in a more cost-effective manner).
Date: April 1, 1980
Partner: UNT Libraries Government Documents Department

TMI-2 Lessons Learned Task Force. Final report

Description: In its final report reviewing the Three Mile Island accident, the TMI-2 Lessons Learned Task Force has suggested change in several fundamental aspects of basic safety policy for nuclear power plants. Changes in nuclear power plant design and operations and in the regulatory process are discussed in terms of general goals. The appendix sets forth specific recommendations for reaching these goals.
Date: October 1, 1979
Partner: UNT Libraries Government Documents Department

Blowdown heat transfer separate-effects program. Quarterly progress report, January-March 1980

Description: Six additional bundle uncovery/recovery tests were performed in the Thermal-Hydraulic Test Facility during January, bringing the total number of these tests to eight. Data taken during the tests were found to be contaminated by numerous spurious spikes. Work to remove the spurious spikes is under way. Posttest analysis of the tests is approx.20% completed. The recovery portion of one of the tests will be analyzed by COBRA/TRAC, currently being developed by Pacific Northwest Laboratories (PNL). Work to debug the code for this application is in progress at PNL. The uncovery/recovery tests apparently caused damage to the 0-rings that form part of the loop pressure boundary. Refurbishment of the 0-ring seal system is being performed concurrently with scheduled loop modifications that include installation of ten in-bundle differential pressure instruments. Design, procurement, and fabrication of the in-bundle gamma densitometer system are continuing on schedule.
Date: June 1, 1980
Creator: Craddick, W.G.; Anklam, T.M.; Bohanan, R.E.; Felde, D.K.; Flanders, R.M.; Hagar, R.C. et al.
Partner: UNT Libraries Government Documents Department

Analysis of steam-line-break transient with RELAP5/MOD1. 5 and comparison to CESSAR. [PWR]

Description: Section 15.1.5 of the Standard Review Plant outlines the areas of review and the acceptance criteria for the analysis of steam system piping failures inside and outside of containment. Combustion Engineering (C-E) submitted analysis of the limiting steam line break (SLB) transient for the C-E 3800 MW nuclear steam supply system as Appendix 15C to the CESSAR FSAR. Conservative initial conditions and analysis assumptions were utilized in the transient calculation which was performed with the CESEC-III computer program. Argonne National Laboratory (ANL), under contract to the Nuclear Regulatory Commission, performed audit calculations of SLB transients presented in the CESSAR FSAR, with RELAP5/MOD1.5 (Cycle 26) and compared them with the CE results. The transient presented here is a large steam line break during full power operation with a concurrent loss of offsite power. The objective of this study was to examine the impact of mixing of cold primary fluid returning to the core from the affected steam generator with the hot fluid returning from the intact side, and its effect upon the potential return to power due to moderator feedback.
Date: January 1, 1983
Creator: Peeler, G.B.; Caraher, D.L. & Guttmann, J.
Partner: UNT Libraries Government Documents Department

Analysis of the ANO-2 turbine trip test

Description: The start-up tests performed with the Arkansas Nuclear One-Unit Two (ANO-2) plant provided an opportunity for studying the validity of certain integral systems codes. In particular, the turbine trip from 98.2 percent full power test was investigated with the RELAP5/MOD1 (cycle 18) ode. A detailed plant model was developed and used to understand the test reports. The early depressurization portion of the transient was reproduced; however, the resultant repressurization was not well represented due to uncertainty in the data and plant response. As a result of these computations and detailed analyses of the test data considerable insight was drawn as to the best way to perform and gather data from such integral systems tests for use in code verification studies.
Date: January 1, 1983
Creator: McDonald, T.A.; Tessier, J.H.; Senda, Y. & Waterman, M.D.
Partner: UNT Libraries Government Documents Department

Detailed analysis of the ANO-2 turbine trip test

Description: A RELAP5/MOD1 (Cycle 18) computer code simulation of the ANO-2 turbine trip test from 98% power level was performed for use in vendor code qualification studies. Results focused on potential improvements to simulation capabilities and plant data acquisition systems to provide meaningful comparisons between the calculations and the test data. The turbine trip test was selected because it resulted in an unplanned sequence of events that broadly affected the plant process systems and their controls. The pressurizer spray valve stuck open at an undetermined flow area, and an atmospheric dump valve remained stuck fully open while several atmospheric dump and secondary side safety valves were unavailable throughout. Thus, although the plant remained always in a safe condition, this transient potentially provided an unusual set of data against which the fidelity of a NSSS simulation by RELAP5/MOD1 along with certain vendor analysis codes might be judged.
Date: January 1, 1983
Creator: McDonald, T.A.; Tessier, J.H.; Senda, Y. & Waterman, M.D.
Partner: UNT Libraries Government Documents Department

Direct heating containment vessel interactions code (DHCVIC) and prediction of SNL ''SURTSEY'' test DCH-1

Description: High-pressure melt ejection from PWR vessels has been identified as a severe core accident scenario which could potentially lead to ''early'' containment failure. Melt ejection, followed by dispersal of the melt by high velocity steam in the cavity beneath the PWR vessel could, according to this scenario, lead to rapid transfer of energy from the melt droplets to the containment atmosphere. This paper describes DHCVIC, an integrated model of the thermal, chemical and hydrodynamic interactions which are postulated to take place during high-pressure melt ejection sequences. The model, which characterizes vessel (or building), is applied to prediction of the Sandia National Laboratory ''SURTSEY'' Test DCH-1 and a (post-test) prediction of that test is made.
Date: January 1, 1986
Creator: Ginsberg, T. & Tutu, N.
Partner: UNT Libraries Government Documents Department

Draft report: a selection methodology for LWR safety R and D programs and proposals

Description: The results of work done to develop a methodology for selecting LWR safety R and D programs and proposals is described. A critical survey of relevant decision analysis methods is provided including the specifics of multiattribute utility theory. This latter method forms the basis of the developed selection methodology. Details of the methodology and its use are provided along with a sample illustration of its application.
Date: March 1, 1980
Creator: Husseiny, A. A. & Ritzman, R. L.
Partner: UNT Libraries Government Documents Department

Measures of risk importance and their applications. [PWR; BWR]

Description: This work is part of a project being conducted for the Division of Risk Analysis (DRA) of the Nuclear Regulatory Commission (NRC). The objectives of the project are to evaluate the importances of containment, the different safety functions, and other various contributers as assessed in probabilistic risk analyses and to identify generic conclusions regarding the importances. Effective display of the importances is an important part of these objectives. To address these objectives, measures of risk importance need to be first identified and then they need to be evaluated for the different risk analyses which have been performed. This report describes the risk importance measures that were defined and were applied to the risk analyses which were performed as part of the Reactor Safety Study Methodology Applications Program (RSSMAP). The risk importance measures defined in this report measure the importance of features not only with regard to risk reduction but also with regard to reliability assurance, or risk maintenance. The goal of this report is not to identify new mathematical formulas for risk importance but to show how importance measures can be interpreted and can be applied.
Date: July 1, 1983
Creator: Vesely, W.E.; Davis, T.C.; Denning, R.S. & Saltos, N.
Partner: UNT Libraries Government Documents Department

An analysis of loss of offsite power with a PWR at shutdown

Description: In many Probabilistic Risk Assessments (PRAs), loss of offsite power (LOOP) when a nuclear power plant is operating was found to be a significant contributor to core damage. The purpose of this study is to provide an analysis of a loss of offsite power event that occurs while a PWR is shut down. The importance of such an analysis was recognized as part of a study to evaluate the core damage frequency due to a loss of decay heat removal capability during an outage. 5 refs., 1 fig.
Date: June 1, 1987
Creator: Chu, T.L.; Yoon, W.H. & Fitzpatrick, R.G.
Partner: UNT Libraries Government Documents Department

Analysis of MSIV-ATWS events with the BNL plant analyzer

Description: There are automatic safety features and operator-initiated emergency procedures which influence the sequence of events until the time when the standby liquid control system (SLCS), or other attempts to get control rods inserted, can effect shutdown of the core. One emergency procedure for a BWR/4 would require the operator to reduce the flow of high pressure coolant injection (HPCI) into the reactor. The core inlet flow rate at this time would be due to natural circulation and the reduced flow would lower the water level in the downcomer thereby reducing the natural circulation flow rate. This effect, and the reduction in core inlet subcooling due to mixing of the emergency feedwater with steam in the downcomer when the level was lowered, cause a sufficient increase in core void fraction so that the power would be reduced. A reduction in pressure might also be called for during this event in order to comply with the PSP heat capacity temperature limit (or possibly to prevent cycling of relief valves). In the past few years there have been several studies of this problem with the emphasis on calculating the power level in the core. In the present study we consider the power level as well as the resulting PSP temperature and take into account different assumptions regarding plant parameters and operator actions.
Date: June 1, 1986
Creator: Diamond, D.J.
Partner: UNT Libraries Government Documents Department

Determining a lateral load specification for downcomers during chugging in a Mark II containment. [BWR]

Description: During a postulated LOCA in a BWR containment, steam is vented to the suppression pool and condensation-driven pressure oscillations may occur. In a Mark II type of containment with vertical downcomers these pressure oscillations may give rise to lateral loads near the downcomer exit which are impulsive in nature and random in direction. Domestic BWR vendors found that a dynamic load having a half sinusoidal shape with maximum amplitude F and duration tau, when applied to an analytic model of a Mark II downcomer, reproduced well the structural response observed in domestic and foreign lateral load tests. Because of the relative time scales involved, this could be thought of as an impulse specification with an impulse of amplitude F. Originally, the highest load observed in a domestic test conducted for the Mark II Owners had been suggested as an appropriate specification. At the request of NRC, Brookhaven National Laboratory and its consultants evaluated the results of several foreign and domestic lateral load tests in order to recommend an appropriate load amplitude.
Date: January 1, 1983
Creator: Lehner, J.R. & Sonin, A.A.
Partner: UNT Libraries Government Documents Department

Generating human reliability estimates using expert judgment. Volume 2. Appendices. [PWR; BWR]

Description: The US Nuclear Regulatory Commission is conducting a research program to determine the practicality, acceptability, and usefulness of several different methods for obtaining human reliability data and estimates that can be used in nuclear power plant probabilistic risk assessments (PRA). One method, investigated as part of this overall research program, uses expert judgment to generate human error probability (HEP) estimates and associated uncertainty bounds. The project described in this document evaluated two techniques for using expert judgment: paired comparisons and direct numerical estimation. Volume 2 provides detailed procedures for using the techniques, detailed descriptions of the analyses performed to evaluate the techniques, and HEP estimates generated as part of this project. The results of the evaluation indicate that techniques using expert judgment should be given strong consideration for use in developing HEP estimates. Judgments were shown to be consistent and to provide HEP estimates with a good degree of convergent validity. Of the two techniques tested, direct numerical estimation appears to be preferable in terms of ease of application and quality of results.
Date: November 1, 1984
Creator: Comer, M.K.; Seaver, D.A.; Stillwell, W.G. & Gaddy, C.D.
Partner: UNT Libraries Government Documents Department

Human factors review for nuclear power plant severe accident sequence analysis

Description: The paper discusses work conducted to: (1) support the severe accident sequence analysis of a nuclear power plant transient based on an assessment of operator actions, and (2) develop a descriptive model of operator severe accident management. Operator actions during the transient are assessed using qualitative and quantitative methods. A function-oriented accident management model provides a structure for developing technical operator guidance on mitigating core damage preventing radiological release.
Date: January 1, 1985
Creator: Krois, P.A. & Haas, P.M.
Partner: UNT Libraries Government Documents Department

Fission product source terms for the LWR loss-of-coolant accident

Description: Models for cesium and iodine release from light-water reactor (LWR) fuel rods failed in steam were formulated based on experimental fission product release data from several types of failed LWR fuel rods. The models were applied to a pressurized water reactor (PWR) undergoing a hypothetical loss-of-coolant accident (LOCA) temperature transient. Calculated total iodine and cesium releases from the fuel rods were 0.053 and 0.025% of the total reactor inventories of these elements, respectively, with most of the release occurring at the time of rupture. These values are approximately two orders of magnitude less than releases used in WASH-1400, the Reactor Safety Study.
Date: July 1, 1980
Creator: Lorenz, R.A.; Collins, J.L. & Malinauskas, A.P.
Partner: UNT Libraries Government Documents Department

Brief account of the effect of overcooling accidents on the integrity of PWR pressure vessels

Description: The occurrence in recent years of several (PWR) accident initiating events that could lead to severe thermal shock to the reactor pressure vessel, and the growing awareness that copper and nickel in the vessel material significantly enhance radiation damage in the vessel, have resulted in a reevaluation of pressure-vessel integrity during postulated overcooling accidents. Analyses indicate that the accidents of concern are those involving both thermal shock and pressure loadings, and that an accident similar to that at Rancho Seco in 1978 could, under some circumstances and at a time late in the normal life of the vessel, result in propagation of preexistent flaws in the vessel wall to the extent that they might completely penetrate the wall. More severe accidents have been postulated that would result in even shorter permissible lifetimes. However, the state-of-the-art fracture-mechanics analysis may contain excessive conservatism, and this possibility is being investigated. Furthermore, there are several remedial measures, such as fuel shuffling, to reduce the damage rate, and vessel annealing, to restore favorable material properties, that may be practical and used if necessary. 5 figures.
Date: January 1, 1982
Creator: Cheverton, R.D.
Partner: UNT Libraries Government Documents Department

Status of the Fire Protection Research (FPR) Program. [BWR; PWR]

Description: The objective of the Fire Protection Research (FPR) Program under the sponsorship of the Office of Water Reactor Safety Research, USNRC, is to confirm the capability of safety features in use or planned for use in nuclear power plants. This objective is to be achieved by full scale testing and experimental and analytical evaluation of fire phenomenology.
Date: January 1, 1979
Creator: Klamerus, L.J.
Partner: UNT Libraries Government Documents Department