skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions

Abstract

A study was performed to assess the effects of aging on the Containment Isolation (CI) functions of Westinghouse Pressurized Water Reactors and General Electric Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research (NPAR) program, sponsored by the U.S. Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. Failure data from two national databases, Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LERs), as well as plant specific data were reviewed and analyzed to understand the effects of aging on the CI functions. This study provided information on the effects of aging on component failure frequency, failure modes, and failure causes. Current inspection, surveillance, and monitoring practices were also reviewed.

Authors:
; ; ;
Publication Date:
Research Org.:
Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research; Brookhaven National Lab., Upton, NY (United States)
Sponsoring Org.:
Nuclear Regulatory Commission, Washington, DC (United States)
OSTI Identifier:
211319
Report Number(s):
NUREG/CR-6339; BNL-NUREG-52462
ON: TI96008079; TRN: 96:010537
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: Mar 1996
Country of Publication:
United States
Language:
English
Subject:
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; WATER COOLED REACTORS; CONTAINMENT SYSTEMS; AGING; RELIABILITY; SYSTEM FAILURE ANALYSIS

Citation Formats

Lee, B.S., Travis, R., Grove, E., and DiBiasio, A.. Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions. United States: N. p., 1996. Web. doi:10.2172/211319.
Copy to clipboard
Lee, B.S., Travis, R., Grove, E., & DiBiasio, A.. Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions. United States. doi:10.2172/211319.
Copy to clipboard
Lee, B.S., Travis, R., Grove, E., and DiBiasio, A.. Fri . "Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions". United States. doi:10.2172/211319. https://www.osti.gov/servlets/purl/211319.
Copy to clipboard
@article{osti_211319,
title = {Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions},
author = {Lee, B.S. and Travis, R. and Grove, E. and DiBiasio, A.},
abstractNote = {A study was performed to assess the effects of aging on the Containment Isolation (CI) functions of Westinghouse Pressurized Water Reactors and General Electric Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research (NPAR) program, sponsored by the U.S. Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. Failure data from two national databases, Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LERs), as well as plant specific data were reviewed and analyzed to understand the effects of aging on the CI functions. This study provided information on the effects of aging on component failure frequency, failure modes, and failure causes. Current inspection, surveillance, and monitoring practices were also reviewed.},
doi = {10.2172/211319},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Mar 01 00:00:00 EST 1996},
month = {Fri Mar 01 00:00:00 EST 1996}
}
Copy to clipboard
Technical Report:
View Technical Report (9.27 MB)
DOI:  10.2172/211319
Save / Share:
Export Metadata
Save to My Library
You must Sign In or Create an Account in order to save documents to your library.

  • ;
    A study of the effects of aging on the Westinghouse Control Rod Drive (CRD) System was performed as part of the Nuclear Plant Aging Research (NPAR) Program. The objectives of the NPAR Program are to provide a technical basis for identifying and evaluating the degradation caused by age in nuclear power plant systems, structures, and components. The information from NPAR studies will be used to assess the impact of aging on plant safety and to develop effective mitigating actions. The operating experience data were evaluated to identify predominant failure modes, causes, and effects. For this study, the CRD system boundarymore » includes the power and logic cabinets associated with the manual control of rod movement, and the control rod mechanism itself. The aging-related degradation of the interconnecting cables and connectors and the rod position indicating system also were considered. The evaluation of the data, when coupled with an assessment of the materials of construction and the operating environment, leads to the conclusion that the Westinghouse CRD system is subject to degradation from aging which, if unchecked, could affect its intended safety function and performance as a plant ages. The number of CRD system failures which have resulted in a reactor trip (challenge to the safety system) warrants continued attention. Ways to detect and mitigate the effects of aging are included in this report. The current maintenance practices for the control rod drive system for fifteen Westinghouse plants were obtained through an industry survey. The results of the survey indicate that some plants have modified the system, replaced components, or expanded preventive maintenance. 33 refs., 35 figs., 18 tabs.« less

  • The isolation condenser system (ICS) is part of the emergency core cooling system in five U.S. boiling-water reactors. In the event that the reactor pressure vessel becomes isolated from the main condenser, the ICS removes decay heat from the reactor. The ICS is important to reactor safety because it is relied on to help mitigate core damage during a loss-of-coolant accident. In support of the U.S. Nuclear Regulatory Commission's Nuclear Plant Aging Research (NPAR) Program, staff from the Pacific Northwest Laboratory researched the aging of the ICS by reviewing available industry databases. Each component of the ICS was evaluated tomore » 1) identify applicable aging issues and also to 2) determine if the component had already been studied as a part of other NPAR assessments. The results of this preliminary study indicate that most of the critical ICS components have been previously evaluated by the NPAR program. The one ICS component that has not been specifically studied is the isolation condenser itself. There is little evidence in the databases to suggest that there have been problems with the isolation condenser. Only one plant, Millstone Unit 1, has ever had an isolation condenser tube failure problem recorded. This instance resulted from events that occurred early in the life of the plant. The problem was remedied through tube replacement. The isolation condenser and the pressurized-water reactor (PWR) steam generator were compared to illustrate that even though the isolation condenser is a heat exchanger, it is not subjected to the same service dynamics as the PWR steam generator. The isolation condenser operates for most of its service life in a relatively benign, static environment, resulting in a comparatively good service record. PNL staff recommend that the results of this research be used to continue studying the ICS to determine if the aging isolation condenser tubes are being adequately maintained. This new study should include an evaluation of the current inspection methods and a verification that they are effective in identifying tube aging degradation, such as intergranular stress corrosion cracking. Continued study may also provide beneficial input into the design of the Simplified Boiling Water Reactor.« less

  • Accumulators are important components used in many systems at commercial boiling water reactors (BWRs) and pressurized water reactors in the United States. The accumulators are vessels attached to fluid systems to provide 1) a limited backup source of stored fluid energy for hydraulic/pneumatic mechanical equipment, 2) a damping effect on pressure pulses in fluid systems, and 3) a volume of fluid to be injected passively into a fluid system. Accumulators contain a gas that is compressed or expanded as the fluid from the system enters or exits the accumulator. The gas and fluid in accumulators are usually separated from eachmore » other by a piston or bladder. In support of the U.S. Nuclear Regulatory Commission's Nuclear Aging Research Program (NPAR), the Pacific Northwest Laboratory conducted an analysis of available industry databases to determine if accumulator components already had been studied in other NPAR assessments and to evaluate each accumulator type for applicable aging issues. The results of this preliminary study indicate that two critical uses of accumulators have been previously evaluated by the NPAR program. NUREGICR-5699, Aging and Service Wear of Control Rod Drive Mechanisms for BUT Nuclear Plants (Greene 199 I), identified two hydraulic control unit components subject to aging failures: accumulator nitrogen-charging cartridge valves and the scram water accumulator. In addition, NUREGICR-6001, Aging Assessment of BWR Standby Liquid Control Systems (Buckley et al. 1992), identified two predominant aging-related accumulator failures that result in a loss of the nitrogen blanket pressure: (charging) valve wear and failure of the gas bladder. The present study has identified five prevalent aging-related accumulator failures: rupture of the accumulator bladder separation of the metal disc from the bottom of the bladder leakage of the gas from the charging valve leakage past the safety injection tank manway cover gasket leakage past O-rings. An additional study of the accumulator subcomponents associated with these failures is recommended, including an evaluation of current inspection programs to verify that they are detecting the aging degradation effects. The study may also provide beneficial input to the design of passive accumulator applications in advanced reactor designs.« less

  • ;
    To verify the use of reduced scale hydraulic models of large scale ratios to demonstrate the performance of containment emergency sumps, in view of concerns regarding possible scale effects, a test program involving two geometric scale models (1:2 and 1:4) of a full size sump (1:1) was undertaken as a part of the total test program towards the resolution of unresolved safety issue A-43, Containment Emergency Sump Performance. The test results substantiated that hydraulic models with large scales such as 1:2 to 1:4 reliably predicted the sump hydraulic performance; namely, vortexing, air-ingestion from free surface vortices, pipe flow swirl andmore » inlet loss coefficient. No scale effects on vortexing or air-withdrawals were apparent within the tested prediction range for both models. However, a good prediction of pipe flow swirl and inlet loss coefficient was found to require that the approach flow Reynolds number and pipe Reynolds number be above certain limits.« less

  • This document is presented as a demonstration of the ENC WREM-II ECCS model calculational procedure applied to a Westinghouse 2-loop PWR with a dry containment (R. E. Ginna plant, for example). The hypothesized Loss-of-Coolant Accident (LOCA) investigated was a split break with an area equal to twice the pipe cross-sectional area. The break was assumed to occur in one pump discharge pipe (DECLS break). The analyses involved calculations using the ENC WREM-II model. The following codes were used: RELAP4-EM/ENC26A for blowdown and hot channel analyses, RELAP4-EM FLOOD/ENC26A for core reflood analysis, CONTEMPT LT/22 modified for containment backpressure analysis, and TOODEE2/APR77more » for heatup analysis.« less