ML20116N944
| ML20116N944 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 10/31/1992 |
| From: | Gore B, Lloyd R, Moffitt N, Vo T Battelle Memorial Institute, PACIFIC NORTHWEST NATION |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| CON-FIN-L-1310 NUREG-CR-5831, PNL-7783, NUDOCS 9211240202 | |
| Download: ML20116N944 (36) | |
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{{#Wiki_filter:- - . . . . . .- - _ - - .. _ - . -- NUREG/CR-5831 PNL-7783 emm e gmen Auxi iary Feecwater System Ris:(-Basec Insaection Guice for :ae Comancle Pea:i Nuclear Power Pian:
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I'repared t>y it. ('. IJoyd, N. II. Moffitt,11.1 Core, T. V. Yo l'acific Northwest 1,aboratory Operated by > llattelle Memorial Institute l'repared for U.S. Nuclear Itegulatory Commission i hhjl240202921031 O ADOCK 05000443 ) PDR l
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Referenced documents available for inspection and copying for a fee from the NRC Public Document Room
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include NRC correspondence and internal NRC memoranda: NRC bulletins, circulars, information notices, inspection and investigation notices licensee event reports: vendor reporti and correspondence; Comrns. sion papers: and apphennt and bcensee documents and correspondence. The fobowing docbments in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC sponsored conference proceedings, international ag?eement toports, grant publications and NRC booklets and brochures. Also available are regulatory guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Cornmission Issuances. Documents available from the National Technical information Service include NUREG series reports and technical reports prepared by other Federal agencies and reports prepared by the Atomic Energy Commis-ston, forerunner agency to the Nuclear Regulatory Cs Wssion. Documents available from pubhc and special technical litwaries include all open 1*terature items, such as books, journal articles, and transactions 4 federaf Hegister notices, Federal and State legislation, and con-pressional tiports can usu lly be obtained from these librarles. Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference pro 4 ceed.ngs are available for purchase from the organizatinn sponsoring the pubhcation cited. Single copisa cf NRC draf t reports are avaltable free, to the extent of supply, upon written request to th0 Office of Admirustration, Distribution and Mail Services Section, U.S. Nuclear Reutatory Commission, Washington, DC 20$$$. - CopV of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library,7920 Norfolk Avenue, Bethesda, Maryland, for use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organ 128 tion or, if they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018 DISCLAIMER NOTICE "This report was prepared as an account of work sponsorod by an agency of the United States GovemmonL Neither the Unitod States Govemmont nor any agency thercof, or any of tror employees, makes any warranty, expressed or implied, or assumes any legal liability of responsibility for any third pr.rty's use, or the rosults of - such use, of any information. epparatus, product or procots disclosed in this report, or represents that its use by such third party would not infringe privately owned rights,
NUltEG/ Cit-5831 PNI.c7783
-1 Auxiliary Feedwater System Risk-Based Inspection Guide for the Comanche Peak Nuc ear Power Plant
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Manuscript Cornpleted: September 1992 Date l'ublished: October 1992 l'repared by , it. C IJoyd. N. !!. Moffitt,11.17. Gore, T. V. Vo Pacific Notthwest laboratory Itichland, WA 99352 l'repared for Division of lladiation Protection and Emergency Preparedness Office of Nuclear Iteactor llegulation U.S. Nuclear Itegulatory Commission Washington, DC 20555 NItC FIN L1310
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O Abstniet in a study sponsored by the U.S. Nuclear Regulatory Commission (NRC), Pacific Northwest laboratory has developed and applied a methodology for deriving plant spreific risk based inspection guidance for the auxiliary feedwater (AFW) systern at pressurlied water reactors that have not undergone probabilistic risk assessment (PRA). This ' methodology uses existing PRA results and plant operating experience information. IIxisting PRA based itupection i guidance information recently developed for the NRC for various plants was used to identify generic component lailure modes. This information was then combined with plant specific and industry-wide compment information and failure data to identify failure modes and failure mechanisms for the AIM splem at the selected plants. Comanche Peak was sclected as one of a series of plants for study The product of this c: fort is a prioritlic4 listing of AIM failurn w hich have occurred at the plant and at otinct PWRs. 'Ihis listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the Comanche Peak plant. S l
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f iii - NURI:G/CR 5831
Contents l l l Abstract .......... .......................... ........ ................................... .... ill lx Summary..................................................................................... 1.1 - 1 I n t r od u c t io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2 Coma nche Peak AITV System . . . . . . . . . . . . . . . . . . . . . . . . . . . .............................. 4.... 2.1 2.1 2.1 Syste m Descript ion . . . . . . . . 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ............................. l 2.1 2.2 S u enss Cri t e r io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3 Sys t e rn De pe n d e n cies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ... .. 2.2
.......................................... 2.2 2.4 Operational Constraints . . . . . . . . . . . . . . . . . . . . . . .
..... . .................... .......... 3.1 '
3 !ns[wetion Guidance for the Comanche Peak AITV Sys, tem 3.1 Risk important AMV Components and Pallures Modes .... ... .... ................... 3.1 l 3.1.1 Multiple Pump Failures Due to Common cause . . . . . . . . . . . . .................... 4.... 3.1 3.1.2 'Ibrbine Driven Pump Falls to Start or Run . . . . . . . . .. . ..... .................... 3.2 3.1.3 Motor Driven Pump 1 or 2 lhils to Start or Run . . . . . . . . . . ... ..................... 3.3 3.1.4 Pump Unavailable Duc to Maintenanec or surveillance . . . . . . .... . ............ ....... 3.4 , 3.1.5 Air Operated Flow C(mtrol %lva Fall Closed . . . . . . . . . . . . .......................... 3.4 3.1.6 Motor Operated Isolation %!ves Ibil Closed ..... . ... .............................. 3.5 3.1.7 Manual Suetion or Discharge Wh'es Fail Closed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5 3.1.8 Leakage of Ilot Feedwater through Ched Wives . . . . .. .... .... ......... .......... 3.6 3.2 Risk Impo. tant AISY System Walkdown Thble ..... ............. ............... .......... 3.6 4 Generic Risk Insights From PRAs . . . . . . . . . . . . . . . ..... . ... . ..... ............. .. 4.1 4.1 Risk Irnportant Accident Sequences imolving Af4V System Failurc . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 4.1.1 less oI Power System . , . . . , . . . . . . . . . . . . . . . . . . . . .... ... ................... ..... 4.1 4.1.2 'llansient-Caused Reactor or % bine 'itip . . . . . . . , . ....... ........................ .. 4.1 4.1 4.1.3 ) ms o f M a i n Feed wa t e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4,1 4.1.4 Steam Generator %be Rupture (SGTR) . . . . . , , , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
..........., .... ... ............... 4.1 4.2 Risk Important Component Failure Modes . . . . . . . . .
5 Failure Modes Detennined From Operating Paperience , . . . .... .... ......... ......... 5.1. 5.1 Comanche Peak Experience . . . . . . . . . . . . . . . . . ... . ...... ... ... ................ 5.1 - 5.2 Industry Wide F.xperience 4 ...... ..... ....... .. .. .. ... ,, , ,... ... . ....... 5.1. v NUREGli 1831
t 5.2.1 Common ca use linit ur es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . 5.1
- 5. 2.2 i l u m a n Er r ors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3 5.2.3 Design 'EngincCring l'roblems and Errors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-5.2M CO!11lO!1CuiIIailurc$ ...............................................................,. $.$
6 RCI0rCDCCS........ ...... ................................... ............................... 6.1 NUREG/CR.5831 vi
mm--i mm- , Figures 2.3 2.1 Comanche l'eak Auxiliary ited wa ter Spte m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
'thbles 3.7 3.1 Risk Importance AISV Splem Walkdown'Thble .... ...........................................
i vg NUREG/CR-$831
0 Summary , I i
'Ihis document presents a compilation of auxillary feedwater (AITV) sptem failure information which has been '
screened for risk significance in terms of failure frequency and degradation of sptem performar.cc. It is a risk priori. tiicd listing of failure events and their causes that are significant enough to warrant mnsideration in inspection plan-ning at the Comanche Peak plant. This information is presenttxt to provide inspectors with increased resources for - inspection planning at Comanche Peak. The risk importance of various component failure rnodes was identified by analysis of the results of probabilistic risk assessments (PRAs) for many pressurited water reactors (PWRs) 110 wever, the component failure categories identi-fled in PRAs are rather broad, because the failure data used in the PRAs is an aggregate of many individual failurcs having a variety of root causes. In 01 der to help inspectors focus on specific aspects of component operation, mainte-nance and design which might cause these f ailurcs, an extensive review of component f.alute information was per-formed to identify and rank the root causes of these component failures. Iloth Comanche Peak and industry wide fall; ' ute information was analy/cd. Failure causes were sorted on the basis of frequency of occurrence and seriousness of consequence, and categorlied as common cause failures, human errors, design problern, or mmponent failures. This information h presented in the body of this document. Section 3.0 provide brict descriptiont.of these risk-impor. tant failure causes, and Section 5.0 presents more extensive dhcussions, with specific examples and references 'the entries in tiie twoections are cross-referenced. An abbreviated system walkdown table is presented in Section 3.2 which includes only components identified as risk , important. This table lists the sptem liacup for normal, standby system operation.
'this information pe;msts an inspector to concentrate on comp (ments important to the prevention of mic damage, llowever,it is _Important to note that inspections should not focus exclusively on these comp (ments. Other compo-nents which perform essential functions, but which are not included because of high reliability or redundang, must also be addressed to ensure that degradation does not increase their failure probabilities, and hence their risk importance.
r s I l ix NUREO/CR 5831
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_ _ . _ _ . - . _ . . . _ _ _ _ _ . _ . . _ _ _ _ _ . _ - - _ - _ . _ _ . . _ . = _ . 1 Introduction This document is one of a series providing plant specific The remainder of the document describes and discusses inspection guidance for auxiliary feedwater (AFW) syv the information used in compiling this inspection guld-tems at pressurired water reactors (PWRs). %Is guld- ance. Section 4.0 describes the risk importance informa-ance is based on information from probabilistic risk tion which has been derived from PRAs and its sources. assessments (PRAs) for similar PWRs, industry-wide As review of that section will show, the failure events - operating experience with AISV systems, plant specific identified in PRAs are rather broad (e.g., purnp fails to AISV system descriptions,and plant 4pecific operating start or run, valve falls closed). Section 5.0 addresses experience. It is not a detailed inspection plan, but the specific failure causes w hich have been combined 1 rather a compilation of AITV sysicm failure information under these broad events. which has been screened for risk significance in terms of failure frequency and degradation of system perform. AITV1 m operating history was studied to identify ance. The result is a risk prioritl/cd listing of failure the various specific failures which have been aggregated events and the causes that are significant enough to war- into the PRA failure events. Section 5.1 presents a sum. Iant consideration in inspection planning at Comanche mary of Comanche Peak failure information, and See-Peak, tion 5.2 presents a review of industry. wide failure infor. mation. The industry-wide information was compiled 4 This inspection guidance is presented in Section 3.0, from a variety of NRC sources, including AEOD analy, following a description of the Comanche Peak AIAV sys- ses and reports,information notices, inspection and en- ' tem in Section 2.0. Section 3.0 identifies the risk impor- forcement bulletins, and generic letters, and from a tant system camp ments by Comanche Peak identifica- variety of INPO reports as well. Some Lkensee Event tion number, followed by brl'f descriptions of each of Reports and NPRDS event descriptions were also . the various failure causes of that compment. These in- reviewed. Finally,informatlon was included from (lude specific human cri' ors, design deficiencies, and reports of NRC. sponsored studies of the effects of plant hardware failures. De discussions also identify where agl ag, w hich it.. ~: quantitative analyses of reported common cause failures have affected multiple, redun- AISV system failures. This industry-wide information dant comp ments. These brief discussions identify spe- was then combined with the plant specific failure infor- , cific aspects of system or component design, operation, mation to identify the various root causes of the broad maintenance, or testing for inspection by observation, failure events used in PRAs, which are identified in records review, training observation, pmcedures review, Section 3.0. or by observation of the implementation of procedures. An Al%V system walkdown table identifying rbk impor-tant components and their lineup for normal, standby system operation is also provided.
.5 1.1 NUREG/CR 5831 r
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2 Conianche Peak AFW System This scetion presents an oven .lew of the Comanche switch activated normally closed motor operated valves. Peak AFW systt m, including a simplified schematic syv Power, control, and instrumentation associated with tem diagram. In addition, the system succe , eriterion, each train are independent from each other. Steam for system dependencies, and administrative operational the turbine driven pump is supplied through 2452-1 and constraints are also presented. -2 f rom steam generators 1 an; 4, from a point upstream of the main steam isolation valves. Each Al W pump is equipped with a recirculation flow system which pre-vents pump deadheading. 2.1 System DeScripilon The discharges of tbc motor driven pumps are normally The AFW system provides feedwater to the steam gen, aligned so that the #1 pump supplies the 1 and 2 steam crators (50) to allow sceandary-side heat removal f rom l'enerators and the #; pamp supplies the 3 and 4 steam the prim 2ry system when main feedwater is unavailable. generators. Cross-connect valves are provided to allow The system is' capable of functioning for extended feeding of any steam generator from either pump. The periods, whkh allows time to restore main feedwater cromonnect valves are closed and administratively flow or to proceed with an orderly coutdown of the plant controlled. The turbine-driven pump can also feed all to . here the residual heat removal (RHR) system can four steam generators, but through separate lines. remove decay heat. A simplified schematic diagram of NormaHy open, moust operated containment isolation the Comanche Peak AFW system is shown in Figure 2.1. valves are h>cated downstream of the feed regulator valves. Manual locked open isolation valves are also lhe AFW system consists of a Condensaic Storage Tank provided for maintenance. Flow control valves (2453 (CST), two rnotor-driven pumps, one turbine-driven A&II,2454 A&II,2459 2462)in each of the eight feed-pump and associated piping, valves, and instrumenta-water lines aic pneumatic. Each line also contains tion. The system is designed to respond to any starting multiple check valves to prevent leakage from the feed-signal and establish flow automatically. All pumps start water lines, en receipt of a steam generator low low level signal. fThe motor-driven pumps statt on low low levelin one The CST provides the normal suction source for the SG, whereas, two SG low low level signals are required AFW system and is required to store water at a level suf-for a turbine-driven pump start.) The motor-driven ficient to maintain the reactor coo! ant sy: tem (RCS) at MD) pumps start for the following conditions: a safety (injection signal a blackout signal, a trip of bothhotmain standby for eighteen hours with steam discharge to annosphere or maintain the RCS at hot standby for four icedwater pumps,and an AMSAC (ATWS Mitigation hours fo!! owed by a cool down to 350 degrecs Fahren-System Aetuation Citeuitry) signal. The single nirbine-heit at a rate of 50 degrees per hour. All tank connec. driven (TD) pump starts on a blackout signal, and an tions except those required for instrumentation, aux-AMSAC signal. ihary feedwater pump suction, and tank drainage are loc, ted above this minimum capacity of 282,540 gallons. The preferred souice of AFW pump suction is from the CST. A common header supplies water through a hicked open valse and a check valve to the suction of both the motor-driven pumps. A sceand header sup- 2.2 SUCCESS CriterlOn plies water to the turbine driven put ps, also through a locked open valve and a check valve. An additional System success requires the operation of at least one back up soutee of water for the AFW pumps is provided pump oplying a mir.imum flow of 430 ppm to at least from the service water system (SSW), through two key two si .im generators. 2.1 NUREO/CR 5831 - - - - - - - - - - - - _ _ _ _ ^ ^ ' ' ~ " - - - - - - - - - - - - - . _ _ _ _ _ _ _ _ _ _ .
Comanche Peak AISV Sptem 2.3 Systein Dependencles twelve hours. With thlec AISV pumps inoperable,cor-rective action to restore at least one pump to operable
. "Ihc AISV sptem depends on AC power for motor- status must be initiated immediately, driven pumps and motor-co.itrolled isolation valves, DC power for control power to pumps, valves, and auto- 'Itc Comanche Peak Rchnical Specifications requires a
' 539, indicated water level to be stored in the CST. With matic actuation signals, and instrument air for AINV flow c4mtrol vahes. Air operated valves are also pro- the CSTinoperable,it must be restored to operable vidc4 with safety class air accumulators to ensure the t.tatus within four hours or the plant must be shut down valves operate as required on a loss of instrument alt. In to hot standby within the next twelve hours. if the addition, the turbine-driven pump also requires steam nuclear service water pond is demonstrated to u opera-availability. -. ble,it may serve as a backup AINV supply for seven days before plant shutdown is required. Comanche Peak tchnical Specifications also require that AISV area 2.4 Operational Consirainis * 3*"' "' (I""d *' ^I'Y P"* P '""*') "" h* maintained less than specified maximum limits. If area temperature is greater than the limits for normal condi-When the reactor is in Modes 1,2,or 3, the Comanche lions for greater than 8 hours, the Licensec is required Peak 'Rchnical Specifications require that all three to preparc a written report to the Nuclear Regulatory AINV pumps and associated flow paths are operabic Commission within 30 days. If area temperature is with each motor-driven pump powered from a different greater than the limit for abnormal conditions,in addl. cmcrgency bus. If one AIAV pump becomes inoperabic, tion to a report, the area temperature must be returned 11 must tvt restored to operable status within 72 hours or to normal within 4 hours or the equipment in the af-the plar must be shut down to hot standby within the fected area is declared inoperable. next twelve hours, if two AF sV pumps are inoperable, the plant must be shut down to hot standby within b NUREG/CR-5831 2.2
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3 Inspection Guidance for the Conianche Peak AFW System in this section the risk important mmponents of the 3.1.1 Multiple Pufnp failures Due 10 Comanche Peak AIM System are identilied, and the im. Common Cause portant failure modes for these components are briefly descrit ed. 'lhese failure modes include specific human 1he follow ng listing summarires the most important errors, design deficiencies,and types of hardware fall- multiple pump failure modes identified in Section 5.2.1, Common Cause Rdlures, and each item is keyed with a urcs which have been observed to occur for these ponents, both at Comanche Peak and at PWRs through com" . 3 digit code to entries in that section. out the nuelcar industry. The discussions also identify where common cause failuro have affected multiple, re'
- Incorrect operator intervention into automatic sys-dundant components. These brief discussions identify tem functioning, including improper manual start-specific aspects of system or mmponent design, oper- ing and securing of pumps, has caused failure of all ation, maintenance,or testing for inspection activities. pumps,includingoverspeed trip on startup, and h4a-Dese activitics include observation, temids review, bility to rotart prematurely secured pumps. CCl.
training observat'on, procedurcs review, or by observa-tion of the implementation of procedurcs. Inspection Suggntion . Observe Abnorm .I and II:nctgency Operating Procedure (AOP/IIOP) simu. Thble 3.1 is an abbreviated AIM sptem walkdown table lator training exercises to verify that the operators which identifies risk important components. This table cornply with procedures durine, observed evolutions. lists the system lincup for normal (standby) system oper* Observe surveillance testing on the AIM system to ation. Impection of the identified components addres* verify it is in strict compliance with the surveillance ses essentially all of the risk associated with AIM sy9 tot procedure. tem operation.
- Valve mispositioning has caused failure of all pumps. Pump suction, steam supply, and instru-3.1 Risk Important AFW Components ment isoladon valves have been involved. CC2.
and Failure Modes impection Suggntion Verify that the system valve alignment, air operated valve control and valve Cornmon cause failures of muidple pumps are the most actuating air prnsures are wrrect ming 3.1 % risk-important failure modes of Alv system cornpo. down 1hble, the system operating procedures, and nents These are followed in importar.cc ty sinric pump operator rounds logsheet. Review surveillance pro-failures, level c<mtiol valve failures, and individual check cedurn that alter the standby alignment of the leakage failure >< AIM splem. linsure that an adequate return to normal section exist.s. The following scetions address each of these failute modes,in decreasing order of risk-importance. They
- Steam binding has caused failure of multiple pumps.
present the important root causes of these component Dk rouhed from leakage of hot feedwater past , failure modes which have been distilled liom historical check valves and a motor operated valve into a com-( records. IIach item is keyed with a three digit code to mon discharge header. CC10. Multiple pump discussions in Section 5.2 where additional information Wam binding has aho resuhed from improper valve on historical events is pr;sented. lineups, and fmm running a pump deadheaded, CC1 3.I NURiiG/CR 5831 v--,+- . , , - ,eme .,_.u m .- - % , -, . -
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Inspection Guidance l Inyiection Suuntion . Verify that the pump dis- suction head (NPS11). CC7. At ILIL Robinson, charge temperature is within the limits specified on design reviews have identified inadequa tely sized
- he operator rounds logsheet. Assure any instru- suction piping which could have yielded insufficient ments used to verify the temperature by the utility NPSil to supp>rt operation of snore than one are of an appropriate range and included in a call. pump. CCH.
bianon program. Verify affected pumps have been vcnted in acmrdance with procedures to ensure Inspection Suuntion Anure that plant condi-steam binding has not occurred. Verify that a tions which could tesult in the bk)ckage or degrada-maintenance work request has been written to re- tion of the suction flow path are addressed by syu palt leaki 3 check valves, tem maintenance and test ptocedures. Examples include,if the AFW system has an emergency source
- Pump control circuit deficiencies or deMgn modifi- from a water system with the potential for bio foul.
cation errors have caused failures of mtatiple pumps ing, then the system should be perkxlically treated to auto start, spurious pump trips during operation, to prevent buildup and toutinely tested to assure an and failures to totart after pump shutdown. CC4. adequate flow can be achieved to support operation incorrect setpoints and control circuit calibrations of all pumps,or inspected to usure that bio fouling have also prevented proper operation of multiple is not occurring. Design changes that affect the pumps. CC5. suction flow path should repeat testing that verified an adequate suction source for simultaneous opera-Impection Suuntion - Review design thange Im. tion of all pumps. Verify that testing has, at some-plementation doeurnents for the post maintenance time, demonstrated simultaneous operation of all testing required prior to returning the equipment to pumps. Verify tlat surveillances adequately test all - servwe. Asture the testing verifics that all poten- aspects of the system design functions, for example, llally impacted lunctions operate correctly, and in. demonstrate that the AFW pumps will trip on low cludes repeatir.g any plant start up or hot functional suction pressure. > testing that may be affected by the design change. a 3.1.2 %rbine Driven Purnp Falls to Start or loss of a vital power bus has failed both the turbine' Ilun I driven and one motor-driven pump due to loss of control power to steam admission valves or to tur-
- Improperly adjusted and inadequately maintained bme controls, and to motor controls powered from turbine governors have caused pump failures. IIE2.
the same bus. CCh. Problems include worn or kiosened nuts, set screws, linkages or cable connections, oil leaks and/or c<m. ' Impection Suuntion - The material condition of '
- tamination, and electrical failures of resistors, the electrical equipment is an indicator of probabic transistors, dhxtes and circuit cards, and erroneous reliability. Review the Preventative Maintenance grounds and connections. CF5. Governor problems, (PM) records to assure the equiprnent is rnalinained bearing wear, low oil,and human et for in making on an appropriate Ircquency for the environment it proper settings have caused failure of the turbine is in and that the PM's are actually being performed driven pump at Commanche Peak.
as required by the program. Review the outstanding Corrective Niaintenance records to assure the defi. Impectfon Suuntion Review PM records to ciencier found on the equipment are promptly. corrected, assure the governor oil is beitC replaced within.the designateu pequency. During plant walkdowns
- careft.ily opat t he governor and linkages for kiose Simultaneouutartup of rnultiple pumps has caused fast:ncrr4 leaks, and unsecured or degraded conduit.
oscillations of pump suction pressure causin8 Ruiew vendor manuals to ensure PM procedures i multiple pump trips on low suction pressure, are performed according to manufacturer's recom-despite the existence of adequate stmic net poshire mendations and good maintenance practices, NUREG/CR 5831 12 _ __ . . _ _u _ _ . . . , _
inspection Guidance
- 'Ibrbines with Woodward Model PG PL of the oilin the turbine, interfering with proper governors have tripped on overspeed when overspeed trip operation. Review training pro-restarted shortly af ter shutdown, unless an cedures to ensure operator training on resetting the operator has h>cally exercised the speed setting TTV h current.
Anob to drain oil from the governor spred set-ting cylinder (per procedure). Automatic oil
- low lubrication oil pressure resulting from dump valves are now asallable through 1erry. heatup due to previous operation has prevented DE4. pump restart due to failre to satisfy the protective interkick. DES.
Insyrtion Suggestion - Observe the operation of the turbine driven Aux 1 ced pump and assure that inspection Suggestion - lew oil ptessure is a trip the governor is reset in accordance with operating that is in service at all timm for the turbine driven procedutes. Assure the turbine is not coasting over, AIM pump. Normally the low oil pressure ocerts whkh can result in refill of the speed setting at approximately 14(K) tpm and serves to protect the cylinder. pump from low RPM operation, howeve-low oil pressure due to a plugged filter will aho cause a trip.
- Condensate slugs in steam lines have caused Review PM record, to assure the filter is replaced turbine overspeed trip on startup. Rsts on the designated frequency.
repeated right af ter such a trip may fail to indicate the problem due to warming and clear- 3.1.3 Motor Driven Punip 1 or 2 l' alls to Start ing of the steam lines. Surveillances should or Itun exercise all steam supply connections. DE2.
=
Control circuits used for automatic and manual Inspytlon huggestion - Verify that the steam traps pump starting are an important cause of motor are valved in on the steam supply line. Ibr steam driven pump failures,as are circuit L aker traps that are on a pressuriecd portion of the steam failures. CI'7. Controlcircuit problems and a line, theck the steam trap temperature (if unlagged) Idown fuse due to overload hase occurred at to assure it is warmer than ambient (otherwise it Comanche Peak. may be stuck or have a plugged line).11 the steam trap discharge is visible, assure there is evidence of Inspection Suggestion - Review corrective main. liquid discharge. tenance records w ben control circuit problems occur to determine if a trend exists. Evciy time a
* ' hip and throttle vahe (TTV) problems which breaker is racked in a PMT should be performed to have failed the turbine driven pump intlude start the pump, assuring no control circuit problems phpicitly bumping it, failure to reset it fol- have occurred as a result of the manipulation of the lowing testing, and failures to verify control breaker. (Control circuit stabs have to make up room indication of reset. I LE2. Whether either upon racking the breaker, as well as cell switch the oserspeed trip or 1TV trip can be reset damage can occur upon temoval and reinstallation without tesetting the other, and unambiguity of of the breaker.)
control room and hical indication of TIV posi-tion and overspeed trip linkage reset status, all , Mispositioning of handswitches and procedural affect the likelihood of these errors. DE3. deficiencies have prevented automatic pump start. IIE3. Inspection Suggestion Caref ully inspect the TI'V overspeed trip linkage and assure it is reset amt in g,.tlon Suggestion Confirm switch position good physical condition. Assure that there is a good using'Ihble 3.1. Review administrative procedures steam isolation to the :urbine, otherv,ise cor.tinued concerning documentation of procedural turbine high temperature can result in degradation i 3.3 NU REG /CR.w31.
1 I Inspection Guidance -{ l deficiencies. I'nsure operator training en broken and dirty cimtacts, misaligned or broken ! proceduralchanges is current. limit switches, ctmtrol power loss, and calibration proolems. Degraded operation has also resulted 3.1.4 Pump Unnvallable Due to Maisteenance frora improper air preuure due to air regulator or Surveillance failure orleaking air hnes.
- Both schulukd and un3chedt. led maintenance Inslation Sugestion Check for control air system remove pumps from operability. Surveillance alignment and air leaks during plant walkdowns, requires operation with an altered line-up, (Regulators may have a small amount of external although a pump train may not be declared bleed to maintain downstream pressure.) Check for inoperable during testing. Prompt scheduling cleanliness and physical condition of visible circuit and performance of maintenance and surveil- clements. Review valve stroke time surveillance for lance minimize this unavailability, adverse trends, especially those valves on reduced testirg frequency. Review air system surveillances In pection Sugestlos. . Review the time the AFW moisture mntent of air is within established limits.
Systern and components arc inoperabic. Assure all naintenance is being performed that can be per-Out of. adjustment electrical flow controllers have formed during a single outage time frame, avoiding caused improper valve operation, af fecting multi-multiple equipment outages. The maintenance ple trains of AISV CCl2 should be scheduled before the routinc surveillance test, so credit can be taken for both post main. Inspecti n Sugestion - Resiew PM frequency and tenance testing and surveillance testing, avoiding records,only u}xm a trend of failure of the excenive testing. Review surveillance schedule for controllers. ficquency and adequacy to verify system operability requirements per Technical Specifications. leakage of hot feedwater through check va!ves ha.* caused thermal binding of flow control MOVs.
- liigh frequency vibration has caused wcld cracking AOVs may be similarly susca ptible. CF2.
at a llange junction to a flow element at Comanche Peak. Inspection Sunestion - Covered by 3.1.1 bullet 3. 3.1.5 Air Opernted Flow Control Valves Fall In dequate torque adjustinent on an air operated sicam supply valve txmnet caused steam to leak Closed through the valve at Comanche Peak. TQPumo'Itaim liv-245R4Nk2461.2462 Inspection Sugestion . Review the AOV test MD PumnTrain: PV2453A.24531(2454A.245411 records to assure the testing and settings are based on dynamic system con'"tions. Overtorquing of the lhese normally closed air operated valves (AOW) valve operator can result in valve damage such as control flow to the steam generators. They fall open on cracking of the seat or disc. Review the program to loss ofinstrument air. assure overtorquing is identified and cntrective
+ actions are taken to assure valve operability follow.
C<mtrol circuit problems have been a primary cause ing an overtorque camdition. Review the p'rogram
- of failures, both at Comanche Peak and elsew her' to assure EO seats are renewed as required during Cl% Valve failures have resulted from blown the restoration from testing to maintain the EO fuses, failure of control components (such as ratingof the AOV.
l- current / pneumatic convertors),diaphram failures. NUREG/CR-5831 3.4 lt
inspection Guidance 4 3.1.6 Motor Operated Isolation Valves rail
- Orcase trapped in the torque switch spring psch Clor,ed of 1.imitorque SMll motor operators has caused motor burnout or thermal oveskiad trip by pre-venting torque switch actuation. CRt.
MD Pump Discharre Isolatiorg IIV-249111.249211.2404A.2493A
'1D Pump Discharce isolation: Insgation Sugestion Resiew this only if the MOV testb ig program reveals deficiencies in this llV.249fA.2492A.249411.249311 area.
Servige Water Suction isolation; H V-2395.2196.2480.248 f.2482
- Manually revening the direction of motion of
%ese MOVs isolate flow to the steam generators and operating MO% hss ove. oaded the motor cir-cult. Operating prondures should provide provide AFW pump suction isolation. The discharge ,
cautions, and circuit designs may prevent isolation valves and CST suction valves are normally reversal before cach stroke is finished. DE7. open and the nuclear service water suction valves ate normally closed. They all fall as.is on loss of power. InspectionSugestion Reviewoperating procedures to ensure adequate precautions are Common cause failure of MOVs has resultea included to prevent reversing the direction of from failure to use electrical signature tracing motion during valve operation. equipment to determine proper settings of torque switch and torque switch bypass switches. Pailure to calibrate switch settings for Space heaters designed for preoperation storage have been found wired in parallcl with valve high torques necessary under design basis acci, motors which had not been environmentallj dent c4mdltions has also been involved. CCl1. qualified with them present. DE8. Inspection Sugestion Resiew the MOV test inspection Sugestion Spot check MOV's during records to assure the testing and settings are based MOV testing to assure the space heaters are on d.vnamic system ctmditions. Overtorquing of the physically removed or disannected. valve operator can result in valve damage such as crackir g of the seat or disc. Rniew the progran. .o 3.1.7 Manual Suction or Discharge Valves Fall assure overtorquing is identified and corrective actions are taken to asmic valve operability fol- Closed lowing an overtorque amdition. Review the pro. gram to assute EQ seals are renewed as required TD Pump 'Itain: during the restoration from testing to maintain the AF 31:41 77.85.97.105.122.l?4.126.128 EQ rating of the MOV. MD pump '!1ains.J.J; AF- 13.23 :54.W74.82.92. l GO.121.123.125.127
- Valve motors have been failed due to lack of, or improper strinr,or use of thermal overload These manual valves are normally locked open. For protective devica ~ passing and oversiring each train, closure of the first valves would bh>ck pump shculd be bass , ver engineering for suction, closure of the second valves would block pump dysignbas.jp 4 :11 ions. CF4. discharge and ch>sure of the third set of valves would hh>ck discharge to steam generators 1,2,3, and 4 '
Inspection Sugestion - Rniew Ihe administrative respectively. controls for documenting and changing the settings of thernialoverk)ad protective desiecs. Assure the informatien is available to the maintenance planners. 3.5 NUREG/CR.5831
Inspection Guidance
- Wlve mispositioning has resulted ia failures of 3.1.8 Leakage of Hot Feedwater through mu!tiple trains of AFW. CC2. It has also been Check Valves the dominant ause of problems identified dur-ing operational readiness inspections. HEl. MD Pump 1 Dain: AF-65.75.83.15U97.200.201 Events have occurred most often durin; e.iainte- MD Pump 2 hain: A7-51.93.101.195.198.199.202 nanee, e dibralion, or system modificatnor ' TD PumriVain: AF-38.78.86.98,106 Irnporta.;t causes of mispositioning include:
On two omasions at Commanche Peak failure of Paihire to provide complete, clear, and specific check valves to properly seat resulted in backflow of procedures for tasks and system restoration hot feedwater from one or more steam generators to
- the condensate storage tank during system pre-Failure to promptly revise and validate pro- operational testing. One of the events resulted in cedures, training, and diagrams following system water hammer damage to a piping support.
modifications Leakage of hot feedwatcr through several check Failure to complete all steps in a procedure valves in series has caused steam binding of multiple pumps. la' age through a closed level Failure to adequately teview uncompleted control valve in series with check valvr- has also - procedural steps aften task completion occurred ris would be required for leakage to reac:e the motor driven or turbine driven Failure to verify support functions after pumps. CC10. restoration
+
Failure to adhere scrupulously to administrative procedures regarding ti.;. 'rol and - Slow leakage past tt Sal check valve of a tracking of valve operations series may not f' /@ check valve closed. Other check valves it ries may leak similarly. Pailure to log the manipulation of scaled valves Piping orientation and valve design are important factors in achieving true series
+
Failure to follow good practices of written task protection. CFl. assignment and feedback of task completion informatior. Inmeetion Suggestion - Covered by 3.1.1 bullet 3.
+
Failure to provide casily icad system drawings, legible valve labels corresponding to drawings and procedures, and labeled indications of local 3.2 Risk ImEortant AFW S) stem valve position Walkdown "Ihble Inspection Suggestion - Review the administrative Table 3.1 presents an AFW system walkdown table controls tha' relate to valve [v itioning and scaling, including only components identified as risk important.
- j. system restoration following mainienance, valve This information allows irapectors to concentrate their labeling, system drawing updating, and procedure efforts on components important to prevention of core l revision, for proper implementation. damage. However,it is essential to note that inspec-1-
tions should not focus exclusively on these components. I Other components which perform essential functions, must also be addressed to ensure that their risk impor-tances are not increased. Examples include the (open) steam lead isolation valves MS 101 and MS 128,and an adequate water levelin the CST. NT IREG/CR-5831 3.6 l l
Inspection Guidance Table 3.1 Hisk importance ADV systern walkdown table for north anna AlW system components Component p;. ~ ' Actual Number Component Name Pt . rf , sed Position Electrical A Motor Driven Pump Racked in/ Closed 11 Motor Driven Pump Racked In! Closed
%1ve AF(WWi CST to TDAITV Pump Isolation tecked Open AF (XX17 CST to MDARY Pumps 1 & 2 locked Open AF (X)13 CST to MDAFW Pump 1 suction isol Locked Open AF 0023 CST to MDAITV Pump 2 Suctica lsol Locked Open AF m31 CST to TDAFW Pump Suction isol Locked Open SW 4395 SW to AFY/ Suction isolation Closed SW 4396 SW to AFW Suction isolatior. Closed AF (X)19 SW to MDAFW Pump 1 Suction isol Closed AF m22 SW to MDAFW Pump 2 Suction Isol Closed AF (K)30 SW to TDAFW Pump Suction isol Closed AF (KMi TL)AFW Pump Discharge isol locked Open AF(X42 TDAFW Pump Rst isol Closed AF(X44 TDAFW Pump Dwnstrm Recirc isol Imcked Open AF (X46 TDAFW Pump Upstream Recire Isol locked Open AF (X)54 MDAFW Pump 2 Discharge isol Locked Open 3.7 NUREG!CR 5831 I
inspection Ouidance Thble 3.1 (Continued) Component Required Actual . Number Component Name Position losed Position AF0055 MDAIAV Pump 2'Ibst isol Closed-AF 0056 MDAIN Pump 2 Dwnstrm Recirc Isol locked Open AF(X158 MDAFW ? ump 2 Upstream Recirc Isol locked Open FV 2457 MDAIAV Pump 2 Recire Flow Control Opcn i AF(XWi MDAFW Pump 1 Discharge Isol Locked Open 1 AF 0067 MDAFW Pump 1 lbst isol Closed _ l AF(X68 MDAFW Pump 1 Dwnstrm Recirc isol Locked Open AF 0070 MDAFW Pump 1 Upstream Recirc Isol locked Open IA'2456 MDAFW Pump i Recite Flow Control Open AF 0074 MDAFW Pump 1 Discharge to SG 1 Upstream Isol Locked Open __ PV 2453A MDAFW Pump I to SG 1 Flow Control Open AF 0121 MDAFW Pump i Discharge to SG 1 Downstream isol locked Open HV 2491B MDAFW Pump 1 to SG 1 Containment Isol Open e AF(K182 MDAFW Pump 1 Discharge to SG 2 Upstream isol Locked Open PV 2453B MDAFW Pump I to SG 2 Flow Control Open AF 0123 MDAFW Pump 1 Discharge to SG 2 Downstream Iso Locked Open HV 2492L MDAFW Pump 1 to SG 2 Containment Isol Open AF(XYA) MDAFW Pump 1 Cross-Tie 601 Closed NUREG/CR-5831 3.8
.~. .. . .- . - .- . . - . - ~ .- -. -. -
~. -
- Inspection. Guidance hble 3.1 (Continued)
Required Actual Component - Component Name . Position losed Position - Numter AF0091 MDAFW Pump 2 Cross-Tie Isol Closed AF 0092 MDAFW Pump 2 Discharge to SG 3 Upstream Isol Locked Open PV 2454A MDAFW Pump 2 to SG 3 Flow Control Open AF 0125 MDAFW Pump 2 Discharge to SG 3 l Downstream isol Imcked Open HV 2493A MDAFW Pump 2 to SG 3 Containment Isol Open AF 0100 MDAFW Pump 2 Discharge to SG 4 Upstream Isol locked Open PV 2454B MDAFW Pump 2 to SG 4 Flow Control Open AF 0127 MDAFW Pump 2 Discharge to SG 4 Downstream Isol Locked Open HV 2494A MDAFW Pump 2 to SG 4 Containment Isol Open j- AF 0077 TDAl~W Pump Dishearge to SG 1 Upstream 1s01 : Imcked Open L l FIV 2459 TDAFW Pump to SG 1 Flow Control Open
'AF 0122 - TDAFW Pump Discharge to SG 1 Downstream Isol - lacked Open .
l i IIV 2491 A TDAF%' Pump to SG 1 Containment l- Open
- Isol AF (KE5 TDAFW Pump Discharge to SG 2 Upstream Isol locked Open ilV 24N) TDAFW Pump to SG 2 Flow Control Open 3.9 NOREG/CR-5831 i
inspection Guidance l. Table 3.1 - (Continued) - l~ Component Rettuired Actual Number Component Name Position losed Position - AF 0124 TDAFW Pump Discharge to SG 2 Downstream Isol Locked Open iIV 2492A TDAFW Pump to SG 2 Flow Ccmt'rol Open - AF0097 TDAFW Pump Discharge to SO 3 Upstream isol locked Opn I HV 2461 TDAFW Pump to SG 3 Flow Control Open .j l AF 0126 TDAFW Pump Discharge to SG 3 Downstream isol 1.ncked Open 1IV 24930 TDAFW Pump to SG 3 Flow Control Open AF 0105 TDAFW Pump Discharge to SG 4 Upstream Isol- locked Open HV 2462 TDAFW Pump to SG 4 Flow Control Open AF 0128 TDAFW Pump Discharge to SO 4 Downstream Isol. 1.acked Open llV 2494B TDAFW Pump to SG 4 Flow Control Open MS 0101 HV 2452-1 Upstream isol Open _ _ . IIV 2452-1 SG 1 Steam Supply to TDAFW Pump Closed MS 0144 ilV 2452 1 Downstream Isol Open MS0711 IIV 2452-1 Bypass ' Locked Closed MS 0128 ilV 2452-1 Upstream Isol Open p HV 2452-2 SG 4 Steam Supply to TDAFW Pump Closed I L MS 0137 HV 2452-2 Downstream isol - Open l NUREG/CR-5831. 3.10 W e-
4 Generic Itisk insights From pitas PRAs for 13 PWRs were analy/cd to identify risk- 4,l.3 laiss of Main Feedwater important accident sequences involving loss of AFW, to identify and risk-prioritize the component failure modes
- A feedwater line break drains the common water involved. The resuiis of this analysis are described in soutcc for MFW and AFW. The operators fail to this section. They are consistent with results reported provide feedwater from other sources, and fail to by INEL and BNL (Gregg et al 1988, and flavis et al, initiate feed.and bleed cooling, resulting in core 198S). damage.
4.1 Itisk Important Accident Sequences
- A loss of main feedwater trips the plant, and AFW IInvolving AFW System Failure fails due to operator enor and hardware failures.
The operators fail in initiate feed-and-bleed cooling,
"" " ""' """ E
4.1.1 loss of Power System A loss of o!Tsite power is followed by failure of AFW. Due to lack of actuating power, the power , g.,.it s followed by failure of AFW. Coolant is operated relief valves (PORVs) cannot be opened ont ik pdntary untd tk refudng water preventing adequate feed and-bleed cooling, and resulting in core damage. 9.nage tan (RW,ST)is depleted. High pressure injection (llPI) fails since recirculation cannot be esta e mm the empty sump, and me damage A station blackout fails all AC power except Vital AC from DC invertors, and all decay heat removal systems except the turbine-driven AFW pump. AFW subsequently fails due to battery depletion or hardware failures resulting in core damage, 4.2 Itisk Important Component Failure Modes A nC bus falls causing a trip and failure of the power conversion system. One AFW motor-driven The generic component failure modes identified from pump is failed by the bua loss, and the turbine- PRA analyses as important to AFW spiem failure are driven pump fails due to loss of turbine or valve listed below in decreasing order of risk importance, control power. AFW is subsequently lost com-pletely due to other failures. Feed-and bleed 1. Thrbine-Driven Pump Failure or Start or Run. cooling fails because PORV control is lost, resulting in core damage. 2. Motor Driven Pump Failure to Start or Run. 4.1.2 Transient-Caused Reactor or'Ibrbine 3. TDP or MDP Unavailable due to Tbst or
'IYip Maintenance-A transient-caused trip is followed by a loss of the 4. AFW System Valve Riilures power conversion system (PCS) and AFW. Feed-and-bleed cooling fails either due to failure of the
- steam admission valves operator to initiate it, or due to hardware failures, resulting in core damage.
- trip and throttle valves 4.1 NUREG/CR-5831
Generic Risk flow mntrol valves
- suction valves.
=
pump discharge valves in addition to indhidual hardware, circuit, or ! instrument failures, each of these failuremodes may .
- l. .
pump suction valves result from common causes and human errors. Com-mon cause failures of AFW pumps are particularly risk ---
- valves in testing or maintenance, important Valve failures are somewhat less important due to the multiplicity of steam generators and mn-
- 5. Supply / Suction Sources - nection paths.11uman errors of greatest risk impor-tance involve: failures to initiate or control system -
crmdensate storage tank stop valve operation when required; failure to restore proper system lineup after maintenance or testing; and failure l hot wellinventory to switch to alternate sources when required. 7 Y f I -: l-l l NUREG/CR-5831 4.2
. , . ~_
5 Failure Modes Determined Froni Operating Experience This section describes the primary root cause of AFW auxiliary feedwater system to the condensate storage system component failures, as determined from a review tank. This flow caused abnormally high temperatures, of operating histories at Comanche Peak and at other thermal stresses exceeding values allowed by code, and PWRs throughout the nuclear industry. Section 5.1 des- damage to piping supports (requiring replacement of cribes experience at Comanche Peak. Section 5.2 sum- one support). Check valve failure was attributed to the marizes information compiled from a variety of NRC improper assembly due to an incorrect vender reassem-sources, including AEOD analyses and reports,infor- bly procedure. Subsequently the procedure was tur-mation notices, inspection and enfortement bulletins, rected and training in procedures and awareness of and generic letters, and from a variety of INPO reports operating events and equipment failures was completed, as well. Some 1.icenwe Event Reports and NPRDS 11aining emphasired the transition from construction to event descriptions were also reviewed Finally,infor- operating aititudes. mation was included from reports of NRC-sponsored studies of the effects of plant aging,which include quantitative analysh of AIM system failure reports. 5.2 Industry Wide Experience This information was used to identify the various root causes expected for the broad PRA based failure events lluman errors, design / engineering problems and errors, ideritified in Section 4.0, resulting in the inspection and component failures arc the primary root causes of guidelines presented in Section 3.0. AFW System failures identified in a review of ind istry wide system operating history. Common cause failures, which dhable more than one train of this operationally 5,1 Cmnanche Peak Experience redundant system, are highly risk sigrjficant, and can result from all of these causes, Comanche Peak started operation in 1990 and during the remainder of that year,only three AFW system re- This section identifies important common cause failure ! lated failure esents were reported. These include, leak- modes, and then provides a broader discussion of the age of a seal on a motor-operated isolation valve, steam single failure effects of human errors, design / supply valve leak through, and a weld crack at a flange engineering problems and errors. and component fail-junction on a high pressure tap to a flow element. Fail- ures. Paragraphs presenting details of thete failure urcs were attributed to insulficient torque of valve bon- modes are coded (e.g., CCl) and cross referenced by net fasteners and high frequency vibration, respectively. inspection items in Section 3. Also a failed diode in the solid state protection system caused a safety injection system actuation during which 5.2.1 Common Cause Faillires the AFW system operated properly. The dominant cause of AFW system multiple train fail. During preoperational testing of the AFW system in ures has been human error. Design / engineering errors April and May of 1989 (prior to fuel loading) partial and component failures have been less frequent, but blowdown of the SGs occurred resulting in water- nevertheless significant,causes of multiple train failures. hammer events. The cause was attributed to operator error in valve operation combined with malfunction of CCl. Iluman error in the form ofincorrect operator check valves (due to improper assembly). Contrary to intervention into automatic AFWsystem functioning proecdures, operators opened a test line isolation valve during transients resulted in the temporary loss of all before closing the pump dhcharge valve while conduct- - safety grade AFW pumps during events at Davis Besse ing an auxiliary feedwater system operability test. Fail. (NUREG ll54,1985) and 11ojan (AEOD/r416,1983). ure of the check valves to seat allowed backflow of high in the Davis Besse event, improper manual initiation of temperature water from the steam generator though the the steam and feedwater rupture control system 5.1 NUREG/CR-5831
Failure Modes l l l (SFRCS) led to overspeed tripping of both turbine- main feedwater. At Zion.2, restart of both motor drisen 1 driven AFW pumps, probably due to the introduction of pumps was blocked by circuit failure to deenergize when condensate into the AFW turbines from the long, un. the pumps had been tripped with an automatic start sig. heated steam supply lines. (The system had never been nal present (IN 82-01,1982). In addition, AFW control tested with the abnormal, cross. connected steam supply circuit design reviews at Salem and Indian Point have lineup which resulted.) In the Trojan event the operator identified designs where failures of a single component l- incorrectly stopped both AFW pumps due to misinter- could have failed all or multiple pumps (IN 87 34, pretaiton of MFW pump speed indication. The diesel 1987). At Indian Point 2 it was found necessary to block driven pump would not restart due to a protective fea- closed the 12 inch valve between the CST and conden-ture requiring complete shutdown, and the turbine- ser, and to utilize a 4 inch valve to ensure adequate driven pump tripped on overspeed, requiring local reset AFW pump suction pressure when a vacuum exists in of the trip and throttle valve. In cases where manual the condenser, intervention is required during the early stages of a transient, training should emphasire that actions should CC5. Incorrect setpoints and c(mtrol circuit settings be performed methodically and deliberately to guard resulting from analysis errors and failures to update . against such errors. procedures have also prevented pump start and caused pumps to trip spuriously. Errors of this type may CC2. Wlve mispositioning has accounted for a signifi- remain undetected despite surveillance testing, unless cant fraction of the human errors failing multiple trains surveillance tests model all types of system initiation of AFW. This includes closure of normally open suction and operating conditions. A greater fraction ofinstru-valves or steam supply valves, and of isolation valves to mentation and control circuit problems has been identi-sensors having control functions. Incorrect handswitch fied during actual system operation (as opposed to sur-positioning and inadequate temporary wiring changes veillance testing) than for other types of failures. have also prevented automatic starts of multiple pumps. Factors identified in studies of mispositioning errors CC6. On two occasions at a foreign plant, failure of a include failure to add newly installed valves to valve balance-of-plant inverter caused failure of two AFW ; checklists, weak administrative control of tagging, res. pumps, in addition to loss of the motor driven pump toration, independent verification, and locked valve whose auxiliary start relay was powered by the invertor, logging, and inadequate adherence to procedures. lileg. the turbine driven pump tripped on overspeed because ible or confusing local valve labeling, and insufficient the governor valve opened, allowing full steam flow to training in the determination of valve position may the turbis. This illustrates the importance of assessing cause or mask mispositioning, and surveillance which the effects of failures of balance of plant equipment does not exercise complete system functioning may not which supports the operation of critical components. reveal mispositionings. The instrument air system is another example of such a system. CC3. At ANO-2, both AFW pumps lost suction due to steam binding when they were lined up tc both the CST CC7. Multiple AFW pump trips have occurred at and the hot startup/ blowdown demineralizer effluent Millstone-3, Cook 1,'Irojan and Zion-2 (IN 87-53, (. .EOD/C404,1984). At Zion-1 steam created by runn- 1987) caused by brief, low pressure oscillations of suc-ing the turbine-driven pump deadheaded for one minute tion pressure during pump startup . These oscillations caused trip of a motor-driven pump sharing the same occurred despite the availability of adequate static inlet header, as well as damage to the turbine. driven NPSH. Corrective actions taken include: extending the pump (Region 3 Morning Report,1/17/90). Both events time delay associated with the low pressure trip, remov-were caused by procedural inadequacies. ing the trip,and replacing the trip with an alarm and operator action. CC4. Design / engineering errors have accounted for a smaller, but dgnificant fraction of common cause fail- CC8. Design errors discovered during AFW system re-ures. Problems with control circuit design modifications analysis at the Robinson plant (IN 89-30,1989) and at I- t Farley defeated AFW pump auto-start on loss of Millstone-1 resulted in the supply header from the CST NUREG/CR-5831 5.2
Failure Modes at Davis Besse, the normally-open AFW isolation valves being too small to provide adequate NPSH to the pumps if more than one of the three pumps were oper- failed to open ab r they were inadvertently closed. The ating at rated flow c(mditions. This could lead to mul- failure was due to improper setting of the torque switch tiple pump failure due to cavitation. Subsequent bypass switch,which prevents motor trip on the high reviews at Robinson identified a loss of feedwater tran- torque required to unscat a closed valve. Previous prob-lems with these valves had been addressed by increasing sient in which inadequate NPSH and Dows less than de-sign values had occurred, to. which were not recognied the torque switch trip setpoint - a fix which failed during at the time. Event analysis and equiptnent trending, as the event due to the higher torque required due to high well as surveillance testing which duplicates scrvice con- differential pressure across the valve. Similar common mode failures of MOVs have also occurred in other sp-ditions as much as is practical, can help identify such tems, resulting in issuance of Generic Letter 89-10, design errors.
" Safety Related Motor-Operated Valve 'Ibsting and Sur.
CC9. Asiatic clams caused failure of two AFW flow veillance (Partlow,1989)." This generic letter requires control valves at Catawba-2 when low suction pressure licensees to develop and implement a program to pro-caused by starting of a motor-driven pump caused suc. vide for the testing, inspection and maintenance of all tion source realignment to the Nuclear Service Water safety-related MOVs to provide as urance that they will system. Pipes had not been routinely treated to inhibit function when subjected to design basis conditions. clam growth, nor regularly monitored to detect their presence, and no strainers were installed. The need for CCl2. Other component failures have also resulted in surveillance which exercises alternative system opera- AFW multi-train failures. These include out-of-tional modes, as well as complete system functioning, is adjustment electrical flow controllers resulting in im-cmphasized by this event. Spurious suction switchover proper discharge valve operation, and a failure of oil H has also occurred at Callaway and at McGuire, although cooler cooling water supply valves to open due to silt W accumulation. F no failures resulted. O q CC10. Common cause failures have also been caused by 5,2.2 Iluman Errors i F- component failures (AEOD/C404,1984). At Surry-2, i both the turbine driven pump and one motor driven HEl. The overwhelmingly dominant cause of problems pump were declared inoperable due to steam binding identified during a series of operational readiness caused by backleakage of hot water through multiple evaluations of AFW systems was human performance, check valves. At Robinson 2 both motor driven pumps The majority of these human performance problems re-were found to be hot, and both motor and steam driven suited from i, complete and incorrect procedures, par-ticularly with respect to valve lineup information. A pumps were found to be inoperable at different times. Backleakage at Robinson 2 passed throagh closed study of valve mispositioning events involving human motor-operated isolation valves in addition to multiple error identified failures in administrative c(mtrol of check valves. At Parley, both motor and turbine driven tagging and tagging, procedural compliance and comple-pump casings were found hot, although the pumps were tion of steps, verification of support systems, and in-not declared inoperable. In addition to multi-train adequate procedures as important. Another study found failuies, numerous incidents of single train failures have that valve mispositioning events occurred most often occurred, resulting in the designation of' Steam Binding during maintenance, calibration, or modification activi-of Auxiliary Feedwater Pumps" as Generic issue 93- ties. Insufficient training in determining valve position, This generic issue was resolved by Generic Letter 88-03 and in administrative requirements for controlling valve (Miraglia,1988),which required licensees to monitor positioning were important causes, as was oral task as-AFW piping temperatures each shift, and to maintain signment without task completion feedback. procedures for recognizing steam binding and for restor-ing system opeability- HE2. 'Ibrbine driven pump failures have been caused by human errors in calibrating or adjusting governor speed CQJ1, Common cause failures have also failed motor control, poor governor maintenance, incorrect operated valves. During the total loss of feedwater event 5.3 NUREG/CR-5831 I
- Failure Modes adjustment of governor valve and overspeed trip link- and the mechanism may lack labels indicating when it is ages, and errors associated with the trip and throttle in the tripped position (AEOD/C602,1986).
valve. TTV associated errors include physically bump-ing it, failure to restore it to the correct position after DE4. Startup of turbines with Woodward Model PG-testing, and failures to verify control room indication of PL governors within 30 minutes of shutdown has resul. TTV position following actuation, ted in overspeed trips when the speed setting knob was not exercised locally to drain oil from the speed setting HE3. Motor driven pumps have been failed by human cylinder. Speed control is based on startup with an errors in inispositioning handswitches, and by procedure empty cylinder. Problems have involved turbine rota-deficiencies. tion due to both procedure violations and leaking steam. 1btry has marketed two types of dump valves for auto-5.2.3 Design! Engineering Problems and matically draining the oil efter shutdown (AEOD/C602, Errors 1986). del. .As noted above, the majority of AFW subsystem At Calvert Cliffs, a 1987 loss-of-offsite-power event failures, and the greatest relative system dcgradation, required a quick, cold startup that resulted in turbine has been found to result from turbine-driven pump fail, trip due to PG PL governor stability problems. The utes. Overspeed trips of Tbtry turbines controlled by short term corrective action was installation of stiffer Woodward governors have been a significant source of buffer springs (IN 88-09,1988). Surveillance had alwap , these failures (AEOD/C602,1986). In many cases these been preceded by turbine warmup, which illustrates the overspeed trips have been caused by slow response of a importance of testing which duplicates service condi-Woodward Model EO governor on startup,at plants tions as much as is practical, where full steam flow is allowed immediately. This over-sensitivity has been removed by installing a startup DEL Reduced viscosity of gear box oil heated by prior steam bypass valve which opens first, allowing a con. operation caused failure of a motor driven pump to start trolled turbine acceleration and buildup of oil pressure due to insufficient tube oil pressure. Lowering the pres-to control the governor valve when full steam flow is sure switch setpoint solved the problem, which had not admitted. been detected during testing.1 DE2. Overspeed trips of Terry turbines have been DE6. Waterhammer at Palisades resulted in AFW line cauxd by condensate in the steam supply lines. Con- and hanger damage at both steam generators. The AFW > densate slows down the turbinc, causing the governor spargers are located at the normal steam generator level, valve to open farther, and overspeed results before the and are frequently covered and uncovered during level ~ governor valve can respond, after the water slug clears. fluctuations. %bterhammers in top feed ringsteam This was determined to be the cause of the loss of-all- generators resulted in main feedline rupture at Maine AFW event at Davis Besse ( AEOD/602,1986), with Yankee and feedwater pipe cracking at Indian Point 2 condensation enhanced due to the long length of the (IN 84-32,1984). cross-connected steam lines. Repeated tests following a - cold-start trip may be successful due to system heat up. DE7. Manually reversing the direction of motion of an operating valve has resulted in MOV failures wherc i DE3. Tbrbine trip and throttle valve (TTV) problems such loading was not considered in the design (AEOD/_ are a significant cause of turbine driven pump failures C603,1986), Control circuit design may prevent this,- (IN 84-66). In some cases lack of TTV position indica, requiring stroke completion before reversal. tion in the control room prevented recognition 01 a trip-ped TTV. In other cases it was possible to reset either DE8. At each of the units of the South 1bxas Project, the overspeed trip or the TTV without reseting the space heaters provided by the vendor for use in pre-other. This problem is compounded by the fact that the installation storage of MOVs were found to be wired in position of the overspeed trip linkage can be misleading, NUREGICR-5831 14
Failure Modes parallel to the Class IE 125 V DC motors for several injection valves resulting frota check valve ic1kage has AFW valves (lR 50-48939-11; 50-499/89-11,1989). 'lhe prevented MOV operation (AEOD/CI43,1986). valves had been environmentally qualified, but nat with the non-safety related heaters energized. CF3. Gross check valve leakage at McGuire and Robinson caused overpressurization of the AFWsuc-5.2.4 Component Failures tion piping. At a foreign PWR lt resulted in a severe waterhammer event. At Palo Verde-2 the MFW suction Generic issue ll.E.6.1, *In Situ Testing Of Valves" was pipingwas overpressurized by check valve leakage from divided into four sub-issues (Beckjord,1989), three of the AFW system (AEOD/C404,1984). Gross check valve leakage through idle pumps represents a potential which relate directly to prevention of AFW system diversion of AFW pump flow. component failure. At the request of the NRC,in. situ testing of check valves was addressed by the nuclear CF4. Roughly one third of ACW system failures have industry, resulting in the EPRI report," Application been due to valve operator failures, with about : qual Guidelines for Check Valves in Nuclear Power Plants (Brooks,1988)." This extensive report provides infor. failures for MOVs and AOVs. Almost hairof the MOV failwes were due to motor or switch failures (Cesada, mation on check valve applications, limitations, and 1989). An extensive study of MOV events (AEOD/ inspection techniques, in-situ testing of MOVs was ad. C603,1986) indicates continuing inoperability problems dressed by Generic letter 89-10
- Safety Related Motor.
caused by: torque switch!!imit switch settings, adjust. Operated Valve 'Rsting and Surveillance * (Partlow, ments, or failures; motor burnout; improper sizing or 19S9) which requires licensees to develop and imple. use of thermal overload devices; premature degradation ment a program for testing, inspection and maintenance related to inadequate use of protectise devices; damage of allsafety-related MOVs. " Thermal Overload Protec. tion for Electric Motors on Safety Related Motor. due to misuse (valve throttling, valve operator hammer. Operated Valves - Generic issue ll.E.6.1 (Rothberg, ing); mechanical problems (kic;ened parts, improper 1988)* conclude, tna; valve motors should be thermally assembly); or the torque switch bypass circuit impro-perly installed or adjusted. The study concluded that protected, ye* in a way which emphasizes system func. current methods and procedures at many plants are not tion owr protection of the operator. adequate to assure that MOVs will operate when needed under credible accident conditions. Specifically, CJil. The common-cause steam binding effects of check a surveillance test which the valve passed might result in valve leakage were identified in Section 5.2.1, entry CC10. Numerous single-train events provide additional undetected valve inoperability due to cornponent failure insights into this problem. In some cases leakage of hot (motor burnout, operator parts failure, stem disc sepa. MFW past multiple check valves in series has occurred ration) or improper positioning of protective devices because adequate valve-seating pressure was limited to (thermal overload, torque switch, limit switch). Generic the valves closest to the steam generators ( AEOD!C404, Letter 89-10 (Partlow,1989) has subsequently required 1984). At Robinson, the pumn shutdown procedure was licensees to implement a program ensuring that MOV switch settings are maintained so that the valves will changed to delay closing the MOVs until after the check valves were seated. At Farley, check valves were operate under design basis conditions for the life of the changed from swing type to litt type. Check valve re. plant. work has been donc at a number of plants. Different valve designs and manuh.cturers are involved in this CFS. Component problems have caused a sio nificant number of turbine driven pump trips (AEOD/C602, problem, and recurring leakage has been experienced, even after repair and replacement. 1986). One group oievents involved worn tappet nut faces, loose cable connections, loosened set screws, CF2. At Robinson, heating of motor operated valves by improperly latched TTVs, and improper assembly. check valve leakage has caused thermal binding and fail. Another involved oilleaks due to component or seal ure of AFW discharge valves to open on demand. At failures, and oil contamination due to poor maintenance Davis Besse, high differential pressure across AFW activities. Governor oil may not be shared with turbine 5.5 N UREG/CR-5831
Failure Modes q lubrication oil, resulting in the need for separate oil relief kits are needed for MOV operators manufactured changes. Electrical component failures included tran- before 1975. At Limerick, additional grase relief was sistor or resistor failures due to moisture intrusion, required for MOVs manufactured since 1975. MOV re-crroneous grounds and connections, diode failures, and Iurbishment programs may yield other changeovers to , a faulty circuit card. EP-0 grease. ED. Electrohydraulic-operated discharge valves have CF9. For AFW systems using air operated valves, al-performed very poorly, and three of the five units using most half of the system degradatlan has resulted from them have removed them due to recurrerst failures. failures of the valve controller circuit and its instrument Failures included oil leaks, contaminated oil, and inputs (Casada,1989). Failures occurred predominantly hydraube pump failures. at a few units using automatic electronic controllers for the flow coatrol valves, with the majority of failures duc CF7. Control circuit failures were the dominant source to electrical hardware. At 'Ibrkey Point 3, controller of motor driven AFW pump failures (Casada.1989). malfunction resulted from water in the Instrument Air This includes the controls used for automatic and man- system due to maintenance inoperability of the air ual starting of the puups, as opposed to the instru- dryers. mentation inputs. Most of the remaining problems were due to circuit breaker failures. CF10. For systems using dicsci driven pumps, most of ' the failures were due to start control and governor speed CF8, ' Hydraulic lockup' of Limitorque SMB spring control circuitry. Half of these occurred on demand,as packs has prevented proper spring compression to. opposed to during Icsting (Casada,1989). actuate the MOV torque switch, due to grease trapped h the spring pack. During a surveillance at 'nojan, CF11. For systems using AOVs, operability requires the failure of the torque switch to trip the'lTV motor re- availability of Instrument Alt, backup air, or backup - sulted in tripping of the thermal overload device, leaving nitrogen. However, NRC Maintenance'Ibam inspec-the turbine driven pump inoperable for 40 days until the lions have identified inadequate testing of check valves next surveillance (AEOD/E702,1987), Problems result isolating the safety-related portion of the lA system at from grease changes to EXXON NEBULA EP.0 grease, several utilitics (Letter, Roc to Richardson). Generic one of only twu greases considered environmentally Letter 88-14 (Miraglia,1988), requires licensecs to qualified by Limitorque. Due to lower viscosity,it verify by test that air-operated saf + -related com-
- slowly migrates from the gear case into the spring pack. ponents will perform as expected .. accordance with all Orcase changeover at Vermont Yankee affected 40 of design-basis events, including a loss of normal I A.
the older MOVs of which 32 were safety related. Grease NUREG/CR-5831 56 I ____ _______ ____________a
6 References Beckjord, E. S. June 30,1989. Closcout of Generic Issue AEOD Reports 11 E.b.1, *In Situ ksting of thlves.* lxtter to V. Stello, Jr., U.S. Nuclear Regulatory Commission, %hshington, AEOD/C4G4. W. D.12nning. July 1984. Tream Binding DC. ofAuxiliary Fredwater Pumps. U.S. Nuclear Regulatory Commission, %hshington, DC Brooks,IL P. I988. Application Guidelinesfor Check itdres in Nuclear A>wer Plants. NP-5479, Electric AEODIC602. C. Hsu. August 1986. Operational Power Research Institute, Palo Alto, CA. ErperienceInvolving 7hrbine Overspeed hips. U.S. Nuclear Regulatory Commission, Washington, DC. Casada, D. A.1989. Auriliary Feedwater System Aging Study. Volume L Operating Enperience and Current AEOD!CfD3. E. J. Brown. December 1986. A Review Atonitonng hactices. NUREGlCR-5404. U.S. Nuclcar of Afotor-Operated Udve Performance. U.S. Nucicar Regulatory Commission, %Sshington, DC Regulatory Commission, Wasinington, DC Gregg, R. E. and R. E. Wright.1988. Appendit Review AEOD/E702. E.J. Brown. March 19,1987. Af0VFail-for (kominant Generic Contributors. BLB-31-88. Idaho ute Due to Ilydraulic Lockup From Excessive Grcase in National Engineering Laboratory, Idaho Palls, Idaho. .5pring Pack U.S. Nuclear Regulatory Commission, Whshington, DC Miraglia, E J. February 17,1988. Resolution of Generic Safety Issue 93,
- Steam Binding ofAuxiliary Feedwater AEODfT416. January 22,1983. Loss of ESF Autiliary Pumps"(Generic Letter 88-03). U.S. Nuclear Rcgulatory Feedwater Pump Capabihty at Trojan on January 22, Commission, Washington, DC 1933. U.S. Nuclear Regulatory Commission, Washington, DC Miraglia, E J. August 8,1988. Irutrument Air Supply System Problems Affecting Safety-Related Equipment Information Notices (Generic Letter SS-14). U.S. Nuclear Regulatory Commission, Washington, DC IN 82-01. January 22,1982. Aardiary liedwater Pump Lockout Resultingfrom Westinghouse 1fC2 Switch Circuit Partlow, J. G. Junc 28,1989. Safety.Related Afotor- Afodification. U.S. Nuclear Regulatory Commission, Operated ialve Testing and Surveillance (Ge.neric Letter Mhshington, DC 6%10). U.S. Nuclear Regulatory Commission, Mhshington, DC IN 84-32. E. L Jordan. April 18,1984. Auriliary Feedwater Sparger aml Pipe llangar Ikimage. U.S.
Rothberg, O. June 1988. ThermalOverload Protection Nuclear Regulatory Commission, WSshington, DC. Jirr Electric 31otors on Safiny-Related Afotor-O;wrated Ihives - Generic issue II.E.b.f. NUREG-12% U.S. IN 84 66. August 17,1984. Undetected Unavailability of Nuclear Regulatory Commission, %Sshington, DC. the Turbine-Driven Auriliary Feedwater Train. U.S. Nuclear Regulatory Commission, Washington, DC Travis, R. and J. Taylor. 1989. Develo;>r, en' of Guidancefor Generic, Functionally Onented PRA. Based 1N 87-34. C E. Rossi. July 24,1987. Single F zitures in Team inspectionsfor BWR Plants-Identification of Rik Auriliary Feedwater Systems. U.S. Nuclear Regulatory important Systems, Components and Human Actions. Commission, whshington, DC. TLR-A-3874-TGA Brookhaven National Laboratory. Upton, New York. l 6.1 NUREGICR-58T1 l
- - _ . -.. ..-. -. _ - _ ._.-..- _..-. . . - ,.- ...- .~ -.- ~-_.-:-_ ..
J References IN 87-53. C E. Rossi. October 20,1987. Auriliary inspection Repm1 Iredwatcr Pump 1 rips Resultingfrom Low Suction
. hessure. U S. Nuclear Regulatory Commission, IR 50-489/89-11; 50-499/89-il. May 26,1989. South Mhshington, DC. Texas Project inspection Report. U.S. Nuclear Regula. .
tory Commission, Washington, DC IN 884N. C E. Rossi. March 18,1988, Reduced Rehabihty of Steam.Drisrn Auritiary Feedwater Pumps NUYn nepart , Caused by Instabihty of IWodward PG PL hpe Governors. U.S. Nuclear Regulatory Commission, NUREG-1154.1985. Loss of Afain and Auritiary Feed.
%%shington, DC. water &cnt at the lhwis Besse Plant on June 9,1985, U.S. Nuclear Regulatory; Commission, Washington, DC.
IN 89 30. R. A. A7ua. August 16,1989. Robinson Unit 2 Inadequate NPSil ofAutiliary Feedwater Pumps. Also, Event Notification 16375, August 22,1989. U.S. Nuclear Regulatory Commission, Washington, DC. t 4 4 4 I i NUREG/CR.5831 6.2 i
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I NUREG/CR $831 1 PNL.778'4 Distribution No. of No. of C.RPicS Col's OFFSITE 4 Comanche Peagesident Irpector Office ll.S Nuclear Rerulatory Commission J. H. 'Ihylor B. K Otimes Brookhaven National laboratory OWIN 9 A2 Bldg.13 Upton, NY 11973 E Congcl OWFN 10 E4 R. 'Irwis Brookhaven National Laboratory A. C. Thadani Bldg.130 OWFN 8E2 Upton, NY 11973 A R. John. son R. Gregg OWFN 14 D1 EG&G Idaho, Inc. P.O. Box 1625 G. D. Holahan Idaho Falls,ID 83415 OWTN 8E2 D. R. Edwards S. M. Long Prof. of Nuclear Engineering OWFN 10 A2 University of Missouri - Rolla Rotta, MO 65401 K. Campe OMTH 1 A2 ONSITE 10 J.Chung 23 Pacific Northwest Laboratory OWFN 10 A2 S. R. Doctor R. H. Wessman L R. Dodd OWFN 14 D1 B. E Gore (10) N. E. Maguirc-Moffitt 2 B. Thomas R. C. Lloyd OWFN 12 H26 B. D. Shipp E A. Simonen U.S. Nuclear Rerulatory Commission . T V.Vo Region 4 Publishing Coordination Tbchnical Report File (5) S. J. Collins T. E Westerman L J. Callan A. B. Beach l l Distr.)
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, 2.m a A a suem a Auxiliary feedwater System Risk-Based Inspection Guide for the 3 oAtt at* Cat 'vous to Comanche Peak Nuclear Power Plant e ...a j October 1992
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- b. Av1MoHtSi 6. TYPE OF REPOM1 R. C. Lloyd, N. E. Mof fitt, B. F. Gore, T. V. Vo Technical i.Peaioocoveaio. .. ,
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- 11. AB5T R ACT f200 ==,a, .r ==a In a study sponsored by the U.S. 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 2nd failure mechanisms for the AFW system at the selected plants. Comanche Peak 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 plant, n e.t y wo a osso csca+ r Oa s a., - ., ,,,- ,, _, - ,,. , ,, , ,. , u ...., .. . 1.i... i e Inspection, Risk, PRA, Comanche Peak, Auxiliary Feedwater ( AFW) ,,,, f",I km
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Printed on recycled paper Federal Recyc'ing Program
NUREG/CR-5831 AUXILIARY FEEDWATER SYSTEM RISK-BASED INSPECTION GUIDE FOR OCTOBER 1992 .. Tile COMANCffE PEAK NUCLEAR POWER FIANT l UNITED STATES l NUCLEAR REGULATORY COMMISSION ARST CLASS MAIL WASHINGTOtJ, D.C. 20555-0001 PCSTAGE AND FEES PAfD uSNRC I PEAMfT No. G-67 i ! OFFICIAL BUSINESS FENALTY FOR PRIVATE USE, $300 J
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