ML20203H534
| ML20203H534 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| 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
0 PDR
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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
9
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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.
11
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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
16
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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
--
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
20
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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.
'
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,
j
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
_
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
{ .
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
t.
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
i
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
l
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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
I
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.
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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:
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
-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
RCP Reactor Ccolant Pump
REV
-
Revision
RG Regulatory Guide
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
TS, Tech. Spec. Technical Specifications
USAR Updated Safety Ana$ysis Report
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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