ML20116A474

From kanterella
Jump to navigation Jump to search
Emergency Feedwater System RISK-BASED Inspection Guide for the Arkansas Nuclear One Unit 2 Power Plant
ML20116A474
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 09/30/1992
From: Gore B, Pugh R, Vo T
Battelle Memorial Institute, PACIFIC NORTHWEST NATION
To:
Office of Nuclear Reactor Regulation
References
CON-FIN-L-1310 NUREG-CR-5828, PNL-7727, NUDOCS 9210290408
Download: ML20116A474 (31)
v • d • e


Text

p ,

=

q

~

r .- .,

~

- NUREG/CR-5828; PNL-7727 Emergency Feedwater System Risk-Based Inspection Guidezfor the Arkansas Nuclear One Unit 2 Power Plant -!

<l l

[.1$g IIF. Gore, T. V. Vo

. Pacific . Northwest Laboratory ,

L Operated by

Battelle Memorial Institute  ;

} Prepared for _

- U.S. Nuclear Regulatory Commission b AOdOk O! boo pg0 8- e

._.m d

ge .+

a; *

e.  :,

ph n

AVAILABluTY NOTICE Avanability of Reference Materials C,ted in NRC Publications -

Most documents cited in NRC publications wRl be avahable from one of the foHowing sources: ,

1, The NRC PubHc Document P%m, ^120 L Street, NW., Lo+ver Level Washington, DC 70555

~

2. The Superintendem of Documents, U.S. Governrnent PrintinC Office, P.O'. Box. 37082. Washbgton.

DC 20013 7082

3. The National Technical information Service, Springfield, VA 22161 A!though the Beting that follows rer. resents the majority of documents citou in NRC pubucations, it is not Intended to be exhaustive.

Referenced documents 'available for inspection and copying for a fee from tne NRC Fublic Doc'anent Room include NRC correspondence and intomal NRC memoranda; NRC tsulterns, circulars, nr.formatica notices,'

inspection and invoutgntion notices; Heensee event reports; vendor reperts and conospondence; Commis-clon papers; and applicant and Ncensee documents and correspondence.

The following documents in the NURiiG series are avaRable for purchase from the GPO Sales *rogram:

formal NRC staff and contractor reports, NRC sponsored conference proceedings, international agreerrwnt .

j reports, grant publications, and NRC booklets and brochures. Also avaHable are regulatory guides NRC regulatkans in the Code of Federal Regulations, and Nuclear Regulato9 Commhslon Issuances,

~

a-Docu' rents avaNable from the National Tuchnical information Service include NUREGaeries reports and =

technical reports crepared by other Fedet al agencies and reports prepared by tM Atomic Energy Commis-s!on, forerunner agency to the Nuclear Regulatory Commission J Documents avaRable from pubac and specla technical Abraries bclude all open 11terature horns, such as books, journal articles. and transactions. Federal Register noticos, Federal end State legislation, 9d con-grossional reports can usually be obtained from these librarle*, ..

Documents such as thesos, dissertations, foreign reports and translations, and non-NRC confer 3nce pro-ceed;ngs are ave 11able for purchase from the organization sponsoring the publiettlon cited.

Shgle' copies of NRC oraft reports are aveBahle free, to the extent of supply; upon written request to the Office of Administration Distrt>ution and Mall Services se, tion,- U.S. Nuclear 8egulatory Commission, Washington, DC 20555.

Copies of industry codes and standards used in a substantive manner in the NRC reDulatory process are maintained at the t'7C Library 7920 Norfok Avenue, Bethesda; MaryW for use by the public. Codes and utendards are usualy copyrighted and may be purchased frorn the orig!nating organization or. If they are -

American National Strudards, frem the American National Standards institute.1430 Proadway, New York, i NY 1C78 f'

c DibOLAIMdR NOTICE -[

g This reporf was prorated as an accour.1 of work sponsored b, an agency of the Untted States Government <

4 . Neither the United States Govemment nor any agency thew, or any of their employees, makes sny warranty, i

~

expes%f or impiad, or assumes any legM P,abHity of responsibility for any third pany's use, or the results of 4' such use, of any information, apparatus. rroduct or process discioced in this report, o, represents that its use by such third party would not infringe gitately ovned rights.

h

>^* .y l' '  ;},  ;}

. g

hUREG/CR-5828 PNL-7727 Emergency Feedwater System Risk-Based Insaection Guide for the Arkansas Nuclear One Unit 2 Power Plant l 1

Manuscript Completed: Aprit 1992 I) ate Published: September 1992 Prepared by IL Pugh. H. F. Gore, T. V. Vo Pacific Northwest laboratory Richland. WA 99352 l*repared for Division of Radiation Protection and Emergency Preparedness Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 NRC FIN L1310

Summary Th!s docurnent presents a compilation of emergency feedwater (T fAV) system failure information which has been screened for risk significance in terms of failure frequency and degrtdation of system performance, it is a risk-prioriti7cd listing of failure events and their causes that are significant enough to warrant consideration in inspection planning at the ANO-Unit 2 plant. This information is presented to provide. inspecto.4 with increased resources for inspection planning at ANO Unh 2.

The risk importance on *us component failure modes was identified by analysis of the results of probabilistic risk assessments (PRAs) for mar.y prem.Td water restors (PWRs). Ilowever, the component failure categories identified in PRAs are rathe, broad, because the tallure data used in the PRAs is on aggregate of many individual failures having a variety of root causes. In order to help inspectors focus on spec!!ic aspects of component opera' ion, ,

maintenancc and desip. which mig'it cause these failures, an extensive review of component failure information was performed to identify and ranic the root causes of these component failures. Both ANO-Unit 2 and industry-wide failure information was analyzed. Riilure causes were sorted on the basis of frequency of occurrence and seriousness of consequence, and categorized as common cause failures, human errors, design probicms, or component failures.

This irdormation is presented in the txxly of this document. Section 3.0 piovides brief descriptions of these risk-important failure causes, and Section 5.0 presents more extensive discussions,with specific examples and references.

The entrics in the two sections are cross-referenced.

An abbreviated sptem walkdown table is presented in Section 3.2 which includes only components identified as risk important. T1 is table lists the system lineup for normal, standby spicm operation.

This information permits an inspector to concentrate on components important to the prevention of core damage.

Ilowever,it is irnportant to note that inspections should not focus exclusively on these components. Other components which perform essential functiotts, but which are not included because of high reliability or redundancy, must also b0 addressed to ensure that dtgradation does not increase their failure probabilities, and hence their risk importance. 1 NURiiG/C4582M l

1

- . . .- - ~ . - . - . . ~ .

k Contents Summary . . . . . . .. .. .. .... ....... . .. . .... ... . .. . .........................,lii _

_q l introduction . . ...... .. . ....... . .. .. .. ... ... ........ .... .... ....... ......... 1.1 2 - ANO Unit 2 EFW Sptem . . . . . ... .. . . ... ......................................2.1' 2.1 System Description ...... .. ... .... . , . ... . ..... . . .. . .. ......,. 2.1 q 2.2 Success Criterion . . .... . .. . . . . .. ... .. . .. . . . . 2.2 2.3 Spiem Dependencies . .. .. . . ...... . . , . . . . .. .. .. ~2.2 2.4 Operational Constraints . . . . . . .. .. .. .. . .. ... .. . . .. . . ... 2.2 3 Inspection Guidance for the ANO Uni: 2 EITV System . .. . . , .. ... . .. . .. . . .. 3.1

- 3.1 Risk Important EFW Componems and Failure Ak> des . . .. ... . .... ....... . . . . . . . . . . . . 3.1

. 3.1.1 Muh.ple Pump Pailures dac to Common cause . . . . . .. .. .. ...... 3.1 3.1.2 'Thrbine Driecn Pump 2P7A Fails to Start or Run . .. .. . . ... .... .. ..... . 3.2 . ,

3.1.3 Motor Dern Pump 2P7B Pails to Start or Run . . . . . . . . . _ . .... .. .. ,, 3.2 3.1.4 Pump 2P7A or 2P7B Unavailable Due to Maintenance or Surveillance . _. .. . . ... . 12 3.1.5 Bilure of Motor Operated Valvese.V.1025,1026,1036,1037.1038,1039,1075 and 1076 =. . , . . 3.2 3.1x Manual Suction or Discharge Wivu Ril Closed . . .. . .. . . . ... . 3.3 3.1.? L.cakage afIlot Feedwater through Check Wives . .. , .. ... . . . . , . .  ; 3.3 '

3.2 Risk Important EFW System Walkdown 'Dble . . . .. .. ,. . . ... . 3.3 4 Generic Risk insights From PRAs . . .. .. . . .. . ... . . . .. . ' 4.1 4.1 Ri.sk important Accident Sequences lavolving AFW System Failure . . . . . . .. .... 4.1 4J. I Loss of Power Spiem . . . .. . . . .. . .

.. .....: 4.1 4.1.2 'llansient. Caused Reactor or Turbine Trip . .. .. .. ... . .. . . 4.1 4.1.3 Loss of Main Feedwater . . . . . . .. ..- 4.1-

4J.4 Steam Generator 'Ibbe Rupture . .. . , .. . . ,. . . ... ... 4.1

~4.2 Risk Important Component Failure Modes. . . . . . . . . - 4.1

5 Failure Modes Desctmined From Operating Experience . , , . . . . . 5.1 t'
5.1 ANO Unit 2 Experience . .. . . ..  ! 5.1 f 5.1.1 EFW Pump Control Logic. Instrumentation and Electrical Failures ... . $J 1

4..

t v NUREG/CR 502s

. , , . . . -- , :: , . . . , + . . --,.L., . , , , , _ , ,

- . . . . . - - ~ . . - - . - - ....- - . . . .- -

L.

i:

1 l;

r 1

- 5.1.2 Pallure of EFW Pump Discharge Flow Control to Steam Generators . . . . . . . . . . . . . . . . . . . . . . . 5.1 ;I 5.1.3 EFW Steam Gerierator Isolation Valve Failures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.1  : 1 5.1.4 EFW 'Ibtbine Stea m lnlet '. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... . ... . .. . . .... . ... 5.1 4 l 5.1.5 Human Errors . . . . . . . . . . . . . . . . . . ..............................................5.2- -!

i 5.2 I n d as t ry Wid e Expe r ien ce . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2 i

5.2.1 Co m mo n Ca use Fa il u r es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2 5.2.2 Human Errors . . . . . . . . . . . . . . . . . ..... .......................................... 5.3 J 5.2.3 Design / Engineering Problems and Errors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . s . ~ 5.4 5.2.4 Compment Failures . . . . .. . . . . . . . . .. . . . . . . . .. .. . . . . . . . .. . . . . . .. . .. . . .. . . ... . . ... 5.5 6 References = . . . . . . . . . . . . . . . . . . . . . .. .. .. ......................................... 6.1 -

'Ihbles 1

3.1 Risk important EFW system walkdown table . . . . . . . . . . . . . . . . ... . . . . . . . . . .. . . .. .. . . . . . .. . . . . 3.5 -

i 1.

l l

l l

+

A l.

l NUREG/CRI5828 ' vi 1

'T M' Mv*- -

.+- I ae -, D 4 e 6 & v 4- w 6 . , ,- q4,e a q_gq4-g. , , pq g_ w ,. ,.j,,ya

1 Introduction his document is one of a series providing plact apecific The remainder of the document describes and discusses inspection guidance for emergency and auxiliary feed- the information used in compiling this inspection water (EFW/ABV) systems at pressurized water reac- guidance. Section 4.0 describes the risk importance tors (PWRs). This guidance is based on information information which has been derived from PIMs and its from probabilistic risk assessments (PRAs) for similar sources. A review of that section will show, the failure PWRs, industry-wide operating experience with ARV categories identified in PRAs are rather broad (e.g.,

systems, plant-specific ARV system descriptions, and pump fails to start or run, valve fails closed). Section 5.0 plant-specific operating experience, it is not a detailed addresses the specific failure causes which have been inspection plan, but rather a compilation of ERV/AFW combined under these categories.

system failure information which has been screened for risk significance in terms of failure frequency and degra- EIAV/AISV system operating histories were studied to dation of s) stem perfermance. The result is a risk-prior. identify the various specific failures which have been ag-itized listing of failure events and their causes that are gregated into the PRA failure mode categories. Section significant enough to warrant consideration in inspec- 5.1 presents a summary of ANO Unit 2 failure informa-tion planning at ANO Unit 2. tion, and Section 5.2 presents a review of industry-wide failure information. The industry widt information was This inspection guidance is presented in Section 3.0,iot- compiled from a variety of NRC sources, including lowing a description of the ANO Unit 2 EFTV system in AEOD analyses and reports,information notices,in.

Section 2.0 Section 3.0 identifies the risk important sys- spection and enforcement bolletins, and I;eneric letMrs, tem components by ANO Umt 2 identification number, and from a variety of INPO reports as wctl. Some Li-followed by brief descriptions of cach of the various fail- censee Event Reports and NPRDS event descriptions ute causes of that component. These include specific ws *e also reviewed. Finally, information was included human errors, design deficiencies, and hardware fail- from ieports of NRC-sponsored studies of the effects of ures. The discussions also identify where common cause plant aging,which include quantitative analyses of re-failures have affected multiple, redundant components. pc.rted EFW/AFW system failures. This industry-wMe These brief discussions identify specific aspects of sys- information was then combined with the plant specific tem or component design, operation, maintenance, or failure information to identify the various root causes of testing for inspection by observation, records review, the PRA failure categories, which are identified in training observation, procedures review, or by observa- Section 3.0.

tion of theimplementatio:.of procedures. An EFW system walkdown table identifying risk important com-ponents and their lineup for normal, standby system operation is also provided.

I 1.1 NUREG/CR4828 '

2 ANO Unit 2 EFW System l

This section presents an overview de,cription of the pump are independent from one another, Steam for the ANO Unit 2 ERV system, including a simplified sche- turbine diiven pump is supplied by each of the two main matic system diagram. .n addition, the system success - steam lines from a point between the containment pene-criterion, system dependencies, and administrative oper- tration and the main steam isolation valves. Each of the ational constraints are also presented, steam supply lines to the turbine has a check valve and a motor-operated steam supply isolation valve. The steam from both supply lines combines and is then directed to the turbine via an isolation valve, trip and throttle valve, 2.1 System Description and governor valve. The motor operated isolation valve, the trip and throttle valve, and the controls to the pver-The ERV system provides feedwater to the steam gener, not are supplied with power from an emergency DC ators (SG) to allow secondary. side heat removal from power source. Each EFW pump discharge is designed the primary system when main feedwater is unavailabic.

with a recirculation flow path to prevent pump dead-The system is capabic of functioning for extended peri, heading. The recirculation flowpath is restricted to a 50 ods, which allows time to restore main feedwater flow or gp 11 wrate by a flow limitin; orifice. The recircula-to proceed with an ordctly cooldown of the plant to tion flowpath is normally lined up to the Start-up and where the shutdown cooling system (SCS) can remove Blowdown Demineralizers.

decay heat. A simplified sc' hematic diagram of the EFW system is shown in Figure 2.1.

Each emergency feedwater pump discharge is provided with a stop check valve. The motor operated pump Ti c EFW system is controlled automatically by an (2P7B) is provided with a locally operated isolation Emergency Feedwater Actuation Signal (EFAS). Initia, valve. The Emergency Feedwater System discharge pip-tion of an EFAS automatically actuates the EFW system ing and valving arrangement is designed to aWx cf;her to provide art EFW supply to the steam generators on pump to supply feedwater to either or both steam gen-low steam generator water level. When an EFAS signal crators. Each supply line to cach steam generator is is generated, the turbine. driven pump (2P /A) and the pr vided with rt jundant control valves to ensure isola-motor-driven pump (2P78) are automatically started. tion of a faulted steam generator and the continued

% deliver flow to the affected steam generator, cmer, feeding of the non-faulted steam generator, The feed-

. gency feedwater control valves and isolation valves re-water valves to the S/Gs associated with the steam driv-ceive an open signal. When steam generator level is re-en EFWP (CV-1076 and CV-1026) have DC motor op-gained and level is greater than the EFAS actuation erators. The feedwater valves to the S/Gs associated setpoint, the control valws will receive a : lose signal.

initiation of a Main Steam isolation Signal (MSIS) _ with the AC motor driven EFWP (CV-1075 and CV-1025) are AC motor operated valves. The down-automatically shuts all remotely actuated emergency stream feedwater isolation valves (CV-1036, CV-1038, feedwater control valves and isolation valves unless an CV-1037, and CV-1039) are electrically operated ball-

. EFAS signal is present. Actuation of both a MSIS and valves that are cross-power supplied from the other aa EFAS automatica! s isolates emergency feedwater train and are normally positioned in the OPEN position.

!!ow to the ruptured steam generator and controls flow CV 1036 and CV-1038 are powered from " Green

power CV-1037 and CV-1039 are powered from

  • Red
  • DC battery power. This arrangement of normal valve The normal EFW pump suction is from condensate .

position and power supply is used to meet single failure storage tanks 2T41 A and B. A single condensate header criteria for emergency feed and main steam isolation supplies a header which cross-connects pump suctions.

Control, and instrumentation associated with each act uations.

2.1 NUREG/CR-5828

ANO Unli 2 EIAV System i

l:

CSTT41B and T41 A are the normal source of water for 2.3 Systerri Dependencies the EFW System ,. are required to store sufficient de-mineralized water to maintain the reactor coolant sys- .The EIAV system depends on AC power for the motor-tem (l(CS) at hot standby conditions for I hour followed driven pump and associated level control and isolation by subsequent cooldown to 350 E Since the CSTis not valves. DC power for motor operated valves associated acismically qualified, an assured source of water must be with the turbine discharge flowpath, control power to available to ensure EIAV operability under all pumps and valves, and an automatic actuation signal. In conditions. He safety grade service water system addition, the turbine-driven pump also requires steam provides this source of water. Service Water Loop 1 availability.

supplies the motor-driven pump,while Service Water

1. cop 2 supplies the turbine-driven pump. The Service Water supply valves (CV-0711-2 and CV-0716-1) will 2.4 Operational Constra.in ts automatically open if EFW pump suction pressure decreases to five psig and an engineered safety features signalis present. De valves are individually c itrolled When the reactor is critical the ANO Unit 2,'Ibchnical by pressure switches installed on the separate pump Specifications require that both EIAV pumps and associ-suction lines, ated flow paths are operable with the motor-driven pump powered from an operable vital bus and the tur-bine driven pump capable of being powered from an perable steam supply syuem. H ne EFW pump 2.2 Success Criterion becomes inoperable,it must be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant must be in hot shut-System success requires the operation of at least one down within the next twc!ve hours.

pump supplying rated flow to at least one of the two steam generators. In this condition, the system is cap. The ANO Unit 2 Tbchnical Specifications require at able of removing 2.9% full power heat load or approxi- least one condensate storage tank (CST) to be operable mately 100 Mwt*

w th a minimum ccmtained water volume of 160,000 gal-lons available for use.

i i

t l

l l

l l

NUREG/CR-5828 2.2

5-5 EFM 1

CU EFu

.. X EIM -

2P7B

, * -6l 43

?s s

14 Wt,tD

'1875-1 ms- 2 g . sw 6716 29 3B F Hurt SERUICE HUTOR E 11RTEg HEftDER ~ S "

D .,IUW p

MnIM DS' FROrt COffDENSHTE 'Cu 73 STORHCE 6703-1

, TnHus - " On Eria

&Tm FLtMSE M h rN CU 1 P EFu 1825-18838-2' 07e7 3 J L SH )

Ena .

q rEH4 1 ift j

Kna EH3 EN '

FRott STRRTUF Bet t 892 CU -

' 1960 RLOuDOtaN 799' l DKMSMEN4 2ER ED3 l l .l r4 EFFLUENT IIIN D . . 7R '

i 6ieiw w

'b SU 4 BD 7 7 I 7N CU CV Ena CU te26-2 1937-1 en SW 8706 . CU K F14 2UM

-9735-2 -N O,

" ns f,so .

f1H IN FDort SEMUICE. g gygg . De

=9 . rM

~

~

m Ea. w ,

-l Em u.

C C.

7 7 s 2P711 mg 3eys.2 CV- n 8U 3e3s-3 0711-2 8205 MnIH --

j SI N N

Cu MS ~

1909-1 39H

~

TURSINE CU 2 MAIN 8336 8332 ErurDRIm O stuw,  ;

ns g

" g -

2 8

% e-2 m

h. tE-2 [

M. .m O r at 3

x Figure 2.1 ANO unit 2 EFW system .m.

(. t 6.

t.

t

= b

.g

~. . _ _

m

3 Inspection Guidare for the ANO Unit 2 EFW System in this section the risk impottant comp <ments of the 11.1 Multiple Pump Failures due to Common ANO Unit 2 EFW system are identified, and the import Cause ant modes by which they are likely to f ail are L ~'y de-scribed. These failure modes include specific human The following le.ing summarizes the most important errors, design problems, and types of hardware failures multiple-pump failure modes identified in Section 5.2.1, which have been observed ic occur for these types of Common Cause Failures, and each item is keyed to components,both at ANO and at PWRs throughout the entries in that section..

nuclear indu3try. The discussions also identify where common cause failures have affected multiple, redun- .

Incorrect operator intervention into automatic sys-dant components. These brief discussions identify speci- tem functioning, including improper manual start-fic aspects of system or component design, operation, ing and securing of pumps, has caused failure of all maintenance, or testing for inspection activities. These pumps, including overspeed trip on startup, and activities include: observation, records review, training inability to restart prematurely secured pumps.

observation, procedures review or by observation of the CCl.

implementation of procedures.

Vahe mispositioning has caused failure of all Table 3.1 is an abbreviated EFW system walkdown table pumps. Pump suction, stern supply, and instru-w hich identifies risk important components. This table ment isolation valves have acn involved. CC2.

hsts the system lineup for normal, standby system opera-tion. Inspection of the components identified addresses ,

Steam bin < ling has caused failure of multiple pumps.

essentially all of the risk associated with EFW system This rew .ed from leakage of hot feedwater pa.st operation. check valves and a motor-operated valve into a com-mon discharge header. CC10. Multipic-pump steam binding has occured at ANO-2 resulting from 3.1 Risk Important EFW Components improper valve lineups and from running a pump allt! Failtire Modes deadheaded. CC3.

Common cause failures of multiple pumps are the most a

Pump control circuit deficiencies or design modifi-cation errors have caused failures of multiple pumps risk-important failure modes of EFW system compo.

to auto start, spurious pump trips during operation.

nents. These I re followed in importance by single pump f ailures, lesel control valve f ailures, and individual check and failures to restart after pump shutdown. CC3 vahe leakage failures, incorrect setpoints and control circuit calibrations have also prevcnted proper operation of multiple The following sections address each of these failure pumps. CC 1.

modes,in decreasing order of importance. They present the important root causes of these component failure Loss of a vital power bus has failed both the turbine-modes which have been distilled from historical records.

driven and m motor-driven pump due to loss of Each item is Lesed to discussions in Section 5.2 which control power to steam admission valves or to tur-Nne controls, and to motor controls powered from present additionalinformation on historical events.

tbc same bus. CCS.

3.1 NURECUCR-SS2x

Inspection Guidance 3.1.2 'Ibrbine Driven Pump 2P7A Fails to driven pump failures, as are circuit breaker failures.

Start or Run CF6. Control circuit failutes have been experienced at ANO.

o improperly adjusted and inadequatc!y inaintained turbine goscrnors have caused pump failures both  ? Mispositioning of handswitches and procedural de-at ANO and elsewhere. HE2. Probleir include ficiencies bwe prevented automatic pump start.

wott or loosened nuts, set screws, linkages or cable H E3.

connecticie,, oil leaks and/or c<mtamination, and cledrical failures of resistors, transistors, diodes and Low lubrication oil pressure resulting from hutup circuit cards, and erroneous grounds and due to previous operation has prevented pump re-connections. CF5. start due to failure to satisfy the protective inter-lock. DES.

'Rrry turbines with Woodward Model EO gour. l nors hase been found to overspeed trip if full steam 3.1.4 Pump 2P7A or 2P7B Unavailabie Due to q flow is allowed on cartup. Sensitivity can be re. Maintenance or Surveillance duced if a startup q oypass valve is sequenced to 1 open first. DE!. -

Both scheduled and unscheduled maintenance re-move pumps from operability. Surveillance requires Turbines with Woodward Model PG-PL governors operation with an altered line-up, although a pump have tripped on overspeed when restatted shortly train may not be declared inoperable during testing.

after shutdown, unless an operator has locally exer- Prompt scheduling and performance of mainte-cise ' the speed setting knob to drain oil from the nance and surveillance minimize this unavailability.

governor speed setting cylinder (per procedure).

Automatic oil dump valves are now available 3.1.5 Failure of Motor Operate 1 Valves CV.

through "Rrry. DE4. 1025,1026,1036,1037,1038i 1039,1075 aml: -

1076 Condensate slugs in steam lines have caused turbine ,

overspeed trip on startup. 'ksts repeated right after gg

, y.

such a trip may fail to mdicate the problem due to g g,

warmingand clearingof the steam lines. Surveil. g gg g 3

( lance should exercise all steam supply corinections.

DE2~

Common cause failure of MOVs has occurred at Tbrbine stop valve (CV.0336) problems which have

    1. # # # E"" "" ## S' fatted the turbine driven pump include physically C9u Pment to dercrmine proper settings of torque bumping it, failure to reset it following testing, and switch and torque switch bypass switches. Failure to g  ; gg, ,g failures to verify control roora indicati m of reset.

tmder design basis accident conditions has also been l HE2. Whether either the overspeed inp or TTV involved. CC8.

trip can be reset without resetting the other, mdica-tion in the control room of Tl'V position, and un- , gj gg g ambiguous local indication of an overspeed trip affect the likelihood of these en ors. DE3. pr per sizing or use f thermaloverload protective devices. Bypassing and oversiting ehould be based on proper engineering for design basis conditions.

3.1.3 Motor Driven Pump 2P711 l' ails to Start CF4. ANO has experienced similar type failures.

[ or Run j'

  • Out.of adjustment electrical flow controllers have Control circuits used for ati.omatic and manual caused improper discharge valve operation, affect.

pump starting are an important cause of motor ing multiple trains of EFW CCl2.

NUREG/Cib5828 3.2

4 Inspection Guidance

Riflure to verify support functions after Limitorque SMB motor operators has caused motor restoration burnout or thermal overload trip by preventing tor-que switch actuation. CF7. - Failure to adhere scrupulousiv to administrative -

procedures regarding tagging, control and track. -

  • Manually reversing the direction of motion of oper- ing of valve operations' ating MOVs has overloaded the motor circuit.

Operating procedures should provide cautions, and -

Riilure to log the manipulation of scaled valves circuit designs may prevent reversal before each .

stroke is finished. DE7. -

Failure to follow good practices of written task assignment and feedback of task completion

  • Space heaters designed for preoperation storage information have been found wired in parallel with va!ve motors which had not been environmentally qualified with -

Failcre to provide easily read system drawings,.-

them present. DER legible valve labels corresponding to drawings and procedures, and labeled indications oflocal -

3.1.6 Manual Suction or Discharge Wives Fail valve position Closed 3.1,7 Leaknge ofIlot Feedwater through TD Pump 2P7A: Wives ERV&EFW-4A MD Check Valves -

Pump 2P7B: %1ves EFW-38. EFW-4B. EFW-6 Between Pump 2P7A and MFW: Wlve EFW-4 A These manual valves are normally lockert open. For Between Pump 2P7B and MFW: %1ve EFW-4B cach train, closure of the first valve listed would block isolate pump suction from all possible sourcesf Closurc

  • Leakage of hot feedwater through several check of the second (or third) valve would block all pump dis- @cs in series has caused steam binding of muitiple charge including recirculation to the start-up and blow pumps.- Leakag; through a closed level control down demineraliters. valve in series with check valves has also occurred, as would be required for leakage to reach the motor
  • Wlve mispositioning har resulted in failures of driven pump 2P7B. CC7<

multipte trains of EFW. CC2. ** has also been the dominant cause of probicms identified during oper.

  • Slow leakage past the final check valve of a ser.ics ational readiness inspections.1IEl. Events have may not force the upstream check valve closc i.

occurred most often during maintenance, calibra- Other check valves in series may leak similarly. Pip-tion,or system modifications. Impo: tant causes of ing orientation and valve design are important fac-mispositioning include: tors in achieving true series protection. CFl.

Failure to provide cornplete, clear, and specifie 3.2 RiskImportant EFW System procedures for tasks and system restoration Walkdown Tal>le Riilure to promptly revise and validate p :c-Thble 3.1 presents an EFW system walkdown table in-dures, training, and diagrams following systent ciuding only components identified as risk important.

modifications The lineup indicated is for normal power operation.

This information allows inspectors to concentrate their Failure to complete all steps in a procedure efforts on components important to prevention of core Rillure to adequately review uncompleted pro-cedural steps after task completion 33 NUREG/CR-5823

Inspection Guidance L-

. damage, However,it is essential to note that inspec- that their risk importances are not increased. Examples tions should not focus exclusively on these comments.' ~ include the an adequate water level in the CST and the Other components which perform essential functions, (cksed) valves cross connecting the discharges of the i but which are absent from this table because of high reli . EFW pumps.

ability or redundancy, must al.so be addressed to ensure 1

-]

s

?

A i

l l

.- NU REG /CR.5828 .

. . -.. . _ - 4

inspection Guidance

'luble 3.1 Risk iniportant EIM' System Walkdow n Table Requireel Actual Comgnment # Component Name location Position Position flectt ical 2P711 Motor breaker 2A311 A3 SWOR RM Racked in on Closing springs CllG Motor Energi<cd EFW i :owpath 2SW.39u loo" 2 Supply tc EFW System 2P7A Ph'P 161 lecked Open 2SW.39A loop 1 Supply to EFW Sy tem 2P711 PMP 'lM i ocked Open 2CV4171r- 1 Service Water to 2P7;l 2P7H PMP " + closed 2CV417112 Scnke Walet to 2P7A 2P7A PMP RM Closed 21EFW.0706 EFW Pumps Suction from SU and 335' El.EV T H Incked*

llD Closed 2EFWSO2 EFW Pump Suction from 2T41 A or 2P711 PMP RM lurked Open DI Emuent

?CV '07 EFW Pump Suction from CST 335' ELEY Til Open 2T41AM 2P711 Flowpath 2EFW.3I; 2P711 Suction Isolation Yalve 2P70 PMP RM locked Open 2 EFW-411 2P711 Discharge Stop-Che(k 2P711 PMP RM incked Open _

2 EFW-6 2P7B Discharge Valve 2P7B PMP RM L , sed Open _

2CV.1075 1 2P711 Dis. charge Control Valve to NPPR Closed _ _ _

S.G. 2E24H 2C%1034 ' 'P7B Discharge to S.G. 2E.'4B NPPR Open _

lsolatio..

2C%10251 2P7B D4 charge Control Yah e to SPPR Closed S.G. 2 E24 A Locked Closed Above 10 Percent Reactor Power NPPR - North Piping Penetration Room SPPR . South Piping Penettelion Room 3.5 NUREG/CR-5828

l l

Inspc< tion (iuidarac

'inble 3.1 (Continued)

Required Actual Comgainent # Cornguinent Narne l xacation Position Pmition 3 Y LOW 2 . '7111)N hatre to S O. .'E24 A NPPR () pen _

P olanon 2 P7A l lowpath 21 I W 3.\ 2P7 A huuton isolanun 2P7A PMP RM 1 m ked Opt n _ _ _ _ _

21:l'W 4 A 2P7A !)i,o arpe Stop-Chc( k 2P7A PMP RM in ked Opt n

?( V 1";t. 2 2P7A thscluope Cantrol %he to NPPR ( 'loso! _

S G. 2E24B 2( Y 10 PA1 27A thst harpe luriation %he t ., NPPit Open _ _ _ .

5 G 2E2111 20 E 10.% 2 2P7A Dmh>ye Control %hc in 5PPP thbed 5 G 21?21 A i

> 2l V !Url-1 2P7A Ihschatre Isol.ition Whc to SPPR Open S G.2H24A Recis c and i 1.- anc 2El?W 1l A Ret ut 1.nic holanon SPPR i n ked Closed _

?lCl W-ll' Ruin I me kolation N1'PR l_w ked Closed __,

21(1 W 10A 2P7A Minimurn Rcurc Fiow (.00u01 NPPR inled Open __

21?l/W- 10B 2P71t Minnnum Rnoc Flow Connol SPPR Ia ked Op ___

t *oss-Connec' %h rs 2CV 07pl l'u..ip Sucuon X TIE 2P;d PMP RM Open _

JCV 07'i$ ' Purup hta tion XTIE 2P7A PMP .<M Open . , _ _

21 FW3A l' ump Dmharge XTIE 2P715 PMP RM Closed 21 l'WAh Pump thstharge NTIE 2P7A PMP RM Closed _

Stearn supply \ ahes 2$V U?hs 5 team Supply Bypass 2PtoA PM RM Closed 2C'-0110 Steam $upply Lhe 2 PICA PMP RM Closed .

2(VOM6 thp and Throttle %he 2P7A l'MP RM Opec __

2CVUwn1

  • A' M on Sicam Supply holanon Main Sicam Open ,_

Pcm hause N t !R EG /CR-M28 3 ri

l Instwction Guidance lable 3.1 (Continued) liegt.'r ed Act u s:1 Cony.onent # Comlmnent Narne I;> cation l'osition Posidon 2CV-10$0 ? *11" Maut Sicam Supply hol.iiton Main $1caru Upen _

Pe n t house i 1/W SG Chett Yahrs 2'l W () 2P711 Discluir re 'Ittuper ature ?P711 PMP RM i50 Deg Maximum 2TWil713 2P7A Disc harre 'ltmperature ?P7A PMP RM 150 Der Maximum l' N U R E G!C R.582S i

4 Generic Itisk Insights From PilAs PRAs for 13 PWRs were analyzed to identify risk- 4.1.3 J. mss of Main Feedwater important accident sequences involving loss < '

EFW/Al W, and to identify and risk-prioritiic the mm- . A feedwater line break drains the common water ponent failute modes invohed. *lhe results of this source for MFW and AEW. The operators fail to analysis are described in this scction. They are consis- provide feedwater from other $.ources, and fall to tent with results reported by INEL and DNL (Gregg et initiate feed and-biecd cooling, resulting in core al.1988;"Itavis et al,19SS). da mage.

  • A loss of ma.n feedwater trips the plant, and AISV 4.1 Itisk Important Accident Sequences f ails duc tu operator error and hardware failutes.

Involvilig AFW System Failure 'D'c operaim fau niinitiate feed and biecd cooling, resulting in core damage.

4.1.1 loss ofl'uoer Systern 4.1.4 Steam Generator *Ibbe Rupture

+ p n is followed by failure of j\ lo s of of fsite p< , gg.g ggg 7 g g, Al% Dec to lack of actuating power, the 1 ORW 10 rom me primary until the RWST is depleted.

/ cannot be opened, preventing adequate Ict d and-a nce rec don cannot be esta%hed bleed cooling, and resulting in core damage.

from the empty sump, and core damage results.

A station blackout fails all AC power except Vital AC from DC invertors, and all decay heat removal s'-1 cms except the turbine-driven dFW pump. 4.2 Itisk Important Component Failure Al W subsequently fails due to battery depktion or Modes hardware failures, resulting in core damage.

The generic component failure modes identified from

+

A DC bas fails, causing a trip and iailure of the PRA analyses as important to AFW systern failure are power conversion system. One AFW moter driven listed below in decreasing order of risk importance.

pump is failed by the bus loss, and the turbiac-drhen pump fails due to loss of turbine or vahc 1. 'Ibibine-Oriven Pump Failure to Start or Run.

conttoi power. AITV is subsequently lost completely due to other failures. Feed-and-bleed 2. poter. Driven Pump Failure to Start or Run.

cooling fails because PORV control is lost, resulting in core damage. 3. TDP or MDP Unavailable due to Test or Maintenance.

4.1.2 busient-Caused Reactor or Turbine

'lyip 4. AFW System Wlve Failures A transient-caused trip is followed by a loss of PCS

  • steam admission valves and AFW.

trip and throttle valve

- Feed and bleed cooling fails either due to tailure of the operator to initiate it, or due to hardware fail.

  • flow controlvalves utes, resulting in core demape.

4.1 NUREO/CR-5828

j Generic Risk insights From PRAs e pump discharge valves in addition to indhidual hardware, circuit, or instru.

ment failures, cach of these failure modes m.sy result i

(

  • pump suction valves from common causes and human errors. Common 1 l cause failurcs of AFW pumps are particularly risk c-
  • valves in te', ting or maintenance. important. Valve failures are somewhat less important due to the multiplicity of steam generators and connec.

5 Supply / Suction Sources tion paths. liuman errors of greatest risk importance involve: failurcs to initiate or mntrol system operation .

  • condensate storage tank stop valve when required; failure to restore proper system lincup, .

af ter maintenance or testing; and failure to switch to l

  • hot wellinventory alternate sources when required.

i

  • suction valves, i

t t

1 P

G NUREG/CR.5828 4.2 -

, __..:.,, _ _ _ , , . z. . u,.. _ _ , . . ..,2.. ., _ _ . . _ . _ , _ _ . _ _ , .

5 Failure Modes Determined From Operating Experience

'niis section desaibes the primary root causes of com- 5.1.2 Failure of EPY Pump Discharge Flow ponent failures of the tiRV sptem, as determined from Control to Steam Genemtors a review of ope ating histories at ANO Unit 2 and at other PWRs throughout the nuclear iedustry. Section 'lkenty eight tailures of the ERV pump discharge !!ow 5.1 de3cribes experience at ANO Unit 2. Secuon 5.2 umtrol and bypass valves were found in the events ex-summarites industry. wide information compiled from a amined. 'Diese resulted from failures of valve control variety of NRC sources, including AEOD analyses and c rcults, valve operators and valve breakers. Fallures tcports,information notices,lapection and enforce' have resulted from DC c mtrol grounds, valve binding, ment bulletms, and generie letters, and f rom a variety of dirty or worn contacts, improper toraue switch oper-INPO reports as well. Some Licensec Event Reports atiors electrica a>mponent failure, frayed wiring, valve (LERs) and NPRDS event descriptions were ab.o re' operator mechanical failure and low hydraulic fluid viewed. Finally, information was included from reports pressure. Failure causes are mechanical wear, contact of NRC sponsored studies of the cficcts of plant aging, oxidatior., inadequate maintenance or te,, ting activities whkh inct e quentitative analyses of ERV system fail- and ircpioper design and/or installation. These valves ure reports. This informathm was used to identify the have also experienax! various packing leaks, as have various root causes expected for the broad PRA-based pump dischat ge check valves.

failure categorie:; identified m Section 10, resulting in the inspection guidelines presented in Section 3.O' 5.1.3 EFW Steam Generator Isolatinct Whc Failures 5.1 ANO Unit 2 Experlenee Ten failures of the ERV steam generator isolation valves were found in the events examined. These fail.

Sixty-eight (68) reports of EFW system equipment fail- urcs resulted from valve binding, solenoid coil failure, ures at ANO Unit 2 between January of 1980 and Octo-fouled torque switch contacts, control power short cir-ber 30 IW9 were reviewed. Rese include failures of the cuits, fonled closing coil motor controlling contactors, ERV pumps, pump discharge flow control valves to and breaker failure. Failure causes are mechanical wear, steam genetators, and pump suction and discharge contact nidation, component aging, and inadequate valves. Failure modes include electrical,instrumenta-g,aintenance or testing activities and original circuit de-tion hardware failures,and human errors.

sign problems.

5,1,1 Elw Pump Control logic, 5.l A EFW'Ibrbine Steans Inlet Ir.strumentation and Electrical Failures Five failures of the ERV steam mlet valve were found in

'Rn failures of the EFW pumps to starr, run, trip when the events examined. These failures resulted fron alve required or achieve rated speed were found in the events binding, acteator motor failure, improper installation of examined. These occu:rences iesulted from failures of valve operator and installation of improperly sized ther-the turbine governor, brealers, relays and contacts, tur- mal overloads. Failure causes are mechanical wear, bine overspeed desice, faulty wiring and power supplies. component aging, contact oxidation or fouling and inad- ,

The failure causes are mechanical wear, corrosien,or equate maintenance or testing activities.

improper design and installation, 51 NUREG/CR 5828

Failure Males 5.l.5 11urnan Errors condensateinto the AF turbines from the long, unheated steam supply hncs. (The system had never Nine events relating ditccdy to signifkant human errors l'cen tested with the abnormal, cross connected steam alfccting the EFW system were found in the esents supply lineup which resulted.) In the 'ltojan event the examined. Wlve operators were wired inwtrectly or oper1to ncorrectly stopped both AIAV pumps due to damaged af ter maintenance activities. Irnproper misinterpretation of MISV pump speed indication. The installation of valve components has tesulted in valve diesel driven pump would no', restart due to a protective binding aaJ motor damage. Miswiring turbine control feature requiring complete shutdown, and the turbine-circuits has rusulted in inability of the turbine to reach driven pump tripped on over* peed, requiring local reset i desired spted. Impreper installation of special test of the trip and throtdc valve. In cases where manual equipment has resulted in loss of turbine control power int-rvention is required during the early stages of a tran-and automatic turbine start capability. sient, training should emphasize that actions should be performed methodically and deliberately to guard againu such errors.

5.2 Intiustry Wide Experience .

CO2. Valve mispositioning has accounted for a signifi.

cani fraction of the human errors failing multiple trains Iluman enots, deso,;n/ engineering problems and errors, of AFW. This includes closure of normally open sucthn and 7mporent failures are the primary root causes of dves or steam supply valves, and of isolation valves to EFW /AFW System f attutes idenuried in a review of in- sensors having control functiom incorrect handsw. itch dustry wide system operating history. Common cause posit uning and inadequate temporary wiring changes failures,which disable more than one train of this oper-hve also prevented automatic st etts of muhlple pumps.

ationally redundant system, are highty nsk significant, Factors identified in studies of misposinoning citors in-and can result from all of these causes' clude failure to add newly installed valves to valve

'"#c Ttus section identifies imponant common cause failure rahon,g, weak adminiuradve independent esmtrol verification, and locked ohagging, testo-valve log-mode', and then provides a Froadet discul.ston of the g r.g,and inadequate adherence to piocedures. lllegible single failure effects of human ettois, design / engineer-or confusine, hical valve labeling. a id insufficient train-inc problems and errors, and component failures. Para- bneh of W pum wm m gn.phs presenting details of these failure modes are mask mispositioning, and surveillance which does not coded (e.g , CCl) an41 crms-reletented by inspection se amplete system furictioning m:iy not reveal 58ctus in ScClion 3.

mispositionings.

5.2.1 Common Cawe l'ailures M Design / engineering errors have accounted for a smaller, but significant traction of common cause fail-

"Jhe dominant cause of AFW .ystem multiple train f ail- i.res. Problems with control circuit design modifications urcs has been human error, Design!coghicer.ng errors at Farby defeated AITV pump auto-start on loss of and component f ailures have been less frequent, but main feedwater. At Zion-2, restart of both motor driven nevertheless sigmficant, causes of multipic train f ailures. pumps was blocked by circuit failure to deenergize wben the pumps had been tripped with an automatic start sig-CCl. Iluman error in the form ofincoricet operator nal present (IN S2-01,1982). In addition, AFW control int' rvention irno automatic EFW system functioning circuit desien icviews at Salem and Indian Poi .t have during transients resulted in the ttmporary lost of all identified designs where failures of a sMgle component e safety-grade AFW pumps during events at Duis Besse could have failet; all or multiple pumps (IN 87 34, (N UR EG- 1154,1985 ) a nd itojan ( AEO D/1416,19S3 h gygg in the Davis Besse event, improper mar ud ut'tiation of

.he steam and feedwater rupture controi system CC.t incorrect setpoints and control circuit settings (SFRG) led to ovetspeed tripping of both turbine-dnt- resu; tint; from analy" Trors and failures to update pro-en AFW pumps, probably due to the introduction of cedures hase also prt .ated pump start and caused NUREOCR-5828 5.2

- ~ . _ _ _ _ _ _ _ __

Pallure Modes l

purnps to trip spuriously. Errors of this type may Auxiliary I'cedwater Putnps' as Generic issue 93. This remain undetected despite surveillance testing, unless generic issue was resolved by Generic Letter 88-03 surveillance tests model a:1 types of system initiation (Miraglia,1988), which req : fred hcensees to monitor i and operating conditions A greater fraction ofinstru- AIW piping teruperatures cach shift, and to maintain mentation and control circuit problems has been identi- procedures for recognizing steam binding and for restor-ficd during actual system operation (as opposed to sur- ing system operability.

veillance testing) than for other types of failures.

CCR Common cause failures have also failed motor op-CCS. On two occasions at a foreign plant, failure of a crated valves. During the votal loss of feedwater event at balance of-plant inverter caused failure of two AIM Davis Besse, the normally-open ATM isolation valves pumps. In addition to loss of the motor driven pump r failed to open e ler they were inadvertently closed. The whose auxiliary start relay was powered by the invertor, failure was due to improper setting of the torque switch the turbine driven pump tripped on overspeed luause bypass switch, wnich prevents motot trip on the high the governor valve opened, allowing full steam now to torque required to unseat a closed valve, Previous prob.  !

the turbine, This illustrates the irnportance of assessing lems with these valves had been addressed by increasing the effects of failurcs of balance of plat.t equipment the torque switch trip setpoint a fix which failed during which supports the operation of critical components. the event due to the higher torque required due to high ,

Tue instrument air system is another exampi: of such a differential pressure across the valve. Similar common '

systern. rnode failures of MOVs have also occurred in other sys; tems, resulting in tssuance of Gencric Letter 8910,

" Safety Related Motor operated Valve Tbsting and Sur. .

CCA Asiatic clams caused hilure of two AIM flow veillance (Partlow,1989)? This generic letter requires umtrol valves at Catawbad when low suction pressure licensecs to develop ard implement a program to pro-caused by starting of a motor-driven pump caused suc- vide fu :he testing, inspection and maintenance of all tion source realignment to the Nuclear Service Water safety clated MOVs to provide assmance that they will system. Pipes had not been routinely treated to inhibit Iurmon w hen subjected to design basis conditions.

clam growth, nor regularly monitored Ia detcet their ,

presence, and no strainers were installed. The need for CH Other component failures have also resulted in surveillance whiQ exercises alternative system opera- AIM multi train failurcs. These include out of. adjust.

  • tional modes, ,weh as tempicie system functioning,is ment clectrical flow controllers resulting in improper emphasized by this event. Spurious suction switchover discharge valve operation, and a failure of oil cooler
  • has also occurred at Callaway and at McGuire, although cooling water supply valves to open due to silt ne failures resulted. accumulation.

CC7. Common cause failure.s have also been caused by 5.2.2 !!untan 14rrors component failures (AEOD/C404,1984). At Surry-2, both the turbine driven pump and one motor driven Iliil. The overwhelmingly dommant cause of problems pump were declare'l inoperabic due to steam binding identified during a series of operational reaviness evalu-caused by backlcakage of hot water through multiple ations of AIM systems was human performance.The check valves. : At Robinson 2 both motor driven pump > majorityof these human performance problems resulted '

were found to be hot, and both motor and steam vriven from incomplete and incorreu prosedurcs, particularly pumps were found to be inoperable at different times- with respect to valve lineup informe. tion. A study of Backlcakage at Robinson 2 passed through closed valve mispositioning events involving human error motor operated isolation valves in addition to multiple identified failures in administrative control of tagging check valves. At Parley, both mctor and turbine driven and logging, procental compliance and completion of pump casings were fuand hot, although the pumps were steps, verification of support systems, and inadequate not declared inoperable. In addition to multi train fail- procedures as important. Another at_idy found that ures, numerous incidents of single train failures have oc- . valve mispositioning events occurred most often during curred, resulting in the designation of

  • Steam Binding of ,

5.3 NUREG/CR-5828

._ m

Failure Modes mainten aco, calibration,or modification activitics. DR1 lbrbine trip and throttle valve (TIV) problems insulficient training i; determining vi lve position, and are a significant cause of turbine driven pump failures in administrative requirements for controlling valve po- (IN MM). In some cases lack of'ITV position indica-sitior.ingwere important causes, as was oral task assir,n- tion in the control room prevented recognition of a trip-ment without task completion feedback. ped TlV in other cases it was possible to reset eitter ibe omspeed trip or the TIY without resetting the 11E7.1brbine drivcn pump f ailures have been caused by other. This problem is compounded by the fact that the human er rors in calibrating or adjusting governor speed position of the overspeed trip linkage can be misicading, umtrol, poor e,0vernor maintemnce, incorrect adjust- and the mechanism may lack labela indicating when it is ment of governor valve and overspeed trip linkages, and in the tripped position (AEOD/CliO2,1986).

errors associated with the trip and throttle valve. 'I'lV-assoc.ated crrors incle physically bumping it, failure DF4. Startup of turbines with Woodward Model PG-to restore it to the correct position after testing, and PL governors within 30 minutes of shutdown has re-fadures to verify control room indication of1*IV posi-  ; alted in overspeed trips when the speed setting knob tion following actuation. was not exercised locally to drain oil fren the speed set-ting cylinder. Speed control is based on startup with an liEJ.

l Motor driven pumps have been fahd by human empty cylinder. Problems have involved turbine rota-errors in mispositiomng handswitc hes, and by procedure tion due to both procedure violations and leaking steam.  !

deficiencies Terry has marketed two types of dump valves for auto, matically draining the oil af ter shutdown ( AEOD/C602, 5,2 3 Design / Engineering Problerns und 1986).

Errors At Cdvert Cliffs,a 1987 loss-of-offsite power event re-del. As noted above, the majority of AINV subsystern quired a quick, cold startup that resulted in turbine trip due to PG PL governor stability problems. The short-failures, and the greatest relative system degradation, ~

has bet n found to icsult from turbine-driven pump fail, term corrective action was installation of stiffer buffer springs (IN SS-W,1988). Surveillance had always been utes. Overspeed trips of'Ibrry turbines con sited by Woodward governors have been a significant source of preceded by turbine warmup, which Ulustrates the im-these failures (AEOD/C602,1986). In many cases these portance of testing which duplicates service conditions as much as is practical.

overspeed trips have been caused by slow response of a Woodward Model EG governor on startup, at plants where full steam now is allowed irnmediately. This over. M Reduced viscosity of gear box oil heated by prior operation caused failure of a motor dri en pump to start sensitivity has been removed by it: stalling a startup due to insufficient tube oil pressure Lowering the pres.

steam bypass vaive which opens first, allowing a control.

sure switch setpoint solved the problem,which had not led turbine acceleration and buildup of oil pressure to been detected during testing, control A governor valve when fullsteam flow is admitted.

DFri. Waterhammer at Palisades resulted in AFSV line and hanger damage at both steam genetators. The AIRV l]El Ose speed trips olleny turbines have been

  • spargers are located at the normal steam generator level, caused by condensate in the steam supply lines. Con.

oensate stom down the turbine, causing the governor and are frequently coscred and uncovered during level fluctuations. Waterhammers in top-feed ring stear.,

valve to open farther, and overspeed results before the generators resulted in main feedline rupture at Maine governor valve can respond, af ter the water slug clears.

Yankee and needwater pipe cracking at Indian hint 2 This was determined to be the cause of the loss-of-all.

AITV event at Dasis 11 esse (AEODNt2,1986), with (IN 84-32,1984).

condensation enhanced due to the long length of the cross. connected steam lines. Repeated tests following a

[2El. Man Jiy reversing the direction of motion of an operating valve has resulted it MOV failures w here cold-start trip may be successful due to systcm heat up.

such loading was not considered in the design 6

NUREG/CR-5828 5..t

Rillute Modes (AEOD/CMG,1986). Control circuit design may pre- problem, and recurring leakage has been experienced,  ;

vent this, requiring stroke completion before reversal. even after repair and replacement. '

DE8. At cach of the units of the South 7bxas Project, CF2. At Robinson, heating of motor operated valves by -

space heaters provided by the vendor for use in prein- check valve leakage has caused thermal binding and fail.

stallation storage of MOVs were found to bewired in ute of AFW discharge valves to open on demand. At i

parallel io th: Claw IE 125 V DC motors for several Davis Desse, high differential pressure across AFW AFW valves (IR 50-489/89-11; 5 499/89-11,1989). The injection valves resulting from check valve leakage has valves had been environmentally qualified, but not with prevented MOV operation (AEOD/C603,1986). r the non. safety-related heaters energized.

CF3. Gross check valve leakage at McGuire and Robin.

5 2,4 Component Fnilures son caused overpressurization of the AFW suction pip-ing. At a foreign PWR it resulted in a severe water-Ocneric Issue ll.IL6.1, .*In Situ Tbsting Of Valves

  • was hammer event. At Palo Verde 2 the MFW suction pip.

divided into four sub-issues (DecHord,1989), three of ing was overpressurized by check valve leakage from the which relate directly to prevention of AFW sptem com. AFW system (AEOD/C404,1984). Gross check valve ponent failure. At the request of the NRC,in situ test. leakage through idle pumps represents a potential diver.

Ing of check valves was addressed by the nuclett indus- sion of AFW pump flow.

try, resulting in the EPRI report,

  • Application Guidc.

lines for Check WM n, Juclear Power Plants (Drooks, CF4. Roughly one third of AFW system failurcs havc 1988)? This extensive report provides information on been due to valve operator failures, with about equal .

check valve applications, limitations, and inspection failures for MOVs and AOVs. Almost half of the MOV techniques. In situ testing of MOVs was addressed by failures were due to motor or switch failures (Casada, bencric tetter h910,

  • Safety Related Motor-Operated 1989). An extensive study of MOV events (AEOD/

Valve Tbsting and Surveillancc* (Partiow,1989) which C603,1986) indica'es continuing inoperability problems requires licensecs to develop and implement a program caused by: torque switen/ limit switch settings, adjust.

for testing, inspection and maintenance of all safety-re. ments, or failures; motor burnout; improper sizing or lated MOW.

  • Thermal Overload Protection for Electric use of thermal overload devices; premature degradation Motors on Safety.Related Motor Operated Valves. related to inadequate use of protective devices; damage Generic issue ll.E.6.1 (kothberg,1988)* cor cludes that due to misuse (valve throttling, valve operator hammer-valve motors should be thermally protected,yc* in a way ing); mechanical problems (loosened parts, improper which emphasizes system function over protecaon of the assembly); or the torque switch bypass circuit improp-operator, erly installed or adjusted. The study concluded that cur- ,

rent methods and piocedures at many plants are not ad.

CFl. The commordcause steam binding effects of check equate to assure that MOVs will operate v len needed valve leakage were identified in %cction 5.2.1, entry under credible accident conditions. Specilically, a sur.

CC10. Numerous single-train events provide additional veillance test which the valve passed might result in un-insights into this problem. In some cases leakage of hot detected valve inoperability due to component failure MFW past multiple check valves in series has occurred (motor burnout, operator parts failure, stem disc sepa-because adequate valve-seating pressure was limited to ration) or improper positioning of protective devices the val as closest to the steam generators (AEOD/C404, (thermal overload, torque switch, limit switch). Generic .

1984). At Robinsen, the pump shutdown procedute was let ter 89-10 (Partlow,1989) has subsequenth required changed to delay closing the MOVs until after the check licensees to implement a program ensuring that MOV '

valves were scated. At Parley, check vahrs were switch settings are maintained so that the valves will changed from swing type to :ift typc Check valve re- operate under design basis conditions for the life of the ,

work has been done at a number of plants. Different plant.

valve designs and manufacturers are involved in this g5% Component problems have caused a significant number of turbine driven pump trips 5.5 NUREG/CR 5828

_ . _ _ -. _ _ _ __. _ _ . _ __. 1 i

Failure Modes

( AEOD/Cui2,1 *,6L One group of esents involved duven pump inoperable for 40 dap until the next sur-worn tappet nut face, h>ose cable connections, loosened veillance (AEOD/E702,1987). Pioblems result irom set screws, improperly latt hed TfV>, and improper a grease thanges to EXXON NEBULA EP-0 grease, one sembly. Another invohed oilleaks due to wmponent of only two greases considered environmentally quali-or seal f ailures, and oil contamination due to poor rnain- ficd by Limitorque. Due to lower viscosity,it slowly mi-tenance activities. Gmernor oil may not be shared with grates from the gear case into the spring pack Grease surbine lubrication oil, resulting in the need for separate changcoser at Vermont Yankee affect:d 40 of the older G(hanget Electrical component f.ulurcs included MOVs nf which 32 were safety related. Grease relief transistor or iesistor f ailures due to moisture intrusion, kits are needed for MOV operators manuf actured erronmus grounds and wnnections, diode f ailures, and beim: 1975 At Limerick, ailitional grea$,c rehef was

- a f aulty circuit card. requiicd for MOVs mandactured since 1975. MOV re-f urbishrnent programs may yield other changeovers to cln Control circuit f ailures were the dominant source EP-0 gicase.

of motor driven AFW pump failuies (Casada,19X9p This includes the controls used for automatic and manu. rFx Ihr systems usi.,g AOVs, operability requires the al starung of the purnps, as opposed to the instrumenta. avauability of instrument Air, backup air,or backup tion inputs Most of the iemaining problems were due nitrogen. Ilowever, NRC Maintenance lam inspec-to circuit bre:er f ailures. tions have identified inadequate testing of check valves isolating the safety related portion of the lA system at Cil *1lydraune lockup" of Limitorque 5MB spring several utilities (Letter, Roe to Richardson). Generic packs has presented proper sprmg compressior. to actu- 1 rtter M-14 (Miraglia,1988), requires lic':nsees to ver.

ate the MOV totque switch,due to grease trapped in the ify by test that air. operated safety-related components spring pack. During a surveillance at ~1tojan, tailure of will perform as expected in acmrdance with all design-the torque switch to trip t he Tiv motor resulted in trip- basis es ents, intiuding a loss of nor mal l A.

ping of the therto oserload device, leaving the turbine s

NURCO CR 5S2X 56

6 Iteferences BecLjord E. S. June 30,1989. C/mcout of Gencric issuc A1:OD Reports

> ll2.6.1, *In Situ hsting oflidecs.' lxtter to V Stello, l Jt., U.S. Nucleat Regulatory Commission, Washington, AEODIC404. W. D. Lanning. July 1984. Sicam liinding  ;

DC ofAurillary Fredwater Pumps. U.S. Nuclear Regulatory  !

" i Commission, Washington, DC 13tooks,lt P.1988 Application Guidchncsfor Check Vcdecs in Nuclear Power Plants. NY-5479, Electric ABODiC602. C iIsu. August 1986. Operational _

Power Rescarch Institute, Palo AI o, California. Opcrience Inroh ing Turbine Orcrspccd Trips. - ~U.S. No- ]

clear Regulatory Commission, Washington, DC Casada, D. A.1989. Annitiary Fc?awatcr System Aging . .

Study: l'olume L Operating Experience and Current AEODlCHO. E.J.11rown. December 1986. A Review Aiunitoring Practices. NUREGICR-5404. U.S. Nucicar ofAfotor-Operated Valve Performance. U.S. Nucleus Ryulatory Commission, Wasnington, DC Pegulatory Commission, Wtshington, DC Gre;*g, R. E. and R E. Wright.19S8. Appendn Revicw AEOD/E702. E. J. Brown. March 19, D87. Af0VFail.

for Dominant Generic Contnbutors. dLU 31~88. Idaho ute Due to Hyuraulic lxckup From Ercessive Grease in - ,

Nationali ..rcering 12boratur,, Idaho Falls, Idaho. 3pringPuck. U.S. Nuclear Regulatory Commission, Washington DC Miraglia, E J. February 17,1988. Resolution of Generic Safety issue 93,

  • Steam Rinding ofAntiliary Fredwater AECD1(416. 3anut'ry 22,1983. Loss afESFAuxiliary 1%mps' (Gcncric Letter 88-03). U.S. Nuclear Regulatory Fredwater Pump Capability at Dojan on January 22, Commission, Washington, DC 1933. U.S. Nucicar Regulatory Commission. ,

Washington, DC Mitaglia, E 3. August 8. l988. Instnament Air Supply System Problems AJJecting Fafety.Related Equipment

- (Generic Lettcr SS14). U.S. Nuclear Regulatory Com. Information Notices mission, Washington, DC IN 82-01. January 22,1982. Auxiliary Fcedwater Pump Partlow, J. G. Junc 28,1989. Safety-Related Afotor-Op- Lockout Resulting[om Histinghouse W-2 Swin h Circuit crated I alve Testing and Sareciliance (Generic Letter 89 Afodification. U.S. Nuclear Regulatory Commission, 10), U.S. Nuclear Regulatory Commission, Washing. Washirgton, DC ton, DC .

IN 84-32. E. L Jordan. April 18,1984. Anriliary lied--

Ro*bberg.O 3une 1988. Thermal Overload Protection water Sparger and Pipe Hangar Damage, U.S. Nud:ar .

for Electric Alourrs on Safety-Related Afotor Operated Regulatory Commission, Washington, DC ihlecs - Gcncric issue llE6.1. NUl'EG-12% U.S.

Nusicar Regulatory Commission, Washington, DC IN 84-66. August 17,1984. Undetected Linavailabihty of the Turbine Driven Autiliary Fredwater Dain. U.a. Nu-Thvis, R and J. Taylor.1989, Development ofGuid- clear Regulatory Commission, Washington, DC k

antefor Gcneric, Ematlonally Or nted PRA-Based Team inspections for BiVR Plants.identipcation of Risk- IN 87- 34. C E. Rossi. July 24,19s7, Single Failures in important Systems, Comp <ments and 1fuman Actions. Auriliary Fredwater Systems. U.S. Nuctaar Regulatory ,

TLR-A 3874-T6A lhokhaven National L2 oratory, Commission, Washington, DC 1 i

Upton, New York.

6.1 NUREG/CR-5828

Refctences IN 87 53. C E. Rossi. October 20,1987. Auriliary inspection Report Fredwater 1%mp Trips Resultingfrom law Suction Pres-sure. U.S. Nuclear Regulatory Commission, IR 50-489/89-11; 50-499/84-11. May 26,1989. South Washington, DC Texas hoject Inspection Byort. U.S. Nuclear Regulato-ry Com nission, Washington, DC IN 88-09. C. E. Rossi. March 18,1988. Reduced Relia-bility of Steam. Driven Auritiary Tecdwater Pumps Caused byInstability ofIt'uodwardiGPL Type Governors. U.S. NUREG Report Nuclear Regulatory Commission, Washington, DC NUREG-1154.1985. Loss ofMain and Auxiliary Feed-IN 84-30. R. A. Azua. August 16,1989. Robinson Unit water Event at the Davis Besse Plant on June 9,1985. ---

21nadequate NPS1/ ofAuri!iary feedwater Pumps. Also, U.S. Nuclear Regulatory Commission, Washington, DC Event Notification 16375, August 22,1989. U.S. Nucle- -,

at Regulatory Commission, Washington, DC 4

NUF EG/CR-5828 6.2

Distribution No. of No. of Copin Copies OITSITII 4 ANO.2 Resident inspector OJJ,Et UE. Nockat Reculatory Commission J.11. Thylor Brookhaven National 1.aboratory B. K. Grimes Bldg. "')

OWlH 9 A2 Upton,NY 11973

! . E Gmgel R.11ravir OWFN 10 E4 Brookhaven Nationallaboratory Bldg.130 _ _

R. Barrett Upto9,NY 11973 OWFN 10 A2 .

J.Bickel A. En Bassioni EG&O Idaho,Inc.

OWFN 10 A2 P.O. Box 1625 Idaho Palls,ID 83415 W.T Russell OWFN 12 E23 Dr. D. R. Edwards ,

Professor of Nuclear Engineering K. Campe University of Missouri- Rolla OWFN 1 A2 Rolla,MO 65401 10 J. Chi..,g ONSITE OWN 10 A2 .

26 MRNjorthwest 1.aboratcry l 2 B. Thomas OWFN 12 H26 S. R. Doctor L R. Dodd

U.S. Nuclear Reculatory Commission - Region 4 B. F. Gore (10) l R. Pugh (5)

. A. B. Beach B. D. Shipp L J. Callan E A.Simonen S. J. Collins T.V.Vo J. P. Jaudon Publishing Coordination .,

R. D. Martir. Tbchnical Report Fi!c (5) l T. R Leika t

Distr.1 NUREO/CR-5828 1

i-.--- .,-....,,-.,+4 ..-g[ .- pe p mg. .,.

y q g w., $. y-w w e g o w #y myg 9,, yyq u g pmr,t .gy g p q,.gs erkmg +gw.nr

Nnc FOAM 335 U1 NUCLI AR REGUL ATOR V COMMI5140N 1. RE P0mi Nung g a

!!abca.  !^.'o".*.*: ".",t,*,*',7g ""-

e nu BICf.lOGRAPHIC D ATA SHEET n,, ,om ,, ,~ ."" '

. NUREG/CR-58?8

2. TIT LE ANU LUpt T L6 pg{,,77g7 Emergency feedwater System Risk-Based Inspection Guide for the Arkansas Nuc. ear One Unit 2 Power Plant a cats REroRT PUsussto w j ...

September 1992

4. FIN OR GRANT NUM6f.R L1310
h. Ad l. TOR (5) IL T YPE OF RtPOR T R. Pugh, B. F. Gore, T. V. Vo Technical

~

r.rsRioucovERso -s,.- e. .> \

12/80 to 10/89 l

. _ __ 1

! g. pg Ra OnMING OHGAf4ZAT LON . f4AML ANQ ADDRLM fn 44C.a=== Domme. On.,, y me, , ,, ap y c - , ene mas y marmee, y a.now,w, em.m j none e.* mean **wmo .l l

Pacific Northwest Laboratorv l Richland, WA 99352

9. 5PONLORING CRG ANIZATION - HAME Atc ADDRf.hs af /vec. enie *1.=< m em..e ;# e.aen.cm. m m ava(&wea,0u A.,,, u A a , s., ,f c.m e.w e.as.ne sawr.esJ i

Division of Radiation Protection and Emergency Preparednesa OffItc of Nuclear Reactor Kegulation U.S. Nuclear Regulatory Commis ion Washington, DC 20555

10. SUPPLEMENT ARY NOTES l
11. A85 TRACT (M =ee w ras 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 pressarized 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 fnr various plants was used to identify generic component failure modes.

~

This information wes then combined with plant-specific and industry-wide compnnent l information and failure data to identify failure modes and failure mechanisms for the -

AFW system at the selected plants. Arkansas Nucicar One Unit 2 was selected as one

( of a series of plants for study. The product of this effort is a prioritized listing-of_AFW tailures 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 Arkansas Nuclear One Unit 2 plant.

j n = E r wo a oi~c uc R.r i oR s <t ., -. w ,-- , =.u . . -. ,

u. ....u..u i v o .a m t-Inspection, Risk PRA, Arkansas Nuclear One Unit ?,

Auxiliary feedwater ( AFW)

,, [ [ ted

,,w,<,,.

Unclassified ur . u Unclassified Ib. NVMS&R 06 PA(,t$

7 6. P RICi

+ ,

( hNU hN hSN

h 6

Printed on recycled paper Federal Recycling Program i

. ___ l SEFIEMBER 1892 ^

NUREG/CR-fB28 EMERGENCY PEEDWATER SYSTEM RISK-BASED INSPECTION GUIDE -

FOR THE ARKANSAS NUCLEAR ONE UNTT 2 POWER PIANT _

-e UPc : ED STATES - nRST CLASS MAR.

POSTAGE AND FEES par 0 NUCLEAR REGULATORY COMMISSION usNac WASHINGTON, D.C. 20555-0001 I I PERMIT NO. G-67

' l Of7tCIAL E .itNESS P7Nr.LTY FOR FHIVATE USE. 5300 l

I 1

l 1

I

)

l l

i l

, 7 g o.i ' O G 31 #

((S p ;n% 13 e. O f:s"E:$d-0E{" svC-  :

f tIraTI^Ng nIV #0I8 '~

T P ';- FO ~ "' R r r. . - r'#r #' l p-71I vC

,gsv1NGT_t a

~ _- i y g ,

Retrieved from "https://kanterella.com/w/index.php?title=ML20116A474&oldid=2808579"