ML18152B085

Monthly Operating Repts 88-07 for Jul 1988 for Surry Power Station Units 1 & 2.W/880815 Ltr

ML18152B085
Person / Time
Site:	Surry
Issue date:	07/31/1988
From:	Cruden D, Warren L VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
88-539, NUDOCS 8808240092
Download: ML18152B085 (28)

Text

.. e e POW 34-04

$()-Uc, VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION MONTHLY OPERATING REPORT REPORT 88-07 APPROVED:

l. 1 SECTION PAGE Operating Data Report - Unit No. 1 1 Operating Data Report - Unit No. 2 2 Unit Shutdowns and Power Reductions - Unit No. 1 3 Unit Shutdowns and Power Reductions - Unit No. 2 4 Average Daily Unit Power Level - Unit No. 1 5 Average Daily Unit Power Level - Unit No. 2 6 Summary of Operating Experience - Unit No. 1 7 Summary of Operating Experience - Unit No. 2 8 Facility Changes Requiring NRC Approval 10 Facility Changes That Did Not Require NRC Approval 11 Procedure or Method of Operation Changes that Did Not Require NRC Approval 18 Procedure or Method of Operation that Did Require NRC Approval 19 Tests and Experiments Requiring NRC Approval 20 Tests and Experiments That Did Not Require NRC Approval 21 Chemistry Report 22 Fuel Handling - Unit No. 1 23 Fuel Handling - Unit No. 2 24 Description of Periodic Test Which Were Not Completed Within the Time Limits Specified in Technical Specifications 25

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.. e OPERilING DAU REPORr DOCICEX No. -5~0.... 8....0____

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COMPLEXED BY L. A. Warren m.EPHONB 804-357-3184 OPERATING S'.tAXUS

1. Unit Name: Surry Unit # 1 Notes

2. Reporting Period: Tuly OI thru 31, 19 88

3. Licensed l'hermal Power (MWt): __,2=-4=-4:..:1:.------

4. Nameplate Rating (Gross MWe): _8_4_7_._5_ _ _ _ __

S. Design Electrical Rating (Net MWe): 788

6. Maximum Dependable Capacity (Gross MWe): _8~2::;..;0;;,____

7. Maximum Dependable Capacity (Net MWe): 781 *

8. If Changes Occur in Capacity Ratings (Items Number 3 1'hrough 7) Since Last Report, Give Reasons:

9. Power Level Xo Which Restricted, If Any (Net MWe):

10. Reasons For Restrictions, If Any:

l'his Month Yr.*to*Date Cumulative

11. _Hours In Reporting Period z~~-a Slll~Q 136823"0

12. Number of Hours Reactor Was Critical ~15.Z 2728~3 87036;0

13. Reactor Reserve Shutdown Hours a~ a o;o 3774~5

14. Hours Generator On-Line aa 1~ a 2628;6 85599~8

15. Unit Reserve Shutdown Roura Q~O o-o 3736:2

• 16. Gross l'hermal Energy Generated (HWB) Z2U09.S 6Q44219;7 19876444L 1

17. Gross Electrical Energy Generated (HWB) 23~1 lQ. O 202879~5 64403968~0

18. Net Electrical Energy Generated (HWB) 221010.0 1926514;0 61081890~0

19. Unit Service Factor ~~.5% 51 ;4% . 62.6%

20. Unit Available Factor ~~.5% 51 ~4% 65~3%

21. Unit Capacity Factor (Using MDC Net) JB.0% 48;3% 57. 7.%

22. Unit Capacity Factor (Using DER. Net) JZ.2% 47;8% 55-. 7%

23. Unit Forced Rate 0~0% 2~9% 17.2%

24. Shutdowns Scheduled Over Next 6 Months (:type, Date, and Duration of Eacbh

25. If Shut Down At End Of Report Period Estimated Date of Startups

26. Units In Test Status (Prior to Commercial Operation)s Forecast Achieved INIXIAL CR.InCALiff INITIAL ELECXRICiff COHHERCIAL OPERAnOH (9/77)

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. COMPLE'IED BY . L. A. Warren TELEPHONE 804-357-3184 OPERATING STAXUS

1. Unit Name: Surry Unit # 2 Notes

2. Reporting Period: July 01 thru 31, 1988

3. Licensed Thermal Power (MWt): -:2=..4.:..4:.:1:..-_____

4. Nameplate Rating (Gross MWe): ---'8:;,.4~7_,._5_______

S. Design Electrical Rating (Net HWe): 788

6. Maximum Dependable Capacity (Gross HWe): _8._2__0____

7. Maximum Dependable Capacity (Net HWe): _..;7..;8_,l"----

8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report, Give Reasons:

9. Power Level ro Which.Restricted, If Any (Net HWe):

10. Reasons For Restrictions, If Any:

This Month Yr.-to-Date Cumulative

11. Hours In Reporting Period 744.0 5111.0 133703.0

12. Number of Hours Reactor Was Critical 744.0 4066.3 88732.3

13. Reactor Reserve Shutdown Hours 0 0 0 0 328.1 lit. Hours Generator On-Line 744.0 403217 87331.0

15. Unit Reserve Shutdown Bour* o.o 0 0 0 0

• 16. Cross Thermal Energy Generated (MWB) 1771099.7 9700719 1 l 20487108118

17. Gross Electrical Energy Generated (HWB) 56801Q 1 Q 320554510 66578369.0

18. Net Electrical Energy Generated (MWB) ~3953Q 1 0 3050089 10 63126566 10

19. Unit Service Factor 1009£1 28,2% 65,3%

20. Unit Available Factor 100% 78,9% 65.3%

21. Unit Capacity Factor (Using MDC Net) 92,9% 26.~% 60,6%

22. Unit Capacity Factor (Using DER Net) 92% 75 7% 59. 9'°.t6

23. Unit Forced Rate 0,0% 21.1% 14 7%

24. Shutdowns Scheduled Over Next 6 Month* (Type, Date, and Duration of !acb)1 Refueling:, 09/09/88, 68 days

25. If Shut Down At End Of Report Period Estimated Date of Startups

26. Units In rest Status (Prior to Coanercial Operation)1 Forecaat Achieved  :..

INirIAL CRITICALiff INirIAL ELEC'lRICiff COHHERCIAL OPERATION (9/77)

UNI'.[ SHU'lDOWNS AND POWER REDUC'.rlONS DOCKE'r No. _s__o__PAGE

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COHPLETED BY L*. A. Warren REPOR'.1: K>N'IH JULY 1988 n:LEPHONE  :::00:4:-:3:s1:-:3:1:04:::::::

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1 2 3 4 F: Forced Reason: Method: Exhibit *c - Instructions S: Scheduled A - Equipment Failure (Explain) 1 - Mamial for Preparation of Data B - Maintenance or Test 2 - Manual Scram. Entry Sheets for Licensee C - Refueling 3 - Automatic Scram. Event Report (LER) File D - Regulatory Restriction 4 - Other (Explain) (NUREG 0161)

E - Operator Training & License Examination 5

F - Administrative Exhibit 1 - Same Source G - Operational Error (Explain)

(9/77) H - Other (Explain)

r'Ali.t:; q UNI'I SHU'IDOWNS AND POWER REDUCTIONS DOClCE'.J: NO. 50-281 UNIT NAME Surry Unit # 2 DATE 08/03/88 r COHPLEnl> BY L.*A. Warren REPORT K>N'IH JULY 1988 TELEPHONE 804-357-3184 C

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C 88-08 07/11/88 s o.o A 1 Unit reduced to 75% power, 560 MWs to clean-water boxes.

88:...og 07 /22/88 s o.o A 1 Unit reduced to 65% power, 490 MWs for PT-29.1 and to clean waterboxes.

88-10 07/23/88 s o.o A 1 Unit reduced to 63% power, 500 MWs..,to work 1A1 MFP and to clean waterboxes.

I 88-11 07 /30/88 s o.o A 1 Unit reduced to 77% power, 570 MWs to allow removal of 13B1 feedwater heater.

e l 2 3 F: Forced Reason: Method: Exhibit G - InstTUCtions S: Scheduled A - Equipment Failure (Explain) l - Manual for Preparation of Data B - Maintenance or 'Iest 2 - Manual Scram. Entry Sheets for Licensee C - Refueling 3 - Automatic Scram. Event Report (LER) File D - Regulatory Restriction 4 - Other (Explain) (NUREG 0161)

E - Operator Training & License Examination 5

F - Administrative Exhibit 1 - Same Source G - Operational Error (Explain)

(9/77) H - Other (Explain)

-- AVERAGE DAILY lJNIT POWER LEVEL PAGE 5 DOCKET NO. 50-280 lJNIT Sura Unit# 1 DilE 8£10l88 COMPLETED BY L.A. Warren TELEPHONE 804-357-3184 MON'XH Illl,Y 1988 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 0 17 0 2 0 18 177

'3 0 19 309 4 0 20 517 5 0 21 690 6 0 22 734 7 0 23 711 8 0 24 749 9 0 25 756 10 0 26 767 11 0 27 769 12 0 28 770 13 0 29 767 14 0 '30 763 l 'i 0 '31 762 16 0 INSTRUCTIONS On this format, list the average daily unit power level in MWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.

_(9/77)

e PAGE 6 AVERAGE DAILY UNIT POWER LEVEL DOCICET NO. 50-281 UNIT Sura Unit# z DAl'E 8£10l88 COHPLEl'ED BY L.A. Warren TELEPHONE 804-357-3184 MONl'H !ULY 1988 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER. LEVEL (MWe-Net) (MWe-Net) 1 768 17 731 2 771 18 724

'3 770 19 722 4 766 20 723 5 763 21 719 6 753 22 680 7 764 23 590 8 752 24 686 9 748 25 738 10 751 26 734 11 703 27 732 12 752 28 732 29

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l '3 751 725 14 747 '30 597 15 735 31 622 16 728 INSTRUCTIONS On this format, list the average daily unit power level in MWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.

(9/77)

e e PAGE 7

SUMMARY

OF OPERATING EXPERIENCE MONTH/YEAR JULY 1988 Listed below in chronological sequence by unit is a summary of operating experiences for this month which required load reductions or resulted in significant non-load related incidents.

UNIT ONE 07-01-88 0000 This reporting period begins with the unit at CSD on RHR.

07-09-88 1548 RCS is> 200°F.

07-10-88 0405 RCS > 350°F /450 PSIG.

1035 Holding RCS temperature, evaluating leakage thru MOV-CS-101 B.

07-11-88 0455 Commenced RCS cooldown, declared NOUE due to LCO condition exceeded, excessive leakage by MOV-CS-101 B.

0912 RCS< 350°F.

1300 RCS < 200°F, terminated NOUE.

07-12-88 1123 RCS> 200°F.

1636 RCS > 350°F /450 PSIG.

07-13-88 0205 RCS at HSD.

07-14-88 0825 Reactor critical.

1117 Commenced physics testing.

1937 Manually tripped reactor due to over boration of RCS.

07-15-88 0329 Reactor critical.

1957 Completed physics testing.

07-18-88 0503 Unit on line.

0602 Holding power at 30%, 165 MWs.

1310 Commenced power increase at 3%/hr.

07-20-88 0100 Holding power at 70%, 520 MWs.

2125 Commenced power increase at 3%/hr.

L, e PAGE 8

SUMMARY

OF OPERATING EXPERIENCE MONTH/YEAR JULY 1988 Listed below in chronological sequence by unit is a summary of operating experiences for this month which required load reductions or resulted in significant non-load related incidents.

UNIT ONE (cont'd) 07-21-88 0645 Holding power at 97%, 785 MWs due to CP-A P.

07-22-88 0900 Unit at 100%, 810 MWs.

07-31-88 2400 This reporting period ends with the unit at 100% power, 800 MWs.

UNIT TWO 07-01-88 0000 This reporting period begins with the Unit at 100% power, 805 MWs.

07-09-88 2122 Commenced power reduction at 150 MW /hr, to allow cleaning of water boxes.

2150 Holding power at 90%, 700 MWs.

07-10-88 0128 Commenced power increase at 150 MW /hr.

0450 Unit at 100% power, 800 MWs.

07-11-88 0236 Commenced power reduction at 150 MW /hr to allow cleaning water boxes.

0514 Holding power at 75%, 560 MWs.

0726 Commenced power increase at 150 MW /hr.

1215 Unit at 100% power, 800 MWs.

07-22-88 1825 Commenced power reduction at 150 MW /hr for PT-29.1.

2256 Holding power at 65 %, 490 MWs.

07-23-88 0618 Commenced power increase at 150 MW /hr.

0938 Unit at 100% power, 780 MWs.

1818 Commenced power reduction at 150 MW/hr for repair of 'A' MFP.

2043 Holding power at 63%, 500 MWs.

e e PAGE 9

SUMMARY

OF OPERATING EXPERIENCE MONTH/YEAR JULY 1988 Listed below in chronological sequence by unit is a summary of operating experiences for this month which required load reductions or resulted in significant non-load related incidents.

UNIT TWO (cont'd) 07-24-88 0245 Commenced power increase at 150 MW /hr.

0706 Unit at 100% power, 800 MWs.

07-30-88 0022 Commenced power reduction at 150 MW /hr to allow removal of

'3B' feedwater heater and to clean waterboxes.

0220 Holding power at 81%, 645 MWs.

0709 Reducing power further to remove 'C' waterbox from service.

0845 Holding power at 77%, 570 MWs.

1058 Commenced power increase at 150 MW /hr **

1145 Holding power at 89%, 655 MWs.

07-31-88 2400 This reporting period ends with the unit at 89% power, 655 MWs due to the '3B' feedwater heater being out of service.

t. PAGE 10 F ACIIJTY CHANGES REQUIRING NRC APPROVAL MONTH/YEAR JULY 1988 NONE DURING THIS PERIOD

FACIIJTY CHANGES THAT DID NOT REQUIRE NRC APPROVAL PAGE 11 MONTH/YEAR JULY 1988 DC 85-10 INADEQUATE CORE COOIJNG SYSTEM UPGRADE UNIT 1 The non-redundant non-class IE incore (core exit) thermocouple system was upgraded to a redundant class IE core exit thermocouple system.

In addition, the existing RVLIS and CCM systems will be integrated with the upgraded CET system into a common system called the ICC system.

SUMMARY

OF SAFETY ANALYSIS This modification upgrades the incore thermocouple and core cooling monitoring systems to class IE systems. It improved and increases safety-related information displays in the control room to inform the operators of an approach to inadequate core cooling. The system is seismically installed.

DC 85-29 SAFETY INJECTION LEAKAGE MONITORING UNIT 1 This design change installed a leakage monitoring tubing between the two check valves in the safety injection piping which connects to each of the Reactor Coolant System cold legs.

SUMMARY

OF SAFETY ANALYSIS The tubing was designed and installed in .accordance with the appropriate codes and standards, and is seismically supported. The modification improves the reliability of leakage monitoring and failure of the 3/4" tubing would result in an insignificant amount of leakage and is within the previous analysis.

DC 86-09 ALTERNATE SERVICE WATER FLOW PATHS UNITS 1 & 2 This design change had two parts.

A temporary flow path, thirty inch pipe line, was installed to provide service water to two of the component cooling water heat exchangers during the period that some of the ninety-six service water valves were being replaced.

A permanent eight inch pipe line was installed to provide a second source of charging pump service water and will allow for the removal of certain f arty-two inch and thirty inch valves for repair or replacement.

e e PAGE 12 FACIIJTY CHANGES THAT DID NOT REQUIRE NRG APPROVAL MONTH/YEAR _ JULY 1988 _

DC 86-09 ALTERNATE SERVICE WATER FLOW PATHS UNITS l & 2

SUMMARY

OF SAFETY ANALYSIS Failure of the temporary line would have resulted in a loss of service water to the CCW heat exchangers and a loss of the component cooling which has been evaluated in the UFSAR. The reactor unit would have been brought to hot shutdown using existing station abnormal operating procedures. The piping was constructed of safety related material and installed in accordance with the piping specification. The system is a moderate energy system, less that 275 PSIG and 200°F, and the postulated failure is limited to throughwall cracks. The eight inch alternate charging pump service water supply was designed and installed consistent with the existing design basis of the charging pump service water system. The piping is seismically analyzed and installed.

DC 86-10 SERVICE WATER AND CIRCULATING WATER BUTTERFLY VALVE REPLACEMENT UNIT I This design change replaced 96 11 , 42 11 , 36" and 30" circulating and service water butterfly valves and expansion joints. The replacement valves will be ductile iron with the wetted portions coated with a liquid epoxy.

The following valves and expansion joints were replaced:

MOV-CW-106A, B, C, D CW inlets MOV-SW-102A, B SW to CC HX l-SW-25, 29, 33, 37 CC HX SW inlets The following valves will be replaced at a later time:

MOV-CW-lOOA, B, C, D CW outlets MOV-SW-101 A, B SW to BC HX l-SW-27, 31, 35, 39 CC HX SW outlets

SUMMARY

OF SAFETY ANALYSIS The replacement valves and expansion joints are one for one replacement of existing equipment and are designed to specification requirements which meets or exceeds the original specifications. The equipment will operate and function identical to existing equipment.

The new valves and expansion joints do not affect or change the basis for any Technical Specification. The replacement valves are seismically qualified.

l" e e PAGE 13 FACII.JTY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR JULY 1988 DC 86-13 CONTAINMENT SPRAY FLOW AND PRESSURIZER HEATER STATUS UNIT 1 To comply with R.G. 1.97, a containment spray flow and a pressurizer heater status monitoring system was added by this design change. An annubar flow element was added to sense containment spray flow and provided an electrical flow signal to the multiplexer system for remote display in the control room. Also, this design change installed new watt transducers to monitor pressurizer heater status for pressurizer heaters H and J. This will provide an independent output to the multiplexer system for remote display in the control room.

SUMMARY

OF SAFETY ANALYSIS*

This modification provides instrumentation to allow the operator to monitor the magnitude of containment spray flow and the power consumption of the backup pressurizer heaters. The installation of the instrumentation does not impact the operation of existing safety related equipment or systems. The instrumentation is environmentally and seismically qualified.

DC 87-01 ACCUMULATOR TANKS PRESSURE TRANSMITTER UPGRADE UNIT 1 This design change installed six new qualified accumulator tank pressure transmitters above the flood level to replace six nonqualified pressure transmitters and mounted below the flood level in the reactor containment.

SUMMARY

OF SAFETY ANALYSIS This modification provides qualified instrumentation to allow the operator to monitor the pressure of the safety injection accumulator tanks. The replacement of the instrumentation (transmitters and power supplies) does not impact the operation of safety related equipment or systems. The new instrumentation is environmentally and seismically qualified. Safety limits, as defined in the Technical Specifications, are not changed.

e PAGE 14 FACIJ.JTY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR JULY 1988 DC 87-10

• REPLACEMENT OF TlllRD AND FOURTH POINT FEEDWATER HEATERS UNIT I The third and fourth point feedwater

• heaters were replaced because of the number of tubes which were plugged and to remove copper alloys from the feedwater system. The new heaters will have tubes fabricated of 304 stainless steel.

SUMMARY

OF SAFETY ANALYSIS The new heaters do not change the operation of equipment important to safety or their ability to perform their safety functions. The replacement heaters are essentially one for one replacements of an enhanced design. The modification is *consistent with the systems design bases and was installed using existing design criteria. The feedWater heaters do not perform any safety functions.

DC 87-22 RECIRCULATION SPRAY COOLER REPLACEMENT UNIT I The four recirculation

• spray coolers were replaced with coolers designed and fabricated in accordance with the requirements of the 1986 ASME vm, Division 1 Code which meets or exceeds the requirements of the 1968 ASME III Class C and ASME vm Codes used to fabricate the original coolers.

SUMMARY

OF SAFETY ANALYSIS The recirculation spray system with the new coolers meet or exceed all original design requirements. The modification was essentially a one for one changeout of the coolers. The change of materials in the coolers did not change the basic design. The coolers are independently supported and seismically qualified.

DC 88-01 INSIDE RECIRCULATION SPRAY PUMP FULL FLOW TEST IJNE UNIT I This design change provided recirculation test piping and a temporary sump to allow full flow testing of the IRS pumps. The piping was designed to ANSI B3 l. l code requirements (with N-Code Cases) with flow and pressure instrumentation to meet ASME XI requirements.

SUMMARY

OF SAFETY ANALYSIS (cont'd)

e e PAGE 15 FACIIJTY CHANGES THAT DID NOT REQUIRE NRG APPROVAL MONTH/YEAR JULY 1988 DC 88-01 INSIDE RECIRCULATION SPRAY PUMP FULL FLOW TEST UNE

SUMMARY

OF SAFETY ANALYSIS The recirculation test piping and temporary sump will be connected/installed during unit shutdown conditions only. This modification will not affect the operation of the safety-related IRS pumps. It will not affect the design basis of the IRS system to mitigate LOCA or SLB accidents. The modified IRS pump discharge piping will meet the original station design criteria. This modification increases the reliability of the IRS pumps since the periodic testing will be used to access the operational readiness of the pumps.

The only permanent change to the IRS system piping installed under this design change is the addition of a flanged spool piece in the pump discharge line. This piping section has been designed to meet safety related and seismic requirements. There are no active components added to the IRS system. There are no valves installed in the pump discharge piping which could be misaligned (closed) and, therefore, prevent initiation of spray flow.

TM Sl-88-108 TEMPORARY MODIFICATION 07/02/88 This temporary modification removed check valve internals from 1-BD-558, installed an adapter and jumper hose to condensate polishing waste neutralization or low conductivity sumps.

The steam generator will be at atmospheric pressure during implementation of this jumper. Therefore, the probability of a steam generator tube rupture is not credible. If leakage on the jumper should occur, the release can be terminated immediately by securing the recirculation and transfer pump. Therefore, no unreviewed safety question exists.

TM SI-88-113 TEMPORARY MODIFICATION 07-07-88 A radiator hose will be connected to the reactor vessel head vent and direct leakage to the transfer canal to prevent leakage of reactor i

coolant system on the reactor head and to CRDM shroud cooling fan coolers.

The installation of a radiator hose to direct leakage to the transfer canal from the reactor head vent will not constitute an unreviewed safety question because both trains of the head vent will remain operable with this jumper installed. If the system is to be activated, the jumper will blow free from the reactor vessel and the system will remain fully operable.

PAGE 16 FACIIJT-HANGES THAT DID NOT REQUIR-RC APPROVAL MONTH/YEAR JULY 1988 EWR 88-177 ENGINEERING WORK REQUEST 07/07/88 This engineering work request installed a manual isolation valve between the cold leg sampling valves and the containment isolation trip valve to facilitate leakage testing.

The change does not constitute an unreviewed safety question. The installation of the valve is for testing purposes only and will not affect normal operation or safety function of the system.

EWR 88-178 ENGINEERING WORK REQUEST 07/07/88 This engineering work request installed a manual isolation valve between the hot leg sampling valves and the containment isolation trip

• valves to facilitate leakage testing.

This change does not constitute an unreviewed safety question. The installation of the valve is for testing purposes only and will not affect*

normal operation or safety function of the system.

EWR 88-182 ENGINEERING WORK REQUEST 07/07/88 An evaluation was performed of sensitized stainless steel piping not listed in the Technical Specifications as requiring flushing. It was determined that flushing of these lines is being performed adequately by existing periodic tests, although this is not necessary as long as R WST chemistry is maintained within its acceptable chemistry specification for chlorides and flourides.

A technical specification change is being prepared to include the identified sensitized stainless steel piping in the Technical Specifications and to delete the flushing requirement unless required by RWST chemistry.

EWR 88-195 ENGINEERING WORK REQUEST 07/07/88 The addition of strainers, as described in the EWR, is required to prevent clogging of the cavitating venturi due to small parts from damaged components.

The addition of the strainers does not represent an unreviewed safety question because they do not change the flow rate of AFW system.

TM Sl-88-114 TEMPORARY MODIFICATION 07/10/88 The signal from containment to the "safety injection - valve out of position" annunciator will be blocked by lifting the lead. Verification that the accumulator isolation valves are open will be conducted following any containment entry.

Lifting the lead from containment to the safety injection valve out of position alarm in the control room will not constitute an unreviewed safety question because prior to lifting the lead and following each containment entry, verification will be made that MOV-186SA, B, and C are open. This temporary modification will not affect the operation of these valves or the accumulators.

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e e PAGE 17 F ACIIJTY CHANGES THAT DID NOT REQUIRE NRC APPROVAL I

MONTH/YEAR JULY 1988 TM Sl-88-119/ TEMPORARY MODIFICATION 07/12/88 TM-S2-88-50 These temporary modifications will maintain the air intake louvers open to the emergency diesel generator room by isolating service air to the louver positioner.

Since the diesel room air intake louvers are open when the diesel is running, the change does not affect operation of the diesel. Since carbon dioxide is heavier than air, maintaining louvers open will enhance carbon dioxide displacement of oxygen in the event of a fire and carbon dioxide initiation. Therefore, the change does not constitute an unreviewed safety question.

TM Sl-88-123 TEMPORARY MODIFICATION 07-21-88 This modification installed. a fitting and flexible hose for ion exchange resin.

This modification does not constitute an unreviewed safety question because its probability of failure is small (due to appropriate pressure rating of components and pre-use testing of installation) and consequences of failure are previously evaluated.

2-MOP-48-1 MAINTENANCE OPERATING PROCEDURES 07-27-88 thru MOP-48-8 These procedures maintained the recirculation spray (RS) heat exchanger (HX) service water (SW) discharge valves 2-SW-MOV-205A, B, C, and D closed and under administrative control to prevent service water ingress to the RS Hx from the discharge tunnel to prevent biofouling the RS HXs.

Since the valves (2-SW-MOV-205A, B, C and D) will remain operable and will be opened prior to auto or manual start of the RS pumps, the change will not affect the operation of the RS system. Therefore, no unreviewed safety question is created.

TM S2-88-54 TEMPORARY MODIFICATION 07-28-88 This temporary modification will lift the lead from SV2XA to maintain the relay de-energized (tripped) making up the trip signal from this relay.

An unreviewed safety question does not exist because the turbine trip signals to trip the reactor on stop valve closure is still functional.

Logic has been reduced to 3/3 valves closed.

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e e PAGE 18 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRG APPROVAL MONTH/YEAR JULY 1988 EWR 88-295 ENGINEERING WORK REQUEST 07/13/88 This engineering work request minimized the precipitation of nickel ferrite corrosion products, reduced core deposits and enhanced ALARA concerns.

The elevated lithium *concentration has no effect on associated mechanical equipment. Therefore, an unreviewed safety question does not exist.

• EWR 88-296 ENGINEERING WORK REQUEST 07/13/88 This engineering work request evaluated operation with the EDG room main air intake louvers failed open.

The change does not affect the operation of the EDGs because the louvers are open while they are in operation. Leaving the louvers open will help to enhance air displacement (ih the event of a fire) if the carbon dioxide system is actuated. Therefore, an unreviewed safety question does not exist.

1/2-0P-8. l.3 OPERATING PROCEDURES 07/14/88 thru OP-8.1.8 These operating procedures provide detailed instructions for shifting from one charging pump in operation to starting and operating a second charging pump and securing the previously running pump.

For conditions >350°F/450 PSIG simultaneous operation of more than one charging pump will have no effect on the listed accidents as a safety injection signal results in an auto start of all available charging pumps. For conditions < 350°F /450 PSIG with OPMS in operation, the procedure places restrictions on the simultaneous running of two charging pumps. These are: 1) a maximum of three minutes in this mode, 2) a maximum of 33% p;m level with a bubble in the P~R, or

3) two PORVs fully operable and unisolated. These conditions ensure applicable operating curves will be met and will prevent exceeding cold overpressurization 10CFRSO, App. J limits and are within the basis of Technical Specification, section 3.1.G.

e e PAGE 19 PROCEDURE OR METHOD OF OPERATION THAT DID REQUIRE NRC APPROVAL MONTH/YEAR ____

JULY 1988 NONE DURING TIIlS PERIOD

PAGE 20 TESTS AND EXPERIMENTS REQUIRING NRC APPROVAL MONTH/YEAR JULY 1988 NONE DURING THIS PERIOD

PAGE 21 TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR JULY 1988 NONE DURING THIS PERIOD

e e PAGE 22 VIRGINIA .POWER SURRY POWER STATION CHEMISTRY REPORT

,IlJT,Y 19 BB PRIMARY COOLANT UNIT NO. 1 ANALYSIS UNIT NO. 2 MAX. MIN. AVG. MAX. MIN. AVG.

Gross Radioact., uCi/ml 6. 77E-1 4.48E-3 l.24E-1 2.51E-1 9.24E-2 1. 78E-1 o.o o.o a.a a.a

.a.a o.o Suspended Solids, 1>1>m Gross Tritium, uCi/ml 7.87E-2 l.89E-2 4.88E-2 3.86E-2 2.lBE-2 2.77E-2 Iodine 131 , uCi/ml 1.84E-2 6.22E-5 l.08E-2 1.22E-3 8.87E-4 l.07E-3 11'31 / 11'31 0.13 0.07 0.10 0.14 0.06 0.09 Hydrogen, cc/ksl. 30.5 21.3 27.2 34.2 25.1 29.7 Lithium, ppm 2.69 0.75 1. 72 0.55 0.28 0.41 Boron-10, ppm* 409.6 196.2 310.3 10.976 0.196 4.102 Oxygen, (DO), ppm 5.00 0.005 0.149 0.005 0.005 0.005 Chloride, ppm 0.010 0.001 0.006 0.010 0.001 0.001 pH@ 25 degree Celsius 6.46 4.64 5.86 8.65 7.18 7.69

• Boron-10

• Total Boron X 0.196 REMARKS: UNIT ONE: started the month at Refuelinq Shutdown and was at 100% by

• the end of the I!!Onth.

• 4. 75 gallons of Hydrazine were added to the RCS for oxygen removal. Lithium was added on 7-1 at 1000, 324q;7-9 at 1030, 850g;7-9 at 1815,

• 325g;7-9 at2125, 325g;7-10 at 0200, 325g;7-11 at 1830, 175g; 7-12 at 0915, 810g; 7-13 at 1415, 1150g;7-14 at 0215, 974g;7-15 at 2035, 325g;7-19 at 0500, 535g; 7-19 at 2130, 400g;7-21 at 1110, 400g;7-23 at0400, 220g;7-25 at 0345, 285g;7-27 atllOO, 360g;7-28 at 2140, 270g; for a total addition of 8053 grams LiOH.

UNIT 'IWO: stC;lrted the m::>nth at 100% and was coasting to Refueling at the end. .The Cat Bed was placed in service for Lithium control ori 7...1. from 1130 until 1720. Lithium additions on 7-24 at 0920, 211g and 7.-27 at 1830, 200g; for total addition of 411grams LiOH.

e e PAGE 23

• UNIT 1 FUEL HANDLING DATE July 1988 NEW OR DATE NUMBER OF NEW OR SPENT SPENT FUEL SHIPPED OR ASSEMBLIES ASSEMBLY ANSI INITIAL FUEL SHIPPINC SHIPMENT II RECEIVED PER SHIPMENT 11 11 ENRICHMENT CASK ACTIVITY LEVEL NONE DU ~ING THIS P 8RI0D

• PAGE 24 UNIT __ 2_ FUEL HANDLING DATE July 1988 NEW OR DATE NUMBER OF NEW OR SPENT SPENT FUEL SHIPMENT fl SHIPPED OR RECEIVED ASSEMBLIES PER SHIPMENT ASSEMBLY fl ,

ANSI INITIAL ENRICHMENT FUEL SHIPPING CASK ACTIVITY LEVEL NONE DUJ ~ING THIS P ~RIOD

,J,L ....., .

e

• PAGE 25 DESCRIPTION OF PERIODIC TEST WHICH WERE NOT COMPLETED WITHIN THE TIME LIMITS SPECIFIED IN TECHNICAL SPECIFICATIONS MONTH/YEAR JULY 1988 NONE DURING THIS PERIOD

.. II ..

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 2326l August 15, 1988 D.S.CRUDEN VICE PRESIDENT-NUCLEAR U.S. Nuclear Regulatory Commission Serial No.88-539 Attention: Document Control Desk NO/DAS:vlh Washington, D. C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 MONTHLY OPERATING REPORT Enclosed is the Monthly Operating Report for Surry Power Station Units 1 and 2 for the month of July 1988.

Enclosure cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.

Suite 2900 Atlanta, Georgia 30321 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station