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Title: Investigation of pressure transient propagation in pressurized water reactor feedwater lines

Abstract

Results are reported of a study for the Nuclear Regulatory Commission (NRC) to provide a general understanding of pressure transient (water hammer) propagation in pressurized water reactor (PWR) steam generator feedwater piping systems. A typical feedwater network is defined, and pressure transient initiation is discussed, as well as the plausible pulse shapes reported. The analysis is performed by using the computer codes PTA and WHAM. Forces are calculated at elbows and valves by using momentum principles. The effects of pipe yielding, pipe wall friction, and elbow and value losses are included. Pipe yielding and elbow/valve effects are found to be important, and pressure magnitudes and forces are substantially reduced when these effects are included in the analysis. Typical pressure and force time histories are also given.

Authors:
Publication Date:
Research Org.:
California Univ., Livermore (USA). Lawrence Livermore Lab.
OSTI Identifier:
5452084
Report Number(s):
UCRL-52265
TRN: 78-001459
DOE Contract Number:
W-7405-ENG-48
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; AUXILIARY WATER SYSTEMS; WATER HAMMER; PWR TYPE REACTORS; MATHEMATICAL MODELS; FEEDWATER; AUXILIARY SYSTEMS; HYDROGEN COMPOUNDS; OXYGEN COMPOUNDS; REACTORS; WATER; WATER COOLED REACTORS; WATER MODERATED REACTORS; 210200* - Power Reactors, Nonbreeding, Light-Water Moderated, Nonboiling Water Cooled

Citation Formats

Sutton, S.B.. Investigation of pressure transient propagation in pressurized water reactor feedwater lines. United States: N. p., 1977. Web. doi:10.2172/5452084.
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Sutton, S.B.. Investigation of pressure transient propagation in pressurized water reactor feedwater lines. United States. doi:10.2172/5452084.
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Sutton, S.B.. Fri . "Investigation of pressure transient propagation in pressurized water reactor feedwater lines". United States. doi:10.2172/5452084. https://www.osti.gov/servlets/purl/5452084.
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@article{osti_5452084,
title = {Investigation of pressure transient propagation in pressurized water reactor feedwater lines},
author = {Sutton, S.B.},
abstractNote = {Results are reported of a study for the Nuclear Regulatory Commission (NRC) to provide a general understanding of pressure transient (water hammer) propagation in pressurized water reactor (PWR) steam generator feedwater piping systems. A typical feedwater network is defined, and pressure transient initiation is discussed, as well as the plausible pulse shapes reported. The analysis is performed by using the computer codes PTA and WHAM. Forces are calculated at elbows and valves by using momentum principles. The effects of pipe yielding, pipe wall friction, and elbow and value losses are included. Pipe yielding and elbow/valve effects are found to be important, and pressure magnitudes and forces are substantially reduced when these effects are included in the analysis. Typical pressure and force time histories are also given.},
doi = {10.2172/5452084},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Jul 22 00:00:00 EDT 1977},
month = {Fri Jul 22 00:00:00 EDT 1977}
}
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Technical Report:
View Technical Report (1.65 MB)
DOI:  10.2172/5452084
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  • During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity.

  • ; ;
    This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficientmore » action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage.« less

  • ; ;
    The response of the Westinghouse Zion-1 pressurized water reactor to transients initiated by loss of main feedwater with auxiliary feedwater unavailable was simulated using the Transient Reactor Analysis Code (TRAC). The normal response mode in which emergency systems perform as designed was first studied to identify critical equipment performance and operator actions necessary for normal recovery. Subsequent analyses were performed to determine the effects of additional equipment failures, such as valves sticking open, and delayed or degraded operation of emergency systems. Strategies were developed for operator actions not covered in existing emergency procedures and were tested using TRAC simulations tomore » evaluate their effectiveness in preventing core uncovery.« less