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in subjects: "210100 -- Power Reactors, Nonbreeding, Light-Water Moderated, Boiling Water Cooled"

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Implications of passive safety based on historical industrial experience

In the past decade, there have been multiple proposals for applying different technologies to achieve passively safe light water reactors (LWRs). A key question for all such concepts is, ''What are the gains in safety, costs, and reliability for passive safety systems.'' Using several types of historical data, estimates have been made of gains from passive safety and operating systems, which are independent of technology. Proposals for passive safety in reactors usually have three characteristi… more

Date: January 1, 1988

Creator: Forsberg, Charles W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Routine and post-accident sampling in nuclear reactors

Review of the Three Mile Island accident by NRC has resulted in new post-accident-sampling-capability requirements for utilities that operate pressurized water reactors and/or boiling water reactors. Several vendors are offering equipment that they hope will suffice to met both the new NRC regulations and an operational deadline of January 1, 1981. The advantages and disadvantages of these systems and projected future-new-system needs for TVA reactors are being evaluated in light of TMI experie… more

Date: January 1, 1980

Creator: Armento, W. J.; Kitts, F. G. & German, G. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Prioritization of tasks in the draft LWR safety technology program plan. Final report

The purpose of this report is to describe both the approach taken and the results produced in the SAI effort to prioritize the tasks in the Sandia draft LWR Safety Technology Program Plan. This work used the description of important safety issues developed in the Reactor Safety Study (2) to quantify the effect of safety improvements resulting from a research and development program on the risk from nuclear power plants. Costs of implementation of these safety improvements were also estimated to… more

Date: May 1, 1980

Creator: Lim, E. Y.; Miller, W. J.; Parkinson, W. J.; Ritzman, R. L.; vonHerrmann, J. L. & Wood, P. J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Light-water-reactor coupled neutronic and thermal-hydraulic codes

An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three st… more

Date: January 1, 1982

Creator: Diamond, D. J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Risk-Based Plant Performance Indicators

Tasked by the 1979 President's Commission on the Accident at Three Mile Island, the US nuclear power industry has put into place a performance indicator program as one means for showing a demonstrable record of achievement.'' Largely through the efforts of the Institute of Nuclear Power Operations (INPO), plant performance data has, since 1983, been collected and analyzed to aid utility management in measuring their plants' performance progress. The US Nuclear Regulatory Commission (NRC) has al… more

Date: January 1, 1989

Creator: Boccio, J. L.; Azarm, M. A.; Vesely, W. E. & Hall, R. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Development of the reactor safety film

This paper summarizes the text and describes the processes followed to develop the first computer-generated film of LASL's Reactor Safety efforts. The 11-1/2 min film with narrative and musical background gives a brief overview of reactor components, of how LASL's Reactor Safety groups develop and verify computer codes to anticipate accidents, and of how these codes were applied to the Three Mile Island accident.

Date: January 1, 1980

Creator: Sheheen, N.N.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Instrumentation needs in LWR severe fuel damage experiments

The Class 9 type nuclear accident is defined and the Three Mile Island type accident and proposed Idaho National Engineering Laboratory experiment series are described in some detail. Different types of severe fuel damage experiments are briefly discussed in order to show typical measurement requirements. General instrumentation needs and problems encountered in Class 9 accident research are outlined. It is concluded that the extremely high temperatures, high nuclear radiation fields, and oxidi… more

Date: January 1, 1980

Creator: McCormick, R.D.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Safety and licensing issues that are being addressed by the Power Burst Facility test programs. [PWR; BWR]

This paper presents an overview of the results of the experimental program being conducted in the Power Burst Facility and the relationship of these results to certain safety and licensing issues. The safety issues that were addressed by the Power-Cooling-Mismatch, Reactivity Initiated Accident, and Loss of Coolant Accident tests, which comprised the original test program in the Power Burst Facility, are discussed. The resolution of these safety issues based on the results of the thirty-six tes… more

Date: January 1, 1980

Creator: McCardell, R. K. & MacDonald, P. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of a partial scram event in a typical BWR/4

A study has been made of core thermal-hydraulic conditions and containment suppression pool water temperature following a partial failure to scram in a BWR 4. This was motivated by the occurrence of such an incident in the Browns Ferry Unit 3. If suppression pool temperature should rise too much, the effectiveness of the suppression pool can be impaired. However, the pool temperature in the present case rose no more than 58 degrees (F), which is considered to be acceptable. In this study the tr… more

Date: January 1, 1982

Creator: Lu, M. S.; Shier, W.; Levine, M. M. & Cerbone, R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in… more

Date: June 1, 1980

Creator: Oak, H. D.; Holter, G. M.; Kennedy, W. E. Jr. & Konzek, G. J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle description

The Nonproliferation Alterntive Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) pr… more

Date: June 1, 1980

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Licensee Event Report (LER) compilation for month of March 1988

This monthly report contains Licensee Event Report (LER) operational information that was processed into the LER data file of the Nuclear Safety Information Center (NSIC) during the one-month period identified on the cover of the document. The LERS, from which this information is derived, are submitted to the Nuclear Regulatory Commission (NRC) by nuclear power plant licensees in accordance with federal regulations. Procedures for LER reporting for revisions to those events occurring prior to 1… more

Date: April 1, 1988

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Probabilistic analysis of fires in nuclear plants

The aim of this paper is to describe a multilevel (i.e., staged) probabilistic analysis of fire risks in nuclear plants (as part of a general PRA) which maximizes the benefits of the FRA (fire risk assessment) in a cost effective way. The approach uses several stages of screening, physical modeling of clearly dominant risk contributors, searches for direct (e.g., equipment dependences) and secondary (e.g., fire induced internal flooding) interactions, and relies on lessons learned and available… more

Date: January 1, 1985

Creator: Unione, A. & Teichmann, T.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Data processing for the 1/5-scale Boiling Water Reactor Mark I pressure suppression experiment

A description is given of methods used for data reduction and post-processing of reduced data for the 1/5-scale Boiling Water Reactor (BWR) Mark I pressure suppression experiment. Output from approximately 200 transducers, recorded onto analog magnetic tape, were reduced to engineering quantities with an analog-to-digital, COBOL-like conversion code. The reduced data were analyzed with conversational FORTRAN codes and mass-processed for reports with batch-processing FORTRAN codes.

Date: January 13, 1978

Creator: Lai, W. & McCauley, E.W.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Assessment of engineering plant analyzer with Peach Bottom 2 stability tests

Engineering Plant Analyzer (EPA) has been developed to simulate plant transients for Boiling Water Reactor (BWR). Recently, this code has been used to simulate LaSalle-2 instability event which was initiated by a failure in the feed water heater. The simulation was performed for the scram conditions and for the postulated failure in the scram. In order to assess the capability of the EPA to simulate oscillatory flows as observed in the LaSalle event, EPA has been benchmarked with the available … more

Date: January 1, 1992

Creator: Rohatgi, U.S.; Mallen, A.N.; Cheng, H.S. & Wulff, W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Source-term evaluations from recent core-melt experiments

Predicted consequences of hypothetical severe reactor accidents resulting in core meltdown appear to be too conservatively projected because of the simplistic concepts often assumed for the intricate and highly variable phenomena involved. Recent demonstration work on a modest scale (1-kg) has already revealed significant variations in the mode and temperature for clad failure, in the rates of formation of zirconium alloys, in the nature of the UO/sub 2/-ZrO/sub 2/ eutectic mixtures, and in aer… more

Date: January 1, 1985

Creator: Parker, G. W.; Creek, G. E. & Sutton Jr., A. L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Transactions of the nineteenth water reactor safety information meeting

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: Article

Partner: UNT Libraries Government Documents Department

open access

Safety System Function Trend Indicator: Theory and Test Application

Methods for formulation, interpretation, and validation of dynamic risk and reliability indicators are studied. The use of these indicators for monitoring various levels of safety performance in nuclear power plants, as identified by probabilistic risk assessments (PRAs), such as safety system unavailability, safety system failure frequency, and core-damage frequency, are explored. Simplified indicators for detecting trends in the unavailability of safety systems in nuclear power plants not req… more

Date: January 1, 1989

Creator: Azarm, M. A.; Carbonaro, J. F.; Boccio, J. L. & Vesely, W. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Systematic methodology for the reduction of uncertainties in transient thermal-hydraulics by using in-bundle measurement data

The development of a systematic methodology for the reduction of uncertainties in transient thermal-hydraulics by using in-bundle measurement data is presented. The adjustment of the system parameters and responses and the reduction in their respective uncertainties is treated as a time-dependent constrained minimization problem. An on-line (i.e., real time) large scale optimization scheme is also outlined. Although formulated within the framework of reactor safety analysis, the proposed method… more

Date: January 1, 1980

Creator: Barhen, J.; Cacuci, D. G.; Wagschal, J. J. & Mullins, C. B.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Void coefficient in a unit cell for U-Pu water breeder reactors

The possibility of breeding with the Pu-U fuel cycle in light or heavy water has been established. For such lattices the void coefficient in a unit cell was calculated. It was found that for light water lattices the void coefficient is negative whereas for heavy water lattices the void coefficient is positive. However, it was shown that by introducing a relatively small amount of light water to the heavy water, a negative void coefficient is obtained while retaining many of the heavy water latt… more

Date: January 1, 1980

Creator: Ronen, Y.; Cojocaru, M. & Radkowsky, A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Steam explosion studies with single drops of molten refractory materials. [PWR; BWR]

Laser heating, levitation melting, and metal combustion were used to prepare individual drops of molten refractory materials which simulate LWR fuel melt products. Drop temperatures ranged from approx. = 1500 to > 3000K. These drops, several millimeters in diameter, were injected into water and subjected to pressure transients (approx. = 1MPa peak pressures) generated by a submerged exploding bridgewire. Molten oxides of Fe, Al and Zr could be induced to explode with bridgewire initiation. H… more

Date: January 1, 1980

Creator: Nelson, L. S.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Iodine chemical forms in LWR severe accidents

Calculated data from seven severe accident sequences in light water reactor plants were used to assess the chemical forms of iodine in containment. In most of the calculations for the seven sequences, iodine entering containment from the reactor coolant system was almost entirely in the form of CsI with very small contributions of I or HI. The largest fraction of iodine in forms other than CsI was a total of 3.2% as I plus HI. Within the containment, the CsI will deposit onto walls and other su… more

Date: January 1, 1991

Creator: Weber, C.F.; Beahm, E.C. & Kress, T.S.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Characterization of debris/concrete interactions for advanced research reactor and commercial BWR severe accidents

The core concrete interaction (CCI) is an important phase of any severe accident where the reactor vessel has failed and core debris is relocated onto the containment basemat. In recent calculations performed at the Oak Ridge National Laboratory (ORNL), CCI has been studied for severe accidents occurring in a commercial Boiling Water Reactor (BWR) and in a high-power density Department of Energy (DOE) research reactor that is currently in the conceptual design stage. Because of differences in t… more

Date: January 1, 1991

Creator: Hyman, C. R.; Taleyarkhan, R. P. & Greene, S. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Fission-product transport analysis for the loss of decay-heat-removal accident sequence at Browns Ferry

No Description Available.

Date: January 1, 1983

Creator: Niemczyk, S. J.; Wright, A. L.; Beahm, E. C.; Weber, C. F.; Lorenz, R. A.; Forsberg, V. F. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department