ML20203H534

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Insp Rept 50-482/97-201 on 971103-980109.No Violations Noted.Major Areas Inspected:Capability of RHR & CCW Sys to Perform Safety Functions Required by Design Bases & Adherence of Sys to Design & Licensing Bases
ML20203H534
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 02/23/1998
From:
NRC (Affiliation Not Assigned)
To:
Shared Package
ML20203H499 List:
References
50-482-97-201, NUDOCS 9803030253
Download: ML20203H534 (43)


See also: IR 05000482/1997201

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OFFICE OF NUCLEAR REACTOR REGULATION

D"cket No.: 50-482

License No.: NPF-42

Report No.: 50 482/97 201

Licensee: Wolf Creek Nuclesr Operating Corporation

Facility: Wolf Creek Generating Station

Location: 1550 Oxen Lane, NE

Burlington, Kansas

Dates: November 3,1997, through January 9,1998

Inspectors: Roy K. Mathew, Team Leader, PECB, NRR

E. Kleeh, PECB, NRR l

A. Bizarra, Contractor * l

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K. Neubauer, Contractor *

A. Rahman, Contractor *

M. Sanwarwalla, Contractor *

R. Sheldon, Contractor *

(' Contractors from Sargent & Lundy)

Approved by: Donald P. Norkin, Section Chief

Specialinspection Section

Events Assessment, Generic Communications,

and Specias inspection Branch

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

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9803030253 980223

PDR ADOCK 05003481

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TABLE OF CONTENTS

l EXECUTIVE S U MMAR Y , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i

Ill . E n g inee rin g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

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l E1.0 CONDUCT OF ENGINEERING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

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E1.1 Inspection Scope and Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

E1.2 Residual Heat Removal (RHR) System ............. ................. 1

E1.2.1 Mechanical Design Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

E1.2.2 Electrical Desigt. Review . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

E1.2.3 Instrumemation and Controls Design Review . . . . . . . . ......... 16

E1.2.4 System lnterface s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

E 1.2.5 System Walkdown . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . 23

E1.3 Component Cooling Water (CCW) System . . . . . . . . . ................... 25

E1.3.1 Mechanical De sign Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5

E1.3.2 Electrical Design Review . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . 30

E1.3.3 Instrumentation and Controls Design Review . . . . . . . . . . . . . . . . . . . 30

, E 1.3.4 System lnterfaces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... 32

l E1.3.5 System Walkdown . . . . . . . . . . . . . .......... ..... .... . . . 32

E1.4 Updated Final Safety Analysis Report (USAR) and Other Document Reviews . . 32

XI Exit Meeting summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35

AP PE NDIX A - OP EN ITEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......A1

APPENDIX B . EXIT MEETING ATTENDEES . . . . . . . . ...........................B1

APPENDIX C - LIST OF ACRONYMS . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . ...........C1

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ZXECUTlW: SUMMARY

A design inspection at Wolf Creek Generating Station was pcrformed by the Events

Assessments, Generic Communications, and Spec!siInspection Branch of the Office of Nuclear

Reactor Reguistion (NRR) during the period November 3,1997, througn January 9,1998. This

inspection included onsite inspections during November 1" " '%cember 15, December 1519,

1997, and January 5 9,1998. The inspection team consisc 2 deem leader from NRR, one

inspector from NRR, and five contractor engineers from Sargent a Lundy Corporation (S&L).

The purpose of the inspection was to evaluste the capability of the salected systems to perform

safety functions required by their design bases, the adherence of tri; systems to their design and

licensing bases, and the consistency of the as-built cor' figuration with the updated safety analysis

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report (USAR). The team selected the residual heat removal (RHR) s*estem, thu component

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cooling water (CCW) system, and their support interface systems for this inspection because of

the importance of these systems in mitigating design basis accidents at Wolf Cr3ek. The team

followed the engineering design and configuration control section of Inspection Procedure (IP)

93801 for this inspection. For the selected systems, the team reviewed the USAR, system

descriptions, calculations, drawings, modification packages, surveillance procedures, snd other

documents.

Except as noted below, the team detcimined that the selected systems were capable of

performing their design basis safeiy functions and that design and licensing bases were

adequately adhered to. The licensee implemented appropriate measures to resolve the

immediate concems identified by the team, and no immediate operability concems exist. For

other issues, the licensee initiated appropriate reviews and evaluations using the corrective

action process or took corrective actions such as revising design documents and changing

procedures.

Tim team identified the following weaknesses in 10 CFR 50.59 evaluations and design changes:

(1) The design change and safety evaluation for the replacement of Class 1E batteries with

AT&T round cell batteries did not address the effect on Technical Specifications (TS). Currently,

TS Gee 'ans 4.8.2.1.e and f regarding battery capacity replacement criteria and battery

degradation criteria appear to be nonconservative because the batteries were sized without

aging factors and the batte.y performance characteristic was changed. The battery design

capacity margin was less than that stated in the staff's safety evaluation report (NUR2G 0881)

and USAR. The NRR staff will review the design change to determine the adequacy of the

existing TS. (2) The design change and safety evaluation for lowering the CCW temperatur6 did

not address the effects of low temperature on the spent fuel pool reactivity and on diesel

generatorloading. (3) A reactor coolant system (RCS) draindown procedure for installing the

nitrogen bottles during plant refueling outages did not have a safety evaluation to address

seismic restraint requirements to preclude potential missiles.

10 CFR 60.59 evaluations generally lacked adequate documented justification. The licensee had

recently revised its procedure to emphasize that sufficient details and basis should be supplied in

10 CFR 50.59 evaluations.

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Surveillance testing of batteries was jeficient because the battery caoacity and service test

procedures did not provide appropria e acceptance criteria. The recent capacity test for the

NK11 battery did not identiff and prog orty evaluate the impact on battery capcity when the

battery load current was not maintained constant during the test.

The team identified calculations with errors or inappropriate or noncons ervative assumptions, in

some cases analysis did not exist to support the design bases. For example: the refueling water

storage tank instrument (RWST) loop uncertainty calculations did not consider density variations

due to temperature and boron concentrations which affected the alarm setpoints, the swapover

setpoin's, and the level indications; the Westinghouse cooldown analysis did not assume correct

essential service water flow; a de voltage drop calculation did not identify the worst case

minimum battery voltage; nonconservative downstream pressures were assumed for some CCW

motor operated valves in valve closure calculations resulting in incorrect des!gn differential

pressures for valves; the calculation to estimate the maximum control circuit wire lengths for

motor control center starter control circuits did not model the auxiliary loads correctly; and no

an.)!yses existed to show that the 120 Vac feeders and control circuits are protected adequatdy

during a fault, and that 120 Vac safety-related loads have adequate voltages.

The team noted that several calculations for each system have similar purposes and/or similar

results. These calculations were not contradictory, but the practice tends to confuse

identification of the design basis,

in two design change packages, the electrical calculation was not reased in accordance with the

licensee's procedure when de load changes occurred. There was an inconsistency between the

electrical load growth control procedure and the engineering screening form regarding the

threshold for analyzing electricalload changes and discrepancies with some de load changes in

the licensee's load data base

Other discrepancies identified included the following: the licensee did not correct a previously

identified design basis requirement discrepancy in operating procedures for the CCW system; the

electrical distribution system was not modeled down to the 208/120 voit level or compared to

field voltage measurements as specified in Branch Technical Position P.B.1; and no

documentation existed to show the RHR pump suction pressure gauges were seismically

qualified to maintain the RCS pressure integrity.

The as-built configurations of the systems ure generally consistent with the USAR. In general,

the availability of the design bases documentation was good, as wts the material condition of the

areas observed by the team. However, the team iderbfied r; number of discrepancies in the

USAR, system descriptions, and other plant documents.

The team's findings indicated an ongoing need to emphasize design and conf;guration-control

related issues in maintaining the design and licensing bcses for Wolf Creek. Wolf Creek had

established a design basis / licensing basis (DB/LB) review program to address these types of

concems. This new DB/LB program had not yet produced widespread or consistent results at

the time of the inspection.

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Ill. Engineering

E1.0 CONDUCT OF ENGINEERING

E1.1 Inspection Scope and Methodoloav

The purpose of the inspection was to evaluate the capability of the selected systerns to perform

safety functions requirrsd by their design bases, the adherence of the systems to their design ani

licensing base *, and the consistency of the as built configuration with the updated safety analysis

report (USAR). The systemt selected for inspection were the residual heat removal (RHR)

system, the component cooling water (CCW) system, and their support interface systems.

These systems were selected on the basiJ of theirimportance in mitigating design basis

accidents at Wolf Creek.

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The inspection was performed in accordance with NRC Inspection Procedure (IP) 93801, " Safety

System FunctionalInspection." The engineering design and configuration control section of the

procedure was the primary focus of the inspection.

The open items resulting from this inspection are included in Appendix A. The acronyms used in

this report are listed in Appendix C.

E1.2 Residual Heat Removal fRHR) System

E1.2.1 Mechanical Design Review

E1.2.1.1 Inspection Scope

The team evaluated the capability of the RHR system to achieve and maintain cold shutdown

and to mitigate the consequences of a smali or large break loss of coolant accident (LOCA).

The team also reviewed portions of the high head and intermediate head safety injection system

that interfaces with the RHR system, the refueling water storage tank (RWST), and portions of

the containment spray system.

As part of this effort the team reviewed the plant design drawing 3, calculations, accident

analyses, the containment flooding analysis, design change packages, the USAR, the system

design description, technical specifications, operating procedures, maintenance and surveillance

tests, information notir'es, generic letters, environmental qualification (EQ) files, and engineering

evaluatinns associated with the system.

E1.2.1.2 Observations and Findings

a. RHR System Flow and Decay Heat Removal Requirements

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The team evaluated the capabilitiof the RHR system to remove the sensible and decay heat

generated in the reactor, arid to achieve and maintain the plant in cold shutdown. The time

required to bring the plant to cold shutdown so that neither the cooldown rate nor the cooldown

time is exceeded is dependent on the RHR flow and the temperature of the component cooling

water (CCW), which again is dependent on the essential service vater (ES N) system that cools

the CCW. In the current Westinghouse analysis, FSDA C 365, *NSSS Upiating Analysis,"

Revision 1, which was performed for power uprate, it was determined that the plant can be

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brought to cold shutdown in about 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> with both RHR trains in operation. The

.iestinghouse calculation used reactor coolant system (RCS), RHR, and CCW flow rates and

temperatur6s that were consistent with design bases. However, this calculation used ESW flow

rates (13500 gpm) to the CCW heat exchangers that were higher than used in licensee

calculation EG-06 W, 'Deterrt.me Flow and Heat Land Requirement," Revision W3 (8800 gpm).

Calculation EG-06-W included flow effects associated with plugging 46 tube r., irs to provide

margin for the future should plugging of a significant number of CCW heat exchanger tubes

beco.,ie necessary. Westinghouse did not use the tube plugging assumption in their analysis.

Currently only two tubes in one heat exchanger are plugged. A preliminary evaluation of g

cooldown time with appropriate assumptions indicated that cooldown will be consistent with

current commitments to achieve ecoldown in 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> as stated in USAR Section 5.4.2.1.7 at a

cooldown rate not exceeding 100 F as specified in the TS. The licensee issued PIR 97-4145 to

resolve the design issue. The team determined that the licensee's design control measures did

not meet the requirements specified in Criterion lll of Appendix B to 10 CFR Part 50 regarding

verifying or checking the adequacy of the design. (Unresolved Item 50-482/97 20101)

The team reviewed licensee calculations SA 90-067, * Calculation of the Decay Heat Load to the

RHR Heat Exchanger Following Normal Shutdown," Revision 0, and SA 92 074, * Decay Heat

Load to the RHR Heat Exchanger Following Normal Shutdown for Uprated Power to 3565 MWt,"

for decay heat removal. Whereas the above Westinghouse analysis determined the cooldown

for the most limiting design case for CCW temperature, the licensee's analysis evaluatad the

cooldown time fer various CCW temperatures. For lower CCW temperatures, cooldown can be

achieved faster, but to maintain the cooldown rate within TS lin,its licensee's procedures SYS

EJ-120,"Startup of a Residual Heat Removal Train," and SYS EJ 121, "Startup of RHR Train in

Cooldown Mode,' are implemented. The team determined that the above procedures are

adequate to maintain the cooldown rate within the Tb li".1it.

The team reviewod licensee calculations SA 89-009, "The Minimum RHR Flow Requirement for

Decay Heat Removal During Mid Loop Operation," Revision 0; AN 93-009, * Minimum RHR Flow

Requirement for Decay Heat Removal During Mid Loop Operation to Support the Power Rerate

Program,' Revision 0; and RE EJ 005, Revision 1, which were performed to dem.nstrate that

low water levol in the reactor would not limit RHR flow for decay heat removal and allow

vortexing and air binding of the RHR pumps. The team also evaluated the licensee's responses

for information notices IN 90-06, " Potential for Loss of Shutdown Cooling while at Low Reactor

Coolant Levels due to Loss of Power to the RHR Flow Control Valve;' IN 87 23,' Loss of Decay

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Heat Removal During Low Reactor Coolant Level Operation (LosJ of Residual Heat Removal)

(Diablo Canyon Event);' and IN 86-101," Loss of Decay Heat Removal due to Loss of Fluid

Levels in Reactor Coolant System," and operating procedures GEN-007, *RCS Drain Down," and

GEN-008, * Reduced Inventory Operations." Based on a review of the above documents the

team concluded that as long as mid-loop opuation was delayed for a minimum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after ,

cold shutdown was initiated, vortexing and air bindmg of the RHR pumps would not occur.

b. RHR System Flow Requirements and Flow Rates for Emergency Core Cooling

To evaluate RHR system capability to provide the lim!Pa flow specified in Westinghouse report

WCAP 13447, *3579 MWt NSSS Rerating Engineering Report," Volume 1, dated October 1992,

the team reviewed the licensee's calculations for RHR system hydraulic resistance, piping

isometric drawings, and pump surveillance test procedures. For the small break LOCA analysis,

no credit is tako) for flow from the RHR system. The team's review for the small break LOCA

was, therefore, essentially limited to verifying the " piggyback' mode of operation of the RHR. __

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system to provide flow to the suction of the intermediate and high head pumps when their suction

is switched from the RWST to the containment sump.

Iniection Phase

The calcu'ation that determines the RHR flow rates during the injection phast of emergency con

cooling s3 stem (ECCS) operation for the large and small break LOCA is documented in licensee

calculation SA 91016, *ECCS Design Basis Flow Rates Resnalysis in Support of the WCGS Re-

rating Project,' Revision O. The calculated RHR flow rates were provided as input to

Westinghouse to use in the LOCA analysis which is summarized in the above Westinghouse

report. The required and calculated total RHR flow injecting to the three intact loops at an RCS

pressure of 0 psig is about 2834 gpm. Calculation SA 91-016 very conservati',ely assumed a

10 percent pump degradation and open pur,ip miniflow lines. The above RHR pump flow is

assured per TS Surveillance Requirement 4.5.2.1, which verifies the total pump flow is equa! to or

greater than 3800 gpm and equal to or less than 5500 gpm, which is the pump runout flow.

QgjiLga Recirculation Phase

No specific evaluation had been performed by the licensee to determine the minimum ECCS flow

i requirement during cold leg recirculation. Guidance provided in Westinghouse document NSAL.95-001, * Minimum Cold Leg Recirculation Flow," dated January 12,1995, roquires that the ECCS

l flow during cold leg recirculation should be at least 1.2 times decay heat boiloff when cold leg

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recirculation is initiated. The team determined that this amounted to about 608 gpm for a cold

l leg switchover time of 0.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The results of licensee calculations SA 92 056, "CCP ar.d Si

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Pump Runout Flowrate in Recirculation Phase," Revision 0, and AN-95-021, * Determine ECCS

Flow Rates in Recire. Phase," Revision 0, which were performed to evaluate the runout flow of

ECCS pumps under " piggyback" operation, provide reasonable assurance that sufficient cooling

flow dunng cold lec recirculation is available from one operating ECCS train (the other ECCS

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train assumed inoperable).

Hot Lea Recirculation Phau

The team reviewed the minimum required cooling flow for a large break LOCA during hot leg

recirculation that was spec.fied in Westinghouse document NSAL 92-010, dated January 9,

1993. The licensee's flow calculations discussed earlier indicated that the RHR and ECCS

pumps had the capability to provide the required flow, assuming the most limiting single failure

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specified in the Westinghouse document.

The team reviewed industry technical information program (ITIP) No. 02178, which was in

response to a 10 CFR Part 21 notification from Westinghouse regarding the potential for boron

precipitation in the reactor core in the long-term cooling mode following a postulated LOCA. This

evaluation confirmed that sufficient flow is available from the intermediate and high head pumps

to satisfy the minimum flow requirement during the hot leg recirculation, assuming a single failure

of the RHR hot leg header isolation valve and a loss of one diesel generator so that only one

train of ECCS is available. The team also reviewed Westinghouse calculation SEC-TSA 3958-

CO, " Wolf Creek (SAP) Hot Leg Switchover Time for Power Uprating," Revision 0, which

deteimines how long after a LOCA a switchover to hot leg recirculation should be initiated to

prevent boron precipitation in the reactor vessel. This calculation determined that the minimum

switchover time to hot leg recirculation is 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Emergency Operating Procedure EMG

ES 13 " Transfer to Hot Leg Racirculation," was also reviewed to verify the switchover time to hot -

leg recirculation. No unacceptable conditions were identified.

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Pumo Performance

The team review 3d the pump quarterfy surveillance test (procedure STS EJ 100A) and trending

data on pump degradation to verify the pump's dility to provide the required flow. The

surveillance test and trending data over the last 4 years indicated no pump degradation. Also, a

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review of procedure STN EJ 100A, 'RHR Pump "A" Reference Curve Detemilnation," showed '

that the reference curve determined in 1993 showed no noticeable degradation in pump

performance when compared to the original pump performance curve.

c. RHR Pump Net Positive Suction Head (NPSH)

The team reviewed calculation ECCS 35, *RWST to RHR pump *A* Suction Mode A." and TS

Section 3/4.5.5 to verify the adequacy of the NPSH for the RHR pumps when taking suction from

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the RWST. The NPSH was determined based on the levelin the RWST at the time of

switchover to the containme it sump. The team determined that the available NPSH was much

greater than the NPSH required. The team also concluded that since the head of water above

the suction piping at the time of switchover to the containment sump was about 14 feet, the

probability of any air ingestion by the RHR pumps was very low.

The team's evaluation of the available NPSH for the RHR pumps when taking suction from the

containment sump was based on the guidance provided in Regulatory Guide (RG) 1.1, * Net

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Positive Suction Head foi Emergency Core Cooling and Centainment Heat Removal System

Pumps," and NRC Generic Letter (GL) 97 04, * Assurance of Sufficient Net Positive Suction Heh

for Emergency Core Cooling and Containnient Heat Removal Pumps." The team reviewed

calculations and analyses EJ 30, *RHR Pumps A&B NPSH," Revision 1; EJ 29, *RHR Flow

Orifice Sizing," Revision 0; FL 18, *LOCA & MSLB Ctmt. Flood Levels,' Revision 1; USAR Table

6.2.2 6; and Containment Recirculation Sump Hydraulic Performance analysis. The review

determined that the NPSH available for the RHR pumps was greater than the NPSH required.

The RHR pump suction lines from the containment sump have vortex breakers installed in the

pipe inlets. The vortex suppressor test results in the ' Sump Hydraulic Performance Report,"

showed that the suppressor was effective in decreasing both vortex activity and swirt. The

licensee's tests indicated no vortex activity for RHR pump operation at the water levels at which

ECCS pump switchover and containment spray pump switchover occur.

In determining the original setpoints for the RWST level instruments, no consideration was made

for the temperature and boron density (see section E1.2.3.2.1). The team noted that any

corrections to the setpoints to account for the temperature and boron density could alter the

volume of water transferred to the containment sump and affect the containment flood level and

the available NPSH for the RHR pumps. Based on preliminary input provided to the team by the

licensee regarding setrat changes, the team determined that the available NPSH would

decrease but not signift mntly enough to affect pump performance.

d. ECC8 Leakage Testing

The team reviewed the provisions made in the system, in compliance with RG 1.i39, * Guidance

for Residual Heat Removal to Achieve ana Maintain Cold Shutdown,' to allow for normalleakage

during long term cooldown without affecting plant safety or violating the radiation limits

established by 10 CFR Part 100.

Contrary to the statement in the USAR Table 5.4A-1, the licensee had not established in the_TS -_

an acceptable leakage limit at which the RHR train is to be declared inoperable. However, as per

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USAR Section 18.3.4, and TS Section 6.8.4, the licensee has established a program to reduce l

leakage from systems outside containment that contain highly radioactive fluids. The program l

requires monitoring and correcting leakages identified during survelliance tests or routine plant

walkdowns. This program meets the requirement of item Ill.D.1.1 of NUREG 0737. The

licensee stated that the existing program met the intent of RG 1.139 and the USAR table will be

revised appropriately. Review of the licensee analysis for leakage rsquirements, Calculation AN-

97 049," Radiological Consequence of a LOCA and Available NPSH Determination Due to the

Leakage from the RHR and CS Encapsulation Tanks," Revision 0. Indicated tnat Wolf Creek can

tolerate a total ECCS leakage of 2 gpm and still be within the limit 1 established by 10 CFR

Part 100, and that long term plant cooling would not be affected. Tho team noted that the staff's

safety evaluation report (NUREG-0881) Section 15.4.5.1 states the maximum operating leakage ,

limit to be 1 gpm. The team reviewed licensee procedures AP 25C-001, "WCGS Leak Reduction l

of Primary Coolant Sources Outside of Containment;" STN BG 001,' Leakage Inspection 1

Program of CVCS;" STN EJ 001, " Leakage inspection Program of RHR;" and STN EM 001,

" Leakage Inspection Program of Sl," to determine if any of these procedures addressed a

requirement for establishing leakage limits or leakage acceptance criteria to determine what is an

acceptable leakage rate.

The team determined that the ebove procedures did not establish acceptance criteria to account

for individual system leakage or cumulative ECCS leakage. The team reviewed leakage data

collected for ths ECCS between the eighth and the ninth refueling outages and the plant

manager report for radioactivity leaks from the last refueling outage (the ninth). The team

concluded that except for a few sightings of boron precipitation during plant walkdowns, no leaks

existed in the ECCS and controls existed for detection and elimination of leakage. The team

noted that the licensee also identified this issue. PIRs 97-3563,97 3138 and 97 3738 were

written to address the above issues. (lnspection Followup Item 50-482/97-20102)

e. Potential for Radioactive Leakage from the RWST to Atrnosphere

The team reviewed the potential for radioacuve leakage from the RWST to the atmosphere

, during the recirculation phase of a LOCA, when radioactive water from the containment sump

could enter the RWST due to leakage through the isolation valves. The team reviewed ITIP

1737, which was the licensee's evaluation in response to information notice (IN) 9156, " Potential

Radioactive Leakage to Tank Vented to Atmosphere." This evaluation determined that to remain

within 10 CFR Part 100 limits, Wolf Creek could tolerate a leakage of 1.8 gpm flowing to the top

of the RWST and 9 gpm flowing to the bottom of the RWST. The team reviewed procedure

STS BN-206, * Borated Refueling Water Storage System inservice Valve Test," Revcon 7, for

back-leakage testing of the RWST isolation valves. This procedure establisned the acceptance

enteria for leakage through valve EJ 8717 at 10 gpm at 600 psig and 4 gpm at 20 psig for each

of the eight FCCS pump suction check valves that isolate the RWST. This leakage would be

lower for valve EJ 8717 at RHR system pressure but could be higher at the higher containment

pressure for the pump suction check valves. Tne team's concem was that with the current

acceptance criteria, the leakage of contaminated radioactive fluid into the RWST and to the

atmosphere at high containment pressure may be greater than the limits established in

ITIP 1737. The team noteo that the existing analysis considered leakage into the RWST for

30 days, which may be conservative since the containment pressure drops below 6 psig within

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and the actual dose may, therefore, be lower than evaluated in ITIP 1737. This was

confirmed in a preliminary analysis dono by the licensee. In addition, a review of valve tests data

showed no significant leakage of any consequence. Therefore, there is no immediate safety

concem. The team noted that the licensee identified this issue and was taking appropriate

corrective actions via PIR 97 4124.

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f. Consequence of Floeding of Containment Sump Isolation Valves

For Wolf Creek, containment isolation for the RHR suction line from the containment sump is

provded by only one containment isolation valve. To contain any radioactive sump fluid leaking

from these valves, the valves are encapsulated in a tank. The team revl3wed the design of the

valve ar.J its encapsulation to determine if ieakage due to any failere or normal wear of the valve

could prevent the valve from performing its safety function.

The encapsulation, or enclosure, for containment isolation valves HV 8811 A & B has no

automatic drain or any level measuring device in the enclosure to indicate any leakage in the

cont.ol room for operator action. The possibility exists for these valves to be flooded due to

leakage from normal wear of valve packings or gaskets. Because of this post.ibility, the valves

and piping have been knalyzed for mechanicalintegrity. Also, a review of electrical schematics

E 13EJ05A and E-13BNO3 showed that flooding of the valve motor operator and its circuit ,vould

not affect the valve safety function to remain open. However, flooding of the limit switch in the

operator may affect the containment isolation indication for Interlocked valve HV-8701 A. The

licensee stated that operators would depend on the Indication provided by the plant computer for

valve HV 8701 A in case of loss of containment isolation indication. However, the team noted

that none of the existing design documents or plant procedures iden'ify the plant computer as an

alternate source for containment isolation indication. The licensee issued PIR 98 0083 to

address this issue.

g. Piping Design Pressure and Temperature

The team reviewed the RHR system Piping and Instrumentation Diagram (P&lD) M 12EJ01,

Revision 19, System Description M 10EJ(Q), Revision S, Piping Class Summary MS-01,

Revision 40, and piping isometric drawings to verify the piping r'. ign pressure and temperature

classification for the suction lines from the RCS, RWST, and containment sump; and discharge,

minimum recirculation, test and tie lines to the intermediate and high head safety injection

systems. The team determined that the design pressure and temperature ratings of the lines

were acceptable. However, minor discrepancies in documentation of norn al service ratings

were noted The licensee issued PIR 98-0002 tc address this issue,

h. RHR Pump Operation in Minimum Recimulation Mode

For a small break LOCA the RHR pumps would start and operate on minimum recirculation flow

for an extended duration, therefore the toam verified the duration for which the pumps could run

without any cooling flow to the RHR heat exchanger without being damaged.

From a review of procedures EMG E-0," Reactor Trip or Safety 4ction;" E 1," Loss of Reactor

or Secondary Coolant;" and ES-11, Post LOCA Cooldown and Oepressurization," the team

determined that the RHR ptmps may be allowed to run for a durailon of 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> before an

operator decision is made to either shut the pump down or initiate cooling flow to the RHR heat

exchanger. This period of 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> was beyond the pump manufacturer's ( Pacific Pumps) time

limit of 30 minutes. The team reviewed licensee calculation EJ-M-018, 'RHR Pump Recire.

'

Operation vs. Time of initiation of CCW flow to RriR Heat Exchanger," Revision 0, which justified

this duration of 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. This calculation essentially determined the time it would take for the

temperature of the circulating water to reach 212 *F, a temperature at which the pump could start

vortexing. The team noted that this calculation assumed an incorrect initial water temperature of

90 'F. As per USAR Table 3.11b, the maximum temperature in the RHR pump rooms is 104 'F.

The basis for the initial wster temperature used in the calculation was not readily apparent and

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was not consistent with the USAR. Even with a higher initial water temperature, i.e.,104 *F, the

licensee determined that sufficient margins existed in the calculation to demonstrate that the

pump operation for 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> without any cooling water would not damage the pump. The

licensee issued PIR 97-4150 to address the issue. The team determined that the licensee's .

design control measures did not ensure that the design was verified or checked adequately in

accordance with Criterion ill of Appendix B to 10 CFR Part 50. (Unresobed item 50-482/97-

201 03)

1. Environmental Qualification (EQ) of RHR Pump and Motor.

The team reviewed the environmental qualification of the RHR pump and motor to determine

whether the as built configuration matched the qualification documents. The team determined

that pumps and motors were qualified for their application. However, the following weaknesses

were identified in EQ documentation:

.

The EQ files for the RHP. mechanic.al pump and electrical motor did not provide a

reference for the normal and accident environmental parameters used to justify

qualification. The normal and accident radiation parameter (1.28x10') used for

qualification was lower than the total integrated radiation dose of 1.82x10' provided in

USAR Table 3.11(b) 2. The pump and motors were qualified to a total radiat:en dose of

5x10' rads, and hence, there was no qualification concem. The licensee issued PIR 98-

0037 to address this issue.

.

The current revision of the EQ file for the RHR pumps indicated that the pumps used lube

I

oil. It was deterW that the pumps did not use any lubrication oil as the pumps have

carbon bearings suoricated by the pumped water. The licensee issued PIR 98 0038 to

address this issue

J. Freeze Protection for the RWST and RHR Suction Lines

The team reviewed the design documents and operating procedure for protecting the RWST tank

and lines exposed to the outside environment from freezing at very low temperatures during

winter (freezing could affect emergency core cooling during an accident). The team determined

that the existing design and operating procedure precluded any freezing of the tank or lines.

However, weakness in a 10 CFR 50.59 evaluation documentation was ident;fied, as discussed

below.

Procedure STN-GP-001, " Winterization Procedure,' was revised to keep the bypass salve across

the automatic control valve open to allow continuous steam flow to the RWST heaters and so

prevent any freezing in the piping and assure that the RWST temperature is maintained above a

nominal 50 'F when temperatures fall below 35*F. The team reviewed the 10 CFR 50.59

evaluation done by the licensee for implementing this procedure change, CCP07251. This

evaluation, 59 97-0008, stated in Section 1 a that "the temperature control salve set point is a

nominal 50'F'; and the proposed USAR change stated that," the RWST is maintained above a

nominal 50 *F set point.' The team verified from the licenses setpoint data base that the cut-in

set point is 65'F and the cut-out is 80 F for the temperature switches used to control the RWST

temperature. The reference to the valve setpoint of 50 'F in 59 97 0008 was incorrect. The

team determined that this was a documentation error and it did not adect the temperature

settings. T he licensee issued PIR 97 3830 to address this issue. Subsequent to the inspection

the licensee revised the document to reflect the setpoint data.

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E1.2.1.3 Conclusion

1

4 The team conciuded that the mechanical design for the RHR syster was adequate to provide l

the required flow knd heat transfer capability to bring the plant to cold shutdown from hot I

shutdown conditions within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> and maintch the plant in cold shutdown, and was also

adequate to provide the required flow for decay heat removal during mid loop operation. The

system was adequately designed to provide the required flow to the RCS to mitigate the i

consequencec of a small or large break LOCA, considering a 10 percent reduction in the pump i

flow capacity. The system provided the required NPSH for the RHR pumps when taking suction

from either the RWST or the conta!nment sump. The team had concems regarding nonexistent

or nonconservative acceptance criteria for normal system or radioactive leakage from ECCS and

fur leakage of radioactive sump fluid to the RWST. The team noted some nonconservative

assumptions used in calculations, one of which was a Westinghouse cooldown analysis.

Discrepancies were observed in a piping class summary document, EQ files, and a 10 CFR

50.59 evaluation.

E1.2.2 Electrical Design Review

)

E1.2.2.1 Inspection Scope

For t:,e electrical design review, the team focused on the essential power supplies to the RHR,

COW, and support-interface systems. The following power supplies were chosen for review:

emergency diesel generators, the 4160 Vac system, the 480 Vac sys :m, the 120 Vac system,

and the 125 Vdc systems. The following attributes for the above areas of review were assessed

by the team: equipment sizing; regulatory and standard compliance; electrical separation; voltage

drops and available voltages; protective device sizing, coordination and setpoints; controls and

interlocks; operating procedures; plant modifications; surveillance tests; and design and

configuration control,

The team reviewed USAR Section 8.0, TS 3/4.8, system descriptions, electrical requirements,

design change packages, surveillance test requirements, and other miscellaneous electrical

cucuments related to the design basis.

E1.2.2.2 Observations and Findings

E1.2.2.2.1 AC System Review -

a. Emergency Diesel Generator (EDG)

The team reviewed the emergency diesel generator (EDG) system description, loading

calculation, elementary diagrams, protective relay setpoints, and TS test requirements. A review

of Drawing E-11005, "I kt of Loads Gupplied by Emergency Diesel Generator," Revision 19,

determined that the E. .. loads under various postulated accident conditions remained within the

EDG continuous rating and had adequate capacity margin. The team evaluated power demands

for major pumps such as ESW, RHR, and CCW pumps, to verify the loads on the above drawing.

The team determined that the licensee property estimated the loads. A review of recent EDG

surveillance test report STS KJ-001 A showed that electrical loads, including ECCS loads, were

sequenced onto the EDG within the required times and that the drop and recovery in output

voltage and frequency were acceptable. The licensee's test met the TS requirements and

complied with Regulatory Guide 1.9, Revision 3.

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b. 4160 Vac System

The team reviewed the system description, drawings, voltage and short circuit analysis, relay

setpoints, and coordination calculations, focusing on the equipment performance under worst-

case voltage conditions and the equipment short circuit duty. The calcult,tions indicated that the

undervoltage setpoints were adequats for proper operation of the electricalloads and for

shedding loads in the event of a loss of offsite p]wer. The breakers' interrupting duty was found

to be adequate. However, the team noted the following discrepancies.

Based on calcutation XX E 006,AC System Arialysis," Revision 3, some equipment fed from

Class IE bus NG01 could experience vo!tage levels slightly higher than their maximum ratings

with tho switchyard M '05% of rated voltage. The calculation stated that this was considered a

ra.e occurrence sin e switchyard voltage was not normally higher than 104 percent; however,

switchyard data fc - i996-1997 showed severalinstances where switchyard voltage had reacheo

105 percent for short durations. The licensee indicated that an evaluation wou!d be dona to

determine the impact of this overvoltage condition (down to the 120 voit level) and calculation

XX E 006 would be revised. The licensee's preliminary evaluation indicated no operability

concems. The licensee issued PIR 98-0035 to address this issue.

A review of calculations XX E 007," Verification of Voltage Analysis at Wolf Creek," Revision 0,

and XX E-006, Revision 1, indicated that the short circuit contribution from the EDG was

corsidered only for one plant scenaric. An cnalysis is required to determine the worst case fault

current, and the consequence, when the EDG is connected to the bus in other scenarios such as

when the safety buses are fed from one startup transformer. The licensee issued PIR 98-0071 to

address this issue.

c. 480 Vac System

The team reviewed the system c; ascription, drawings, voltage drop and short circuit calculations

for the 480 Vac unit substations and motor control centers (MCr.s) focusing on the equipment's

short circuit duty and the equipmint's performance under minimum available voltage conditions.

The review determined that adequate voltage would be provided at the terminals of all Class 1E

equipment. The switchgear breakers' interrupting duty for both the instantaneous and short time

delay trips was found to be adequate. The interrupting ratings of the 480 Vac molded case -

circuit breakers at the MCC levelwere also found to be adequate, except that the availabte short

circuit current at the buses of MCCs NG01 A and B for 460 Vac EF3 type breakers exceeded their

interrupting rating. The licensee indicated that a ciesign change package, PMR 03907, was

replacing these breakers. A number of these breakers had already been replaced by breakers

with a higher interrupting rating and the rerraining EF3 type breakers will be replaced during the

current fuel cycle.

d.170 Vac Systems

The team reviewed the drawings and voltage drop calculation EB 10, " Determine Voltage Drop in

MCC Control Circuits," Revision 3, for control circuits driven by control power transformers

(CPTs) located at i .CCs. The above calculation, which determined the maximum permissible

circuit wire lengths to allow prcper pickup of motor starters, was found to have used assumptions

which were not conservative. For example, all auxiliary loads were erroneously assumed to be at

the location of the CPT. The team picked one clicult (valve BBPV8702A circuit) which had an

actualinstalled circuit wire length of 4228 ft for review. The team's review indicated that

sufficient voltage would not hav6 been available for the circuit if the wire length of 5374 ft allowed

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In the calculation was used. The licensee's review confirmed that allowable wire length for that

circuit to operate proper 1y should be 5189 ft rather than 5374 ft. Based on the sample review, l

the team determined that if the maximum permissible circuit wire lengths in the calculation for

other motor starter circuits (with different starters and loading configuration) weru used, the

potential existed that the starter coils, in some cases, might not pickup. The team raised this

concem. In response the licensee performed an operability evaluation that was supported by an

analysis. The licensee's analysis and operability evaluation concluded that the actual control I

circuit lengths installed were lower than the allowables and were acceptable and would perform

their safety related function. The team noted the licensee's calculation for allowable maximum

wire length was still nonconservative and needed to be revised. The licensee issued PlR 97-

4159 to revise the calculation. The team determined that the licensee's design control measures

did not meet the requirements specified in Criterion lli of Appendix B to 10 CFR Part 50 regarding

verifying or checking the adequacy of the design. (Unresolved item 50-482/97 20104)

No analysis existed to show that 120 Vac feeders and control circuits are proter"ed adequately

during a fault. Also, no voltage drop analysis existed to show that all 120 volt a ty related

loads had adequate voltage at their terminals. The licensee's preliminary eva. .tlon of some

sample circuits indicated that there were no operability concems. In addition, the team noted

that the licensee's electrical system modeling did not include Class 1E buses down to 120/208

volt level. Power System Branch Technical Position PSB -1, positions 3 and 4 (licensing

confirmatory issue B.19), require electrical system modeling down to 120/208 volt buses,

venfying the modeling by actual tests and comparing the difference between measured and

modeled values with the equipment rating. The licensee was not able to dernonstrate or provide

documentation to show how it met the PSB-1 commitments for the 120 volt system. It was noted

that the licensee met the PSB 1 requirements for voltages up to 480 volts. The licensee issued

PIRs 97-4041 and 97 4032 to address the above issues. (Inspection Followup item 50-482/97-

201 05)

6. Evaluation of Plant Modifications

Three modification packages (DCP 07195,05588 and PMR KN 84-0044) were randomly picked

for evaluation, with particular emphasis on the 10 CFR 50.59 evalua' ion and procurement of

components. The 10 CFR 50.59 evaluation conclusions were adtouate, and the design changes

were consistent with the design basis. However, the team had a concem with DCP 05588,

which covered, in pari, procurement of an overexcitation relay for the EDG. This component was

installed as a safety-related item through the commercial grade dedication process conducted by

the supplier (Farwell & Hendrickt, Inc.). The analysis done as part of the dedication process

established the qualified shelf life of the relay to be 16 years. Monitoring of degradation is

requi red to determine the level of degradation and to establish frequency of replacement of the

relay or any of its parts to maintain its qualification in accordance with Certificate of

Conformance, #62152.1. Documentation of the methodologies used to meet these requirements

could not be provided by the licensee during the inspection period, nor could any documentation

of surveillance results. The licensee issued PG 98-0085 to address this issue. (Inspection

Followup item 50-482/97 201-06)

f. Cable Ampacity and Short Circuit Rating

The team's review of cable ampacity calculations F02, * Cable Sizing," Revision 0, and F03

" Cable Sizing," Revision 4, revealed that cables in covered trays were given a derating based on

96% of the ampacity of cables routed in open trays (4 percent derating factor in addition to the

derating factor used for cable tray fill). While this is a not in accordance with current industry

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design practice (18 27 percent derating), the team could not identify a licensing or regulatory

basis that would require a greaterlevel of derating. A review of a sample of as installed RHR -

system cable installations revealed that the cables were adequately sized even for the current

industry design derating value.

The team also reviewed calculation F09, " Determine the minimum cable sizing based on

maximum AC short circuit rating,' Revision 3. The review determined that the cables were sized

adequately to withstand the maximum short circuit current.

g. Protective Coordination

The relay setting tabulation, E 11023, * Relay Setting Table and Coordination Curve," Revision 4,

was reviewed, and setiings were verified for RHR, CCW, CS, and ESW pump motors. The relay

settings were consistent with values indicated in the coordination diagram. Coordination was

found to be adequato among the following: motor fullload amperes (FLA) at 75 percent and

100 percent, motor locked rotor amperes (LRA) at 75 percent and 100 percent, motor thermal

limit, feeder cable thermallimit, and relay characteristic curves. To ensure that changes in relay

settings had been adequately transferred to the coordination diagram, the team r0 viewed an

instantaneous trip setting that had been revised in the relay setting tabulation, it was verified that

the characteristic curve on thn coordination diagram had been appropriately revised to reflect the

changes in settings.

l

h. Electrical Penetration Protection

The team reviewed the protection of electrical penetration modules and conductors that were

shown in calculation A-6-W, " Thermal Capability of Electncal Penetration Assemblies (EPAs) vs

Dual Short Circuit Protection to Satisfy Reg. Guide 1.63,* Revision W2. Several circuits covering

sizes #12 AWG, #10 AWG, #8 AWG, #6 AWG, #4 AWG, #2 AWG,2/0 AWG,250MCM,

350MCM, and 500MCM cables were selected for review. The team determined that the

penetration and circuits were adequately protected with primary and secondary protective

hvices and the licensee had complied with RG. i.63,

i. Class IE 480 Vac Molded Case and 4160 Vac Switchgear Breakers - Test and Surveillance

The tearn reviewed the licensee's surveillance testing and maintenance procedures on selected

480 Vac penetration molded case circuit breakers and 4160 Vac switchgear breakers for the

RHR and CCW systems. The procedures reviewed were STS-MT 024 and MPE E 0090-02.

The procedures provided detailed instructions for testing and inspection of breakers in

accordance with the manufacturer's instructions. No discrepancies were identified. The team

noted that the testing program was previously reviewed by NRC.

J. Calculation Documentation Discrepancy

During the review of calculations for cable sizing the team noted that calculation F 03, " Cable

Sizing," Revision 4, referenced calcuWion F 08, " Electrical- Provide Essential Service Water

Pump Cable Sizing," Revision 0, which has been superseded. The licensee stated it would

review calculation F-03 for any potentialimpact via a PIR.

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E1.2.2.2.2 125 Vdc System Review

a. &ctrical Penetrations

The team reviewed the electrical prutection for the do circuits directed through penetrations, as

presented in calculation A 6-W," Thermal Capab;lity of Electrical Penetration Assemblies (EPAs)

vs Dual Short Circuit Protection to Satisfy Regulatory Guide (RG.) 1.63," Revision W2, and

determined that the licensee had complied with RG.1.63. Each de circuit routed throt.ch an

electrical penetration hrd proper overioad protection by suitably sized redundant fuses in its

positive and negative circuit sides. The calculation, however, did not contain a complete listing

of all safety-related de circuita routed through electrical penetrations, including related schematic

drawings, and all worst case time-current plots of relevant de overcurrent devices and associated

thermal capability curves of de electrical penetrations. The ' ansee initiated PIR 97 3910 to

evaluate and resolve this issue.

b. Sizing of Class 1E Batteries

The team evaluated the sizing of the new AT&T round cell batteries in accordance with

calculation NK E 002, * Class 1E Battery Sizing," Revision 3. The new batteries were installed in

March of 1996. The team was able to verify that the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> load profile (load values and running

times) for battery NK11 was correct except for the following errors: assumed amperage and

duration of EDG field flash circuit (first attempt), exclusion of some inrush currents during ist

'

minute, and total value of coatinuous loads for entire battery discharge cycle (likewise, for NK12,

the exclusion ofinrush current for valve FCHV0312). The licensee issued PIRs 97 3988 and 97-

4063 to address the discrepancies and to validate the portions of battery load profiles in

question. (Inspection Followup item 50-402/97 20107)

TN team also reviewed DCP $846, which implemented the ir'stallation of the new AT&T round

cell batteries in early 1996. The licensee concluded that the Technical Specifications (TS) were

not affected by the change. Neither the licensee's 10 CFR 50.59 evaluation nor design change

package provided any basis for its conclusions. Currently, TS Sections 4.8.2.1.e and f (battery

capacity replacement criteria and Lattery degradation criteria) appear to be noncor.servative

because the batteiies were sized without any aging factors and the battery performance

characteristic was changed. The team determined a TS change was warranted as indicated

below.

The 80 percent battery capacity replacement cherion used in the TS is in accordance with

Institute of Electrical and Electronics Engineers Inc. (IEEE) 450-1975. Section 1 of lEEE 450-

1975 states that battery sizing is one of several applications that are beyond that document's

intended scope. In Section 6, the same IEEE standard states that "the timing of the replacement

is a function of the sizing criteria utilized and the capacity margin available, compcred to the load

requirements." It further states that a battery should be replaced within 1 year of when its

capacity deteriorates to 80 percent.

Since the sizing criterion was the basis for the replacement of the batteries, the licensee should

look for the bases of the replacement criterion outside of IEEE 450. IEEE 4851978 describes its

scope as defining de loads for generating ststions and sizing the batteries to supply those loads.

Section 6.2.3 of IEEE 4851978 recommends that an aging factor of 125 percent of the e/pected

load demand be utilized in order to meet the battery replacement criterion of 80 percent used in

._

IEEE 450-1975, which is therefore also applicable to TS Section 4.8.2.1.e and USAR Section

8.3.2.1.2. The licensee stated that IEEE 485 is not within their licensing basis. The team noted

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that calculation E 3 W, " Class 1E Battery System," Revision WO, for the original square cell

batteries (Gou'd) was performed in accordance with IEEE 485-1983 with respect to aging, design

margin, and temperature. The present battery sizing calculation (NK E 002) also referenced

IEEE 485 but specified an aging factor of 1.00, At 80 percent of its nominal rated capacity, a

battery can stih supply its rated 100 percent demand load if a 1.25 aging factor is used. Since en

aging factor was not considered in the design of AT&T round cells, et 80 percent of rated

capacity a Wolf Creek round cell batt( / could possibly not be capable of meeting 100 percent of

the demand load. The existing design capacity margin of 25 percent as stated in USAR 8.3.2.1.2

is intended for load growth and could be used up at the battery's end of life. Further, the licensee

has experienced some problems with recognizing totalload on the de buses (see Section

E.1.2.3.2.e).

The vendor manual states that the round cell battery capacity under float conditions increases

with age. The design change package also states that the round cell battery capacity will

increase with age during the life of the battery. Neither the vendor manual nor the design change

package discussed how a round cell battery would perform under discharge conditions. The

team inquired about the testing performed by the manufacturer or the licensee to verify the

battery degradation rate. The team noted that there is not enough test data or historical data

available to statistically validate a specific performance criterion during its service life. During the

inspection the licensee provided some test data from the discharge tests performed by the

vendor which indicated that the cells may experience some capacity loss over time. The

licensee stated that the loss would be gradual and the capacity may increase if rect.arged

properly after the discharge. The licensee believes that the round cell batteries would degrade

similai to the typical square cell batteries under discharge conditions The team noted that how

a round cell battery would degrade, whether gradually or suddenly, once degradation started was

not known at the time the battery was installed and stillis not known because sufficient test data

is not available. The licenses failed to question whether the capacity performance requirements

in the TS were exacting enough to measure adequately the performance of the round cells.

The team determined that since the battery was sized without any aging factors and the

performance characteristics of the battery were changed for the reasons stated above, the TS

was affected by the des;gn change. The licensee's 10 CFR 50.59 safety evaluation failed to

identify the effect on the TS Presently, TS Sections 4.8.2.1.e and f appear to be

nonconsuvative. The licensee presently disagrees with the team's contentions; thus this matter

is being referred to the NRR staff for further review and resolution. (Inspection Followup Item 50-

482/97 201 08)

The team also reviewed Section 8.3.2.2, " Battery Capacity," of the NRC safety evaluation reports

(SERs) NUREGs-0881 and-0830, and determined that the licensee had failed to comply with the

staffs position on battery capacity for the standardized nuclear unit power plant system

(i NUPPS) plants. The licensee had sized its new AT&T round cell batteries for a 25 percent

margin as stated in USAR Section 8.3.2.1.2, but not with the 50 percent margin stipulated by the

NRC staff since th3 correct values of applicable battery sizing factors were not utilized. The

staffs SER states that the TS are written assuming a 50 percent greater than-required capacity

for each Class 1E battery. Presently, the new AT&T round cell batteries are sized with the

following margins, as statad in PMR 03845, Revision 0, and battery sizing calculatior NK-E-002:

NK11 32 percent; NK12 35 percent; NK13- 65 percent; and NK14- 35 percent. However, the

l

team found errors in the way the design capacdy of the batteries was calculated in the battery

sizing calculation. The licensee used an incorrect discharge rate in determining the percentage

l of ampere-hours required for each cell. The team calculated the design capacity margin for the

I

NK11 battery as 23 percent,instead of 32 percent as stated in the above documents. Similar

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results may be expected for other batteries. The team determined that tha Class 1E batteries

, are not sized with appropriate design capacity margin as stated in the staff's GER and USAR

, Section 8.3.2.1.2. This issue is being refereed to the NRR staff for further review along with the

battery TS issue mentioned in the above paragraph. (Inspection Followup Item 50-482/97

201 09)

c. DC Fault Cortribution

The team reviewed calculation NK E-003, " Class 1E 125 Vdc batteries short circuit study," *

Revision 0, and dolermined that all de buses and associated cabling were conservatively sized

for the available short circuit currents. Fuses provided the correct overload and fault protection

for the DC system distribution circuits, and the correct sizing of fuses ensured the requisite

selective coordination between fuses in series when upplicable. The licensee originally assumed

that the AT&T round cell batteries had a lower fault contribution and did not perform an analysis

before declaring them operable in enriy 1996. !n response to the team's questions, the iscensee

performed a quick calculation during this inspection to verify its assumed lower fault current

contribution f) de buses from the AT&T round cell batteries. In addihn, the licensee has

decided to clarify assumptions 3.4 and 3.6 in the above calculation both to achieve cons;stency

and to state which motors provide a fault contribution to the total fault current under PIR 97 4063

d. DC Load Flow / Voltage Drop

The team reviewed calculation NK E 001, " Class 1E DC Voltage Crop," Revision 1, and

determined that it adequately demonstrated that all available de components would have

sufficient voltage to property operate, except those components for which the licensee assumed

a value of 100 Vdc in the load data base of the above calculation because the vendor did not

stipulate a minimum value. The team questioned the licensee about this apparent discrepancy in

this data base. In response, the licensee issued PIR 97-4180 which indicated that all de

equipment procured having an operating voltage of 140 Vdc would function adequately. For

those devices assumed as having a minimum operating voltage of 100 Vde, the licensee has

decided to evaluate them under PIR 97-4043. Preliminarily, the licensee determined that tnere

were no operability concems and that it would either analyze each circuit on a case by case

basis or devise a generic solution applicable to all the affected circuits. Assuming 100 Vic for

devices with no specified minimum voltage is an example of an unverified assumption. The

licensee failed to demonstrate the adequacy of this design. This is contrary to Criterion 111 of

10 CFR Part 50, Appendix B, which requires that design control measures verify or check the

adequacy of a design. (Unresolved item 50-482/97 20110)

The team also reviewed the basis for battery NK11's minin'um terminal voltage and questioned

the licensee about the justification for the stated value of 107.3 Vdc for the worst case minimum

battery voltage. This voltage of 107.3 Vdc was originally Trived by using the minimum operating

voltage of 105 Vdc for the Class 1E inverter and then a c. .servative voltage drop between the

inverter and the battery's terminals. The I;censee initiated an effort under PIR 97-4185 to

evaluate the minimum required voltage for each Class 1E battery and determined that the end

voltage for each battery's discharge cycle, for either station blackout (SBO) or LOCA, would have

to be raised. The licensee determined that the worst case is NK11, whose required end voltage

is to be raised from 107.38 Vdc to 111.659 Vdc to operate the worst case loads 15 PS and

48 PS. No operability concerns were identified by the licensee. The team noted that the battery

discharge current during a battery performance test envelops the current drawn during the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

service test, and the minimum battery voltage based on analytical results is higher than 111.659

Vdc either for the SBO or LOCA conditions. The licensee failed to establish the conect design

14

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basis inforraation in surveillance procedure STS MT 021, and calculations NK E 001 and NK E-

002 in regard to minimum requireo terminal voltages for all the batteries. This failure is contrary

to requirements specified in Criterion ill of 10 CFR Pari 50, Appendix B, which requires that the

design basis be correctly translated into specifications, drawings, procedures, and instructions

and that design control measures verify or check the adequacy of a design. (Unresolved item

50-482/97 201 11)

e. DC Load Control

The team reviewed procedure Al 05-00," Electrical Load Growth," Revision 0, and determined

that Step 8.5 categorizes all de load growth changes (positive or negative) as significant. Step

6.6 of Al 05-006 requires the Load Growth Coordinator to reviso the applicable de calculations

when significant change occurs to the de system or during a refueling outage, whichever comes

Crst. The licensee has not always operated in this manner in the past. For instance, DCP 5248

and PMR 4304 listed soms de calculations, including NK E 002 for battery sizing, as affected

documents. To date the c:.lculations have not been revised to incorporate the respective de load

changes of those two deOgn change packages, even though the systems affected by DCP 5248

were declared operable in the last refueling outage in November of 1997 and the systems

affected by PMR 4394 even eariier. The licensee has issued PIR 97 4123 to address this issue.

The licensee in th,s instance failed to adhere to the procedure on electricalload growth. This is

contrary to Criterion V of 10 CFR Part 50, Appendix B, which requires that activities affecting

quality shall be accomplished in accordance with instructions, procedures, or drawings.

(Unresolved item 50-482/97 201 12)

In addition, the team reviewed the current de load data base which enables the Loaa Growth

Coordinator to ascertain the total cumulative value of outstanding de load changes. Step 6.3 of

procedure Al 05-006 requires an Electrical Data input Sheet for each pemianent modification that

has an effect on (adds, deletes, or changes a load of) the electrical distribution system. This

load data is to be confirmed by an independent verifier and sent to the Load Growth Coordinator.

The team ascertained the following errors in the data sheets submitted for DCP 5248. The team

noted that relay 52XX (page 8 of the load data base) and relay 52YY (page 18 of the load data

base) in the breaker control circuit of the battery charger were stated for deletion when in fact

they were not. After the inspection, the licensee provided a revised load data base for just the

load changes due to DCP 5248. The revised data showed additionalloads that were not

accounted for and that the totalload had increased (momentary loads by about 7 amperes and

continuous loads by 0.25 ampere) Thus, there are a number of discrepancies in the original

data base being maintained by the licensee. The licensee is evaluating this issue under PIR 97-

4125. (Second example of Unresolved item 50-482/97 20112)

The team also noted inconsistencies between the load growth p.- dure Al 05-006 and

engineering screening form APF 05-002 01. Presently, engineering screening form APF 05-002-

01 a!!ows Design Engineering to evaluate electricalload changes only for increases or decreases

of 10 kW or more when other screening conditions for load changes (such as addition of cables,

separation problems, or increased imperage in existing cables) are not applicable. Procedure

Al-05-006, Step 6.5, invokes different load change values to initiate evaluation and requires

different actions. The licensee is evaluating this under PIR 97-4123. (Third example of

Unresolved item 50-482/97-201 12)

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f. Surveillance Tests

The team reviewed surveillance procedures STS M1 -021, " Service Test for 125 VDC Class 1E

Batteries," Revision 10, and STS MT-022, "5 Year 125 VDC Discharge Battery Test," Revision 9.

The lic.ensee did not fully incorporate the requirements and acceptance limits contained in

app'icabin design basis documeris into the above surveillance procedures. The acceptance

criterion in Ptep 6.1 of STS MT-Oh is that the test be successfully completed with no other,

riiore definiti'te requirements. The licensee did not incorporate the design basis requirer;.e

contained in calculations NK E 001 " Class 1E DC Voltage Drop," Revision 1, and NK E 002, l

  • Class 1E Battery Sizing," Revision 3 pertaining to whether the battery discharge current adhered

tc the load profile and wnether final battery terminal voltage was greater than the minimum

allowable value for the buttery being tested. These parameters are not being verif'ed by the

licenses. Similarly, the acceptance criterion in Step 6.1 of STS MT-022 is only that the battery

being testod show no signs of degradation with no details on how to successfully complete the

test. IEEE 450-1975 in Section 5.4.1(2) coquires that a constant discharge rate be maintained

until battery terminal voltage falls to a value equal to the minimum specified average voltage per

cell (t.75 Vdc per Section 8.6 of ST S MT-022, or 105 Vdc for 60 cells). A nondetectable failure

in ble load bank would allow a battsry's capacity to be incorrectly evaluated as adequate, still

mee'ing the TS requirements end needing no corrective actions. However, battery terminal

voltage or discharge current could deviate from their acceptable test ranges without being

detected by technicians. As an example, in the latest capacity test for NK11, completed in the

last refueling outsge, the decesing current was not detected by technicians or reviewers (see

l

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the following peragraph for more details). In addition, Step 4.7 of STS MT 021 and Step 4.11 of

STS MT-022 are somewhat unclear about corrective actions to be taken for test deviations or for

nonadherance to the less than thorough acceptance criteria in these procedures. The licensee is

evaluating these issues under PIR 97 3941 for STS-MT-022 and PIR 97 3989 for STS-MT-021.

These surfeillance test procedures did not meet the test control measures specified in Criterion

XI of 10 CFR Par 150, Appendix B, which require tests to be conducted in accordance with

written test procedures that incorporate requirements and acceptance limits contained in

applicable design documents. (Unresolved item 50-482/97 201 13)

The licensee performed a capacity test for Class 1E battery NK11 on November 11,1997, in

, accordance with surveillance procedure STS MT-022. The load bank failed with approximately

I 20 minutes remaining in the test and before the battery terminal voltage reached a final voltage

of 105 Vdc. The licensee performed an evaluation of the results of the test and decided to

terminate the test. The minimum value of evailable battery capacity as verified by actual test

was 95 percent; post test analysis put it at 104 percent, which satisfied the TS requirement. The

inspection team reviewed the results of the test and determined that the battery discharge

current gradually decreased for about the last 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the test. In doing the capacity test, the

licensee did not follow IEEE 450, which requires that battery discharge current be maintained

constant during the test and the test be continued until the final voltage, typically 105 Vdc for a

60 cell battery, is reached. The licensee did not notice the decreasing battery current until

questioned by the team and subsequently during the inspection had to determine by analysis

what the battery capacity would have been if the test had been completed. The actual test va!ue

remainod at 95 percent, but the proposed analytical value, based on en ongoing analysis at the

time inspection was completed, decreased to 100 percent from the 104 percent stated above.

The licensee failed to consider declining battery discharge current during its initial analytical

determination of the expected capacity of the NK11 battery and failed to take appropriate

, corrective actions. The licensee is evaluating this issue under PIR 97-3941. It appears that the

licensee took improper or incorrect corrective actions because the analysis for battery capacity

did not consider the decreasing battery current. The licensee's corrective action measures did

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not meet the requirements specified in Criterion XVI of 10 CFR Part 50, Appendix B.

(Unresolved item 50-482/97 201 14)

g. Fuse Control

The team reviewed the licensee's fuse control program. Procedure AP 03A 001," Fuse ,

Verification and Control," Revision 1, govems the licensee's program for controlling both ac and

de fuses. It is supplemented by the fuse list document, WCRE-08. The program seeks to design  !

and field-verify all safety related fuses by the end of 1998 or during the next refueling outage. I

This effort is more than 90 percent complete. The team selected a sample of installed de fuses

and of fuses depicted on electrical schemes. The licensee was able to demonstrate that all

fuses in the sample were correctly assigned and labeled in the present fuse list,

h. Battery Charger

The team noted the following discrepancies in a 10 CFR 50.59 evaluation and the design basis

,

for the Class 1E battery chargers:

1. The 10 CFR 50.59 evaluation for rnooification DCP 5248 (swing battery charger addition)

stated that there would be no changes in the operating parameters at Wolf Creek. The

team noted that this statement is not true because there is a smallincrease in operating

parameters (amperes and kilowatts) for both ac and de buses due to additionalloads

from the new swing charger. The licensee's analyses adequately supported the design.

But the safety evaluation did not refer to these analyses or discuss the impact on

operating parameters. The team determined that the licensee's safety evaluation

documentation was weak. The team also determined that this problem did not have an

adverse effect on the results of the 10 CFR 50.59 evaluation.

2. Load data tables on drawings E 11NG01 and E 11NG02 show an ac input current value

of 59 amps when the de output of the train A Class 1E charger is 300 amps. Actual ac

input current, taking into account power factor and inefficiency, is 81 amps. The licensee

issued PIR 97 4044 to address this issue.

3. Procedure AP 05-002, Step 6.4.5.3, requires that calculat'ons affected by a design

modification be listed as affected documents in that design change. However, calculation

NK EW 001 was not listed as an affected document in DCP 5946 (design change for

addition of swing battery chargels) untilits absence was pointed out to the licensee by

the team. It was listed as a reference document instead. The licensee took prompt

measures to address this issue.

E1.2.2.3 Conclusion

The team concluded that generally the essential power supplies for th3 RHR, CCW, and support

systems were capable of performing their safety functions as required by their design bases.

The team identified deficiencies in surveillance testing and siziag of batteries and load growth

control. The majority of electrical calculations were adequate but sufficient errors,

nonconservative assumptions, and omissions have been identified to warrant a review of all

critical electrical calculations to reverify that their design basis is accurate and consistently

applied. in some cases analysis did not exist to support the design bases. The team identified

weaknesses in some 10 CFR 50.59 evaluations and design changes. -

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E1.2.3 :nstrumentation and Controls Design Review

E1.2.3.1 Inspection Scope

The scope of the instrumentation and controls (l&C) design assessment was to review RHR

system design documents such as the USAR, TS, system descriptions, setpoint documents,

setpoint calculations, instrument loop uncertainty calculations, specifications, maintenance,

surveillance, and operating procedures, design drawings, modification packages, and

miscellaneous l&C documents.

E1.2.3.2 Observations and Findings

The system design oocuments reviewed by the team were consistent with the design bases

except for the items identified in the following subsections.

, a. RWST LevelInstrumentation

The team reviewed the RWST levelinstrumentation design with respect to RG 1.105,

,

" Instrument Setpoints," Revision 1.

TS Section 314.5.5 requires a minimum contained RWST volume of 394,0 , aallons with a 2400

to 2500 PPM boron concentration for the ECCS function. Four redundant level instrument loops

(LT 930 through LT 933) provide input for indication, initiate RHR pump sudion switchover from

the RWST to the containment recirculation sump on low-low level, and a ' arm on low level to

signal that the RWST is approaching the TS limit. Level setpoint bases for the RWST are

provided-under USAR Figure 6.3-7 and Bechtel calculation BN 20,'RWST Setpoints,"

Revision 1, as follows:

Setpoint Contained

Function Volume _ Heloht

Hi alarm 413,000 gal. 529'

LO alarm 400,000 gal. 513"

LO-LO alarm & 236,200 gal.* 208'

switchover

Empty alarm 54,600 gal. 53'

' Minimum volume required for RHR rx 1 switchover

Sarveillance procedures STS IC-508A, "RWST Level Transmitter Calibration," Revision 4, and

STS IC-508B, " Calibration of RWST Level Instrumentation," Revision 6, and the setpoint

document translated the above values to 99.8 percent,96.88 percent, and 36 percent for high,

low, and low-low level setpoints, respectively, using the level transmitter taps (located 24" above

tank bottom) as the zero reference point. The team also reviewed calculation SA 90 056,

" Reactor Protection System ESFAS Channel Error Allowances," Revision 0, which calculated the

RWST levelinstrument loop uncertainty. Based A che review, the team found the following

discrepancies:

1. Calculation SA 90-056 did not consider density variation due to temperature and boron

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concentration in determining the RWST levelinstrument loop uncertainty. As a result,

- previously calculated instrument inaccuracies were incorrect. This could affect alarm

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setpoints, the RHR suction switchover setpoint, and the control room level indicators that

maintain Technical-Specification-required inventory. The calculation did not fully comply

with RG 1.105, Section C 4, which requires consideration of environmental effects on the

instrument setpoint determination. The new estimated inaccuracies had the following

impact:

Control room indication (RG 1.97 Type A variable) The actual RWST level could be

2.93 percent less than the indicated value; therefore, an indicated minimum Technical

Specification value of 94 percent would actually be 91.07 percent, corresponding to

383,962 gallons (the TS value is 394,000 gallons). This variable is monitored in the

control room per procedure STS CR 001,' Shift Log for Modes 1,2 and 3," Revision 35,

and SDTS CR-002," Shift Log for Msdes 4,5 and 6," Revhion 22. The team noted that

these procedures do not include Lny correction for instrument loop inaccuracy. The

licensee gave the team a copy of a letter sent to NRC (SLNRC 84 0089, dated May

31,1984) justifying the use of indicated readings (without regard for instrument

uncertainties) to satisfy TS surveillance requirements. The licensee could not find

documentation of the NRC's acceptance of the licensee's position. However, preliminary

evaluation by the licensee indicated that there is adequate margin in the NPSH analysis

to compensate for level indication inaccuracies up to 17 percent.

Low-low level switchover setpoint With an estimated inaccuracy of 3.24 percent, the

switchover point would be reduced from 36 percent to 33.89 percent, corresponding to a

tank volume of 154,657 gallons. This would reduce the volurr.e available for injection

between the minimum indication level of 91.07 percent and the corrected swapover

setpoint of 33.89 percent, which is 383962 - 154657 = 229,305 gallons. This is less than

the required volume of 236,200 gallons, as specified in USAR Fig. 6.3 7 and calculation

BN-20. On the basis of a preliminary evaluation by the licensee, the team considered this

reduced injection volume did not impact the pump NPSH.

Low alarm setpoint - With an estimated inaccuracy of 2.51 percent, the low level alarm

could be as low at 94.57 percent, which is very close to the minimum TS reading of

94 percent. This condition would reduce the margin for operator response ir. case of an

actual leak in the RWST.

Empty Alarm - As a result of the transmitter error and u .._ble inaccuracy in the

instrument loop, it is estimated that the ' empty" alarm cou'd drift 14" below the existing

setpoint of 53". This levelis approximately 3' above the RWST suction pipe. The effect

would be a late alarm, which would reduce the margin for operator response to protect

pumps that are taking suction from the RWST from loss of NPSH.

2. Calculation BN 20 had assumed instrument inaccuracies of 1 percer,t for bistables and

3 p treent for totalloop error to establish the existing RWST level setpoints. The team

was unable to find uncertainty calculations, supporting these values.

The licensee's preliminary evaluation indicated that the above issues did not constitute an

operability concem. The licensee issued PIR 97-3974 to address the instrument setpoint and

indication inaccuracios. The team determined that the requirements defined in Criterion lli of

Appendix B to 10 CFR Part 50, which requires that the design basis be correctly translated into

specifications, drawings, procedures, and instructions and also verifying or checking the

adequacy of the design were not followed for the RWST levelinstrumentation defgn.

(Unresolved item 50-482/97 201 15)

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b. Seismic Qualification of the RHR Pump Suction Pressure Indicators

A review of P&lD M12EJ01, Revision 15, showed the PHR pump suction pressure gauges

PI 601 and PI-602 as normally valved open. The Q-List shows these instruments as safety-

related for pressure boundary only. These instruments are the original equipment fumished by

Westinghouse as commercial grade !! ems. They were later qualified through engineering

judgment (ref. Westinghouse document RCS/ CIEL (89) 299, dated 07/10/89) and commercial

grade dedication (ref. Package 028 P0015, Rev 0). Based on review of the commercial grade

dedicauon package, pressure integrity was verified through hydrostatic testing but there was no

verification to ensure that the pressure boundaryis maintained during . seismic event. To satisfy

IEEE 3441975, ' Recommended Practices for Seismic Qualification of Goa ill Equipment for i

Nuclear Power Generating Stations," 0:id RG 1.100, " Seismic Qualification of Electric Equipment l

for Nuclear Power plants," Revision 1, both of which the licensee has committed to, the

instruments should also be seismically qualified for pressure integrity. How6ver, the licensee had

not dor,e an adequately documentad analysis to establish the seismic qualification of the installed

units. The licensee provided a seismic qualification report, Farwell & Hendricks Report 50089.6,

for a pressure gauge of similar make and model to demonstrate that the installed gauges are

qualifiable, but in order to validate this report for the pressure gauges that were furnished by

Wesiinghouse, a similanty analysis was required. Subsequent to the inspection, the licensee

issued calculation XX F010 to establish the qualification. The team did not review this

l calculation. The team determined that the licensee's design control measures did not verify or

l check adequacy of the pressure gauge design in accordance with Criterion ill of Appendix B to

10 CFR Part 50. (Unresolved item 50-482/97 201 16)

c. RHR Instr, .. lent Loop Accuracy and Setpoint Calculations

The team reviewed the licensee's setpoint methodology, uncertainty calculations, a.:d related

calculations for various RHR instrument loops to verify that adequate tolerance for instrument

errors had been incorporated in the design. The team reviewed documents CWS-SNP-470C,

"SNUPPS Flow Switch Setpoints"; M-74' 00025-W35," Precautions, Limitations and Setpoints

for Nuclear Steam Supply Systems," Revision 4; J2A02, ' Accuracy, Foxboro Bistable,"

Revision 0; J2F01, " Accuracy of Standard Orifice Plates," Revision 0; AN 96-074, *RWST Water

Level Necessary to Supply Adequate NPSH for the ECCS Pumps," Revision 0; XX E003, "RHRS-

RCS ISO-VLV Open Setpoint," Revision 0; J K-GEN, ' Instrument Loop Uncertainty Estimates,"

Revision 0; and J1 GEN," Instrument Loop Uncertainty Estimates," Revision 1. The team's

review determined that the calculations adequately demonstrated loop accuracies and setpoints,

d. System Modifications

The team reviewed five l&C modification packages for the RHR and RWST systems:

1. PMR 02820," Containment Sump Level Indicator Scale Replacement'

2. PMR-03142, " Addition of Flow Indicators for EJ System"

3. CCP-05804, " Containment Sump Level Indication Modification"

4. PMR 03529, * Deletion of RWST LevelIndicator BNLIS0001"

5. PMR 04637, "RHR Hest Exchanger Outlet Valve Nitrogen Backup"

Based on the review, the team concluded that the design,10 CFR 50.59 evaluation, and

document closeout were performed adequately for these modifications. However, the team

noted that 10 CFR 50.59 evaluation documentation could be improved to provide sufficient

details and basis for an independent reviewer to verify the change without performing an in-depth

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review of the design change package. The licensee had recently revised its procedure to

emphasize that sufficient details and basis be supplied in 10 CFR 50.59 evaluations.

E1.2.3.3 Conclusions

The l&C design for the RHR system and the interfacing portion of the RWST was considered

adequate. The team was concemed with the level instrument loop uncertainty determination,

which failed to consider dent!!y changes due to environmental effects on the RWST. Unverified

instrument accuracles were used in the RWST level setpoint determination and incomplete

seismic qualification documentation was observed for the safety-related pressure gauges.

E1.2.4 System interfaces

E1.2.4.1 Inspection Scope

The team inspected mechanical aspects of the emergency diesel generat:rs (EDGs), the

'

essential service water (ESW) system, and the emergency core cooling system (ECCS) area and

battery room coolers to ensure that these systems properly supported functioning of the RHR and

l CCW systems.

l The review of interfacing system attributes involved the USAR, TS, calculations, drawings, test

l procedures and results, and licensing commitments. System walkdowns and discussions eth

licensee personnel were also ennducted.

E1.2.4.2 Observations and Findings

a. Emergency Diesel Generators (EDGs)

The team performed a walkdown of the *B' EDG room, the fuel oil system and the

combustion / ventilation air intake / exhaust system and found no material condition or physical

configuration problems. Sampling procedure STS CH-015 for the fuel oil tanks and fuel oil

shipments was reviewed and found to meet the requirements of TS Sections 4.8.1.1.2.d & e.

Surveillance test procedures STS JE-003A and -004A were also reviewed and found to conform

with the periodic sampling and draining of water from the fuel oil storage tanks required by TS

Sections 4.8.1,1.2.b & c.

The team verified that the missile protection design of the EDG ventilation / combustion air intake

structures was in accordance with USAR Section 9.4.7.2.3 by reviewing calculation 06-05 F and

structural drawings M-1G052, M 1G054, C-1C5311, and C-1C5904.

To verify that adequate cooling and combustion air are available to the EDGs, the team

reviewed USAR Section 9.4.7; calculation GM-320, " Diesel Generator Building HVAC - Required

Air Flow,' Revision 0; and EER 90-GM-02, Revision 2. The team found that EDG

combustion / ventilation air sizing was adequate and that the system was designed such that

combustion air is available even if the ventilation supply fan and system dampers fail.

EDG room design basis temperature was verified by reviewing calculation GM-M-002, " Diesel

Generator Building Minimum Room Temperature," Revision 1.

The team noted that a portion of the EDG exhaust stacks was exposed to the plant exterior 3

environment. The stacks were therefore reviewed for protection against design basis events.

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During this review it was determined that the stacks had not been analyzed for design basis wind

loads. A close review of the Wolf Creek licensing basis (USAR Sections 3.3,3.5.2.6, and 9.5.8

and the corresponding sections of the SER (NUREG.0881)) indicated that this analysis was not

specifically required; however, the licensee agreed that it would bo prudent engineering desig7

practice to include wind loads in the current analysis, which is found ir, calculation P 311.

Preliminary analysis indicated that the existing support st7uctures were adequate for design basis

wind loading. PIR 98-0060 was issued to track this item to closure.

Electrical aspects of the EDGs inspected are discussed in Section E1.2.2.2.1

b. Essential Service Water

The team verified the ESW system flow and heat transfer capabilities during normal plant

shutdown and accident conditions. The team's review is discussed in Sections E1.2.1.2 and

E1.3.1.2. The team determined that the ESW system could provide the required flow oi

7150 gpm at an ultimate heat sink (UHS) worst case temperature of 95'F to achieve and

maintain safe shutdown conditions. To verify the pump has adequate NPSH, the UHS minimum

level stated in calculation EF M-014, * UHS Thermal Analysis Review for Pawer Rerate,'

Revision 1, was reviewed against the ESW pump submergence requiremrmt given on ESW

pump curves M-089-K043-02 and -03, dated Augu:t 8,1979, and July 26,1979, respectively.

l The review determined that UHS minimum level was greater than the required level by a small

l margin (about 4 inches), and was found to be acceptable.

The team verified that the missile protection design was in accordance with USAR Section

9.4.8.2.3 by reviewing calculation K 20-05 F and structural drawings M-KG080, M-KG081,

C KC304, C-KC305, C-KC306, and C KC309.

The team performed walkdowns of the ESW intake structure and ESW/SW distribution area at

the 1974' elevation of the control building and found no material condition or physical

configuration problems. Both ESW trains were found to share space in a common room at

elevation 1974' of the control building, This room also contained non safety related piping and

equipment. For safety related ESW equipment, the team checked compliance with fire

separation, seismic ll/l and flood protection criteria and found it adequate.

ESW self assessment report SEL 96-055 was reviewed by the team. This self assessment,

performed during the first half of 1997, was found to be adequate and appeared effective in

initiating both system specific and programmatic changes. Several generic design basis and

licensing basis issues were identified by the licensee during the ESW self assessment. Not all of

the issues identified by the nelf assessment had been closed out at the time of this inspection.

c. ECCS Area and battery Room Coolers

The team reviewed USAR Section 9.4.3 and capacity calculations for the ECCS area and battery

room coolers to determine whether the coolers had the capability to maintain temperatures below

maximum design. For the ECCS areas the team reviewed calculations GL-04 W, *RHR Pump

Rooms 1109 and 1111, Heat Loads," Revision 1; GL 03 W, " Auxiliary Building HVAC." Revision

W 1; and GD 234," Essential Service Water Pumphouse, Cooling and Heating Requirements,"

Revision 1. To ensure adequate capacity for the battery rooms, the team reviewed calculations

GK M-001, " Safety Related Control Room Building HVAC Capabilities During Accident

Conditions,' Revision 2; and GK M-004, ' Loss of Ventilation During Normal Operating Conditions

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Battery and SWBD Rooms Control Building,' Revision O. The team concluded that the coolers

could maintain the design basis temperatures under the worst case ESW temperature.

E1.2.4.3 Conclusions

,

Interfacing system attributes reviewed were found to adequately support RHR and CCW system

l design basis functions.

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E1.2.5 System Walkdown

E1.2.5.1 Inspection Scope

The team conducted a walkdown of the RHR system and the plant areas, including the RHR

pump rooms, the RHR heat exchanger rooms, the RWST, the control room and penetration area,

the switchgear rooms, the battery and inverter rooms, and the cable spreading room. The team

focused on comparing system configuration to the design basis documents and the USAR. The

team also looked closely at equipment condition, area cleanliness, tagging, and the means used

to avoid potential hazards such as missiles, flooding, fire, and pipe rupture,

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E1.2.5.2 Observations and Findings

The team determined that the overall material condition of the plant areas was good. The

equipment sampled matched the design documents. However, the team identified the following

issues.

l a. RWST Valve House Room

During a walkdown of the RWST valve house room,9102, the team observed that the auxiliary

steam line used to supply the heater coils wrapped around the RWST was not seismically

supported. Two level transmitters, LT-930 and 931, were close to the auxiliary steam line and a

rupture in the steam line could potentially affect the function of these transmitters. These

transmitters are identified in USAR Section 7.4.1 as being required for plant safe shutdown.

Contrary to the USAR Section 3.6.2.1.2.3, no hazard analysis existed that determined the effect

of a failure of the steam line on the RWST system. The team was concemed that failure of this

steam line, which could go uridetected for 6 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> until detected during operator rounds

(procedure CKL ZL-001, " Auxiliary Builoing Reading Sheets"), could affect plant safe shutdown.

The licensee issued PIR 97 3896 to address the team's cancem. The licensee performed a

hazard analysis during the inspection and determined that a failure of the non-safety-related

auxiliary steam piping would not affect plant safe shutdown. This analysis also determined that

the RWST level transmitters, cui ently identified in USAR Section 7.4.1 as being required for

plant safe shutdown, were not listed in USAR Table 3.11(b)-3 as being required for cold or hot

shutdown. The licensee also determined that emergency boration procedure OFN BG-009 did

not re'erence use of the RWST level parameter as required for the proper and rapid insertion of

negative reactivity to achieve plant shutdown. The licensee has determined that the RWST level

transmitters are not required to bring the p! ant to a safe shutdown and PIR 97 3958 has been

initiated to delete reference to the RWST level transmitters in USAR Section 7.4.1. The hazard

analysis was reviewed by the team and found acceptable. The team determined that the above

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issue represents a weakness in the licensee's design and configuration control process._

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b. Control Room RG 1.97 Instrumentatior.

l The PHR and RWST instrumentation on control room pielt 017,018,019, and 020 was

veaified to comply with RG 1.97 and with USAR Appendix 7.4., which documents the licensee's

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commitment to RG 1.97. Per the RG, specialidentification tage are required for Type A, B, and

O' C variables, Category 1 & 2 indicating and recording instruments in the main control room.

k Specialidentification is not required for Type D Category 2 instruments. During the walWwn the

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team noted that RWST level indicators LI-930 through LI 933 and RCS pressure indicators Pi-

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934 through PI-937, classified as Type A variables, do not have unique markirgs to identify theri

es RG 1.97 instruments, however, the instrument nameplates are color coded according to the

engineered safety feature (ESF) group' .g and separation, which the licensee considers a method

of identification for post-accident usw. USAR Appendix 7A, Section 7A.1, wMch states %trict

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compliance to the many piescriptive recommendations (of RG 1.97) is not provided in all cases,"

appears to be the basis for not identifying RG 1.97 instrumentation by special tags. Based on

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the team's review, tho existing RG 1.97 oesign for the RHR and RWST is considered adeo

c. Equipment Tagging

The inspection team noted missing tags for RHR pumps and motors PEJ01 A/B and DPE~ iA/B

and sump pump level switches LS0021, LSH0008, and LSH0022. The licensee took prompt

measures to correct this problem.

d.10 CFR 50.59 Eva!uation for RCS Drain Down Procedure.

1

L A walkdown of the RHR heat exchangcr rooms was performed during the refueling outage and

the inspection team observed that nitrogen bottles were temporarily installed inside the rooms.

The bottles were tied to support steel by #9 wire. It appeared that they were not seismically

restrained and the condition could pose a petantial missile hazard. Those nitrogen bottles were

' installed in ucordance with procedure GEN 00-007, "RCS Drain Down Procedure," Revision 19,

to provide backup air 'or the RHR heat exchangor outlei valve operators (EJHCV-606 and 607)

=

during the refueling outage.

Based on a review of this procedure and its 10 CFR 50.59 screening (GEN 00-00719, No.59,

approved 9/9/94), the team noted that the 10 CFR 50.59 screening did not check the screening

I question " yea" to generate a new 10 CFR 50.59 safety evaluation or develop guidar,ce on

seismically rt straining the nitrogon bNtles. Before the establishment t.a this procedure, backup

air and nitrogen bottles were installed through the temporary plant modification process. The

team reviewed those temporary modification packages and >he'.r 10 CFR 50.59 evaluations and

found that the bottles needed to be restrained by chains, wire rope, or tube frame stanchions

instead of the #9 wire that was used. It appears that the seismic restraining requirements for the

_ nitrogen bottle.s under the temporary modification wem not adequately translated during the

development of the RCS drain down procedure or in the 10 C6R 50.59 screening review. The

licensee's 10 CFR 50 59 screening raview failed to perform a ufsty evaluation as required by

10 CFR 50.59. Since the bottles were removed at the end of the outage, this condition did not

constituto an operability concem. The licenses %ued PIR W-3961 to initiate a revision to

procedure GEN 00-007 and to generate a 10 CFR 50.59 safety evaluation. In addition to

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CrFerion il! of Appendix B to 10 CFR Part 50, the requirements of 10 CFR 50.59 were apparently

not met. (Unresolved item 50-482/97-201-17)

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E1.2.5.3 Conclusions

in general, the RHR system design observed during walkdown was consistent with the design

basis requirements. However, the team identified several h. sues: the absence of a hazard

evaluation for the auxiliary steam line break in the RWST room, missing tags on some

equipment, failure to perform a 10 CFR 50.5g safety evaluation and lack of procedural guidance

for the nitrogen bottle installation in the RHR pump room.

E1.3 Component Coolina Water (CCW) System

F.1.3.1 Mechanical Design Review

E1.3.1.1 Inspection Scope

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The team evaluated the mechanical aspacts of the CCW system to determine its ability to

perform the design duty and safety functions during normal power operation and accident

conditions. The evaluation included a review of the system descriptions, USAR, TS, draw ...-,

calculations, modifications, operating and surveillance testing procedures and test records,

information notices, generic letters and environmental qualification (EQ) files.

E1.3.1.2 Observations and Findings

a. System Flow and Heat Removal Capability

The team reviewed the following calculaticns:

1. EG-06-W, " Component Cooling Water System Calculation," Revision W-3

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2. EG-09-W, " Tube Plugging for CCW Heat Exchangers EEG01A/B Max. CCW Temperature-

LOCA," Revision 0

3. EF-10-W, " Essential Sen, ice Water Flows at 90'F," Rev!: ion 1

4. EG-02-W, " Component Coolirig Water Pumps PEG 01A, C & D Performance," Revision 0

5.- EG-11, Component Cooling Water Heat Exchanger and By-Pass Flow," Revision 0

6. EG-11-W " Component Cooling Water Heat Exchanger and Bypass Pressure Drop

Evaluat,an," Revision 0

7. EG-18, CCW Circulation Time via RHR HX," Revision 0

8, EG-27, "Effect ci Diesel Ger stator Frequency Degradation on CCW Pump Operation."

Revision 0

9. SA-92-006," Updated Heat Rejection Rate to the UHS," Revision 1

10. SA-90-042, " Heat Rejection to the Ultimate Heat Sink During LOCA," Revision 0

11. SA-91-085, " CONTEMPT-LT Component Cooling Water (CCW) Heat Exchanger,"

Revision 0

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12. SA-89-017," Evaluation of CCW & RHR Heat Exchanger Performance for the Extended

Fuel Operating Cycle (18 Months)," Revision 0

The above calculations documented that adequate flow and heat removal capability were provided

for varior plant operating modes and poriutated design basis events. However, the team identihed

the following discrepancies:

1. USAR Tables 9.2-9,10, and 11 list flow of 40 gpm to a removed reverse osmosis unit. The flow is y

mainta:ned in the piping in prevent corrosion. This flow is not accounted for in calculation

EG-06-W. The reverse osmosis unit was removed and the changes to the USAR were made as

part of DCP 03406. Calculation EG-06-W was overiooked. The licensee initiated PIR 97-3857 to

resolve this flow discrepancy.

2. USAR Tables 0.2-9,10 and 11 and calculation EG-06-W Tables IA,18 ar.d IC do not agree on

total f)ow and heat load duty and some minor discrepancies exist on individual heat load 2. For

example, tha total CCW flow and heat load during normal operation are given in the USAR as

8

9,974 gpm and 75.15 X 10 BTU /hr, respectively. Calculation EG-06-W gives this flow and heat

load as 10,011 apm and 8. 32 X 10s BTU /hr, respectively. The licensee initiated PIRs 97-1341

and 97-3983 to resolve this item.

3. Calculation SA-89-017, "Evalustion of CCW & RHR Heat Exchanger Perrormance for the *

Extended Fuel Operating Cycle (18 Months)," Revision 0, deterraine that CCW temperature

reaches 126"F. Calculation EG-06-W Table IC is based on a CCW nmperature of 120 F. The

CCW system has been analyzed for a temperature of 130 F; therefore this is only a

documentation discre.pancy. The licensee initiated PIR 97-4052 to resolve this item.

l b. Pump Net Positive Suction Head (NPSH) and S3 stem Transients

Calculations EG-5, " Component Cooling Water System," Revision 0; EG-10," Calculation of Available

NPSH for CCW Pump," Revision 0 ; and EG-M-016. " Time Delay for Isolation of CCW Flow to RCP

Thermal Barriers," Revision 0, determined adequate NPSH for CCW pun,p: The licensee also

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property evaluated the system transients in calculations EG 12, " Component Cooling Water System

Pipe Break," Revision 0, and EG-24, "CCW Nuclear Aux. Component Train Switchover Single Valve

Failure Analysis," Revicion 1. However, the team noted nonconservebe asumptions and errors in

some of the calculations as follows:

1. In calculation EG-5, the licensee used the height of water in the CCW surge tank at the % level to

determine that adequate NPSH is available at the CCW pump. This was nonconservative

because the CCW surge tank could be at the lowest nonnal level. A preliminary, conservative

ar,alysis was provids which showed adequate CCW pump NPSH is provided at the lowest

normal level. The licensce initiated PIR 97-3837 to resoive this item. After the inspection. the

licensee revised calculation EG-5, confirming that adequate CCW pump NPSH was ava. ute if

the surge tank level was at the bottom of the tank. The NPSH available at the CCW pumps is 37

ft as ccmpared to 43 ft in the original analysis. The reouired NPSH is 12 ft.

2. In calculation EG-M-016, the 1 censee determined that a 10 second time delay for isolatbn of the

CCW high flow from the MP thermal barriers was acceptable. A smaller break that results in a

flow lower than the comrnon header flow element setpoint (210 gpm), with no action taken unt ,

the surge tank is at the high level alarm, was not evaluated. A preliminary, conservative analysis

was provided which showed the small break has worse consequences than the large break;

however, the radiological consequences are t -nded by a chemical and volume control system

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(CVCS) break with a loss of B460 gallons (analyzed in USAR Section 15.6.2). The licensee

initiated PIR 97-3837 to resolve this item. Subsequsnt to the inspection, the licensee revised I

calculation EG-M-016, confirming the results of the preliminary analysis. The loss of reactor

coolant outside of containment is 1178 gallons, compared to 202 gallons in the original analysis.

3. In calculation EG-12, the licensee assumed that the isolation valve (for non seismic CCW piping

to Radwaste) closure time was linear with respect to valve position and then used this assumption

to determine that the average flow will be half the starting flow. This is an incorrect extrapolation

of the original assumption. Attachment 1 to the calculation s,howed that the valve flow coefficient

(Cv ) is not linear with respect to valve position. Additional?/ , the change in valve Cv shodd be

added to the total system resistance to determine the effect on flow. The team noted that

instrument error was not accounted for in the calculation. A preliminary, conservative analysis

was provided by the licensee which showed adequate surge tank level (i.e., CCW pump NPSH) is

maintained. The licensee laitiated PIR 97-3837 to resolve this item. Subsequent to the

inspection, the licensee revised calculation EG-12, confirming the results of the preliminary

analysis. The remaining inventory in the surge tank is 636 gallons, compared to 1278 gallons in

the original analysis, in addition, in the above calculation the licensee assumed the guillotine

break as the worst case. A smaller break that results in a flow lower than the flow e'ement

setpoint (4500 gpm) and does no' initiate a valve dosure signal until the surge tank is at the low-

lowlevel trip setpoint was not evaluated. The above PIR also addressed this issue. The revised

calculation with the smaller break determined that more water was lost in this scenario ad that

the remaining inventory in the surge tank is 5 gallons. Adequate NPSH is still maintained for the

CCW Pump.

c. Motor-Operated Valve Design

The team evaluated the adequacy of CCW containment isolation valves to meet their design basis

requirement by reviewing motor-operated valve (MOV) design document E-025-00007(Q)-W10, *MOV

Design Conf,garation Document," Revision 9W, snd calculation EG-M-007, " Motor Operated Valve

Bounding Conditions Determination," Revision 2. The team noted that in pages 218 and 230 of

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design document E-02500007(Q)-W10 the differential pressure to close CCW valves

EG-HV-062/132 is identified as 1120 psi. These valves are required to close against reactor coolant

pressure resulting from a RCP thermal banier break. A nonconservative asscmption, that the

downstream pressure was the average of the pressure before and after closing the MOV (1130 psig),

was used and resulted in a lower than actual differential pressure. The team determined that the

downstream pressure would be 22 psig based on the static head from th CCW surge tank and the

differential pressure for the valves to close would be 2228 psid. The lice see performed a review and

determined that the only other valves affected are the BB-HV-0013/14/15/16 valves. These valves

are closed on limit switch control. There is no operability concem as the licensee provided a

preliminary cnalysis which showed that the motor operators have sofficient thrust to close the valves

against the required differential pressure. The licensee initiated PIR 97-4054 to resolve this item. The

licensee's design control measures did not ensure that motor operated valve design was adequately

verified or checked in accordance with Cr" srion ill of 10 CFR Part 50. Appendix B. (Unresolved item

50-482/97-201-18)

d. Othu Discrepancies h Cdeulaticns

The team noted the follov6 s.eaknesses in calculations:

1. Calculation EG-13, " Component Cooling Water Radiation Monitor Flow Orifice Calc. " Re vision 1,

analyzes the ioss of CCW fmm a break in nonseismic piping to the radiation monitors RE9 and

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RE10. This calculation uses an acceptance criteria of 30 minutes available for operator action to

isolate the leak after receiving a low level alarm. No basis is provided for the 30 minute criteria in

the calculation. In response to the team's question the licensee performed en analysis which (;,

showed that the operators would have approximately one hour to isolate h leak which is

adequate.

2. The licensee cculd not establish the tasis for the required capacity of CCW heat exchanger relief

valves EG-V-027 and EG-V-052 or for the set pressure of CCW surge tank relief valves EG-V-159

and EG-V 170. The CCW surge tank relief valves p7 vide overpressure protection for the CCW

t.ystem. During the inspection the licenses issued calculatbns EGM030 and EG-M-031, whicn

shawed that these valves 5m adequate capacity and proper setpoirits. The team noted that the

surge tank vacuum relief valve was sized adequately in accordance with calculation EG-04-W,

" Determine Acceptable Surge Tank Vacuum," Revision O.

e. Modi'ications and Safety Evaluations

The team reviewed seven modification packages and nine unreviewed safety question determination

(USQD) evaluations to ensure that the system design basis was being maintained and that no

unreviewed safety questions existed. 'lhe team determined that the system design basis was being

maintained and that no unreviewed safely questions existed. However, incomplete design change

and 10 CFR 50.59 (USQD) reviews had been performed * the following instances. The team also

determined that the licensee's 10 CFR 50.59 evaluatior. , enerally lacked documentation of sufficient

justification.

Design et nge PMR 4380, "CCW Temperature Change," Revision 2, and its associated safety

evaluation,59 92-0216, Revwn 0, was reviewed. This PMR changed the allowable CCW minimum

temperature from 60 F to 32 F. During normal plant operation, only one train of CCW is in operation.

The re'. undant CCW train is on standby with no flow through the CCW side. However, there is always

flow on the ESW side. Upon initiation of a safety injection signal, CCW pumps in 'he standby train

start operating, thereby urculating cold water through various CCW componer s. The operating CCW

train also experiences cold temperatures, as the nonessential heat loads are dropped off the system

ar.a flow through tne CCW Heat Exchanger is no longer being regulated because the air supp'y to the

bypass flow control velve is not safety related. The valve is designed to fail closed. The des:gn

change assumed, conservatively, that CCW is at the same temperature as the lake water,32 *F. The

following items were not adequately addressed either in the design change or in the safety evaluation.

.

The lower CCW temperature causes lowe lubricating oil temperature for several motus, resulting

in higher power requirements. The increased EDG loading was not addressed in either the PMR

or the unrevi ewed safety question determination (USQD). There is no operability concem as the

loading increase is small and the diesel generator has a large loading margin. The licensee

initiated PIR 97-3978 to resolve this item.

.

The lower CCW temperature results in a lower spent fuel pool (SFP) water temperature. The

lower SFP temperature effect on reactivity was not addressed in either the PMR or the USQD.

The minimum temperature for which the SFP reactivity was analyzed is 60 F (USAR

Section 9.1 A). The licensee has stated that the OFP temperature could approach within 4 F of

the CCW temperature. Thera was no current operability concem as On The Spot Change 97-

0898 to procedure CKL ZL-003, " Control Room Daily Readings," was issued to place an

administrative limit of 65 F on minimum SFP temperature until a rezctivity analysis at lower

temperatures was completed. The licensee initiated PIR 97-4062 to resolve this item.

I Subsequent to the inspection the licensee completed an analysis which determined that lowering

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the SFP temperature from 60'F to 35'F would reduce the reactivity. In addition, the lower

temperature has no adverse eflect on the solubility of boron because the SFP boron concentration

of 2000 - 2500 ppm is well be!ow tne sa;uration curve at 35'F.

The team determined that contrary to 10 CF~150.59, the licensee's safety evaluation did not

completely verify the absence of an unreviewed safety question and that, contrary to Criterion 111 of

Appendix B to 10 CFR Part 50, the design was not adequately verified or checked to ensure that

spent fuel pool desig.1 was not affected by the design change. (Unresolved item 50-482/97-201 19)

f. Operating Procedures

The team reviewed the CCW system operating procedures to ensure that the system was being

operated in accordance with its design basis and the commitments contained in the USAR. The

review determined that operating procedures were consistent with the CCW design basis. However,

one concem was ioentified which requited resolution.

The USAR Section 9.2.2.2.3 states that CCW flow to the spent fuel pool heat exchanger is reduced or

terminated at start of cooldown at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The licensee's procedures EJ-120, "Startup of a Residual

Heat Removal Train," Revision 32, and EJ-121, "Startup of a RMR Train in Cooldown Mode,"

Revision 11, only contain a caution to monitor Spent fuel pool temperature if cooling is secured.

Furthermore, there is no requirement to reduce or terminate CCW flow in procedure EMG ES-11,

,

" Post LOCA Cooldown and Depressurization", Revision 12. The above procedures did not implement

!

the USAR requirements. This discrepancy was previously reported in PIR 951167 but the procedures

were not corrected. The licensee initiated PIR 97-3897 to resolve this item. This is contrary to the

corrective action measures specified in Criterion XVI of Appendix B,1? CFR Part 50. (Unresolved

item 50-482/97 201-20)

g. Testing Procedures

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The team reviewed system flow verification procedures TMP-EN171, "ESW Trair A Post-LOCA Flow

( Balance," Revision 1; TMP-EN173, "ESW Train B Post-LOCA Flow Balance," Revision 1; STN EG-

001 A, " Train A Component Cooling Water System Flow Verification," Revision 0; and STN EG-001B,

" Train B Compotut Cooling Water System Flow Verification," Revision 0; local leak rate test (LLRT)

valve lineup procedures STS PE-174, "LLRT Valve Ur'eup for Penetration 74," Revision 0, STS PE-

175, "LLRT Valve Uneup for Penetration 75,' Revision 0, and STS PE-176, "LLRT Valve Lineup for

Penetration 76," Revision 0; component cooling water pump inservice pump test procedure STS EG-

1008, " Component Cooling Water Pumps B/D inservice Pump Test," Revision 13; heat exchanger

flow and differential pressure trending procedure STN PE-037, "ESW Heat Exchanger Flow and DP

Trending," Revision 11; and heat exchanger performance test procedure STN PE-033, "CCW Heat

Exchanger Performance Test," Revision 4. The reviews included th. latest results and trending data.

The team's review indicrJed that the CCW system valve leak rate testing was being performed

property as were check valve testing, pump testing, heat exrbnger tiow, diffential pressure and

performance testing in accordance with NRC Generic Letter 89-13 " Service Water System Problems

Affecting Safety-Related Equipment," and system flow balancing. No adverse trends were noted in

the test results.

E1.3.1.3 Conclusions

The team concluded that the mechanical aspects of the CCW system could perform the design

functions of cooling the safety-related equipment during the normal operating mode and in post-

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accident conditions. The team noted nor. conservative assumptions and errcrs in calculations,

discrepancies in 10 CFR 50.59 and design changes, and an inadequate corrective action for operating

procedures.

E1.3.2 Elect ical Design Review

i ne electrical design of the CCW system is reviewed in Sechon E1.2.2.

E1.3.3 Instrumentatinn and Controls Design Review

E1.2.3.1 Inspection *.> cope

The scope of the instrumentation and controls design assessment was a review of the CCW system

documents such as Chapters 7 and 9 of the USAR, Section 3/4.3 of the Technical Specifications (TS),

system descriptions, calculations, setpoint documents, design specifications, procedures, drawings,

and modification packages.

E1.3.3.2 Observations and Findings

a. Isolation of Nonseismic, Non-Safety-Related Portion of the CCW System

The CCW system instrumentation was assesst.d to verify its capability to isolate the non-safety-

related portion of the CCW system, consisting of those lines downstream of valves HV69A and

HV70A and upstream of valves HV69B and HV708. These valves ciose on low-low surge tank level,

high CCW flow, or a safety injection signal. USAR Sections 7.6.8 and 9.2.2.1.1 provide the design

bases for this isolation function. The team reviewed drawhgs M 12EG02, M-12EG03, J-02EG03,

J-12EG02, J-12EG08A, and J-12EG088, schematic diagrams, and setpoint documents and venfied

that tha design meets the design requirements as described in the USAR. TS 4.7.3 requires that a

channel operational test of the surge tank level and flow instrumentation for the isolation logic be

performed every 31 days, with charnel calibration and valve actuation verified every 18 months.

Based on a sample review of the sun aillance test data for procedures STS IC-916," Channel

Ca"bration CCW System Automatic isolation of Non-Nuclear Safety Related Components," Revision

0, and STS IC-915A, "ACOT A Tr Component Cooling Water Sys. NSSR isolation," Revision 1, the

team determined that the tests met the TS requirements.

b. Radiation Manitoring interlock With Surge Tank isolation Valve

The radiation monitoring system was assessed to verty its capability to perform the required isolation

function. The team reviewed calculation EG-13, drawings M-12EG01, M-12EG02, J-12EG02, and J.

12EG03, schematic diagrams, setpoint documents, and vendor drawings and determined inat the

desi0n meets the system functional requirements. Since the monitors were non-safety-related, the

team verified that proper seoaration from 1E circuits was maintained. Wa!kdowns were also

performed to verify locations of sampling and retum points. The team noted that PIRs 96-1129 and

97-0457 had previously identified a recurring problem conceming a reduced sample flow through the

radiation monitors, which might affect the values that are listed in USAR Table 11.5-5. The licensee

determined that the reduced sample flow did not affect the monitoring function of the radiation

monitors. The team noted that the problem was corrected and both PiRs were closed. As a result of

the review, the team cone.,luded that the design of the CCW radiation monitoring system interlock was

adequate.

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c. CCW Instrument Loop Accuracy and Seipoint Calculations

The team reymwed the folkwng licensee setpoint methodology documents, uncertainty calculations

and related calculations for various CCW instrument loops to verify that adequate tolerance for

instrument error had been incorporated in the design:

1. J-K-GEN, " Instrument Loop Uncertainty Estimates," Revisions 0 and 1

2. J-K-EG01, " Instrument Uncertainty Estimate and Safet/ Related Setpoints, System EG,

Loops 1 and 2," Revision 1

3. J-K-EG03, " Instrument Uncertainty Estimate and Safety Related Setpoints, System EG,

Loop 62," Revision 1

4. J-K EG04, " Instrument Uncertainty Estimate and Safety Related Setpoints, System EG,

Loops 77 and 78," Revision 1

5. J-K-EGOS, " Instrument Uncertainty Estimate and Safety Related Setpoints, System EG,

Loops 107 and 108,* Revision 1

6, J-EG01," Stress Analysis of Instrument Unes, EG Component Cooling Water Surge Tank A,"

Revision 2

7. J-EG08, " Stress Analysis for Instrument Tubing from RHR HX 18 to Accum. Inj. LP 3 & 4,"

Revision 4

The team's review determined that the above calculations adequately demonstrated the capability of

the instruments to perform their intended function.

d. RG 1.97 Instrumentation for CCW System

USAR Appendix 7A provides the desigr. bases for the CCW RG 1.97 instrumentation. The required

instrumentation consists of local indication for CCW flow to the engineered safety features (ESF)

systems and main control room indication for CCW inlet temperature to the ESF systems. The

I;censee took exception to the RG 1.97 requirement for main control room indication of CCW flow,

which had previously been evaluated and found acceptable by tha NRC. Attemate indication is

provided by the local indicators and the plant computer.

Four local flow indicators are provided for CCW flow to the ESF systems (loops FT-95 through FT-98).

CCW heat exchangers 1 A and 1B outlet temperature indicators (loops TE-31 and TE-32), located in

the main control room, provide indication of CCW inlet temperature to the ESF systems. RG 1.97

identifies these instrument loops as Type D, Category 2 variables with a reliable power source. Based

on the team's revie'v of USAR Appendix 7A. design documents, and the as-built condition, both the

CCW ilow and temperature indicators are in accordance with the design bases.

E13.3.3 Conclusions

The instrumentation and controls design for the CCW system was considered adequate. All

instrumentation setpoints that were reviewed have adequate margin and the technical specification

limits were met.

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E1.3.4 Systeminterfaces

Gystem interfaces are reviewed in Section E1.2.4 of this report.

E1.3.5 System Walkdown

E1.3.5.1 Inspection Scope

The team conducted a walkdown of the CCW system and the plant areas, including the CCW pump

and heat exchanger room, surge tank room, control room and penetration area, but excluding

containment, that housed the CCW system. The team compared system configuration to the design

documents and the USAR and looked closely at equipment corxktion, area cleanliness, tagging, and ,

means used to avoid potential hazards such as missiles, fire, and pipe rupture.

E1.3.5.2 Observations and Findings

The team determned that the overall material condition of the plant areas was good. The equipment

sampled matched the c;4 sign requirements. No concoms were identified concoming configuration,

equipment cordtion or potential hazards. The System Engineer demonstrated a good knowledge of

the system and its components and exhibited good " system ownership."

,

l During the walkdown, the team noted that valves EG-V003, V313, V016 and HV059 were missing

l equipment identification tags. The licensee took prompt measures to correct this condition.

I

i

E1.3.5.3 Conclusions -

The team determned that generally the CCW system design observed during the walkdown was

consistent with the design basis requirements.

E1.4 Updated Final Safutv Analysis Report (USARI and Other Document Revews

E1.4.1 Inspection Scope

The team reviewed applicable USAR sections for the RHR, CCW, EDG, ESW, auxiliary building

HVAC, instrumentation, and electrical systems. The team also reviewed the system descriptions,

drawings, calcu'ation u)ntrol, and program plans concoming design basis and licensing basis issues.

- E1.4.2 Observations and Findings

a. USAR Review

- The team identified the following discrepancies in the USAR.

.

Different values were referenced for the RWST water volumes in the documents listed below.

PIR 97-4018 was initiated to addross the inconsistencies.

(1) USAR Section 6.3.2.2 (page 6.34) stated that the minimum RWST volume "available" or-

" assured" for ECCS injection mode operation is 394,000 gallons. Another paragraph in the same

USAR section refers to "usableNiume. However, TS 3/4.5.5 specified the 394,000 gallons as

the mi,1imum contained water volume; (2) USAR Table 6.3-1 listed 419,000 gallons as maximum

32

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. *.

a

volume,407,000 gallons as normal capacity, and 394,000 gallons as assured water volume.

These three RWST volumes are also shown in USAR Figure 6.3-7 and system descripJon,

M 10BN(Q), Figure 1; (3) USAR Table 6.2.1-5 listed RWST weter volume of 370,000 gallons for

containment analysis; and (4) USAR Table 6.310 listed 326,860 gallons as RWST volume for

ECCG cooling.

. NUREG-0881 (Wolf Creek), Section 0.3.2.1.2, refers to the same section in NUREG-0830

(Callaway) for a discussion of the NRC staffs position on battery capacity. That section of

NUREG-0830 states that the licensee revised the USAR in Revision 6 to stats that batteries are

sized in excess of the 50 percent margin required. Callaway USAR was revised, but the Wolf

C,reok's USAR has not been revised to reflect similar changes. (Refer Section E.1.2.2.12.b)

.

USAR Section 9.2.2.2.2 states, "The normally closed parallel sets of containment isolation valves

will allow the operator to establish cooling water to the reactor coolant pumps and the excess

letdown heat exchanger under emergency conditions, with a single failure." However, USAR Table

3-11(B)-3 listed the motor operators for these valves as category C, EQ not required. The

currently installed n.otor operators are class RH, that is, environmentally qualified (reference

EQWP-Umitorque, Checklist 1, Supplement 15). The licensee initiated PIR 97-4126 to resolve

this item.

.

Calculation EG-06-W, "Compor ent Cooling Water System Calculation," Revision W-3,

determined that the CCW heat exchanger heat transfer coefficient was 190 Stu/hr-ft 2-F based

on the revised ESW flow of 7150 gpm. The USAR stated that the transfer coefficient was 193

Blu/hr-ft2 p,

.

Calculation SA-89-017. " Evaluation of CCW & RHR Heat Exchanger Performance for the

Extended Fuel Operating Cycle (18 Months)," Revision 0, dctormined that CCW temperature

reaches 126 *F. Howeve:, USAR Table 9.2-11 is based . a CCW temperature of 120 F. The

, licensee initiated PIR 97-4052 to resolve this item.

.

USAR Fig. 6.4-8 showed suction for RHR pumps A and B as coming from RCS hot leg loop 4,

whereas system description M-10EJ(Q) and P&lD M-12EJ01 showed loop 1 for pump A and Loop

4 for pump B. The licensee initiated PIR 97-3823 to resolve this item.

.

USAR Section 6.3.5.3, " Flow Indication," stated that the flow from each RHR subsystem to the

RCS cold legs is recorded in the main control room. This contradicted USAR Table 7.5-1 and

P&lD M-12JE01 (Loop FT-988), which showed this parameter as being indicated (instead of

recorded)in the main control room. The licensee issued PIR 97-4179 to update the USAR.

USAR Table 7A-3 showed a range of 0-60 psig, whereas, the actua? :nstalled range was 0-69

psig. PIR 98-0062 was issued to update the USAR.

.

USAR pages 6.3-6,9.2-43,9.2-45, and 9.2-48 incorrectly described the control function of the

hWST auxiliary steam heating system with respect to winterization procedure STN GP-001.

USAR Change Request Log No.97-044 was issued to update the USAR.

.

USAR Section 8.3.2.1.2 stated that a Class 1E battery is to supply the loads in Tables 8.3-2 and

8.3-3 for 200 minutes where it should actually be for 240 minutes.

33

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.

.

.

USAR Sechon 7,4.1 states that the RWST level transmitters nre required for safe shu'down.

However, Table 3.11(b)3 does not list these transmitters as required for hot or cold shutdown.

Tne above discrepancies had not been corrected and the USAR updated to assure that the

information included in the USAR contained the latest material as required by 10 CFR 50.71(e).

(Unresolved item 50-482/97 20121)

b. System Description Review

The following system description discrepancies were identified:

  • USAR Section 5.4.7.2.1 stated that no motor-operated valves (MOVs) in the RHR system are

sut' ject to flooding. The RHR system description, M10-EJ(Q), stated that the only MOVs subject

to flooding were the suction isolation valves. The suction ; solation valves were above the 2007'-

11" elev? tion in containment, whereas, as per USAR Section 6.3.2.2, the maximum flood level in

containment post-LOCA was 2004'-6." The licensee issued PIR 97-3782 to revise the system

description.

.

RHR system description, M10-EJ(Q), stated that, "A leaktight sealis provided so that neither the

pressure vessel nor the guad pipe is connected directly to the sump or containment atmosphere."

This statement contradicted drawings M-109A-00015, M 03EJ05 and C-1L2311. These drawing.,

l

indicated that the guard pipe is connected to the containment sump and that

'

the outer diameter of the Ope does come in contact with the containment Sump as it enters the

sump. The licensee issued PIR 97-3805 to revise the system description.

Chemistry Specification Manual, AP 02-003, Revision 6, gave difforent chemistry parameters than

the CCW system description M 10EG(Q), Revision 2. Tho licensee initiated PIR 97-3466 to

resolve this item.

.

The USAR stated that one of the safety design bases of the CCW system was to provide heat to

maintain the ESW inlet trash racks from being blocked with frazilice. This safety design bas.s

was not discussed in system description M-10EG(Q), Revision 2. The changes to the USAR were

made as part of DCP 06349, which apparently overtooked the CCW system description. The

licensee initiated PIR 97-3885 to resolve this item.

.

Load profiles in the Class 1E battery system description did not agree with those in calculation NK.

E-002 for the Class 1E batteries; in addition, incorrect values were stated in the system

descript."n for Class 1E batteries' minimum voltage, amp-hour rating, etc. The licensee initiated

PIR 97-4190 to resolve this item.

.

USAR Tables 9.2 9,10, and i1 and CCW system description M-10EG(Q), Revisio,12, Table 2 did

not agree on total flow and heat load duty. The licensee initiatec' PlR 97-3983 to resolve this item.

c. Drawing Review

The team identified the following discrepancies:

.

Drawing M-12EJ01 showed loop FT618 low flow alarm with a PAL (low pressure alarm)

designation, which appeared to be in error. This et JIicted with drawing E-03EJ12 which shows

34

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, reia.

.

FAL(low flow alarm). The noensee concurred that M 12EJ01 is irk error. PIR 97-3824 was written

to update the drawing

. Drawinga J-110-31g and 320 for the RWST temperature monst. ing system, indicated

connections to the control room temperature irdcators with ungrounded shield or with no shield.

These connechons were not consistent with the boonsee's standard winng design for

instrumentation. The hoensee determined that the as t,uilt winng to the instruments were

shielded sind that the drav.ings were in error. PIR 96-0063 was. issued to update the drawingrs

d. Calculation Control Review

Wolf Creek inspechon activities i,wolved the review of over 130 calculations. The inspection team

observed that there were many active calculations for each system. Some of the calculations

reviewed did not ferm a part of the design basis, malung the design basis difficult to identify in some

cases. Further, severa! casce. eons for each system had similar purposes and/or similar results.

Although hose calculations wu not found to be contradictory, the prachce tended to further confuse

identification of the design basit Finally, the licensee apparently preferred supplementing existing

caiculations with new calculatic is to superseding, archivhg, or reviting existing calculations. This

,

practice increased the populaun of active calculations. The inspecdon team was concemed that the

'

current situation could cause nadvertent use of non-design-basis data in future design chynge or

analysis activities.

i

e. Design Basis / Ucensing Basis (DB/LB) Review Program Review

The inspection team noted that in early 1997 Wc7 Creek staff established a design basis / licensing

basis (DB/LB) review program to address the types of concems ideriified by this inspection.

Examples of the seven initiatives embodied in the licensee's DB/LB program are a USAR 'Tidelity"

- review project, a plan for penodic safety system functional self-assessments (the first of which was

conducted on the essential service water system during the first half of 1997), and the establishment

of an Engineering information System. The latter was expected to help track design basis information.

During an interview with the program manager for the USAR fidehty review effort, the team noted that

~

the licensee was identifying mmy discrepancies and was taking required actions to update the USAR.

While this DB!LB program had not produced widespread or consistent results at the time of the

inspection, the initiatives were highly encouraging. The scope of this inspection did not inclede an

__ effectiveness review of the DB/LB program.

E1.4.3 Conclusions

The team concluded that in severa! instances the USAR was not revised in accordance with 10 CFR

50.71(e) requirements. Discrepancies were also identified in system descriptions, drawings, and the

identi5 cation of design basis data in calculations.

XI Exit Meeting Gummary

After completing the onsite inspechon, the team conducted an exit meetin( with the licensee on January 9,

1998. During the meeting the team presented the results of the inspection. A list of persons who attended

the exit meeting is contained in Appendix B. Propnetary material was reviewed during this inspection but

this report contains no proprietary information.

t

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APPENDIX A

OPEN ITEMS

This report categorizes the inspection findings as unresolved items and inspection follow-up items in

c:cordance with NRC Inspection Manual, Manual Chapter 0610. An unresolved item (URI) is a matter

about which more in'oimation is required tu determine whether the issue in questior, is an acceptable iter.1,

a deviation, a nonconformance, or a violation. The NRC Region IV office will issue any enforcement action

resulting from their review of the identified URls. An inspection followup item (IFl) is a matter that requires

further inspection because of a potential problem, because specific licensee or NRC action is pending, or

because additional information is needed that was not available at the time of the inspection. The URIs

and OlFis found in this inspection are listed below;

Item Number Finding Title

LP_e

50-482/97-201-01 URI Cooldown Ar.alysis (Section E1.2.1.2(a))

50-482/97-201-02 IFl ECCS Leakage (Section E1.2.1.2(d))

50-482/97-201 03 URI RHR Pump Operation in Minimum Recirculation Mode

(Section E1.2.1.2(h))

r

l 50-482/97-201-04 URI Motor Control Center Circuit Length (Section E1.2.2.2.1(d))

50-482/97-201-05 151 120 Vac Short Circuit and Voltage Drop Analysis (Section

E1.2.2.2.1(d))

50-482/97-201-06 IFl Procurement of EDG Relay (Section E1.2.2.2.1(e))

50-482/97-201-07 IFl N*tery Load Profile (Section E1.2.2.2.2(b))

50-482/97-201-08 iFi TS Change for Batteries (Section E1.2.2.2.2(b))

50-482/97 201-09 IFl Battery Sizing (Section E1.2.2.2.2(b))

50-482/97-201-10 URI DC Voltage Drop Calculation (Section E1.2.2.2.2(d))

50-482/97-201-11 URI Minimum Battery Voltage (Section E1.2.2.2.2(d))

50-482/97-201-12 URI Load Growth Control (Section E1.2.2.2.2(e))

>

504c2/97-201-13 URI Acceptance Criteria for Battery Test (Section E1.2.2.2.2(f))

50-482/97-201-14 URI Corrective Action for Battery Test (Section E1.2.2.2.2(f))

50-482/97-201-15 URI RWST Level Instrumentation (Section E1.2.3.2(a))

50-482/97-201 16 URI Seismic Qualificetion (Section Ei.2.3.2(b))

A-1

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.

.. .. .

, as e

.

50-482/97-201 17 URI NPro9en Bottle Installation (Section E1.2.5.2(d))

50-482/97-201 18 URI MOV Differential Pressure (Section E1.3.1.2(c))

, 50 482/97-201-19- URI - CCW Low Temperature (Section E1.31.2(e))

50 482/97-201 20 URI Corrective Action for CCW Operating Procedure (Section

E1.3.1.2 (f))

50-482/97 201-21 URI USAR Discrepancies (Section E1.4.2(a))

A-2

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APPENDIX B

EXIT MEETING ATTENDEES

Wolf Creek Nuclear Operatina Corporation

O. Maynard, President and CEO

C. Warren, Vice President , Operations / COO

R. Muench, Vice President, Engineering

T. Garrett, Manager, Design Engineering

C. Younie, Manager, Opeiatioris

R. Sims, Manager, Systems Engineering

M. Angus, Manager, Licensing and Corrective Action

W. Norton, Manager, Performance Impronment and Assessment

C. Fowler, Manager, Integrated Plant Scheduling

N. Hoadley, Manager, Support Engineering

R. Flannigan, Manager, Nuclear Safety and Licensing

R. R. Osterrider, Supervisor, Safety Analysis

B. Smith, Lead Engineer, Design Engineering

J, Yunk, Senior Engineering Specialist

J. Stamm, Supervisor, Safety Analysis

R. Rietmann, System Engineer

M. Blow, Superintendent, Chemistry

T. Damashek, Supervisor, Licensing

R. Holloway, Project Engineer

B. Masters, System Engineer

L. Solorio, Design Engineer

W. Eales, Design Engineer

W. Selbe, Project Engineer

M. Guyer, Operations

U S. Nuclear Reaulatori Commission

R. Mathew, Team Leader, NRR

S. Richards, Chief, PECB, NRR

T. Stetka, Acting Chief, EB, RIV

W. Johnson, Chief, Project Branch, RIV

F. Ringwald, SRI, NRC

K. Neubauer, Contractor, S&L

M. Sanwarwalla, Contractor, S&L

A. Rahman, Contractor, S&L

G. Bizarra, Contractor, S&L

R. Sheldon, Contractor, S&L

B-1

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APPENDIX C

l.lST OF ACRONYMS

,

AC, ac Altemating Current

AMPS Amperes

AWG American Wire Gauge

BTU British Thermal Unit

CCW Component Cooling Water

CFR Code of Federal Regulations

CPT Control Power Transformer

CS Containment Spray

-Cv Valve Flow Coefficient *

DB- Design Basis

DC, de Direct Current

DCP Design Change Package

DP Differential Pressure

ECCS Emergency Core Cooling System

EDG Emergency Diesel Generator

EOP Emergency Operating Procedure

EPA Electrical Penetration Assemblies

EQ Environmental Qualification

ESF Engineered Safety Features

ESW Essential Service Water

"

Fahrenheit

FE Flow Element

FLA Full Load A:aperes

ft, FT Feet or Foot

gal Gallons

f GL Generic Letter

gpm- Gallons Per Minute

HVAC Heating, Ventilating, and Air Conditioning

l&C Instrumentation and Controls

IEEE. Institute of Electrical and Electronics Engineers inc,

IFl . Inspection Follow-up Item

IN Information Notice

IST Inservice Testing

ITIP Industry Technical Information Program

kVA Kilovcit-Ampere

kV Kilovolt

kW Kilowatt

LB Licensing Basis

LC Locked Rotor Current

LCO Umiting Condition for tpaiation

LER Licensee Event Report

LLRT- Local Leak Rate Test

LOCA Loss-of-Coolant Accident

LOOP Loss-of-Offsite Power

LT LevelTransmitter

MCC Motor Control Center

MOV Motor Operated Valve

.

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NPSH Net Positive Suction Head

NRC Nudear Regulatory Commission

NRR Nuclear Reactor Regulation, Office of (NRC)

NSSS Nuclear Steam Supply System

P&lD Piping & Instnamentation Diagram

Pl Pressure Indicator

PIR Performance improvement Request

PMR Proposed Modification Request

ppm, PPM Parts Per Million

PS8 Power System Branch

psi, PSI Pounds per Square inch

psia, PSIA Pounds per Square Inch Absolute

psid, PSID Pounds per Square Inch Differential

psig, PSIG Pounds per Square Inch Gauge

RC Reactor Coolant

RCP Reactor Ccolant Pump

RCS Reactor Coolant System 3

REV

-

Revision

RG Regulatory Guide

RHR Residual Heat Removal

RWST Refueling Water Storage Tank

SBO Station Blackout

S&L Sargent & Lundy

SER Safety Evaluation Report

-

SFo Spent Fuel Pool

St Safety injection

SIS Safety injection Signal

SNUPPS Standardized Nuclear Unit Power Plant System

STP Sumeillance Test Procedure

SW Service Water

TS, Tech. Spec. Technical Specifications

USAR Updated Safety Ana$ysis Report

UHS Ultimate Heat Sink

URI Unresolved item

USQ Unreviewed Safety Question

USQD Unreviewed Safety Question Determination

Vdc Volts DC

Vac Voits AC

1 W Watts

WCAP Westinghouse Containment Anai> sis Program

WCGS Wolf Creek Generating Station

WCNOC Wolf Creek Nuclear Operating Corporation

c-2

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