ML20137Z506

Washington Nuclear Plant-2 Annual Operating Rept 1985

ML20137Z506
Person / Time
Site:	Columbia
Issue date:	12/31/1985
From:	Powers C WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
NUDOCS 8603130058
Download: ML20137Z506 (54)

Text

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WXP-2 AXXUAL io OPERATING l

REPORT

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0 lO ANNUAL REPORT OF 0

WNP-2 FOR 1985 0

DOCKET 50-397 LICENSE NPF-21

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Washington Public Power Supply Syster.1 0~

3000 George Washington Way Richland, Washington 99352 O

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TABLE OF CONTENTS JC) 1.

INTRODUCTION.....................

1 - 1 A.

1985 Power History Graph............

1-4 l

2.

REPORTS.......................

2-1

O A.

Annual Personnel Exposure and Monitoring Report.

2 - 2 B.

Relief Valve Challenges.............

2 - 3 C.

Summary of Plant Operation...........

2-5

+

-(/

i D.

Summary of Significant Safety Related Maintenance Performed

............2-15 E.

Indications of Failed Fuel 2 - 30

()

F.

Plant Changes and Tests.............

2 - 31 1.

Plant Changes...............

2 - 32 A.

Design Changes............

2 - 32 iC)

B.

Procedure Changes...........

2 - 45 1

2.

Plant Tests................

2 - 46 1

4

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C) f D

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1. INTRODUCTION O

The Annual Report of the Washington Public Power Supply System's Plant Number 2 (WNP-2) for 1985 is submitted in accordance with the requirements of Federal Regulations and Facility Operating License NPF-21 and as a supplement to the Monthly Operation Report.

Other required reports are included for ease of reference and completeness.

O WNP-2, a 3323 MWT BWR/5, began commerical operation on December 13, 1984. The plant was base loaded at or near 100 percent power for the first three months of 1985.

In early April, power was reduced to 50 percent due to excessive vibration of Reactor Recirculation Pump B.

A two month maintenance outage began in May.

Originally the outage was planned to start in April, but was o

delayed due to regional generation requirements.

In July, after the mainten-ance outage, the plant was restarted and ran in single loop operation mode (Recirculation Pump A only) at or.near 72 percent power for the remainder of the year. Refer to Figure 1-2 for 1985 Power History Graph.

The year was punctuated by a series of major accomplishments requiring sig-O nificant effort on the part of the plant staff to successfully complete them.

A brief sumary of the major items gives an indication of the variety and depth of these activities:

o A significant reduction was accomplished in total hours of water chem-istry falling outside of Technical Specification Limits.

In 1984, the t tal number of hours was 219.95, as compared to 8.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> in 1985. This O

reduction was accomplished due to attention on the part of Plant Manage-ment and the dedication of necessary resource personnel, o

WNP-2 successfully completed i ts first 100 days of consecutive power operation.

The plant ran continuously from August 4 to November 13, with an average power level of approximately 72 percent limited by single loop g

operation, o

WNP-2 personnel completed the year with an excellent safety record.

For every 200,000 manhours worked, there were 4.76 injuries reported, as com-pared to the electric utility industry average of 5.66.

O WNP-2 reduced the number of Licensee Event Reports (LER's) generated this o

year.

A total of 63 LER's were written in 1985 as compared to 131 in 1984 o

WNP-2 successfully completed its first major maintenance outage.

Major maintenance activities of the outage included:

Reactor Recirculation "B" O

seal and bearing replacement; installation of new design seals in both Reactor Feed Pumps; a design change to the feedwater heater level control system-successful performance of 42 Logic System Function / Technical Surveiilance procedures; successful completion of the Local Leak Ra te Testing (LLRT) program as required by Appendix J of 10CFR50; and turbine work following the initial 100*. run included removal of startup strainers O

from the throttle valves, inspections of the bearing oil system, moisture separator reheater internals, turbine by-pass valves and resolution of the jack shaft vibration problem.

1 -1 0

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A number of significant problems were resolved during the year in which the potential implication relative to nuclear safety was the overriding factor in O

selecting the course of action.

Some of these activities carried a signiff-cant price tag.

Certain decisions directly impacted generation output, others levied their cost by adding complications to the already routinely complex task of generating power in a safe effective manner.

Some major examples of these events include:

O o

Reactor Recirculation (RRC) Vibration Problems - Due to unusually high vibration levels emanating from RRC-P-18, the Recirculation System was at first operated with both RRC pumps at low speed, ifmiting plant power to 57%, and then operated in single loop with RRC-P-1A only, at high speed, ifmiting plant power to 72%.

The desire to minimize the potential for small bore line breaks, creating a small LOCA, was the primary motivation O

for the decision to Itmit plant power level.

This decision was also predicated and consequently resulted in minimizing further equipment degradation.

o Containment Atmospheric Control (CAC) Isolation Valve Elastomer Seal Replacement - Seals for these valves were replaced based on the determin-O ation that quali fied lifetime during accident conditions did not meet design requirements.

Although no visual evidence necessitated replace-ment, Plant Management conservatively chose to assure reliability by replacement of the seals with qualified new parts.

o RWCU Pumo Suction Subcooling Modification - A subcooling line from the

'O nonregenerative heat exchanger outlet was added to the comon RWCU suc.

tion line to increase net positive suction head available to the pumps.

This significantly lengthened pump seal life and consequently greatly reduced radiation dosage previously received during seal maintenance, 1

o Plant Service Water (TSW) Modification to RCC - The supply to the suction O

of the TSW pumps was modified to receive cool water directly from the Columbia River in an effort to decrease the temperature of the Reactor Closed Cooling Water System used to cool the drywell atmosphere.

The motivation for the decision to modify this system was the desire to decrease average drywell temperature in an effort to maintain the average atmosphere temperatures within the limits assumed by the containment O

pressurization analysis during a LOCA event and to maintain temperatures within the equipment qualification limits.

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.O 1-2 0

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The actual and adjusted capacity factors for the year are listed in the following table.

Adjusted capacity factors are based on the limitation of power output caused by reactor recirculation pump vibration problems.

MONTH CAPACITY FACTOR ANUSTED CAPACITY FACTOR January 73.6%

73.6%

February 60.0%

60.0%

March

81. 0".

81.0%

)

April 52.4%

75.2%

May

• 3.6%

6. 3".

June **

0.3%

0.53%

)

J ul y ***

45.4%

72.8%

August 67.6%

93.9%

September 70.7%

98.2%

)

October 70.3%

97.7%

November 56.3%

78.2%

December 65.8%

91.4 %

4 OVERALL 53.92%

69.06%

Started Maintenance Outage May 3, 1985.

Erided Maintenance Outage June 29, 1085.

• • • Entered Single Loop Operation.

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2. REPORTS The reports provided in this document meet the requirements of Federal Regulations (10 CFR 50.59) and the WNP-2 Operating License.

Complete data for the year 1985 has been included, i

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a 2.8 MAIN STEAM LINE SAFETY / RELIEF VALVE CHALLENGES This section contains information concerning relief valve and safety valve challenges for calendar year 1985 in accordance with the requirements of NUREG 0737. Item II.K.3.3.

TYPE OF PLANT REASON FOR ACTUATION CON 0! TION ACTUATION RX POWER ASSOCIATED DATE COMPONENT 10 (CODE)

(CODE)

(CODE)

LEVEL LER 01/31/85 MS-RV-lC A

E A

1001 85-007 01/31 /85 MS-RV-58 8

0 E

0%

01/31/85 MS-RV-1A B

D E

01 01 /31/85 MS-RV-2C B

D E

0%

01/31/85 MS-RV-40 B

D E

0%

01 /31/85 MS-RV-4B B

0 E

0%

01 / 31 /85 MS-RV-4A B

0 E

0%

01 /31 /85 MS-RV-4C B

D E

0%

01 /31/85 MS-RV-48 8

0 E

01 01/31/85 MS-RV-3C B

D E

0%

01 /31 /85 MS-RV-4C B

0 E

0%

01 /31 /85 MS-RV-20 B

0 E

0%

01/31/85 MS-RV-2A B

0 E

0%

01 /31 /85 MS-RV-5B B

G E

0%

01/31/85 MS-RV-30 B

G E

0%

01/31/85 MS-RV-SC B

G E

0%

01 /31 /85 MS-RV-48 8

G E

0%

01/31/85 MS-RV-3B B

G E

0%

Manual actuation following a reactor scram.

05/03/85 fiS-RV-1 A 3

E C

15/%

05/03/85 MS-RV-10 B

E C

15/1 05/03/85 MS-RV-1C B

E C

15/%

05/03/85 MS-RV-4C B

E C

15/1 05/03/85 ftS-RV-5C B

E C

15/1 05/03/85 MS-RV-40 B

E C

15/1 05/03/85 MS-RV-20 B

E C

15/1 05/03/85 MS-RV-2C B

E C

15/1 05/03/85 MS-RV-1B B

E C

15/5 2-3

MAIN STEAM LINE SAFETY / RELIEF VALVE CHALLENGES (continued)

TYPE OF PLANT REASON FOR ACTUATION CONDITION ACTUATION RX POWER ASSOCIATED DATE COMPONENT ID (CODE)

(CODE)

(CODE)

LEVEL LER 05/03/85 MS-RV-4B B

E C

15/%

05/03/85 MS-RV-4A B

E C

15/%

05/03/85 MS-RV-5B B

E C

15/%

05/03/85 MS-RV-30 B

E C

15/%

05/03/85 MS-RV-2A B

E C

15/%

05/03/85 MS-RV-3A B

E C

15/%

05/03/85 MS-RV-2B B

E C

15/1 05/03/85 MS-RV-3C B

E C

15/5 05/03/85 MS-RV-3B B

E C

15/5 Performed required Technical Specification surveillance on acoustic monitors.

06/30/85 MS-RV-1A B

E C

24/%

06/30/85 MS-RY-SC B

E C

24/5 06/30/85 MS-RV-20 B

E C

24/%

Performed required Technical Specification surveillance on acoustic monitors following repair.

CODES:

Tyce of Actuation A.

Automatic B.

Remote Manual C.

Spring Plant Condition A.

Construction B.

Startup or Power Ascension Tests in Progress C.

Routine Startup D.

Routine Shutdown E.

Steady State Operation F.

Load Changes During Routine Operation G.

Shutdown (Hot or Cold)

H.

Refueling Reason For Actuation A.

Overpressure B.

ADS or Other Safety C.

Test 0.

Inadvertent (Accidental / Spurious)

E.

Manual Relief 2-4

2.C. SimetARY OF PLANT OPERATION INCLUDING UNIT SHUTDOWNS / POWER REDUCTIONS GENERATOR OUTAGE OFF-LINE CAUSE SHUTDOWN LER DATE TYPE HOURS CODE METHOD NUMBER SYSTEM COW ONENT CAUSE AND ACTION TO PREVENT RECURRENCE 01/01/85 F

22.88 E

3 85-002 EA TRANSF Failure of a potential transformer in the 500 KV substation in the proximity of gen-erator breakers resulted in operation of ground 0.C. relays which tripped the main generator and caused a reactor scram from 100% power.

Time elements on the group relays were extended from.25 seconds to

.5 seconds.

Other corrective actions are to be implemented as conditions permit.

.y u.

01/17/85 F

58.33 G

3 85-006 EB CKTBRK A scram occurred at 1001 power due to an accidental trip of a feeder breaker and the resultant loss of "A" RPS Bus while a half scram existed on the opposite RPS channel during troubleshooting activities.

01/17/85 F

Not on Line G 3

85-005 IA INSTRU During surveillance testing of APRM channel "A"

prior to a reactor startup, APRM channel "B"

was inadvertently placed in "Staney" causing a full RPS trip.

01/25/65 F

43.95 E

3 85-003 EA RELAYXX Generator trip from 100% power was attri-buted to perturbations in the microwave link between generator and 500 KV gener-ator breaker which gave a signal that the generator breaker was open when, in actu-ality, it remained closed.

The signal was subsequently shifted to a more reliable microwave channel.

l

1 SUtttARY OF PLANT OPERATION INCLUDING UNIT SHUTDOWNS / POWER REDUCTIONS (Continued)

GENERATOR OUTAGE OFF-LINE CAUSE SHUTDOWN LER DATE TYPE HOURS _

CODE METHOD NUMBER SYSTEM COPPONENT CAUSE AND ACTION TO PREVENT RECURRENCE 01/31/85 F

135.49 H

3 85-007 EB RELAYXX Generator tripped from 100% power as a result of failure of a unit lockout relay with the resultant tripping scheme causing a partial loss of off-site power.

The cause of trip was investigated, trans-formers were tested and modifications to relays were implemented to prevent similar future occurrences.

? 02/03/85 F

Not on Line H 1

85-011 CF PIPEXX During a routine inspection, two 3/4" lines CH were found to have leakage.

These 3/4" drain lines formed part of the primary system pressure boundary.

The reactor was shutdown and the lines were repaired per ASE requirements and returned to service.

02/06/85 S

0.42 B

1 N/A N/A N/A Tripped main generator via unit emergency trip to verify revision of relaying logic.

02/13/85 F

107.15 G

3 85-016 CA INSTRU Perturbation in RPV water level reference and variable legs as a result of an error while performing a Tech Spec instrument surveillance on a level transmitter re-suited in a reactor scram at 100% power.

The plant was returned to service after a4 day delay due to other unrelated maintenance activities.

SlMtARY OF PLANT OPERATION INCLUDING UNIT SHUTDOWNS / POWER REDUCTIONS (Continued)

GENERATOR 00TA2E OFF-LINE CAUSE SHUTDOWN LER DATE TYPE HOURS CODE METHOD NUMBER SYSTEM COW ONENT CAUSE AND ACTION TO PREVENT RECURRENCE 02/14/85 F

Not on Line G 3

85-014 IA INSTRU During plant startup, a high pressure reactor scram was initiated due to inadvertent manual actuation of the turbine bypass valve 7" Hg low condenser vacuum interlock.

This closed the bypass valves (BPV) and switched the turbine to "BPV Manual".

This resulted in loss of Reactor Pressure Vessel (RPV) pressure control and caused a RPS actuation due to High RPV pressure.

y 03/22/85 F

102.9 A

3 85-024 HA INSTRU Reactor scrammed at 100% power when plac-ing the turbine DEH control system in the monitoring mode.

Trouble shooting revealed:

1) A faulty computer memory board

2) A faulty EIS computer interface board

3) A faulty PROM logic card in the gover-nor valve circuit.

Items 1 4 2 were replaced and item 3 was repaired.

The DEH control system was then tested and the unit returned to service.

04/01/85 F

53.5 A

1 N/A CB PENETR Plant was shut down due to failure of a weld and rupture of the hydraulic actuator Ifne to "B"

recirc FCV.

The line was

(

repaired and plant returned to service.

04/10/b5 F

0.0 A

5 N/A CB PUWXX "B"

Recirculation pump /rmtor developed a high vibration problem which necessitated going to 15 Hz operation.

The resultant l

power reduction existed for remainder of the month.

StM4ARY OF PLANT OPERATION INCLUDING UNIT SHUTDOWNS / POWER REDUCTIONS (Continued)

GENERATOR OUTAGE OFF-LINE CAUSE SHUTDOWN LER DATE TYPE HOURS CODE METHOD NUMBER SYSTEM COW ONENT CAUSE AND ACTION TO PREVENT RECURRENCE 05/03/85 S

1364.2 H

1 N/A N/A N/A Plant shutdown as scheduled for annual maintenance outage (M3).

06/29/85 F

20.9 A

1 85-046 CH PUWXX Reactor was manually scrammed because of an oil fire on the outboard bearing of the "B"

Reactor Feedwater Pump (RFP) due to bearing failure.

The feedwater pump was repaired by replacing the entire rotating

element, pump bearf r.gs,

seals and by 7

performing extensive rework of the out-board bearing housing.

06/30/85 S

2.3 B

1 N/A HA TURBIN Generator was taken off-line for overspeed testing of turbine.

07/01/85 F

29.6 A

3 85-047 HA VALVEX Reactor scram during startup due to fail-ure of limit switches on main turbine throttle valves 1 and 4 to reset when valves were opened.

The switches were recalibrated and returned to service.

07/01/85 F

0 A

5 N/A CH PUWXX "B" RFW Pump out of service due to bearing failure with resultant pump damage which limited power output to a maximum of approximately 82%.

St# MARY OF PLANT OPERATION INCLUDING UNIT SHUTD0lalS/ POWER REDUCTIONS (Continued)

GENERATOR OUTAGE OFF-LINE CAUSE SHUTDOWN LER DATE TYPE HOURS CODE ETH00 NUfBER SYSTEM COW ONENT CAUSE AND ACTION TO PREVENT RECURRENCE 07/03/65 F

0 A

5 N/A CB PUWXX The "B"

RAC Pump failed due to motor current transformer failure and high vibration.

This limited power output to SOS until 7/19/85 when an Emergency Technical Specification change authorized increase power output during single loop operation as determined by core stability which pemitted approximately 705 to 721 power output.

The plant ran at a reduced power level for the remainder of the year.

y v

07/07/85 5

8.2 A

1 N/A C8 PUWXX Plant was shutdown because of a current transformer failure on "8"

RRC motor due to a failed termination.

Repairs were made and plant returned to service.

However, excessive vibration prevented operating pump on 60 Hz.

07/16/65 F

43.4 A

1 N/A CH VALVEXX Plant was shutdown due to excessive unidentified leakage in drywell.

The leakage was identified as a feeduater check valve hinge pin cover gasket.

Repairs were completed and the plant was returned to operation.

07/18/65 F

0.8 G

9 N/A HJ INSTRU Generator tripped during startup due to high MSR level.

Upon investigation it was detemined that the level controller was not in the proper mode.

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

i l

SLR94ARY OF PLANT OPERATION INCLUDING UNIT SHUTDOWNS / POWER REDUCTIONS (Continued) i GENERATOR OUTAGE OFF-LINE CAUSE SHUTDOWN LER

+

DATE TYPE HOURS CODE ETH00 NUMBER SYSTEM C(DFONENT CAUSE AND ACTION TO PREVENT RECURRENCE l

l 08/04/85 F

11.86 A

3 85-053 IA INSTRU p* actor scrammed at 715 power while per-t forming an instrument surveillance of ATWS RRC Pump Trip feature, due to a faulty i

instrument isolation valve. The valve was replaced.

?

11/13/85 F

112.37 A

3 85-059 EG INSTRU Turbine trip and reactor scram at 515 power.

RPV high water level, which resulted in the turbine trip, was caused I

m 1.

by failure of an inverter uhich caused a o

transient of the instrument power supply and blew a fuse in the fee &ater control circuitry.

The inverter was repaired and a design change initiated to revise the power supply circuitry for the fee &ater control system.

l t

11/17/85 F

Not on Line G 3

85-061 IA INSTRU During a reactor cold startup, with hot excess reactivity near maximum for the i

first fuel

cycle, high rod worth on l

control rod withdrawal caused a Hi Hi IRM trip and subsequent reactor scram.

12/06/85 5

13.63 8

1 N/A C8 POW XX Reactor power was reduced and generator taken off-line during the drywell entry for installing a power supply to RRC Pump l

"B" vibration test instrumentation.

t e

e 3__

7 --.

e-

-m

St#NutY OF CODES OUTAGE TYPE F-Forced S-Scheduled

~

CAUSE CODE A-Equipment failure 8-Maintenance or Test C-Refueling D-Regulatory Restriction

.- 2 E-External Cause F-Administration G-Personnel Error H-Other SHUTD06Al ETH00 1-haual 2-knual Scram 3-Auto Scram 4-Continued 5-Reduced Load 9-Other

SYSTEM CODE STANDARD CODE SYSTEM DESCRIPTION CA Reactor Vessels & Appurtenances C8 Coolant Recirculation Systems f. Controls CF Residual Heat Removal Systems a Controls CH Feedwater Systems & Controls IA Reactor Trip Systems EA Offsite Power Systems & Controls EB AC Onsite Power Systems & Controls EG Other Electric Power Systems a Controls HA Turbine Generator & Controls EU Other Features of Steam & Power Conversion Systems (not included elsewhere)

COW ONENT CODE Coff0NENT TYPE / CODE CON ONENT TYPE INCLUDES:

Circuit Closers / Interrupters Circuit 3reakers (CKTBRK)

Contactors Controllers Starters Switches (other than sensors)

Switchgear Instrumentation and Controls Controllers (INSTRU)

Sensors / Detectors / Elements Indicators y

Differentials C

Integrators (Totalizers)

Power Supplies Recorders Switches Transaitters Computation Modules Penetrations, Primary Containment Air Locks (PENETR)

Personnel Access Fuel Handling Equipment Access Electrical Instrument Line Process Piping Pipes Fittings Pipes (PIPEXX)

Fittings Pumps Pumps (PUWXX) 1

CopFONENT CODE (continued)

C0090NENT TYPE / CODE C0ft0NENT TYPE INCLUDES:

Relays Switchgear (RELAYXX) l Transformers Transformers (TRANSF) t Turbines Steam Turbines (TUR8IN)

Gas Turbines Hydro Turbines Valves Valves (VALVEX)

Dampers

{

t I

2.D. SIGNIFIC4NT MAINTENANCE PERFORE D ON SAFETY RELATED EQUIPE NT EQUIPENT REQUIRING MRINTEMANCE SYSTEM PROBLEM ACTION TAKEN MSLC-MD-1B Main Steam Valve would not operate.

Replaced clutch sleeve re-installed operator, teakage Control tested and returned to service.

E-IN-1 Uninterruptible A ground existed in inverter Replaced a capacitor and a power diode, Power Supply supply capacitor bank.

checked ground relay and ground indicating lights. Ground cleared, tested and returned to service.

y HPCS-MO-12 High Pressure Valve would not go full Replaced torque switch and inspected valve Core Spray closed.

internals. Valve was found to be mechanically bound. Valve worked properly af ter re-installation, tested and returned to service.

MS-PS-238 Main Steam Installed pressure switch Replaced pressure switch with direct over sensitive to replacement, recalibrated and returned pressure fluctuations.

to service.

RPS-PS-SC Reactor Switch would not actuate as Switch would not move until it was actuated Protection required on low pressure.

by hand. Re-tested after manital actuation System and actuated at praper setpoint.

I M3-RIS-610A Main Steam Radiation Monitor produced Replaced internal heater thermostat assembly, spikes up and down scale.

Performed radiological calibration and returned to service.

9 O

l SIGNIFICANT MAINTENANCE PERFORED ON SAFETY RELATED EQUIPENT (Continued)

EQUIPENT REQUIRING MAINTENANCE SYSTEM PROBLEM ACTION TAKEN DMA-M-FM/II Diesel Gen.

Fan did not auto start Replacement relay coil not immediately Bldg. HVAC when Diesel Generator #1 available. Modified plant procedure to l

is running.

manually start fan if DG #1 runs more than I

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> while awaiting parts to repair /

replace relay.

DMA-FN-11 Diesel Gen.

Fans failed to auto start Found blown fuses in timer control circuit.

DEA-FN-11 Bldg. HVAC when DG output breaker was Replaced fuses, ran operability test and closed.

returned to service.

~*

HPCS-tO-12 High Pressure Valve failed to open as re-Replaced sheared drive key, set torque switch, Core Spray quired during operability test. tested, and returned to service.

RHR-FT-15B Residual Heat Flow transmitter not Replaced sensing module, calibrated instrument Renoval operational.

and returned to service, i

RHR-MO-47A Residual Heat Motor operator blew fuses Replaced fuses and torque switch. Reset i

Removal at the motor control torque and limit switches, tested and returned l

center.

to service, i

1 l

f l

DLO-P-3A2 Diesel Lube Pump did not provide Replaced pump and check valve DLO-V-14A2.

l Oil significant flow to main-Re-aligned pump / motor, tested and returned tain lube oil lines full to service.

l l

or lube oil at the proper l

temperatures.

l l

_g

--,-w--

SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continued)

EQUIPMENT REQUIRING MAINTENANCE SYSTEM PROBLEM ACTION TAKEN SGT-EllC-1 A1 Standby Gas lleating coil burned out.

Replaced bad heating coil, tested and returned Treatment to service.

RCIC-P-1 Reactor Core Vibration readings found to Disassembled and replaced inboard and outboard Isolation Cooling be in the alert range.

bearings. Tested and returned to service.

LPCS-PIS-S Low Pressure Alarm present with system Replaced high side switch and recalibrated the Core Spray pressure in normal range.

instrument, tested and returned to service.

ro L

LD-TS-608B Leak Detection Switch found to be Replaced defective alarm module, calibrated, inoperable.

tested and returned to service.

CAC-Ell 0-FCV/4A Containment Valve operator required Removed existing spring and replaced with Atmospheric spring replacement in order the recommended replacement spring per ITT Control to comply with OER 80-048 General Controls. Verified stroke and proper and manufacturer's closing force, returned to service.

recommendations.

DSA-PS-34 Diesel Starting Pressure switch failed, Installed and calibrated a new pressure switch.

Air would not turn off air Functionally tested and returned to service.

compressor on high pressure.

MS-RIS-610D Main Steam MS Line Radiation monitor Replaced logarithmic feedback element, per-found out of tolerance formed radiological calibration and returned during source check.

to service.

SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continued)

EQUIPMENT REQUIRING imINTENANCE SYSTEM PROBLEM ACTION TAKEN llS-RIS-610D Main Steam MS Line Radiation monitor was Installed new detector which had been found to be defective.

modified with a water proofing kit. Performed resistance cable checks and Health Physics source checks. Returned to service.

RRC-P-1B Reactor Pump showed signs of excessive Replaced pump radial bearing and seal assembly.

Recirculation vibration.

(For detailed information on maintenance activities and tests performed, see the last page of this section.)

(O

'.o RFW-AO-328 Reactor Valve actuator would not Honed cylinder, replaced seal tube end, Feedwater stroke properly.

"0" Ring, seal kit and seal piston per manufacturer's instructions. Tested and returned to service.

SW-V-212 Standby Service Valve would not close upon A.C. feed to rectifier burnt causing loss Water demand.

o' continuity. Replaced full wave bridge rectifier in solenoid valve, tested and returned to service.

CRD-V-111/2631 Control Rod Leaking valves resulted Replaced with valves which had been rebuilt, CRD-V-lll/4655 Drive in accumulators being unable tested and returned to service.

to hold N2 pressure.

SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continued) l EQUIPENT REQUIRING

(

MAINTENANCE SYSTEM PROBLEM ACTION TAKEN HS-POE-1A Main Steam Acoustic Monitoring Channel Replaced valve flow monitor card and replaced for MS-RV-1A did not respond broken connector for sensor. Tested and during surveillance testing.

returned to service.

DilA-TIC-32/2 Diesel Mixed Terperature indicating con-Replaced broken instrument, calibrated, tested Air troller would not calibrate, and returned to service.

l l

l REA-DPIC-1B Reactor Bldg.

When REA-FN-1B is running Calibrated all instruments which provide input l

?

HVAC and REA-DPIC-1B is in the to REA-DPIC-18. Found wires reversed on 3

auto position, the reactor REA-E/P-1B and mechanical dampers partially building overpressurizes.

closed. Returned system to designed configur-ation, tested and returned to service.

MS-LR-615 Main Steam Level recorder will not Replaced burned amplifier, replaced inoperable respond.

servo motor, calibrated and returned to service.

LD-TRS-611 Leak Detection RilR Room differential temper-Replaced two selector switch boards, cleaned ature points alarm spuriously.

slide wires, adjusted gain, calibrated and returned to service.

SW-V-212 Standby Service Solenoid valve not closing Removed and replaced rectifier.

Water as required per surveillance Tested and returned to service, procedure.

SIGNIFICANT llAlllTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continued)

EQUIPMENT REQUIRING MAINTENANCE SY STEM PROBLEM ACTION TAKEN DEA-M-FN/32 Diesel Gen.

Fan would start but wouldn't Adjusted the setting on the under voltage Bldg. IIVAC run after switch was taken relay to allow relay to seal in.

to the auto position.

RilR-PS-39A Residual lleat Wrong pressure switch was Replaced presently installed pressure switch RilR-P S-39B Removal installed while implementing with the correct model per the design change a design change package.

package. Calibrated and returned to service.

CSP-SPV-78 Cont. Supply Air blowing from the regu-Replaced regulator for solenoid valve and 8

Purge lator, possible diaphram returned to service.

problem.

SGT-ESil-2A Standby Gas Strip heater appeared to be Replaced defective heaters and defective SGT-ESil-2B Treatment burned out.

fuses. Tested and returned to service.

OG-RIS-601A Off Gas Radiation indicating switch Replaced preamplifier and all three G-M tubes.

had spiking problem.

Performed radiological calibration and returned to service.

LD-TS-608B Leak Detection Switch found to be Replaced defective alam module, calibrated inoperable.

and returned to service.

l l

SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continued)

EQUIPENT REQUIRING MAINTENANCE SYSTEM PROBLEM ACTION TAKEN CAC-Ell 0-FCV-4A Containment Valve operator required Removed existing spring and replaced with the Atmospheric spring replacement in recomunended replacement spring per ITT General Control order to comply with Control s.

Verified stroke and proper closing

)

OER 80-048 and manufacturer's force, returned to service.

recommendations.

REA-DPIC-1B Reactor Bldg.

Controller did not aoe-Replac.ed associated solenoid valve, tested and HVAC quately control reactor returned to service.

l building pressure.

l ra CAC-LT-1A Containment Failed to meet tolerances as Replaced linearity resistor, calibrated and re-Atmospheric specified on surveillance turned to service.

Control procedure.

j OG-RIS-601B Off Gas Random spiking on Off Gas Replaced three G-M tubes and detector / preamp radiation indicating switch.

assembly. Performed radiological calibration and returned to service.

i CMS-LR/PR-4 Containment Suppression Pool Level Indi-Actual level had remained constant. Replaced Monitoring cation has dropped 4 inches.

servo amplifier board, calibrated and returned to service.

DLO-P-2Al Diesel Lube D.C. Lube Oil Pumps needed Removed pump / motor assemblies and associated DLO-P-2A2 011 to be changed out to perform piping. Transferred piping to new pump / motors routine maintenance.

and re-installed in the system. Tested and returned to service.

i

l l

SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPENT (Continued)

EQUIPENT REQUIRING MAINTENANCE SYSTEM PROBLEM ACTION TAKEN llPCS-FIS-6 High Pressure Switch was found to be broken Replaced switch, calibrated instrument and Core Spray during channel calibration.

returned to service.

PJIR-SYS-1 Residual Heat A leak was found between Lines were repaired per ASE requirements and Removal RilR-V-1128 and RHR-V-508.

returned to service.

SGT-FR-2A2 Standby Gas Flow recorder quit working Installed new flow recorder, calibrated and Treatment after Div. II inverter returned to service.

l y

failed.

U l

MS-PT-SIB Main Steam The originally installed Removed Bailey transitter and replaced with a instrument was not environ-Rosemount.

Inspected tubing, bolt torquing, mentally qualified. Replace performed pressure check and operational with QC I instrument.

calibration. Returned the instrument to service.

DLO-PS-6Al Diesel Lube 011/

Originally installed Replaced ASCO pressure switches with Static DLO-PS-6A2 Starting Air pressure switches had 0-Ring pressure switches under the direction DSA-PS-9A a high failure rate.

of a plant design change. Calibrated the instruments and returned to service.

DLO-PS-6Bl Diesel Lube 011/

Originally installed Replaced ASCO pressure switches with Static DLO-PS-6B2 Starting Air pressure switches had 0-Ring pressure switches under the direction DSA-PS-9B a high failure rate.

of a plant design change. Calibrated the instruments and returned to service.

l l

SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continued) l l

EQUIPME NT REQUIRING MAINTENANCE SYSTEM PROBLEM ACTION TAKEN DSA-PS-15 Diesel Lube Dil/

Originally installed Replaced ASCO pressure switches with Static l

DSA-PS-16 Starting Air pressure switches had 0-Ring pressure switches under the direction l

DLO-PS-8 a high failure rate, of a plant design change. Calibrated the

(

instruments and returned to service.

I IIPCS-RLY-GlDC/DG3 liigh Pressure Rewired relay contact logic Rewired logic from normally open contacts to llPCS-RLY-G2DC/DG3 Core Spray to increase reliability.

normally closed cor. tact because the relay was not covered and the system engineer felt a normally closed contact configuration was more reliable.

O HPCS-B1-DG3 High Pressure The HPCS Battery had Replaced HPCS battery, tested and returned Core Spray problem with low specific to service.

gravity readings, replaced battery.

CRD-SPV-9 Control Rod GE/Valcor suggested Installed silicone rubber tubing on leadwires CRD-SPV-182 Drive modification after cracks of CRD-V-9 and CRD-V-182 per manufacturer's appeared in the Tefzel instructions and per associated plant design insulation and on the change. Tested valves and returned to service.

leadwires.

CMS-AY-3 Containment Time delay relays for Replaced time delay relays for CMS-SR-13 & 14.

CMS-AY-4 Monitoring CMS-SR-13 & 14 defective.

Verified operability and returned to service.

SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continued)

EQUIPMENT REQUIRING l

MAINTENANCE SYSTEM PROBLEM ACTION TAKEN l

l RilR-PIS-22A Residual lleat Hi/Lo Reactor Pressure Replaced low pressure switch, annunciator Removal annunciator in solid, cl eared. Calibrated and returned to service.

believed to be caused by a defective low pressure switch.

MS-PS-23C Main Steam Intermittent half scrams Replaced faulty pressure switch calibrated generated from the "A" RPS and tested. Prcblem was resolved, returned Trip System.

to service.

,o HPCS-PS-12 High Pressure Pressure switch is leaking Direct replacement switch was unavailable.

Core Spray internally.

Replaced with temporary switch until qual-fled replacement is received from manufac-turer. Calibrated, tested and returned to service.

ldlR-FT-15B Residual lleat Installed flow transmitters Replaced GE transmitters with Rosemount RilR-FT-15C Removal were unreliable and are not transmitters, re-tubed upstream cf manifold, environmentally qualified.

performed hydro-test. Recalibrated instru-ments and returned to service.

MS-DPIS-88 Main Steam Switch will not repeat Replaced micro-switch and installed new 3 way trip points on demand.

block valve which was plugged. Recalibrated dnd returned to service.

i I

SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPE NT (Continued)

EQUIPl1ENT REQUIRING MAINTENANCE SYSTEM PROBLEM ACTION TAKEN l

CMS-TR-4 Containment Temperature recorder was Repaired pen amplifier by replacing output Monitoring not reading accurately.

transistor. Calibrated and returned to service.

DG-GEN-1B Diesel Sparks emitted from generator Disassembled generator, modified generator Generator bearing.

Insulated bearing bearing insulation per a plant design change, failed.

tested and returned to service.

LD-TS-6200 Leak Detection Switch failed to maintain Replaced defective temperature module, m

calibration within tolerances.

calibrated and returned to service, a

m REA-RIS-609D Reactor Bldg.

Detector failed upscale.

Replaced G-M tube. Performed radiological HVAC recalibration and returned to service.

CRD-V-Il Control Rod Valve failed to open when Replaced top piston "0" ring, cleaned Drive scram logic was reset.

cylinder and straightened bent indication position bracket. Tested and returned to service.

SW-M-V-187A Standby Removed existing unqualified Installed new Class IE motors, bumped SW-M-V-1878 Service motors and replace with motors to assure proper direction of SW-M-Y-188A Water Class lE qualified motors.

rotation, tested and returned to service.

SW-ti-V-188B This task is in support of the FPC modification.

SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continued)

EQUIPMENT REQUIRING MAINTENANCE SYSTEM PROBLEM ACTIDN TAKEN SGT-FN-182 Standby Gas Air damper pivot shaft Aligned pivot shaft and welded to base of Treatment broken from base, damper.

DLO-M-P/10 Diesel Lube Diesel soak back pump Disassembled motor, checked resistance.

Oil would not run.

Replaced brushes, took vibration readings and returned to service.

MS-LIS-37C Main Steam Input of reactor level to RCIC Replaced switch 2A which provides input to was found to be inoperable RCIC. Re-ran surveillance procedure and no f;

during surveillance procedure.

returned to service.

l FDR-V-3 Floor Drains Closing time exceeded Cleaned and lubricated exposed parts of valve Technical Specification stem. Removed control air tubing and blew out requirements, with clean dry air. Disassembled and cleaned solenoid valve. No improvement noted.

Installed tegorary tubing and time tested valve within limits. Requested eengineering evaluation concerning possible shortening of tubing run or other permanent solution.

MS-RI S-610D Main Steam Hi-Hi trip setpoint Repaired defective rate meter, calibrated and failed during re-calibration.

returned to service.

SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continued)

EQUIPMENT REQUIRING

!!AINTENANCE SYSTEtt PROBLEtt ACTION TAKEN REA-RIS-609C Reactor Bldg.

Reactor Bldg. Exhaust Plenum Replaced defective 24V relay with a 36V relay liVAC Monitor 24V relay burnt out.

per a field change request.

ItS-M0-1 Main Steam Reactor vent to equipment Interlock contact on open coil was sticking.

drains would not stroke open.

Adjusted and cleaned contact, functionally tested and returned to service.

RFW-V-65B Reactor Valve would not seat during Disassembled valve and found seat cracking.

Feedwater Local Leak Rate Test.

Repaired by performing a weld overlay'and regrind. Retested per ILRT and returned to m4 service.

N RFW-V-10A Reactor Valve would not seat during Changed vulcanized seat elastomers. Retested RFW-V-10B Feedwater Local Leak Rate Test.

per Integrated Leak Rate Test procedures.

RFW-V-31A RFW-V-328 Reactor Valve would not seat during Disassembled valve and found backup seat Feedwater Local Leak Rate Test.

cracking. Repaired by performing a weld overlay and re-grind, changed vulcanized seat elastomers. Retested per Ingegrated Leak Rate Test procedures.

l

SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continued)

EQUIPMENT REQUIRING f1AINTENANCE SYSTEM PROBLEM ACTION TAKEN REA-E/P-1A Reactor Bldg.

Moore E/P converter had a Replaced Moore E/P converter with a Conflow to llVAC history of being unreliable, increase reliability per a plant design change.

Tested and returned to service.

REA-E/P-1B Reactor Bldg.

Moore E/P converter had a Replaced Moore E/P converter with a Conflow to ilVAC history of being unreliable.

increase reliability per a plant design change.

Tested and returned to service.

I RWCU-M0-4 Reactor Water Close stroke time too slow Replaced the existing motor pinion and worn j

m k

Cleanup to remain within the envelope shaf t clutch gear. Verified operability, and for environmental profile decreased closing time. Returned to service.

curves.

RPS-RLY-K14A Reactor The Supply System had com-Replaced originally installed scram contactors through 11 Protection mitted to the NRC to replace with a qualified series CR305 contactor.

System scram contactors by the Verified operability and returned to service.

I first Refueling Outage.

RCIC-H0-59 Reactor Core The motor for the RCIC test Replaced the motor with one that was Isolation return line isolation valve qualified. Tested, verified operability, Cooling not environmentally qualified and returned to service.

and needed to be replaced.

SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RE! ATED EQUIPMENT kontinued)

The following is a more detailed description of the problems and corrective action taken concerning the "B" Reactor Recirculation Pump.

In April,1985, a reactor recirculation flow control valve hydraulic line weld failed.

The line was repaired and the subsequent ftIlure analysis attributed the failure to excessive piping vibration.

Dat4 collected and analyzed from vibration instruments installed on recirculation pump "B"

indicated the cause to be excessive pump vibration.

During the M-3 outage the pump was partially disassembled and damage was found in the radial bearing and seal assembly.

The pump was reassembled and tested while in cold shutdown on the 60 Hz power supply.

The vibration data collected dur-ing the test was evaluated and determined by the pump technical represen-tatives and GE Nuclear Engineering to indicate that the vibration had been reduced to an acceptable level.

No further repair was deemed necessary.

July 7, during power escalation following the M-3 outage, vibration read-ings associated with loop B began increasin On July 8 the vibration levels of the pump exceeded the manufacturer'g.s recommended shutdown ifmits and the pump was subsequently secured.

Between July 8

r. d July 16, the plant operated in single loop while engineering options f or repair were considered.

On July 17, a shutdown was performed which allowed visual examination of the pump and installation of a snubber on the pump suction valve to be completed.

It was felt that the addition of the snubber would dampen the sympathetic vibration caused by the piping system and isolate the pump induced vibration, thus facilitating data analyses.

The plant was then returned to power and the punp was tested at 60 Hz with unsatisfactory resul ts.

It was concluded that pump repair must be accomplished.

On July 17,1985, due to power limitations imposed by single Recirculation Loop operation, an emergency technical specification change was necessary to permit continued operation in single loop at a power level greater than the 50% limit.

An emergency technical specification change was granted by the Commission on July 19, 1985, allowing the Supply System to operate in single loop operation at or near 72% power for the remainder of tne year.

2-29

f 2.E. INDICATIONS OF FAILED FUEL

)

Plant operation in 1985 produced no evidence of any fuel failures.

No indication of fuel failure was detected from main steam line or off gas radiation monitors during the course of plant operation.

Also, no indications of fuel failure were observed through normal Plant chemical or radiological analysis.

This information is supplied in accordance with requirements as set forth in Regulatory Guide 1.16.

l 2-30

2.F. CHANGES, TESTS, AND EXPERIMENTS Federal Regulations (10 CFR 50.59) and the WDPSS Nuclear Project No. 2 Operating License NPF-21 allow changes to be made to the facility as described in the safety analysis report and tests or experiments to be conducted which are not described in the safety analysis report, without prior Nuclear Regulatory Comission (NRC) approval, unless the proposed change, test, or experiment involves a change in the Technical Speci fi-cations incorporated in the license or an unreviewed safety question.

In accordance with 10 CFR 50.59, summaries of the changes performed and tests and experiments conducted in 1985 are provided, included are summaries of the safety evaluations.

2-31

2.F.1.A PLANT MODIFICATIONS AND DESIGN CHANGES The following plant modifications and design changes were completed in 1985 and reported in accordance with 10 CFR 50.59.

These modifications were evalu-ated and it was determined that they did not:

(a) increase the probability of occurrence of an accident or malfunction of the equipment important to safety, as previously evaluated in the WNP-2 updated Final Safety Evaluation Report l

(FSAR), (b) create the possibility of an accident or malfunction of a differ-ent type than previously evaluated in the FSAR, or (c) reduce the margin of safety as defined in the basis for any WNP-2 Technical Specification.

During the WNP-2 licensing process and particularly the FSAR review process, the NRC requested the Supply System to evaluate Secondary Containment Pressur-ization events and the consequences thereof.

From that effort it was con-cluded that to mitigate 1)the consequences to safety related equipment in the Reactor Building caused by the fuel pool boiling, 2) the potential impact of the ECC systems from flooding and 3) exceeding the capability of the Standby Gas Treatment units to handle the saturated conditions, the Fuel Pool Cooling (FPC) system needed to be upgraded to Seismic Class I and Quality Class I.

A brief description of FPC plant design changes, which led to a modification of the FSAR, are also included in this section.

PLANT DESIGN CHANGE 83-0037-1A Plant Design Change 83-0037-1A was initiated because experience at other nuclear power plants has shown that gas pressure develops in sealed B4C enclosures causing distortion of the spent fuel rack cavities and binding of spent fuel assemblies.

The design of the spent fuel racks at WNP-2 did not have provisions for the venting or sampling of the off-gas produced from this reaction.

This plant design change added tubing and sample valves necessary for moni-toring off-gas pressure, sampling of off-gas and venting to relieve pressure in the spent fuel assemblies.

This rodification did not result in a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question.

2-32

PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)

PLANT DESIGN CHANGE 83-0048-0A/0D Plant Design Change 83-0048-0A was initiated to complete the original design of the Reactor Feed Turbine Control System and make operable the " Speed Lockup" feature for the Reactor Feedwater (RFW) Turbines which would cause the units to remain at their last required speed and flow output following loss of input signal to the Reactor Feedwater Turbine Governors.

This feature was designed and partially installed as part of the original General Electric system Asign, but was not completed or connected to the balance of plant components by the Architect Engineer.

As plant testing and transient analysis studies dictated, modification of the original design change package 83-0048-0A was incorporated through,Setpoint package 83-0048-00, which incorporated Automatic Reactor Vessel Level Setdown" upon reactor protective trip and transfer of the Reactor Feedwater Control System and Drive Turbine Governors to the critical power supply busses.

The modifications to the original change package were intended to prevent a vessel feedwater overfill condition upon reactor automatic trip and increase the system availability.

This modification did not result in a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 84-0365-0A Plant Design Change 84-0365-0A was initiated because the pressure indicator located at the suction of the High Pressure Core Spray pump was not designed for the large pressure surges to which it was subjected.

This olant design change removed the local pressure indicator from service.

This instrument provided no input to control logic of the High Pressure Core Spray system.

Local indication can be obtained from HPCS-PIS-3 which is located approximately 18 inches from the location of the removed pressure indicator.

This modification did not result in a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question.

2-33

PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)

PLANT DESIGN CHANGE 84-426-0A Plant Design Change 84-426-0A was initiated because originally the High Pressure Core Spray (HPCS) Diesel Generator was not designed for idle speed operation with automatic return to rated speed during an accident condition.

Due to monthly surveillance starting and running requirements, this modi fi-cation is expected to increase generator reliability and reduce the amount of corrective maintenance needed on the diesel generator.

This plant design change modified the electrical circuit of the HPCS Diesel Generator to allow auto acceleration from idle speed or slow speed to normal running speed during a Loss of Coolant Accident or a Loss of Power Accident.

This modification did not result in a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 84-542-0A Plant Design Change 84-542-0A was initiated because Reactor Recirculation Pump seals, which were replaced, had a higher design leakage than the originally installed seals.

The associated flow instruments were not the proper range for the new pump seals.

This plant design change replaced flow instruments for the Reactor Recircu-lation Pump seals with instrumentation that had a higher range, corresponding with the higher design leakage of the new pump seals.

This modification did not result in a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question.

PLANT DE3IGN CHANGE 84-0927-1 A Plant Design Change 84-0927-1 A was initiated because solenoid valves tempo-rarily installed in the Standby Service Water system wera not environmentally qualified.

This plant design change provided the engineering review necessary to upgrade these solenoid valves to class lE and authorizes the continued use of these valves.

This plant design change also revises the start delay time of the Service Water Pumps from 30 seconds to 20-40 seconds.

This allows for slower than initially anticipated opening times of interlocked valves.

This modification did not involve a change to the WNP-2 Technical Speci fi-cations or involve an unreviewed safety question.

2-34

PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)

PLANT DESIGN CHANGES 84-1065-0A/0B Plant Design Changes 84-1065-0A/0B were initiated because the Fire Protection l

System for the records storage area in the service building had been requir-ing excessive maintenance.

These plant design changes replaced the halon fire protection system in the records storage area with heat sensitive automatic sprinklers.

These modifications did not result in a cnange to the WNP-2 Technical Speciff-cations or involve an unreviewed safety question.

I PLANT DESIGN CHANGE 84-1210-3A Plant Design Change 84-1210-3A was initiated as a portion of the design for the Fuel Pool Cooling modification and upgrade.

This particular plant design change installed three motor operated Reactor Closed Cooling Water valves into the system.

These valves are the primary source of cooling water to the Fuel Pool Cooling Heat Exchangers.

They also serve as isolation valves from the Standby Service Water System, the backup source of water to the Fuel Pool Cooling Heat Exchangers.

This modification resulted in a change to the WNP-2 Technical Specifications but did not involve an unreviewed safety question.

PLANT DESIGN CHANGE 84-1210-4A Plant Design Change 84-1210-4A was initiated as a portion of the design for the Fuel Pool Cooling modification and upgrade.

This particular plant design change moved the control switches and indicating lights for the wetwell containment isolation valves from Board "N" to the new Fuel Pool Cooling Boards, both located in the main control room. This design change also added an additional interlock in the control circuit to prevent these valves from re-opening prior to the FAZ signal clearing.

This modification did not result in a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question.

2-35

PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)

PLANT DESIGN CHANGE 84-1210-6A i

Plant Design Change 84-1210-6A was initiated as a portion of the design for

{

the Fuel Pool Cooling modification and upgrade.

i This particular plant design change added six motor operated Standby Service Water valves to the system.

Four of these valves serve as isolation for the Standby Service Water System, the backup source of water to the Fuel Pool Cooling Heat Exchangers.

The other two valves serve as isolation for the Standby Service Water System, the backup makeup source of water to the spent fuel pool.

This modification did not result in a change to the WNP-2 Technical Speciff-cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 84-1211 -4A Plant Design Change 84-1211-4A was initiated as a portion of the design for the Fuel Pool Cooling modification and upgrade.

This particular plant design change added Fuel Pool level switches with con-trol room annunciation.

These level switches provide indication of either high or low fuel pool level.

In the event of a fuel pool low level alarm, these level switches supply the auto closure input for the isolation valves between the reactor building and the radwaste building in an attempt to isolate a potential break in the non-seismically designed portion of the system.

This modification did not involve a change to the WNP-2 Technical Specifi-cations or involve an unreviewed sa'aty question.

PLANT DESIGN CHANGE 84-1211-7A Plant Design Change 84-1211-7A was initiated as a portion of the Fuel Pool Cooling modification and upgrade.

This particular plant design change was initiated because the original design of the Fuel Pool Cooling System provided local fuel pool temperature indica-tion but did not provide annunciation in the main control room.

This olant design change added temperature switches to the circuitry to allow control room annunciation of fuel pool high temperature.

This modification did not result in a change to the WNP-2 Technical Speci fi-cations or involve an unreviewed safety question.

2-36

PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)

PLANT DESIGN CHANGE 84-1211 -9A Plant Design Change 84-1211-9A was initiated as a portion of the design for the Fuel Pool Cooling modification and upgrade.

This particular plant design change provided the necessary radiation exposure calculations for equipment designed to be installed into the Spent Fuel Pool System.

These calculations were a prerequisite to writing purchase specifica-tions for the new plant equipment due to equipment qualification requirements.

This modification did not result in a change to the WNP-2 Technical Speciff-cations or involve an unreviewed safety question.

ptANT DESIGN CHANGE 84-1212-0A Plant Design Change 84-1212-0A was initiated as a portion of the design for the Fuel Pool Cooling modification and upgrade.

This particular plant design change provides material specifications and installation instructions for equipment necessary to upgrade the Fuel Pool Cooling Heat Exchangers from Quality Class II to Quality Class I, and Seismic Class I.

This modification did not involve a change to the WNP-2 Technical Speciff-cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 84-1212-1 A Plant Design Change 84-1212-1 A was initiated as a portion of the design for the Fuel Pool Cooling modification and upgrade.

This particular plant design was an addendum to the contract specification for the procurement of the Fuel Pool Cooling Heat Exchangers.

The modifications were as follows:

o Increase the shell side design pressure from 150 psi to 270 psi.

o Revise the pipe stress calculations for nozzle loads, o

Allow the ASME Boiler and Pressure Vessel Code.1980 Edition, to be used to qualify existing and design new component parts.

This modification did not involve a change to the WNP-2

• cchnical Speci fi-cations or involve an unreviewed safety question.

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PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)

PLANT DESIGN CHANGE 84-1229-0A Plant Design Change 84-1229-0A was initiated because the location of the plant welding facility was moved from the service building machine shop to another building on site.

This plant design change removed a Service Building Exhaust Air fan and associated dust collection equipment from the service building to provide i

additional space for storage and spare parts.

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This modification did not result in a change to the WNP-2 Technical Speciff-cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 84-1233-0A Plant Design Change 84-1233-0A was initiated because the low pressure alarm for the Residual Heat Removal System would not reset when system pressure was normal.

The cause for this problem was that the instrument range was too broad for this particular application and the low pressure alane would not reset.

This design change replaced the low pressure switch with one that was the proper range and that had a smaller deadband.

This modification did not involve a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 84-1270-0A Plant Design Change 84-1270-0A was initiated because the high delta flow alarm for the Reactor Water Cleanup System gave the operators no advance warning prior to the system isolating.

This plant design change added a 45 second time delay to the alarm circuit, allowing the control room operators time to rectify the flow mismatch problem prior to isolating the Reactor Water Cleanup System.

This modification did not involve a change to the WNP-2 Technical Speci fi-cations or involve an unreviewed safety question.

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PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)

PLANT DESIGN CHANGE 84-1277-0A Plant Design Change 84-1277-0A was initiated because, with the reactor at l

power, the Reactor Water Cleanup System has experienced excessive pressure transients following system isolation with blowdown established.

With the inboard and outhcard containment isolation valves closed, the system rapidly depressurizes through the blowdown flow control valve, resulting in reactor water flashing to steam.

This plant design change interlocks the inboard and outboard containment iso-lation valves with the blowdown flow control valve, so that if the isolation valves are not full open, the blowdown flow control valve will automatically close.

This modification did not result in a change to the WNP-2 Technical Speciff-cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 84-1279-0A Plant Design Change 84-1279-0A was initiated to allow routine maintenance to be performed on permanently installed control and service air compressors.

This plant design change provided a portable air compressor to ensure adequate control and service air was available for plant loads while maintenance was being performed on permanently installed compressors.

This modification did not result in a change to the WW-2 Technical Specifi-cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 84-1279-1 B Plant Dcsign Change 84-1279-1B was initiated to void the previously written design change, which provided a temporary control and service air compressor.

This plant design change removed the temporary air compressor and returned the system to its original design configuration.

This modification did not result in a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question.

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PLANT MDDIFICATIONS AND DESIGN CHANGES (Continued)

PLANT DESIGN CHANGE 84-1296-0A Plant Design Chaages 84-1296-0A was initiated because, after a review of Appendix R, it was determined that the Architect Engineer failed to include vertical thermolagging for ten cables identified as affecting safe shutdown.

This plant design change provided the design to re-route, run conduit and therinolag these cables.

This modification, in conjunction with other modifi-cations and plant programs, indicates WNP-2's efforts to comply with 10CFR Appendix R requirements.

This modification did not result in a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question.

PLANT 9ESIGN CHANGE 84-1298-0A Plant Design Change 84-1298-0A was initiated because the hydrogen analyzers for the Off Gas System were tripping intemittently and were generally unreliable.

This plant design change made the following modificat!ons:

o Modified circuitry so that a high vacuum alarm would isolate the hydrogen analyzer from the main condenser.

o Added a convenience feature that would automatically reset the venting solenoid valve when the reset button on the front of the panel was pushed.

o Added a short time delay in the circuitry to allow for intermittent high vacuum conditions without tripping the hydrogen analyzer.

o Added a test tee for access to the sample filter.

These modifications did not involve a change to the WNP-2 Technical Speciff-cations or involve an unreviewed safety question.

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PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)

PLANT DESIGN CHANGE 84-1393-0A/0B Plant Design Changes 84-1393-0A/0B were initiated because Regulatory Guide 1.97 requires the radiatiqn detectors for the reactor building elevated release duct to read to 1 x 105 uc/cc of Xe 133.

The origina1Ty installed radiation detectors did not read at this level.

These plant design changes shielded the detectors with 0.020" thick lead foil to ensure the specifications of Regulatory Guide 1.97 are met.

These modifications did not result in a change to the WNP-2 Technical Speciff-l cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 84-1399-0A Plant Design Change 84-1399-0A was initiated because the originally installed turbine building supply fan discharge dampers were constructed of aluminum and with under designed actuator shafts.

These shafts were bending due to flow induced forces.

This plant design change replaced the aluminum dampers with steel gravity backdraft dampers which improves the reliability of the system to perform its intended function.

This modification did not result in a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 84-1449-0A Plant Design Change 84-1449-0A was initiated because the Fire Protection System for the Turbine Building station #56 was not performing reliably.

This plant design change replaced the deluge open nozzle system with a wet system, using standard automatic sprinkler heads.

This design change also installed a terminal box for the new system and removed ti.e existing alarm panel.

This modification did not result in a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question.

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PLANT MDDIFICATIONS AND DESIGN CHANGES (Continued)

PLANT DESIGN CHANGE 84-1457-0A Plant Design Change 84-1457-0A was initiated because the filter septa for the Reactor Water Cleanup (RWCU) domineralizers required replacement because they were partially plugged and did not allow uniform resin precoating.

Spares were not readily available for replacement using the original design and would not have been cost effective to obtain.

This plant design change replaced the filter septa for the RWCU domineralizers with higher efficiency filter septa increasing the removal of soluble and insoluble impurities from the system.

This modification did not result in a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 84-1537-0B Plant Design Change 84-1537-08 was initiated because the reactor vessel indi-cating transmitter switches were not Class IE qualified and historically had not provided the desired reliability.

The plant design change replaced the combination level transmitter / switch with two separate instruments.

Separating the level indicating switch and level transmitter functions allows increased accuracy and reliability in addition to fulfilling the equipment qualification requirements.

This modification did not result in a change to the WNP-2 Technical Specift-cations or involve an unreviewed safety question.

PtANT DESIGN CHANGE 84-1552-0A Plant Design Change 84-1552-0A was initiated because the Reactor Water Cleanup Pumps were cavitating and tripping as a result of depressurization following reactor scrams, which significantly shortens pump seal life.

Seal replacement activities accounted for the highest exposure levels during the year.

This effort was also pursued due to an aggressive ALARA program.

This plant design change added a subcooling line from the nonregenerative heat exchanger outlet to the connon Reactor Water Cleanup suction line to increase the net positive suction head available to the pumps.

This modification did not involve a change to the WNP-2 Technical Speciff-cations or involve an unreviewed safety question.

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PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)

PLANT DESIGN CHANGE 84-1622-0A Plant Design Change 84-1622-0A was initiated because, after a review of this modification package was made, it was determined that there was no technical justification for qualifying the HVAC intake filters to the diesel building as I

Seismic and Quality Class I.

This plant design change downgraded the Seismic and Quality Class from Class I to Class II to be more consistent with overall plant design configuration.

l This modification did not result in a change to the WNP-2 Technical Speciff-l cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 84-1717-0A Plant Design Change 84-1717-0A was initiated because the heat trace on the Main Steam Leakage Control System did not compensate for different temperature variance along the pipes. This caused the auxiliary heat trace thermoststs to interrittently control pipe temperatures above designed operating temperatures.

This plant design change deleted all auxiliary thermostats and modified the temperature controller to be a single removable themocouple.

This modification did not involve a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 85-008-0A Plant Design Change 85-008-0A replaced the manual outboard containment iso-lation valve to the wetwell with a motor operated valve.

This modification allows for operation of the suppression pool cleanup loop during plant operation.

This modification did not result in a change to the WNP-2 Technical Speciff-cations or involve an unreviewed safety question.

The conversion of this valve from manual to motor operated generated a Technical Specification change which was covered on another Design Change Package which was not completed in 1985.

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PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)

PLANT DESIGN CHANGE 85-203-0A Plant Design Change 85-203-0A was initiated because the Containment Instru-ment Air programmers which were originally installed in the system, were not environmentally qualifled.

This plant design change replaced originally installed prograssiers with equip-ment which was environmentally qualified, and updated appropriate drawings and documentation to reflect these changes. Although Plant Design Change 85-203-0A did not actually initiate modifications to the FSAR, Plant Design Changes 84-0966-0A/B were worked in conjunction with 85-203-0A and provided the docu-mentation to change drawings and modify the FSAR as required.

These modifications did not result in a change to the WNP-2 Technical Speciff-cations or involve an unreviewed safety question.

PLANT DESIGN CHANGE 85-0324-0A Plant Design Change 85-0324-0A was initiated to request an engineering review to detennine if there will be a future need for the deferred Chemical Waste Processing equipment which has been installed but is not operational, and the disposition thereof.

This plant design change identifies the operational boundaries for the Chemical Waste Processing System, revises the associated drawings to reflect these boundaries and provides the authorization to use inactive equipment as spare parts if needed.

This modification did not result in a change to the WNP-2 Technical Soeciff-cations or involve an unreviewed safety question.

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2.F.1.8 CHANGES TO PROCEDURES Procedures described in the WNP-2 Final Safety Analysis Report (FSAR) are used by the Plant Operating Staff and by various offsite support organizations.

In 1985, the Plant Staff made changes to procedures in accordance with 10 CFR 50.59, and concluded that none of the changes involved unreviewed safety questions.

Changes to procedures were generally either administrative or technical in nature.

Administrative changes consisted of title, organizational and editorial changes, while technical changes were the result of system or component modifications, or improvements in procedural processes.

A safety evaluation was conducted for each change in accordance with 10 CFR 50.59, was reviewed and approved by the Plant Operations Cossnittee and are avail-able for audit as necessary.

The review concluded that the probability of occurrence or consequences of an accident or equipment malfunction were not increased, there was no reduction in any plant safety margins, and the possibility of an accident or malfunction not previously evaluated was not increased.

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n-2.F.2 PLANT TESTS The following plant tests were performed in 1985 and are reported in accordance with 10 CFR 50.59:

Temporary Plant Procedure 8.3.22 Reactor Recirculation Pump 1B Uncoupled Motor Vibration Testing:

This temporary test procedure was written and performed to gather baseline vibration data following maintenance which was performed on Reactor Recir-culation Motor 1B prior to coupling the motor to the pump.

This test was run prior to but in conjunction with TP 2.1.7, Recirculation Pump Vibra-tion Testing.

This test did not involve an unreviewed safety question or a change to the WNP-2 Technical Specifications.

Temporary Plant Procedure 2.1.7 Recirculation Pump Vibration Testing:

This temporary plant procedure was written and performed to provide guid-ance for operation of Reactor Recirculation (RRC) Pump 1B with the Plant in a Shutdown Condition. This test allowed evaluation of pump performance following the pump bearing replacement but prior to re-starting the plant.

This test did not involve an unreviewed safety question or a change to the WNP-2 Technical Specifications.

Temporary Plant Procedure 8.3.24 Reactor Recirculation Pump at Power Testing:

During the sumer maintenance outage additional vibration measuring equip-ment and a vibration damper were installed on the "B" Loop of the Reactor Recirculation System.

This procedure provided the directions to be used for further diagnosis of the vibration as well as provide the data for establishing operating limits for the vibrating loop upon the completion of the test., This test did not involve an unreviewed safety question or a change to the WNP-2 Technical Specifications.

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CECElVED n

s Imc Washington Public Power Supply System P.O. Box 968 3000 George Washington Way Richland. Washington 993py g'09),372 5000

!!! II: Ip 4

9ECICK VW Docket No. 50-397 February 27, 1986 Mr. J.B. fiartin Regional Administrator Region Y U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, CA 94596

Subject:

NUCLEAR PLANT NO. 21985 ANNUAL REPORT

Reference:

1) Title 10, Cod? of Federal Regulations, Part 50.59(b)

2) WNP-2 Technical Specifications, 6.9.1.4 and 6.9.1.5

3) Regulatory Guide 1.16, Reporting of Operating Information -

Appendix A In accordance with the above listed references, the Supply System hereby submits the Annual Report for calendar year 1985.

Should you have any questions or comments please contact M.R. Wuestefeld, WNP-2 Plant Engineering Supervisor, Reactor Systems.

Very truly yours, Ok C.M. Powers Plant Manager CIP:MRW:tw Attachments Y ~.W )

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