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The effects of aging on Boiling Water Reactor core isolation cooling system

Description: A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling system in commercial Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research program sponsored by the US Nuclear Regulatory Commission. The failure data, from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failure causes.
Date: June 1, 1994
Creator: Lee, Bom Soon
Partner: UNT Libraries Government Documents Department

Digital control systems in nuclear power plants: Failure information, modeling concepts, and applications. Revision 1

Description: This report briefly describes some current applications of advanced computerized digital display and control systems at US commercial nuclear power plants and presents the results of a literature search that was made to gather information on the reliability of these systems. Both hardware and software reliability were addressed in this review. Only limited failure rate information was found, with the chemical process industry being the primary source of information on hardware failure rates and expert opinion the primary source for software failure rates. Safety-grade digital control systems are typically installed on a functional like-for-like basis, replacing older analog systems without substantially changing interactions with other plant systems. Future work includes performing a limited probabilistic risk assessment of a representative DCS to assess its risk significance.
Date: June 23, 1993
Creator: Galyean, W. J.
Partner: UNT Libraries Government Documents Department

Auxiliary feedwater system aging study

Description: The Phase 1 Auxiliary Feedwater (AFW) System Aging Study, NUREG/CR-5404 V1, focused on how and to what extent the various AFW system component types fail, how the failures have been and can be detected, and on the value of current testing requirements and practices. This follow-on study, which will be provided in full in NUREG/CR-5404 V2, provides a closure to the Phase 1 Study. For each of the component types and for the various sources of component failure identified in the Phase 1 Study, the methods of failure detection were designated and tabulated and the following findings became evident: Instrumentation and Control (I and C) related failures dominated the group of failures that were detected during demand conditions; many of the potential failure sources not detectable by the current monitoring practices were related to the I and C portion of the system; some component failure modes are actually aggravated by conventional test methods; and several important system functions did not undergo any function verification test. The goal of this follow-on study was to categorize and evaluate the deficiencies in testing identified by Phase 1 and to make specific recommendations for corrective action. In addition, this study presents discussions of alternate, state-of-the-art test methods, and provides a proposed Auxiliary Feedwater Pump test at normal operating pressure which should do much to verify system operability while eliminating degradation.
Date: December 31, 1992
Creator: Kueck, J. D.
Partner: UNT Libraries Government Documents Department

Factors which affect fatigue strength of fasteners

Description: Axial load cycling fatigue tests of threaded fasteners are useful in determining fastener fatigue failure or design properties. By using appropriate design factors between the failure and design fatigue strengths, such tests are used to establish fatigue failure and design parameters of fasteners for axial and bending cyclic load conditions. This paper reviews the factors which influence the fatigue strength of low Alloy steel threaded fasteners, identifies those most significant to fatigue strength, and provides design guidelines based on the direct evaluation of fatigue tests of threaded fasteners. Influences on fatigue strength of thread manufacturing process (machining and rolling of threads), effect of fastener membrane and bending stresses, thread root radii, fastener sizes, fastener tensile strength, stress relaxation, mean stress, and test temperature are discussed.
Date: November 1, 1992
Creator: Skochko, G. W. & Herrmann, T. P.
Partner: UNT Libraries Government Documents Department

Materials performance in prototype Thermal Cycling Absorption Process (TCAP) columns

Description: Two prototype Thermal Cycling Absorption Process (TCAP) columns have been metallurgically examined after retirement, to determine the causes of failure and to evaluate the performance of the column container materials in this application. Leaking of the fluid heating and cooling subsystems caused retirement of both TCAP columns, not leaking of the main hydrogen-containing column. The aluminum block design TCAP column (AHL block TCAP) used in the Advanced Hydride Laboratory, Building 773-A, failed in one nitrogen inlet tube that was crimped during fabrication, which lead to fatigue crack growth in the tube and subsequent leaking of nitrogen from this tube. The Third Generation stainless steel design TCAP column (Third generation TCAP), operated in 773-A room C-061, failed in a braze joint between the freon heating and cooling tubes (made of copper) and the main stainless steel column. In both cases, stresses from thermal cycling and local constraint likely caused the nucleation and growth of fatigue cracks. No materials compatibility problems between palladium coated kieselguhr (the material contained in the TCAP column) and either aluminum or stainless steel column materials were observed. The aluminum-stainless steel transition junction appeared to be unaffected by service in the AHL block TCAP. Also, no evidence of cracking was observed in the AHL block TCAP in a location expected to experience the highest thermal shock fatigue in this design. It is important to limit thermal stresses caused by constraint in hydride systems designed to work by temperature variation, such as hydride storage beds and TCAP columns.
Date: November 21, 1992
Creator: Clark, E. A.
Partner: UNT Libraries Government Documents Department

Risk-based configuration control: Application of PRA in improving technical specifications and operational safety

Description: The objective of risk-based configuration control is to detect and control plant configurations form a risk perspective. The configurations of particular interest involve components which are down during power operation. Controlling plant configurations from a risk-perspective can provide more direct risk control and also more operational flexibility by allowing looser control in areas unimportant to risk.
Date: December 31, 1991
Creator: Samanta, P. K.; Kim, I. S. & Vesely, W. E.
Partner: UNT Libraries Government Documents Department

On monitoring nuclear power plant emergency diesel generator reliability

Description: If offsite power is interrupted, the availability of onsite alternating current power supplies is a major factor in assuring acceptable safety at commercial light-water-cooled nuclear power plants. To control the risk of severe care damage during station blackout accidents at a given plant, the reliability of the emergency diesel generators (EDGS) to start and load-run upon demand must be maintained at a sufficiently high level. The minimum EDG reliability, which we denote by RT, is targeted at either 0.95 or 0.975 per nuclear unit consistent with the reliability level that the plant operator assumed in the coping analysis for station blackout. In 1992 the US Nuclear Regulatory Commission (NRC) considered an amendment that would require licensees to test and monitor EDG reliability against performance-based criteria that indicate possible degradation from the EDG target reliability levels. They originally proposed the following set of fixed sample-size triggers for use in monitoring EDG reliability. The purpose of this report is to compare the performance of the proposed triggers with corresponding alternative sequential variable sample-size triggers which potentially permit earlier detection of EDG reliability degradation without significantly increasing the false alarm rate. The comparison is to be done in a simulated use environment by means of Monte Carlo simulation. We are also interested in the inverse conditional probabilities of reliability degradation given that a trigger has occurred.
Date: August 11, 1993
Creator: Martz, H. F.; Tietjen, G. L.; Kvam, P. H. & Abramson, L. R.
Partner: UNT Libraries Government Documents Department

Fuel-element failures in Hanford single-pass reactors 1944--1971

Description: The primary objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation dose that individuals could have received as a result of emissions since 1944 from the US Department of Energy`s (DOE) Hanford Site near Richland, Washington. To estimate the doses, the staff of the Source Terms Task use operating information from historical documents to approximate the radioactive emissions. One source of radioactive emissions to the Columbia River came from leaks in the aluminum cladding of the uranium metal fuel elements in single-pass reactors. The purpose of this letter report is to provide photocopies of the documents that recorded these failures. The data from these documents will be used by the Source Terms Task to determine the contribution of single-pass reactor fuel-element failures to the radioactivity of the reactor effluent from 1944 through 1971. Each referenced fuel-element failure occurring in the Hanford single-pass reactors is addressed. The first recorded failure was in 1948, the last in 1970. No records of fuel-element failures were found in documents prior to 1948. Data on the approximately 2000 failures which occurred during the 28 years (1944--1971) of Hanford single-pass reactor operations are provided in this report.
Date: July 1, 1993
Creator: Gydesen, S. P.
Partner: UNT Libraries Government Documents Department

Probabilistic based design rules for intersystem LOCAS in ABWR piping

Description: A methodology has been developed for probability-based standards for low-pressure piping systems that are attached to the reactor coolant loops of advanced light water reactors (ALWRs) which could experience reactor coolant loop temperatures and pressures because of multiple isolation valve failures. This accident condition is called an intersystem loss-of-coolant accident (ISLOCA). The methodology was applied to various sizes of carbon and stainless steel piping designed to advanced boiling water reactor (ABWR) temperatures and pressures.
Date: May 1, 1993
Creator: Ware, A. G. & Wesley, D. A.
Partner: UNT Libraries Government Documents Department

An empirical investigation of operator performance in cognitively demanding simulated emergencies

Description: This report documents the results of an empirical study of nuclear power plant operator performance in cognitively demanding simulated emergencies. During emergencies operators follow highly prescriptive written procedures. The objectives of the study were to understand and document what role higher-level cognitive activities such as diagnosis, or more generally `situation assessment`, play in guiding operator performance, given that operators utilize procedures in responding to the events. The study examined crew performance in variants of two emergencies: (1) an Interfacing System Loss of Coolant Accident and (2) a Loss of Heat Sink scenario. Data on operator performance were collected using training simulators at two plant sites. Up to 11 crews from each plant participated in each of two simulated emergencies for a total of 38 cases. Crew performance was videotaped and partial transcripts were produced and analyzed. The results revealed a number of instances where higher-level cognitive activities such as situation assessment and response planning enabled crews to handle aspects of the situation that were not fully addressed by the procedures. This report documents these cases and discusses their implications for the development and evaluation of training and control room aids, as well as for human reliability analyses.
Date: July 1, 1994
Creator: Roth, E. M.; Mumaw, R. J. & Lewis, P. M.
Partner: UNT Libraries Government Documents Department

The deterministic prediction of failure of low pressure steam turbine disks

Description: Localized corrosion phenomena, including pitting corrosion, stress corrosion cracking, and corrosion fatigue, are the principal causes of corrosion-induced damage in electric power generating facilities and typically result in more than 50% of the unscheduled outages. Prediction of damage, so that repairs and inspections can be made during scheduled outages, could have an enormous impact on the economics of electric power generation. To date, prediction of corrosion damage has been made on the basis of empirical/statistical methods that have proven to be insufficiently robust and accurate to form the basis for the desired inspection/repair protocol. In this paper, we describe a deterministic method for predicting localized corrosion damage. We have used the method to illustrate how pitting corrosion initiates stress corrosion cracking (SCC) for low pressure steam turbine disks downstream of the Wilson line, where a thin condensed liquid layer exists on the steel disk surfaces. Our calculations show that the SCC initiation and propagation are sensitive to the oxygen content of the steam, the environment in the thin liquid condensed layer, and the stresses that the disk experiences in service.
Date: May 1, 1993
Creator: Liu, Chun & Macdonald, D. D.
Partner: UNT Libraries Government Documents Department

A probabilistic analysis of a catastrophic transuranic waste hoist accident at the WIPP

Description: This report builds upon the extensive and careful analyses made by the DOE of the probability of failure of the waste hoist, and more particularly on the probability of failure of a major component, the hydraulic brake system. The extensive fault tree analysis prepared by the DOE was the starting point of the present report. A key element of this work is the use of probability distributions rather than so-called point estimates to describe the probability of failure of an element. One of the authors (MAG) developed the expressions for the probability of failure of the brake system. The second author (TJS) executed the calculations of the final expressions for failure probabilities. The authors hope that this work will be of use to the DOE in its evaluation of the safety of the waste hoist, a key element at the WIPP.
Date: June 1, 1993
Creator: Greenfield, M. A. & Sargent, T. J.
Partner: UNT Libraries Government Documents Department

Melt behavior of aluminum clad rods

Description: Since the Li-Al alloy cores in control rods used to control production reactors are susceptible to corrosion by heavy water, they were clad with Al. This paper reports results of an experimental and numerical study of the behavior of control rods heated to the point of clad and rod-core failure. Results show that the core of the rod melts first; the clad fails only after significant additional heating. Once the rod breaks and drops to the bottom of the quartz tube in the furnace, the lower section of the rod fails by ``poker-chipping`` downward as the topmost portion fails before the portion below it. Part of the core in the remaining top of the rod relocates immediately after rod separation, leaving a hollow tube of Al which also melts upon further heating.
Date: August 1, 1994
Creator: Geiger, G. T.; Long, T. A. & DeWald, A. B. Jr.
Partner: UNT Libraries Government Documents Department

Evaluation of MHTGR fuel reliability

Description: Modular High-Temperature Gas-Cooled Reactor (MHTGR) concepts that house the reactor vessel in a tight but unsealed reactor building place heightened importance on the reliability of the fuel particle coatings as fission product barriers. Though accident consequence analyses continue to show favorable results, the increased dependence on one type of barrier, in addition to a number of other factors, has caused the Nuclear Regulatory Commission (NRC) to consider conservative assumptions regarding fuel behavior. For this purpose, the concept termed ``weak fuel`` has been proposed on an interim basis. ``Weak fuel`` is a penalty imposed on consequence analyses whereby the fuel is assumed to respond less favorably to environmental conditions than predicted by behavioral models. The rationale for adopting this penalty, as well as conditions that would permit its reduction or elimination, are examined in this report. The evaluation includes an examination of possible fuel-manufacturing defects, quality-control procedures for defect detection, and the mechanisms by which fuel defects may lead to failure.
Date: July 1, 1992
Creator: Wichner, R. P. & Barthold, W. P.
Partner: UNT Libraries Government Documents Department

The NUCLARR databank: Human reliability and hardware failure data for the nuclear power industry

Description: Under the sponsorship of the US Nuclear Regulatory Commission (NRC), the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) was developed to provide human reliability and hardware failure data to analysts in the nuclear power industry. This IBM-compatible databank is contained on a set of floppy diskettes which include data files and a menu-driven system for locating, reviewing, sorting, and retrieving the data. NUCLARR contains over 2500 individual data records, drawn from more, than 60 sources. The system is upgraded annually, to include additional human error and hardware component failure data and programming enhancements (i.e., increased user-friendliness). NUCLARR is available from the NRC through project staff at the INEL.
Date: May 1, 1993
Creator: Reece, W. J.
Partner: UNT Libraries Government Documents Department

Test quality

Description: This document discusses inservice testing of safety-related components at nuclear power plants which is performed under the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (the Code). Subsections IWP and IWV of Section XI of the Code state test method and frequency requirements for pumps and valves respectively. Tests vary greatly in quality and frequency. This paper explores the concept of test quality and its relationship with operational readiness and preventive maintenance. This paper also considers the frequencies of component testing. Test quality is related to a test`s ability to detect degradation that can cause component failure. The quality of the test depends on several factors, including specific parameters measured, system or component conditions, and instrument accuracy. The quality of some currently required tests for check valves, motor-operated valves, and pumps is also discussed. Suggestions are made to improve test quality by measuring different parameters, testing valves under load, and testing positive displacement pumps at high pressure and centrifugal pumps at high flow rate conditions. These suggestions can help to improve the level of assurance of component operational readiness gained from testing.
Date: September 1, 1992
Creator: Hartley, R. S. & Keller, A. E.
Partner: UNT Libraries Government Documents Department

Most likely failure location during severe accident conditions

Description: This paper describes preliminary results from which finite element calculation results are used in conjunction with analytical calculation results to predict failure in different LWR vessel designs during a severe accident. Detailed analyses are being performed to investigate the relative likelihood of a BWR vessel and drain line penetration to fail during a wide range of severe accident conditions. Analytically developed failure maps, which were developed in terms of dimensionless groups, are applied to consider geometries and materials occurring in other LWR vessel designs. Preliminary numerical analysis results indicate that if ceramic debris relocates within the BWR drain line to a distance below the lower head, the drain line will reach failure temperatures before the vessel fails. Application of failure maps for these debris conditions to other LWR geometries indicate that in-vessel tube melting will occur in either BWR or PWR vessel designs. Furthermore, if this melt is assumed to fill the entire penetration flow area, the melt is predicted to travel well below the lower head in any of the reference LWR penetrations. However, failure maps suggest the result that ex-vessel tube temperatures exceed the penetration`s ultimate strength is specific to the BWR drain line because of its material composition and relatively large effective diameter for melt flow.
Date: December 31, 1991
Creator: Rempe, J. L. & Allison, C. M.
Partner: UNT Libraries Government Documents Department

Proceedings of the Third NRC/ASME Symposium on Valve and Pump Testing. Session 1A--Session 2C: Volume 1

Description: The 1994 Symposium on Valve and Pump Testing, jointly sponsored by the Board of Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the Nuclear Regulatory Commission, provides a forum for the discussion of current programs and methods for inservice testing and motor-operated valve testing at nuclear power plants. The symposium also provides an opportunity to discuss the need to improve that testing in order to help ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants results in the discussion of a broad spectrum of ideas and perspectives regarding the improvement of inservice testing of pumps and valves at nuclear power plants. This document, Volume 1, covers sessions 1A through session 2C. The individual papers have been cataloged separately.
Date: July 1, 1994
Partner: UNT Libraries Government Documents Department

Overview of ORNL/NRC programs addressing durability of concrete structures

Description: The role of reinforced concrete relative to its applications as either safety-related structures in nuclear power or engineered barriers of low-level radioactive waste disposal facilities is described. Factors that can affect the long-term durability of reinforced concrete are identified. Overviews are presented of the Structural Aging Program, which is addressing the aging management of safety-related concrete structures in nuclear power plants, and the Permeability Test Methods and Data Program, which is identifying pertinent data and information for use in performance assessments of engineered barriers for low-level radioactive waste disposal.
Date: June 1, 1994
Creator: Naus, D. J. & Oland, C. B.
Partner: UNT Libraries Government Documents Department

Aging assessment of nuclear air-treatment system HEPA filters and adsorbers. Volume 1, Phase 1

Description: A Phase I aging assessment of high-efficiency particulate air (HEPA) filters and activated carbon gas adsorption units (adsorbers) was performed by the Pacific Northwest Laboratory (PNL) as part of the US Nuclear Regulatory Commission`s (NRC) Nuclear Plant Aging Research (NPAR) Program. Information concerning design features; failure experience; aging mechanisms, effects, and stressors; and surveillance and monitoring methods for these key air-treatment system components was compiled. Over 1100 failures, or 12 percent of the filter installations, were reported as part of a Department of Energy (DOE) survey. Investigators from other national laboratories have suggested that aging effects could have contributed to over 80 percent of these failures. Tensile strength tests on aged filter media specimens indicated a decrease in strength. Filter aging mechanisms range from those associated with particle loading to reactions that alter properties of sealants and gaskets. Low radioiodine decontamination factors associated with the Three Mile Island (TMI) accident were attributed to the premature aging of the carbon in the adsorbers. Mechanisms that can lead to impaired adsorber performance include oxidation as well as the loss of potentially available active sites as a result of the adsorption of pollutants. Stressors include heat, moisture, radiation, and airborne particles and contaminants.
Date: August 1, 1993
Creator: Winegardner, W. K.
Partner: UNT Libraries Government Documents Department

Auxiliary feedwater system risk-based inspection guide for the McGuire nuclear power plant

Description: In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. McGuire was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the McGuire plant.
Date: May 1, 1994
Creator: Bumgardner, J. D.; Lloyd, R. C.; Moffitt, N. E.; Gore, B. F. & Vo, T. V.
Partner: UNT Libraries Government Documents Department

The deterministic prediction of damage functions to low pressure steam turbines

Description: Localized corrosion phenomena, including pitting corrosion, stress corrosion cracking, and corrosion fatigue, are the principal causes of corrosion-induced damage in electric power generating facilities and typically result in more than 50% of the unscheduled outages. Prediction of damage, so that repairs and inspections can be made during scheduled outages, could have an enormous impact on the economics of electric power generation. To date, prediction of corrosion damage has been made on the basis of empirical/statistical methods that have proven to be insufficiently robust and accurate to form the basis for the desired inspection/repair protocol. In this paper, we describe a deterministic method for predicting localized corrosion damage. We have used the method to illustrate how pitting corrosion initiates stress corrosion cracking (SCC) for low pressure steam turbine disks downstream of the Wilson line, where a thin condensed liquid layer exists on the steel disk surfaces. Our calculations show that the SCC initiation and propagation are sensitive to the oxygen content of the steam, the environment in the thin liquid condensed layer, and the stresses that the disk experiences in service.
Date: June 1, 1993
Creator: Liu, Chun & Macdonald, D. D.
Partner: UNT Libraries Government Documents Department