ML17285B080

WPPS-2 Annual Operating Rept for 1989

ML17285B080
Person / Time
Site:	Columbia
Issue date:	12/31/1989
From:	Powers C WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
NUDOCS 9003140332
Download: ML17285B080 (192)

Text

ACCHZMTZD D)STR1SUTtON; DKMONSTRZT1ON SYSTEM REGULAT INFORMATION DISTRIBUTIOYSTEM (RIDS)

ACCESSION NBR: 90033:49332 DOC. DATE: 8+j.'Mj'3?

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"ti WASHINGTON PUBLIC POWER SUPPLY SYSTEM F.O. Box 968

~ 3000 George Washington Way

~ Richland, Washington 993$2 Docket No. 50-397 February 28, 1990 Mr. J.

B. Martin Regional Administrator Region V

U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, California 94596

Subject:

NUCLEAR PLANT NO.

2 ANNUAL REPORT

Reference:

1)

Title 10.,

Code of Federal Regulations, Part 50.59(b) 2)

WNP-2 Technical Specifications, 6.9.1.4 and 6.9.1.5

~

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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 1989.

Should you have any questions or commerits please contact G. L. Gelhaus, WNP-2 Assistant Plant Technical Manager.

Very truly yours, C

M. Powers WNP-2 Plant Manager

/bc Attachments voosxeass2 s9123i

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.OPERA REPORT 1989

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TABLE OF CONTENTS

1.0 INTRODUCTION

1 1.1 1989 Power History Graph for MNP-2'........

4 2.0 REPORTS

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2.1 2.2 Annual Personnel Exposure and Monitoring, Report Main Steam Line Safety/Relief Valve Challenges.............

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

11-2.4 Summary of Significant Maintenance Performed on Safety-Related Equipment 2.5 Indications of Failed Fuel 2.6 Plant Modifications....

16 35 42 2.6.1 Plant Design Changes

43 2..6.2 Lifted Leads and Jumpers 2.6..3 FSAR Amendment Evaluations....

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59 2.6.4" Other

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2.7 Plant Tests and Experiments

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60 74 2.8 2.9 Plant Procedure Changes Reactor Coolant Specific Activity Levels.

75

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79 2.9.1 MNP-2 Dose Equivalent Iodine Graph 2.10 Diesel Generator Failures

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~ 80

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81 2.11-Fire Protection Program Changes.........

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ANNUAL OPERATING REPORT OF WNP-2 FOR 1989 OOCKET NO. 50-397 FACILITY OPERATING LICENSE NO.

NPF-21 Washington Public Power Supply System 3000 George Washington Way

Richland, Washington 99352

1.0 INTRODUCTION

The 1989 Annual Operating Report of Mashington Publ ic Power Supply System Q

Plant Number 2

(MNP-2) is provided as a

supplement to the Honthly Operation Report.

This report is submitted in accordance with the requirements of Federal Regulations and Facility Operating License NPF-21.

It should be noted

that, for ease.

of reference and completeness, additional reports are also included.

Plant MNP-2 is a

3323 HMt, BMR-5, which began commercial operation on, December 13, 1984.

8 On January 30, 1989 the reactor scrammed due to Turbine Control Valve Fast Closure actuation of the-Reactor Protection System (RPS) logic.

The RPS logic

'as actuated when the main generator 500Kv output breakers tripped as a result of high currents created when a porcelain insulator, on the -output side of the 25/500Kv main transformer, shorted to ground.

On February 2-,

1989, following a three-day outage to replace and clean several insulators, the Plant returned to normal operation.

However, due to a problem with opening one of the four outboard. Hain Steam Isolation Valves, power output was limited to 78 percent.

The Plant continued to run in this configuration, with the valve closed and steam supplied through three of four main steam lines, until the Plant was shut down for the annual maintenance and refueling outage.

e-J From April 28, 1989 until July 2, 1989 the Plant was, in a shutdown condition as.scheduled -for the annual maintenance and refueling outage.

Following the outage; the Plant was restarted and operated until August 6,

1989 when a

reactor scram occurred due to the trip of a reactor feedwater pump caused by a problem in the= feedwater pump control oil system.

On August 9, 1989 the Plant was restarted; The Plant was shut down and an Unusual -Event was declared on August 11, 1989 as a result of declaring six Class 1E, 480 volt A.C., Motor Control - Centers (HCCs) inoperable due to the discovery of a

design defi-ciency.:All--of the affected HCC power supply circuit breakers were replaced with jumper cables of equal capacity with the exception of

one, which was

replaced, with-. a fused disconnect of equal capacity..

During

restart, on August 17,

1989, another reactor scram occurred as a--result of=a surveillance being performed on a'eactor level instrument associated.

with-the Automatic Oepressurization System (ADS).

On August 18, 1989 the reactor was restarted but power output was limited to 70 percent due to removing a reactor feedwater pump from service because of a bearing failure.

One of two reactor pumps was repaired and the Plant returned to-full power on September 13, 1989.

The Plant essentially remained at 100 percent power until September 21,

1989, when it was-shut down due to two leak-ing condenser

.tubes and two ruptured bellows connections on a

low-pressure steam extraction line.

Repairs were made and the Plant was restarted on September 29, 1989 and ran at or near 1005 capacity for the remainder of the year (94 days).

5I Ouring 1989,:there were several examples of major accomplishments which required significant effort on the part of Supply System personnel to complete.

The following is a summary of those efforts.

(a)

The fourth refueling outage was successfully completed.

Significant activities included:

o Preventive maintenance on the eight Main Steam Isolation Valves (MSIVs), and a major overhaul on the valve, that limited'lant power

.output to 78 p'ercent of capacity.

Four of the valves were repaired to reduce the potential for valve binding caused by galling in the

-cylinders.

Valve pistons of a

new 'design by

Rockwell, the MSIV

supplier, were installed in those MSIVs.

o

=

Overhaul of one of the two Reactor Feedwater Pump Turbines.

The turbine was dismantled and the rotors were cleaned and inspected for-cracks or other defects.

o Inspection of two of the three Oiesel Generators.-

This task included replacing power assemblies in the two engines.

o

- Inspection of one of three Low-Pressure Turbine Rotors.

Non-: destructive examination of the rotor confirmed 17 crack indications and the blades.

were replaced.

Subsequent evaluations determined that the problems were limited to this single-rotor.

e-

'00'aintenance on 40 Control Rod Orive Mechanisms (CROMs).

This activity included removing, replacing and rebuilding the CROMs.

Removal of a

radioactive "hot spot" in the vessel dr ain to the

-Reactor Mater Cleanup System.

This activity required the installa-tion-of a temporary bottom head drain plug in the Reactor Pressure

.Vessel.

The plug was installed from the top= of the vessel which required removal of four fuel assemblies, a control rod blade, guide tube and associated support pieces.

Removal of spent fuel assemblies and refueling the reactor.

The refueling activity included replacing 136 fuel assemblies, using a

fuel shuffle scheme.

(b)

In terms of electrical

output, MNP-2 delivered 6.1 billion kilowatt-hours to the. Bonneville Power Administration, surpassing the previous year'

record by more than 117 mi 1 lion kilowatt-hours..

In

addition, the

- capacity factor for 1989 was a Plant record 63.78 percent (up from 62.38 percent in 1988).

-(c).

A new monthly generation mark was established in

Oecember, when 780 million kilowatt-hours were generated.

f

(d)

In

December, MNP-2 celebrated five years of commercial operation.

Since

1985, the 'lant has provided more than 28 million megawatt-hours of electricity to the Bonneville Power Administration.

In 1989, total radiation exposure at the Plant was 492 man-rem, as compared to the 1988 level of 352 man-rem.

(The Institute for Nuclear Power Operation

.(INPO) has set;460 man-rem as the 1990 industry goal for BMRs.)

Contributing to this increase were the following activities:

0 Removal and replacement of 40 Control Rod Drive Mechanisms (CRDMs).

During work on the first 20

CRDMs, the-man-rem exposure was 13.005.

As a:result of that exposure, temporary shielding was put into place and the man-rem exposure was reduced to 5.635 for the remaining 20-CRDMs.

o Removal of the "hot spot" in a

reactor bottom head drain line

.elbow.

Before replacement, the "hot spot" area was reading between 2,000 and 3,000R.,

Total man-rem exposure for this activity was 18.836.

e o

Modifications to the Control Rod Drive Rebuild Room and flushing of Low-Pressure Core Spray (LPCS)

System lines.

These actions were taken to reduce man-rem exposure in the future.

During the year NNP-2 received 23 'Notices of Violation (NOVs):

One (1)

Level III, twenty-one (21)

Level IY and one (1)

.Level V.

The Level III violation. was:-associated with commercial grade dedicati.on issues and included a proposed

$50,000 civil penalty.

Also during-1989, a total of 45 Licensee Event Reports (LERs) were written and submitted pursuant to the requirements of 10CFR50.73.-

-.. The-1989 capacity factors, based upon net electrical-'nergy

output, are listed in the following table.

Month Ca acit Factor January February March April

. Hay June July August September October November December Overall 76.17 68.25 73.02 68.32 0

0 88.47 52.94 58.39 88.80 95.84

95. 74 63.78

" Started Maintenance/Refueling Outage

~ Ended Maintenance/Refueling Outage

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90 80 70 E

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I-40 65 30 0

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Refuel Outag 0 January February March

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p)y'.: ','ugust'( 'September October November December 1989 Data based on average power generated per da

. Therefor, recovery from a scram that ocurred within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period,willnot indicate a zero percent power level.

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I

2.0 REPORTS The -reports provided in this section meet the requirements of Federal Regula-r

~

tions (10CFR50.59) and the MNP-2 Operating License.

Complete, data for the year 1989 has been included.

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2.1 ANNUAL'ERSNINEL'EXPOSUfIE ANO NONI@RING REPORT RER-020 I

I I WASHINOTOl C POWER SUPPLYI SYSTEH RADIA EXPOSURE RECORDS

'ORK AND JOB FUNCTION REPORT /'. 16 APPENDIX A' I ~ ~

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NUCLEAR PLANT NO. 2

~ I NUHBER OF PERSONS RECEIVIN4 OVER 100 ilREH REPORT FOR CALENDAR YEAR. 1989 TOTAL HAN-REtf 8TATION EtPLOYEES UTILITY CONTRACTORS STATION EtMILOYEES AND OTHERS EtlPLOYEES UTILITY EtlPLOYEES CONTRACTORS AND OTHERS

~ J'l II ~I OPERATIONS fl SURVEILLANCE t&INTENANCE PERSONNEl-

68. 105 OPERATINO PERSONNEL

42. 537 HEALfH PHYSICS PERSONNE~&. 767 SUPERVISORY PERSONNEL

15. 073 ENOINEERINO PERSONNEL

14. 164

0. 073

55. 753

0. 000

0. 000

40. 102

32. 471

0. 027

0. 000 0:037 l5. 620

22. 13D 0."004

1. 816

0. 283

h. 090

f. 472

9. 165

13. 565

4. 036

4. 399

31. 527

0. 000

8. 943

0. 089

h. 090 ROUTINE tlAINTENANCE L"I HAINTENANCE PERSONNEL

23. 592 OP ERATINO PERSONNEL

i. 916 PHYSICS PERSONNEE.=7. 561 SUPERVISORY. PERSONNEL

2. 3&8 ENOINEERINO PERSONNEL

3. 619

0. 139

0. 000

0. 000

0. 000

2. 498

0. 330

4. 002

0. 980

1. 052

23. 78$

14. 824

0. 000

1. 571 I. 266

i. 303

0. 000

0. 884

0. 103

0. 669

0. 054

11. 054

0. OOO O. OOO

0. 000

1. 213 40'II

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54I 34I INSERVICE INSPECTION

'AINTENANCE PERSONNEL

2. 790 OPERATINO PERSONNEL

2. 101 HEACIH PHVSICE1'ERSONNEL~IE SUPERVISORY PERSONNEL

~

0. 715 ENDINEERINO PERSONNEL

3. 313

0. 184

0. 010

3. 116

9. 977

0. 487

1. 222

0. 000

7. 869 1'. 799

0. 000

0. 000

i. 512 0; 000

0. 589 1: 597

0. 000

3. 919 0 000 0 000

0. 000=

0: 622

0. 179

0. 004

2. 316

f. 120 SPEC IAL HAINTENANCE

5:

I HAINTENANCE PERSONNEL 116. 436

1. 776 161. 313 OPERATINO PERSONNEL

1. 493

0. 000

0. 000 ALfH PHYSICS PERSONNEL

~07

0. 000

30. 639 SUPERVISORY PERSONNEL

3. 313

0. 000

3. 386 ENOINEERINO PERSONNEL

)0. 462

6. 805

13. 340

89. 675

0. 747 9'P. 971

0. 555 D. 000

0. 000 6.723 D: 000

30. 502

2. 484

0. 000 '. 857

2. 766

5. 127

3. 848 Il~

4 WA8TE PROCESSINO tlA'IHTCNANCd'PERSdf4NEL' I'

lf0 458'f I l f"1 f fO'.'OOd HI OOO OPERATIN4 PERSONNEL

0. 070.

0. 000

0. 000

5. 319

0. 022

0. 000

0. 000

0. 000

0. 000 34 HEALTH PHYSICS PERSONNEL 0; 467 0.000'. 700 0:523 ': 000 2: 797 SUPERVISORY PERSONNEL

0. 000

0. 000

0. 991

0. 000'. 000

i. 902 ENOINEERINQ PERSONNEL

'O. 000 ' '

'0. 000 '

0. 000

0. 000

0. 000

0. 000 7

~ H R EFUELINQ HAINTENANCE PERSONNEL 15. 235 '

0. 013

1. 220

13. 147 OPERATINO PERSONNEL

1. 009

0. 000

0. 000

0. 714 HEATH PHYSICS PERSONNEf.

0. 528 0, 000

2. 596

0. 723 SUPERVISORY PERSONNEL'. 542

0. 000

o. Ooo

0. 210 ENOINEERINO PERSONNEL

1. 336

0. 604

0. 220

0. 358

0. 004

0. 000

0. 761

0. 000

0. 000

~

1: 257

0. 000

. 0. 000

0. 206

0. 067 TOTAL HAINTENANCE PERSONNEL 234. 613

2. 001 249. 940 1&4. 866

0. 832 147. 232

~ I

0. 000

0. 000 OPERATINO PERSONNEL

'9. 126

0. 000

o. ooo

37. 245 HEALTH PHYSICS"PERSONNEL

40. 672 0.037'3. 410 35."499 '"0. 004 "45."334 SUPERVISORY PERSONNEL

22. 011

2. 000

5. 000

10. 251

'. 651

3. 955 ENOINEER INO PERSONNEL

32. 894

22. 188

41. 104

10. 516

9. 375

14. 269

>>>>>>ORAND TOTAL>>>>>>

379. 316

26. 226 34'9. 454 258. 377

11. 862 210. 790

II I

2.2 MAIN STEAM LINE SAFETY/RELIEF VALVE CHALLENGES

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This section contains information concerning main steam line safety/relief valve challenges for calendar year 1989 in accordance with the requirements of NUREG 0737, Item II.K.3.3, and as required by WNP-2 Technical Specifications, Administrative Controls section, paragraph 6.9.1.5(b).

FFiF 1

TYPE OF PLANT REASON FOR REACTOR ACTUATION CONDITION ACTUATION POWER ASSOCIATED 9TE E ~~~

EEL 01/30/89-"

- HS-RV-18 01/30/89'.

- HS-RV-lc A

A 6X -

89-002 65 -

89-002 D

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D D

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D D

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C ~,

17.8%

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17.8%

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17.85 04/29/89 04/29/89 04/29/89 04/29/89 04/29/89 04/29/89 04/29/89 04/29/89 04/29/89 04/29/89 04/29/89 MS-RV-58

-- HS-RV-1C HS-RV-2C

-- HS-RV-3C

'S-RV-4C

- HS-RV-5C HS-RV-10 HS-RV-2D

- MS-RV-30

-- HS-RV-4D HS-RV-1C 8

8 8

8 8

8 8

8 C

C -'.17.85 C--

17.&X-C'"

17.&X C

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- The January'30; 19&9 actuations were in response to a turbine trip - reactor scram transient:

~d 8

I 04/29/89

-- HS-RV-lA B

C -

178$ :

04/29/89

- MS-RV-2A 8

04/29/89

- HS-RV-3A 8

04/29/89 HS-RV-4A 8

17. 8$.'-

. 04/29/89 HS-RV-18 8

C -

-',7.&X 04/29/89

-- MS-RV-28 8

04/29/89

-'. HS-RV-38 8

C--;

17. 85', "

04/29/89 HS-RV-48 8

e 2.2 MAIN STEAM LINE SAFETY/RELIEF VALVE CHALLENGES LContinued)

TYPE OF PLANT REASON FOR REACTOR ACTUATION CONDITION ACTUATION POWER ASSOCIATED

~E E<<E 04/29/89 04/29/89 04/29/89 04/29/89 04/29/89 04/29/89 MS-RV-48 HS-RV-28 HS-RV>>38 E

MS-RV-58

-- HS-RV-2C HS-RV-18 C

C' C-C

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05 The April 29, -1989 manual actuations were in response to valves being cycled

.to test acoustic monitors.

The April'9, 1989 -spring actuations were in response to the valves being "simmered" four times for in-situ setpoint verification testing.

06/26/89 HS-RV-3A 06/26/89

-- MS-RV-18 C

C C.-

1 5$

C',.'. 5X The June -26;r-1989 actuations were in response to the valves being "simmered" two times for in-situ setpoint verification testing.-.-

06/27/89

-- MS-RV-4'A

.06/27/89 HS-RV-28 C

C, C:

15f-C C

C..

1.55

-:The June,A-1989 actuations were in response to the valves. being "simmered" two times for in-situ setpoint verification testing.,-

C'6/28/89 HS-RV-18 8

C C==

13.0X-06/28/89 MS-RV-40 8

C C

'3.5%

.06/28/89

-- HS-RV-5C 8

C C

'3 5$

06/28/89

-= MS-RV-4A 8

C EC.

13. 5X

.06/28/89

-. HS-RV-30

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C 13.5%

06/28/89 HS-RV-48 8

C 13.5f 06/28/89 MS-RV-4C 8

C C

13.5%-

06/28/89 HS-RV-58

'8 C

13.55 06/28/89 HS-RV-18 8

C C

1'3;5f

,06/28/89 MS-RV-2C 8

C C-135X-06/28/89

-- HS-RV-2A 8

C.

C-

-13 5C-I The June 28, 1989 actuation of MS-RV-18 was in response to the valve being

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manually actuated,to clear seats and reseat to reduce leakage.

The remainder of the June 28, 1989 actuations were in response to the valves being cycled to verify operability and to test the acoustic monitors. r

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2.2 MAIN STEAM LINE SAFETY/RELIEF VALVE CHALLENGES (Continued)

TYPE OF PLANT REASON FOR REACTOR 3I ACTUATION CONDITION ACTUATION P ONER ASSOCIATED

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08/16/89 HS-RV-5C 8

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10K The August 16,.1989 actuation was in response to the valve being cycled to test the acoustic monitor.

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~. 2.2 MAIN STEAM LINE SAFETY/RELIEF YALVE CHALLENGES (Continued)

T e of Actuation A.

Automatic B.

Remote Manual C.

Spring Plant Condition A.

B.

C.

0.

E.

F.

G.

H.

Construction Startup or Power Ascension Tests in Progress Routine Startup Routine Shutdown Steady State Operation Load Changes During Routine Operation Shutdown (Hot or Cold)

Ref uel ing Reason for Actuation e

A.

Overpressure B.

AOS or Other Safety System C.

Test 0.

Inadvertent (Accidental/Spurious)

E.

Manual Relief NOTES:

-1)

= Remote-.manua1 actuations occurred in supjiort of acoustic monitor

.position indication calibration testTng required by Tecbnica1 Specifications LCO 3/4.4.2.

2)

Spring set-testing was performed in accordance with ASME

-Section XI and Technical Specifications requirement in applica-bility paragraph 4.0.5.

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>slit 2.3 SUHHARY OF PLANT OPERATION INCLUDING SYSTN GENERATOR OUTAGE OFF-LINE CAUSE SHUTDOWN>JN LER DATE TYPE HOURS CODE HETHOD NUHBER l)s

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COHPONENT OS

.A D ACTIO TO PREVENT REC RRENCE.

. 1/7/89 S

82.9 HC HTEXCH-D The Plant was shut down to correct a

condenser tube in-leakage problem.

Repairs were performed and unit returned to service.

1/30/89 F

2/2/89 F

2/18/89 S

3/17/89 F

'3.2 10.4 0

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3 89-002

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5 89-006

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t EB HF I >ll I" ( I >i RB AA ELECON ZZZZ CONROD ZZZZZ The generator tripped at lOOX power due to a fault on an insulator between the main step-up transformers and gen-erator disconnects.

The insulator was replaced and inspection/cleaning was performed on remaining insulators prior to startup.

Generator was removed from service due to main condenser vacuum problems.

Low vacuum was caused by high conden-sate temperature as a result of cool-'ng tower problems.

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(tss Reduced power to perform a

scheduled control rod sequence exchange.

Reduced load due to an engineering analysis which indicated post loca potential 'integr'dted 'ose rate " to control room personnel through venti-1'ation'ystem would exceed Tech Spec Limitations.'fter additional evalua-tion, errors were discovered in cali-brat ion methodology which alleviated the finding.,of the previous analysis.

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SUMMARY

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

GENERATOR OUTAGE OFF-LINE DATE TYPE HOURS I

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thru 6/28/89 6/29/89 S

1456.8 33.5 C

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89-028 RC FUELXX HECFUN The plant was shut down as scheduled for the annual refueling and mainte-nance outage.

Generator was removed from grid to perform overspeed tests on turbine.

A reactor scram occurred prior to com-pletion of tests.

6/30/89 S

I 8/6/89 F

31.7 59.6 3

89-031 CH HECFUN TURBIN Generator was removed from grid to complete overspeed testing of turbine and perform scram time testing.

Reactor scram from lOOX power on Low RPV Level.

Initiated by trip of "8" reactor feedwater drive turbine on low lube oil pressure during testing of I

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8/11/89 F

8/17/89 F

115

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1 89-034

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Plant was shut down to resolve and correct electrical fuse coordination and separation issues on safety-related low voltage motor control centers.

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Reactor scram from 67%

power due to inadvertent actuation of an RPV Low Level switch dur'ing execution of a

Tech Spec Surveillance.

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I II(II 2.3 SU MARY n

Ii i, GENERATOR OUTAGE i OFF-LINE DATE TYPE I

HOURS 9/21/89 F

29

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r j-Plant was manually shut down because of rapidly increasing conductivity due to condenser tube in-leakage.

Two damaged tubes were plugged and plant remained

.down for repair of steam extraction lines.

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<I' iii'I il.:iii OF PLANT OPERATION INCLUDING UNIT SHUTDOWNS/POWER REDUCTIONS (Continued) ij}

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CAUSE SjHUTDOWN LER i}P< i CODE

.NETIIOD NUHDER SYSTEN i

D PONENT CAUSE AND ACTION TO PREVENT RECURRENCE

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HC HTEXCH-D i>>lhi 9/22/89 S

10/23/89 S

12/12/89 S

167.2

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H3 RB PIPEXX-E Plant remained down for repair/

replacement of two failed steam extraction line expansion bellows and condenser baffle repair.

After com-pletion of repair to all damaged components, the plant was returned to service.

CONROD Reduced power to perform a

scheduled control rod sequence exchange.

'"'CONROD """Red'uc'e'd 'PoQer'operform a control rod sequence exchange.

CAUSE 'CODE j

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'TOTAL FOR '1'989 5

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115.0 17.3 10.4

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2.3 SINHARY OF PLANT OPERATION INCLUDING UNIT SHUZDDMNS/PSKR REDUCTIONS (Continued)

SINSNY OF CODES OUTAGE TYPE CAUSE CODE SYSTEH SHUTDOMN HETHOD CODE SYSTEM DESCRIPTION F - Forced A Equipment Failure S Scheduled B Haintenance or Test C Refueling D Regulatory Restriction E External Cause F - Administration G - Personnel Error H Other l Hanual 2 Hanual Scram 3 Auto Scram 4 Continued 5 Reduced Load 9 - Other, AA Air Conditioning, Heating, Cooling 4 Ventilation Controls CH Feedwater Systems 4 Controls EB AC Onsite Power Systems 4 Controls HA Turbine Generator 8, Controls HC Hain Condenser Systems

!L Controls HF Circulating Mater Systems 4 Controls N

Other Features of Steam 4 Power Conversion Systems (not included elsewhere)

IA Reactor Trip Systems RB Reactivity Control Systems RC Reactor Core

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2.3 SUHHARY OF PLANT OPERATION INCLUDING UNIT SHUTOQSS/POWER REDUCTIONS (Continued)

SUHHARY OF COHPONENI'ODES COHPONENT TYPE/CODE Control Rod Drive Hechanism (CONROD)

Electrical Conductors (ELECON)

Fuel Elements (FUELXX)

Neat Exchangers (NTEXCH)

Instrumentation and Controls (INSTRU)

- Hechanical Function Units (HECFUN).

COHPONENT l'YPE INCLUDES:

Control Rod Drive Hechanism Bus Cable

.Hire Condensers Coolers Evaporators Regenerative Heat Exchangers Steam Ge'nerators Fan Coil Units Controllers Sensors/Detectors/Elements Indicators Differentials lntegrators (Totalizers)

Power Supplies Recorders Switches Transmitters..

Computation Hoduies'echanical Controllers Governors Gear Boxes

'aridrives Couplings COHPONENT TYPE/CODE Pipes, Fittings (PIPEXX)

Turbines (TURBIN)

Codes Not Applicable (zczzz) g

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COHPONENT TYPE INCLUDES:

Pipes Fittings Steam Turbines Gas Turbines Hydro Turbines

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While working in the

area, an electrician noticed that the charger was making an unusual noise (with the Plant in normal operation).

Silicon Control Rectifier (SCR) firing board was found defective.

Defective silicone card.

Rectifier caused damage to firing board.

Replaced SCR firing board with new like board.

Performed operability test and verified charger operated proper ly.

Replaced damaged firing board with same.

Replaced defective silicone card rectifier with same.

DSA-C-1C

~ I HPCS Power-Nhile performing Diesel Starting preventive maintenance Air (DSA) on DSA High Pressure

" 'C'oV4"S+8$ diesel '"'

'enerator during Plant operation, it was noted that 'the'unloader valve fitting and'ischarge valve were damaged.

Previous installation or repair had apparently

" over't'ightened 'va I v'e "

and fitting resulting in cracking.

Replaced unloader valve and discharge valve with same.

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- SAFET RELATED E UIP ENT co tinued EQUIPHENT REQUIRING HAINTENANCE SYSTE I ))

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CAUSE ACTION TAKEN DSA-C-'2C HPCS Power-Diesel Starting Air (DSA)

While on tour, an operator noticed air leaking from the air intake on the backup starting ai.r. compr essor for the'PCS diesel generator DG-1C.

Condensation in the line from the after cooler caused rust and corrosion on the valve seats, which created a leak path.

Replaced the suction and discharge valves with same type.

Verified compressor operation.

CRD-CB-PlA Control Rod Drive

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II)I'solation Cooling

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With the reactor. in Hode 1 at a 100 percent power, the racking mechanism was found-broke for the 4160 Volt circuit breaker.

This was discovered while attempting to return the Control Rod Drive "A" to service after repairing a minor seal leak.

'Wl'<le 1tkddbl 6 Ishobti'rig

'<'he failure of the RCIC minimum flow return valve It8"the Suppression Poh't to"op'e'n"br to

close, an undervoltage

'r'elegy in the valve motor starter was'ound'burned up." 'The'robl'em was identified 'during'OVATS test'ing.'

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'u'n'kn'o'w'n",t probably wearout.

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Replaced broken racking mechanism with same type new racking mechanism.

Torqued hold-down bolts to 41 ft/lbf.

Observed racking in and verified pump motor operated.

Replaced undervoltage y'clay.

Tested satisfactorily.

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- SIGNI CAN H IN ENANCE PER ORHED ON SAFET E ATED E UIPHEN cont n ed EQUIPHENT REQUIRING AINTENANCE RPS-EPA-3D SYSTE Reactor Protection System PROBLEH DESCRI TION During performance of-surveillance channel functional test, a

failure occurred in the undervoltage trip functional check for the Reactor Protection System electrical protection assembly.

The undervoltage trip has a setpoint of 110.7 to 108.5 Volts and an allowable value of 108 Volts.

The undervoltage trip occurred at 50.1 volts.

CAUSE Cause of failure traced to defective circuit board.

Cause of defective circuit board is unknown.

C ION AKEN Replaced the circuit board with the same type.

Hodification request issued to request new design of circuit boards to be implemented at next scheduled outage.

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'i cont'acts"dn"K7'relay'.

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ENANCE ERFORMED 0 SAFE RELA ED UI ENT co ti ued EQUIPMENT REQUIRING MAINTENANCE S

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PROBLEM DESCRI TION CAUSE ACTION AKE E-IN-3 SGT-ESH-IA

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Fuses have blown repeatedly.

The electric strip heater bank 2A on the "A" Standby Gas Treatment train low temperature alarm would pot clear.

Pnnunciation

'f'kemper'atur'e's'n'hain Control Room.

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This failure was due to wearout and aging.

The heater wire where it connects to the heaters had frayed insulation causing the wire to short

, against the,)eater "b'okcover"." "'

Replaced static switch gate firing module with same and tested functional.

Replaced shorted heaters with spare heaters.

Taped frayed insulation test satisfactorily.

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WNP-2 found one RHR pump which had hold down bolts that did not meet specifications.

Cause of loose bolts on the RHR pump was attributed to probable thermal

.cycling causing the bolts.to relax.

Retorqued the bolts on RHR pump No. 2'o specification.

Initiated preventive maintenance measures to verify adequate torquing on hold down bolts on all emergency core cooling systems.

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.The pump was running at 15 Hertz at the time..

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input to pump 60 Hertz logic was not

,, paking, up, y)th valve in'losed'osition. "

Inserted spacers between limit switch mounting bracket and limit switches.

Replaced mounting

'bolts with longer bolts to provide sufficient thread eppagemept, to prevent bracket loosening...

,PeI'formed, yoltage check to verify

'ropep operability.'

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CANT HAI E ANCE PE ORHED 0 SA E

ELA ED. E UI HE cont ued EQUI PHENT REQUIRING HAI TENA CE HS-RLY-K81D SSEH Hain Steam PROBLEH ESCRIP ION Hanual isolation relay exhibited excessive noise when running.

CAUS Normal wear expected in GE "HFA" relays.

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TAKEN Replaced with rebuilt relay.

Removed relay to be rebuilt and reused.

RCIC-DT-1 RFM-DT-1B

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i>fs Reactor Core Isolation Cooling Reactor Feedwater Check of suspected leakage on inboard gland seal showed leakage in the interface area between turbine casing and gland seal upper housing.

During unrelated maintenance on Reactor Feedwater Pump '"B", oil was observed leaking in the vicinity of the

)ifdraul)c trip assembly.

Carbon glands degraded.

Sealant between gland seal housing and casing halves showed wear and deterioration.

Infra-red examination of hydraulic trip assembly piping showed leak through of the in-line check

, valve probablg, due to

"- wea'r'out.'

'eplaced carbon glands.

Removed old sealant and applied tempfl ex joining compound.

Performed in-servi ce leak test.

The check valve was replaced with the same type.

Visual inspection of the hydraulic trip assembly was

'p'erformed to verify no leakage;

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2 SIG ICA HAI ANCE PERFORMED ON SAFE Y RE ATED E UI HEN co t ued EQUIPHENT REQUIRING HAIN ENANCE CRD-FCV-2A HS-V-28A S

S Control Rod Drive Hain Steam PROBLEH DESC P

ION During routine observation, the indicated flow through the Control Rod Drive flow controller valve 2A was observed,to be oscillating between 5

and 10 gpm when the controller was in automatic.

Hain Steam Isolation Valve 28A failed locaj leak rate test.

CAUSE The reason for flow oscillations was attributed to a faulty valve positioner.

The cause of faulty valve positioner was normal wear..

Leak by main body seat due to surface imperfections.

ACTIO AKE Replaced faulty valve positioner with spare.

Recalibrated valve and tested satisfactorily.

The main body seat was machined The valve was retested and leakage was acceptable.

PSR-V-X77A/1 Post-Accident Containment isolation PSR-V-X77A/2 Sampling valve for post-accident Radioactive sampling system failed

' "" 'l'6Lil 'le'ak" rath 'test'in(.'

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Probably due to valve disc wear.

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Hating

'surfaces were lapped to remove minor rough spots and burrs.

Valve wa'scasse'mbl ed and leak tested satisfactorily.

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K RFW-V-10B RFW-V-65A Reactor Feedwater Reactor Feedwater During local leak rate surveillance testing,'he reactor feedwater swing check valve 108 would not seal.

During performance of local leak rate testing, Reactor Feedwater supply isolation valve showed excessive leakage.

The Stillman EP soft seat seal had excessive wear.

Valve seat and disc scratched due to unknown causes.

Replaced the Stillman EP soft seat seal.

Established four year equipment qualification schedule for soft seal replacement.

Valve seat and disc lapped.

Valve repacked and torqued.

Local leak rate test performed satisfactorily.

SW-V-165B Standby Service Normal observation found Water the 18-inch spray. pond "B" ring by-bass valve

'""'I'eikirig'6y'thj(,'e'at."'

Normal seat wear.

Replaced seat and thrust collar.

Valve tested 1 l I jj ( ( (1 1

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't'emp'erature'.

1 Standby Service Water valve which supplies co'o'I'in'g water to diesel engine heat

xchanger

.fatl'ed to'pen due to'disc

.separating from stem, most likely due to wearout.

Valve was temporarily re'mov'ed from'erviceand re'p)aced 4ith a spool'iece until a n'w valve can be purchased and ins'tailed.

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STE PROBLEM OESCRI ON CAUSE 2 4 S

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.SM-V-220A HPCS-MO-15 Standby Service Mater High Pressure Core Spray Attempted to operate Standby Service Mater makeup valve to "A" diesel engine cooling water.

Valve would not operate.

During surveillance

testing, Suppression Pool suction valve would not operate via the motor.

Hotor ran, but valve would not move.

Broken stem nut and lock nut resulted in stem assembly failure and valve could not be operated.

The worm shaft clutch gear assembly fell apart due to missing split spacer which acts as a seat for shaft set screws.

The spacer was not installed during manufacturing.

Replaced stem nut and lock nut.

Valve retested satisfactorily.

Replaced entire clutch assembly and worm shaft.

Revised procedures to inspect assembly and re-stake set screws if required.

Notified Limitorque of possible quality control

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Report).

Verified operability of all safety related valves wi'th Limitorque motor

'op'orator's of similar design

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Low Pressure Core Spray Reactor Core Isolation Cooling PROBLEH DESCR P

ION During performance of the annual stroke. times surveillance on the Low Pressure Core Spray

System, valve operator 012 indicated closed-with 3500 gpm through the valve.

Reactor Core Isolation Cooling turbine steam supply valve trips overloads when operated.

CAUSE Limit switches on the valve operator, Rotor were found out of adjustment.

Cause of adjustment problem was unknown.

Limit switch 816 found closed and rotor locked due to attempting to stroke valve closed when it was 98% closed.

AC ION AKEN Limit switches were adjusted to close at.5% open.

Valve functionally tested satisfactorily.

Adjusted limit switch and verified it opened during valve closing.

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. test, motor operator for Reactor Feedwater high pressure heater 6B

" '4'gtlet'Aolationivalge blew all three line

- fuses during open stroke;"'Valve remained (j );

in 'par ti'al'1'y'open position.

The Plant was operating't.

71% power.

l <gi Rotors 1 and 2 on valve position limit switches were out.

of'djustment.

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'close valve.

The valve was retested satisfactorily.

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testing, Standby'Service Mater inlet valve to diesel generator 82 blew a fuse in one of the

')hike li663"dkt iiIII.'ttte ' close cycle. Rotor Ol did not move when valve closed. I<) l h I I i} } 'I'I)) ]ll l } } Hotor leads Tl and T3 we9e" n'ieked by improper installation of motor cover.'djusted number 1 rotor to open with handwheel 3.5 turns off of seated position. " "Tempted for proper operability. The'ires were splic'ed an'd Tl was relugged. /! I ~ r

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2 SIG CANT N E ANCE ERFORHE ON SA E REL ED E UIPMEN cont ued EQUIPHENT REQUIRING AINTENANCE SGT-ESH-1A SGT-ESH-2A DCM-TS-12Al SSE Standby Gas Treatment Diesel Cooling Mater PROBLEH DESC I TIO The electric strip. heater low temperature alarms would not clear. Annunciation of temperature is in the Hain Control Room. During the Spring refueling outage, while testing the 1A Emergency Diesel Generator, the Diesel Cooling Mater Temperature Sensing Element was found leaking. CAUSE The heater wire where it connects to the heaters had frayed insulation causing the wire to short against the heater box cover. Unknown-Probably wearout. C ION TAKEN Replaced shorted heaters with spare heaters. Taped frayed insulation test satisfactorily. Replaced and tested temperature sensing switch. (((((I ( 't l !(((tl(t~(lt l i(( l>>l((i i( i ll >(I I; l>ilhlIi( (((t((l"<<(( ( ("'l ~ ((li ~ (

g

'CQ ~ ) <<))u ) ~, i l i)<< e ~ l<) l < I ~ I I)l <))l I)i< ) <<i<)) I I I'<.l) I <)< )'< ' I ),< 2.4 SIGNIFICAN INTENANCE ERFORNED.O SAF RELATED E UIPHEN co tinued EQUIPHENT REQUIRING HAINTENANCE LD-TS-619B SYS EN Leak Detection PROBLEN DESCRIPTION Mith the plant operating routine observation found the output relay chattering on temperature trip unit 619B of the leak detection system on the main steam line to the turbine building. Contacts on the temperature switch were worn due to normal wear. CAUSE Replaced the temperature trip unit 619B on the leak detection system on the main steam line to the turbine building. ACTION T KE HS-LIS-24B. HS-LIS-24D ! <)-'<'l) l'I I Hain Steam System During performance of the monthly 'surveillance test in the normal operating mode the level. indicating switches , (LIS) would not trip as Vequir'eg" 'Thyme 'gwitche's provide input to the Reactor Protection System'."" " I>> l )I <ll Contacts on the switch were worn due to normal wear. )<i i bl j () l <i)llIill l<l Changed wiring to use a spare switch. Performed surveillance testing. ) ~ ~ ill) l), I<l ~ l ) ))l'a) ) i ~I~ Ilsuil st !i'I,) s,, I INTENANCE E FDRilED'5N SAFE Y REL ED E UI NEN co t I v ue'd AC IO AK HS-LIS-31C Hain Steam System During the Spring Refueling Outage while performing surveillance testing on Reactor Low Level 2 it was noted that the level 'ndicating switch (LIS) had considerable bounce. These switches have a Replaced switch history of failure with the same due to age. type. HS-LIS-37A Hain Steam System During the Spring Refueling Outage while filling the reactor vessel level trip unit exhibited considerable bounce. The microswitch in the Level Indicating Switch (LIS) was not functioning properly causing excessive movement. Probable cause is normal wear. Replaced and tested the micro switch. HS-RIS-6IOD i I la .r ' II I 1 s Ig I'-/I I II I ~ s Radiat)on, e, Dur)ng,ge perforpanre, 'onito'ri'ng' ' '"'o'4"rotitiiie'serve'))lance testing with the plant at power, the main steam (HS) 'line ',r',ad,iation monitor (RIS)'hannel ,"D" downscale and Re'p6'tpr Prot'ect'ion,, Sys'em tr'ips woul'd'no't clear. ~, I sl The swltc would not re's'e't diie to'ormal aging and cyclic fatigue. < hll Removed defective 'drawer and replaced with the same type spare dt aper, Su'rveillance test performed. I ~ I ~ ~

0

SIG IF CAN ]}}}iaa} } l,)}i i I-.. i"!ll }I l '.} I lii!'t~ i ~ l (!<' l }}ll!l } liig}} is) AI E A LE E ORI}ED'0 S I:E } i '>> x i' ~ I, >>! i" i }. j} l I l } <I ~ i ~ ~ s ~ RELA E E UIPHEN cont nue EQUI PHENT REQUIRING MAINTENANCE RHR-PS-16A SSE Residual Heat Removal ialiil e }i PROBLEN ESCRI IO Mith the plant operating normal observation revealed the Residual Heat Removal (RHR) Pressure Switch (PS) which is an Automatic Depressurization System (ADS) permissive failed. CAUSE Cause unknown-probably wearout. C IO AKE Replaced pressure switch with like kind and performed test. SGT-TS-2A11 Standby Gas Treatment Mith the plant at power, the annunciator for a carbon adsorber strip heater low temperature on the standby gas treatment system came on and stayed in. The strip heater would not heat. Contacts for the temperature switch (TS) showed open at ambient temperature. Cause most 'likely due to age or wearout.'eplaced temperature switch and performed a satisfactory test. SLC-LS-600 'tandby'Li'qu'id ' "" Control }~}'~~'}! }.'i} }~- 'Ah'll'e 'pei'for'll}jhg'la 'urveillance test, with the plant operating at power",'the'Standby Li)kaid Cont'r'o'I (SLC) tank level indicating 'met'er'- was'tickihg at'id 'seal'e. i I I i I tab'sQ 'i's'u'n'k'n'own'. 'robably wearout of Level Switch (LS). 'Replaced level indicator and level switch. Completed su'rveillance'test.

4

S G CA T NTE A CE OR ED 0 S E E A EO E UIPHEN continued EQUIPHENT REQUIRING AINTENA CE SW-PS-1A SSEH Standby Service Water PROBLEH DESCRI T ON During Spring Refueling

Outage, while performing preventative maintenance on the Service Water Pump the pump discharge pressure switch was found to be full of water.

CAUSE This was caused by a ruptured tube most likely caused by wearout. AC ON AKEN Replaced pressure switch and performed test. RFM-LIC-620 Reactor Feedwater With the plant at power, during routine observation of the Reactor Feedwater (RFW) startup valve Level Indicating Controller (LIC) meter went to zero when "tapped" by fi'ngers. The performance of the system was ~ <:( igngffycted (.'>l l The meter had an "open" in the circuit. Cause was unknown. Replaced the Level Indicating Controller (LIC) and performed satisfactory retest. ~ ( 1'

A 1 4

0 % ~r (()> ~;t*l I.(1 1 !( l.I (} 1 ~ >1> '((((( A ( i 1 1 1 l((i (>>(l t>l SIGN C N AI A FE Ek (%(> = I I} !gal 1 }.>~ ( 1 fl(I( ](!1(t(,I >>/I( 1 }II ORMED 0 SA E (g I >I>t> I ((,( } (>(1 trt( I 1>.ttl ' ( '>1 ~ Il I'i>l( I I 1 ~ I "(~,(! 1 ( I ( 1 t>I ~ 1%%. '> > ~ I ~ ~ I'ELA ED E U PHEN co t e ~ ~ (% ~ ~ EQUIPHENT REQUIRING A NTE ANCE S S E HS-TR-614 Hain Steam HS-RIS-610B Radiation Honitoring FO I ( ( ~ % ~ Il (lt > ~ i ( ( l(l l(I( -!(II (ii( 1 $ > I 1 Ij>%}lit'ROBLEH 1"(IDE'<CRI 0 '( '(I Ii ' While performing surveillance testing of the Hain Steam (HS) Relief Valve Discharge Temperature Recorder (TR), with the plant at power, it was noted the CAH operated alarm switch actuated but the control room alarm did not annunciate. During a surveillance test, with the plant at power, Hain Steam (HS) Line Radiation Indicating Switch (RIS) would not calibrate. CAUSE Found a loose connection on the switch which actuates the annunciation. I The radiation monitor . drawer circuit was open. This was most likely due to age. 'AC IO AK Tightened loose connection and completed the surveillance test. Replaced drawer with a spare and performed satisfactory test. DLO-H-P/2A2 <<((((( '( }11 ~ (I( > I((( (' I( r, I Diesel((L1jbe.Oil}( }DOHng'r'Uutiiid( "'1 '(1 ( 'bservation, with.the plant operating at power',("it 'was found that thd(}motor'o'p6rating the soak back pump for 'diehel( geherator engine 82'n'the-stahdbg AC pbweY system was 'running but the'impeller has not turning."'I r " 1Thh( 'so'ak'b'a'ck((pump 'ad a broken shear rpin on the motor shaft. Failure appe'ared to be normal wear. 1 ~ '-I'%.' (.I - ~ % ~ ~ ~ ~ I ~ 'Replaced pump motor coupling and shaft shear pin and returned to ser0ice."'% I

h I

EQUIPHENT REQUIRING HAIN NA C ~', GNI C N I S STE (I(!1 lii~ i (i ~ I >>i i } } ~ I (>> I ~

} }lI II ( ll ( ( l ill 11!}lli le! ( ~ l(s, ( Ie l>> E>>IA EE P RFOR EO'0 S PROBLEH ESCR T ON IIllI ! I ~ Itl }I (l il A f3 E CAUSE 'il !( I l '>> ~ ~ ( I l ll l }>> U E co 't I ue C 0 KE DLO-H-P/6 RCIC-H-P/3 Diesel Lube Oil Reactor Core Isolation Cooling The motor driven oil pump (P/6) on the High Pressure'ore Spray Diesel was observed to have high vibration. The Reactor Core Isolation Cooling (RCIC) water leg pump was observed running noisy and vibrating. The vibration was caused by a bearing failure caused by the previous failure of the pump to motor coupling.. The pump motor shaft = was out of alignment. Replaced motor with like kind. Realigned Pump and replaced motor

bearings, RPS-H-HG)

Reactor Protection A loss of power was experienced on Reactor Protection System (RPS) bus IIAII~ A motor bearing on the motor-generator set (RPS-HG-1) failed probably due to normal wear. Replaced motor with spare and tested. DLO-P-10 I (i(! I >> I "l (I ( >>II l ( ~ I ll l(! I II >>li ~ I" l!l I '}l>>}(( } j ( (l( (>>('(}}}l (i(} fl! I I l Diesel Lube Oil During normal observation it was noted the }}}5'diesel generator lu58 oil"soak'ack pump would trip on overload. Ii, ~ >> ~ ', 'iI, (I I N I! } Ii(( (I l Il>>((>>,l(.'l( t, A motor brush was excessively worn causing reduced brush te'ns'ion which resulted in brush to 2'ommutator'arcing. (i Replaced motor with spare and tested sal't)sfactorily.

f I l II t P

I 'ill ) i li .,ol ) ~ \\'I ~ I ~ a ~ ) ~ 'I l i, I I ~ ~ .4 SIG I C N AIN NA CE ERFORHEO 0 S FE EL EO E U EN cont e EQUIPHENT REQUIRING HAIN ENANCE S S H PROBLEM DESCRIPT ON CAUSE C IO AKEN RCC-P-18 Reactor Building Closed Cooling An operator on tour noted the outboard 'earing on Reactor Building Closed Cooling (RCC) Pump "1B" was running hot. The cause was unknown but thought to be excessive lubrication or expected end of life for the bearing. Bearing was replaced. RFW-P-1B Reactor Feedwater During retest of the Reactor Feedwater Pump 1B (RFW-P-1B) following maintenance the thrust bearing overheated. The inboard oil seal ring had been improperly installed and caused uneven loading on the bearing. Replaced bearing and associated oil seal ring. RFM-P-18 !>>~'< ~ ('il / I I! 'With the plant at power a high vibration alarm was received on Reactor Feed Pump 1B;(RFM-P-IB) ~ < / ( ( J c ~ < I I 1 I 4~ I ff.I ~ I f I 'J t 'll ( I <>III t I >>~ il .A small orifice that. admits oil to the thrust bearing was ,found plugged> Tfle caUs'e of the plugging" was believed to be filter particles caused by filter ch'a'ngeout. Bearing was replaced and lubrication drained. The procedure was changed to require draining prior to filter,changeout. I 'L 1 I II ~ I /I l <<,i('. I ( I ~ ~ lt ( ~

2.5 INOICATIONS OF FAILEO FUEL INTROOUCTION - In accordance -with the commitment and requirements described = in the MNP-2

FSAR, Section 4.2.4.3, a

visual inspection of discharged fuel from MNP-2, Cycle 4 was'erformed on October 5-10, 1989. The purpose of -the inspection was to= veri fy assembly and fuel rod structural.- integri ty. In addition, although . not. a commitment, a visual inspection of -selected discharged fuel channels was performed at the same time.

SUMMARY

OF INSPECTION RESULTS A. total of ten-assemblies and two channels discharged at the end of cycle 4 were inspected:

No evidence of mechanical

damage, geometric distortion or rod

-;-.bow were observed.

All rods inspected appeared properly seated in the lower , tie;plate. All: spacers appeared to be in proper position. The fuel exhibited nodular corrosion which covered portions of the clad: on fuel rods which were cleaned for =-.inspection. The extent of coverage did sot.appear to be markedly changed from= previous inspections although some 'instances""of clad surface . :roughness .-were observed on profile. The - assemblies.; uncleaned, generally ~conformed to-.General Electric (G.E.) visual standard,-2.

However, in the

==.single instance where a fuel rod was cleaned of.surface

crud, the observed

~ nodular= corrosion was substantially less than the 100$ coverage associated . with visual standard 2. Based on comparisons, with end of cycle 3 fuel, it appears that-.nodular corrosion is still taking place=.but the rate of growth appears to be low. l' Fretting marks . appeared on several assemblies, particularly in the span 6

- region.

(See Results. of Fuel Examination. section-for span location defini- -,tion). Investigation as to the cause of the scratches in. the 6th span has --:- determined that they are caused by contact with the-upp'et bracket of the WP-2 south -fuel preparation machine during de-channeling following discharge. The -,--=-,- other-scratches are assumed to be caused by foreign-'bjects or by rubbing -.--"===-against-the spent fuel storage locations during fuel movement.:-None of the -'scratches . appeared to have sufficient depth to be of -concern. Two of the inspected assemblies appear to have contained foreign material. Some scratches were noticed on the bottom of the lower tie plate on some assemblies which might be indicative of a slight bundle-rotation. On some assemblies, the tie rods are apparently growing faster than adjacent fuel rods. This is causing an apparent loss of tension on the tie rod hex nuts. One instance--of finger spring damage was recorded on a-photograph. The observed damage was most probably caused by fuel handling after--de-channeling although spring relocation would have the same affect. - The missions of the springs while in the core was not impacted. Fuel rod E-4 of Fuel Assembly LT3 511, which is the non-spacer/capture water ~ ~ ~ ~ ~ ~ ~ ~ ~

rod, has what is either a clad imperfection or a foreign object wrapped around it at least for 270'.

From the photographs, it is impossible to be differen-tiate further;

The phenomenon occurs low on the rod in what is the natural enrichment zone of the core. t t

The inspected channels all exhibit a coating of flake-like oxide material. Some miscellaneous scratches were observed. There was no evidence of mechan-ical damage, holes in the channels or control rod shadowing effects. SELECTION OF ASSEMBLIES AND CHANNELS Our.ingT the spring 1989 refueling

outage, 136 original core f.uel assemblies were, discharged-.

Ten of these assemblies and two channels were selected for visual=- inspection. The ten assemblies represent greater than-5 percent of the = discharged fuel and are representative of the highest-burnup assemblies in the discharged batch. Visual examination of the peripheral fuel rods of these assemblies included observation for cladding defects, fretting, fuel rod

bow, missing components, corrosion, deposition and geometric.distortion.

The selected assemblies are all high enriched (2.19 =:weight. percent U-235; initial). -The two channels selected were representative of. the highest exposed channels discharged. Some-characteristics of the selected assemblies and -. channels are shown in Table l. TABLE 1.0 CYCLE 4 DISCHARGED FUEL ASSEMBLIES SELECTED FOR EXAMINATION FUEL ASSEMBLY 'DENTIFICATION LJT 522 LDT 770 LJT 398. LJT 525 LJT 414 LJT 713 LJT 604 LJT 511 LJT 737 LJT 794 CHANNEL IDENTIFICATION 71895* 71473+ 25,817 25,217 26,026 26,006 25,961 26,209 25,991 25,971 26,083 22,33& X -X -X MET ULTRA EXPOSURE - SIP SONIC ~HWD/NT TEST TEST SUSPECT CELL X "The channel has the same exposure as the assembly, it was on." - 36

l t 4

- The ten assemblies inspected have exposures ranging from,22,338 to 26,209 HMO/HT. The inspected assemblies include assemblies which were sipped

and, in some

cases, ultrasonically tested for fuel leaks during the R-4 outage.

In

addition, some assemblies were located in fuel cells suspected of containing fuel leaks as determined from flux tilt testing.

INSPECTION TECHNI UE - -; The - poolside -visual examination was performed with an underwater.periscope system. with results of the fuel inspection being recorded on the Nuclear Fuel =--'-'ransfer L'ist in addition to the inspectors working= notebook. Two sides of each -fuel 'assembly were viewed. Photographs of selected points of interest -were taken.' total of eight-eight photographs of the examined fuel and chan- '=nels. were taken. Fifty-four of these photographs-were successful. As the . Nuclear Fuel'ransfer List log and accompanying notes constitute the permanent " :record of the'nspection, successful photographs oF.all inspected locations - -::.are 'ot required although certainly desired. =.The inspection procedure " =.involved moving the selected selected fuel assembly-in a vertical direction -past the: fixed periscope; This was accomplished by. raising the fuel assembl'y 'ut'f the spent fuel rack with the fuel. handling mast on the refuel bridge.

'Channel

'inspection was performed in a similar manner;= A ptece of abrasive material was"used to remove the heavy layer of red-.'.colored surface crud from some of the edge-fuel;rods in order to assess the rate-of nodular growth. INSPECTION CRITERIA -..=Visual inspection of the selected fuel assemblies was performed according to ~ the following criteria: -o Proper rod seating in the lower tie plate o Rod bow and spacing o Spacer location and perpendicularity o Finger spring condition a Condition of ti.e rod hex nuts and other structural components o Nodular corrosion and crud scaling o Fuel rod fretting The channels -were inspected for spallation, wel'd failures, cracks and other structural

failures, and buildup of oxidation.

The results are discussed below. r

t

RESULTS OF THE FUEL EXAMINATION ~ ~ ~ ~ ~ With one possible exception, the inspected fuel assemblies exhibit good apparent integrity. The upper tie plates were level, fuel rod springs had ample compression

space, the rod nuts appeared snug except in one or two instances and all the fuel rods observed were properly seated in the lower tie

=- plate.- The -spacers appeared perpendicular to the fuel rods and were properly located. Host-finger spring sets observed displayed no damage.. Hinor finger - spring damage -was observed in isolated cases. The grid.spacers in general exhibit -a heavy nodular buildup. Exceptions to the above statements along with =specific phenomena observed on specific assemblies are discussed below on an assembTy basis. The channels inspected displayed no 'instances of spalla-tion, ':cracking -'r other loss of integrity. They did exhibit a heavy oxide corrosion covering on all non welded surfaces. 'uring the inspection, the Nuclear Fuel Transfer List was maintained, field .notes were obtained and photographs were taken. Developed photographs were '=; - -. not obtained for all inspection points. During the inspection activities, it .- was discovered that the, photograph taken of the first.5 assemblies were not correctly exposed. No attempt was made to re-examine and photograph these assemblies except for those cases where the inspection -notes indicated a potential-.anomaly. The following description of the specific assembly inspec- .tion-is. based-on the Nuclear Fuel Transfer List, the field notes

and, where available, the photographs.

,In. discussing specific fuel assembly observations, the following convention will.be used. With the threaded post of the assembly in the upper left

~-corner, -the top is side A, the right side 8, the bottom side C and the left

~.side= D, (See Figure 1). Fuel pin locations are identified as'ollows: With the threaded=- post in the upper left corner, the fuel rod columns are labeled = from left to right A,B,C,D,E,F,G and H. The fuel rod; rows are 1,2,3,4,5,6,7,8 from-top to bottom. Fuel spacers are numbered 1 througlr 7 beginning at the lowest spacer and the regions between

spacers, called

spans, are numbered 1

through 8 beginning at the bottom of the fuel assembly; Nodular corrosion was observed to some degree on all of the-assemblies that were inspected.. The assemblies, uncleaned, generally conformed to G.E. visual standard =-2; However, in the single instance where a fuel rod was cleaned of surface crud; the observed nodular corrosion in no way approached the 100$ nodular or sheet coverage associated with visual standard 2. fuel Assembly.LJT 552 was inspected on sides A and B. No unusual mechanical features-were. observed. Scratch marks, later determined to be associated with de-channel.ing in the south preparation machine, 'were observed on both sides of the-bundle just below grid 6 as has been seen before. The outer appearance of the fuel generally conforms to G.E. visual standard 2. Fuel Assembly -L1T 770 was inspected on sides A and D. No mechanical anomalies were observed. The uncleaned fuel conforms in appearance to G.E. visual standard 2. . Preparation machine scratches were observed on both sides A & D just below grid 6. A small foreign object was observed at the span 4 height on fuel rod A-5 (side D).

Fuel Assembly LJT 398 was inspected on sides A and D. No mechanical anomal'ies were observed.. The uncleaned fuel appears to conform in appearance to G.E. visual standard 2. Scratches made by the south fuel preparation machine were observed below grid 6 on side.D. Fuel Assembly.=LJT 525 was inspected on sides A and .C. No unusua') mechanical .features were: observed. The appearance of the uncleaned fuel assembly appears to be consistent with G.E. visual standard 2. Scratches were observed on side A, span 1 and span 5, which could have been caused by the spent fuel racks. Side A.was the-second side inspected after assembly de-channeling. Scratches were observed on the nose cone of the lower tie plate; side A, which appear to have been caused by a slight rotating movement of the-. assembly. Fuel Assembly LJT 414 was inspected on sides C and D. No mechanical damage was, observed;. The appearance of the uncleaned fuel= appears to be consistent with G.E. visual standard 2. A foreign object was observed on the finger springs on side 0. Specific views were made of the lower tie plate, grid 5 and-upper tie plate of side C and the lower tie plate "span 1 and span 4 of =- side 0. The assembly was later re-inspected. On re-inspection, the foreign object was missing. Photographs were obtained of the lower tie plate span 5 and upper tie-plate of side C and the lower tie plate, span 1 and span 4 of -side -0. .Some finger spring damage was noted on side B as seen from side C. - The type. of damage observed is caused either by contact with a fuel rack after de-channeling or by a loss of spring tension. In either case, the mission of the. finger springs while the fuel in the core was not impacted. Some evidence of minor bundle rotation can be seen on this photograph. Fuel Assembly; LJT 713 was inspected on sides A and. 0. No mechanical damage was observed.. The appearance of the uncleaned fuel';most closely matched G.E. ~ ~ visual standard 2. Preparation machine scratches were observed below grid 6 on span 6 on'ides A and 0. A hex nut is shown backed off on Side A. The -- backed off.appearance is most probably caused by greater differential fuel rod growth-of the.tie rod. There does not appear to be a concern for loss of the =-= nut. Photographs were taken of span 1, span 4, span 6 and the upper tie plate of side A and span 2 and span 6 of side 0. Fuel Assembly-LJT 604 was inspected on sides A and 0. No-mechanical damage was observed.. A hex nut on a tie rod is shown backed off on the view of side 0. The appearance of the uncleaned fuel most closely matched G.E. visual standard 2.. Preparation machine scratches were observed on span 6 of side A. Rotational type scratches can be observed on the lower tie plate view on side C. Photographs were taken of the lower tie plate and span 6 of side A and of the lower tie plate, span 3 and the upper tie plate of side D. Fuel 'Assembly LJT 511 was inspected on sides A and 0. The appearance of the uncleaned fuel most closely resembled G.E. visual standard 2. No mechanical damage. was observed other than the specific phenomena discussed below. Oe-channeling scratches were observed in span 6 on both sides A and 0. A roughness of the clad surface perhaps associated with enhanced corrosion may be observed, in span 6 of side A. A foreign object-or.clad bulge was observed between fuel rod columns E and F as viewed from side A on rod E-4. This is the water rod which is not the spacer capture rod in the G.E. fuel design for the -MNP-2 initial core. When viewed from side D, this same phenomenon can be -39" r

I ,I f>

seen between -fuel rod rows 3 and 4 and fuel rod rows 4 and 5. This object occurs in the span 1 region near the bottom of the lower end cap. This loca-tion appeared to be in the natural uranium blanked region of the core. Photo- . graphs of the lower tie plate, span 3 and span 6 of side A (two photographs of the lower tie plate region) and six photographs of span 1, span 6 and the upper. tie plate (four photographs of'pan

1) were taken.

This;fuel assembly --was located in a suspect cell during cycle 4 as.determined by flux tilt testing. --Fuel Assemb1y--LJT 737 was inspected on sides A and D. -No mechanical damage - was observed. The uncleaned fuel rods most closely conform to G.E. visual . standard 2. Some clad roughness can be observed in. profile on span 4 of side A. -.Preparation machine scratches can be observed on span-6 of side A. One of'- the: hex. nuts appears to have backed off a very small amount in the upper tie plate view: of side 0. Photographs were obtained for span 1, span 4 and span 6 of side A and span 1, span 5 and the upper tie plate of side D.-- -Fuel Assemb1y-:LJT 794 was inspected on sides A and 0. No. mechanical damage =. was observed The fuel (uncleaned) most closely resembled G.E. visual .- standard 2.. Some hex nuts loosening may be present;. Photographs were taken . of span:1,-. span 4 and the upper tie plate region of side A and. the lower tie - -- plate, span 4 and the upper tie plate of side 0. Th'en, a section of span 7 of side:- D was cleaned with an abrasive material (Scotch brite). After cleaning, --:"some ophite. oxide nodules could be observed on the clad surface. The nodular -. coverage - is=.estimated at less than 30$ which is less -than the -lOOC coverage usually associated with visual standard 2. ~ Channel 71895,. which has been resident on fuel assembly LJT 737 since initial --=~ startup. .was inspected on sides A and 0. ,This channel has an exposure of

---

.- 26,083-NNO/NT-and was measured for bow just prior to inspection.

The channel =-- passed the-measurement criteria. No mechanical anomal.ies were observed on the channel. =-,It--was covered with a heavy uniform oxide =layer, white in appear- = - -ance-, except--for the weld

seam, visible on-side A,.-which exhibited occasional

--..: whi,te oxide -nodules. Photographs top of side A and the 'bottom and middle of side 0 were obtained. Channel 71473,. which has been resident on fuel assembly LJT 794 since initial

startup, was inspected on sides A

and D. This channel has an exposure of - 22,338 HNO/NT-and was measured for bow just prior to inspection. The channel passed the measurement criteria. No mechanical anoma'l=ies were observed on the channel.= It was covered with a heavy white oxide layer except for the seam weld region ,The seam weld visible on side A, had occasional white modules ~ -but was. mostly--clean of oxidation. Photographs of the top, middle and bottom regions of both side A and side 0 were obtained. /l

~ ~ SPAN GRID 5 5 CHANNEL FASTENER A 8 D E 3 -0 FIGURE 1. FUEL ASSEMBLY MAP SHOHING LABELING CONVENTION - 41,

ll t h t f I

2. 6 PLANT MOOIFICATIONS Federal Regulations (10CFR50.59) and the Facility Operating License (NPF-21)

.allow changes; to be made to the facility and procedures. 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 .=- Commission (NRC)

approval, unless the proposed change,,

test or experiment =.= involves a change in the Technical Specifications incorporated in. the license or an:wnreviewed safety question. In accordance with 10CFR50.59, -summaries of .-the" permanent-. design changes and temporary plant-. modifications completed

in-1989are provided.

Included are summaries of the safety evaluations.

~h

2.6.1 PLANT OESIGN CHANGES The'following plant design changes were completed in 1989 and reported in accordance with 10CFR50.59. These modifications were -evaluated and it was determined that they did not (a) increase the probabi.lity of occurrence of an accident or malfunction of. the equipment important -to safety, as previ-'usly evaluated in the WNP-2 updated Final Safety Analysis Report (FSAR), (b) create the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR, (c) reduce the:margin of safety as'efined in the basis for any MNP-2 Technical Specifications, or (d) require a change to the MNP-2 Technical Specifications = and as

such, prior NRC approval was not required.

e 43-

Plant Desi n Chan e 84-0190 t Plant Design Change 84-0190 was initiated to modify breaker control logic to allow-operation of one Plant Service Mater (TSW) pump during a LOCA when off-site power is available. This modification will minimize loss oF TSW pumps and facilitate plant recovery from a LOCA. To prevent an undesirable bus transfer due to voltage transients caused by large motor starting during a LOCA with power supplied. from the startup trans-former (TR-S),- this design change provided ten second-time delays to the auto- -'atic 'starts of ECCS pumps on a LOCA initiation. However, the start interval between ECCS pumps on the same division remained the same (i.e., 5 seconds). The safety analysis bounding times are unchanged. In addition, the automatic-trip of SH-75 and SH-85 on a LOCA signal was defeated and automatic shedding - of SH-72 and SH-82 on a LOCA signal was provided. Automatic trip of SH-75 and SH-85 on loss of offsite power was retained. As a result, this allows for continued operation of the TSW pumps during a LOCA and with offsite power available.. Also, an electrical interlock was provided to prevent starting of the second TSW pump during a LOCA. =-= This modification did not result in a change to the MNP-2 Technical Specifica-Nions or involve an unreviewed safety question because the margin of safety - ---= was not reduced or the possibility of a different malfunction as defined in the -basis-- for -any Technical Specification was not increased. Redundant safe shutdown-equipment and systems will always 'remain operational-and the required .system

response, times were not affected.

gualitatively, the probability of a . successful shutdown following a LOCA with offsite power available and a TSW =pump available. was increased, which qualitatively decreases the overall core damage risk. Plant Desi n Chan e 84-0623 --= - Plant Design Change 84-0623 was initiated to modify-two Reactor Mater Clean-up (RMCU) valves. to decrease their stroke times for containment isolation. Pre-vious.ly,- the--valves were 'blocked" from stroking full'.open to reduce the stem --travel--required to close. Blocking provided stroke times within the maximum allowable valve closure times. This modification changed the design of the Limitorque operators to the RMCU-V-1 and RMCU-V-4 valves to increase the stroke speed of the valve. This -design satisfied the maximum al'Jowable stroke time without limiting the valve opening. This modification did not result in a reduction in the margin of safety to the MNP-2 Technical Specifications 'r result in an unreviewed safety question because the valve closure times for the two RMCU containment isolation valves remained within the maximum allowable closure time. 44

'l

Plant Desi n Chan e 84-1360 Plant Design Change 84-1360 was initiated to allow maintenance to be performed ~ ~ on one fire protection system in a given area without removing the fire 3 protecti on alarm capab i 1 ity of a redundant system. This reduces the pos-sibility of an undetected fire in areas where safety equipment is located during periods of maintenance on fire protection equipment. This modification removed cross-connections between fire control panels FP-,CP-FCP1 and FP-CP-FCP2. When an alarm is activated on one fire suppression system for a given area due to actual conditions or maintenance activities, the alarm covering the same area will remain functional to alarm on actual conditions only. This-modification did not involve a change to the WNP-2 Technical Specifications or involve an unreviewed safety question because: (1) the margin of safety in Technical Specifications was not

reduced, and (2) this change provides for increased fire protection during maintenance activities.

Plant Desi n Chan e 85-0093 ,Plant Design Change 85-0093 was initiated to reduce the maintenance frequency on the Diesel Starting-Air (OSA) system for the High Pressure Core Spray (HPCS) diesel engine and increase the reliability of the OSA to the HPCS diesel engine. Two high maintenance valves were removed from the OSA system which reduced the overall DSA maintenance. 'lso, OSA piping was rerouted to make redundant engine starting equipment=completely independent. As a result, the overall reliability of the HPCS system was increased. This modification 'removed the crosstie between air receivers OSA-AR-1C and OSA-AR-2C, which included a globe valve (DSA-V-5) and a check valve '(OSA-V-6), respectively. A, new 2-inch line was added to the line coming from air receiver OSA-AR-lC, another 2-inch crosstie line was

removed, and a

block valve (DSA-V-84) was added to an existing crosstie line to make redundant OSA equipment completely independent. This modification did not result in a change to WNP-2 Technical Specifications or involve an unreviewed safety question because: (1) the modification increased the reliability of the OSA which increased to overall reliability of the HPCS

system, and (2) the boundary conditions for the FSAR evaluations remained unchanged.

'e e

Plant Desi n Chan e 85-0184 Plant Design .Change 85-0328 was'nitiated to increase the reliability of the portion of the leak detection system that monitors leakage from the reactor coolant pressure boundary. Monitoring is performed by sensing temperature -..increases and = initiating alarms and isolations. .The-. previous hardware had been-causing 'n inordinate number of system isoIations:.caused by spurious trips. The-. old system (Riley Model 86) was replaced with a General:..Electric NUMAC system (LD-MON-1A, LO-MON-18, LO-MON-2A, and LO-MON 28). In.addition, the system recorders (LD-TRS-608, LO-TRS-611, LD-TRS-622-,= and -LD-TRS-624) were replaced. with more reliable equipment; The new, monitors provide = automatic self-testing -.every 30 minutes that test all channels--and functions of the monitor.- In addition, there is constant-monitoring-for-power failure and open '/C signal.. .The isolation logic and devices external -to the temperature -monitor units -were not changed. A preoperational-.test was-performed on the new equ'ipment prior to return to service. The change of hardware involving the leak detection system'id not result in a ..-change-- to.= the NNP-2 Technical Specifications and:the - unreviewed safety -..-="...-question: evaluation concluded: (1) the function and-performance-of the Leak Detection, did not change, (2) the margin of safety:provided in the technical .- specifications was not changed, and (3) the boundary-conditions for the FSAR evaluations were not changed. - 46

Plant Desi n Chan e 85-0328 Plant Design Change 85-0328 was initiated to remove a highly radioactive scc-a ~ tion of-piping ("hot spot") in the drywell under the reactor pressure vessel. Removal of the "hot spot" significantly reduced radiation exposure to person-nel performing..maintenance activities in the immediate

area, particularly on the control rod drives.

This modification removed a two-inch drain line (2" RRC(51)-1) between Reactor - Recirculation (RRC) line 4"RRC(51)-4-3 and the Equipment ;-Drains Radioactive (EDR) -system:header 4"EOR(47)-1 including valves RRC;V and RRC-V-30. Caps were installed -on the tees from the 4-inch RRC and ZDR 1-ines to maintain the reactor pressure boundary and seal off the opening to. the EDR header, respec-tively.. =This line served no useFul function during operation or shutdown. .=.-The line would ease draining of the reactor vessel during decommissioning but it is not required to achieve this draining. .Implementation of this modification was done through =the-use of a reactor .,= ;: vessel bottom head drain plug for isolation between the reactor vessel and the 2-inch RRC drain line. The bottom head drain plug must also perform as a pressure bouhdary for hydrostatic testing of the:spool piece ,welds in the --. 4-inch RRC 1:ine to 1172 psig. The plug was back pressure tested to 1400 psig

=--.=-Co demonstrate-. acceptability.

A 10CFR50.59 evaluation determined there was no unresolved safety question related to the implementi.ng =activities or the plant - configuration-. during implementation oF this modification because: (1) the boundary-conditions of the FSAR evaluations were not: changed because a 2-inch leak through-the bottom head drain at reactor shutdown conditions under atmos- -pheric pressure are well within the postulated design -basis conditions for the Smal.l Break <OCA, and (2) the implementing activiti~s did-'not rj.duce the mar-gin of safety in the MNP-2 Technical Specifications.-

--.- This modification did not result in a change to the MNP-2 Technical Specifica-tions or.involve an unreviewed safety question because:

(1) removal of the --. drain= line -and--valves reduces the possibility of an'inadvertent leak from the -.reactor pressure

vessel, and (2) the boundary conditions of the.

FSAR evalua-tions were not changed. Plant Desi n Chan e 85-0360 Plant Design Change 85-0360 was initiated to modify Class 1E and some non-lE .4.16 KV and 6.9 KV Mestinghouse circuit breakers. A failed spot weld in the breaker linkage allowed the linkage to, decouple. This; had the effect of ren-dering electrical control circuits as well as anti-pump circuits (to prevent multiple breaker closures during faulted conditions) inoperable in affected breakers. = This design change fabricated and installed new linkage =in all Class 1E 4.16 KV and 6.9 KV Reactor Recirculation Pump Mestinghouse breakers. The new link-age piece used a pivot pin assembly that was plug. welded instead of spot welded. g-This modification did not result in a change to the MNP-2 Technical Specifica- . tions or, involve an unreviewed safety question because this design change cor-rects a potential problem with auxiliary switch linkages for 4.16 KV and 6.9 KV Mestinghouse

breakers, and thereby,.

reduces the probability of a malfunction of equipment important to safety.

Plant Desi n Chan e 85-447 & 86-0557 Plant Design Changes 85-0447 and 86-0557 were initiated to increase the time delay in the ground fault relay settings to prevent spurious alarms from power transients. -This.modifica'tion changed the time delay on the GRC= type ground fault relays for motor control centers, and 4.16 KV and 6.9 KV switchgears from 2 cycles to 30 cycl'es. The ground fault relays provide alarm only and do not perform any safety function. This modification did not. result in a change to the MNP-2.Technical Specifica-tions or-result in an unreviewed safety question because: (1) the margin of 'afety was not reduced in the Technical Specifications, and (2) the boundary conditions of the FSAR evaluations remained unchanged:-- Plant Desi n Chan e 86-0218 Plant Design. Change 86-0218 was initiated to eliminate one of two fire sup-pression manual-pull stations in the Communications .Room (525-ft level) of the Radwaste Building because the station is inaccessible= =-=The one remaining pull station in the area is much more accessible than the-one removed. This-modification removed fire protection manual pull-station FPHPS-28/31. ---.-This modification did not result in a change to the MNP-2 Technical Specifica-tions or involve an unreviewed safety question because this is not a safety- . related

system, the modification has no affect on safety

systems, and removal

. of'-the jul.l,station did not reduce the margin of safety in the Technical Specifications. Plant Desi n Chan e 86-0332 . =-Riant Design-.Change 86-0332 was initiated to provide .increased assurance

-=against..overpressurization of the diesel fuel day tank-for each of. the three d.ivisions-due..to failure of the transfer pump to stop on'igh level in the day tank.-.- This- -increases the reliability of the diesel fuel oil system, thereby increasing the reliability of the diesel-generators.

,This modification provided a passive overflow drain line. between the diesel fuel oil day tank and its associated underground storage tank for each division. This modification did not result in a change to the MNP-2 Technical Specifica-tions or involve an unreviewed safety question because: (1)'he overall reliability of the diesel-generators was increased, and (2) the boundary conditions of the FSAR evaluations remained unchanged. - 48

u t I t

Plant Desi n Chan e 87-0031 Plant Design Change 87-0031 was initiated to modify motor-operated valve interlocks on the Residual Heat Removal (RHR) System to minimize the probabil-ity of inadvertent partial draining of~ the . reactor pressure vessel to the suppression pool. The existing design did not pose. a safety threat of com-pletely draining the reactor pressure vessel because the water level would not drop below the top of the jet pumps. This. design change provided additional electrical control interlocks of the - auppression pool spray and,test, return valves, RHR-V-24A & -24B and RHR-V-27A & =278, respectively, with the RHR suction valves, RHR-V-6A & -6B. The exist-ing interlock 'prevented opening of an RHR suction line valve if a suppression' pool. spray - or: test return valve in the same division is not fully closed. This modification provided interlocks against the reverse process. That is, the -RHR suppression pool spray and test return valves in a given division are prevented from opening if the suction valve in the same division is open. There were"-no modifications to the WNP-2 Technical Specifications as a result of this design.change. This change did not involve an unreviewed safety ques-tion: because the probability of maintaining a safe shutdown condition is increased. and =the margin of safety in the Technical 5pecifications was not reduced.. Plant Desi n Chan e 87-0114 Plant Design Change 87-0114 was initiated as a human-factors improvement to minimize the possibility of operator error by changing the physical location of. selected=power bus control switches. Changing -the-switch -locations made =lineup of the'witches relative to the sequence of manual operation consistent with all'ther-similar power bus control switches, 'improving the human-to-control board =interface. This modification reduces the possibility of a reactor scram-due to operator

error, which could occur if the switches are operated out of sequence.

.This modification exchanged location of the following two pairs of switches on =- -- Boa'rd- "C- -that control power between the startup:- transformer. TR-S and bus SH-6; and the-normal transformer TR-N2 and SH-6: '(1) synchronizing selector switches CB-S6 and CB-N2/6 exchanged locations, and (2) startup feeder CB-S6 and normal feeder CB-N2/6 exchanged locations. . This design change did not result in a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question because: (1) all wiring

==- remained the same; (2) only the location of switches

changed, and (3) this modification-reduces the possibility of a reactor scram due to operator-error.

Plant Desi n Chan e 87-0316 Plant Design Change 87-0316 was initiated to provide annunciation to the main Control Room-operators when the transfer switch for the second of two Residual Heat Removal (RHR) boundary isolation suction valves.from the Reactor Pressure Vessel-(RPV) is not in its required position of "Emergency". The new annunci-ator will alert operators that the RHR System is=-incorrectly - lined-up and could lead to an overpressurization of the RHR System. This. modification changed existing wiring to energize an annunciator when the transfer switch for RHR-V-8 (switch number E-RHS-ARST24) is not in its required position of "Emergency" during Modes 1, 2, or 3. The transfer switch - for RHR-V-8 must be in the "Emergency" position during normal. operation to prevent-it from inadvertently opening simultaneously with RHR-V-9. during cer- 'ain postulated: accident conditions. Simultaneous opening of the two valves - -during normal operation would lead to overpressurization-of-the RHR-System. -". This modification did not result in a change to the WNP-2 =Technical Specifica-tions or involve an. unreviewed safety question because: (1) the overall reliability of the RHR System was

improved, and (2) the boundary conditions of the FSAR evaluations remained unchanged.

Plant Desi n Chan e 88-0038 Plant-Design Change 88-0038 was initiated to replace selected high maintenance -=. . -radiation and turbidity recorders with low maintenance recorders. Extensive manhours and spare parts were required to maintain the mechanical type recorders. -=. This=modification replaced five existing recorders with Yokogawa recorders and - installed-one-- additional new Yokogawa recorder.- The five replacement ,=

-.-;. recorders

consisted of one. turbidity recorder on the-Reactor Feedwater (RFM) -- system:-(RFM-TBR-622) and four radiation recorders on the Area Radiation Mon- - itor (ARH)- system (ARM-RR-600); Off-Gas (OG) system (OG-RR-601, OG-RR-604), and Reactor: Building Exhaust Air (REA) system (REA-RR.-603).= The new recorder , was installed-for radiation recording on the Standby.=Service Mater (SM) system (SM-RR-2). These modifications did not result in a change to the WNP-2 Technical Speci-fications or involve an unreviewed safety question because: (1) the maintain-abilityy and rel iabi 1 ity of the recorders were incr eased; (2) the boundary conditions of-.the FSAR were not changed, and (3) the margin of safety in the MNP-2 Technical Specifications was not reduced.. 0

Plant Oesi n Chan e 88-0056 Plant Oesign Change 88-0056 was initiated to add over pressure relief for the Reactor Building Outside Air (ROA) System. This prevents destructive over-pressurization of the reactor building as had occurred on February 14, 1988. This modification installed a relief damper for the ROA Heating and Ventila-tion Unit (HY) ROA-HV-1. In the event the Reactor Building Exhaust Air (REA) System fails -to start or initiation lags that of the ROA System could result =-in increased reactor building

pressure, the back draft damper.

provides a relief path back to the fan suction to prevent reactor-building.overpressuri-zation. The -relief damper is on the ROA fan intake and downstream of the ROA . supply valves. ROA-V-l and ROA-V-2 that close under LOCA or radioactive release . conditions."- Thus, this modification does not compromise Secondary Containment. - This modification did not result in a change to the WNP-2. Technical Specifica-tions or involve an unreviewed safety question because it did not affect a safety=related

system, the modification to the ROA

- system did not change the boundary conditions used in the

FSAR, and the WNP-2 Technical Specifications were not affected.

Plant Oesi n Chan e 88-0306 Plant. Oesign Change 88-0306 was initiated to provide increased assurance of =:appropriate.Control Room HVAC System operation following a design basis Loss of . Coolant Accident (LOCA), thus ensuring Control: Room personnel post-event radiation doses remained within acceptable levels. 'uring a;LOCA, the normal fresh -air intake for the Control Room HVAC is isolated and two remote air intake lines are opened. Each remote air intake line has two isolation valves

wHh. one -valve powered from Oivision I and the other valve powered from Oivi-

. sion II;- In. the unlikely event of a special single.'failure (i.e., "hot short" - or "smart short") in a power division, a valve in each remote air intake line

could isolate.'ith the loss of all fresh air input, the Control Room HVAC would continue-to

operate, but in the recirculation. mode.

In.the recircula-ti.an-.-mode, the Control Room would not remain pressurized with respect to sur- - rounding==areas. Acceptable post-event radiation dases to Control Room per- --sonnel could not be assured because operating post-LOCA in this mode was not analyzed. (This condition was discussed in LER 88-031.) -This modification replaced the motor operators on the four remote air intake isolation valves (WOA-V-51A, -518, -52A, and -52B) with manual operators. This.allowed one remote air intake line to be open continuously, thus assuring that-a single failure could not cause operation in the recirculation mode.

==Also, post-event manual transfer could be made to the other remote air intake ,-path if the currently open remote air intake path-reached 'unacceptably high radiation levels. This modification did not involve a change to the WNP-2 Technical Specifica-tions=- or involve an unreviewed safety question because: (1) changing the -- ',remote air intake valves to require manual operation eliminates the possi-bi-lity of a single failure and ensures that the Control Room continues to meet the licensing design basis for analyzed radioactive dose rates; (2) no new event important to safety was creyted by this

change, and (3) the margin of safety in th'e Technical Specifications was not reduced because one path will

-.- always be operational and the time required for Operator action has minimal .dose impact. I

Plant Oesi n Chan e 88-0430 Plant Oesign Change 88-0430 was initiated to prevent premature failure of maintenance drain lines from two Main Steam (MS) trap stations. Orain valves were removed from each trap station and the lines capped to minimize flow -induced vibration forces that were causing maintenance line fatigue failures. The drain lines were uncapped and the valves were-replaced during the R-4 maintenance outage. Additional supports were provided to reduce vibration forces to acceptable levels. This - modification removed two Main Steam valves (MS-V-239 5 MSV-2388) from Trap Station g2 drip leg piping and two Main Steam valves (MS-V-118C 8 MS-V-238C) from Trap Station g3 drip leg piping, and welded a cap to each of the respective drip legs. This resulted in temporarily disabling the drain capability of the trap stations until the maintenance outage. Oraining of the trap stations can only be performed during shutdown and is normally done during the maintenance outage to remove built-up debris. As a result, this modification did not impact safety-related equipment nor increase the potential to degrade related equipment (e.g., main turbine). Temporary removal of the valves did not require a modification to the WNP-2 Technical Specifications. This change did not involve an unreviewed safety question because the potential failure of the drip legs from vibration induced fatigue was reduced making the Main Steam system more reliable. Plant Oesi n Chan e 89-0141 - Plant Oesign 'Change 89-0141 was initiated to ensure a Reactor Building pressure-of -0.6 inch water gage within the existing:.HVAC system capability. The modification maintains adequate ventilation and. cooling within all areas of the Reactor Building. =This modification changed the pitch of the blades of the Reactor Building Outside. Air HVAC System fan ROA-FN-1A from a supply.flowrate of 90,000 cfm to . 70 000;:.elm. =.With a bui lding in-leakage of 5,000 cfm at -0.6 inch water gage =--"- -: and-a.nomina1-exhaust fan flowrate of 91,000 cfm for REA-FN=, 18, the new supply -. fan configuration assures appropriate building pressure, even - with moderate

winds, without creating excessive loads on the exhaust fans.

This 'xtends equipment life and increases overall Plant reliability.

Also, a

building ~ pressure of -.0;25 inch water gage can be maintained 'under design basis condi- - tions. - Although the new air balance configuration reduces total ventilation flow-.-below design, adequate HVAC is still provided for all areas of the

== Reactor Buildi;ng. This is because the capability of-HVAC system with the new ai.r balance configuration exceeds the actual building heat load which was determined to be less than the design building heat load. In addition, the requi-rement -to draw air from areas of minimum contamination through areas of higher contamination was satisfied. ~ -= = This modification did not result in a change to the WNP-2 Technical Specifica-

..tions or invo1ve an unreviewed safety question because:

(1) supporting.calcu-lations -determined the HVAC system will meet the design basi's requirements as described in the FSAR; (2) the boundary conditions-of the FSAR evaluations t were not

changed, and (3) the margins of safety in the WNP-2.Technical Specifications were not reduced.

1l

Plant Desi n Chan e 89-0178 8 Plant -Design Change 89-0178 was initiated to reduce the time to energize the emergency buses (SM-7 SM-8) from the backup emergency diesel-generators (OG). - The relays that provide the contact permissive in the diesel-generator - output breaker control circuit were electromechani.cal with marginal per-formance'. More consistent OG start and load times-can=,be realized with solid state relays. This modification replaced the existing electromechanical GE relay DG-RLY-59 OG1/OG2 -with = an ASEA (ITE-27N) relay of solid state design to improve per-formance of -the OG voltage permissive interlock for--output breaker closure. -'This change util improve pickup voltage repeatability and provide faster and more consistent diesel start-to-load acceptance times. This modification did not result in a change to the'NP-2 Technical Specifica-tions or involve an unreviewed safety question because the reliability and time to energize emergency buses from backup power was-improved. Plant Desi n Chan e 89-0200 - Plant Design--Change 89-0200 was initiated to minimize =the =-possibility of con-tainment -liquid bypass leakage through the Control: Rod-Drive (CRO) System. Given --failure of the CRD

pumps, the existing design=

had the potential of 'releasing radionuclides in excess of the 10CFR 100 guidelines. This was based upon the design basis post Loss-of-Coolant Accident (-LOCA) radiation dose ca1culations;::-This condition was identified as a result of a commitment made , in; LER 88-012 to evaluate NNP-2 for possible unmonitored release paths. . The. design change installed two check valves (CRD-V=524.&.-525), a globe valve .(CRD-V-526), -and three vent lines and valves ups'tream -and between the two check valves and globe valve in the 2-inch CRO supply-. line upstream of two CRO filter units-.(CRO-FU3A & -38). The check valves .perform the safety-related function of preventing bypass leakage from the reactor vessel to the area out-side of the reactor building during post-LOCA conditions.. - This modification did not result in a change to the NNP-2: Technical Specifica-tions or involve and unreviewed safety question because: (1) the probability of an unmonitored release from the CRD system was reduced; (2) the boundary conditions-used in the FSAR evaluations were not affected, and (3) the margin of safety in the Technical Specifications was not reduced.

2.6.2 LIFTED LEADS AND JUMPERS The following are summaries of. noteworthy changes made. in the facility by use ~ ~ of the Lifted Lead and Jumper (LLJ) Procedure (PPM 1.3.9) as required by 10CFR50.59. Each change was evaluated and determined not to. represent an unreviewed 'afety question nor require a change - to the WNP-2 technical specifications. LLJ 289-207 (Change to.SM-7 and SM-8 Minimum Bus Voltage Annunciation) Problem Descri tion A. review being performed in response to an Operational=-Experience Report (OER) discovered.- a problem with equipment powered from some -distribution panels. This condition - whould occur during plant conditions where bus voltage was slightly higher than the degraded voltage relay.pick-up. An urgent Plant Modification Request was immediately processed to provide. a permanent fix to this condition (See PMR 89-0159 under the Plant Modification Section of this report). Discussion and Corrective Action A -.Justification for Continued Operation was prepared which::-recommended that -=-.- plant operators be made aware of the changes in the--degraded-voltage relay protection'equirements. A Lifted Lead and Jumper-Temporary Modification was -.= approved which. changed the degraded voltage protection-to-provide annunciation .: "=: in the 'control room upon the occurrence of the minimum acceptable voltage of 93%. -In - addition, plant annunciator procedures '(PPMs 4;800.C1-2.4 and 4.-.800;C5-. 2:4) were modified to require operator action-,if the alarm occurred. =-:-"A-. 50;59: evaluation was performed to support this temporary change= in the plant electrical: -configuration. The operation of the plant with this temporary power-'upply in place did not result in a. change to=.the-WNP-2 Technical Speci-fications or involve an unreviewed safety question-because: =(1) the overall - -'.operation-.of;the undervoltage electrical protection..-.met-minimum -.requirements, .(2) the margin of safety provided in the technical specifications was not -.changed, -.and (3) the boundary conditions for the:FSAR..evaluations were not changed. J

LLJ 289-0221= (Temporary Power Provided to Division II 24YOC Battery Chargers) Problem Oescri tion During the -Spring 1989 refueling outage transformers E-TR-8/81 and. E-TR-8/83 needed to be-.taken out of service for Division II= ma.intenance. Mith these transformers out of service the primary source of power to the Division II 24VOC system would be lost. Discussion and Corrective Action A Jumper and Lifted Lead request was processed and approved which allowed a temporary. power supply from a non-Division II source-to .be.connected to bat- --'". tery chargers E-CO-2A and 28. This allowed continued .operation. of control

. room instrumentation as desired by plant operations =and..prevented the bat-teries from dischar in durin the Division II outa e. g g g g =A. 50;59'. evaluation was performed to support this temporary change in the plant electrical configurati on. The operation of the plant with thi s temporary power supply..in place di d not resul t in a change to the MNP-2 Techni ca 1 =-.=Specifications or involve an unreviewed safety question because: (1) main-taining the power to the Division II 24VDC system during -the Di.vision II out-age did-not-change the function of the system (2):the-. margin of safety pro- =vided in the. technical specifications was not

changed, and (3) the boundary conditions for the FSAR evaluations were not changed;-

LLJ 289-0222. ,(Temporary Chargers) Problem Oescri tion Power Provided to Division II: 125VOC Battery During 'the -Spring 1989 refueling outage transformers E-TR-8/Bl. and E-TR-8/83

-.needed-.-to -be-. <aken. out-of-service for maintenance.-- Mith these transformers out 'of service the primary source of power to the Division-II-125VOC system would be lost.

Discussion and Corrective Action A Jumper and Lifted Lead request was processed and approved which allowed a temporary power supply to be connected to distribution =panel OP-Sl-2 which al.lowed continued operation of control room instrumentation needed to monitor =. the-safe shutdown status of the plant and prevent the 81-2 batteries from dis-charging.-. This was done by providing a jumper between the 81-7 and Bl-2 bat-teries to allow the 81-7 charger to carry the load. - A 50.59 evaluation was performed to support this temporary change in the plant electrical configuration. The operation of the plant with this temporary power supply in place did no't result in a change - to the MNP-2 Technical --: Specifications or involve an unreviewed safety question because: (1) main-taining the power to the Division II 125VOC system during the Division II outage: did not change the, function of the system (2) -the margin of safety pro-vided in the technical specifications was not

changed, and (3) the boundary conditions for the FSAR evaluations were not changed.'

55-

I 1 1

LLJ 289-9999 (Temporary Power Provided to MC-7A) Problem Descri tion During the Spring 1989 refueling outage transformer E-TR-7/73 needed to be taken out of service for Division I maintenance. Nith-this transformer out of ser vice the source of power to Motor Control Center MC-7A would be lost. This motor control center provides power to Division I plant monitoring instrumen-tation. -Plant operating personnel requested power to allow monitoring activi-ties to continue during the transformer outage. 'iscussion and Corrective Action A Jumper and Lifted Lead request was processed and approved which allowed a temporary power supply from a non-Division I source= to be connected to MC-7A via SL-73. This maintained power to battery chargers Cl-1 and C2-1 and allowed for continued operation of control room instrumentation as desired by plant operations. A 50.59 evaluation was performed to support this temporary change in the plant electrical configuration. The = operation of the plant with this temporary power supply.in place did not result in a change to the MNP-2 Technical Specifications.or involve an unreviewed safety question because: (1) main-taining the power to the Division I MC-7A during the= Division I outage did not change the-function of the system (2) the margin of safety provided in the technical specifications was not changed, and (3) the boundary conditions for the FSAR evaluations were not changed. LLJ -289-0224 (Change to Make "Bridge-Over-Core" Interlock Functional) LLJ 289-0300 Problem Descri tion -.. During the.Spring 1989 Refueling Outage conductors 1,84 and 186 in the refuel-ing. bridge takeup reel were found broken on two di-fferent = occasions. With = these conductors broken'he interlock for the "Bridge-Over-Core" was not functional. Discussion and Corrective Action -Jumper and Lifted Lead requests were approved which allowed the use of power from a spare: cable (SP-1) in place of the broken conductors. This provided --.power to the activity control unit logic to determine when the refuel bridge -is in "Over-The-Core" status for implementation of refuel mode interlocks. A 50.59 evaluation was performed to support these temporary changes in the plant electrical configuration. The operation of the plant with this tem-porary power.supply in place did not result in a change to the MNP-2 Technical -- Specifications - or involve an unreviewed safety question because: (1) the - overall operation of the Refueling Bridge and its logic=and "interlocks did not .change, (2) the margin of safety provided in the technical specifications was not changed, and (3) the boundary conditions for the FSAR evaluations were not changed. I 4 h I

LLJ -289-0225-

(Temporary Power Provided to Source and Intermediate Range Neutron Monitor Logic)

Problem Descri tion 'uring the Spring 1989 Refueling Outage maintenance was required on the ,Division II safety-related Switchgear (SM-8).

This, in turn, would cause a

loss of power= to the Source and Intermediate Range Neutron Honitoring (SRM and IRH) logic circuits. This was unacceptable since refuel mode surveilla'nces were required which called for operation of the SRH/IRM logic. Discussion and Corrective Action =-A Jumper " and= Lifted Lead request was approved which allowed for temporary power to be supplied from a convenience outlet at -:the 522 foot. level of the Reactor Building to the SRM/IRM logic. This allowed the control rod block to be cleared and the refuel mode surveillances to proceed during the outage. A 50;59 evaluation was performed to support these temporary changes in the plant'lectrrca1 configuration. The oper~tion of the plant with this tempo- - rary-power supply. in place did not result in a change to= the MNP-2 Technical =Specifications'--or involve an unreviewed safety question -because: (1) the overall operation of the SRH/IRM and its. logic and -interlocks did not change, - (2) 'the margin of safety provided in the technical specifications was not . changed;- and -(3) the boundary conditions for the FSAR evaluations were not changed.

LLJ 289-316

-.(One .Main Steam Relief Valve Declared Inadequate Air Supply) Problem Descri tion .-During.'he Spring 1989 refueling outage the flex-hose air to Main Steam-Relief Valve (MS-RV-2D) was found damaged flex-hose could not be replaced prior to plant startup; Discussion and Corrective Action Inoperable Due to supply (CIA-FLX-1C) beyond repair. The A Lifted Lead and Jumper request was approved which removed the flex-hose from the.=relief valve and replaced it with a blind flange. - Thus, the manual relief (air-actuated)- function of the valve was not operational. The safety (spring lift) function of the valve i s sti 1 1 operational., .A-50.59 evaluation was-performed to support this temporary change in the plant mechanical configuration. The operation. of the plant with this temporary flange in place did not result in a change to the MNP-2 Technical Specifica-tions or involve an unreviewed safety question because: (1) the overall oper-ation-and function of the relief and safety valves for the primary pressure boundary-did not change since only twelve of the eighteen valves are required to.;be operational ( In addition, this valve is not one of the Automatic Oepres-surization System Valves), (2) the margin of safety. provided in the technical specifications was not

changed, and (3) the boundary conditions for the FSAR evaluations were not changed.

t

LLJ 289-469 (Defeat of Alarm "Remote Shutdown or. Alternate-Remote Shutdown Transfer Switch Activated" ) r Problem Descri tion . During the Spring 1989 refueling outage two fans in the Radwaste Building (WHA-FN-52B and MHA-FN-538) were being operated from Fire Remote Transfer Panel 1 (FRTP1) because of degraded voltage concerns,. When the fans'ontrol switch -is 'placed in EHERGENCY to operate the fans,. an alarm is generated to signal '-the cont'rol switch is not in NORHAL. This masks all other alarms that could be -generated if other NORHAL/EHERGENCY control switches were placed in EHERGENCY. Discussion and Corrective Action The a1arm-was-occurring because Fire Remote Transfer-Switch -(E-RHS-FRTS-5) was in emergency -=to allow operation of the two fans from-FRTPl. A jumper and lifted lead request was approved which deactivated the alarm-from E-RHS-FRTS-5. The defeat of'he alarm from E-RHS-FRTS-5 did not result in a change to the MNP-2 Technical Specifications or involve an unreviewed safety question because: "(1)- restoring the annunciator to a usable.state met all requirements and-allowed-monitoring of the remaining remote and.alternate remote shutdown panel switches;- (2) the margin of safety provided in =the technical specifica- -tions was-'ot

changed, and (3) the boundary conditions for the FSAR evalua-tions were not changed.

LLJ 289-0493 -(Temporary Jumper to Allow Standby Service Mater Loop "A" to .: - -;--Remain Operational With the, Pump Discharge Valve (SM-V-2A) Non-operational in the Full Open Position) Problem Descri tion While =starting -the Standby Service Water-System "A" the pump discharge valve ~ (SM-V-2A)--failed to open. Further investigation found the valve operator motor still running but no longer engaged to the-operator. The valve was manually placed in the full open position and Standby. Service Mater Loop "A" continued. to operate.

However, in this condition, if.the service water pump (SW-P-lA) were to trip (e.g.,

loss of off-site power)--the pump would not be able to start since it needs a "SW-V-2A CLOSED" permissive.to start. Discussion and Corrective Action I A Jumper/Lifted Lead request was approved which defeated the "SWV-2A CLOSED" permissive. The use of the Standby Service Mater Systems without the "SM-V-2A CLOSED" permissive did not result in a change to the MNP-2 Technical Specifi-cations -and the unreviewed safety question concluded'. (1) the performance of the Service Mater System met all requirements, (2)-;the margin of safety pro-vided -in the technical specifications was not

changed, and (3) the boundary conditions for the FSAR evaluations were not changed.

2.6.3 FSAR AMENDMENT EVALUATIONS The following are summaries of changes made to the FSAR in Amendment 40 which were not initiated as a.result of a plant modification. As part of the process of submitting an FSAR

change, an analysis is performed in accordance with 10CFR50.59 to ensure the proposed modification does.not involve an

=-unreviewed - safety question. The following summaries represent -changes in 'system operation, clarification and/or updates of system descriptions, clari- =- -.- fication of Supply System positions

and, in some

cases, changes to-commitments previously made in the FSAR.

Cha ter 9 Standb Service Mater =-MODIFICATION=-This revision to the-FSAR changes the requirement for minimum cooling water-flow to the Residual Heat Removal (RHR) Loop C pump (RHR-P-2C) 5 seal from 9 gpm to 0 gpm. Basis For Chan e This change was based in part an the =similarity in design

== -- and operating conditions during a design basis accident between the RHR Loop C = pump and'he Low Pressure Core Spray (LPCS) pumps. The design and size of the

seals-are very"similar between the'PCS and RHR pumps. (i.e;--,

3=.5 inch OD shaft =-. versus 3.75 =inch OD shaft, respectively). The LPCS-.-pump and RHR Loop C pump -=-- seal flushing operating temperature conditions are-.:the. same (i.e., maximum .=... -.narmal operating water temperature is 120 F with a--peak temperature of 212'F -=. -for accident temperature). During the Reactor Pressure Vessel (RPV) cooling -:mode, the; RHR -Loops A and B pump seal flushing line suction water temperature -reaches 335'.F.. Since the LPCS specification does not,-require cooling water for its-seal and the RHR Loop C pump seal does not experience the high temper-ature=-;fluid that Loops A and B do, - the Loop C RHR-.pump does. not require any cooling water flow.

2.6.4 OTHER The . Plant Problems-Plant Problem Reports Procedure (PPH 1.3.15) provides ~ ~ ~ ~ ~ ~ ~ ~ . instructions for the disposition and documentation of 'lant problems. An "imnediate disposition using the "Use-As-Is" or "Repair" options is considered a.-"change.'ithin the definition of 10CFR50.59. . Each.item below has been evaluated to:.provide assurance that the disposition:. does. not.involve a change to the technical specifications or an unreviewed safety question. -'CR 288-0356 NCR 288-0357 (Maintenahce of Secondary Containment~me ative Pressure) Problem Oescri tion The FSAR (6.5.1-and (9.4.2) and the Technical Specifications (3/4.6.5) requi re . the-.secondary containment (reactor building air space) to be less than .25 "-..inches. of vacuum water gauge. The pressure devices which measure this limit =..- did not-;compensate for the environmental effects of. differential temperature .- and:could-have-resulted in a situation where the ='vacuum limit would not be maintained.-for-secondary containment. In addition;- there is a documented =. .-. -concern--.regarding the re-establishment of secondary= containment: differential '9 pressure following a design basis accident. Discussion and Corrective Action A Justification-for Continued Operation (JCO) was prepared which concluded the existing setpoint for the pressure measuring devices of.60 inches of vacuum 9', water--gauge -would accommodate the environmental.effects and maintain the .: required: vacuum in the secondary containment during =-normal .operation. -A ':.- second -JCO. -was., prepared to justify operation of the: plant while calculations -":..:are.completed. on both offsite and onsite doses during postulated design basis accident. conditions with new secondary containment assumptions. -,.- -,-- As= a consequence of the evaluations performed in preparing the second JCO, the

Standby..'Gas.=Treatment System (SGTS) flow was increased=and.-.credi.-t "was taken for-,building,: inleakage less than the Technical Specification limits resulting

...:in-an = unreviewed safety question. The unreviewed s'afety -question-evaluation . concluded =the -function of the secondary, containment'ould be maintained and

al.lt study, calculations show the offsite and onsite doses to be below 10CFR100 limits following design basis accidents.

Ultimate'esolution of this problem -.- will involve:Technical Specification and FSAR-changes and require-significant cal'culational updates. 4'&'

- PER 289-0009 (Emergency. Lighting Failure During Annual Discharge Test) ~ ~ F. Problem Oescri tion Several eight-hour emergency battery lights failed their annual. discharge test being conducted by plant surveillance procedure (PPH J0.25.63). , Discussion and Corrective Action / Battery--units were replaced to the extent permitted--by available spares. A - -=='JCO was prepared which concluded that sufficient . emergency lighting was =-available for operation', access and egress. This-included an evaluation of physical 1:ighting installed -arid functional, and a: drawing review to ensure' lighting was provided in all necessary areas. The-.disposition of this item was "Use-As-Is". The use of-'he. Emergency Lighting System as-is did not. result in a change to the WNP Technical Specifications and the unreviewed safety question evalua- .='ion concluded: (1) the function and performance wf: the Lighting System did '- not change-, (2) the margin of safety provided in the technical specifications wa's not changed, and (3) the boundary conditions for.; the;= FSAR:-evaluati'ons were 8 not changed. PER 289-019-( Ident1ffcat1on of Four New Fai1ure-Medeas.for the Containment Nitrogen System) Problem Oescri tion . Four; -.new failure modes for the Containment Nitrogen. (CN) System were =.identified ='.that should have been analyzed as part.- of the - plant design

- including

.=-(1.), A postulated break in the Auxiliary"-Steam piping, (2) Swamping the -low flow vaporizer, (3)

Design Basis Tornado;.and.- (4) Rupture of the Nitrogen Storage Tank or its associated piping. Discussion and Corrective Action Each.of.the -identified failure modes. were analyzed -and a -Justification for-Continued'peration was completed. The immediate disposition of this item was "Use-.As-, Is".with a deviation to two plant procedures -and the placement of a .-..portable alarming oxygen monitor in the control room: under certain conditions. A

.-- The. use. of the.Containment Nitrogen System as designed and constructed did not

- result.-in a: change to the WNP-2 'Technical Specifications and the unreviewed safety question evaluation concluded: (1) the potential for damage to plant equipment-.and--the containment was very low, (2) the margin of safety provided . in -- the technical specifications was not

changed, and (3) the boundary conditions for the FSAR evaluations were not changed:

F l J t t, l I II f t

PER= 289=020: --(Secondary Fuse Covers Installed

==in Safety-Related Motor , Control Centers Without Proper Design ControT) Problem Oescri tion G - -Per'sonnel-safety secondary fuse covers were instal.led in selected safety-relat'ed motor control centers without proper design'control-. The covers were installed to alleviate a personnel shock hazard in some 480 volt motor control centers:. 'No Plant Modification was processed and no;10CFR50.59 evaluation was performed to evaluate the change. Discussion and Corrective Action A Maintenance= Work Request (MWR) was initiated to inspect.and -record the type of'overs used in each location. A 10CFR50.59 evaluation was performed which - concluded: -'1') that the probability of occurrence-or the. consequences of an - accident or malfunction of equipment as evaluated= in--the FSAR would not be --- -'-'increased because the fuse covers were fabricated. of-insulation-material which '-.cannot" present any electrical failure mode and -,-,their ight construction prevented the possibility of seismic

concerns, (2) there-was no possibility of

= creating='an =accident or malfunction of a different=-type-.than evaluated pre- - viously:sn the-FSAR because the covers provided added insulation in the area of-the fuse blocks which add to the safety of the design by reducing.-the pos- 'ibiTity of-failure during accident conditions, and-(3)= the margin of safety. ~==- as defined in the Technical Specifications was not reduced.-- In'ddition, the process for performing plant modification was changed to clearly r'equire a formal modification before physical. changes.-.are initiated 'in the plant. - 62'-

PER 289-026 (Gasket of Incorrect Thickness Installed in. Several Main Steam e Safety/Relief Valves) Problem Oescri tion ~ .The vendor for the Main Steam Safety/Relief Valves =specifies a 0.125 inch -'-="thick eductor (bonnet) gasket. The Supply System's Materials Management System incorrectly specified a 0.250 inch thickness for the gasket and these were procured 'and installed in the plant. Mhen this;problem.was discovered in January 1989 eight incorrect gaskets were still in--stock -in the warehouse. = Records showed that at least 10 of the incorrect gaskets had been withdrawn from spare stock previously, of which four were wi-thdrawn-From. the warehouse and =installed on spare

valves, and six were withdrawn from-the-warehouse and installed on in-service valves in the plant.

Oiscussion and Corrective Action The immediate -disposition for this item for the installed. in-service valves was =-to "Use-As-Is" based on the following actions: (1) the vendor (Crosby) was contacted to identify potential concerns associated with the incorrect gasket thickness, (.2) it was concluded that the impact= on-the setpoint would

==-" =be -in the conservative direction, (3) blowout of.=- the. gasket - would not be 11kely because . of a groove machined in the

body, (4) any increased leakage would not *be a

problem as it would be identified and dispositioned in accor-danc'e with existing procedures, (5) misalignment of.=-the body-to-bonnet joint --would not-be a problem since the alignment't the joint is controlled by the diametral fit of the eductor in the body, and (6) valve function using the air =actuator would-not be affected by the thicker gasket; The incorrect installed - gaskets will be replaced per the normal preventative:maintenance schedule. -'ther corrective actions were taken as follows: ('1)- the --.incorrect "Material 'Code:"- has-been -deleted and replaced by the correct

code,

=-(2) new gaskets were

ordered and-the incorrect gaskets were

scrapped, and- (-3) the correct gaskets were installed in. the spare valves.

The use-of. a -gasket of incorrect thickness did not result in a change to the MNP-2 Technical Specifications or involve = an unreviewed-safety. question because: (1) the valve function and performance did. not -change, (2) the mar-gin of safety'rovided in the technical specifications was not

changed, and (3)- the boundary conditions for the FSAR evaluations were not changed.

'e

PER 289-029

(Limit Switches and Connectors on Containment (Wetwell-Drywell) Vacuum Breaker Valves Installed Without Proper Seismic and t Environmental Qualification Review of the Design Change) Problem Descri tion <imit switches.'and connectors for nine containment=-(wetwell-drywell) vacuum breakers were installed without proper seismic and environmental qualification review of -the-design change. The plant modification changed the type and mounting of -the position switches and added a CONAX connector for the wires .: exiting the valve between the two discs. The connectors constitute part of the wetwell/drywell isolation as they penetrate between the dual disks of each wetwell/drywell vacuum breaker. Discussion and Corrective Action -The.immediate.disposition for this item was "Use-As-Is"; The design change was -reissued;,and reviewed to quality class I requirements The -review showed the new switches were mounted with two more bolts than the original switch and the new switch had less mass than the old. Since the switch itself has no 'safety.-.related= function (they are for indication -only)- the evaluation was --. limited to a seismic review of the mounting which met all -requirements. -The connectors themselves were found to be specified to= quality class I require- -:-;ments;. The-seal uses ceramic separators with Grayfoil packing to prevent leakage and leakage tests were performed on the seals;- The use of-the -limit switches and connectors did not result in a change to the WNP-.2 Technical Specifications or involve an unreviewed safety question because". (1);-the wetwell-drywell vacuum breaker function and performance did

not change, (2), the margin of safety provided in the technical specifications was not changed, and (3) the boundary conditions for the FSAR evaluations were not changed.

ti r, e I

PER 89-033: (Residual Heat Removal Heat Exchanger Thermal Relief Valve Install ed Backwards) Problem Descri tion The Residual Heat Removal Heat Exchanger (RHR-HX-1A) service water (tube side) thermal relief valve (SW-RV-1A) was found to be = installed backwards. The relief valve inlet was bolted to the discharge piping and the relief valve -discharge -was bolted to the heat exchanger tap. . SW-RV-1A provides thermal

- overpressure protection to RHR-HX-1A if the tube side =is isolated by its block valves (RHR.-.V.-14A and RHR-V-68A).

This heat exchanger is used-for shutdown cooling and'lso functions as part of the Emergency Core Cooling System (ECCS). Discussion and Corrective Action The -immediate;disposition for this item was "Use-As-Is" until the system was available to reposition the valve during the next outage. Until that time thermal. relief.protection was provided by tagging open the service water iso-lation valve (RHR-V-14A). With this valve tagged open thermal overpressure of the heat exchanger could not occur. --.The use of..SW-RY-1A in the backwards configuration did not result in a change to-the WNP-2 Technical Specifications "or involve an unreviewed safety question because: .'(1)--the RHR-HX-1A function. and performance did not change, (2) the margin=- of.-safety'rovided in the technical specifications was not changed, and (3) the boundary conditions for the FSAR evaluations were not changed. PER 289-038 (Average Power Range Monitor Channel F. Placed in Bypass) Problem Descri tion -.. In January 21, ~ 1989 the Plant received a half scram caused by a loose K18 -=-:.-= .-- relay socket; The relay socket had been loosened by.repeated removal of the - relay in accordance with Average Power Range Monitor (APRM) surveillance procedures. Discussion and Corrective Action A =10CFR50.59 Safety Evaluation was performed to allow APRM Channel F to be -bypassed until the next outage when the defective -relay socket was replaced. The APRM bypass selector switch for channels B, D, and F was caution tagged to select APRM F for bypass except while performing surveillance tests on APRM B or APRM -D. Operation with a single APRM bypassed was consistent with the FSAR and the :Technical Specifications. A plant shutdown occurred on January 30, 1989 and the K18 relay socket was replaced on January.31, 1989. The use of the.APRM System with Channel F placed in Bypass did not result in a change to the=- MNP-2 Techn.ical Specifications or involve an unreviewed safety question because: (1) the APRM function and performance did not change, (2) the margin of..safety provided in the technical specifications was not chang'ed, and (3)..the boundary conditions for the FSAR evaluations were not changed. /II L

PER 289-0041 - (Check valves found installed backwards . in Diesel Starting System) Problem Oescri tion In January 1989 seven check valves were found installed backwards in the =.- Diesel Starting. Air Systems for DGl and DG2. The. valves and attached lines were painted :in line and appeared to have been installed incorrectly in the factory. . The - valves are spring-ball check valves and are located in the

.bypass line,that connects the air supply line to the=. starter pinions and to

-the line leading between the pinions and the air, start relay valve. The . incorrect valve direction was noted by guality Assurance on a routine walkdown of the system. Discussion and Corrective Action .- A Justification For Continued Operations (JCO) was prepared:and the immediate disposition-was "Use-As-Is". The JCO showed that the check valves in the air start logic provided no essential function and did -not-impact the operability of the diesel.generator units. This was due to the vent path of the external air port of the-upper air start motor pinion. In addition, there-was no indi-cation of any=.:malfunction in the air start system 'during several hundred starts-performed in the factory and during plant startup and"operation. The check; valves were installed in the correct orientation during the next refuel-ing outage in May 1989. . The use of..the.- Diesel Starting Air check valves in the; backwards orientation 'id not'result-. in a change to the MNP-2 Technical Specifications or involve an, unreviewed-- safety question because: (1) the Diesel =Generator.'function and

.; --.- -. performance did not change, (2) the margin of safety=provided in the technical ---specifications. was not

changed, and (3) the boundary conditions for the FSAR evaluations were not changed.

=-- =.:PER 289-0094 =-(Failure of a Damper Motor in the Dieej.Generator Heating and Ventilating System)

Problem Descri tion A damper motor,, in the Diesel Mixed Air System (OMA-AO-51) failed in the -- recirculation position. This damper is an outdoor mixing damper for the air handling unit which cools Division II cable and equipment in the -corridor during diesel operation. Discussion and Corrective Action This item was dispositioned "Use-As-Is". A JCO was prepared which included a calculation by Engineering which showed that adequate cooling was provided with the damper in the failed position. --- -The,use -of-.the Diesel Mixed Air System with the failed-damper did not result i'n a change. to the. NNP-2 Technical Specifications or involve.an unreviewed Qt safety question because: (1) the performance of the system met all require-

ments, (2) the margin of safety provided in the technical specifications was not changed, and (3) the boundary conditions for the FSAR evaluations were not changed.

lt

PfR 289-0098-(Standby Service Mater Pumphouse Ash Filters Not Installed in Accordance Mith Design Requirements) Problem Oescri tion 9 ~ ~ ~ -A review. of.the Standby Service Mater Pumphouse Heating and Ventilating System - -revealed three deficiencies with regard to the ash fall-filters-needed to pro-tect against the design basis volcanic event. MNP-2 has two standby service water. pumphouses (an "A" and a "B" pumphouse) and each pumphouse has two sets . of, fi.lters. - For one set of filters in each pumphouse-(PRA-FL-2A/2B) the ash-fall. filter boxes were not accessible as they were. blanked off by installed sheet metal. = The second set of filters in each pumphouse (PRA-FL-lA/1B) were actual;ly installed in the filter boxes contrary to: design. requirements which call for filter installation only under abnormal conditions (ashfall). The third deficiency was the requirement in the plant -procedures for replacement of the filters every three hours or when the delta P indication across the filters - exceeds a predetermined value. The delta P -indicators were never installed. Discussion and Corrective Action The immediate: disposition for this i tem was "Use-As-Is";-;.For PRAFL-2A/2B the sheet . metal -was removed and the filter boxes are-- now. accessible. For PRA-FL-1A/1B the filters were removed and placed in='tandby status. The requirement for delta P indication was removed from -.the -plant procedure. The " plant operators are required to replace the filters every-three hours in the - event of an ashfall. Calculations show that the three hour changeout time is very conservative. =. The use of. %he Standby Service Mater Pumphouse ashfall filters in the "as-is" - configuration:-did not result in a change to the MNP2 Technical Specifications -.= or involve arr-unreviewed safety question because: -(1) the performance of the --Service Mater -System was not degraded, (2) the margin of safety provided in the technical=.specifications was not changed, and (3) the boundary conditions for the FSAR evaluations were not changed. I

- PER 289-0179 (Calculated Non-Conservative Doses to Control Room Operators NCR 288-0403 Post-LOCA) Problem Oescri tion Under post-LOCA conditions engineering calculation NE-02-88-27 (performed in 'support of NCR 288-0403 took credit for 100% mixing of primary containment leakage within the reactor building volume before postulating a release to the -. environment-through the standby gas treatment system. This assumption was in

conflict with -Regulatory Guide 1.3 and resulted in.

a non-conservative dose estimate for the control room under accident conditions-. -. Oiscussion and Corrective Action - After an evaluation by Plant Management the Shift Manager declared an Unusual Event and-a controlled shutdown of the plant commenced; After three hours the shutdown was-halted at 52 percent power consistent with..the Engineering analy-sis indicating that the associated reduction in source term was adequate to - ensure habitability while the calculation problem was being. resolved. Compen-satory measures were defined to assure control room habitability, including the: requirement that both control room remote air intakes - remain open to ..assure-control room habitability, an operator be dedicated - to respond within '0 'inutes to close one of the remote air intakes-.in the case of high radiation, and= that system operating procedures be modified to reflect the new

- restrictions;

- This problem was resolved by improving the calculational methodology and removing. unnecessary conservatism.

A-. 50.59 evaluation was performed to support the continued operation of the plant 'at full=gower.

The operation of the plant with= the; revised-analysis and

the compensatory measures for control room ventilation--operation did not require a

change to the NNP-2 Technical Specifications or involve an " unreviewed safety question. Resolution of the associated NCR (288-0403) on single failure =vulnerability of the control room remote intakes did necessi- .'= tate a -Technical Specification change to exit the action statement requiring the control room pressurization mode of operation-. - 68,-

L V

PER 289-0487 LLJ 289-0353 LLJ 289-0376 Problem Oescri tion (Temporary Removal of Service Water Valves Associated with Diesel Cooling Mater) p ,Surv'ei:llance'esting on the Division I Diesel Generator-showed increasing high temperature in the diesel cooling water. This event was traced, to the failure of serv'ice water inlet isolation valve SW-V-214 to properly.open-. This valve is -'in -.the line that sup'plies water to one of two Diesel Cooling Water (OCM) heat exchangers. Discussion and Corrective Action A root cause analysis of the failure of SW-V-214 determined that the disc to shaft'ap'er. pins had corroded and subsequently worked loose. The recommended a'ction.*was:to:remove this valve and the other three.valves -.of the same design 'and application (SM-V-215, 216, and 217) to preclude the potential for similar - - fai-lures. 'in the future. Jumper and lifted lead requests were -approved which replaced each -of the four valves with straight "spools".. Other service water

valves (SM-V-4A and SW-V-48) will be used for heat exchanger isolation.

The. use of -the Diesel Cooling Water and Standby Service Mater Systems without the--four-.isolation valves did not result in a change to the WNP-2 Technical - Specifications'r involve an unreviewed safety question because: (1) the per-formance. of the Division I and II Oiesels were unaffected, (2).the margin of =.'safety provided in the technical specifications was not

changed, and (3) the boundary conditions for the FSAR evaluations were not-changed.

. - PER 289-0573 (Temporary Change to Allow Repairs On=the. High Pressure Core LLJ 289-0429 Spray (HPCS) Air Compressor Diesel (DSA-ENG-C/2C)) PDF 289-0590 Problem Descri tion The-air start diesel (OSA-ENG-C/2C) on.the HPCS diesel Starting Air System (OSA)- was not functioning correctly as it would not shut down after an auto start. Discussion and Corrective 'Action A jumper. and. lifted lead request was approved which disabled the compressor and. plant procedures were deviated to allow for operation of the air start diesel in an emergency. The use, of.the Diesel Air Start System with the Diesel Compressor disabled did not result in. a change to the WNP-2 Technical Specifications or involve an unreviewed safety question because: (1) the performance of the HPCS met all requirements,. (-2) the margin of safety provided in==.the--technical specifica-tions was not

changed, and (3) the boundary conditions for the FSAR evaluations were not changed'.

PER 289-0588 (Inadequate Service Water Flow Through Critical Switchgear Air Handling Unit) Problem Descri tion tQ - Service water flow through critical switchgear air handling unit cooling coil WMA-CC-5381 could not be adjusted to a value greater than-58 GPM. FSAR Table 9.2-5 requires': a minimum flow of 60 GPM. Partial .blockage . of the piping and/or cooling coils is indicated. Discussion and Corrective Action = A'.justification-for continued operation was prepared and a review of the Engi-neering calcul'ation showed a minimum flow of 54 GPM=- would provide adequate cooling. The, operation of WMA-'C-5381 with slightly reduced flow. did not result in a . change -'to,the WNP-2 Technical Specifications or, involve an unreviewed safety

=-.'question because:

(1) the cooling of the critical:.,switchgear rooms met min- '-imum requirements, (2) the margin of safety provided-in the technical specifi- -.- cations was-'not

changed, and (3) the boundary. conditions for the FSAR evaluations were not changed.

.--.- -'; PER 289-0649 -(Recirculation Flow Control Valve Penetration Transmitted Vibration and Noise) Problem Descri tion

At-'.the

-83%-=.open position recirulation flow control--valve (RRC-FCV-60B) '::penetrations

transmitted vibrations and noise.

The-noise was noted in the northwest-" corner of the 501 foot elevation of the reactor building and the "-.:. " hydraulic lines. to Recirculation Cooling Pump "8" were-vibrating and noisy. Q Discussion and Corrective Action -The:valve was=opened to the full open position and-the vibration and noise stopped. Flow.'nd power traces obtained from the Transient Data Acquisition -. --'ystem were reviewed by Engineering. Copies of the.data'traces -were submitted . to -General Electric for review. At the next outage. entry was, made into con-tainment and the valve and the area around the valve was inspected. All equipment appeared to be undamaged and operated normally; Plant; operation: with the noise and vibration did not result in a change to the WNP-2 Technical Specifications or involve an unreviewed safety question . because: -(1.).-the performance of the Recirculation. System met all.. require-ments;:.(2) the margin of safety provided in the technical specifications was not changed; and (3) the boundary conditions. For the FSAR.evaluations were not changed.

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POF 289-0653 ISCR-937 Problem Oescri tion PER 289-0650;, (Change to Reactor Recirculation Flow Control Val ves Runback e Limit Setpoint) A reactor scram occurred from 100K power when one-of the Reactor Feed Pumps (RFN-P-18) tripped. The scram occurred on low water. level since the remaining feed. pump was:not able to maintain vessel level. The problem was traced to an inappropriate Reactor Recirculation (RRC) runback setpoint; = Discussion and Corrective Action The procedure -.was devi'ated and the setpoint for the RRC. flow control valves (RRC-FCV-60A/8)- was changed from the incorrect 30$ -open position to the -correct 20$ -open position. This setpoint had been--veri. fied during plant

==.startup as the correct value to allow for recovery from. a. feedpump trip. Plant operation with the revised flow control valve-setpoint did not result in a-.change'o -the NNP-2 Technical Specifications or involve an. u'nreviewed safety = question.because: (1) the performance of the Recirculation and.Level Control .- Systems-met."all-requirements, (2) the margin of safety provided in the techni-

".-cal "specifications was not

changed, and (3) the boundary conditions for the FSAR evaluations were not changed.

PER 289-0736 (Incorrect Duty Cycles for Safety-Related

-125VOC Batteries) Problem Oescri tion The Supply System's internal Safety System Functional Inspection.(SSFI) dis-covered an i:ncorrect assumption in the calculation of duty cycles for the Division -I and II Safety-Related 125VDC Batteries. 'hen calculations in

breaker-actuation sequencing were made it was incorrectly.assumed that the

.-.:-"==.- spring -:charging motors associated with the 480VAC:-..switchgear. were energized --...=after -"closing". as is the case with 4160VAC switchgear.

However, the 480YAC switchgear motors are energized after breaker

". Trip". -. Discussion and Corrective Action A Justification for Continues Operation was prepared and approved. The 480VAC breaker. closing spring charging motors added a 10 second load during the first minute of. battery discharge of 50 Amps for Battery B1 1 and 60 Amps for Battery" 81-2.. The capacity requirement for these batteries -is determined by the steady state loads (two-hour) not by the first minute. loading. Therefore, adding the 480VAC breaker spring charging motors to -the first minute load did not change the battery capacity requirement. Plant :.operation with the existing Division I and -II 125VOC Batteries did result- -in a change to the NNP-2 Technicql Specifications and resul.ted in an unreviewed safety question evaluation w8ich showed: - (l) the existing bat- ' teries are capable of supplying the updated battery-duty cycles, (2) the margin of safety provided in the technica) specifications was not changed, and (3) the-boundary conditions for the FSAR evaluations were not changed. - 71

1 I

PER 289-0747 ( Inadequate Electri ca 1 Sepa rat ion and Non-Fail sa fe Oes ign of the Reactor Building Exhaust Air Radiation Monitoring System) Problem Oescri tion Ouring the preparation oF a Plant Modification three discrepancies. were dis- -'overed in the Reactor Building Exhaust Air (REA) radiation monitoring system. They = consisted of inadequate physical separation in Control Room

cabinets, routi'ng of failsafe cable in non-Failsafe raceways outside of the control

room, and a non-failsafe design response of:-the radiation monitors to inoperative/downscale conditions.

Discussion and Corrective Action A justification for continued operation was prepared and approved by the Plant Manager. The failsafe circuits routed in non-failsafe raceways were placed on an hourly.-fire -tour to minimize the probability of:a fire that could cause a . circuit 'fault '- and the REA radiation monitor downscale annunciato'r response procedure was revised to require operator action to place the affected trip monitor in -a. tripped condition upon receipt of a.valid downscale condition. -An-engineering -evaluation and a plant modificaion are being prepared to pro-vide a permanent change to correct the problem. Plant operation with the REA radiation monitoring.system -"as-is" did not result in a -change to the WNP-2 Technical Specifications or involve an unreviewed safety question because: (1) the performance of the Radiation Monitoring System met all requirements, (2) the margin of safety provided in the-technical,.specifications was not changed, and (3) the boundary conditions for the FSAR evaluations were not changed. II

PER 289-0860 (Plant Operation Mith RHR-V-40 Tagged in the Closed Position) PDF 289-0868 Problem Descri tion Generic letter 89-10, "Safety-Related Motor-Operated -Yalve Testing and Sur-veillance", caused.a review of motor operated valve operability. The results of this review showed that one of the Residual Heat Removal (RHR) Loop "B" Discharge Valves from the Supression Pool (RHR-V-40)-.to the main turbine condenser had a-motor operator that did not provide;-sufficient=starting. torque at degraded voltage to operate at the maximum design. differential pressures. Discussion and Corrective Action A Justification. for Continued Operation was prepared and -approved which placed RHR-V.-40 in a: closed danger tagged position. Manual-handwheel closure of the -.- valve is performed after each opening. In addition, RHR-RLY-80/Y40, the relay

that acti vates: the Bypassed and Inoperable Status.

Indication (BISI) for a series of motor operated valves was removed to clear=the BISI alarm associated with this valve. . Plant. operation with RHR-V-40 in a closed danger tagged position did not result -in-a=- change to the NNP-2 Technical Specifications and the unreviewed . safety question evaluation concluded: (1) the .overall .operation of the .= Residual Heat= Removal System did not change, (2) the-margin of safety provided

in the.technical specifications was not

changed, and (3)

. the boundary con-a.~ ditions for the FSAR evaluations were not changed. 'ER-.289-0869.-=(Diesel Generator Room Overtemperature Conditions Postulated Accident Conditions) Problem Descri tion During Calculation-of-diesel generator room ambient temperatures exceeded values +:-.----..stated.in--the FSAR based on new diesel heat loads determined from a 24-hour test. =-Limiting temperatures were based on postulated accident conditions involving a Loss of Offsite Power during ashfall conditions: Discussion and Corrective Action A justification-for continued operation (JCO) was performed which showed that the plant could operate until the Spring 1990 refueling outage (through April) without changes-to the diesel room cooling system. The JCO was based on the -fact that the-hot weather that provides the limiting condition For the temperatures will not occur during that time period.-- Plant operation with the Diesel Room cooling 'as-is" did not result in a change =to the NNP-2 Technical Specifications or involve an unreviewed safety question because: (1) the performance of the diesel-room cooling systems will meet al-4 requirements during winter and spring conditions, (2) the margin of safety provided in the technical specifications was= not

changed, and (3) the boundary conditions for the FSAR evaluations were not changed during the period of operation.

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2.7 PLANT TESTS AND EXPERIMENTS This section of the report covers WNP-2 Plant tests and experiments not described in the Safety Analysis Report as required by 10CFR50.59. ARTIAL.DRAINING OF THE SPENT FUEL STORAGE POOL TEMPORARY PROCEDURE 2.8.14 A temporary procedure was written to lower the spent fuel pool level 19 inches to allow check valve maintenance to be performed; The. fuel pool diffuser check val ves-FPC-V-1 46A and FPC-V-1468 required -modifi cat i ons to their internals to implement an Equipment Modification Specification. The controlled lowering of the spent fuel pool level by-19 inches did n'ot result in =a:-change to the WNP-2 Technical Specifications or involve an unreviewed safety question because: (1) the performance of the fuel pool cooling systems met all requirements, (2) the margin -of-safety provided in the technical specifications was not changed, and (3) the -boundary conditions For the FSAR evaluations were not changed during the period of operation. WIDE RANGE NEUTRON MONITOR FINAL TESTS PPM 8.3.123 AND 8.3.74 e ~ Final testing w'as completed on the Wide Range Neutron-Monitor System installed per the requirements of Licensing Condition 16 and Regulatory Guide 1.97. The pressure integl ity of the in-containment cable assemby was verified using test procedur e PPH 8.3.123. During the subsequent startup= the system was cali. brated and an operabi1 ity check was performed 3n - accordance with plant procedure PPH 8.3.74. The performance of this final test did not result in.a change to the WNP-2 Technical. Specifications or involve an unreviewed. safety question because: (1) the -performance of the Wide Range Neutron Monitoring -System met all requirements,.'(2) the margin of safety provided in-.the -technical specifica-tions was not =-changed, and (3) the boundary conditions for the FSAR eval-uations were not changed. ROD WORTH MINIMIZER RWM PREOPERATIONAL TEST PPM 8.6.11 A preoperational test was performed on the Rod Worth Minimizer (RWM) following the modifications made by Computer Change Request (CCR) 001. The test ver-ified the operation of the replacement Rod Position Information System (RPIS) interface boards and the software change made to the existing Input/Output subroutines. The modifications resolved the problem: on the RWM with the . Select Error. Indication in the Transition Zone (greater than Low Power Set Point (LPSP) and less than Low Power Alarm Point (LPAP)). The performance-of this preoperational test did not result in.a change to the WNP-2 Technical Specifications or involve an unreviewed'afety question = because'. .(1.)-.;the performance of the Rod Worth Minimizer..met all requirements, .(2) -the margie of safety provided in the technical specifications was not

changed, and (3) the boundary conditions for the FSAR evaluations were not changed.

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2.8 PLAN

T. PROCEDURE

CHANGES The Plant Procedure control program requires a 10CFR50.59 evaluation whenever a procedure is'hanged which provides assurance that the disposition does not involve. a..change to the technical specifications.- or an unreviewed safety questi'on. -;The:-following are summaries of significant--Plant Procedure changes. processed during 1989: POF 289-0152 . (Procedure Change to Allow the Reactor =.Mater Cleanup System -'.(RWCU) to Operate During Modes 4 and 5 With Division I Power (SL-73) Out-of-Service) Problem Oescri tion A Division I =Power Outage is normally required for maintenance activities . during 'each refueling outage. When Safety-Related bus-SL is taken out of

=service.

power - is, lost to the Reactor Water Cl.eanup System (RWCU) non- .regenerative heat exchanger outlet temperature switch-=.(RMCU-TIS-8). This, in . turn, closes the outboard containment isolation valve (RWCU-V-4) which iso-lates the RWCU-System. During refueling there is a: need:to keep RWCU opera-tional to maintain ~ater quality. Discussion and Corrective Action - The; Plant Procedure on Removing SL-73 from Service (PPH 2;7.14) was changed to allow the: installation of a Lifted Lead and Jumper.:to deactivate the tempera-ture switch (RMCU-TIS-8) if preferred during an outage-;..

A. 50-.59-.evaluation was performed to support this change in plant. procedures.

The operation of the plant with this jumper in pTace.would not result in a . change =to.the.MNP-2 Technical Specifications or involve an unreviewed safety

-- question because:

(1) the Reactor Water Cleanup .System can operate safely without the. temperature switch during Modes 4 and 5,.-(.2) the margin of safety p'rovided=.in-.the technical specifications was not changed;=.and (3) the boundary conditions for the FSAR evaluations were not changed. --'- 9

PDF 289-0289 -(Procedure Change to Allow Refuel Bridge. Operation With Node Switch In Shutdown) Problem Descri tion The. Reactor Manual Control System (Rf<CS)- logic was not designed correctly and can send incorrect signals to the mode switch for the Refueling Bridge. Discussion and Corrective Action A. procedure change (PPM 2.14.1) was processed to allow the installation of a jumper to defeat the bridge s logic input from RHCS on mode switch position. This will allow refuel bridge operation while the reactor mode switch is in shutdown. The jumper will simulate the reactor mode switch being in the refuel position thus allowing continued bridge .operation over the core. 'mplementation of this change during the Spring 1989 refueling outage was not required. Implementation at any time in the future-will =require a Lifted Lead and Jumper. A 50.59 evaluation was performed to support this change in plant procedures. The operation of the plant with this jumper in pTace.would not -result in a change -to the.WNP-2 Technical Specifications or involve an unreviewed safety question because: (1) the overall operation of the:refueling platform and its restrictions -.would not

change, (2) the margin of=-safety= provided in the

-'- technical-specifications was not changed, and (3) the boundary conditions for the FSAR evaluations were not changed. -PDF 289-0394 = (Operation of the Residual Heat Removal "B". Pump Without the Suction Valve Full Open Interlock) Problem Descri tion During-the Spring 1989 refueling outage maintenance 'was being performed on the Limitorque Operator for the valve (RHR-V-9) that provides isolation'o the suction-of Residual Heat Removal Pump "8" (RHR-P-2B)..;.-RHR-V-9 -has a full.open interlock that=.-prevents the pump from starting if the.-valve is not fully open and-.trips the pump when the valve starts to close. It was desired "to have an additional shutdown cooling method available by using RHR loop "8" during this phase of the outage. Discussion and Corrective Action A 50.59.. Safety Significance review was performed to install an electrical - jumper and a.:temporary procedure deviation was approved to change the operat-ing procedure -(PPH 2.4.2) to allow operation of RHR-P-28 without the RHR-V-9 full open interlock during the short period of time while maintenance was being performed on the Limitorque operator on RHR-V-9-.

- This-change did not result in a

change to thh WNP-2 Tech'nical Specifications . or-.involve--an;unreviewed safety question because: (1) the overall operation -of the Residual Heat Removal System was admipistratively controlled and did not change, (2) the margin of safety provided'n the'echnical specifications was not changed, and (3) the boundary conditions for the FSAR evaluations were not changed. 1,

s I ISCR-920.;(Change to Reduce Temperature Range of'Accident Honitoring Recorder) .0 Problem Descri tion The Residual Heat Removal (RHR) Heat Exchanger Outlet Temperature had a range of 0 to 600 degrees F. This recorder is part of the accident monitoring .=instrumentation-and monitors the RHR and Fuel Pool Cooling -temperatures. The readability of the chart recorder needed to be improved. Discussion and Corrective Action -The -recorder- +ange was reduced to 0 to 500 degrees F.

• The reduced span pro-.

-:-"vided better resolution in both the operating and accident-temperature ranges. This change'id not result in a change to the WNP-2 Technical Specifications or.invol've'.an unreviewed safety question because: (1) the overall operation of -the -accident monitoring instrumentation did not:change,. (2). the margin of ---; -.safety provided-'n the technical specifications was:not

changed, and (3) the boundary conditions for the FSAR evaluations were not changed..

-0 (R t C t 1 4 li I ~tbl / Pgp 289-0949 and Inoperable Status Indication Change) Problem Descri tion 'e, . - Two Reactor; Core 'solation Cooling Valves (RCIC-V-22 and.RCIC-V-59) were taken out-.of.-service to effect repair of RCIC-V-22. These two-valves are the test return: to Condensate Storage Tank flow control and stop valves, respectively. Deenergizing=-:and tagging out ihe valves caused the - Bypass. and - Inoperable Status '.Indication (BISI) system to activate masking=.any signals that may be present on-eleven other RCIC valves associated with..annunciator 4.601.A4-6.8, '-. '!RCIC DIVISION:-I OUT-OF-SERVICE" and the "MOTOR OPERATED VALVE NETWORK POWER LOSS/OVERLOAD" BISI. Discussion and Corrective Action

Relays RCIC-RLY-80/22 and RCIC-RLY-80/59'ere pulled to allow the remaining motor;operated=valves to be monitored by the BISI system.

Plant procedure PPH 4.601.A4-6.8 was deviated to show the change. The -.oper.'ation of the plant with the BISI system modified as noted above did not. result, in-a change to the WNP-2 Technical Specifications and the unre- --= viewed safety.question concluded: (1) the performance of the BISE System was. improved,'.(2): the margin of safety provided in the technical specifications was not-changed', and (3) the boundary conditions for..the FSAR evaluations were not changed.

0

POF 289-964'. -(Modification of the Installation) ~ ~ ~ ~ Problem Descri tion Q The plant procedure (PPM 10.25.61) used required-clari:fication on the practice out of a multi-conductor cable. 4 Discussion and Corrective Action Plant Procedure-. on Wire Marker to mark cables.and wires in the plant oF marking the wires which are broken = The. plant procedure (PPM 10.25.61) was changed to -allow. the use of plastic sleeves or. Brady markers which have black lettering on -a white background as an approved installation method. The operation.=of the plant with the wire marking procedure. modified as noted above did;-not" result in a change to the WNP-2 Techni.cal Specifications and the unreviewed. safety 'uestion concluded: (1) the wire marking criteria and performance was not

changed, (2) the margin of safety-provided in the tech-nical specifications was not changed, and (3) the boundary conditions for the FSAR evaluations were not changed.

0

2.9 REACTOR COOLANT SPECIFIC ACTIVITY LEVELS This section contains information relative to reactor coolant cumulative iodine levels, iodine spikes and specific activity of all~isotopes other than iodine. The specific activity of the primary coolant was significantly. less than 0.2 microcuries per gram dose equivalent I-131 as set-forth in NNP-2 Technical Specification LCO 3/4.4.5 and paragraph 6.9.1.5 'c., (see 1989 cumulative iodine graph;= attached). The specific activity. of. the primary coolant was routinely 'sampled and analyzed as required by MNP-2-,Technical Specifications, and was in all cases, less than or equal to 100/E microcuries per gram. A graph - showing cumulative iodine dose equivalent for.the calendar year 1989 fol 1 ows.

() c~ ~ e ucX/gm 0.030 C9 Q (dl ) EACTOR DOSE.EQUIVAL'EN'T WNP-2 llll l I IODINE Z2DEQ PFlX S I ) I <)1 (3 1 () (i/i lO 4 ()tl; il I lil I'I'I 0.025 I:)r)i):)r yr 1 1 <1tl)f I t) t>)-) )i)>>f<< 1 f, ftftlfl 0.080 0.0%6 O.oio 0.005 44* Il l I )III )II ll P( I(>lI i I I I (llli~'I )N)gi ll)lbllllll +) I I ii.,<<j ilies 0.000 W.ODBOf-Ol 03-02 05-01 '6-30 08-29 %0-28 - Januar y I, %989 to December 3i, %989 42 27

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2.10 REPORT OF OIESEL GENERATOR FAILURES ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ This section contains information pertaining to the reporting of diesel generator failures, valid and nonvalid, in accordance with the requirements of MNP-2 Techqical Specification 4.8.1.1.3. This report provides the =:.-.information;-required by Regulatory Position C.3;b - of, Regulatory GUide 1.108, Revision 1, August 1977. - 81

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Diesel Generator Failure Number One 1. Identity of diesel generator unit and date of failure: Division Three Emergency Diesel Generator (OG-3) Hay 12, 1989 2. Number designation of failure in last 100 valid tests,. Not applicable. This was a nonvalid failure. The unit was inoperable for maintenance overhaul activity. 3. .Cause of failure: The-exact cause of the failure was not able to be determined. During --a slow start test (nonvalid test) run of OG-3, the unit started at . full.speed (900 RPH) rather than slow speed (400= RPM). as designed. Subsequent tests were not able to duplicate the failure.-. 4. Corrective measures taken: None. 5. Length of time diesel generator unit was unavailable: -- e'ot applicable for this nonvalid failure. Current surveillance test interval: Thirty-one days. 7. Verification of test interval: " The surveillance test interval of thirty-one days is in conformance with NRC Reg. Guide 1.108 position C.2.d.

Diesel Generator Failure Number Two Identity of diesel generator unit and date of failure: Division Three Emergency Diesel Generator (OG-3) Hay 13, 1989 2. Number designation of failure in last 100 valid, tests: Not applicable. This was a nonvalid failure. The= test was. a nonvalid test-because it was testing a feature which was not-,a. part of the defined di.esel generator unit design., 3. Cause of fai,lure: The-failure was the result of 'a nonvalid test performed to discover if DG-3 would start with one of two starting air headers .. isolated. The ability to-start on one air header is not part of:-the OG-3 design. The cause of.'the failure to start was insufficient.capacity of only one 'tarting air header. The design of this system calls for two starting air headers. 4. Corrective measures taken: None. Length of time diesel generator unit was unavailable:. Not applicable. This Has a nonvalid test. 6. Current surveillance test interval: Thirty-one days. 7. Verification of test interval: -..The, surveillance test interval of thirty-one days is in conformance with NRC Reg. Guide 1.108 position C.2.d. 0

Diesel Generator Failure Number Three l. Identity of diesel generator unit and date of failure: 3 ~ ~ ~ Division One Emergency Diesel Generator (DG-1) May 18, 1989 2. Number designation of Failure in last 100 valid tests: Not applicable. This was a nonvalid failure. The unit was inoperable for maintenance overhaul activity. 3. Cause of failure: -No definite cause of failure was able to be identified.= The unit tripped during an end of maintenance warranty run prior to declaration oF . operability. A thorough investigation was unable to identify a definite cause.- -The fault trip was not able to be repeated during follow-up testing. 4. Corrective measures taken: f = =-==The 18=month overhaul procedure was modified to include a specific check of the manual overspeed trip mechanism. 5. Length of time diesel generator unit was unavailable:-. Not applicable for this nonvalid failure. J 6. Current surveillance test interval: Thirty-one days. 7. Verification of test interval: -.--. The'urveillance test interval of thirty-one days is -in. conformance with NRC Reg. Guide 1.108 position C.2.d. /a

Diesel Generator'Failure Number Four Identity of diesel generator unit and'date of failure. Division Two Emergency Diesel Generator (DG-2) May 20, 1989 2. Number designation of failure in last 100 valid tests. Not-applicable. This was a nonvalid test failure. Per WNP-2 Technical Speci fication Tab 1 e 4.8.1.1. 2-1, with the except ion of the semiannual fast start, no starting time requirements are required to meet the valid test requirements of NRC reg. Guide 1.108. 3. Cause of failure: During performance of the 18 Month Logic System Functional Test DG-2 Loss of Power Test, the diesel generator did not attain rated speed within ten seconds of receiving a start signal. The cause of the failure was originally isolated to a broken pneumatic boost line which supplies the Woodward speed governor unit. This prevented the start boost signal from being received by the actuator and would have resulted in a decrease in fuel supply to the diesel during fast start. 4. Corrective measures taken: The pneumatic line was repaired and the unit was retested.. The retest did not demonstrate acceptable starting time. (See Diesel Generator Failure 45.) 5. Length of time diesel generator unit was unavai.lable: Not applicable for this nonvalid test. 6. Current surveillance test interval: Thirty-one days 7. Verification of test interval. The surveillance test interval of thirty-one days is in conformance with NRC Reg. Guide 1.108 position C.2.d.

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Diesel Generator Failure Number Five .e'dentity of diesel generator unit and date of failure: Division Two Emergency Diesel Generator (DG-2) May 24, 1989 2. Number designation of failure in last 100 valid tests: Not appli:cable. This was a nonvalid test failure.

Per, WNP-2 Technical Specification Table 4.8.1.1.2-1, with the exception of the semiannual fa'st start, no starting time requirements are required to meet the valid test requirements of NRC reg.

Guide 1.108. 3. Cause of failure: During performance of the 18 Month Logic System Functional Test DG-2 Loss of Power -Test, the diesel generator did not attain rated speed within ten seconds of receiving a start signal. The cause -of the failure was isolated to the voltage permissive relay DG-RLY-59/DG2 which provides a permiss.i-ve: signal to close the DG2 output breaker when generated voltage is - high - enough. The relay actuation setpoint calibration tolerance was found to'llow sufficient variation to affec't the 10 second start time under certain conditions. 4. Corrective measures taken: e-The relay was recalibrated to obtain sufficient setpoint performance to -ensure--obtaining a maximum 10 second start time.

This mechanical relay was -.later replaced'ith a solid state relay which could be calibrated to perform consistently within the required tolerance.

5. Length of time diesel generator unit was unavailable: Not applicable. This was a nonvalid failure. 6. Current surveillance test interval: Thirty-one: days. 7. Verification of test interval: The surveillance, test interval of thirty-one days is in conformance with NRC Reg. Guide 1.108 position C.2.d. - 86

Oiesel Generator Fai.lure Number Six 1. 'Identity of diesel generator unit and date of failure: Oivision Three Emergency Oiesel Generator (OG-3) ~ ~ ~ June 2, 1989 2. Number designation of failure in last 100 valid. tests: Not applicable. error. This was a nonvalid failure as it was. due to personnel 3. Cause of failure: Ouring performance of the Logic System Functional Loss of Power Test, the diesel'perator did not apply sufficient load soon enough after synchronization with the power grid to prevent a reverse power trip of the OG unit. 4. Corrective measures taken: e, The operating procedure was'valuated for correctness and:-found to be acceptable. The operator was counselled. Length of time diesel generator unit was unavailable: OG3 was'navailable for approximately ten minutes while the protective relays were being reset. Current surveillance test interval: Thirty-one days 7. Verification of test interval: -.--. The.surveillance test interval of thirty-one days fs =in conformance with NRC Reg. Guide 1.108 position C.2.d. 0

. Diesel Generator Failure Number Seven Identity of diesel generator unit and date of failure: Division One Emergency Diesel Generator (DG-1) June 9, 1989 2. Number designation of failure in last 100 valid tests: -'ot app'licable. This was a nonvalid failure. This was not.a valid test . failure.- .Per MNP-2 Technical Specification Table 4.8.1.1.2-1, with the exception of the semiannual fast start, no starting-.time-requirements are required to meet the valid test requirements of NRC Reg.- Guide 1.108. 3. Cause oF failure: ='uring performance of the 18 Month Logic System functional Test DG-1 Loss of Power Test, the diesel generator did not attain rated speed within ten seconds -of receiving a start signal. The cause. of the Failure was incorrect connection of the start pneumatic boost signal to the Moodward -'- speed -governor unit. This resulted in insufficient;:fuel.-supply to the diesel during fast start to ramp speed at the required-,- rate.,=.=- 4. Corrective measures taken: Th'.'neumatic line was connected .to the correct port--on = the governor actuator. The unit was then retested successfully: The other diesel units were inspected for similar fault. 5. Length of. time diesel generator unit was unavailable: h Not applicable to this nonvalid start. 6. Current surveillance test interval: Thirty-one days 7. Verification of test interval: . The surveillance test interval of thirty-one days is in conformance with - NRC Reg. Guide 1.108 position C.2.d.

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Oiesel Generator Failure Number Ei ht Identity of diesel generator unit and date of failure:. Oivision One Emergency Oiesel Generator (OG-1) June 10, 1989 2. Number designation of failure in last 100 valid, tests: Not applicable. This was a 'nonvalid failure. The failed valve is not a part of the defined diesel generator unit design. 3. Cause of fai.lure: 4 ~ The. OG-1 -output circuit breaker was tripped by protective relay actuated by high engine cooling water temperature.

This, in turn, shutdown the diesel= generator.

The high engine cooling water temperature was caused by failure of the Standby Service Mater System cooling water inlet valve which:..supplies cooling water to the engine cooling water.heat exchanger. The valve-. disk separated from the operating stem and remained in the

closed,

= position blocking cooling water flow..The-. pneumatic valve operator,;.however, stroked fully open showing a ful:1::open valve position indication. P Corrective measures taken: The faulty. valve was removed. The possible generic'mplications of this failure;; -were inves tigated. These va 1 ves on ihe rema ining Oi ese 1 Generator units were removed. (See PER 289-0487.) Length'f time diesel generator unit was unavailable: Not applicable to this nonvalid failure. 6. Current surveillance test interval: ,Thirty-one days 7. Verification of test interval: The surveillance test interval of thirty-one days is in conformance with NRC Reg. Guide 1.108 position C.2.d. - 89

1

2.11 FIRE PROTECTION PROGRAM CHANGES The following changes were made to the fire protection program during the calendar year. - These revisions were all made to plant procedure 1.3.10, Fire Protection

Program, in which the procedure is included as part of the FSAR by reference.

The procedure was revised to require all detectors in vital areas to .be operational at all times. If they are not operational compensatory measures must be taken. The revision is more restrictive than the previous requirements. 2. Detector maintenance activities were removed from plant procedure 1.3.10 and moved to . volume 15 procedures. Haintenance activities , will.be performed in accordance with the applicable NFPA standards, as well as insurance company and manufacturer recommendations. Maintenance activities will be scheduled via the Scheduled Maintenance System (SMS). 3. The minimum requirements for fire protection system pump and water = supply operability were clarified. The requirement is: two 2000 gpm pumps and the 2500 gpm pump must be operable at all times. 4. Fire Suppression Hater System Maintenance activity descriptions were removed from plant procedure 1.3.10 and moved to volume 15 procedures. Fire Suppression Hater System Maintenance activities

will be performed in accordance with the applicable NFPA standards, as well as insurance company and manufacturer recommendations.

Maintenance activities will be scheduled via . the Scheduled Maintenance System (SMS). 5.= Operability requirements for hydrants within the protected area were changed to require operability of only those hydrants that provide protection for equipment that, is required to be operable. This is more restrictive than the previous requirement. .6. The compensatory actions associated with an impaired hydrant or hose -.house were changed in that a hose is now required to be placed at an adjacent hydrant/hose house in 24 hours instead of one hour. The basis-for the change is that a van is used by the fire brigade when responding to fires in the protected area. The van has hose on board and can be used to rapidly lay the hose for fire fighting. 7-. Maintenance activities associated with hydrants were removed from 'plant procedure 1.3.10 and moved to volume 15 procedures. Maintenance activities will be performed in accordance with the applicable NFPA standards, as well as insurance company and manufacturer recommendations. Maintenance activities will be scheduled via the Scheduled Maintenance System (SMS). 90-

IH

8. Operability requirements for hose stations were changed to require all hose stations in the Corridors,

Turbine, Reactor, Radwaste and Diesel Buildings be operable anytime the equipment which the hydrant provides protection for is required to be operable.

This is more restrictive than the previous requirement. " 9. Hose station maintenance activity descriptions were removed from plant= procedure 1.3.10 and moved to volume 15 procedures. Hose station maintenance activities will be performed in accordance with - the 'pplicable NFPA standards, as well as insur'ance .company and manufacturer recommendations. Maintenance activities will be scheduled via the Scheduled Maintenance System (SMS). 10; The requirements for a fire watch to be stationed for inoperable control room halon systems was removed since the control room is manned 24 hours per day and there are halon extinguishers available. A fire protection impairment must be issued. ll: Halon-fire protection maintenance activity descriptions were removed from plant procedure 1.3.10 and moved to volume 15 procedures. Halon fire protection maintenance activities will - be performed in '-accordance with the applicable NFPA standards, as well as insurance company 'and manufacturer recommendations.. Maintenance activities will be scheduled via the Scheduled Maintenance System (SMS). e '2. .The'equirements for compensatory measures associated with =suppression systems were changed to require a fire impairment . permit. The requirement of a fire watch is needed only if a system is inoperable and the associated detection system is inoperable. If the detection system is operable an hourly fire tour must be established. 13. Various valve and fire protection equipment maintenance activity descriptions were removed from plant procedure 1.3;10 and moved to volume 15 procedures. Valve and fire protection maintenance activities will be performed in accordance with the applicable NFPA standards, as well as insurance company and manufacturer recommendations. Maintenance activities will be scheduled via the Scheduled Maintenance System (SMS). 14.. The-requirements for various inspection activities were changed to be ;in compliance with NFPA standards and insurance company and manufacturer recommendations. It additionally states that the maintenance will be scheduled via - the Scheduled Maintenance System (SMS). The.. modifications to the Fire Protection Program as - described above did not r.esult 'in a change to the NNP-2'echnical Specifications since the Fire Protection, Te'chnical Specifications had been removed.-per Amendment 67. The unreviewed safety , question evaluation . concluded:- -:(1');- -the function and ~ performance of the Fire Protection Program did not" change; (2) the margin of safety provided in the technical, specifications was.not

changed, and (3)'he boundary conditions for the FSAR evaluations were not'changed.

'91.- '

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