ML18152A305

Monthly Operating Repts for Oct 1989 for Surry Power Station Units 1 & 2

ML18152A305
Person / Time
Site:	Surry
Issue date:	10/31/1989
From:	Arney C VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18152A306	List: ML18152A305 ML18153B991 ML18153B992
References
NUDOCS 8911280507
Download: ML18152A305 (29)

Text

e ATTACHMENT 1 MONTHLY OPERATING REPORT OCTOBER 1989

POW 34-04 VIRGINIA ELECTRIC AND POWER COMPANY SORRY POWER STATION MONTHLY OPERATING REPORT REPORT #89-10 APPROVED:

TABLE OF CONTENTS 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 Requiring NRC Approval 16 Procedure or Method of Operation Changes that Did Not Require NRC Approval 17 Tests and Experiments Requiring NRC Approval 18 Tests and Experiments That Did Not Require NRC Approval 19 Chemistry Report 23 Fuel Handling - Unit No. 1 24 Fuel Handling - Unit No. 2 25 Description of Periodic Test Which Were Not Completed Within the Time Limits Specified in Technical Specifications 26 i

OPERATING DATA REPORT DOCKET NO.: 50-280 DA TE : --.,-No_v_e_m,_be-r----=;7-,-:1~9=s=9--

COM PL ET ED BY: C.A. Arney TELEPHONE: ----,.-,(8'"""0~4)'"""3"""'57=-_-3-=-1s"""""'4,-x--,3=5=5~

OPERATING STATUS NOTES

1. Unit Name: Surry Unit 1

2. Reporting Period: Oct. 1 - Oct 31, 1989

3. Licensed Thermal Power (MWt):2441

4. Nameplate Rating (Gross MWe):847.5

5. Design Electrical Rating (Net MWe): 788

6. Maximum Dependable Capacity (Gross MWe): 820

7. Maximum Dependable Capacity (Net MWe): 781

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

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

10. Reason For Restrictions, If Any: -----------

THIS MONTH YTD CUMULATIVE

11. Hours In Reporting Period 745 7296. 0 147792. 0

12. Number of Hours Reactor Was Critical 745 2833.0 91311.6

13. Reactor Reserve Shutdown Hours 0 0 3774.5

14. Hours Generator On-Line 745 2781.7 89387.1

15. Unit Reserve Shutdown Hours 0 0 3736.2

16. Gross Thermal Energy Generated (MWH) 1795742.0 6481089.0 207652356.0

17. Gross Electrical Energy Generated (MWH) 610,275 2165190.0 67368863.0

18. Net Electrical Energy Generated (MWH) 580,566 2049841.0 63890244.0

19. Unit Service Factor 100% 38.1% 60.5%

20. Unit Available Factor 100% 38.1% 63.0%

21. Unit Capacity Factor (Using MDC Net) 99.8% 36.0% 55.8%

22. Unit Capacity Factor (Using DER Net) 99.0% 35.7% 54.9%

23. Unit Forced Rate 0% 61.9% 21.9%

24. Shutdowns Scheduled Over Next 6 Months (Type, Date , and Duration of Each):

25. If Shut Down at End of Report Period Estimated Date of Startup:,--___,,.'=""""~="'""-

26. Unit In Test Status (Prior to Commercial Operation): FORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 1

~

1.

• OPERATING DATA REPORT DOCKET NO.: 50-281 DATE: -=-No_v_e-mb.-e-r--=-7,~1-=-98=9:----

COMP LETE D BY: C.A. Arney TELEPHONE: --r:,(8=0-::-4)<"'=3=57=-_--=3-=-:18==-=4-x--:3=5= OPERATING STATUS NOTES

1. Unit Name: Surry Unit 2

2. Reporting Period: Oct. 1 - Oct. 31, 1989

3. Licensed Thermal Power (MWt):2441

4. Nameplate Rating (Gross MWe):847.5

5. Design Electrical Rating (Net MWe): 788

6. Maximum Dependable Capacity (Gross MWe): 820

7. Maximum Dependable Capacity (Net MWe): 781

8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last

~eport, Give Reasons:

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

10. Reason For Restrictions, If Any: -----------

THIS MONTH YTD CUMULATIVE

11. Hours In Reporting Period 745 7296. 0 144672. 0

12. Number of Hours Reactor Was Critical 275.3 603.2 90297.5

13. Reactor Reserve Shutdown Hours 0 0 328.1

14. Hours Generator On-Line 270.5 459.5 88752.5

15. Unit Reserve Shutdown Hours 0 0 0

16. Gross Thermal Energy Generated (MWH) 550555.5 7383114. 0 207478747.5

17. Gross Electrical Energy Generated (MWH) 180425. 0 227150.0 67363394.0

18. Net Electrical Energy Generated (MWH) 170864.0 211695.0 63859073.0

19. Unit Service Factor 36.3% 6.3% 61. 3%

20. Unit Available Factor 36.3% 6.3% 61.3%

21. Unit Capacity Factor (Using MDC Net) 29.4% 3.7% 56.6%

22. Unit Capacity Factor (Using DER Net) 29.1% 3.7% 56.0%

23. Unit Forced Rate 63.7% 55.5% 15.4%

24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):

25. If Shut Down at End of Report Period Estimated Date of Startup: 11-23-89

26. Unit In Test Status (Prior to Commercial Operation): FORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 2

DOCKET NO.: 50-280 ~*

UNIT SHUTDOWN AND POWER REDUCTION UNIT NAME: Surry Unit 1 DATE: November 7, 1989 COMPLETED BY: C.A. Arney REPORT MONTH: October 1989 TELEPHONE: 804-357-3184 x355 METHOD OF LICENSEE CAUSE &CORRECTIVE DURATION SHUTTING EVENT SYSTEM COMPONENT ACTION TO PREVENT NO. DATE TYPE(l) (HOURS) REASON(2) DOWN REACTOR(3) REPORT# CODE(4) CODE(5) RECURRENT N/A 10-20-89 s 0 B,A 4 N/A SB TRB Schedule Ramp Down ~

1-PT-29.l Governor Valve Free-dom Test. While at reduced power an upper quadrant power tilt was indicated due to excores being out of calibration. Flux map was performed which verified no tilt and that hot channel factors were in spec. Cleaned all 4 water boxes while at reduced power.

(1) (2) (3) (4)

F: Forced REASON:

• METHOD:

S: Scheduled. A - Equipment Failure (Explain) 1 - Manual Exhibit G - Instructions for B - Maintenance or Test 2 - Manual Scram. Preparation of Data Entry Sheets C - Refueling 3 - Automatic Scram. for Licensee Event Report (LER)

D - Regulatory Restriction 4 - Other (Explain) File (NUREG 0161)

E - Operator Training & License Examination F - Administrative (5)

G - Operational Error (Explain)

H - Other (Explain) 3 Exhibit 1 - Same Source

DOCKET NO.: 50-281 ~*

UNIT SHUTDOWN AND POWER REDUCTION UNIT NAME: Surry Unit 2 DATE: November 7, 1989 COMPLETED BY: C.A. Arney REPORT MONTH: October 1989 TELEPHONE: 804-357-3184 x355 METHOD OF LICENSEE CAUSE &CORRECTIVE DURATION SHUTTING EVENT SYSTEM COMPONENT ACTION TO PREVENT NO. DATE TYPE(l) (HOURS) REASON(2) DOWN REACTOR(3) REPORT# CODE(4) CODE(5) RECURRENT 10/02/89 F 0 4 N/A SJ MO,P Ramped down to remove 2-FW-r~

from service for alignment.

Realigned pump and motors.

C-10-3 10/12/89 F 468 hrs. A 1 S2-89-13 AB RV 2-RC~SV-25518 leaking by seat.

44 min. Placed unit in Cold Shutdown -

sent all three safety valves out for recalibration.

(1) (2) (3) ( 4)

F: Forced REASON: METHOD:

S: Scheduled A - Equipment Failure (Explain) 1 - Manual Exhibit G - Instructions for .

B - Maintenance or Test 2 - Manual Scram. Preparation of Data Entry Sheets C - Refueling 3 - Automatic Scram. for Licensee Event Report (LER)

D - Regulatory Restriction 4 - Other (Explain) File (NUREG 0161)

E - Operator Training &License Examination F - Administrative (5)

G - Operational Error (Explain)

H - Other (Explain) Exhibit 1 - Same Source 4

.- *1 e

AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.: --=---~---

50-280 UN IT NAME: Surry Unit 1 DATE: November 7, 1989 COMPLETED BY: C.A. Arney TELEPHONE:(804)357-3184 x355 MONTH: October 1989 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 788 17 783 2 788 18 782 3 783 19 782 4 786 20 782 5 787 21 623 6 784 22 789 7 782 23 788 8 787 24 788 9 786 25 786 10 786 26 782 11 785 27 783 12 785 28 784 13 781 29 819 14 783 30 784 15 782 31 782 16 782 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.

5

AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.: -=--~..--,-~--

50-281 UN IT NAME: Surry Unit 2 DATE: November 7, 1989 COMPLETED BY: C.A. Arney TELEPHONE:(804)357-3184 x355 MONTH: October 1989 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 268 17 0 2 450 18 0 3 442 19 0 4 476 20 0 5 602 21 0 6 772 22 0 7 775 23 0 8 781 24 0 9 784 25 0 10 785 26 0 11 784 27 0 12 546 28 0 13 0 29 0 14 0 30 0 15 0 31 0 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.

6

e

SUMMARY

OF OPERATING EXPERIENCE MONTH/YEAR: October 1989 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 10-01-89 0000 Reporting period begins with unit at 100% power.

10-20-89 2252 Start ramp down for 1-PT-29.1.

10-21-89 0042 Stop ramp (74%, 600 MW) 1702 Commenced ramping up 2011 Unit at 100%, 825 MW 10-31-89 2400 Unit ended reporting period at 100% power.

7

(

e

SUMMARY

OF OPERATING EXPERIENCE MONTH/YEAR: October 1989 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 10-01-89 0000 Unit at hot standby, 1 x 10- 5 amps.

0439 Unit on line, ramping up.

0600 Stop ramp to shift main feed pumps.

1300 Commenced ramp up (40%, 290 MW).

1520 Stop ramp (49%, 330 MW) for PT 2.1.

1858 Commenced ramp up.

2029 Stop ramp (52%, 370 MW) - Recommencing PT 2.1.

2135 Start ramp up.

10-02-89 0445 Stop ramp (70%, 500 MW) for flux map.

1735 Commenced ramping down.

1811 Stop ramp (60%, 480 MW) removing 2-FW-P-18 from service for realignment.

10-03-89 2320 Commenced slow ramp up.

10-04-89 0145 Stop ramp (63%, 520 MW).

10-05-89 0521 Commenced ramping up. 2-FW-P-18 returned to service.

0815 Stop ramp (70%, 580 MW) due to polishing building D/P.

0845 Commenced ramping up.

1230 Stop ramp (81%, 660 MW) - Unit entered 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> clock to hot shutdown due to MOV-2350 line being declared out of service. Hydro not done after repairs.

1440 Start ramp up (6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> clock to hot shutdown terminated@

1415, based on ASME code case and discussions with NRC).

1834 Stop ramp (92%, 750 MW) due to receiving high hydrogen gas temperature alarm on the main generator.

1955 Start ramp up - H2 Gas alarm clear.

8

J

• e e UNIT TWO (cont'd) 10-05-89 2300 Stop ramp (99.4%, 810 MW) - Received high delta T alarm.

2303 Unit stable@ 99%, high delta T alarm clear.

10-08-89 1000 Slowly increased power from 99% to 100%, 825 MW.

10-12-89 0355 Start ramping down (100%, 825 MW) due to leaking pressurizer safety valves (leaking by to the PRT).

0620 Stop ramp@ 68% to adjust IRPI s.

1 0635 Restart ramp down.

1103 Unit off line.

1116 Tripped reactor manually for normal shutdown.

2012 Mode change - commenced cooldown to cold shutdown.

10-13-89 0345 Stop cooldown at 345° F for Chemistry Hold.

1014 Recommenced cooldown to Cold Shutdown.

1520 Unit at Cold Shutdown.

10-31-89 2400 Unit ended reporting period at Cold Shutdown.

9

e FACILITY CHANGES REQUIRING NRC APPROVAL MONTH/YEAR: October 1989 JC0-89-100 JUSTIFICATION FOR CONTINUED OPERATION (JCO) 19-89 This JCO, including a Safety Evaluation (#89-0002), was prepared to support discretionary enforcement for Unit 1 operation with the pressurizer safety valve lift settings potentially outside Technical Specification tolerance limits. This condition could result due to the testing methodology utilized in setting the valves. Test data shows that Surry's valves may have a tolerance shift in one direction between 3.5% and 5.0%.

The analyses performed to support the JCO indicate that the current UFSAR accident analyses remain bounding for valve lift settings below 5.4%. Since test data on Surry valves show a lift setting shift below 5.4%, no unreviewed safety questions exists. The JCO also provides compensatory measures to ensure that any overpressure transient remains within design limits for lift setting shifts in excess of 5.4%. Six weeks of discretionary enforcement was granted by NRC on 10-19-89.

10

\

e e FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: October 1989 DC 86-10 SERVICE WATER AND CIRCULATING WATER BUTTERFLY VALVE REPLACEMENT -

NIT 1 This design change replaced 96 11 , 42 11 , 36 11 and 30 11 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 and reported in July 1988.

MOV-CW-106A, B, C, D CW inlets MOV-SW-102A, B, SW to CC HX 1-SW-25, 29, 33, 37 CC HX SW inlets The following valves have now been replaced which completed the design change.

MOV-CW-lOOA~ B, C, D CW outlets MOV-SW-lOlA, B SW to BC HX 1-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 speci-fication requirements which meet or exceed the original specifications. The equipment will operate and function identically 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.

SCAFFOLDING REQUEST 03-89 Temporary scaffold erected in Unit 1 Cable Vault to work Cable Tray Covers.

Installation of this temporary scaffold was reviewed for effect on Accident Analyses and Equipment Operability/ Function. It is concluded that assumptions, bases, and probabilities of accident analyses and equipment malfunctions are not affected.

11

• ~

e e SCAFFOLDING REQUEST 05-89 Erection of temporary scaffolding located in 2-CH-P-lC Pump Cubicle.

Installation of this temporary scaffolding was reviewed for effect on accident analyses and equipment operability/func-tion. Conclusion is that assumptions, bases and probabili-ties of accident analyses and equipment malfunctions are not affected. Scaffolding will not be erected over the other Unit 2 charging pumps or the cross tie from Unit 1 while installed over 2-CH-P-lC.

SCAFFOLDING REQUEST 12-89 Erection of temporary scaffolding located in Unit 2 Safeguard Alleyway to work Unit 2 Safeguards Gutters.

Installation of this temporary scaffolding was reviewed for effect on accident analyses and equipment operability/func-tion. Conclusion is that assumptions, bases and probabili-ties of accident analyses and equipment malfunctions are not affected.

SCAFFOLD REQUEST 10-16-89 Erection of temporary scaffolding located in Unit 2 Contain-ment at penetration area to calibrate remote standpipe transmitter.

Installation of this temporary scaffolding was reviewed for effect on accident analyses and equipment operability. The conclusion is that assumptions, bases and probabilities of accident analyses and equipment malfunctions are not affected.

TM-S2-89-127 TEMPORARY MODIFICATION (SAFETY EVALUATION #089-000lA) -

10-17-89 Temporary modification to remove check valve internals from 2-BD-588 and install adapter and jumper hose to Condensate Polishing Waste Neutralization Sump to allow release of the contents of Steam Generators.

The Steam Generators will be depressurized during the time this jumper is installed; also the Reactor Coolant System (RCS) will be depressurized and at Cold Shutdown.

Therefore, the secondary heat sink is not required. In addition, any leak that could occur in the jumper could be isolated by closing the Slowdown Trip Valves. Therefore no unreviewed safety question exists.

12

SCAFFOLDING REQUEST 17-89 This temporary scaffold will be erected to support W.0.#

3800086812 to repair/replace valve 02-SI-79.

Installation of this temporary scaffold was reviewed for effect on accident analyses and equipment operability/func-tion. Conclusion is that assumptions, bases and probabili-ties of accident analyses and equipment malfunctions are not affected.

SCAFFOLDING REQUEST 18-89 Request to temporarily erect scaffold to support the repair of valve 2-MS-116 under W.0.# 3800086546.

Installation of this temporary scaffolding was reviewed for effect on accident analyses and equipment operability/func-tions. Conclusion is that assumptions, bases and probabili-ties of accident analyses and equipment malfunctions are not affected.

TM-S2-89-129 TEMPORARY MODIFICATION (SAFETY EVALUATION #089-0003) -

10-20-89 Temporary modification to maintain a component cooling water flowpath operable while maintenance was performed on trip valve operator.

Since containment integrity will not be required while change is in effect and redundant Residual Heat Removal loops will remain operable, no unreviewed safety question is created.

SCAFFOLDING REQUEST (SAFETY EVALUATION 089-0004) 20-89 Request for er*ection of temporary scaffolding to support inspection and repairs to valve 2-SS-SOV-501C located in Unit 2 Containment annulus area at elevation(-) 3 1 -6 11 near penetration number 38, per Work Order Number 3800077473.

The temporary scaffold is required for safe performance of work. Installation of scaffolding constructed per SUADM-ADM-07 has a high confidence level against failure and was reviewed for effects on accident analyses and equipment operability/function. It is thus concluded that assumptions, bases and probabilities of accident analyses and equipment malfunctions are not affected.

13

e TM-S2-89-130 TEMPORARY MODIFICATION 24-89 TM-S2-89-131 TM-S2-89-132 Temporary modification to allow starting of reactor coolant pumps under cold reactor coolant system conditions without RCP locked rotor motor protection due to normal cold reactor coolant system starting amps.

With the installation of this Temporary Modification the protection provided for a Reactor Coolant Pump locked rotor is defeated. However, independent verification of relay reinstallation shall be provided. Also, upon Reactor Coolant Pump start, Control Room Operators must verify that increased Reactor Coolant System Loop Flow is indicated within ten (10) seconds after breaker closure. If flow indication is not shown, then the pump breaker shall be tripped open. During this installation, the RHR cooling loops will remain operational to provide necessary decay heat removal. Finally, an unreviewed safety question is not created because the Reactor Coolant System will remain at Cold Shutdown Conditions during this installation. Thus a loss of reactor coolant flow will not lead to a loss of core heat removal.

SCAFFOLDING REQUEST (SAFETY EVALUATION# 089-0011) -

10-27-89 Scaffolding request - W.O. # 3800069908 - for the temporary erection of scaffolding to support replacement of power cables on spent fuel handling crane 1-FH-BRDG-13 and hoists 1-FH-H01-13A, B, located in Fuel Handling Building.

The temporary scaffold is required for safe performance of work. Installation of scaffold constructed per SUADM-ADM-07 has a high confidence level against failure and was reviewed for effect on accident analyses and equipment operability/

function. It is thus concluded that assumptions, bases and probabilities of accident analyses and equipment malfunc-tions are not affected.

SCAFFOLDING REQUEST (SAFETY EVALUATION# 089-0012) -

10-27-89 Request for erection of temporary scaffolding located in Unit 2 Safeguards Building just above line 30 11 -SHP-101-601 and around valve 02-MS-NRV-201A per W.O. #3800087100. The 8 1 x8 1 x6 1 high scaffold will be erected from the grating at elevation 50 1 -0 11 and will be supported from the grating and structural steel in the area.

The temporary scaffold is required for safe performance of work. Installation of scaffold constructed per SUADM-ADM-07 has a high confidence level against failure and was reviewed for effects on accident analyses and equipment operability/

function. It is thus concluded that assumptions, bases and probabilities of accident analyses and equipment malfunc-tions are not affected.

14

l SCAFFOLDING REQUEST (SAFETY EVALUATION# 809-0013) -

10-27-89 Scaffolding request for erection of temporary scaffold to support the maintenance work on the leaks of valve 02-MS-NRV-201B per W.O. # 3800087101.

The temporary scaffold is required for safe working.

Installation of scaffold constructed per SUADM-ADM-07 has a high confidence level against failure and was reviewed for effects on accident analyses and equipment operability/func-tion. It is thus concluded that assumptions, bases and probabilities of accident analyses and equipment malfunc-tions are not affected.

EWR 89-137 ENGINEERING WORK REQUEST 10-27-89 EWR required to improve the reliability and maintainability of rising stem safety related motor operated valves (MOVs).

Accident Analysis reviewed Loss of Coolant Accident (LOCA) and High Energy Line Break (HELB) are not affected and the probability of increasing the chances for a LOCA or HELB does not occur. The changes serve to ensure proper operation of the MOVs.

SCAFFOLDING REQUEST (SAFETY EVALUATION# 089-0014) -

10-29-89 Scaffolding request to erect scaffolding located in 1 A1 Reactor Coolant Pump cubicle to work 2-SI-91 (Hot Leg SI Check Valve)

Installation of temporary scaffolding was reviewed for effect on accident analyses and equipment operability/func-tion. Conclusion is that assumptions, bases, and probabili-ties of accident analyses and equipment malfunctions are not affected.

15

PROCEDURE OR METHOD OF OPERATION CHANGES REQUIRING NRC APPROVAL MONTH/YEAR: October 1989 None during this reporting period.

16

J PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

• October 1989 2-TOP-2023 TEMPORARY OPERATING PROCEDURE (SAFETY EVALUATION# 089-0001) -

10-16-89 Temporary Operating Procedure to allow maintenance on the three pressurizer safety valves simultaneously.

The potential for overpressurizing the RCS while at shutdown was considered and it is concluded that the filter/screen placed over one of the safety valve openings meets the intent of having one safety valve operable when the head is on the Reactor. Therefore an unreviewed safety question does not exist.

17

,,-------~-------~------------~

TESTS AND EXPERIMENTS REQUIRING NRC APPROVAL MONTH/YEAR: October 1989 None during this reporting period.

18

r

• ~ \.~

TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: October 1989 1-ST-237 SPECIAL TEST 17-89 This Special Test documents completion of test procedures that were deve 1oped to test 11 W bus Engineered Safety Features ( ESF) actuations concurrent with an undervoltage(UV) signal. These procedures incorporate previous station Periodic Test 18.2A and Periodic Test 8.5A. In addition the post modification testing for Design Change (DC) 88-31 EOG load sequencing was included.

Station equipment was operated and tested as designed and analyzed. In addition, since the EOG load bank was used within Emergency Diesel Generator (EOG) design, an unreviewed safety question was not created.

ST-238 SPECIAL TEST 24-89 ST-239 ST-232 A special test procedure was developed to test 11 J 11 bus ESF actuations concurrent with an UV signal. This procedure incorporated previous station PTs 18.28, 8.5A. In addition 88-31 testing was included (EOG load sequencing). On the test inserting an UV signal 5 minutes after the ESF actuation a load bank was used to simulate actual emergency bus load. The use of the load bank was to provide greater assurance of proper load sequencing and EOG operation during a HI-HI-CLS actuation. The periodic test and D.C. testing are being combined for ease of performance.

Station equipment was operated and tested as designed and analyzed. In addition, since the EOG load bank was used within EOG design, an unreviewed safety question was not created.

2-ST-270 SPECIAL TEST 27-89 This special test documents completion of the NRC Generic Letter 88-14 requirement that fail safe design basis testing be performed on safe shutdown air activated components which are operated by backup air accumulators upon loss of the normal air supply.

The components were operated as designed. The Technical Specifications Limiting Conditions of Operation (LCOs) were met prior to commencement of test. Therefore, an unreviewed safety question was not created.

19

I

• --1\...

2-ST-253 SPECIAL TEST 10-89 This Special Test was written to document pressure/stroke testing of the Auxiliary Feedwater (AFW) MOVs against the motor driven auxiliary feedwater pump discharge.

The reactor was in Cold Shutdown and decay heat removal was provided by the Residual Heat Removal system. Core Cooling via the Feedwater/Main Steam (FW/MS) system was not required per Technical Specifications. Primary pressure/temperature was controlled by Residual Heat Removal. Therefore, cycling an AFW valve does not present an unreviewed safety question.

2-ST-264 SPECIAL TEST 15-89 This Special Test was completed in response to NRC Generic Letter 88-14, and documented the walkdown of the Unit 2 Safe Shut- down air operated valves to ensure that the air supply components are installed properly and free from damage or deformation.

No components or equipment were operated or manipulated. This was a walkdown and visual inspection. Therefore, this Special Test did not produce an unresolved safety issue.

2-ST-271 SPECIAL TEST 15-89 This Special Test verified the power supply for safety related miscellaneous components.

This test verified design basis power supply integrity for safety related components, therefore no unreviewed safety question was created.

1-ST-236 SPECIAL TEST 09-89 This Special Test documents completion of test procedures that were developed to test 11 W bus ESF actuations concurrent with an undervoltage signal. These procedures incorporate previous station PTs 18.2A and 8.5A. In addition, the post testing for DC 88-31 EOG load sequencing was included.

Station equipment was operated and tested as designed and analyzed. In addition, since the EOG load bank was used within EOG design an unreviewed safety question was not created.

1-ST-242 SPECIAL TEST 09-89 This Special Test completion documents that Unit 1 11 811 train emergency, vital, and DC buses were de-energized to verify that only 11 811 train components are powered from these buses.

The required plant conditions combined with the ability to energize 11 811 train components in a timely manner ensured an unreviewed safety question was not created.

20

i

'¥~ \.,

2-ST-246 SPECIAL TEST 09-89 This Special Test completion documents that Unit 2 11 811 train emergency, vital, and DC buses were de-energized to verify that only 11 811 train components are powered from these buses.

The required plant conditions combined with the ability to energized 11 811 train components in a timely manner ensured an unreviewed safety question was not created.

1-ST-240 SPECIAL TEST 11-89 This Special Test documents completion of test procedures that were developed to test 11 H" bus ESF actuations concurrent with an undervoltage signal. These procedures incorporate previous station PT s 18.2A and 8.5A. In addition, the post testing for 1

DC 88-31 EOG load sequencing was included.

Station equipment was operated and tested as designed and analyzed. In addition, since the EOG load bank was used within EOG design, an unreviewed safety question ~as not created.

1-ST-241 SPECIAL TEST 11-89 This Special Test completion documented that Unit 1 11 A11 train emergency, vital, and DC buses was de-energized to verify that A train components was powered from these buses.

11 11 The required plant conditions combined with the ability to energize 11 A11 train components in a timely manner ensured an unreviewed safety question was not created.

ST-241 SPECIAL TEST 11-89 ST-245 Special Test to document that Breaker 15J3 was placed in the test position to support Unit 2 logic testing. The breaker was maintained under Administrative Control and was racked into the connect position in case of Unit 1 11 J 11 Bus UV/DV.

The affected breaker was under Administrative Control and would have been immediately racked into the connect position and closed in case of Unit 1 UV/DV. The unit (Ul) was verified to not be in an LCO for 11 A11 train components prior to racking the breaker in test.

2-ST-247 SPECIAL TEST 13-89 This Special Test provided a method to obtain base line data for trending of the service water header isolation valve leakage.

This test did not constitute an unreviewed safety question, nor a change to Technical Specifications. This conclusion was based on a review of Technical Specifications, UFSAR and Test Procedure performance requirements.

21

2-ST-274 SPECIAL TEST~ 09-13-89 This Special Test completion verified the ability of 12 selected valves to operate under differential pressure.

No unreviewed safety question was created because the test did not create an unanalyzed condition or increase the possibility or consequences of an analyzed accident and did not increase the probability of or consequences of an equipment malfunction.

ST-251 SPECIAL TEST 13-89 This Special Test measured the leakage through each of the condenser inlet and outlet motor operated valves.

This Special Test did not constitute an unreviewed safety question nor require a change in Technical Specifications. This conclusion was based on the review of applicable Technical Specifications, applicable UFSAR sections and Test Procedure Performance requirements. The equipment affected by this test was operated in its normal manner.

ST-277 SPECIAL TEST 20-89 This Special Test verified that the addition of an instrument air bypass line with a regulator valve (needle valve) to the volume booster relay of main steam turbine dump bypass valve stabilizes the cycling action of the valve.

An unreviewed safety question was not created by this special test of a non-safety related valve because the valve was isolated from the main steam system and its operation was not required to achieve or maintain the safety shutdown of the power station.

The valves are normally required for up to 50% station load rejection. The remaining valves not in test had sufficient capacity to perform this function.

ST-276 SPECIAL TEST 31-89 This special test provided a controlled test method that was used to develop baseline data for plant periodic tests that will verify future pump operability.

The pumps and piping in question was operated in a normal manner and met the Technical Specification Requirements. Therefore, an unreviewed safety question was not created.

22

.r . ~

• *--* ,* -..,r '**

VIRGINIA POWER SURRY POWER STATION CHEMISTRY REPORT MONTH/YEAR: October 1989 PRIMARY COOLANT UNIT NO. 1 UNIT NO. 2 ANALYSIS MAX. MIN. AVG. MAX. MIN. AVG.

Gross Radi oact., µCi/ml 8.35E-1 5.45E-1 6.85E-1 2.47E-1 2.37E-3 4.49E-2 Suspended Solids, ppm 0.0 0.0 0.0 0.0 0.0 0.0 .

Gross Tritium, µCi/ml 4.44E-1 3.31E-1 3.59E-1 1. 09E-1 4.70E-2 7.80E-2 Iodine-131, µCi/ml 4.33E-3 2.47E-3 3.23E-3 1. 81E-3 5.67E-5 3.03E-4 Iodine-131/Iodine-133 0.18 0.08 0.11 0.12 0.07 0.09 Hydrogen, cc/kg 31.5 25.2 28.1 33.2 1. 9 15.2 Lithium, ppm 2.32 2.09 2.21 2.38 0.85 1. 91 Boron - 10, ppm* 144 128 136 451 196 357 Oxygen, (DO) , ppm ~ 0.005 z 0.005 .z. 0. 005 8.5 j: 0.005 1. 65 Chloride, ppm 0.011 0.007 0.009 o. 011 0.002 0.009 pH@ 25 degree Celsius 6.89 6.69 6.67 6.52 5.54 5.95

• Boron - 10c Total Boron X 0.196 REMARKS:

UNIT ONE: Unit one started October at 100% power with BMB in service. On 10/1/89 at 1421 the CAT bed was placed in service and at 1542, the CAT bed was removed from service. On 10/20/89, the unit ramped down to 75% power to clean water boxes.

On 10/31/89, letdown was secured due to a leak on the non-regenerative heat exchanger. Unit one ended the month at 100% power.

UNIT TWO: On 10/1/89, the unit came on-line and BMB was placed in service. A 115 gm lithium addition was also made on 10/1/89. Unit two reached 100% on 10/6/89.

The CAT bed was placed in service on 10/8 and 10/11/89 for lithium removal.

On 10/12, the unit began ramping to CSD for repairs to the pressurizer safety.

On 10/26 and 10/27 a total of seven (7) gallons of Hydrazine were added to the RCS for plant heatup. A 1700 gm lithium addition was also made on 10/27/89.

Unit 2 ended the month at CSD.

23

UNIT FUEL HANDLING DATE: October 1989 NEW OR DATE NUMBER OF NEW OR SPENT SPENT FUEL SHIPPED ASSEMBLIES ASSEMBLY ANSI INITIAL FUEL SHIPPING SHIPMENT# OR RECEIVED PER SHIPMENT NUMBER NUMBER ENRICHMENT CASK ACTIVITY LEVEL None during this reporting period.

24

UNIT - FUEL HANDLING DATE: October 1989 NEW OR DATE NUMBER OF NEW OR SPENT SPENT FUEL SHIPPED ASSEMBLIES ASSEMBLY ANSI INITIAL FUEL SHIPPING SHIPMENT# OR RECEIVED PER SHIPMENT NUMBER ~UMBER ENRICHMENT CASK ACTIVITY LEVEL None during this reporting period.

25

DESCRIPTION OF PERIODIC TEST WHICH WERE NOT COMPLETED WITHIN THE TIME LIMITS SPECIFIED IN TECHNICAL SPECIFICATIONS MONTH/YEAR: October 1989 None during this reporting period.

26