ML20118B241

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High Pressure Core Spray Sys Risk-Based Insp Guide
ML20118B241
Person / Time
Site: Clinton Constellation icon.png
Issue date: 08/31/1991
From: Travis R, Villaran M
BROOKHAVEN NATIONAL LABORATORY
To:
Office of Nuclear Reactor Regulation
Shared Package
ML20118B240 List:
References
CON-FIN-A-3875 A-3875-T5A, NUDOCS 9210010228
Download: ML20118B241 (111)


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- Technical Report ,

A 3875 T5a lilGil PitESSUltE COllE SPitAY SYSTEM ,

IllSK.llASED INSI'EC110N GUIDE i

Prepared by M. Villaran and it. Travis llrookhaven National 1.aboratory Department of NucIcar Energy Engineering Technology Division Upton, NY 11973 August 1991 Prepared for -

U.S. Nuclear llegulatory Comminion Office of Nuclear Reactor llegulation -

Washington, DC_ 20555 IIN A.3875 i

i-l-

l l ..

I

. . - , , ~ , , . _ m___ _

CONIENIS l

l'agg 1-1 1 INTitODUCllON .. .. .. . .. . . .

Purpose . . . . .. 1-1 1.1 ... .....

1.2 Scope ... . ,, . ... . . . . .. ... . .. 11 13 Application to inspections . .. . . . ......, . 11 2 GENEllAL 1IPCS SYSTIIM DESCitil'IION ..... .. . . 21 3 ACCIDENT SEQUENCis DISCUSSION . . . .. . . 31 3.1 loss of liigh Pressure injection and Failure to Depressurire . ... ... .... . . . 31 -

3.2 Station illackout (Silo) with Intermediate Term Failure of liigh Pressure Injection . 32 3.3 Station illackout with Short Term Failure of Iligh Pressure injection . . ... . 3-3 3.4 NIWS With Failure of ItPV Water Level Control at liigh Pressure .. .. .... ... . 3-4 3.5 Unisolated 1.OCA Outside Containment . ... . . . . 35 3.6 Overall Assessment of IIPCS Importance in the Prevention of Core Damage . . 3-6 4 Pila-IIASED llPCS FAILUltE MODES .. . . 41 5 OPEllATING EXPEltlENCE IttiviEW .

5-1 5.1 IIPCS Failure No.1 - Pump Fails to Start or Itun . . ., . 5-3 5.2 ilPCS Failure No. 2 System Unavailable Due to Test or Maintenance Activities . 59 -

5.3 IIPCS Failure No. 3 - IIPCS injection vahe Font Fails to Open .

. 5-11 5.4 IIPCS Failure No. 4 - CST / Suppression Pool Switchover logic Fails . . . 5-11 5.5 llPCS Failure No. 5 - Suppression Pool Suction Valve F015 Fails to Opei. .

5 12 5.6 IIPCS Failure No. 6 - Manual Injection Valve is Plugged / Clogged . . .. . 5-12 5.7 IIPCS Failure No. 7 - Minimum Flow Valve F012 Fails to Open 5-13 5.8 IIPCS Failure No. 8 - CST Suction Check Valve F002 Fails to Open . 5-13 5.9 IIPCS Failure No. 9 - CST Suction Manual Valve is Plugged / Closed . . . 5-14 5.10 llPCS Failure No.10 CST Suction Valve I 001 Fails Closed 5-14 iii

CONTENTS (Cont'd) _

bze 5.11 IIPCS Failure No.11 Pump Discharge Check Valves (F024 or F005) Fail to Open . . . . . . . . . . . . . . . . . . . . . . . . . 5-14 5.12 IIPCS Failure No.12 Suppression Pool Suction Check Valve F016 Fails to Open . . . . . . . . . ............. -5 15 5.13 IIPCS Failure No.13 - False low Suction Pr e ssu r e Trip . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 15 5.14 IIPCS Failure No.14 System Actuation logic Fails . . . . . . . . . . 5 15 5.15 IIPCS Failure No.15 - Suction Strainer Fails to P a ss Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 5 16 5.16 Comparison of Operating Experience to PRA-Based Ranking......................................... 5 16' 6 OTilER SYSTEM CONSIDERATIONS . . . . . . . . . . . . . . . . . . . . . . . . . 61 _

6.1 I l u m a n Error . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2 IIPCS Support Systems . . ........ ., ...... . .... 6-3 6.2.1 IIPCS Diesel Generator (DG) and Division til AC Pow e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i>-4 6.2.2 Division ill DC Power . . . . . . . . . .................. 6-5 6.2.3 Room Cooling and Ventilation . . . . . . . . . . . . . . . . . . . . . . 6-6 6.2.4 System Actuation Instrumentation . . . . . . . . . . . . . . . . . . . 6-6 6.3 IIPCS Systems Interactions . . . . . . . . . ................., 6-6 6.4 Simultaneous Unavailability of Multiple Systems . . . . . . ... .. 6-7 6.5 Valve Failures in liigh Pressure injection Systems . ....... 6-8 6.6 LOCA Outside Containment . . . . . . . . . . . . , . . . . . . . . . . . . . . 6-8 7

SUMMARY

. , .. ....... ........ ...... ... .... . ... 7-1 8 REFERENCES ....... ... ...... .. ... ..... . .... .. 8-1 APPENDIX A Plant Specific System Informition . . . . ................... A-1 Al Clinton Pow:r Station llPCS System Details . . . . . . . . . . . . . . . . . A1-1 A2 LaSalle County Station Units 1 & 2 IIPCS System Details . . . . . . A21 A3 Nine Mile Point - Unit 2 IIPCS System Details . . . . . . . . . . . . . . - A3-1 A4 Perry Nuclear Power Plant IIPCS System Details . . . . . . . . . . . . . A4-1 A5 River llend Station - Unit 1 IIPCS System Details . . . . . . . . . . . A5-1 A6 Washington Nuclear Plant No. 2 IIPCS System Details . . . . ... A6,1 A7 Grand Gulf Nuclear Station - Unit 1 IIPCS System Details . ... A71 APPENDIX B Proposed Inspection Plan for Diesel Generators . . . . . . . . . . . . . . 1141 iv

?

FIGURES

' Fleure No. fage 21 Simplified IIPCS Flow Diagram . . . . , . . . . . . . . . . . . . . . . . . . -. 22 ,

6-1 LaSalle County Station, Unit 1.- 125V DC ESF Division III . . . . . . 6-5 62 Percent of valve and valve operator failures reported to l LER and NPRDS databases . . . . . . . . . . . . . . . . . . . . . . 6-9 T/illLES Inble No. f. 2EE 4-1 IIPCS PRA flased Failure Summary . . . . . . . . . . . . . . . . . . . . . . -42 51 Summary of Itisk Significant LElls . . . . . . . . . ._ . . . . . . . . . . 52 5-2 Illustrative lixamples of Risk-Important lil'CS Failures . . . , . , . . 5-4 61 Summary of Generic Valve Failures llPC1/IIPCS . . . . . . . . . . 6-10 7-1 IIPCS System RIG Summary . . . . . . . . . . . . . . . . . . . . , . . . 72

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  • _ . . _ . - - -- ~ _ . - - .-. - .- - _ . - - - - -.. . . . ~ . - - . - . . - - - . _ . . - . - . ,
1. INTi(ODUCTION 1.1 Pumose This llPCS System Itisk.Itased Inspection Guide (S IllG) has been developed as an aid to NitC inspection activities at the llWlV5 and llWit!6 plants utilizing ilPCS as a means of high pressure injection into the reactor vessel. 'the document presents a risk-based discussion of the llPCS role in accident mitigation and provides Pila-based IIPCS failure modes. Most PI(A-oriented inspection plans end here and require the inspector to rely on his experience and knowledge of plant specific and ilWil operating history.

'lhe system IllG goes a step further, however, by using industry operating experience, including illustrative examples, to augment the basic Pit >A failure modes. 'Ihe risk based input and the operating experience are combined to develop a composite llWit ilPCS failure tanking. This information nmy then be used to optimite NI(C resources by alk,cating proactise inspection effort based on both risk and industry experience.

1.2 Scorse The Iligh Pressure Core Spray (llPCS) system has been examined from a risk perspective. Ilollowing a brief generalized description of the llPCS system for background information (Section 2), common llWit accident sequences in which ilPCS function is required are discussed in Section 3 both to review the system accident ' mitigation capability, and to identify system unavailability combinations that can greatly increase risk exposure. Section _4 describes the prioritization of the Pila.bastd llPCS failure modes for inspection purposes, and the results of a review of itWit operating experience are presented in Section 5 to illustrate these failure modes. Tliis inspection guide also provides additional information in related areas such as 11

... .-. ., . _ _ _ - . - . , - _...J.. . - - - , , - . - , . . , _ . - - . , - - - , , _ . . . . - . . . . . . ~ - - , . - -

p ilPCS support systems, human errors, cystem interactions, and valve failures (Section 6), A summary and ,

list of references are provided in Sections 7 and 8, respectively. Modified (based upon risk) IIPCS system walkdown tables specific to the 1(PCS systems found in each individual plant are provided in Appendices A.1 through A.7. A proposed inspection plan for diesel generators at nuclear power plants is given in Appendix D as a guideline for inspections which encompass the IIPCS diesel generator.

1.3 Annlication to Intnections This inspection guide can be used as a reference for both routine inspections and for identifying the significance of component failures. The information presented in Section 5 can be used to prioritize day to-day inspection activities and the illustrative IIPCS failures can provide multiple inspection perspectives.

The system RlG is also useful for NRC inspection activities in response to system failures. De accident sequence scenarios described in Section 3, in conjunction with the discussion of multiple system unavailability (Section 6), provide some insight into combinations of system outages that can greatly increase risk. Within the context of the IIPCS system, the operating experience review provides examples of several failure mechanisms (together with the corrective actions implemented and potential areas for inspect!ca) which are useful for the review of licensee response to a system failure. The system RIG can 3

also be used for trending purposes. Table 5-1 provides a summary of the IlpCS operating experience, in -

particular the industry wide distribution of IIPCS failures. The summary section presents a compilation of the industry experience with IIPCS up to mid-1990. Rose !!PCS failure modes which account for a larger fraction of the IIPCS system failures are candidates for increased in>pection activity. Since the plant-specific failure distribution is expected to vary over time, the summary includes a mechanism to update and trend each individual plant's liPCS experience, in comparison to the more static industry experience.

1-2 4

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i 2, GEMERAL IIPCS SYSTEM DESCRIITION The High Pressure Core Spray (HPCS) system is a single train, high pressurc injection syt, tem used in BWR/5 and BWR/6 plants. It includes a single electric motor-driven pump, associated valves, piping, and instrumentation to provide cooling water to the reactor core. A simplified flow diagram is provided in Figure 2-1. The HPCS System, in conjunction with other ECCS, is designed to cool the reactor core sufficiently to prevent fuel cladding temperatures from exceeding 2200*F following any -

break in the nuclear ,ystem piping. The HPCS System is designed to pump w"er into the core over a wide range of pressures from SRV lift pressure (typically up to 50 psi, differential) to below 200 psid. ,

For small breaks (less than one inch diameter), HPCS is capable of maintaining reactor water level above the top of the core, and preventing actuation of the Automatic Depressurization System (ADS).

For large breaks up to and including the Design Basis Accident, HPCS plus either Division I ECCS -

(RIIR A and LPCS) or Division II ECCS (RHR E and C) is capable of providing adequate core cooling.

HPCS also serves as a backup to the Reactor Core Isolation Cooling (RCIC) System to supply makeup water to the reactor vessel in the event of a reactor isolation.

The HPCS system takes suction from either of two sources: the Condensate Storage Tank (CST), or the suppression pool. The primary source of water for the HPCS System is the CST. The suction line taps into the CST via a locked open manual valve at a los.er elevation than any other CST suction lines to ensure an adequate minimum reserve capacity for the llPCS system, as well as the RCIC system, which taps off of the HPCS suction line. The normally open-HPCS CST suction line isolation valve F001,8 and CST suction line check valve F002 are located downstream of the RCIC tap r off.

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8 Standard GE valve designations are utilized throughout this system description.

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The alternate source of water to the HPCS system is the suppression pool. A stdnless steel

. mesh strainer is provided at the source of the suppression pool suction pipe. This strainer is designed to maintain adequate NPSH for the HPCS pump with up to fifty percent of the strainer surface area l l

plugged. The suppression pool suction flow passes through isolation valve F015 and a checkvalve F016 -1 to the HPCS pump.

( Upon system initiation the CST suction line isolation valve F001 will open, providing a source ,

from the CST. Should a CST low level condition occur HPCS, suction will switch over to the suppression pool automatically after a brief (approximately two second) time delay. The time delay prevents inadvertent shifting due to the pressure transient when the HPCS pump starts. Once the suppression pool suction isolation valve F015 is fully open, the CST suction isolation valve F001 will close. Automatic switching of the suction source will also occur upon a high level condition in the s suppression pool.

HPCS pump discharge flow to the reactor vessel passes through a normally shut injection isolation valve F004, krated outside the containment, and a pneumatically testable check valve F005 and a locked open injection stop valve (F036 in BWR/6 or F038 in BWR/5), located inside the drywell.

The piping enters the reactor vessel above the shroud and separates into two lines which curve around' the inside of the vesselin opposite directions, The two lines turn downward 180* apart and penetrate the shroud just above the top of the core,; The two semi-circular spargers contain nozzles which direct the flow to the top of the fuel bundles in the reactor core to remove decay heat following a postulated loss of coolant accident. The HPCS system will initiate automatically on either high drywell pressure L.

or low reactor water level (level 2). In the event the HPCS system is in any mode other than standby when an automatie initiation signal is received, all valves will realign for- the injection mode-of .

L operation. HPCS system injection into the reactor vessel is automatically terminated-when a high -

reactor vessel level (level 8) is reached by automatic closure of the injection isolation valve (FON).-

2-3 d

i i ---+-. - - -._._, .- .

_ , . , .. ~.-_ ,-._.,_ ,, _ .___. . . - ,_ _ , ,_,,,,.m.,

When reactor vessel level again dropa to level 2 the injection valve automatically opens to commence -

injection into the reactor.'

- A minimum flow line is provided to assure that the minimum flow requirements for the HPCS pump are met to avoid impeller damage. The pump minimum flow line passes flow through restricting orifice D001 and minimum flow control valve F012 to the suppression pool. The flow control valve F012 position is controlled at most plants by signals from flow element FE N007, and Pressure Transmitter l'r N051.

Two full flow test lines provide the capability to test the HPCS system while discharging to either the CST, through the CST test valves F010 and F011, or the suppression pool, through the test to suppression pool valve F023. A restricting orifice (D004, D005) is installed in each of these test lines to simulate the back pressure seen when pumping into the reactor vessel. Valve interlocks prevent establishing a test flow path from the suppression pool to the CST.

The HPCS system utilizes a " Keep Fill," " Jockey," or " Water Leg" pump to ensure that the piping between the injection valve F001 and the HPCS pump discharge check valve F024 is maintained full of water. This minimizes the occurrence of water hammer (due to the voiding of system piping) and minimizes the time it takes for the llPCS system to actually deliver water to the reactor vessel

. following initiation. - Normal power for the HPCS ' system is provided from the Division 3 bus. In the event of a total loss of normal power, this bus is powered by the llPCS diesel generator. The llPCS system w.'ll deliver rated flow into the vessel within 27 seconds following receipt of an initiation signal.

Plant specific details for the llPCS system are provided in Appendix A and References 40-63.

24

3. ACCIDENT SEQUENCE DISCUSSION s

The role of the llPCS system in the prevention of reactor core damage during abnormal plant conditions provides valuable information that can be applied to normal day to-day inspection activities.

If a plant has its own Probabilistic Risk Assessment (PRA), this information is readily available. Not only are plant specific design and operating nuances considered, but the accident sequences, sy.tems, and component risk importances are generally quantified and prioritized.

Since most plants do not currently have PRAs, the application of risk insights is less straight-forward. - An ongoing PRA-based Team inspection Methodology for the Risk Applications Branch of NRR has developed eight representative BWR accident sequences based on a review of the available ORAs [1]. Because of design and operational similarities, generic risk insights from these representative accidents can be applied to other BWRs for risk based inspections. This information can be used to alk>cate inspection resources commensurate with risk importaace. In addition, if single or multiple systems are degraded or unavailable, this methodology can he used to designate those accident se-quences that have become more critical due to the unavailability of a key system (s). This allows the inspector to focus on the remaining systems / components within a sequence to assure continued avail-ability and minimize plant risk. Five of the eight representative sequences require the llPCS system to function for mitigation or as a potential initiator. These five sequences are discussed below.

3.1- Loss of Hich Pressure Iniection and Failure to Depressurire l

- This sequence is initiated by a general transient (such as turbine trip with subsequent MSIV-

- closure, MSIV closure and loss of condenser vacuum, loss of main feedwater, inadvertent SRV opening l with MSIV closure, or a loss of offsite power), or a small break LOCA. The reactor successfully l

31

__- _ ~ . _ _ _ . _ - - - _ . _ _ . . - . _ _ _ . . _ _ _ . _ . _ _ _ . _ _ _ . _ . . _ . -

.a scrams. The power conversion system, including the main condenser,is unavailable either as a direct .

result of the initiator or due to subsequent MSIV closure. The high pressure injection .,ystems (HPCS/ --

RCIC) fail to inject into the vessel. The major causes of HPCS/RCIC unavailability include hardivare failures (primarily pump faults) and system outages for test or maintenance activities. The CRD hy-draulic (CRDH) system can also be used as a source of high pressure injection (HPI), but the failure of the second CRD pump or unsuccessful flow :ontrol station valving prevents sufficient RPV injection.

The operator attempts to manually depressurize the reactor pressure vessel (RPV), but a common cause failure of the safety relief valves (SRVs) defeats both manual and automatic depressurization of the reactor vessel. The failure to depressurize the vessel after HPI failure results in core damage due to a lack of vessel makeup.

3.2 Station Blackout (SBO) with Intermediate Term Failure of Hich Pressure Iniection This sequence is initiated by a loss of offsite power (LOOP). The division I and 11 emergency diesel generators (EDGs) are unavailable, primarily due to hardware faults. Maintenance unavailability is a secondary contributor. Support system malfunctions include EDG room or battery /switchgear room HVAC failures, service water pump, or EDG jacket cooler hardware failures.

The high pressure injection systems can provide inventory makeup until:

,

  • the Division til (HPCS) diesel generator fails to continue to run

. the systems fait due to environmental conditions; i.e., loss of room cooling and ventila-tion, high lube oil temperatures, or high RCIC turbine exhaust pressure due to t. high-l suppression pool temperature and pressure, or l 3-2 L

l i

l i

l_

l t -. __ _ - _ - . - _- _ __ . - _ . - -_

. the RCIC high area temperature logic isolates the system or long term exposure to high temperaturcs disables the turbine driven pump.

Plant procedures address means to maintain the HPCS diesel generator for as long as possible, to maintain DC power for as long as possible, to assure a continued source of water to the HPCS or RCIC, to ensure adequate tube oil cooling, and to provide contingency measures (such cs supplying fire water via RHR system)if the SBO progresses until reactor pressure (decay heat) can no longer support RCIC. The plant procedures should be consistent with the BWR Owner's Group Emergency Procedure __

Guidelines.

3.3 Station Blackout with Short Term Failure of Hich Pressure Iniection This SBO sequence is similar to the previous sequence except the high pressure injection sys-tems fail early. HPCS unavailability is dominated by pump failures and maintenance unavailability.

The sources of emergency AC power, i.e., the emergency diesel generators (EDGs) including the HPCS diesel generator fait primarily due to hardware failures. Secondary contributors are: output breaker failures and EDG unavailability due to test or maintenance activities. Support system malfunc- _

ticas, such as service water failures in the EDG jacket cooling water train, battery /switchgear room IIVAC failures, or test ar,d maintenance unavailability are significant contributors to the loss of all AC powe r.

Station battery failures (including common mode) are an important contributor to this sequence by definition, because the EDGs and high pressure injection systems are DC dependent. Core damage occurs shortly after the failure of all injection systems.

3-3

3.4 ATWS with Failure of RPV Water Level Cantrol at Hich Pressure .

1 f

This sequence is initiated by a transient with initial or subsequent MSIV closure and a failure

, of the reactor protection system. Attempts to manually scram are not successful, however the Standby Liquid Control System (SLCS) is initiated. By definition, the condenser and the feedwater system are o

unavailable. The BWR Owner's Group Emergency Procedure Guidelines (EPGs) recommend RPV water level reductions to control reactor power below 3% and the BWR representative sequence was based on that philosophy.

t This sequence postulates a failure to ensure sufficient RPV makeup at high pressure to prevent -

cot damage. The high pressure injection (HPCS) system fails, primarily due to pump failure to start or testing and maintenance (T&M) unavailability. Injection or minflow valves, suction switchover, or loss of electrical power are other system failures, HPCS pump failure to start or run, pump unavailabil.

4 ity due to testing and maintenance activities, and Service Water EDG jacket cooler inlet or return valve

failures are the major system failures, 3

At this point in the sequence, once HPCS has failed, with ADS inhibited, the remaining high

] pressure injection systems cannot keep the core covered at ATWS power levels. The operator fails to -

i manually depressurize in a timely fa3hion, and core damage ensues.

The continued operability of HPCS during an ABVS event is critical. Within the context of this accident sequence, (i.e., time available for success) the licensee capability to perform the logic bypasses should be evaluated periodically. With regard to HPCS system availability, the remaining sections will discuss system failures and availability evaluation.

3-4

'3.5 Unisolated I.OCA Outside Containment t-An interfacing LOCA initiator is defined as the initial pressurization of a low pressure line which results in a pressure boundary failure, compounded by the failure to isolate the failed line. The failure is typically postulated i 2 low pressure portion of the low pressure core spray (LPCS) system, the LPCI, shutdown cooling and (to a lesser extent), the IIPCS or RCIC pump suction, or the head spray line of RIIR rystem.

The unisolated LOCA outside containment results in a rapid loss of the reactor coolant system (RCS) inventory, eliminating the suppression pool as a long term source of RPV injection. Piping i

failures in the reactor building cas. also result in unfavorable environmental conditions for the ECCS.

Unless the unaffected ECCS systems or the condensate system are available, long term RPV injection  ;

is suspect and core damage is likely.

There have been several llPCI pump suction overpressurization events, primarily during surveil-i lance testing of the normally closed motor-operated liPCI injection valve [4]. This is of particular concern for the discharge configuration with a testable air-operated check valve in addition to the normally closed MOV because of the valve's history of back leakage. There is the potential for a simi- j lar situation to develop in the llPCS system which also utilizes a normalh closed motor-operated injec-I tion valve F004-in series with a testable air-operated check valve F005 in the injection line.

l L 'Several possible interfacing system LOCA precursors are included in Section 6, Other System Considerations. i f

35

-. . . _ _ . - . - . . . - . - . - . - . ~ . - - . - . . . - . . - - . - . - . - - - - .

3.6 Overall Assessment of HPCS Imnortance in the Prevention of Core Damace ,

As previously stated, the high pressure injection function (HPCS/RCIC/CRDH) is important in five of the eight representative BWR acciden' sequences. The various system failures and their impor-tances in all eight sequences were prioritized by their contribution to overall core damage frequency (using a normalized Fussell Ver.ely importance measure). The high pressure injection function,-in aggregate, was in the high importance category. Other high risk important systems are Emergency AC Power and RPS. The HPCS system itselfis of medium risk importance, because of the multiple systems that can successfully provide vessel makeup at high pressure. For comparison, other systems with a medium risk importance are: Standby Liquid Control, Automatic / Manual Depressurization.-Service -

Water, and DC Power.

3-6

4. PRA BASED llPCS FAILURE MODES PRA models are often used for inspection purposer .a ptioritize systems, components, and hu-1 man actions from a risk perspective. This enables the inspection effort to be apportiones based on a prioritization measure called risk importance. The HPCS failure modes for this system Risk Based Inspection Guide (System RIG) were developed from a review of BWR plant specific RIGS [6-10) and the PRA-Based Team inspection Methodology [1]. The component failure modes are presented in Table 4-1, grouped by risk significance. The potential human errors associated with HPCS system -

l availability are discussed separately in Section 6.

PRAs are less helpi.! in the determination of specific failure modes or root causes and do not generally are not intended to provide detailed inspection guidance. This makes it necessary for an inspector to draw on his experience, plant operating history, Licensee Event Reports (LERs), NRC Bulletins, Information Notices and Generic Letters, INPO documents, vendor information and similar sources to conduct an inspection of the PRA-prioritized items. To accomplish this task, the next section presents the results of a HPCS operating experience review. The aforementioned sources of HPCS information are correlated by PRA failure mode to provide illustrative examples which help to focus

. inspection efforts.

4 d

4-1

. I

. - - .- -- -. . - - . - - - . . - - - .~ - . - . - - -. - - , . . . . .

Table 4-1 HPCS PRA Based Failure Summary -

1 l

COMPONENTS 2 l

)

Hich Risk immrtance' Pump Fails to Start or Run l Svstem Unavailable Due to Test or Maintenance Activities i IIPCS Pumn injection isolation Valve F004 Fails to Open I Medium Risk Imrottance*

CST / Suppression Pool Switchover Logic Fails Suppression Pool Suction Valve F015 Fails to Open Normally Open HPCS Manual Injection Line Stop Valve (F036 in BWR/6, F038 in BWR/5) is Plugged or Closed Minimum Flow Valve r912 Fails to Open lower Risk immrtance' CST Suction Line Check Valve F002 Fails to Open l CST Suction Line Manual Valve is Plugged Normally Open CST Pump Suction Valve F001 Fails Closed or is Plugged Pump Discharge Check Valve F024 Fails to Open or Air Testable Check Valve F005 Fails to Open Suppression Pool Suction Line Check Valve F016 Fails to Open False Low Suction Pressure Alarm System Actuation Logic Fails Suction Strainer Fails to Pass Flow 2

See Section 6 for a discussion of IIPCS human errors.

3 The' Fussell Vesely importance Measure is used to rank the system components. This measure combines the risk significance of a failure or unavailability with the likelihood that the failure / unavailability will occur.

4-2

5. OPERATING EXPERIENCE REVIEW An HPCS system operating experience review was performed in order to compare actual industry operating experience with the PRA-derived failure modes for HPCS. At the seven BWR/5 and BWR/6 plants utilizing HPCS systems, seventy-five HPCS Licensee Event Report (LERs) were identified (through mid-1990) via the Sequence Coding Search System (SCSS). These were reviewed -

for applicability to the 15 PRA failure modes for HPCS; 23 LERs documented such HPCS faults or degradations. Table 5.1 presents a summary of the LERs categorized by failure mode, Fifteen LERs involved problems with the dedicated HPCS (Division 3) emergency diese! generator (EDG), five reported problems with the Division 3 AC bus, and one incident was associated with the Division 3 DC bus. In addition, twenty-four LERs reported problems such as: generic valve failures of HPCS valves, HPCS support systems, human error, simultaneous unavailability of multiple required ECCS systems, and potential LOCA situations. These events are described in more detail in Section 6. The remainder of the LERs documented occurrences such as successful system challenges, administrative deviations, and seismic or equipment qualification concerns.

The BWR/5 and BWR/6 plants which utilize the HPCS system have only been in operation since the early 1980s. The:e are only seven plants of this vintage, and devo.d have only just begun operating. For these rt. uns, the operating experience accumulated by the nuclear industry for the r

HPCS system is very limited as compared to the HPCI system, and the quantity of HPCS-related LERs-reported to date is not very large.

In order to enhance the comparison of the industry's HPCS LER operating experience with the -

PRA derived failure modes, the Nuclear Plant Reliability Data System (NPRDS) was also searched for .

HPCS-related events. NPRDS incidents which are related to or are potential precursors of the PRA -

j.

derived failure modes are highlighted in the following subsections (by definition, the most significant events are reported as LERs). By examining the NPRDS data, additional insights can be gained into-1 I

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. . 1 failure mec;.auisms or other problems which might impact risk important components in the HPCS system. This supplements the information provided by the limited LER experience base available at this time and provides greater confidence that the most critical items have been identified. l

.- 1 Each of the fifteen PRA-based failure modes. that has corresponding industry failures is  !

discussed below. Selected LERs and NPRDS failure events identified during the operating experience review are summarized to illustrate typicai failure mechanisms, applicable inspection methods, and potential corrective actions. Where applicable, other sources of background information are cited including NRC IE Bulletins, Information Notices, inspection Repo_rts, NUREGs, and AEOD Reports.

Selected illustrative examples of corresponding industry failures are provided in Table 5.2 along with details of the root cause, method of detection, corrective action taken, and potential inspection areas that could identify and prevent similar problems. This information can help inspectors by providing -

valuable insight into the types of problems which their assigned plants may already be experiencing, or anticipating risk-significant problems which have developed in plants of similar design to their own.

5.1 HPCS Failure No.1 FPCS Pumo Fails to Start or Run The major contributor to HPCI system unavailability, both from a risk and operational i

viewpoint,is the failure of the electric motor driven pump to start or continue running. The probleru areas which can lead to this failure mode can be grouped into four categories: 1) HPCS pump circuit breaker problems,2) motor / pump instrumentation and control (I&C) problems,3) motor and pump _

problems, and 4) loss of 4160V AC power to Division III. Table 5.2, Section 5.1 contains descriptions of five events associated with problems of circuit breakers and _I&C.

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P TaNe 5.2 Illustrative Examples of Risk-Important ilPCS Failures {

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f FRA Denved 14teetha Falwe Male Deecepkm of Ezestee Fatwe Method Ross Came Corrective Amion reamtid leepecekm Anas t gNcae I)  !

51 Purap Fails to 125dle 2 IIR 374 84 005TCS 8KI6t20446 IST Breaker PW4an switch (12LSk Restacal breaker p=en Obsesve sedode HPCS enneswtw teming.

Siart er Run A LIR 374 84 0181CS 84W2ttuut. Iknns made enables treaker desing smetch (5215) Otmerve gertale breaker mentenance atd spters innerdre aesting, pasp dettet brehker arriet when tweaker is rnaed tip Analymi faded switA 6nspeckus fakd to dine. 6nto putimi, was %We insreded breaker poet 6an Rewlew postaantenance sesong resgruns for j adpumedtne erwect}y 6nstaDed seitdies in all Divleian III deddcal,_ u .A 4160V switchgear.

Grand Gulf 1 Ira 416 84 ottCS IST hms tweaker was eq=ned due to Cormsed brea6er puities. . Revtew eq=pnent tyndt pwedwes.

I 84031h059.While tagging exa llPCS dead incoruct inforreation m eledricd Curuded eleddcal kneup Redew terrimed RO and tederddart traning in for mantenance cuntn.4 peer to the 18FC3 knear for itPCS deed ciperating shees. tagoas and retten to scrutce pmcedwes and pasp =as hat. Anmewskm. indrpendent vedSestka of safety seemed

  • eqdprnent puities.

IIPCS-NPRDS 5,0499118}Dudsig , ST To be determined fetwa duassemNy Regenced racsor. Verify elecedcol samntensnwieming procedwe in surmiLance testmg. stefks <+serwd shwamg inspectiora. thsasacraMe nunor to acconsance alth IEEE 112 1984 ma frors utger part c4 HKS pianp mmor, deternene cause of fatwe Observe HPCS insea4= surwi3mna semig i instat for adwese errvisonraent: bankary.

detAlust. emer or debne farms fnwa abwe rarter, esc. i

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ST ILgh dropout cmtacts in T* stase Regeac=d the relay. Obserw feriode maintenanceAestinginspedien for  ;

IIPCS NPRDS 9/197 [18). Dunns swwitace testing. orerators received itPCs c=erenred relay for HPC$ purap ptsealw rdaying en safety c isted sysrerna,  !

inspwt for maverse ansenewment at suutedsgear. i purup anator <werewrent alarm and IIPCS muor weve sustared to have tan purnp metrte rasnual <=ernde nians. midung domed intermittenty detA$at. high henidre. wrepensure oil vapor. [

  • satt nar, etc.

Note 1 Abbreviatione for deted6on reethod are as faHms.

15T-laseMce Testing er InseMa inspectum ST- Surmaance Testing SI ' Si edal inspedim AIRM - Ado er Wsual alana er amudater f RO - Resine Observatson OA - Chiermianal abnormdity CM -Cormtive voaintenance FM - Prewnt.ative maintenmice h 'f i  !

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i Table 5.2 Illustrative Examples of Risk.Irnportant IIPCS Failures (Cont'd) l 4

f PRA De,twd Deeeessa ,

fWure Male Desempkm of Earnde Falore Meth<mf Rott Cause Cared.we Aaion rumtial lastweim Areas (Nde 1) I 5.2 Sysaces tinavat- Grand Outf 1 11R 416 82 091DCS SI Inadequate seseming sJter wahe RMeed retest cewrul ' IsMew mms 8cmion swoordure for saferywisted able due to Tes 821}090163 andifR 416 82 09tOCS pontam indcaror modGcatim; gmwdaare, eqsaginent, '

s M4iaenan e S2llionen inmicquase pneenod&casim falure to rernowe info tag warning itern(aed infoneati<= ean Re*6e* P"m for Sedfying pmfer ECCS retenting folkming maiGcaska of a StPCS operwrw noe so way on postaan Resemed litCS synere sycens sequired by teerdest siwaScatan vak pcultion inacasar resuhed in Irwem:t ladcasor ifrC$ vatwe bneup dudrig sorwiBance test

-(

anJ tednical speci6catie violatim Ia$alte 21FR 374 84 018DCS 840920F2L IST Breaker pasition setts ($33h Repawd tvreaker petim Oberrve performane of ciecadca! mantenance  !

Dudng sptere enservice testing, Pamp civetet whi4 enaNes tweaker skning satich (533) erspletkm wd$caticut smedure f twaker fahd to clase. circuit eemi breaker is raded up Anadymed fakd sett&_ Rewtew parensiesenance testing p=grarm fw into pntim, was om correctly inspect breaker petka alcedeal st9sntenance. ,

i inw alled. se4 tees in eD Db4ske IU j 4160V seit6 gear.

  • Perry 1.IfR 440 88 027OCS 88fMMC84. AlJtM !sakage of =mer into e.ansnruce Repaced tranarwner. Evalusse effedr=eneas of cormalwe edism. I IIPCS drdami inop dae to emmeous IIPCS ' foritPCS Ene twak enordtoring Scaled cedult to transmitter kne break alane. Inertsment. Source of miser could to
tp sum be determineti gwweic . smer ladrumm tn River Bend 1 11R 458 86 0540CS Si liman error Unknoan Rewtee operator trardng on Rosernowne anakg [

8103120091 While urut at full prmer. an trip systern IIPCS lewt 2 initamim insanament was Venfy that pnxedures for Nains inoperaNe .{

dedsmi lacperaNe, and las asex$ated maner maner%lswe edy undts into ingyd condtion  !

trtp unit oss ytaced in the tripped creations are mJequase, [

per ted specs. This also made inoperable an RMes wrrmaer trening ees ECCS and inciation ,

HfCS Icwel 8 slave tdp unit also led frors thas namnenente ted speca. t maner trip unit. IIPCS should then hswe ,

been dedared 6noperaNe. taa was act [

ST Mator og erator berat sm4t&cs neve l

5.3 Iniccion Vak IIPCS NPit DS, 3!2a48 [le) 1 haring Adjusted Enut su4tsee, kew6ew PM pregrare for safety-velated vatve t Itaid Fais ta survn8ance test alwe Ftuh4 musor oreraner out of mi;usarment. revested opermares. [

Oprn, tripped on thennal ovettomt then ardung and rettarned to sen4ce Vevify analyus er trendng of surwe Bance sees open. results i for safety.: elated vehe egerators.

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IST . insen4ce Tennas or Innervice laspecaim ST . Servettiance Tesing St . Spedal lespectkwi A1JtM - Audo or visual alane er annwintor  !

RO . Routine Observatim OA - Opersional almormahry CM . Cormsive emnt . ance FM . Preventaniw maintenance a

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i 5-8

I No LERs were repnted concerning the IIPCS pump, however, experience on llPCI pumps has indicated that time depend e g idation mechanisms such as fatigue, high stress, wear, and erosion can lead to pump failure, noe most common pump failure . mode was low injection flow l11). Accordingly, ilPCS pump surveillance testing and AShtE XI in service testing would be the most likely means of detecting l{PCS pump problems.

                                                                                                                                          ]

The fourth category affecting failure of the llPCS pump to start and run is loss of 4160V AC r power to the Division ill bus. The causes of loss of AC power are quite diverse and are outsid: the  : scope of this inspection guide. The special case of the llPCS (Division lii) diesel generator is discussed briefly in Section 6.2. - 5.2 IIPCS Failure No 2 - System Unavailable Due to Test or hiaintenance Activities in addition to component failures, the system may not be functional due to testing or maintenance (T&ht) activities. In a single train system like 11PCS, test and maintenance activities on one compment usually disable the entire system. It is important to keep the time spent in IIPCS T&hi , nethity as low as possible because ofits direct contribution to system unavailability. This unavailability is defined as the number of hours that the system is unavailable due to T&ht activitu divided by the number of houts the system is required to be operable. I The development of a plant specific system unavailability model and subsequent tracking of unavailability trends is strongly recommended, since the variation in system unavailability from plant to , plant (even between seemingly similar plant designs operated by the same utility) can be significant (see Reference l64] for an example of trending safety system function via risk based performance indicators). As a related example, while conducting an inspection of the Limerick Generating Station, Unit I [13l, l Region 1 inspectors determined the IIPCI Cycle 1 T&ht unavailability was 6.4E 2 compared with a 5 l l i.

i

                                                                                                                                         . t PP A assumption of 1.0E 2 using previous Peach Bottom experience Based on the inspector's finding,                     .

the licensee began the development of a methodology for tracking a measure of the system total  ; unavailabilities and for evaluating, on an ongoing basis, the effects of increased unavailability on total  ; core damage frequency (CDF). The root sources of excessive llPCS T&ht unavailability were examined as part of this operating experience review. The 11 }{PCS LER examples of test or m.atenance errors were dh!ded into three categories: 1) inadequate mnintenance or post maintenance testing (6 occurrences),2j human crror that I . inadvertently or incorrectly disables the llPCS system (2 events), and 3) system inadvertently disabled during testing activity (3 occurrences). In eight of the LER events (73%) associated with T&hi activities, the affected plants were i operating at power. The detection of the problems was by means of special reviews or inspeeth ts 45% of the time, and by operational abnormalities or incidental observation 27% of the time. A umilar ! distribution of detection means and system status at the time or the occurrence was noted for six l l NPRDS events in the T&hi area. In summary, the T&ht component of system unavailability must be continuously monitoreit by the inspector to assure it is as low as possible. The licensee should be administratively limiting the number of times that the liPCS system is in test or maintenance during operation. System restoration should be vigorously pursued: 11PCS should not be down for days, if it can reasonably be repaired in hours. Whenever it is feasible, pardons of the system should be tested during outages. In addition, ilPCS unavailability can also be minimized by adequate root cause analysis and effective corrective action to prevent the recurrence of system outages due to the same types of failuies, and thorough, efficiut work planning to avoid unnecessary removals from service or inadvertent system isolations during calibration or surveillances. 5-10 l l

l 5.3 ljPCS Failure No 3 Initetion Valve FDn4 Fails to OpeD The llPCS injection valve F004 is a normally closed, AC motor operated valve. It opens automatically upm system initiation. It closes whenever reastor vessel water le -l rises above the level 8 setpoint. The failure of this valve to open disables llPCS injection into the reactor vessel. One IIR (416 85 050/DCS SI/14070413) reputed the failure of the injection valve (F004) to open automatically when required at Grand Gulf. A reactor scram had occurred when vessel knel dropped to level 3 following the trip of all condensate and feedwater pumps. Operators manually initiated ilPCI and RCIC to restore level. The llPCS injection valve failed to open automatically because of a failed Agastat model CR 0095 relay base; however, the operators opened the valve with the control room handswitch. There were four related occurrences reported on NPRDS [18),2 of which are contair-d in Table 5.2, Section 5.3. Selected illustrative valve failures that have occurred in high pressure injection systems, induding root cause and corrective actions, are presented in Section 6.5. 5.4 }{PCS Failure No,4 - CST /Supprewinn Pool Switchover t ocie Fails in the standby mode, the llPCS pump is normally aligned to take suction from the condensate storage tank (CST). Upon receipt of a low CST level signal, or a high suppression poollevel signal, the suppression pool suction valve F015 automatically opens with subsequent closcre of the CST su. tion valse F001. System operation then continues with the 11PCS pump suction from the suppression pool. As indicated in Table 5.2 Section 5.4, failures of suppression pool level instrumentation and suppression pool high water level occurrences indicate that this area should continue to be monitored by NRC inspectors. 5-11

_______ _ _ - - - ~ _ _ - - - - - - .-

                      $.5       llPCS Failure No. 5 Surpress:on Pool Suetion Valve F015 Falls to Goen                                               ,

j The llPCS pump suppression pool suction valve is a normally closed, rnotor operated valve which is in series with the 11PCS pump suppression p>ol suction check valve F016 in the alternate suction line to the llPCS pump. The liPCS pump is normally aligned to the CST via the CST suction l valve F001. Upon receipt of a low CST level signal or a high suppression pml level signal, the CST suction valve F001 automatically closes and the suppression pool suction valve F015 automatically opens. The irnportance of this IIPCS failure mode has been diminished by the current emergeng procedure guidelines which emphasize the continued use of outside injection sources. This requires operator action to bypass the 11pCS suppression pml switchover logic to prevent the opening of the suppression pool suction valve Fols This is especially true for the decay heat removal (non ATWS) F sequences where it is likely that the CST makeup can be maintained. There has been only one LIIR noted in the industry survey for 11PCS which involved the failure of the suppression pool suction valve F015 to open. 5.6 IIPCS Failure No. 6 Manual Iniection Valve is Plucced! Closed . The injection Stop Valve (F038 in BWR/5 or F036 in UWR/6) is a locked open manual valve in the llPCS injection line located within the drywell. If the valve were to become plugged or somehow closed mistakenly,11PCS injection could not take place. There were no L11*ls or NPRDS events noted for this failure category in the llPCS industry survey. Since it is located inside the drywell, the valve is not readily accessible to verify its bcked open position. Position indicator lights are provided on the main c<mtrol room reactor core cooling panel [ p601. 5-12

                                    --....,.,,._-,.I..-.-_-.---                             .    ..-      . ..- - .           _ . - - - _ . . - .

5.7 11PCS Failure No 7 - Minimem Flow Vahe F012 Fail- to Open The minirnum flow bypass line is provided for pump protection. The bypass valve F012 automatically opens on a low now signal when the pump discharge pressure is greater than a set level which indicates that thc pump is running. When the bypass is open, flow is directed to the suppression pool. The valve automatically closes on a high now signal. With regard to system operation and testing in the minimum flow mode, the licensee response to IE IMletin 88-04[16] should be reviewed to determine if the design of the minimum now bypass line is adequate. There were four LERs related to failure of the : ;iimum now v,lve F012 to open.Two of thern, were associated with the Dow switches. IE Information Notice 86-47 and IE thlletir: 86-02 [15l describe erratic performance problems in Series 102 and 103 differential presare . witches manufactured by SOR. Incorporated, formerly known as the Static *O' Ring Pressure Switch Company. There were 15 events noted on the NPRDS related to failure of the minimum now valve to open. Three involved the motor operator; these included a burned out motor, improperly crimped lugs at the valve motor operator, and improperly adjusted limit switches which caused the valve shear drive key to break. Two were caused by air in the sensing lines to the 11PCS high Dow trip instrument. The remainder were due to excessive instrument setpoint drift on llPCS pump discharge pressure instrumentation (4) and llPCS pump How instrumentation (6) which control the minimum now valve. _ 5.8 llPCS Failure No. 8 - CST Suction Check Valve FTHQ Fails to Open There have been no LERs and no NPRDS events found in the llPCS operating experience survey reporting a failure of the CST suction check valve F002 to open. S-13 l

_ _ _ _ _ _ _ . _ _ _ . _ _ ~ _ - _ _ V l 5.9 UPCS Fallure No 9 - CST Suction Manual Valve is Flucced' Closed . There have been no incidents noted in the llPCS operating experience sutvey in which the . [ normally locked open CST suction manual valve was plugged or improperly closed. 3.10 llPCS Failure No.10 CST Suetion Valve F001 Falls Clmed The CST suction valve F001 is a normally open, motor operated valve in the llPCS pump suction line from the CST, downstream from the RCIC tap off. The primary source of water for the llPCS system is usually the CST (LaSalle is an exception); upon system initiation iU01 opens automatically provided the llPCS pump suction from suppression pool valve F015 is not fully open.  ; F001 will close when Fol$ is fully open, and is prevented from opening when Fol5 is fully open. There were no LERs in the llPCS survey dealing with the CST suction valve F001 failing closed. There was one related NPRDS incident l18) in which the valve would not open with the handswitch, but could be manually. 5.11 llPCS Failure No.11 Pomo Discharce Check Valves (Fn24 or F005) Fail to Orien There are two check valves in the llPCS injection line between the llPCS pump and the reactor vessel. The first, pump discharge check valve F024, is a conventional check valve and is located just downstream of the llPCS pump. The other is the air operated testable check valve F005 hicated within the d,well in the injection line downstream from the llPCS injection valve F004 and upstream from the normally hxked open liPCI injection manual stop valve. In the event of a rupture just outside the drywell, the air testable check valve F005 would act to limit primary coolant loss. To enable veriiication of valve operability, a pneumatic actuator is i i attached to the valve dise pivot arm. The air actuator cannot prohibit disc opening. By energizing a _ 5-14

f l l solenoid controlled air supply to the actuator, the valve can be forced to lift off its seat. A light on the liPCS portion of the control roorn panel P601 verifies the operability of the stem. ) There were no incidents of failure of either check valve F005 or F024 to open reported on  ! LERs. There was one incident reported to the NPRDS in which the testable check valve F005 was binding during surveillance testing Less than one quarter of the valve's full stroke could be attained. i 5.12 IIPCS Failure No.12 Surpression Pool Suetion Chc.ejtValve F016 Falls to Onen , There were no reported failures of the suppression pool suction check valve F016 to open found f in either the LER data t-ase or NPRDS. 5.13 IIPCS Failure No 13 - False Imw Suction Pressure Trin llPCS pump suction pressure is monitored in the main control room by a highAow pressure alarm and with a local pressure indicator. There are no automatic actions associated with these instruments. Note that the llPCS design minimites inadvertent or false low suction pressure trips as compared to liPCI in which the tutbine/purnp trip is automatic, liPCS design inserts an additional step;i.e., operator incorrectly evaluates the alarm and trips liPCS pump. There were no LERs in this category. 5.14 flPCS Failure No 14 System Actuation Locie Fails , Startup and operation of the liPCS system is automatically initiated upon detection of either low low reactor vessel water level in the reactor vessel or high drywell pressure. The 11PCS system can -

also be manually initiated by arming and then depressing the manual initiation switch in the control room.

I 5-1b i

 - . . , . .     -,-.__-..-_..,_._.m.--._.___,_                                                                        .m.,....-,-.,,.----._-.._,,,_..~
           . - ~ . - . - . . -                             - . - - - - . . - _ . . - - - . - - - -                                                                        -

i I i There were two LERs which described failurer of flPCS low level initiation instrumentation. . and ten incidents related to failure of actuation logic instrumentation and controls identified in the NPRDS [18). Five of the failures described in NPRDS were associated with the level transmitter which

initiates llPCI upon low reactor water level. These occurred between Februaiy 1988 and March 1990 1

and involved Rosemount transmitters. There is a generic concern associated with Rosemount l transmitter due to sensing element oil lealrages that affect instrurnent accursey. In February 1989, Rosemount issued a Part 21 notification amccrning the loss of oil problem in sensing cells of some of their Model No.1153 and 1154 transmitters. 5,15 HPCS Failure No.15 - Suction Strainer Fails to Pass Flow A suction strainer is located within the suppression pool at the source of the llPCS suppression pool suction line.The strainer is located off the bottom of the suppression pool to reduce the possibility of clogging by sediments which inevitably collect on the suppression pool bottom. The strainer mesh P I is designed to prevent passage of particles which are large enough to cause clogging of the HPCS sparger nozzles, and the strainer area is oversized so that even with 50% of its surface blocked, the minimum required net positive suction head will be provided to the HPCS pump. There were no reports of failure of the HPCS suction strainer to pass flow in either the LER a search or the NPRDS data base. 5.16 Comna.ison of Oneratine Fxnerience to PR A Bued Rankinc As discussed earlier in this section, the total industry operating experience accumulated thus far is still somewhat limited because the plants have not been operating that long. Nevertheless, the PRA-based prioritization of IIPCS failures correlates well with the cetual industry failure experience. Two 5-16 l

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Salient exceptions within the LER data base are llPCS Failure No. 7 minimum flow valve F012 fails to open, and llPCS Failure No.14 system actuation logic fails. These were designated as " medium

  • and
  • low" risk importance respectively in the PRA based ranking developed in Section 4. The NPRDS I

data corroborated the larger nurnber of problems asmeisted with IIPCS Failures Nos 7 and 14 in the l LER data. As more llPCS operating experience is gained, the operating experience review will be updated to reflect the long term tiends. I t l. 5 17 4

6. OTIIER SYSTEM CONSIDERATIONS in addition to the failures and problems specine to or originating within the llPCS system as discussed in Scction 5, there are other factors which can inhibit the successful functioning of IIPCS.

Amoni; these are human errors of an operational nature affecting 11PCS, failures in systems which support the operation of IIPCS, interactions with systems that can impact upon llPCS function, and generle valve failures. These topics are discussed in the following sections. 6.1 Iluman Error The potential for human error pervades every aspect of nuclear power plant operation. Any task _ _ that requires human intervention such as maintenance, calibration, surveillance and, of course, opera-tion has the potential for human error, in Probabilistic Risk Assessments, operator errors are included both in fault trees (system failure diagrams) and in the event trees to account for human interactions. As such, human error events are usually actions that can fait e complete system or components required to preserve system function. Typical PRA-based itPCS human errors are:

1. Failure to manually statt the high p essure injection system if automatic actuation falls.
2. Failure of the operator to transfer pump suction from the CST to the suppression pool after a pump trip on low suction pessure <!ue to CST unavailability.
3. Failure to provide makeup to the CST during an ATWS event.
4. Failure to transfer pump suction from the suppression pool to the CST during an event with a high suppression pool temperature. There are two eases when this must be performed, one during

. an ATWS event'and one during a non ATWS event with the failure of suppression pool cooling. 6-1 mai g ukui i. .

  • i
                                                                                                                                                                                      )
5. Operator recovery from initial failure of IIPCS, ,
6. Miscalibration of IIPCS sensor (s) disables system actuation, high RPV level isolation or results in false isolation signals.
7. Failure to reset the llPCS system for opetation after testing or maintenance.

With the exception of the last two entries, these human errors are either: (a) conditional, that -1 is, tney must be considered within the context of an llPCS failure or isolation (errors 1 and 2), or (b) event specific (items 3,4 or 5). These requiremer ts make dhcet observation unlikely.The potential for these human errors car be evaluated indire~ly by a review of the licensee procedures, operator training program, and observation of operator performance at a simulator P The last two human errors can occur during normal operation and are therefore more inspectable, Resident inspectors routinely examine surveillance, calibration and maintenance practices and procc-dures, and perform ECCS contre! room and plant lineup verifications, itPCS operability is confirmed by checking the pump suction and discharge lineups, and the control function settings (hand / auto sta-tion in automatic). There is a second source of human error that is not readily discernible in most risk assessments because it is not considered as a separate failure. It is the human contribution to component unavail-ability. The comp (ment failure estimates are developed from plant specific experience,if enough data exists, or from other, more generic, data sources, in either case, the unavailability estimate of a standby _ component is based on the number of fallares ner tos -emands. This estimate inherently includes all failures caused by human error,13ased on the operating experience review, it is estimated that nearly-62 . t

                                                                                                                                                                                     ?

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l l i half of the llPCS LER failures have a human error contribution.The more unusual human errors have been included with the illustrative examples of Section 5. As previously indicated, the examination of licensee practices and procedures, as well as the application of industry experience, can help reduce that portion of the IIPCS unavailability that is due to human enor. In the reactive mode, a thorough root cause analysis and suitable corrective measures can prevent similar occurrences in the future. 6.2 IIPCS Support Sutems r 4 The high pressure core spray system is dependent on other systems (support systems) for success-ful operation. These systems include. . AC Power: For llPCS injection pump, keep fill pump, and valve movement. Loss of off site power requires HPCS (Division Ill) Diesel Generator and its support systems: fuel oil, Jacket cooling water, tube oil, starting air, DC control power, room cooling DC Power: For system control (125V DC)  ; Condensate Storage (CST) and Transfer System: Primary source for injection and system Dushing. Suppression Pool: Alternate source for injection and discharge reservoir for-minimum Dow. and test pathways. Room Cooling: For HPCS pump room cooling to support long-term operations. This function-t requires service water (for cooling) and AC power for the fan motor.

                     - 12 JCS Actuattun:                   RPV level and primary containment pressure instrumentation for system initia-tion and shutdown.

6-3

The Condensate Storage Tank (CST) and Suppression Pool, as the primary and alternate sources , for liPCS injection, are an integral part of system functional success. A sufficient volume of water is maintal- d in them as required by Technical Specifications to assure an adequate supply for emergency core cooling systems such as llPCS and RCIC. In accident sequences requiring extended use of IIPCS, the condensate transfer system or other means of replenishing the CST water volume such as fire water pumps can be more important. The compressed air or nitrcgen system is often considered an 11PCS support system since it provides compressed air to the pneumatic actuator of the testabic check valve F005. It is not required for successful operatic . of the llPCS system, however, since it supports a test function to verify valve operability. 6.2.1 IIPCS Diesel Generator (DG) und Division til AC Power AC power is critical to the success of IIPCS operation which utilizes a single 4160V AC motor-driven pump and 480V AC motor operated valves. HPCS is a Division III electrical system which is usually supplied via normal off site AC power. This source is backed up by the Division til HPCS  : diesel generator, , inspection of the off site power system and the Di ision III (HPCS) diesel generator lie outside r the scope of this inspection guide. Many detailed studies have been performed specifically for these -

                        - components, particularly in the area of reliabdity/ availability of en.ergency dicsci generators at nuclear power stations, and include guidelines for on site inspections [22). Inspectors should consult these documents for inspections concentrating on off sitt. power systems and emergency diesel generators.

A proposed inspection plan for diesel l'nerators at nuclear power plants is provided in Appendix B (22) as a guideline for those inspectors who wish to investigate aspects of the ~ 'CS diesel generator in more '

                          - detail.                                                                                                         .

6-4 l. l l

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6.2.2 Disislon til DC Power The Division ill llPCS DC Distribution System is an independent 125 volt DC system consisting of a battery, one or more static battery chargers (one preferred and one back up), and a DC panel-board. The llPCS DC Distribution system supplies llPCS Equipment and is physically separated from Division I and 11 to provide the utmost reliability. Figure 6-1 illustrates this configuration. During normal operation, all load current required by the system is supp!!cd by the charger with the battery fully charged and on 00at. The battery will only supply current to the llPCS DC Bus when larger loads are started. On loss of all AC power to the chargers, the battery is sized to provide power to all 11PCS DC loads for at least 2 hours, AC power outages should only be of short duration as the chargers are fed frorn the Division til AC Bus and will be re-energized when the llPCS diesel starts and re energizes the Division til AcBus. (llfall 44ttllt tg I d) ' iO> (t97 3) lli V. SC tut til

                                                                                                                        ")     ... n)                       9....

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                                                                                                                                 .::         V
                                                                                                                                                                  -!'i',ii TTfICat t88948 Figure 61. laSalle County Station. Unit 1,125V DC ESF Disision Ill.

65

6.23 Rour, Cooling and Ventilation - The cliect that the loss of room moling would have upon continued operation of the llPCS system is not as straightforward as the previously discussed support systems. Since llPCS and all its auxiliarica and support systems are powered from the same bus, guwer to the llPCS buses would imply power to the cooling units and fans. loss of an indiddual cooler or fan while llPCS injection contin-ued would be confined to individual equipment failures or feeder faults. The licensee should have evaluated the effects of temperature on long-term operation of IIPCS upon loss of room cooling and have proecdural provisions instituted as required to assure long-term IIPCS operability, 6.2A Splein Actuation lustrumentation The reactor pressure vessel level and high drywell pressure instrumentation required to actuate multiple ECCS functions via the Rosemount 510 dun 10DU Trip / Calibration System is shared by llPCS. Consequently, the operating experience review speci0cally for llPCS revealed only a few inci-dents invohing actuation instrumentation, and these have been discussed previously in Section 5.14. In summary, support system problems can sometimes impact ilPCI operation in a less than straightforward manner. In the context of specine accident sequences these support systems may be more prone to failure. The insptetor should verify licensee awareness of these interaction relations and con 0rm that ecmpensating measures are adequate. 63 LIPCS Systems Interactions Systems interactions refer to failures or problems arising in other plant systems unrelated to llPCS that can result in the disabling of liPCS. There were no specific examples of this in the 1-IPCS operating experience review. liowever, an operating experience review covering 1980 to mid 1989 for ilPCI, a system which has accumulated a much more extensive amount of operating time than llPCS _ did have some system interaction incidents that are applicable to llPCS. All of these were 6te protec-tion system malfunctions that disabled liPCI. 6-6 l l y g-wrw Type f - g 77-v~ Wy3-N-- wiygg y-w -g s'-p myyy-i,-w--s my g ,qr i tr'T'P*-***"P'*M-fvt2-'t---* 4'~N'+^*M n'* --

                                                                                                                                 'W**. ' r---' ' * = ' " *       *
                                                                                                       ~ . _ . _ _ _ _                  - _ _ _

0 9 1 There was one event at River llend (LER 458 86 054/DCS 8703120095) invoking IIPCS and the . 1 analog trip system which highlights an area with the p*ntial for adverse system interactions, in that case, a level 2 IIPCS initiation instrument was declared inoperable, and its master trip unit was placed into the trlpped condition per Technical Specifications. It was not until later on that the operators ! realized that the same rnaster trip unit controlled an llPCS level 8 slave trip unit, which therefore made the level 8 slave unit inoperable as well.11PCS should then have been declared inoperable as well.  ; t Licensee procedures and training should demonstrate an awarenos that the analog trip system master l trip units can affect multiple systems and functicns via slave trip units, and that incidents involving loss of power, restoration of power, and system power fuse replacement for the analog trip system cabinet

potentially can irnpact several unrel
o ed systems at the same time.

6A Mmultaneous tina $nllability_ of Multinle Systems Multiple system unavailability is of concern because of the increased risk associated with contin. ued operation and the increased common cause potential it may imply [65). Although Technical Speci. fication 3.0.3 tends to limit the tisk exposure somewhat, the !!censee should avoid planned multiple . system outages,if possible. The operating experience review of IIPCS LERs identified seven incidents , involving simultaneous unavailability of multiple ECCS systems. Within the context of the accident sequences discussed previously (Section 3), certain combina. tions of system unavailability result in a much greater risk of core damage. Among the seven LERs , mentioned above, c'.e of the incidents at Clinton (LER 46189_041/DCS 9001250005) reported the simultaneous unavailability of IIPCS and RCIC, During this period, the probability of core damage is greatly increased for accident sequences that require HPCS and RCIC for mitigation. This would _ include all the sequences described in the_ Accident Sequence Descriptions _ except 'Unisolated LOCA Outside Containment? l Most of these LER examples of multiple system unavailability were initiated by one or more l 1

random failures, and often follov ing licensee decisions to remove a system from service for mainte.

L l 6-7 1

u. . _ _ _ _ _ .._ _ . _ _ ._____ _ . . _ _ _ _ _ _

nance or surveillance when another critical system is not operable. Five of the seven LERs also in. volved some type of human error, attributable to inexperience, unfamiliarity with Technical Specifica-tions, and proceduralinadequacies Nevertheless, plant configurations involving simultaneous multipic system inoperability, particularly the conhgurations mentioned, should be avoided unless absolutely necessary since habitual entry into Technical Specification 3.0.3 greatly lucreases the risk of core dam-age. 6.5 Vulte INilures in illah Prepure_ inleethin Sntems Many of the valve failures encountered in the llPCS or ilPCI Systems whether they involve the pump discharge valve, the minimum flow valve, the turhine steam admission valve, etc. can he consid. ered generic, i.e., not related to a specific application. Tabic 61 presents a summary of generic valve failures gleaned from the operating experience reviews for liPCS and llPCI, A failure description, root cause end corrective action is provided for information and lotential inspection applications, it is worth noting that a recent study entitled

  • Aging Study of Boiling Water Reactor High Pres-sure injection Systems" [ Reference 11] reported that valves and valve operators made up more than 25% of allllPCI failures and nearly 29% ofIIPCS failures in the LER database, in the NPRDS data-i base, the study found that valve and valve operators made up over 44% of IIPCI failurcs and more than 27% of IIPCS failures (see Figure 6-2).

References [24 30] provide additional information on generic valve failures. , 6.6 LOCA Outshte Contninment Unlike the HPCS failures of Section 5 which describe the unavailability of the system for core-I damage mitigation, events have occurred where the high pressure injection system is a potentialinitiator of a LOCA outside containment. There were two LER incidents involvinF 'he llPCI system, in which inadvertent pressurizations of the system low pressure piping occurred.' Due to the similarities of the discharge portions of the llPCS system and the llPCI system, the potential of this type d aishap also exists for ilPCS. 6-8 i

l Percent of failures 50 O VALyt oPER ATORS HPCI O VAlvts HPCS 3o .

                                                                                                                                                                  .HPCI~~~'~ - ---                                         - - HPCS         _
                                                                                                                                                                                          '                    \

2 0 '- - - ' - - - - -

                                                                                                                                                                                                                    \-

10 --

                                                                                                                                                                                 ,                                              1 o

LERs NPRDS ligure 6-2. Percent of valse und satse operator failures reported to llit und NPilDS databases [11] One event was attributed to the slow closure of purnp discharge lift check valve (RX)$), while the second 11PCI overpressurization event was due to the two pump discharge MOVs (RX)6 and F007) being left partially open. In addition, the upstrearn check velve (RX)S) was not properly seated. A similar RCIC overpressurization occurred on April 12,1989 and is the subject of Inspection Report 50-293 89-80 131]. 6-9

  )~                                                                                                                                                                                         !

h - Tabic 6-1. Summary of Generic Valve Failures -IIPC111PC5 i t i" f i I Category radore Descriptim Ront Cause Corr. Meas. (linrnents f " l

Valve O;erators - 1IPO. MR 324 87 0011)CS 8705130141. Failure attributed to pwsiNe Va!ve motor replaced. Note 1 i
   ,            Motor & Starvr          -lirunswick 2 -INmp discharge vahe           heat related tveskdown d valve

) Failed To Open after scram motor internals. 7 4 i IIPCI. MR 298 55 011lDCS 8511260500 Switch damage resulted from Smitch replaced, similar smitches [ j; Cwp r - Min flow vahe ermiperaNe due owr-travel of the operating inspected. Possible redesign er [ to damaged motor starter drsommert handie due to pur suisch desigrt replacement comidered. i i switch. i Valve Operators - IIPO. ER 293 89 0131)CS 890$eme. Taque switch ad ustment l screms Vahe reparred. screws cor*ectly. l Pttg im -Turbine steam admission vahe k=me which affected the torpe torqued. Other safety related  ! Torque Smitch 4 Failed To Ogn. Valve operator matar setting and damaged the vahe. valves impeced. torque sattch ( j mindmgs failed Tamiter plates imtaPeed and [ procedures revised E IIPCS. NPRDS [18). 8/3/89 - Test valve Dirty contaas on the tmque Contacts were cleaned amt vahe [ n to suppressson p=d stopped at 25% cpen switch. tested. } during survedlance test with plant at l 4

          ?

10tM pmcr. 1 I o i Valve Oprators - IIPCS, NPRDS [18). 3C&M -Infecticri Limit switches were out d Adjusted rotor such that hmit  : Ijmit Switch valve motor operator tripped on thermsl adjustment. switch opens to stop vahe i j everload when going to opert between 60 - 1074 d travel in [ open position to grevent i overload in backseat pwitson. . IIPCS, N"RDS [18) 7/17!86 -Test return Ogn knut switch was trig 5=ng 12mit switches were adjusted to valve o the CST would not return to full prior to the torque switches aIkw the inque switches to fu!!y open psition as indecated on txth local during openmg cf valve. Position egen the valve. Valve puition , j an j remote indicators with plant at 80'"c indicators were giving false indicators were a4usted. [ l pwns reading when valve mas closed. } l Valve Operators - IIPCS. NPRDS l18). 2/1aN - Worn shaft clutch gear assembly Replaced entire clutch assemNy Ucensee notified the i . Internals Suppression pud suction valve wondo not fell apart due to unissing spist amt avrn shaft. Revised manufacturer. l ! operate via the motor. Motcr ran but spacer utnch acts as a seat in procedures to irppect assemNy Umitnque, of pws Ne ,

valve wouldn't noe. shaft set serems_ Spacer was ntA and re4take set screws if OC deficiences (10 GR  ;

[ imtailed during manufacturing. requrred Verified operability d Part 21 Repst) J aTI Unntorque motn cyrators { i of similar design. [ I Note 1: It is imclear if the awrective action addressed the mot crese.  ! i b .

_ __ .__ -_ _ _ _ m . - j :: . q- - L Table 6-1. Summary d Generic VaIve Failures -IIPClifPCS (Gmt'd) l i Category . Failure Description Rcvt Cause Carr Meas. Gynments i ! i IIPCS, NPRDS (18). 4/22/87 - Test t$pass Vahe operativ gearing and Re;4 aced valve operator housing, to surpression pool valve would not stay bearings mere wxn out. Iceer bearing usp. mm gear.

torqued into its seat when clewed Attributed to ncemal weer_ worm bearmg. and housmg cap

] clectricaDy during normal operation. screw. f Valve Operator to IIPCI. IlIR 259 87 027/DCS 880RIN1227 Valve disk and shaft amnected Vahe repaired, similar vshes Associs ed Control i Valve Connectimt lirowns Ferry 1 - Suppression pool by a key which is held in g4 ace by m3ected. Valws were Equi; mar Inc. hand {' suction line manual valve stem separated a caer secured by four bolts. scheduled for irspection and contrc4 vahr. from the disk. 1hree bolts fai;cd due to tensile testing during the next refueling l overload. outage. j IllQ NPRDS (18). 7/1/89 - Te=t return Gasket tetween motor and Reg 4 aced gaskets between motw l 3 vahe to the CST had an oilleak at the motor operator had deteriaated ami dutch housing.  ! f mating surlace bermeen the motor and due to '-+noan causes_ l i motor operator. r t ,- IfPC1.111R 254 87 DJs DCS 8703313497 Slight disc.h=Jy misalignment 0;en limit smitdt adjusted to j- 7 j

           -                      Quad Oties 1 - CST suction valve failed    occurred mer time due to               valve did not open quite as far.                                          ;

i to fully dose during testing horizontal vahe mounting.  ;

Increased resistance caused [

i torque switch te stop dosure. [ L Seat / Disk teaks 11PCS, NPRDS {l8]. 4WB8 -licensee Valve seat was dirty and the Vaht internais were deaned, unable to establish p essure boundary flexitalic gasket was spbt as a gasket was replawd and vahe j during leak rate testing of dsscharge to result of agsg and cydic fatigue. retested. , surgvession pool valve. i {~ IIPCS, NPRDS [18).1G21187 - During Valve seat and disc had stress Valve seat and dise deaned; drse j IIPCS system venting and fi!Iing while cracks as a remit of incorrect remsta!!ed 150" rotated fran i

. unit was in a refuesing outage. the torque su,tch settmgs which original pmition and "Nue  !

j minimum ikw vahe was leaking by when inacased stem thrust higher than checked". LLRT sucressfuny the vahe was ckwed necessary to seat the vahe.. perfwmed. Torque suitches i adjusted to lower equivalent stem , thrust to IO W to 13N0 toends.

Packing I eaks IIPCS. NPRDS {l8). 6/5117 - licensee Worn packmg. Vahe was replaced during next j health physicist founJ packing icak from refueling outage, leak tested. and
j. suction valve from CST was causing a stroke time tested. .

spread of contaminatiort 6 P 4 1 ,

l J 5 T.ble 6-1. Summary of Generic Vahe Failures -IIIC,1IPCS (Oct*d) i Category Failure Descriptim RM Cause Qvr. Meas. Camrnems IIIQ NPRDS [18) 5/11t9 -Injectim Old and own packing Packing in the salve uns l stop valve was fimnd Icaning during replaced.

namal operator rounds.

d l Mechanical IIPCS NPRDS [18).108M -Injectron Valve stem needed lutvicatiert Idricated valve stem enh Damagellinding valve was bmding in the intermediate Molykove GN. , -! pmition causing drcuit tweaker to trip _ Vahr stem anti-rotatim damp Valve IT221V11 was repaired. Prim to thrs inodent, the ] IIIQ NPRDS l18).1/28N - Broken

                        *ahr yoke was discovered m the seemd       set screw had ionsen ami alkmed   Design change iratiated to instaII bcensee staked the act test return valve to the C5T (IE22F011). the damp to slide below the key.  "I *-shaied stem kep m toh         screws foe the Th s allowed stem to rMate end    test return to cit valves          antirotation damp on drive stem keys into the damp     1E221VII and IE221V10.             these va3ves because of resultmg in an unequal load                                          similar prt+1 ems at othee being appired to the valve ytte.                                     nudcar rimer plants
     &                                                             After an unkmms number cf N                                                             cydes in this cmfiguration. the                                                                          ,

vane yoke cracked at its base. [ i' IIPCS. NPRDS [IS). I'2n8 - As found Stem to disc coupling not threads Rerdaced coupling urut, valve Note I thrmt load m the vaht stem cf the were stretched due to stem.and valve drse. Retested discharge isolation valve exceeded rated undetermined root cause. Dese valve satisfactorily. - levels- scored due to normal wear, j Differential IIIU.LER 265 85 02TDCS 8601300019 Large differential pressure Open MCCcmtsctw dcmed. Nde1 l

- Pressure Acnss Quad Oties 2 - Gintainment Isolation created a high torque combtioet vahe stroked to verrfy i j De Valve Vahe failed to open (110) after packing Bypass limit suitch was set under gerabilty.

replaced zero dcIts P and cyened prematurely. IIPCS.1.ER 373 83 0%DCS 8307220193 Insufficxnt sprina, tensim of the Spring tension to the rack and Vahes made by W-K-M arwl IER 373 83 067DCS 8307220201 actuator assemN f orf the t 3 pass gear assembly was increased. Dission of AG Imi 1,asafic 1 -Testable check valve 1E22- Sabe IE22-I'334 to dose the Inservice test (IST) procedure IV05 failed to dose due to failure of its valve. was revned. bypass vaht IE22-F354 to close , i preventing a differential dming force to seat the valve. l. 4 1  !

!~

{ I . 4 i u . _

j- More recently, (October 31,1989) Dresden 2 declared ilPCI inoperable due to elevated piping temperatures in the pump discharEe line. The 260T temperature was caused by feedwater back leakage through the closed injection valves. Discharge piping supports were damaged, attributable to water. hammer caused by steam void collapse upon system initiation. In addition to the potential for piping damage, steam binding of the pumps is also a consideration. Information Notice 89 36 [32) provides 4

                          - additional information on elevated ECCS piping temperature, in general, the high pressure injection systems LOCA outside containment initiator is a very small contributor to total care damage. The examples presented above are potential areas of inspection to l

assure that plant design or operation does not increase the potential for this initiator. 4 4 Y t b 6 13 P 4

                                '                                                       *   * "                       T' ' * ' * -' +7"T'r'        '*"Y'+ ' "

( *'-*m q

i

7. SU$th1ARY This System Risk Based Inspection Guide was developed as an aid to BWR !!PCS system inspec-tions. The document presents a risk-based discussion of the !!PCS role in accident mitigation t.r.d provides PRA based llPCS failure modes (Sections 3 and 4), hiost PRA oriented inspection plans end here and require the inspector to rely on his experience and knowledge of plant speci6c and OWR operating history.

The system RlG uses industry operating experience, including illustrative examples, to augment the basic i'RA failure modes. The risk-based input and the operating experience have been combined in _ Table 71 to develop a composite BWR llPCS failure ranking. This information can be used to opti-mire NRC resources by allocating and focusing proactive inspection effort based on risk and industry experience. In conjunction, the more important or unusual component faults are summarized in Section 5 and provide potential areas of NRC oversight both for routine inspections and the " post mortems" b conducted after significant failures. A summary of the risk-important operating expetience failure events is presented in Table 7-1 as a composite of the applicable LERs and NPRDS failures at all of the BWR/5 and BWRh plants (Clinton, Grand Gulf, LaSalle Unit 1, LaSalle Unit 2, Nine hiile Point 2, Perry, River Bend, and WNP 2). Although individual plant speci6e experience is limited, the composite identification of IIPCS failures in Table 7-1 serves as a guide to inspectors and a standard against which the performance of individual plants may be compared. Those components at a specific plant exhibiting a higher propor-tien of failures than the industry experience are candidates for more extensive inspection activity. As the HPCS plants mature, operational experience is accumulated by the plant statfs and plant procedures are refined, thereby decreasing the number of inadvertent HPCS incidents attributable to surveillance and calibration activities. Conversely, as the equipment gets older, aging related faults are expectcd to become a contributor to a plant's llPCS failure statistics {l1]. Inspectors should thus be 71

                                                 -.--         . - -                  . _ . _ . _ _                                      _ - - ~ .                      . - . - . ..-                          - . - -

Table 71. IIPCS Systern RIG Summary , l All 11WW5 & llWR5 Plants Plant Specific Summar)* l'ailure Dewription  % of Total  % of Total l'ailur e fiumber  % of trit Comments l ofi1:R IJ.R l'ailure NPRDS llanking' l'ailure Contribution' I ailure l'ailures Conteibuteon l Contribution' ( 5 [ Purnp l' ails to Slart or Itun 13 0 11.3 1 Systern Unnvailable Due to 47 8 11 3 2 lest & Maintenance injection Valve IVM l' ails 0 94 3 to Ogen i Gl/Suppreninn hxil 4.3 15.1 4 Switchover logic l'aits 43 0 $ 8 Suppression Pool Suction Vahe 0015 Iails to Orien f 28 3 6 6 Minimum flow Vahe l'012 17.5 I ails in Open Manual Injection IJne Stop 0 0 7 Vah-c is PluggedCowd  : t s GT Suction dieck Vahe 0 0 8  ; 1102 I ails to Ogen CST Suction Manual Vahe 4.3 0 9 in PluggedCowd GT Suction Yahe f(on 0 1.9 to I ails Gosed Pump Digharge Vahes 0 IN 11 1U24 or 10.)$ I mil to Open M8 18.9 12 6, 7 System Actuation 1cgic l' ails . 0 0 13 8 , Suppreuion Pool Suction > Oseck Yahe I016 l' ads to Open l'alw inw Suction Prenure 0 1.9 14 Manual 'T rip 1 Suction Sirainer l' ads to 0 0 15 Paul h - TOTAL 1(o% 100 % - l l 72 f b "e'ewa'- ,ew=rTu-_-ep-----hmiw'-w '-M*" -M'gs ' r -NWM9tvuWM T rr' tty 1 M T p'TFt t'NTV

  • W ymeWW t s$T
  • T"t'YT*"TWWW $ y- r-$N'g-M'+'tv$ ('"
s. Table 71 Notes 4
1. Failure contribution is expressed as a percentage of all significant 11PCS LER failures as developed by the Operating Experience Revicw. Current up through inid.1990.
2. Failure contribution is expressed as a percentage of all significant related or precursor HPCS failures on NPRDS. Current up through 10!!/90.
3. Failure ranking is a subjective prioritization based on PRA and operationalinput, s.covery poten- ,

tial, current accident management philosophy and conditional failures, as applicable. t

4. The plant specific summary is to be completed by the inspector initially usirig the LERs provided in Table 5-1, and updated periodically with any additional LER events that may have occurred in the various categories. Categories with significantly higher failure contribution percentages than the industry experience are candidates for enhanced insrection attention.
5. Loss of offsite power and failure of the Division 111 (HPCS) diesel generator to start and run or loss of the Division til bus would also result in failure of the HPCS pump to start and run. How-ever, these events are beyond the scope of the operating experience review for this inspection guide.
6. Failure importance was upgraded from the PRA based ranking of Table 4-1. ,
7. The actuation logic arrangement (one out-of two twice) diminishes the importance of a single instrument to reliable system operation. At least two low RPV level or two high drywell pressure sensors must fail. As shown in Section 5. unavailability is more dependent on control power.
8. The latest BWROG Emergency Procedure Guidelines deemphasize the suppression pool as an injectbn source.

1 l l 7-3 1 l

                                                                                                                                    ,,.,,_.m.......               ....~._.mm_....    . , . _ . . . . , , _ _ . ._ . .... . m . .-v

sware of the effects of time-dependent factor. such as these upon the HPCS failure categories at each , individual plant. The HPCS fciture statistics for each plant may be tabulated in the appropriate columns of Table 71 and can be periodically updated as the plant matures. This report includes all HPCS LERS and NPRDS failures through mid-19. Wquent LERs can be correlated with the PRA failure cate-j gories, used to update the plant specific HPCS failure contribution, and compared with the more static i . industry HPCS failure totals. The industry operating experience statistics are developed from the eight BWR/3 and BWR/6 listed pieviously and are therefore expected to exhibit less fluctuation with time than a single plant. This information can be trended to provide guidance as to where additionalinspec-tion focus is warranted as the plant matures. p 4 7-t

8. REFERENCES
1. Brookhaven National Laboratory (BNL) Technical Letter Report, TLR-A-3874-T6a," Identification of Risk Important Systems Components and 11uman Actions for DWRs," August 1989.
2. NURi 3/CR 5692, " Generic Risk Insights for General Electric Boiling Water Reactors," R.Travis and J. Taylor, Brookhaven National Laboratory, hiay 1991.
3. Shoreham Nuclear Power Station Probabilistic Risk Assessment, Docket No. 50-322, Long Island Lighting Co., June 1983.
4. NRC Case Study Report, AEOD/C502, "Overpressurization of Emergency Core Cooling Systems 3 in Boiling Water Reactors," Peter Lam, September 1985.
5. NUREG/CR 5124, " Interfacing Systems LOCA: Boiling Water Reactors," T.L. Chu, et al., Febru- _

ary 1989.

6. Brookha.en National Laboratory (DNL) Technical Report A-3453-87-5 " Grand Gulf Nuclear Sta-tion Unit 1, PRA-Based System inspection Plans," J. Usher, et al., September 1987. ,
7. BNL Technical Report A-3453-87-2, " Limerick Generating Station, Unit 1, PRA-Based System inspection Plans," A. Fresco, et. al., hiay 1987.
8. BNL Technical Report A-3453-87-3,"Shoreham Nuclear Power Station, PRA-Based System inspec-tion Plans," A. Fresco et al., May 1987.
9. BNL Technical Report A-3864 2, " Peach Bottom Atomic Power Station, Unit 2, PRA-Bned Sys-tem inspection Plan," J. Usher, et al., April 1988.
10. BNL Technical Report A 3872-T4,"Brt.aswick Steam Electric Plant, Unit 2, Risk-Based Inspection Guide," A. Fresco, et al, November 1989.
11. NUREG/CR-5462 (Draft)," Aging Study of Boiling Watt - Reactor High Pressure Injection," D.A. _

Conley, Idaho National Engineering Laboratory (INEL), September 1990.

12. NRC AEOD Report E402," Water Hammer in BWR liigh Pressure Coolant Injection Systems,"

January 1984.

13. NRC Inspection Report 50-322/88-24, Limerick Generating Station, Unit 1,1988.
14. NUREG/CR-5051, " Detecting and Mitigating Battery Charger and inverter Aging," W.E. Gunther, et al., Broci iaven National Laboratory, August 1988.
15. NRC IE Bulletin 86-02, " Static "O" Ring Differential Pressure Switches," July 18,1986.
16. NRC IE Bulletin 88-04, " Potential Safety Related Pump Loss," May 5,1988.
17. NRC AEOD Engineering Evaluation Report E407, " Initiation and Indication Circuitry for liigh Pressure Coolant Injection (HPCI) Systems," March 26,1984.

8-1

18. Nuclear Plant Reliability Data System (NPRDS), Institute of Nuclear Power Operations (INPO), ,

llPCS GE and ilPCS Power-GE,10/5/90. .

19. NRC Inspection Report 50-416/85 28, Grand Gulf 1, August 1986, Section 5d.
20. Illinois Power Company Letter, F.A. Spangenberg to the NRC (W.R. Butler), Docket 50-461, "Clinton Power Station RCIC and ilPCS Low Temperature Automatic Transfer Feature," Septem-ber 3,1985.

2L NRC Inspcetion Report 50-354/86 30, llope Creek, July 1986, Section 3.3. L -

  • A4440, "A Review of Emergency Diesel Generator Performance at Nuclear Power Pla liggins and M. Subudhi, Brookhaven National Laboratory, November 1985,
                                   ~5119, " Aging Assessment of Instrument Air Systems at Nuclear Power Plants," M.                 -

al., Brookhaven National Laboratory, January 1990. LD Case Study Report C603,"A Review of Motor Operated Valve Performance," Earl

                   ..    ..an, December 1986,
25. NRC AEOD Engineering Evaluation Report E702,"MOV Failure Due to liydraulic Lockup from Excessive Grease in the Spring Pack," E.J. Brown, March 19,1987.
26. NRC IE Bulletin 85-03," Motor Operator Valve Common Mode Failures During Plant Transients Due to improper Switch Settings," November 15,1985.
27. NRC IE Information Notice 88M2,"NRC Generic Letter No. 8910," Safety.Related Motor Oper-ated Valve Testiag and Surveihance " Jur,e 28,1989.
28. NRC NUREG/CR-4234 Vol. 2," Aging and Service Wear of Electric MOVs Used in Engineered Safety Feature Systems of Nuclear Power Plants," II.D. Haynes, Oak Ridge National Laboratory.
30. NRC Memorandum, from Jack W. Roe to James E. Richardson, " Check Valve issues," July 25, 1989.
31. NRC l'.spection Report 50-293/89-80, Pilgrim 1, May 1989.
32. NRC Information Notice 89-36, " Excessive Temperatures in Emergency Core Cooling Systmn Piping Located Outside Containment," April 4,1989.
33. Gulf States Utilities Company letter, J.E. Booker to the NRC (Document Control Desk), Docket 50-458, special report concerning invalid failure of the Division 111 diesel generator, October 10, 1988.
34. Gulf States Utilities Company letter, J.E. E or to the NRC (Document Control Desk), Docket 50-458, special report concerning valid faih a the Division 111 diesel generator, May 26,1989.
35. Gulf States Utilities Company letter, J.E. Booker to the NRC (R.D. Martin), Final Report, "DR 175/llPCS Diesel Batteries," April 26,1985.

82 l

36. Niagara Mohawk Company letter, C.V. Mangan 'to the NRC (R.W. Starostecki), " Problem Con-cerning HPCS Battery Terminations," February 22,1985.
37. ' Gulf States Utilities Company letter, J.E. Booker to the NRC (R.D. Martin), " Final Report DR-222 Voltage Drop in 125 Vdc Cables," May 16,1985.
38. Techriical Report A 3875 T4-2, " Quad-Cities Station, Units I and 2, High Pressure Coolant injec.

tion System Risk-Dased Inspection Guide," July 1990,

39. Clinton Power Station, Illinois Power Company, Final Safety Analysis Report (NRC Docket No.

50-461), Sections 63 and 83, Amendment 32, December 1984.

40. LaSalle County Station Ur.its I and 2, Commonwealth Edison Company, Updated Final Safety Analysis Report (NRC Docket Nos. 50-373 and 50 374), Sections 63,73,8.1, and 83, Amendme"?

64, March 1984.

41. LaSalle County Station Units 1 and 2, Commonwealth Edison Company, "High Pressure Core Spray System Operability Test," Procedure LCS-HP MI, Revision 6, Novetuber 17,1986.
                - 42. Lr.Salle County Station Units I and 2 Commonwealth Edison Company, LaSalle System Discrip-non Chapter 35,"High Pressure Core Spray (HP) (E22)," Revision 2 June 1988.
43. LaSalle County Station Units 1 and 2, Commonwealth Edison Company," Filling and Venting the High Pressure Core Spray System," Procedure LOP-IIP-01, Revision 7 February 13,1986.
44. LaSalle County Station Units 1 and 2 Commonwealth Edison Company," Preparation for Standby Operation of High Pressure Core Spray System (IIPCS)," Frocedure LOP-HP.03, Revision 8, November 17,1986. ,

( 45. LaSalle County Station Unit 1. Commonwealtn Edison Company, Drawing No. M-95, "P&lD High , Pressure Core Spray (HPCS)," Re'ision AA, March 1,' 1988.

46. LaSalle County Station Unit 1, Commonwealth Edison Company, Facility Operating License Ap-pendix A, Technical Specifications, Sections 3/4.5.1,3/4.5.2, and 3/4.53, Amendment No. 59. .)

47, Perry Nuclear Power Plant, The Cleveland Electric illuminating Company, Final Safety ..nalysis Report (NRC Docket Nos. 50-440 and 50-441), Sections 3.11, 63, 73, and 83, Amendment 17,- March 1985. l 48. Nine Mile Point Nuclear Station Unit 2, Niagara Mohawk Power Corpo.ation, Operatmg Procc-dure No. N2-OP-33, "High Pressure Care Spray System," Revision 4, September 28,1989. l

49. . Nine Mile Point Nuclear Station Unit 2. Niagara Mohawk Power Corporation, Facility Operating -

License Appendix A Technical Specifications, Sections 3/4 5.1 and 3f4.5.2. l 50. Nine Mile Point Nuclear Station Unit 2,-Niagara Mohawk Power Corporation, Operations Tech-I nology Lesson Plan, "High Pressure Core Spray," Chapter 12-CSH, Revision 5, May 29,1990 I-i 1 83 i a ' .- , .-- -- .. - - . - - -- -~.:-- , . - . -

 - . . - .~ -, . - - . -.                                        .   - - - - . - _ _ . - - - - -                         --

y SL -Nine Mile Point Nuclear Station Unit 2, Niagara Mohawk Power Corporation, Operating Proce- . , dure No. N2-OP 7.4B, "HPCS 125V DC System," Revision 2, September 12,1988.

                - 52. Nine Mile Point Nuclear Station Unit 2, Niagara Mohawk Power Corporation, Unit 11 Operations -

Les.;on Plan No. N2-OLP-12, "High Pressure Core Spray," Revision 4, April 7,1988.

53. River Bend Station, Gulf States Utilities Company, Station Operating Procedure No. SOP 0030, "High Pressure Core Spray (Sys 203)," Revision 6, 8/16/89.
54. River Bend Station, Gulf States Utilities Company, Nuclear Training Department Licensed Opera-tor Training Manual (LOTM), Document LOTM 18, "High Pressure Core Spray (HPCS) System,"

Revision 3,6/22/89.

55. River Bend Unit 1, Gulf States Utilities Company, Drawing No. PID 27-4A, " Engineering P&I Diagrarn System 203 HPCS System," Revision 15, 11/8/89.
56. Washington Nuclear Plant No. 2 Washington Public Power Supply System, Drawing No. M520,
" Flow Diagram HPCS and LPCS Systems-Reactor Building," Revision 53,8/20/86.

57, Washington Nuclear Plant No. 2, Washington Public Power Supply System, System Operating Procedure No. *2.4.4, "High Pressure Core Spray Synem," Revision 4, 11/10/86.

58. Washington Nuclear Plant No. 2, Washington Public Power Supply System, Final Safety Analysis Report (NRC Docket No. 50-397), Sections 6.3 and 7.3, Amendment 37, August 1986.
59. Washington Nuclear Plant No.2, Washington Public Power Supply System, WNP-2 Systems Train-ing Handout No. 82-RSY-0902-T3, "High Pressure Core Spray System (HPCS)," June 1986.
60. Grand Gulf Nuclear Statien, Mississippi Power and Light Company, System Description No. SD-E22, "High Pressure Core Spray (HPCS) System," Revision 2.
61. Grand Gulf Nuclear Station, Mississippi Power and Light Company, System Operating Instruction No. 04-1-01 E221, "High Pressure Core Spray System," Revision 22, 1/8/87.
62. Grand Gulf Nuclear Station, Mississippi Power and Light Company, System Operating Instruction No. 04-1-01-R21-2, "4.16/6.9KV System," Revision 17, 12/9/86.
63. Grand Gulf Nuclear Station, Mississippi Power and Light Company, System Description No. SD-R21/P75/P81, " Engineered Safety FeaWre Power Distribution," kevision 0, knuary 1979.
64. NUREG/CR-5323, " Validation of Risk-Based Performance Indicators: Safety System Function Trends," J.L Boccio et al., Brookhaven National Laboratory, October 1989.
65. NUREG/CR 5641 (draft)," Study of Operational Risk-Based Configuration Control," P.K. Samanta et al., Brookhaven National Laboratory, September 199(1 8-4
i i

APPENDIX A Plant Spedlic System Information Appendices Al through A7 provide simplified ilPCS system flow diagrams for each of the BWR plants utilizing this system. These diagrams provide plant-specific information and nomenclature to supplement the generic HPCS system dewription provided in Section 2. In addition, each of the Appendices Al through A7 provides a plant specific walkdown checklist which has been modified to include the most risF-significant valves and components as determined in the RIG. This is intended to help the inspector to optimize his efforts by concentrating on the most risk significant components

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                                                                             . APPENDIX Al.

Clinton Power Station IIPCS System Details  ! l I l Al-1 l l c -- , .. _ , , . __

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l Table A.1-1 Modified HPCS System Walkdown - Clinton Power Station I. Electrical Lineup Dreaker ID No./ Description location Required Position Actual Position 1E22 C001 IIPCb INmp 4160VAC llus IC1 Cull 1(M Racked in 1E22-C003 IIPCS Water !rg Pump All MCC IC Cull 2C On 1E22 F001 Storage Tank Suetion valve AB MCC IC CUIl D On 1H22 F010 First Test Valve to RCIC Storage All Mt.C IC Cull 3D On Tank 1E22 F011 Second Test Valve to RCIC Stor. All MCC 1 CUB 3E On age Tank IE22-F012 Min flow to Supp. Pool Valve AD MCC IC CUD 4D On 1E22-F015 Supp Pool Pump Suction Valve AD MCC IC Cull 4C On 1E22 F023 !!PCS Test Valve to Supp. Pool All MCC 1C CUB 4D On IE22 F004 IIPS to Containment Outboard AB MCC IC Cull 2E On isot. Valve llPCS Area 120VAC AB MCC 1C Cull 3A(R) On 125 VDC MCC 3C Supply from Dinsmn 3 125 VDC MCC IC (CII-6) On 125 VDC Dattery Divaion 3125 VDC Uus Supply from 125 125 VDC MCC 1C (Cil 8) On VDC MCC 1C 4.16 KV Switchgear Dus 1C1 Dreaker Control 125 VDC (CD.11) On Power Control Room Panel Ill13 P001 125 VDC (Cil-17) On NSPS Div 3 logic. Test, and Xmtr Power - NSPS Power Distr. Panel C On Supply illl3-P663 (C71 ImolC) (CD-7) 1.ogic, Test, and X:ntr Power Supply 111134 NSl*i Power Distr. Panet D - On Po64 (C71 P001B) (CIl 17) ECCS IIPCS Pump Room Supply Fan A All MCC ICI (1 AP78E) On (IVYOSCA) ECCS IIPO Pump Room Supply Fan B All MCC ICI (1 AP78E) On ( IVYO8C11) - , Al-3 v v *ety+ m e vw ~ v e e e -w e s ---key - -- ---t w w w + +ove--,+m,-+-v-+w -mes-er w ev -+wweke- -W:--+e- -w e-s*<-we'

t i Table A.1-1 Modified HPCS System Walkdown Clinton Power Station (Cont'd.) .

               !!.           Valve Lineup                                                                                                                        l Valve ID No.                Description                Ircation                         Required Pai.           Action Position tion htOV                        llPCS buppression Pool     Control Room Panel III13 P601   .C osed IE22-0015                   Suction                    (itP Cub)

MOV llPCT Pump Mmimum Cmtiol Room Panet Illl3-P601 Cosed E22 l'012 11w to Suppr. Pool (llP Cub) MOV llPCs to Containment Control Room Panel Ill13-P601 Cosed 1E22 FtXM Outboard Isolation (Fuel Bldg. 755') MOV llPG Test to Suppres- Control Room Panel Illl3-P601 Cosed 1E22 F0:3 sion Pool (IIP Cub) MOV llPCS First Test Valve Control Room Panel Ill13 Ptol Omed 1E22 F010 RCIC Storage Tank (llP Cub) MOV IIPCS Second Test Valve Control Room Panel Illl3-Pt01 Cosed IE20 F011 RCIC Storage Tank (I(P Cub) AOV IIPCS Testable Qieck Control Room Panel 11!!3-P601 Omed 1E22 F005 (Drywell. 775') MOV llPCS Storage Tank Control Room Panet 11113-PtR11 Open 1E22 F001 Suction (IIP Cub) 1E22 0034 IIPCS Water 1.eg INmp Fuel llidg . 71T ImcLed Open Suction isolation 1E22 fW6 IIPCS Water leg Pump Fuel itidg , 710' incked Open Discharge Stop Check IE22-F3:3 IIPCS Itmp Suetion Fuel Biog., 710' open Pressure lustrument Root 11122-F3:5 ' IIPCS Pump Discharge Fuel illJg., 710' Open Pressure Instrument Rmt IE22-D 4A IIPC3 Pump thw thgh f uel Bldg., 71T Open Side Instrument Root for FT-1E22-N005 & N056 IE22 F3:4 11 IIPCS Pump 11w 1.ow Fuel Bldg., 712' Open Side instrument Root for IT-1E22-N005 & N056 1E22-F314 IIPCS Suetion isolation Fuel Bidg., 712' locked Open from Suppr. Pool 11122-F329 Suppression Pool Level Fuel Bldg., 737' Open Instrument Root 1E F331 Suppression Pool level Fuel D.'dg., 71T Open Instrument Root A14 a

Table A.11 Madified HPCS System Walkdown - Clinton Power Station (Cont'd.) Valve ID W Description location Required Pcsi- Action Pcullion tion 1E22-F38111 Suppression Pool !> vel Fuel 111dg , 712' locked Cosed instrument Test Corm. 1F?,2 F318A Suppression Pool 1.evel Fuel Bldg., 712' Oosed instrument Test Conn. 1E22-F380  !!PQ IJne Dresk Detee- Rx Illdg., containment Open tion trat. RNL 1E22-F00' IIPO Ilushing Water Fuel Illdg., 755' locked Cosed Supply

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1E22 F036 11PCS Manual Injection Rx. Pldg., Drywell 770' locked Open ljne Stop IE22-F318 IIPCS Suction from RCIC Storage Tank Room locked Open RCIC Storage Tank 1E22 F336 RCIC Storage Tank level l Fuel Bldg.,737' Open Instrument Root 1E22 F337 RCIC Storage Tank Level Fuel Illdg., 737' Open Instrument Root 11122-F335 RCIC Storage Tank Level Fuel Bldg , 737' Cosed Instrument Standpipe Vent 1E22 F342 RCIC Storage Tank level Fuel illdg., 737' Cosed Instrument Standpip Vent 1. Al-5

 . . . . . - . . .   .  . . - . . - - . - . - . - _ . - - . - . ~ - . _ . . . ~ ~ - .
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Table A.'t 1 Modified ilPCS System Walkdown - Clinton Power Station (Cont'd.) , III. HPCS Diesel Generator Refer to Table B 1," Proposed Inspection Plan for Diesel Generators at Nuclear Power Plants." Al-6

                                                                                                                    ]

APPENDIX A2 LaSalle Caunty Station Units 1 & 2 IIPCS System Details k e k d N A2-1

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Table A.2-1 Modified HPCS System Walkdown - LaSalle Country Station Unit 1

1. Electrical Lineup Breaker ID No./ Description location Required Position Actual Position IE22 CD01 !!PG Pump lius 143 CUB OtM Racked In i

IE22-5003 IIPO 1ransformer to MCC 10-1 Bus 143 CUB 005 Racked in IE1".-GJ03 Standby Water leg INmp MCC 143-1 CUB 2C On JE22 IT/.11 Cond. Starage Tank Suction Pump Valv;: . ,.C-143-1 CUB 2D On IE22-l'010 Full Flow Test Upstream Stop to CST - MCC 1431 Cull 2E On If!22 F011 Full How Test Downstream Stop to CST MCC-143-1 CUB 3A On 1E22 F012 Supp. Pool Min How Bypass Stop MCC-143-1 Cull 311 On 1E22 F015 Supp. Pool Pump Suction Valve MCC-143-1 CUB 3C On 1E22-F023 Full Flow Test Stop to Supp. Pool - MCC-143-1 Cull 3D On IVYO2C llPG Pump Room Ventilation Area MCC 143-1 Cull 6B On Cooler Supply Fan 1-IE22 FotM IIPCS Injection Discharge Stop MCC-1431 CUB 7C On IIPQ Instrument Power MCC 1431 CUB 2A On Hus 113 Supply from U-1 llattery 125 VDC Bus 113 On (Cil 6) l Ilus 113 Alternata Supply from U-2 llattery 125 VDC llus 113 Off j (CII-7) Switcbgeat 143 Breaker Control Power 125 VDC Bus 113 On (CD 11) - Panel till3-P625 IIPCS Helay logic and Pump 125 VDC Bus 113 On Motor Control (CB-16) Control Room Panel Illl3-P601 Valve Position Ind. 125 VDC Bus 113 On

j. (CB-17) l ===-

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9 Table A.21 Modified ilPCS System Walkdown . LaSalle Country Station Unit 1 (Cont'd) , II. Valve Lineup Valse ID No. Description location Required Position Actual Position MO IIPCS Punip Suction Control Roost Panel Open 11224V15 from Suppression 111131%01 (EL 67J' (Note 1) Pool Rx. Bldg. South of Supp. Pool) MO IIPCS INmp Control Room Panel Closed 1E22-Fol2 Minimum flow Stop till3.P601 (F1673' RL Bids. South of Supp. Pool) MO IIPCS ledection Stop Control Rimm Penel Closed if22-1M4 11113-I%01 (EL 761' Rs. Bldg. Almve & to the Side of C!tD g Filters) MO IIPCS Test Control Room Panel Closed IE22 tT)23 Discharge to 11113-I%01 (EL 694'. Suppression Pool 6 Rx. Bldg. Souni of Supp. Pool) MO IIPCS Test Control Room Panel Closed IE22.FD10 Diwharge to CST 111131%01 (EL 694'.

                                    , stream Stop         6" Rs. Bldg. South of Supp. Pool)

MO IIPCS Test Control Room Panel Closed IF22 d ll Discharge to CST lill3-I%01 (EL 694'. Dimnstressi Stoi. 6' Rx. Bldg. South of Supp. Poch AO 11105 Testable Control Room Panel Closed IE22 1TW5 Check 11113-I% 01 MO lifts INmp suction Control Himm Panel Closed - ll224Tkit from CST lill3-l%01 (EL 673* tNote 11 Rx. Bldg. South of Supp. Pimi) 1E22 F034 IIITS %eter les llPCS Room 673' Open Punip Suction Step level IW22-F006 IIICS Water leg IIPCS Room 673' Open Pump Discharge Ix*el Stop NOTE: 1. - Suppression ;xml is - e preferred HPCS source at LaSalle due to the biological corrosion problem discussed in Section 5.9. A2-4 l l

Table A.21 Modified HPCS System Walkdown - LaSalle Country Station Unit 1 (Cont'd) Velve ID No. 1)euription lustion Required Position Actual Position If221343 11PCS INmp Suction ll!CS Room 673' Opra Pressure Instrunwnt lesel Rm>t for 1%1D2 N003 & R001 1122 1332 IIICS Pump Illts Room 673' Open Diwharge Pressure level Instrument Root for 1%1122.N012A.B & IT 1C2 N004 I C2.F344 IIICS Pump IIPCS Hm>m 673' Open Dixharge Pressure level Instrunwnt Root for PIS 1D2.N013 IE22 1330 IIICS Pump flow 1110S 673' level Open Iligh Side South of Supp. Pool Instrunwnt Root for IT.lD2-N005 & N006 II22 l'331 11PCS Pump flow IIIC S 673' Iesel Oswn low Side Instrument South of Supp, Pa>l Root for ET.lE22 N005 & N006 1f221328 Suppression Pool SW Rx. Bldg. 695' Open Water level level Instrument Root for 13111R2-N002A & N002B 1D2 D29 Suppression Pool SW Rs. Bida. 695' Open Water level level Instrunwnt Root for ISil lE22 N002A & N002R { IE22-DN8 Instnnnent Test Stop S,W. Rs. Bldg. 695' locked Closed l fer 1311-lC2.N002A le el in Raceway

                         & N002B               against Supp. Pm>l IE22 Dtl9         Instrument Test Stop  S.W. Rx. Illdg. 695' Closed I                                         Level in Raceway for 1311 lE22 N002A
                         & N002B              Against Supp. Pool IE22 D90          lustrunwn Test Stop   S.W. Rx. Bldg. 695'  locked Closed far ISil.lE22 N002A   14=el in Raceway
                         & N0028               against Supp. Pool IE22 D91          Instrunwnt Test Stop  S.W. Rs. Hidg. 695'  Cin.ed l                         for 1311.lE22 N002A   level in Raceway
                         & N002H               against Supp. Pmil A2-5 i

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Table A.21 Modified HPCS System Walkdown - LaSalle Country Station Unit 1 (Cont'd) . Velve ID No. Description lustion Required Positien Actual l'osition li221303 tilCS se RL Veswl S.W. RL Itidg. 710' 0;wn' d'p Inst. Rawd for inel liy Drywell 15' l' DIS.ll22.N009 Overland 315' 18121.F347 Differentiall'ress re S.W. RL 111d3 710' Open Indiretor PDIS. Icel l>y Drywell 15' il22.N009 Overiwed 3 5' il22 11W13 nushing Water to S.I'. RL Ilids. 761' Closed til*CS trdettion 1Jne inet Stop Intmard it22-H)3M 111t$ Manual Containnwnt. South locked Open ladettion IJne Stop 7H3' -- 1E221302 tilts Suction from NE Side of CST luked Open CST ICYO33tl CST Instrunwnt Til 710' S.W. Cortwr Open Rand (IS.122 N(W)1II) n ICYO33C CST Instrunwn 111 710' S.W. Corner open Root (13122 N(Wil A) ICYO34B CST Instrunwnt lit 710' S.W. Corner Open Rimt (13122 N(Willl) ICYO34C CST Instrunwnt Til 710' S.W. Corner O pen R<ma (13122 N001A) { s b A2-6

             - Table A.1-1 Modified HPCS System Walkdown - LaSalle Country Station Unit 1 (Cont'd) 111. IIPCS DIESEL GENERATOR 1B Refer to Table B-1,' Proposed Inspection Plan for Diesel Generators at Nuclear Po.< cr Plants."

A2-7

Table A.2 2 Modified 11PCS System Walkdown LaSalle Country Station Unit 2 ,

l. Electrical Lineup lircaker ID NojDescriptian Incation Required Poution Actual Position _

21(22-(110 1 IIPO Pump ilus 243 Cull (04 Racked in 21i22 St03 IIPO Transformer ilus 243 Cull tx)$ Racked In 21!22-1:015 Sappreuion Pool Pump Suction Valve MCC 243-1 Cull 3C On 21122 l't101 Cond. Storage Tank Suction Pump Vahe MCC 243-1 Cull 2D On 21i22-1404 IIPO Injection Dncin:ga Stop MCC 24bl Cull 7C On 21222-10101:uli llow Test Ulstream Stop to CST MCC 24bl Cull 211 On 21!22 1:011 I ull l' low Test Dowmtream Stop to GT MCC 24bl Cull 3A On 21!22-1:012 Suppresuon Pool Min.1-kw Ilypass Stop MCC 243-1 Cull 311 On 2E2217023 l'ull Ilow Test Stop to Supluenion Pool MCC 243-1 Cull 3D On 2E22-QU3 Standby Water 1rg Pump MCC 243-1 Cull 2C On 2VYO2C IIPO Pump Room Ventilation Arca MCC 24bl Cull 611 On Cooler Supply l'an ilus 213 Supply frorn U 2 !!attery 125 VDC lius 2:2 On (Cil 6) llus 213 Alternate Supply from U l Itattery 123 VDC lius 213 OtT (Cil 7) S*1tchgcar 243 liteaker Control 123 VDC llus 213 On (Cil ll) Panel lill3 P625 llPCS Relay logic and Pump 123 VDC lius 213 On Motor Cnntrol (Cil 16) Control 14oom Panel 2ill3 lW)1 Valve Poution Inil. 125 VDC llus 213 On (Cll 17) 4MO VAC Power Supply to 123 VDC Ilittery MCC 2431 OJll 211 On O arger - Al-8

f. Table A.2 2 Modified HPCS System Walkdown - LaSalle County Station Unit 2 (Cont'd) II- Valve ljncup

                                                                                                                                                                                                                                                                      ~

Valve ID No. Dewription laation Required - Actual Pmition Paition MO IIPO Pump Suction from Control R om Panel 21113- Open 2E 2 l'015 Suppression Pool P601 (EL 673' Rx. Bldg. Soutn (Note 1) of Supp. Pool) MO IIPO Pump Minimuin Control Room Panel 21113- Omed - 2E22 F01: Flow Stop P601 (EL 673' llPG Pump  ! Room Pump Discharge) MO IIPO Inje: tion Stop Omtrol Romn Panel 21113- Omed 2E22-lMt P601 (EL 761' R Bldg. Above & to the Side of CRD Filters) MO IIPCS Test Discharge to Control Room Pa .cl 21113 . Omed 2E22 F023 Suppreuio Pad P601 (EL 604* 6* Rx. Didg. South of Supp. Pool) MO HPG lest Discharge to Control Room Panel 21113 O med 2E22-F010 GT Upstream Stop P601 (EL 694'-6* Rx. Bldg. South of Supp. Pool) MO IIPG 1est Dacharge to Control Room Panel 21113- G osed OE22-F0ll GT Dovmstream Stop 1%01 (EL 694*-6" Rx, Bldg. South of Supp. Pool) AO IIPO Testable Check Control Room Panel 21113 O med 2Li:2 F005 P601 MO IIPCS Pump Suction from Control Room Panel 21113- Omed OE22 F001 GT lYal (EL 673' Rx. Bldg South (Note 1) of Supp. Pool)- 2E22-FlXI3 F-:ushing Water to !!PG S E. Rx. Bldg 761; level by _ Cosed t injection Line Stcp CRD llydraulics Inboard 2Cm4 GT Supply to IIPG RB 761' S.E. Between IICU's O med . Plushmg Stop and RD Ilydraulics 2E: -F038 IIPO Manual Injection Containment South 783* 233* Imked Open U ne Stop 7 over head 2E22 F303 IIPO to Rt. Vessel d!p S V!. Rx. UlJg. 710* above Open Inst. Root for PDIS-2E22- DWEDS Pump Roorn

                                       - NW
ll:1-F347 Differential Pressure S.W. Rx. Oldg_ 710' above Open Indicator PDIS- E22 Nom DWEDS Pump Room NOTE: L Suppression pool is the preferred HPCS source at LaSalle due to the biological corrosion problem discusses in Section 5.9.-

A2 9 O ' - - _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ - _ _ - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ - _ _ _ _ _ . _ . _ _ _ _ _ . _ _ _ . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ . _ _ __m______

 . _---                              - . .. - -. - - - - - . .                                                          - . - .    . - - - - . ~ . -                         - . - .     - -

Table A.2 2 Modified ilPCS System Walkdown - LaSalle County Station Unit 2 (Cont'd) . I Valve ID No-' Description 12 cation Required Actual Position Position 2E22 F328 Suppression Pool b'ater S W. Rx. Illdg. 695' level Open level Instrument Root for Agsinst S Pool 1311-2E22-N002A & N00211 2E22-F329 Suppression Pool Water S.W. Rx, illdg 695' Irvel C pen LeSel Instrument Root for 13112E22 N002A & N00211 2F22-FTA Instrument Test Stop for S W. Rx. tildg 695' Level in locked Omed 1311- E22-N012A k Raceway against Supp. Pool N00211 2E22-F389 Imtrument Test Sky fr. S W. Rx,111dg 695' Level in Coced ISil 2E22 t'T12A & Raceway against Supp. Pool N00211 2E22090 Instrument Test Stop for S W. L.t. Illdg. 695' level in locked Cosed 13112E22 N002A & Raceway against Supp. Pool N002E 2E22-F391 Instrument Test Stop for S.W. Rx. Bldg. 695' I.rvel in C ased 13112E22-N002A & Raceway against Supp. Pool N00211 2E22-1 Gin llPO Water leg Pump IIPG Room 673' level Open Discharge Stop 2E22 F034 IIPG Water Ixg INmp ilPG Room 673' Ixvel Open Suction Stop i 2E22 F343 IIPG INmp Suction  !!PCS Room 673' level Open Pressure Instrument Root Section Side for PS-2E22-N003 & R001 2E22-F330 IIPCS Pump Flow liigh IIP 0 673" level South c.f Open Side Instrument Root for Supp. Pool South Side of IT 2E22-N005 & N006 - Station in Raceway 2E22-F331 IIPG Pump ilow Low , itPG 673' level South of Open Side Instrument Root for Supp. Pool South Side of TT-2E22 N005 & N006 Station in Raceway 2E22-F332 IIPG INmp Discharge llPCS Room 673' Irvel Oper. Pressure Instrument Root DiscL.rge Side for PS 2E22-N012A.- li & 1"T 2E22 N0lM 4-OE22-F346 IIPO INmp Discharge IIPO Room 673' level Open Pressure Instrument Root for Pts 2E22 N013 2 2E22-F302 IIPO Suction from GT SE Side of GT Incked Open l A2-10

Table A.2-2 Modified ilPCS System Walkdown - LaSalle County Station Unit 2 (Cont'd) Vahe ID No. Docnption location Required Actual Position Pmition 2CYO3311 CST Instrumt nt Root (IS 111710' 5.W. Corner Open E22 N(OID) 2CYO33C CbT Instrument Root (IS 111 710' S W. Corner Open E22-N(01 A) 2CYO34 D GT Instrument Root (13 Til 710' S W. Corner Open E22-Nu0111) 2CYO34C CST Instrument Root (IS. 111 710' S.W. Corner Open li'.2 N(olA) e bumi A2-11

m_ . . . . _ . _ . . . . _ _ . _ . . _ - . , _ _. . . _ _ . _ _ _ . ~ . ~ _ ._ _ . _ . . _ . ._ _ _ _ _ _ _ _ _ _ . - _ . Table A.2 2 Modified HPCS System Walkdown - LaSalle County Station Unit 2 (Cont'd) .

                          !!!.             HPCS DIESEL GENERATOR 2B Refer to Table B 1," Proposed Inspection Plan for Diesel Generators at Nuclear Power Plant".

i-t: l l l l l i l l t 5 I l. l l --- l I A212

          ,                                                                      .i a

APPENDIX A3 Nine Mlle Point - Unit 2 IIPCS System Detalla i A3-1

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Table A.3-1 Modified HPCS System Walkdown Nine Mile Point - Unit 2 l

1. Electrical Lineup Dreaker ID No/ Description location Required Pmition Actual (Note 1) Pmition 2Gll*P1 (CUO1) IIPCs Pump i 2 ENS'SWG102 DKR 4.7 Racked in  !

2CSil*P2 (CD03) IIPO Water 12g Pump 2 2 Ells'MCC201 Cull 5A On 2CSil'MOV101 (F001) IIPO Pump Suction 2 Ells *MCC01 Cull 3A On from GT 2C311'MOV105 (F012) IIPCS Min Flow 11ypass 2 Ells'MCC01 CUB 3B On- , 2CSII*MOV107 (F004) IIPO Pump 1 injection 2 Ells'MCC01 CUB 2C On 2CSil'MOV111 (F023)IIPCS Test Return to 2 Ells'MCC01 Cull 10D On Supp. Pool 2CSil'MOV110 (F010) IIPCS Test Return to 2 Ells'MCC01 Cull 6B off CNT (Note 2) 2bil'MOV112 (F011) IIPO Test Return to 2 Ells'MCC01 CUB 7B On CST 2CSil'MOV118 (F015) IIPO Pump Suction 2 Ells'MCC01 CUI) 3C On from Supp_ Pcd 2CSilN10 Relay logic 2CES'IPNLA14 BKR 16 On 2CSIIN11 Relay Logic 2SCV'PN1200P UKR 17 On 2C311N12 Valve Position Indication 2CES'!PNL414 IlKR 17 On 2CSIIN13 Valve / Relay legic 2SCV'PNIl00P llKR 16 On NOTES: 1. Standard GE component numbers given in parenthesis.

2. Assures compliance with 10 CFR 50 Appendix R requirements.

A3-3

Table A.31 Modified HPCS System Walkdown Nine Mile Point - Unit 2 (Contid) , II, Valve Lineup Valve ID No. Description location Required Paition Actual Position (N<xe 1) 2Gll*llCV120 Injection Manual Viv Gmtrol Ruwn Panet P601 texked Open (lin8) (Rx. Illdg.) 2Gll'MOV112 'lest Return to Condensate Gmtrol Room Panel 1%01 Shut (F011) n ( Rx.111dg.) 2CSil'AOV108 Injection Testable Occk Gmtrol Rm Panel 1%01 Shut (F005) Viv (Rx.111dr 2Gll'MOV110 Test Return to Omdensate Omtrol Rcxwn Panel f%01 Shut (1 010) n (Rx. Illdg ) (Note 2) _ 2Gil'MOV107 Pump 1 Injection Viv Omtrol Roorn Panel 1%01 Shut (l'Ot)4) (Rx. Illdg.) 2CSil'MOV111 Test Retum to Suppression Omtrol Room Panet 1%01 Shut (F023) Pool (Rx, Illdg.) 2Gil'MOV101 Pump Suct from Cnds Tk Omtrol Room Panel 1%01 Open (F001) (Rx. Illdg.) 2Gil'MOV105 Mmtmum lhw flypass Viv Control Room Panet I%01 Shut (F012) (Rx. lindg ) 2GII'MOV118 Pump Suet from control Rmm Panel Iwol Shut (l'015) Suppression Pool (Rx- Illdg ) 2Gil,V37 2CNS-1K111 Outlet OT 111 12xked Open 2Gll'V107 *LT3A inst Root Isolation GT 1D Open 2Gil'V108 '111311 Inst Rmt isolaton GT IB Open 2Gll'V59 Pl Suction Check Valve llPG Room Installed 2011'"" 'P2 Suction imi ilPG Room Open

 .         2G11*V17             *P2 Discharge Check                       IIPCS Room             Installed                                                            R i

2CSil'V55 'P2 Discharge Occk llPCS Room Installed . 2CSII'V15 'P2 Discharge isol IIPC Room Open 2 Git *V96 *P2 Rectre IJne Throttle tex ked ' throttled (Nixe 3) NOTES: 1. Standard GE component number given in parenthesis. ,

2. Assures compliance with 10 CFR 50 Appendix R requirments.
3. With *P2 running, licensee throttles 2CSH*V54 and 2CSH*V% as required to clear annunciators M11719 and 601720 while maintaining HPCS system pressure > 65 psig.

A3 4 1 i I

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r i TaNe A.31 Modified llPCS System Walkdown Nine Mile Point . Unit 2 (Cont'd) Valve ID No. Description laation Required Pmtion Actual Paithm l (Note 1) 2Gil'V34 'P2 Recire line Throttle taked ' throttled f (Note 3) 2Oll'V18 'I' llc 2, Pflui Iml Open 2Oll'V27 *17103, Pil:8Isol Olen 2 Oll'V9 'PI Discharge Occi installed Mit *V23 '17104,103 isal Open i / 2Oll*V26 'ITHM,10$ Ind Open 2Oll'V23 'P1115 Inst Hoot lio! Open _ 2Gil'V32 Suction hping Drain to Shut Rad *aste Isol 2Oll'Y16 Supp hul duction hping Installed , (F016) G eck l i 2 Oll'V37 *l3143 Inst Hoot Isol Oren 2 Git *V56 *l.S143 trat Drain Shut l 2 0ll'V94 '1510 Inst I' tain , Shut and Capped 2 Gll'V58 'll .O Irnt Vent Shut 2Oll'V93 'I A43 Inst Vent Shut and Capped 2Gil'V31 Condensate Makeup taked Shut isolation 2C- 30 Condensate Makeup I -Acc shut li.olation 2 Oll'V45 'PD'll(W trut Root Iml Open 2Oll'V36 1.T123,124 Iml Open __ 2Oll'V99 LT Test Connection Shut 2Oll'V100 LT lest Connection shut and Capped 2Gil'V79 1.T124 Vent Shut and Capped 2 0ll'V80 LT123 Vent _

                                                                                                                                          ,      Shut 6j Capped NOTES: 1, Standaid GE component i. W ;iven in parenthes s.                                                          i
                                             - 2. Assures compliane: with 10 f.J h 50 Appendix R requirment'
3. With 'P2 running, licensee throttles 2CSil'V54 and 2CSil*V96 as required to clear annunciators 601719 and 601720 while maintaining flPCS system pressure > 65 psig.

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i i Table A.31 hiodified itPCS System Walkdown . Nine Mile Point Unit 2 (Cont'd) - 4 i Valve ID No. Descrigtson 14cauon Required l'oeition Actual limitum j (Note 1) l Dren l 2Gil'V82 1.T124 isolation I l 2 G11'V76 L1123 Isolation Oren 2Gil'V77 LT123 ladation Dren 2Gil'V73 LT124 laolation Dren i 2Gil'V78 1.T124 Dram Shut and Cariged 2Gil'V81 1.T123 Drain Shut and Carled . 2 Gil'V101  !!!' Test Conn shut 1 2GII'V102 I.T Test Conn Shut and I' lugged 2Gil'V35 'Lil23.124144 Ol en l NOTE: 1. Standard GE component number given in parenthesis. e l NOTES: 1. Standard GE component number given in parenthesis.

2. Assures compliance with 10 CFR $0 Appendis R requirments,
3. With *P2 running, licensee throttles 2CSil*V54 and 2CSIl*V96 as required to cle-annunciators 601719 and 601720 while maintaining ilPCS system pressure > 65 r i-A3 6 .
                                                                                                                                                                                )

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1 i 9 i Table A.31 Afodified itPCS System Walkdown - Nine hiite Point Unit 2 (Cont'd) ill. IfPCS DIESEL GENERATOR  ; I Refer to Trble B 1.

  • Proposed Inspection P;an for Diesel Generators at Nuclear Power Plants".

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I i Table A.41 Modified ilPC5 System Walkdown Perry Nuclear Power Plant ]

1. Electrical Lineup lireaker ID No/ Description 12 cation Required Ptmition Actual Pmithin l 11122- 0 101 IIPCS Pump Dus 111113 flKR Racked in 1:111304 l 11!22WR13 dit31rarnfortner tillR 1.E to MCC flus Elll3 IIKR Rncied in I!It 1 El Elll303 1 1E22.UK)3 Standtiy vater !rr Pump MCC El 1111 O med Oneonneet Switch C 1E22 l'001 Cond. Storage Tank Suction Pump Valve MCC El:1E.1 Oceed Disconnect Switch D 1E221:010 I ull llow Test Upstream Stop o C3T MCC III 11!-1 Omed Disconnect Switch G li2217011 !!ull l'hw lest Dowintream Stop to CST MCC El:11b1 Omed Disconnect Switch 11 Il!22 l'012 Supp. Pool Min.110w D 3pass Ship MCC EF1E.1 Omed Daconnect Switch J 11!22f015 Supp. Pool Pump Juction Valve MCC EF1E.1 Ocned Daconnect Switch K 1E22 l'023 Full flow Test Stop to Supp. Pml MCC EF1E 1 Omed Disconnect Switch L llPCS Pump Room Ventilation Area Cooler Supply MCC El III-I Omed I an Disconnect Switch R 11!221;(EM llPG injection Discharge Stop MCC EI'1E-1 Cosed Disconnect Switch F IIPCS Instrument Power MCC 1431 CTil 2A On ,

i IIS VDC HUS Ed 1-C Supply from U 1 Itattery 125 VDC Dus 113 On-(CD-6) l . Switchgear itun Ell.13 Dreaker Conitol Iwer 125 VDC Distr. Panel On (0111) -r !. Panel 11113 P625111'0 Relay 1.cgic and Pump 12$ VDC Distr. Par On Motor Control (0416) Control Room Panel 11113-Po01 113 VDC Dntr, Panel On (Cil.17) 4h h A4 3

 - _-. _ .-. _ - . _ _ - _ _ -... - .-.-                                             . -           - - . .       --.-.a_-...           - _ _ _ ,

Table A.41 Modified llPCS Systern Walkdown Perry Nuclear Power Plant (Cont'd) ,

11. Valve Lineup Description location Required Actual l\*ition Valve ID No.

Position M O li!22 l'015 IIPO INmp Suction from Omtrol Roorn Snel O(sed 1111?1441 Ri llidg Sultression Pml IIPCS INinp Minimum 11o* Omtrol Room Panel Ocsed MO 1122 l'012 Stop 11113-1Y01 Rt. Illdg MO il:22-110t IIPc Injection Stop Control Roorn Panel amed 11113-14 0 1 Rx. Illdg. MO 11021023 - llPO 1es' Discharge to Omtrol Rmm Panel Q *ed Seppression Pml 1111341Yo1 Rs. Illdg MO 111221 010 11P0 *le$t Discharge to GT Omtrol Rmm Pane! Owed U lstream Stop 11113-14 01 Rt. Illdg. I!PO Test Discharge to CST Omtrol Room Panet Ocmed MO 11021011 Dmstream Suy 11113 P601 Rx. Illdg. A O 1LQ2 1005 IIPO lestable Cieck Omtrol Rmm Panel dosed Ill13 Pt01 MO 1E22l'to) IIPG INmp Suction from CS1 O ntrol Rmm Panel Open ill13 lYiot Rx. Itidg. Il22in14 IIPCS Water leg INrnp Suction llPG Room 673' Open Stop level , 11021 tui IIPO Water Irg INmp llPO Room 67.t' Open Discharge Stop l evel ll:22 F305 IIPG INmp Suction Pressure llP O R m m Open Instrument Rmt for PT-1E22 05: & Pl.1102 R001 1122 I:506 ll.*CS Pump Dacharge Pressure llPG Room Open - Imti iment Root for PI 1122 051 A PIS 1122-N651 ill22 F51:A IIPCS Pump ilow liigh Side 11P0 673' level Open Instru sent Root far IT ili:2- Smth of Supp Pool Nto$ i N056 itC2 F51:fi ilP' $ INmp 110w low Side llP O 673'Ievel Open in .rument Root for 1 4 11122- South of Supp. Pool N 05 & N056 11122 1 '.uppress on Pool Water Irvel Rt. Bldg Open Instrument Rmt for 1.1S-1102-N655C A 655G di:21: Suppression Pool Water Irvel Rt. tildg. Open Instrument Root for 1.lS-1E 2 N655C 6550 b A4-4

Tatile A 41 Modified lil'CS Sysicm Walkdown - l'ctry Nuclear Power Plant (Cont'<l) Wlve ID No De gript hin laation Requirtd Actual I'mitnin I'mition 1122I- Instrument test Shy for I1% Rt itidg lesked O mtd 11122.N65?C A N6*?O lii22 F 1mtrument lest Stop for 1.15 Rt IUJg 0:med 21:22 N633C A N63?O II7 22-I- Instrument lent Stop for 1.15 R4 Itidg O med 21i:2 N635C & Ne550 1E22l' instrument Test Stop for !IN lu Inds Omed 11:22.N655C & N6550 11:22-12 IlPO to Rt Woel d'p Inst Rx Itidg 0;rn _ Rcut for PDIN11122 N[m lii211- thfferential Preuuic Indicator Ha lilJg Ogwn PDIN11:22.Ntm 11:22 im1 Ilushing Water to IIPO lu Ittdg O med Iniection IJnc Stop Intmard 11:22-10 % llPO Manual Ingetion ljne Rx il1Jg , Pnmary in ked OFn Stop Oint ainme nt _. 11:2 2 I IIPO Suction f rom o f GT tesked Open OT level T ransnutter Cf ' Ope n Instrument lunt *.'alves

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E .1 Table A.41 Modified flPCS Sptetu Walkdown . Perry Nudcar Power Plant (Cont'd) ,  : i l 111. IIPCS DIESEL GENERATOR EIC itefer to Tame 111,'Prormed inspection Plan for Diesel Generators at Nuclear Power Plantv,  ! t F t t i i I L 1 t I F r h O e i

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Table A.51 Modilled lil'CS Systcm Walkdown River llend Station - Unit i

1. I!!cctrical Lineup

} lireaker or Switch ID No 1)eu ription 11x ation lleqwred Pmition Actual l'osition 11:22 l'OU1 IIPC3 Motor I ceJ SWGR 1:::*9 M llKR 02 Racked in Open tNote 1) li 22 Pond Sutantem till "urnp MCC 1:: *Nrt2 ItKR 2C On 1122 MOVitol Cond Storage lartk Pump MCC 122'9n2 ItKR :D On St;dion Valve II:22 MOVI 010 CnnJ $'orage lank fest MCC in:'No: liKR 3D On thpau Valve _

          !!22 MO\T011 Cond Storage lank lent     McC !!22' Nun IIKR 31t                                                                   On livpaa Valve III: MOV194 IIP (3 Pump Discharge       MrC I:22'NO2 ItKR 2h                                                                     On Valve                                 _

1E22 MOV1023 Suppreuinn l'ool Test MPC I22'No2 ItKR 4D On flypara Valve IE 2-MO\T01. 'Supprenton Pool Pemp MCC 12 '902 IIKR 4C On Suetinti Va!ve li22 MOV101: Suppreuntn Pml Min MCC 1:2 '90: HKR 411 On Ih w flypau Viv 11?:2'P(in1311PCS ljne 1 ill Pump lil13 Pr 01 Neutral Af ter stop 1120*POU1 IIPCS Pump supply itKR 11113 l'rol Neutral After 11:22 Arlto: lnp N O Tli- 1. Check fuses installed, charging motor switch on, and charging springs indicate charged. A5-3

Table A.51 Modified llPCS System Walkdown - Iliver llend Station Unit 1 (Cont'd) .

11. Wlve Lineup IW Vahe ID No. Description l a atm n stequired Position Actual Paitico ll!:2l'013 llPG Pump buttressmn Pool 1111LlY.01 Auto After Cae Suction Whc Ili:21 tol llPG Purnp Cf Suction 11113-IYo1 Auto After Open Valve 11:22l'012 11P0 Min Ihwc Valve to 11113 P601 Auto Afbr Ome 6 Supprtssion Pool 1E22-1:023 IIPG Test iteturn to 11113 P60) Auto After Ose suppression Pmi 1E:21010 IIPG Test livpass to GT lill3 P601 Auto After Owe 111 2Ioli llPG 'lest Return Vahe to lill3-P601 Auto After Case GT IE22luu llPG inicction isolation Vahe ill13 P601 Auto After nose 11.22' AOVITIO* llPG 1estable Geck Vahe llI1Ll%01 Chmed 1Gll'V1 IIPG and RCIC GT 5uctmn Auxiliary liudding. I!!. Inked Open Imlation Valve '?4 6 1Gil'V3 {1E22 I"IN05: & PiltColl Root Auxiliary lludding. EL Ope n Valve 74' 1Gil'V6 {lE22 l'IN051] Root Vahe Auxihary lludding. liL Ope n llPG INnip Diuh Press 748 1E :'\ I 034 IIPO fill Pump Suction Yahe Auxihary Ituilding. PL Open 84i 1E22*Vilut IIPCS Fill Pump Dischange Auxiliary liudding. EL Open Valve 841 1 Gil'V24 [1E22 l'EN(un] Root Valves Autilisry fluilding. El. Scaled Open IC511* V2 IIPG Ilow 955 .

II'22 * \T003 IIPO Disch line I' lushing Auxiliary liuilding. EL O wed 4 Water Supply 1461 1E20'MOVi OO4 IIPO Injection to km Auxiliary Iluilding. EL Osed 1468 1 011*V28 (IE31-PD1N081] Irak Containment. EL 1:75 O p:n Detection isolation Valve A/.286 E :*\T036 1[PG Disch to Rx, Manual Drywell EL 1465 AZ Laked Open isolation Vahe 270 1Gil* V17 [lE51 L'INO35A] Root Valves Puct lluildmg hping Open 1Gil'Y 16 GT level ^!unnel A5-4 1

I . l Table A.5-1 Modified llPCS System Walkdown River Bend Station - Unit 1 (Cont'd) Valve ID No. Dewrirtion Incation Required 1%ition Actual lhition 1Gil'V18 [11!.22 !!!NO34C) Root Valves f uel Iluilding l'ipng 0;en 1 0ll'Vl9 G T Irvel Tunnel E A3 $

Table A.51 Modified ilPCS System Walkdown - River Bend Station Unit I (Cont'd) , Ill. HPCS DIESFl. GENERATOR IC Refer to Table B 1,' Proposed Inspection Plc. for Diesel Generators at Nuclear Power Plants."- AS-6

Al'PENI)lX A6 Washington Nuclear Plant No. 2 Ill*CS Sptern Details A6-1

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4 Table A.6-1 Modified ilPCS System Walkdown War.hington Nuclear Plant No. 2

          !,          Electrical Lineup                                                                                               l lireaker ID No/ Description                location                 li tred Paithm      Actual Pmition (Note 1)                 (Note 2) i             lilo P.! liigh Pter,sure Core Spray        SM 4                     Racked in                                           !

Pump llPO P 3 IIPO Water leg Pump MC4 A Cull IC O med IIPO V 1 IIPCs Suction from OT MC4A Cull 2D O med llPCS-V-4 IIPO Injecthe Wlve MC-4 A Cull Sil Ocoed i IIPO V-10 llPO Intmard Return to CFr MC-4A CUB 2fi Omed IIPO V.11 IIPO Outtonard Return to MC-4A Cull 3A Omed 2 CFI' I IIPO V 12 IIPO Minimum flow MC-4A Cull 3D Omed e i llPO-V.15 IIPO Suppresskm Pool MC-4A Cull 3C O med Suct on IIPCS V41 IIPO 'lest Valu MC-4 A Cull 3D Ormed ,

                                                                                                                                    +

IIPO V 3. IIPCS V-Si PP4AC119 O med PP 4A nt 11 O med llP O_ V5 l'id (f!?.2 Nm4. Pump Disch. Press-) PP 4 A Ot 8 O med FT 5 fli22 N(o$-System flow) IIPO logic 125 VDC Ot D 7 Omed 1100 Dist IIPCS V.10. IIPO V.11 Pmition 125 VDC C11 D 9 Mored Indication llPO Dist t t I NOTES: 1. Electrical equipment physicali> kuted in DG Room, t

2. 'I.icensee personnel are instructed that if these breakers are open, not to clo.c them until directed by the Control Room Operator, and to see filling ar i venting insi uctions. .

A63 A

 . _ .m            .. _ _ _._ - .- _ _. -                                 . _ _ _ _ _ _ _ _ _ _ _ _ _ -.. _ .._ .- .._.- _ _ _ _ _ _ ._ _ .

i

                                                                                                                                                                                                  '         I Table A 61 Modified ilPCS System Walkdown . Wash;ngton Nuclear Plant No. 2 (Cont'd)                                                                              .
                                                                                                                                                                                                            )

II. VALVE LINEUP l Valve ID No. Denenptam location: llid;. Required Actual Pmitice (Note 1) Islev. Pt* tion .- 1 suction from OT MOV Ril-422 Open IIPO.V.1 (1101) (Not.e 2) IIPC5-V-701 Rad Valve for PIS-3 (ii.22- Ril.422 Open N(r9) Suctum Pressure j llM5.V49 IIPO.P.1 Seal Drain Hin-422 Open (N4e 2) IIPO-V41 IIPO-P.1 Seal Drain 11i1-422 Ol en (Note 2) til*G-P.1 Mmimum lhw RIl-422 G osed f IIPO.V 12 (I'012) MOV . IlPO-V-53 Minimum lhw IJne isolatum Ril 422 tecked opn (Note 2) IIPO.V.34 (1034) IIPG-P 3 Suction isolation R11422 '3prn IIPG-P-3 Minimum ikw RB-422 Open llPO.V.77 (1033) 17PCN V4 (flu ri) IIPG.P.3 Stop O eck RIl-421 Open llPG V 7ml Rmt Valve for PIS 13 (water Hil-422 Ol en les pump discharge press.) (Note 2) IIPO-V.709 A Rmt Valves for IT.5 (12h HH 444 Open V-710 Nm5), l'154 (II22.Nmri), & l'). (Note 2) f03 (li22 R603) IIPG.V 15 (1015) Sqpresuon Pool Suction MCW 1111 444 Gosed IIPc suction 'lic to RIIR Ril-444 takt , Omed 1[PO.V.19 (I'019) itPO V.10 (ibl0) IIPO P.1 ' lent to GT MOV Ril-444 Conco llPO V.11 (1:011) IIPCs-P.1 Test to GT MOV Ril444 Ocv.ed llPO-V43 (F023) IIPO-P 1 Test to Suppienion RH-644 C osed , 1%d MOV llPG Pl VX.732 IIPC-DPl5-9 (45*) Ril.536 Open DW.547 Omed llPG.V-3 POS) l Testable Occk Valve h NOTES,1. Standard GE component numbers given in parentheses.

2. Licensee requires independent verification of this valve position by its personnel. .
3. Valve capped.

A6-4

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' Table A.6-1 hiodified llPCS Systern Walldown Washington Nt:lcar Plant No. 2 (Cont'd) Valu ID No Descnption laatum illJg . Itequired Actual l'esition (Note 1) 1:1cv. l'ositum IIP (NV 4 1 uture St C Connection DW 547 Omed Isolation (Note 2) (Note .1) IIP ($V 17 Test Connedian isolatu,n Yahe DW 547 (b.e d ( $ $ l'. 2%*) (Note 2) _ IIP (NV 38 lest Connectiim Imla:km Valve DW 547 O<med ( $ 51'. 2.%") (Note 2) lil'0NV 51 (I OV) Ingetion lane Imlation Vahe DW 547 laked Open (551'. 24t_r) IfP(EV 3 (Im3) Conderaation alushing bupply 1(14 522 la ked Otwed __ Imlation llPCNV.4 (f uW Injection lane MOV ll11522 Otwe d COND V VA  !!Ptb Suetion from GI (31 Area laked Open Imlation Vahe (Nde 2) COND V Vli IIP (3 Suction frorn (.51 GT Area Inked Ogn isolation Valve (Note 2) NOTES: 1. Standard GE component numbers given in parentheses.

2. Licensee requires independent verification of this valve position by its personnel.
3. Valve capped.

A6-5

Table A.61 Modified llPCS System Walkdown . Washington Nuclear Plant No. 2 (Cont'd) , 111. IIPCS DIESEL GENERATOR Refer to Table Il 1. 'Prolx) sed Inspection Plan for Diesel Generators at Nuclear Pe er Plants." t a e A6-6 l

_ _. ._. . . _ _ _ _ _ . . _ , _ . _ . _ _ _ . . _ _ _ _ _ . _ -.___ _ _ . _ ._ _ . _ . _ _ _ __.. _ _ ._ _ __ m P r 0 , i', e APPENDIX A.7 t Grand Gulf Nuclear Station . Unit ! IIPCS Sptem Details

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e Table A.71 hiodified llPCS Systern Walldown - Grand Gull Nuclear Station Unil 1

1. Electrical 1.ineup litcaker !!) No ;1)nenptnin in atem Required Pmition Adual Pinitnin (101 llP(51%mp Motor ilUS 17AC ltKR 152 170 Ra(bed in (Note 11 om IIP (5 Siandt.y water irg th. Ley) MCC 171VJ1 ilKit $217v105 Orwed Pump ___

(Note 2) 1001 IIP (5 Pump Surtnwi from (5'l MUC 171601 ItKR 52170)(r. Omed I(nl llP(5 Injectnin Shutoff Valve MCC 17160114KR 52 l0101 Omed I(119 llP(5 int.oard lest Iteturn to WI MCC 171501 ItKR $2170107 Omed - 101118P(5 Outt.oard lest iteturn to (51 MUC 17'101 llKR 52170ltat ( Leed W rol: IIP (3 Mm Itw to Supp Pml MCC !?tini liKR ? .1701m O ued 1015 llP(3 Purnp Suction frorn Supp MCC 171501 ItKR $2170110 Omed Pml l'023 lil'CS 1rit utturr. to $.ipp l'ool MCC 171t01 ItKH *2171111 Omed 113111001C llP(5 l' ump itoorn ('im der MCP 171401 IlKR 52170117 Ocord Ian (101 llP(5 Pump Motor Sparr lleater PNI.17P11 IlER $21P71122 Omed Panel IlllllWil Wlve I'mition l'NI. til22.P11R 7211C15 Omed indsation . Panel 1111L107 PNI. Ill:2 PilR 7211017 Omed l'anci 11113 li.2%11P(5 l ogic/r introl PNI. Ill:2 P118 7211C18 Omed Volt a ge Ground 't ransformer lius Relapng and PNI.18122 l'Il8 72 11C19 n med Voltme te r ammmmm n-r . w ~;a_n _ _.ww -aw. -- NOTES: 1. 152-1702 will be racked in during the fill and vent instructions,

2. 52-170105 the llPCS 1 >ckey Pump breater will be closed during the fill and vent instructions.

A7-3

i Table A.7-1 Modified llPCS Sptern Walldrun - Grand Gulf Nuclear Station Unit 1 (Cont'd) s ll. Valve Lineup I Descriptitm la athm Required P<w. tion Actual Pm:00n

    ;  Valve ID No 111'(3 Purnp Sudion fitwo                                              Illl3 Pr.01 IIS-M'ol(S 1)      Auto Ogn luil LST 10'2                IIP (.3 Min 110 to Supp                                               11113 IWil llS Mfo$ (S 12)     Auto (hme d Pml                                                                                                                    .

l 015 IIP (3 himp Sudion from Illl3-PtiO! IIS-Mfm (S-15) Auto O(sed Supp Ptul 1023 IIP (3 lest Return to Supp. 11113-lYhl llS Mrdb (SE) Auto (hmed tw4 lud llPC3 triection Shutoff lillhProl llS Mf.01 (S-4) Auto Chaed Valve 1010 llPC3 Intoard 'lest Return Illl3-Pf>01 IIS-M'i(17 (S-h)) Auto Ossed to ml - 1 011 IIPC3 Outhoard ~lest Heturn 11113 lYol 11S My* (511) Auto Ch*cd to (3T IIPr3 lestable Ched Valve Illl3-lYol llS M'02 (S 5) (h med 1015 IP111021 (31 Supply to llPWHC'lC ;31 Dike Area I nked Oren 1 034 llP(3 Jockey Pump Suction Area R llev 9T laked Open l i n v. IIP (3 Ja key Pump Area 8 Ilev 93' Ilandwheel Open Discharge Stop (heck 1-'210 Supp Pml Suct Ijne lest Area 8 lilev 9T (1med Connection Shutoff l'Ul? Supp Pool Suct ime 'lest Aica M lilev 9T CA,3ed & (hwed Connection Imlation I'019 IIPCS l' lush to liq Arca H 1 lev 9T laked Omed Radwaste Surge Iank l 019 Stop (lictk Around 1010 Area 81 lev 9.Y llandwheel Open i XO50 PP N404 Area 8 lilev 9T Capped Omed nrm 13002 P1 N050. Jaley Pump Pren Area 8 Ilev 93 Open 4 l'X(n)3 A Hoot Valves for !!22 l*] Area 8 Elev 9T Open 1% s u N005 & l'I N056 (llPC3 Pump Diwh ) I'X(n i$ l'! N051. Pump Dncharge Area M l. lev VT Open Preu IN ut Pl Hunt Area R IJev 9T Open 1*X(un PI H(o Area 8 llev 9T ( >1e ri A74 I i i i

e Table A.71 Modified llPCS Systern Walkdown Grand Gulf Nuclear Station Unit 1 (Cont'd) Valve ID No. Descrigxion Instion itequired Paition Actual Paition FXO20 1 P N401 Area F llev 93' O(ned I'XO22 PP N403 Area 8 lilev 93' O(wed I'X019 PP N4(O Area 8 lucy 9.1' Omed 1 031 Ilushing Wtr Supply Shutoff Area 8 Elev 119' Inked Omed 1103 11ushing Wtr Supply Imlation Area 8 lilev 119' locked Omed IXO3R LT N054C & G CST Level Area 7 Elev.119' Ogen ll21 IX(r.6 t #T NOR) ( Atxwe Core Area 11 Elev 147' Ogn ate Tap) IIPO l.rab Detection i E31 IX025 PDT N081,IIPO trak Area 11 Elev 147' Olen Detect. - 4. I:0% IIPO Injecten to 16 Ares 11 Elev 147' incked Open isolation S I 1 i A7 5 l

                                       . . .     . . - . . .          . _ . . .        . ~ ~ - . .    - .-           ,_   _ _ - . _                - ,        ._         . . - , .., .., .-- .,

1 i Table A,71 Modified llPCS System Walkdown - Grand Gulf Nuclear Station Unit 1 (Cont'd) , 111. IIPCS DIESEL GENERATOR C Refer to Table Il 1,' Proposed inspection Plan for Diesel Geacrators at Nuclear Power Plants." i A7-6

e Al'I'l;NDIX 11 l'roposed inspntion 1*lan for Dieul Generators m B-1

- ._ -- _ - - - . - . . . - - . - ~.. - - _ Table iM Proposed Inspection Plan for Diesel Generators at Nucicar Power Piants , A. Obiectives To review and evaluate Diesel Generator design, operation, an.1 maintenance at NPPs to ensure that the DGs will be available when needed to power safety systems D. Details

1. The inspection of the following items should focus oa DG auxihary systems as follows:

Fuel injection System, Turbocharger, Startirg System, Speed / Load Control, Jacket Water, Cooling Water, Lube 0;l, Fuel Oil, Ccmtrol and Monitc. ring Systems, and Generator.

2. Using the LER,50.55e, and Part 21 systems cotoputer printout, select 3 recent failures (within 2 years) for followup at the NPP. When at the plant select an additional 2 failures from the internal systems, Evaluate the licensee's response a these failures for proper failure analysis, corrective action, notification of vendor, Part 21 evaluation and documentation.
3. Maintenance: Refer to IE 1.P.s 62700 and 62702, as they apply to DG n.. .enance.

Additionally, does the NPP have, and have they implemented the DG vendo6 maintenance recommendations (especially those recommendations unique to nuclear service DGs such as Colt's described in NSAC-79)? Are maintenance personnel specially trained on DGs? Is failure information fed back into maintenance program?

4. Design Change Control: Select two DG modifications and verify nroper .

implementation. Utilizing information from DG vendor inspection on modific.tions recommended. verify thL NPP is receiving all pertinent information in this area from the vendor. (Reference IE I.P. 37700).

5. Spare Parts and Procurement: Review how spare pi.rts and services are purchased nd parts stored, both from DG vendor and direct from sub-vendor. Verify adequate Part 21 and OA, particularly when vendors, are only supplying commercial grade parts and services (e.g., Woodward Governor und Stewart and Stevenson). _ Verify ASME code specified where appropriate. Tour spare parts storage area. (Reference IE I.P. 38701 and 38702).
6. Training: Ensure appropriate DG Apecific training given to maintenance, operations, OA, and management yrsonnel. Are there adequate documents to describe DG operatien onsite (both main engine and auxiliary system)? (Reference IE I.P. 41700).
7. Observe DGs in operation. Ensure they run smoothly and are operated per procedure.

Imok far abnormal vibration and leaks (air, fuel oil, or lube oil). Check that reding are within specified limits. Are limits per DG vendor recommendations? Are recommendatioc clearly specified? Is air- quality in DG room satisfactory without excessive rut? Are control cabinets properly gasketed? Are instruments calibrated? Is trending of operating data performed to detect degradation early? B-2

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  • 8. Is NPT receiving all appropriate sersice information from vendor: design, maintenance, operational, etc? This is specially important for General Motors DG owner (verify they receive " Power Pointers" from GM).
9. Review site practices to limit DG cold fast starts.
10. Reliabiity records and calculations: Check logs, procedur :s, and calculations versus Reg. Guide 1.108 criteria.
11. Easure that pertinent studies on DG performa: .a have been reviewed and recommendations implemented as appropnate (e.g , N t.!F k 7/CR-0660 and NSAC-79).
12. Torquing: Ensure plan bas - ' equate sper;" :.tivus for all torquing. Ensure it is documented and done with c+ abrated equ:pment. Observe re-torquing ifin progress.

Snute; [22) J.C. Higgins and M. Subudhi, "A Review of Emergency Diesch ('.enerator Performance at Nuclear Power Plants " NUREG/CR-4440, Brookhaven National Laboratory, November 1985. B-3 Y}}