Text

Monthly Operating Repts for Jul 1981
	ML20010C484
Person / Time
Site:	Surry
Issue date:	08/05/1981
From:	Dunn S; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML20010C483	List: ML20010C482; ML20010C484;
References
	NUDOCS 8108200124
	Download: ML20010C484 (49)
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VIRGINIA ELECTRIC AND POWER CG!PANY SURRY POWER STATION MONTHLY OPERATING REPORT REPORT NO. 81-07 JULY. 1981 t

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D 8108200124 010817 PDR ADOCK 05000200 R PDR

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CONTENTS Section h Operating Data Report - Unit #1 i .

4 Operating Data Report - Unit #2 2 Unit Shutdowns and Power Reductions - Unit il 3 Unit Shutdowns and Power Reductions - Unit #2 6 Load Reductions Due to Environmental Restrictions - Unit #1 7

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load Reductions Due to Environmental Restrictions - Unit #2 . 8 Average Daily Unit Power Level - Unit #1 9 i Average Daily Unit Power Level - Unit #2 10 Sumary of Operating ' Experience 11 ,

Amendments to Facility License or Technical Specifications 14 Facility Changes Requiring NRC Approval 15 Facility Changes That Did Not Require NRC Approval 15 Tests and Experiments Requir'ing NRC Approval 23 Testr., and Experiments That Did Not Require NRC Approval 23 Other Changes, Tests and Experiments 24 l

Chemistry Report 25 Description of All Instances Where Thermal Discharge Limits 26 Were Exceeded ,

Fuel Handling 27 Procedure Revisions That Changed the Operating Mode Described in 28 the FSAR Description of Periodic Tests Which Were Not Completed Within the 29 Time Limits Specified in Technical Specifications l Inservice Inspection 30 Reportable Occurances Pertaining to Any Outage or Power Reductions 32 Maintenance of Safety Related Systems During Outage or Reduced Power 33 Periods - Unit #1 - Mechanical Maintenance ;

Maintenance of Safety Related Eystems During Outage or Reduced Power 36 Periods - Unit #2 - Mechanical Maintenance 5

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Haistenance of Safety Related Systens During Outage or Reduced Power 38 Perioda - Unit #1 - Electrical Maintenance Maintenance of Safei, Related Systems During Outage or Reduced Power 41 Periods - Unit #2 - Electical Maintenance Maintenance of Safety Related Systems During Outage or Reduced Power 42 Periods - Unit il - Justrument Maintenance

' Haintenance of Safety Related Systems During Outage or Reduced Fower 44 Periods - Unit #2 - Instrument Maintenance 45 -

Health Physics Sunnary ,

1 Procedure Deviations reviewed by Station Nuclear Safety and 46 Operating Committee after Time Limits Specified in T. S.

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1 OPERATING NTA EPORT DOCKET NO. 50-280 DATE 05 AUG 81 CONPLETED BY SUE D. DUNN TELEPHONE 804-357-3184 DPERATING SIATUS

1. UNI't NAME SURRY UNIT 1

2. REFORTING PERIOD 70181 10 13181

3. LICENSED THEb!AL POWER (WT) 2441 l~ l

4. NAMEPLATE RATING (GROSS WE) 847.5 l MOTES l

5. DESIGN ELECTRICAL RATING (NET nee) 188

6. NAXIMUM DEPENDABLE CAPACITY (GROSS WE) 811

7. NAXIMUM DEPENDABLE CAPACITY (NET kvE) 7'15

8. IF CHANGES OCCUR IN CAPACITT RATINGS NJA (ITEMS 3 THROUGH 7) SINCE I.AST REPORT. GIVE REASONS

9. PWER IEVEL TO WHICH RECIRICTED, IF ANY N/A (NET WE)

10. REASONS FOR RESTRICTIONS, IF ANI N/A THIS MONTH IR-TO-DATE CUMULATIVE

11. HOURS IN REPORTING PERIOD 744.0 5087.0 15455.0

12. NUMBER OF HOURS REACTOR WAS CRITICE 556.1 556.7 43094.6

13. MACTOR RESERVE SHUTDWN HOURS 0.0 0.0 3731.5

14. HOURS GENERATOR ON-LINE 505.0 305.0 42173.8

15. UNIX RESERVE SHUTDOWN HOURS 0.0 0.0 3736.2

16. GROSS THER4AL ENERGY GENERATED (WH) 1061380.0 1061380.0 97450781.0

17. GROSS ELECTRICE ENERGY GENERATED (WH) 342480.0 342480.0 31644223.D

18. NET ELECTRICE ENERGX GENERATED (WH) 320551.0 320551.0 30020475.0

19. UNIT SERVICE FACTOR 67.9 */* 9.9 */* 55.9 */*

20. UNIT AVAILABILTTY FACTOR 67.9 */* 9.9 */* 60.8 */*

21. UNIT CAPACITX FACTOR (USING NDC NET) 55.6 */* 8.1 */* 51.3 */*

22. UNIT CAPACIT! FACTOR (USING DER NET) 54.7 */* 8.0 */* 50.5 */*

23. UNIT FORCED DUTAGE RATE 9.8 */* 9.8 */* 26.2 */*

24. SHUTDWNS SCHEDULED OVER NEXT 6 NONTHS 2/19/82 - SPRING NAIET (TIPE DATE.AND DURATION OF EACH) APPROI.10 DAYS

25. IF SHUT DNN AT END OF REPORT PERIOD.

ESTIMATE DATE OF FIARTUP

26. UNITS IN TEFI STATUS FORECASE AC'!IEVED (PRIOR 70 COMMERCZE OPERATICE)

INITIAL CRITICEITT l INITI E ELECTRICITY COMMERCIAL OPERATION

- OPERATING DATA REPORT DOCKET NO. 50-281 DATE 05 AUG 81 CO4PLETED BY 0.J. COSTELLO TELEPHONE 804-357-3184 OP_ER4 TING STATUS

1. UNIT NAME SURRY UNIT 2

2. REPORTING PERIOD 70181 TO 73181

3. LICENSED THERMAL POWER ( W T) 2441 l l

4. NAMEPLATE RATING (GROSS WE) 847.5 l NOTES l

5. DESIGN ELECTRICAL RATING (NET WE) 788

6. NAXIMUM DEPENDABLE CAPACITY (GROSS WE) 811

7. MAXIMUM DEPENDABLE CAPACITX (NET WE) 775

8. IF CHANGES DCCUR IN CAPACITX RATINGS N/A (ITE45 3 THROUGH 7) SINCE LAST REPORT, GIVE REASONS

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9. P W ER LEVEL TO WHICH RESTRICTED. IF ANY N/A

(#ET WE) i 10. REASONS FOR RESTRICTIONS. IF ANY N/A l

HIS MONTH 1R-70-DATE CUMULATIVE

11. HOURS IN REP.)RTING PERIOD 744.0 .5087.0 72335.0

12. NUMBER OF HOURS REACTOR WAS CRITICAL 729.8 4823.8 42607.8

13. REACTOR RESERVE SHUTDOWN HOURS 0.0 0.0 0.0

14. HOURS GENERATOR ON-LINE 725.9 4785.3 41921.9

15. UNIT RESERVE SHUTDOWN HOURS 0.0 0.0 0.0

16. GROSS THERMAL ENERGY GENERATED (WH) 1760531.5 11591227.5 98045888.5

, 17. CHOSS ELECTRICAL ENERGY GENERATED (WH) 554770.0 3755050.0 31994044.0

18. NET ELECTRICAL ENERGY GENERATED (WH) 524711.0 3558890.0 30337378.0

19. UNIT SERVICE FACTOR 97.6 */* 94.1 */* 58.0 */*

20. UNIT AVAILABIL:T! FACTOR 97.6 */* 94.1 */* 58.0 */*

l 21. UNIT CAPACITY FACTOR (USING NDC NET) 91.0 */* 90.3 */* 54.1 */*

l 22. UNIT CAPACITX FACTOR (USING DER NET) 89. 5 */*- 88.8 */* S3.2 */*

23. UNIT FORCED OUTAGE RATE 2.4 */* 1.4 */* 18.0 */*

24. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHE 8/27/81 - SNUBBER INSP-3DATS (TXPE.DATE.AND DURATION OF EACH) 11/13/81 - REFUELING APPROX - 42 DAYS -

l 25. IF SHUT D0dN AT END OF REPORT PERIOD.

ESTIMATE DATE OF STARTUP

26. UNITS IN TEST STATUS FORECAST ACHIEVED (PRIOR TO CMMERCIAL OPERATION)

INITIAL CRITICALITX IhITIAL ELECTRICITY

COMMERCIAL OPERATION

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UNITSHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 50-280

, UNIT NAME _Surry One DATE Au m e 7_ 10A1 COMPT.ETED BY S. D. Dunn REPORT MONTH JULY' 1981 TELEPHONE 804 357-3184 i -

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No. Date H

( 7. g jE E i 2s5 Event Repor #

g'2 wO g{g Action to Prevent Recurrence j ;p, g 6

81-7 07-01-81 S 184.4 H 1 Continuation of sihutdown for steam generator replacessent which counsenced 09-14-80.

81-8 07-09-81 F 50.0 11 2 81-026 /03L-0 Manually tripped turbine & reactor du e to increasing containnient pressure and sump level caused by secondary t system leakage through blown gasket o 1 's#

auxiliary feed water line flow restricting orifice flange. Repaired flange and inspected other prior to unit recovery.

i I 2 3 4 F: Fo ced Reason: Method: E<hibit G. Instructions S: Scheduled A Equipment Fallu.e(Explain) 1 Manual for Preparation or Data B. Maintenance of Test 2. Manual Scram. Eniry Sheets for Licensee C.Rerueling 3. Automatic Scram. Event ReporI (LER) File INUREG-D Regulatory Restriction 4.Other (Explain) 0161) .

E-Operator Training & License Examination

F. Administrative S G-Operational Error (Explain) Exhibit I Same Source 19/77) Il-Other (Explain) ,

o e C UNITSHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 50-280 UNIT NAME surry N.

DATE Anyme 7 10A1_

COMPLETED BY t n . man REPORT MONTH JULY. 1981 TELEPHONE nru tsv. tina "i -

.IE 3 .h Ik Licensee E3, b Cause & Corrective N". Datt k 58 .E s 5 Event u '2 Action tu F

$E 5 jgg Report Jr vY0 [k.,

O Present Recurrence d

81-9 07-15-81 F 0.0 H 4 A malfunction in the "up-down counter" circuit caused the valve position limiting circuit of the Electro-Hydraulic control system to run the turbine back to approximately 60% power. Runback stopped when control room operator shifted turbine L control to manual. Control was left in turbine manual mode and pc .r increased.

I 2 3 4 F: Forced Reason: Method: Exhibit G Instructions S: Scheduled A Equipment Failure (Explain) 1-Manual for Preparation of Data B Maintenance of Test 2-Mar,ual Scram. Entry Sheets foi Licensee C Refueling 3-Automatic Scram. Event Report (t .:R) File INilREG-D-Regulatory Restriction 4 Other (Explain) 0161)

E. Operator Training & Ucense Examination F Administrative 5 G-Operational Error (Explain) Eshibit I-Same Simrce .

19/77) flother (Explain) ,

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o UNI) *HUTDOWNS AND POWER REDUC 110NS DOCKETNO. 50-280 UNIT NAME SurrY One DATE -August 7. 1981 N ETED p S. A Nnn RCPORT MONTH JULY. 1981 TELEPHONE PO& M7-11R4 ,

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Licensee E*,, g" ., Cause & Corrective !

Er. Date { Eg 5 2gE Event g7 ^'"'" '

H jE E Report # m0 Eu j if, g g Prevent Recurrence 6

81-10 07-16-81 F 4.6 H 2 A inalfunction in the "up-down counter" circuit caused the valve j:

limiter circuit to run the turbine i.

back to a "no-lo'ad" condition. The l:

control room operator manually :

tripped the turbine siid reactor. The problem was caused by a " bad card" 8 l

in the "up-down counter" circuit. Y '

The card was replaced. '

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1 2 3 4 F: Forced Peason: Method: Exhibit G Instructions 5: Scheduled A-Equipment Failure (Explain) l Manual for preparation or Dats B. Maintenance of Test 2 Manual Scram. Entry Sheets for IAensee C-Rerueling 3 Automatic Scram. Event Report (LER) File (NUREG-D Regulatory Restriction 4-Other (Explain) 0161)

EOperator Training & Ucense Examination . .

.. F-Administrative 5 GOperational Error (Explain) Exhibit I Same Source ,

, 19/77) Il-Other (Explain) ,

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e DOCKET NO. 50-281 UNIT SHUTDOWNS AND POWER REDUCTIONS UNIT NAME Surry Two DATE A'a-in s t 7- 19Al NETED BY S. D. Dunn

' REPORT MONTH JULT. 1981 TELEPHONE AO& 157 ~11AA 3

.!'E 3g ,hIh Licensee E*, ,

Cause & Corrective l

No. Date F

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Action to Prevent Recurrence j ;f, g 6

I 81-8 07-17-81 F 14.4 H 3 A safety injection occurred on a spurious high steam header to steam ,

line differential pressure signal.

Cause of spurious signal is believed ,

to have been vibration.

1.0 3 A tuttine trip-reactor trip occurred i.

81-9 07-18-81 F ~d on "A" S/G Hi-Hi Level signal due to i '

leakage through tihe skin . feed ,

t regulating valve while increasing -

t power. ,

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81-10 07-18-81 F 2.7 H 3 Same as 81-9 above.

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3 4 I 2 Reason: Method: Exhibit G Instructions F: Fneced for Preparation of Data A Equipment Failure (Explain) 1-Manual 5: Scheduled Entry Sheets for Licensee B. Maintenance of Test

- 2-Manual Scram.

3 Automatic Scram. Event Report (LER1 File (NUREG-C-Refueling D Regulatory Restrktion 4-Other (Explain) 016l)

E-Operator Training & Ucense Examination F Administrative G-Operational Error (Explain) Eshibit I Same Source 19/77) ll Other (Explain)

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LOAD REDUCTIONS DUE TO ENVIRONMENTAL RESTRICTIONS E T NO.2 MONTH: JULY. I981 DATE TI!E HOURS LOAD. MW REDUCTIONS. W . MWH msg

- NONE DURING IIIS REPORTING PERIOD.

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LOAD EDUCTIONS DUE TO ENVIRONMENTAL ESTRIC7 IONS UNIT NC.1 MONTH: JULY. 1981 5

HCJRS LOAL. W EDUCTIONS. W MWH EASON DATE TI!E NONE DURLNG THIS REPORTIN PERIG). ,

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DOCK M #0 50-280 UNIT SURRY I DATE 8-1-81 CONPLPIED BY Sue D. M an AVERACE DAIU UNIT PWER LEVEL NONTHs JUU 81 ,

AVERAGE DAIU POVER E2 VEL AVERAGE DATU POVER GVEL i DM . (WE-NET) DM ( WE-NET) i 0.0 17 350.9 2 0.0 18 701.1 '

3 0.0 19 747.1 ,

4 0.0 20 752.0 5 0.0 21 753.2 6 0.0 22 754.8 7 0.0 23 755.5 _

8 51.6 24 7 54.5 9 205.7 25 748.2 10 0.0 26 740.1 11 4.5 27 748.8 12 257.5 28 745.2 [

13 280.5 29 747.1 14 601.5 30 747.9 15 592.9 31 736.3 16 579.5 DAIU UNIT POVER LEVEL F0P*! INSTRUCTIONS ON THIS FORN. UST THE AVERACE DAILY UNIT POWER LEVEL IN hWE-NRT FOR EACH DAT IN THE REPORTING MONTH. THESE FIGURES VILL BE USED TO PWT A GRAPH FOR EACH REPORT-ING MONTH. NOTE THAT Bf USING M!.XIMUN DEPENDABLE CAPACITI FOR THE N:7 ELECTRICAL RATING OF THE UNIT THERE MAY BE OCCASIONS VMEN THE DAILY AVERAGE FWER EXCESDS THE 100 e/* UNE (OR THE RESTRICTED PWER EZYEL UNE). IN SUCH CASES THE AVERAGE DAZU UNIT PWER QUTPUT SHERT S80:12 BE FOOTNOTED TO EXPLAIN THE APPARENT ANONALT. '

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,V DOCKET NO 50-281 UNIT SURM II DATE 8-1-81 CONPEEEED E Sue D,Dunn AVERAGE DAIU UNIT PG/ER UVEL N.nNTHs JUU 81 AVERAGE DAIU POWER MVEi, AY: RAGE DAIU POWER MVEL DM (WS-MET) DAI (WE-NET) 1 734.2 17 349.3 2 735.5 18 463.0 3 730.3 19 725.8 4 721.2 20 738.0 5 714.3 21 739.8

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6 710.4 . 22 736.0 7 708.1

. 23 738.5 '

8 711.6 24 739.8 9 712.2 25 737.8 10 708.7 26 738.6 11 705.3 27 734.4 12 708.0 28 734.0 13 719.2 29 735.4 14 715.6 30 737.3 15 724.5 31 . 723.6 16 735.1 D4IU UNIT PWER MVEL FOM INSTRUCTIONS ON THIS FOM. UST THE AVERAGE DAIU UNIT PCVER LEVEL IN WE-NET FOR EACH DM IS TH: REPORTING MONTH. THESE FIGURES L'ILL BE USEO TO PLOT A GRAPH FOR EA:8 REPORT- '

ING MONTH. WOTE TllAT M USING MAHNUM DEPE4DABM CMACITI FOR THE RET ELECTRICAL RMING OF THE UNIT. THERE NAX BE OCCASIONS WHEN THE DAIU AVERAGE F0lER EX:ESD3 THE 100 */* LINS (OR THE RESTRECTED FWER MVEL UNE). IN SlCH CASES. THE AVERAGE DAIU UNIT P&!ER DUTPUT SHEET SHOULD BE FORNOTED TO FXPLAIN THE APPARENT ANONAU.

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11 SUIMARY OF OPERATING EXPCLIEN2 Month / Year July, 1981 l

l Listed below in chronological sequence by mit is a sumary of '

operating experience for this month which required load reductions or resulted in significant non-load related incidents.

4 UNIT 1 July 1 2his reporting period begins with the mit at cold shutdown with preparations for unit recovery in progress. !

July 3 Baated primary system above ccid shutdown (200 0 F) at 0915. Es tablished bubble in pressuriser t.t 1230.

July 4 Exceeded 350 F primary system :egerature at 0650. '

July 5 Established hot shutdown conditions at 0455. -

July 6 De reactor was taken critical at 1712 and low power physics testing comence d.

July 7 Rod control system urgent failure occured at 0040. De Ins trument Lehnicians investigated and determined roblem was a blown fuse in the "1BD" Power Cabinet. Rey replaced the fuse and returned rod control system to service ar 0320.

At 1020 a reactor trip occured when the Ive:ruar.nt 'Dechniciacs simulated i

e fix:t stage turbine pressure >10% without the turbine la tchedi he reactor was critical at 1045. ,

1 i July 8 he generator was synchronized to the line at 1623. Power was increased to 38% and held for resolution of steam generator chemistry problem.

i July 9 S/G chemistry within specifications and power increase to 50% power at 3% per hour bega.n at 1100. Unit reached 50% power at 1440. At 2030 commcaced a rampdown from 50% power due to increasing containment partial pr:ssure and containment sump level. At 2054 the decision was made to manually trip the turbine and reactor ad commence a plant .

conidown. Se cooldown cousenced at 2100 and a safety injection initiated i

at 2127. Se SI was a result of the control room operator failing to block the header to line AP SI signal prior to comumencing the cooldown.

July 10 Se primary sys tem temperature was <350*F at 0200 and <200 F at 0607. .

July 11 De primary system temperature exceeded 200*F at 0018 and 350'F at 0545. The mit. reached hot shutdown conditions at 1215. Se reactor was !

critical at 1758. A header to line AP SI occurred at 1817. It is believed the SI occurred as a result of pressure swing on the main steam he ader when the turbine was latched. he reactor was critical at 2114 and ~

the generator was synchronized to the line at 2252. Stopped power increase at 43% at 2400 to clean g steam generator chemistry.

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l July 12 Commenced power increase to 50% power at 3% per hour. Cosamence d S/C stability tests at 50% power at 2325.

July 13 Completed 10% load reject test at 0235. Test was satisfactory and 4 unit at 40% power. At 1325 courr. aced increasing power at '3% per

. hour. Stop power increase at 2145 at 66% power for flut mapping.

July 14 Started powcr increase to 90% at 3% per hour. Power at 90% at 1031. Started power increase at 3% per hour. Power at 100%

at 2120.

i 4 July 15 ' At 1138 a malfunction in the "Up-Down" counter circuit of the electro-hydraulic control system caused a turbine runback. Runback stopped when operator shifted turbine control to manual. Conditions were stabilized at 60% power and 470 Nie. Stsrted power increase at 3%

i per hour from 63% power at 1735.

July 16 thit reached 100% power at 0710. At 1952 the reactor was manually tripped whan a malfunction in the "th-Down" comte circuit of the EHC system causeu the valve limiter to reduce turbine power to a 3 no load condition. De rea tor was critical at 2214.

h July 17 Performed a turbine over speed test at 0006. De turbine tripped at 1900 RPM. De generator was synchronized to the line at 0028.

and power increased to 40%. Power was held at 40% for clean up of i

} S/G chemirtry. At 1027 S/G chemistry was within specifications and

! a power increase was commenced. At 1120 stopped power increase at

65% until "B" main feed ptsop is retutsed to service. "B" NFP returned to service and power increase comumenced at 3% per hour at 1730.

i July 18 Step power increase at 89% power due to increasing cation conductivity

! indication. Started power increase at 3% per hour at 0328. Stoppe d pcser increase at 95% at 0600. Started power increase at 0818 and reached 100% power at 1020.

j July 20 Integrated 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> at 100% power as of 0900.

July 25 he steam generator carryover test was cogleted at 1950.

! July 31 At 1233 a turbine rmback occurred when a Daniels Construction Cogany i electrician accidently opened the breaker supplying the N-43 power range instrument. De electrician was replacing the. panel coruer on i VB-1-I when the cover slipped and tripped the breaker. De runback I terminated at 86% power when the breaker was reclosed. At 1311 a power increase began and the thit was at 100% power at 1441. Bis reporting period c.nds with the unit at 100% power.

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l INIT 2 July 1 Mis reporting period begins with the unit at 100% power.

July 14 At 2230 the power was reduced to 96% to correct a condenser vacusa problem caused by removing the "2D" waterbox from service. j July 15 he "2D" waterbox was returned to service and power increased to 100% at 0415.

July 17 A reactor trip and safety injection occurred at 1122. De SI and reactor trip were caused by a header to line AP signal, which was determined to be spurious.

July 18 he reactor was critical at 0032 and the generator on the line at ;

0150. At 0200 a turbine trip - Reactor trip occurred due to a !

hi-hi level in the "A" S/G. De hi-hi level was a result of leakage thru the main feedwater regulating i alve, ne reactor was critical at 0223 and the generator ca the line at 0302. At 0325 a turbine trip - reactor trip occurred due to a hi-hi level in the "A" S/G. De reactor was critical at 0409 and the generator on the line at 0606. De unit reached 100% power at 1221.

July 31 his reporting period ends with the smit at 100% power.

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i AMEIOMENTS TO FACILITY LICENSE l

OR TECHNICAL SPECIFICATION JULY, 1981 i

v NONE DURING THIS REPORTING PERIOD.

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FACILITY CHANGES REQUIRHfG NRC APPROVAL NONE DURING THIS REPORTING PERIOD.

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FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL JULY. 1981 Ei1T_

DC-76-14 - Contaiment Insturment Air System Modification 1 This design change installed new compressors of larger capacity. The location of the new compressors is in the lower elevation of the saf tguards building 1.e.,

11'6" elevation. Access for the suction and discharge piping to t!.e containment penetretions will be through existing holes between auxiliary and safeguards build-ings.

Summary of Safety Analysis This modification requires revision to the FSAR and no changes to Technical Specifications. There are no safety implications as a result of this dr.tsign change.

DC-77-09 - Containment Spray System Modification 1 This design change meets the following objectives:

a) increased spray coverage b) Decreased caustic spray transit time to the spray nozzles.

c) Addition of caustic solution to the spray water at an essentially constant rate.

d) Provide caustic spray until the contaiment is depressurized under all operating modes of the safeguards system.

Sunanary of Safety Analysis This modification requires no change to Technical i Specifications. A review of the FSAR indicates a !

change to the function of the affected systems are i mot requried. An unreviewed safety question does not exist.

DC-77-323 - Contaitanent CoolinR System Modification 1-This design change installed new chilled water units and piping into the new system. ,

Sununary of Safety Analysis There f.s no change in the operation of safsty related equipment as a result of this design change.

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FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL (CONTINUED)

UNIT DC-78-08 - Reactor Contairanent Fire Protection 1&2 Portions of this design change' involving installation of a Smoke Detection System (78-8E) and Hose Station in Contalment 1 (78-8J (1)) were implemented.

Summary of Safety Analysis This modification increases the protection afforded the safety related equipment from fire and reduces the l- chance of a fire related accident.

DC-78-10 - Condensate Polishing Addition 1&2 Portions of this design change involving Major Mechanical Installation and Erection of Pipes and Hangers (78-10K, Unit 1) Lighting Eystem (78-10SA) and, Blowdown Connecting Piping (78-10U) were implemented.

Sumary of Safety Analysis This design change neither constitutes an Unreviewed Safety Question nor requires a change to Technical Specifications.

DC-78-21 - Condenser Rebuild W;d 1 The system was rebuilt to the original design criteria and no affect is imposed on the operation of safety related equipment.

DC-78-37C - Recirculation Spray and Lov Head Safety Injection NPSH Modification 1 This design change included various changes made to assure adequate NPSH performance from the recirculation spray and low head safety injection pumps.

Summary of Safety Analysis This modification is entirely passive requiring no operator action to function. Therefore, the operation of the CS, SI and RS Systems are not affected by this modification.

DC-78-40 - RCS RTD Isolation Valve Removal 1 Tb4s design change removed four manual isolation valves in each RCS RTD bypass loop. The valves which incorporate an adjustable stem packing have proven to be a source of uncontrolled RCS leakage through the stem packing. The valves were originally installed such that the RTD mani-folds could be isolated for maintenance purposes. However, because of the leakage problem associated with the valves, they were more of a maintenance problem than the RTD's.

The four valves in each RCS RTD bypass loop were replaced with straight pipe.

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FACILITY CHANGES THAT DID NOT REQUIRE NRC /PPROVAL (CONTINUED)

_ UNIT Susanary of Safety Analysis The removal of the RTD bypass system isolation valves reduces the amount of maintenance performed on the RCS.

Consequently, the number of personnel exposures to high radiation is also reduced. This further insures the safe and efficient operation of the Unit.

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DC-78-44 - Steam Generator Elowdown Treatment System _ 1 Porti,ons of this design change involving Piping (78-44B),

Electrical (78-44C), Justrumentation (78-44D) and Sample Point Relocation and Instrumentation were implemented.

Summary of Safety Analysis This modification has improved the overall safety reliability and performance the steam generator blowdown system. The design specifications have met or exceeds the specifications of the existing system. The system was designed to meet the NRC guidelines presented in Standard Review Plan (10.4.8) for steam generator blow-down systems. The overall effects of radiological releases to the environment will be significantly reduced by removal of activity in the domineralizers.

DC-78-48 - RTD Relocation and Installation 1 This design change replaces the presently installed RTD's with ones newly calibrated, relocating them and four new ones with computer points.

Summary of Safety Analysis Containment integrity was not affected and no safety implications created with this design change.

. DC-79-14 - Replacement of NAMCO Model D2400X Stem Mounted Limit 1 Switches This design change was initiated to replace the originally installed limit switches with those which have the required documentation as to environnent qualifications.

Summary of Safety Analysis The change out of these limit switches that performs latch-in function from unqualified to environmentally qualified limit switches will not affect station '

operation, but will assure proper operation of the safety !

related equipment.

l DC-79-49 - ILRT Air Pressurization System 1 This modification provides for installing a temporary {

containment air pressurization system for the type A

ILRT.

Sununary of Safety Analysis Since this system was installed only during the test and the unit was shutdown at the time, there was no effect on Teehnical Specifications.

FACILITY CHANGES THAT DID NOT REQUIRE URC APPROVAL (CONTINUED) )

UNIT I l

DC-7 9-50 - Contaiment and Recirculation Spray System Isolation i Valve Replacement This design change removed the welded gate valves and replaced them with flanged butterfly valves.

Suunnary 'of Safety Analysis q

The contaiment spray and recirculation spray capability and operational readiness will not be affected. There are no safety implications created by this design change.

, The effect of this modification on NPSE has been incor-porated into the design basis for the NPSH modification.

i DC-79-60 - Contaiment Pressure Indication 1

This design change installed a redundant pressure

j. transmitter that will be capable of measuring a pressure

range of three times the contaiment design pressure, 0-180 psig. The transmitter will provide a continuous i indication of contaiment pressure in the control room.

i l Summary of Safety Analysis

! This modification does not affect the operation of any 1

safety-related equipment and there are no safety 1 problems caused by this modification.

DC-79-61 - Contaitunent Water Level Indication 1&2 This design change installed redundant wide range and redundant narrow range level transmitters for each unit.

l' The wide range transmitterF are Capable of 30asuring

! from the bottom of the contaiment to an elevation of i

9 ft. in the contaiment which is equivalent to a

~

600,000 gallon capacity. The narrow range transmitters are capable of meas ring' from the batt= ta the ton of the contaiment sump which is a distance of about 22

, inches.

$ Sununary of Safety Analysis This contaiment level modification provides a redundant

safety related indication of contaiment level conditions for both wide and narrow ranges at all times.

l DC-79-67 - Charging Pump CrosWuneet 1&2 This design change provides alternate shutdown capability independent of cables, system or components in the area.

Charging Pump Cross-connect shall consist of two locked closed manual valves installed between the Unit 1 "C" i charging Pump discharge and the Unit 2 "C" charging pump discharge. ,

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FACILITI CHANGES THAT DID liI7I REQUIRE NRC APPROVAL (CONTINUED) tMIT Summary of Safety Analysis

The modification does not affect normal power operations or the mitigation of accidents. The cross-connect will only be used in the event that a fire renders all three charging pumps of one unit unavailable for safe shutdown of that unit. The addition of a cross-connect line between the "C" charging pap discharges of Unit 1 & 2 does not create a safety problem.

DC-79-68 -

Fire Protection Modification Remote Monitoring Panel For 1 Alternative Shutdown Capability (ASC)

This design change provides an alternative reactor shutdown !

capability in the event that a fire disables control room and auxiliary shutdown panel monitoring and control devices.

Summary of Safety Analysis The installation of the remote monitoring panel and its associated field equipment and materials does not create any safety problems.

DC-79-76 -

RCP Oil Collection 1 This design change reduces the potential for fire in the '

containment by using a collection system on the reactor coolant ptsap motors. The system collects and tenporarily stores any lubricating oil which leaks out of the RCP' motor lube oil system.

Sununary of Safety Analysis RCP motor oil collection system consists .of a leak-proof can under oil bearing components to contain laaks in pressurized lines. The modification does not affect the operation of any safety-related equipment; therefore, there are no safety problems caused by this modification.

DC-80-29 -

Reactor Coolant Vent System 1 ;

This design change installed a remotely operated high point vent, to provide venting capability of the primary l coolant system. The need for remote venting capability was identified upon review of the installed systems response to the accident conditions encountered at THI-2.

Summary of Safety Analysis The effect of this design change on station operation has not been clearly defined since specific procedure and functions are not available at this time. This design change will not have any adverse impact on the operation '

of any safety-related equipment. .

DC-00-37A -

Auxiliary Feedwater Redundant Level Indication and Alara 1 In Control Room This design change installed instrument and electrical equipment for the addition of Auxiliary Feedvater redundant level indication and alarm capability in the Control Room.

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i FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL (CONTINUED)

UNIT Summaary of Esfety Analysis This modification doec not affect the Technical Specifications and the FSAR. Installation of the safety grade redundant level indication and alarm capability for the AFW System Condensate Storage Tanks does not create an unreviewed safety question as defined in 10CFR 50.59.

DC-80-42 - Control and Relay Room AC and Charging Pump Service Water 1&2 Piping Modification The design of the system change enhances the reliability of the system, increases the seismic margin of safety, and provides complete documentation to substantiate the engineering analysis. The modification has no adverse effect on the safety-related equipment.

- Redundant Wide Range Pressure Imop 1 DC-80-44 f This design change upgrades the existing reactor coolant system wide range (0-3000 psig) pressure transmitter which is located in the reactor coolant loop C. In addition, a redundant reactor coolant system wide range (0-3000 psig) pressure transmitter was installed into reactor coolant loop B. The transmitters are a safety grade Category I, Rosemount 1153 Series D, Pressure Transmitters.

Summary of Safety Analysis The addition of the reactor coolant system wide range -

pressure loop provides the operator with two independant sources for indication of reactor coolant system wide range pressure measurement making it possible to provide redundant inputs of this measurement to the subcooling monitors.

DC-80-60 - Replacement of Post-Accident Sample System Valves 1 This design change replaced air operated trip valves for the RCS + RER samples with direct acting solenoid valves.

The function of the valves have not changed. An additional direct acting solenoid va' r2 was added to the RER sample line inside contairmnent .no functions as a contairnment isolation valve.

1 Summary of Safety Analysis This modification does not affect normal station operation or the operation of any safety related equipment. Replacement of these containment isolation valves improves the capability of the valves to open af ter an accident to obtain required j samples.

DC-80-73 - Relocation of Chemical and Volume Control Pipina 1&2 This design change relocated line 3" DG-41-152 and 3" C11-99-152. Also included was the installation of an analytical anchor on line 2" DG-242-152. This allows analyses of these lines as required by letc Bulletir.79-14B.

FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL p (CONTINUED)

UNIT Sumary of Safety Analysis This modification does act affect the 'rechnical

. Specifications, requires no change to FSAE and, does not create an unreviewed safety questien.

DC-80-75 - AFW System Cavitating Venturi Addition 1 This design change added cavitating venturis in each ,

of the 3 in. AFWS lines to each steam generator. The i venturis will restrict the flow of feedwater to the affected steam generator (i.e., loop with a MSLB or i MFLB inside contaiment) permitting greater flow to the intact loops.

i Sumary of Safety Analysis !

This modification does not directly change the operation ;

of the Auxiliary feedwater system although operating procedures will br.se to be modified to reflect the new auxiliary feedwater flow rate.

l DC-80-78 - Post Accident Sample System Contaiment Return Line 1&2 l

This design change provides a drain path to contaiment sump for post accident sample system drains. Pumping to the contaiment sump provides an alternative to pumping ,

j radioactive liquid to the high and low level waste drain '

tanks and therefore reduce the activity levels inside the 4

auxiliary building.

Sumary of Safety Analysis This modification has no effect on normal station operation or operation of any existing safety related equipment.

DC-80-79 - Removal Pressurizer Cubicle Wall C-47-4-6 1 l This design change replaced existing concrete block portion of the biological shield wall around the pressurizer with precast concrete panels to comply with IE Bulletin 80-11.

Sumary of Safety Analysis The completed modification will have no effect on the

) operation of safety-related equipment.

4 DC-80-85 - Turbine Driven Auxiliary Feedwater Pump Automatic Control 1 !

This design change fulfilled the requirement that at least i one AFW system pump and its associated flow path and essential instrumentation automatically initiate AFW i System flow and is capable of being operated independently of any AC power Source for at least two hours.

Summary of Safety Analysis .

This modification allows operation of the steam driven AFW pump independent of any AC power source. Replacement of a

, motor operated valve with an air-operated valve will not affect the safety function of this system. .No adverse safety implications will result from this design change.

FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL (CONTINUED)

UNIT

+ DC-80 - Modification of Masonary Wall 1 a

Portion of this design che/oge involving Hasonary Wall IC-47-4-6 (8086H), replaced existing block wall with precast. concrete panels to comply with IE Bulletin 80-11.

Stammary of Safety Analysis The completed modification will have no effect on the e

operation of safety related equipment.

4 DC-80-96

- Reactor Vcssel Level 1 f This design change installed redundant microprocessor-i based monitoring systems to monitor Reactor Vessel Level.

l The system utilizes differential pressure measuring transmitters which measure water level and relative void I content of the circulating primary coolant system fluid and thus provides direct readings of the reactor vessel level.

t Sucuna ry of Safety Analysis

} This modification provides redundant safety-related indication of reactor vessel water 3evel thus providing

. an indication of inadequate core cooling.

DC-80-101 - Control and Reley Room Air Treatment Modification 162

! This design change provides a means for injecting and mixing of tracer gasas and aerosoles with the intake air for periodic testing of the control and relay room j air treatment (pressurization) system.

Summary of Safety Analysis i The operati.'n of the station and of the control and relay L room air treatment system remains unchanged.

! DC-81-06 - Electrical Penetration Replacement 1 l This design chang replaces Amphenol spare contaissnent

. penetrations, with Conax penetrations. This provides the new penetrations with pigtails on the feed-throughs long i enough to permit splicing to the corresponding cable in the cable trays. Amphenol terminal conaectors soldering of terminals delayed installation in the field and were difficult to procure with a 26 week delivery.

Summary of Safety Analysis This modification has no effect of the proposed change on station operation and the operation of safety-telated equipment.

- Reroute of Pressurizer Relief Tank Vent Sample Line 1 DC-81-24 This design change rerouted sample line 3/8"-SS-99-ICN9 on the pressuriser vent free penetration #55 to penetration #57.

Stannary of Safety Analysis This modification does not create any safety problems.

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TESTS AND EXPERIMENTS REQUIRING NRC APPROVAL NONE DURING THIS REPORTING PERIOD.

TEST AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL 3pecial Test No,. Unit Title Completed ST-127 1&2 Hechanical Snubber Stroke Test 07-16-81 ST-117 1 Pressurizer Spray Flow Verification 07-16-81 ST-119 1 Reactor Coolant Systent Cooldown 07-23-81 ST-104 1&2 Class IE Electrical Equipment Environmental 06-25-81 Qualification Verification Inspection Program ST-113 1 Atmospheric Steam Dump Valves 07-03-81 ST-116 1 Pressurizer Relief Tank 07-01-81 ST-131 1 Subsequent Testing for Periodic Test 8.5A 07-01-81 e

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CHDf1STRY REPORT July , 19 81 T. S. 6. 6. 3.d PRIMARY C00LAh7 UNIT NO. 1 UNIT NO. 2 ANALYSIS MAXIMUM MINIMUM AVERAGE MAXIMUM MINIMUM AVERAGE Cross Radioact., UCi/ml 8.73E-1 5.06E-3 1.99E-1 3.36E-1 1.22E-1 2.51E-1 Suspended Solids, ppe 0.1 0.1 0.1 0.2 0.1 0.1 4

Gross Tritium UCi/mi 1.24E-1 7.89E-3 6.56E-2 1.19E-1 3.50E-2 7.78E-2 (cj (c) (n) (h Iodine-131, WCi/a1 9.79E-2 4.90E-4 3.52E-2 1.74E-1 1.91E-3 1.70E-2 I-131/I-133 .5587 .2434 .3817 1.3445 .7940 1.1692 II I

! Hydrogen, cc/kg 49.7

25.6 34.9 58.2 32.9 44.2 Lithium, ppm 2.14 .00 ID) .58 1.64 .31 I) 1.07 Boron-10, ppm + 606 181 279 127 48 58 -

Oxyg u-16, ppm 2.3 (E) .000 .'371 ( ) .000 .000 .000 Chloride, PP" <.05 <.05 <.05 <.05 <.05 <.05 pH f 25*C 6.75 4.50 5.69 7.33 6.49 7.02

+ Boron-10 = Total Boron x 0.196 MON-RADICAGIVE CHEMICAL

. RELEASES, FOUNDS (,y*) . -

T. S . 4.13. A. 6,.

Phosphate Doron 1421 Sulfate Chromate D.12 50% NaCH Chlorine- l Remarks: (A) Initial criticality Efter SGRP on 7-6-b1 at 1712 hrs.

(B) 7 trips on 7-9 at 2054, 7-11 at 1815, and 7-16 at 1952 (C) High value due to suspected fuel failure (D) Placed mixed bed (LiOH) in service 7-21-81 (E) High values ref3ect pre-start-up conditions

l. (C) Rx trips on 7-17 at 1122, 7-18 at 0159, and 7-18 at 0325 (H) High values due to Rx trip (I) Bigh values due to suspected pin-hole leaks in fuel (8** N (J) These voltanes of chemicals are be14eved to have no major adverse environmental effects.

(K) LiOH added on 7-2 and 7-22

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DESCRIPTION OF ALL INSTANCES VEERE THERMAL DISCHARGE LIMITS WERE EXCEEDED Month / Year JULY. 1981 Due to the 'pairment of the circulating water system on the following days, the thermal discharge limits were exceeded as noted.

July 14. 1981 Exceeded 15 F AT across station

July 15, 1981 Excesdad 15 F AT across station

July 16,1981 heeeded 17.5 F AT across station

July 18, '.981 Exceeded 15*F AT across station

July 19, 1981 Exceeded 15 F AT across station

July 20, 1981 Exceeded 15 F AT across ststion*

July 21, 1981 Exceeded 15*F AT across station

July 22, 1981 Exceeded 15 F AT across station

July 23, 1981 Exceeded 15 F AT across station

July 24, 1981 Exceeded 15 F AT across station

July 25, 1981 Exceeded 15 F AT across station

July 26, 1981 Exceeded 15 F AT across station

July 27, 1981 Exceeded 15 F AT across station

July 28, 1981 Escarded 15 F AT across station * ~

July 29, 1981 Exceeded 15 F AT across station

July 30, 1981 Exceeded 15 F AT across station

July 31, 1981 Exceeded 15 F AT across station *

Indicates dates where station AT was 15.0 F or less across the statian for sometime during the day.

The AT excursions were allowable undet Technical Specification l 4.14.B.2. There were no reported instances of adverse envirotunental impact.

The temperature charge at the station discharge exceeded 3 F per hour on: July 9,1981, due to a reactor trip on Unit 1; on July 16, 1981, due to e reactor trip on Unit 1; and on July 17, 1981, due to a reactor trip on Unit 2.

These events were allowable in accordance with Technical Specification 4.14. There were no reported instances of adverce environmental impact.

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. TUEL EAIIDLING JULY, 1981 HONE DURING Tills REPORTING PERIOD.

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PROCEDURE REVISIONS THAT C3ANGED TE l OPERATING MODE DESCRIBED IN THE FSAF, JULY 1981 1 e

PROCEDURE NO. UNIT TITLE CHANGE PT 18.9 1 Boric Acid Pump Change acceptance Operability and criteria and other Performance information due to installation of new 4

pump.

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_ DESCRIPTION OF, PERIODIC TESTS WHICH UERE 100T CORPLETED ITIDiIN THE TIME LIMITS SPECIFIED IN TECHNICRI. SPLClFICATIONS JULY. 1981 NONE DURING THIS REPORTING PERIOD.

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_30 INSERVICE INSPECTION JULY, 1981

_ Unit One The Westinghouse program of inservice inspection has been completed with no major reportable indications being noted.

A system hydrostatic test was performed on all of the ASME Class I systans prior to unit start up. Only one indics: ion was noted on a small sample valve off C-loop cold leg. This was a boruiet to body leak and was repaired and reinspected with acceptable results.

The preservice inspection program of the unit one steam

. generator replacement has been completed with no indications being recorded.

The ASFE Class III Support and Bangar Inspection is complete.

Minor indications were recorded on this inspection and repairs are being effected, as station scheduling permits, with approved maintenance procedures.

Unit Two No inservice inspection vork this month.

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NONE DURING THIS REPORTDIG PERIOD.

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REALTH PHYSICS JULY 1981 l There was no single release of radioactivity or radiation exposure specifically associated with an outage that accounted for more than 10% of the allowable annual values in 10CFR20.

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PROCEDURE DEVIATIONS' REVIEWED BY STATION ?UCLEAR SAFETY AND OPERATING COmi1TIEE AFTER TIl!E LIMITS SPECIFIED IN TECENICAL SPECIFICATIONS JULY 1981 DATE SNSOC PROC. NO. UNIT TITLE DATE DEVIATED REVIEWED OP 8A 1 CVCS System - Valve Checkoff 06-24-81 07-13-81 OP-48.5 1 Station Vacuum Priming System - 07-02-81 07-23-81 Valve Checkoff OP-51A 1 Chilled Water System - 07-03-81 07-23-81 Valve Chekcoff ST-127 1, 2 Mechanical' Snubber 07-03-81 07-23-81 Stroke Test ST-130 1 Recirculation Spray 06-28-81 07-23-81 Heat Exchanger Service Water Foreign Material Evaluation PT-11 1 Reactor Coolant Integrity 07-05-81 07-23-81 Test Following Opening PT-16.3 1 Reactor Containment Building 06-25-81 07-13-81 Integrated Leak Rate Test PT-28.11 ' 1 Nuclear Design Check Tests 07-07-81 07-23-81 g a & 'SvV

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