ML18152A383

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Monthly Operating Repts for June 1990 for Surry Power Station,Units 1 & 2.W/900716 Ltr
ML18152A383
Person / Time
Site: Surry  Dominion icon.png
Issue date: 06/30/1990
From: Stewart W, Warren L
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
90-439, NUDOCS 9007190314
Download: ML18152A383 (23)
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e e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 July 16, 1990 U. S. Nuclear Regulatory Commission Serial No.90-439 Attention: Document Control Desk NO/RPC: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 MONTHL V OPERATING REPORT Enclosed is the Monthly Operating Report for Surry Power Station Units 1 and 2 for the month of June 1990.

Very truly yours, W. L. Stewart Senior Vice President - Nuclear Enclosure cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.

Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station

- -I i

POW 34-04 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION MONTHLY OPERATING REPORT REPORT fl 90-06 APPROVED:

e 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 9 Facility Changes That Did Not Require NRC Approval 10 Procedure or Method of Operation Changes that Did Not Require NRC Approval 13 Tests and Experiments That Did Not Require NRC Approval 14 Chemistry Report 18 Fuel Handling - Unit No. 1 19 Fuel Handling - Unit No. 2 19 Description of Periodic Test(s) Which Were Not Completed Within the Time Limits Specified in Technical Specifications 20

e OPERATING DATA REPORT DOCKET NO.: 50-280 DATE: -0=7~;-=-05=-;=9-=-o-----

COMPL ETED BY: L.A. Warren TELEPHONE: (804)357-3184 x355 OPERATING STATUS NOTES

1. Unit Name: Surry Unit 1
2. Reporting Period: June 01-30, 1990
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 720.0 4343.0 153599.0
12. Number of Hours Reactor Was Critical 689.3 4084.3 96835.1
13. Reactor Reserve Shutdown Hours 0 0 3774.5
14. Hours Generator On-Line 680.4 4075.4 94898.6
15. Unit Reserve Shutdown Hours 0 0 3736.2
16. Gross Thermal Energy Generated (MWH) 1641700.0 9820850.5 220937653.5
17. Gross Electrical Energy Generated (MWH) 544085.0 3309710.0 71855113. 0
18. Net Electrical Energy Generated (MWH) 516488.0 3150436.0 68161366.0
19. Unit Service Factor 94.5% 93.8% 61.8%
20. Unit Availability Factor 94.5% 93.8% 64.2%
21. Unit Capacity Factor (Using MDC Net) 91.8% 92.9% 57.3%
22. Unit Capacity Factor (Using DER Net) 91% 92.1 56.3%
23. Unit Forced Outage Rate 5.5% 6.2% 21.1%
24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):

Refueling/Snubber Outage scheduled to begin 10/05/90, 58 days

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

e e OPERATING DATA REPORT DOCKET NO.: 50-281 DATE: 07/05/90 COMPLETED BY: L.A. Warren TELEPHONE: -(.,..,,8-0-4)~3-5~7--3-18-4,---x3"""5=5~

OPERATING STATUS NOTES

1. Unit Name: Surry Unit 2
2. Reporting Period: June 01-30, 1990
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 720.0 4343.0 150479.0
12. Number of Hours Reactor Was Critical 674.1 4197.0 95395.6
13. Reactor Reserve Shutdown Hours 0 0 328:1
14. Hours Generator On-Line 669.0 4186.3 93835.2
15. Unit Reserve Shutdown Hours 0 0 0
16. Gross Thermal Energy Generated (MWH) 1625852.5 9994947.3 219605282.1
17. Gross Electrical Energy Generated (MWH) 539325.0 3344855.0 71425454. 0
18. Net Electrical Energy Generated (MWH) 512156.0 3181995.0 67722954.0
19. Unit Service Factor 92.9% 96.4% 62.4%
20. Unit Availability Factor 92.9% 96.4 62.4%
21. Unit Capacity Factor (Using MDC Net) 91.1% 93.8% 57.7%
22. Unit Capacity Factor (Using DER Net) 90.3% 93% 57 .1%
23. Unit Forced Outage Rate 7.1% 3.6% 15.2%
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*---=-A=cH'"'"'I=E=v=ED=--

INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 2

DOCKET NO.: 50-280 UNIT SHUTDOWN AND POWER REDUCTION ' UNIT NAME: -s=-u-r-ry--=""un-i~t-1_ __

(Equal To or Greater Than 20%) DATE: 07/05/90 COMPLETED BY: L.A. Warren REPORT MONTH:

- -JUNE

- -1990

TELEPHONE: 804-357-3184 x355 METHOD OF LIGENSEE CAUSE & CORRECTIVE DURATION SHUTIING EVENT* . SYSTEM COMPONENT ACTION TO PREVENT NO. DATE TYPE(l) (HOURS) REASON(2) DOWN REACTOR{3) REPORT# CODE{4) CODE(5) RECURRENT 06/01/90 F 39.6 A 2 1-90-005 EL XFMR The Unit was at ~

shutdown as a result of-111" electrical fault in Unit 1

'A' main transformer that occurred on 05/22/90.

(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 Z (Equal To or Greater Than 20%) DATE: 07/05/90 COMPLETED BY: L.A. Warren REPORT MONTH: JUNE 1990 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) C0DE(5) RECURRENT 06/01/90 F 51.0 A 2 2-90-003 SJ FCV The Unit was at ~

shutdown as a result of the closure of 2-FW-FCV-2478A ( 1 A1 main feed regulator valve) which occurred on 05/31/90.

(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) 4 Exhibit 1 - Same Source

AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.: 50-280 UNIT NAME: Surry Unit 1 DATE: 07/05/90 COMPLETED BY: L.A. Warren TELEPHONE:(804)357-3184 x355 MONTH: JUNE 1990 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 0 17 775 2 267 18 773 3 706 19 771 4 769 20 772 5 773 21 768 6 776 22 770 7 775 23 743 8 775 24 745 9 774 25 751 10 774 26 743 11 777 27 747 12 779 28 748 13 777 29 759 14 777 30 773 15 778 31 16 777 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: 07/05/90 COMPLETED BY: L.A. Warren TELEPHONE:(804)357-3184 x355 MONTH: JUNE 1990 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (M~Je-Net) (MWe-Net) 1 0 17 772 2 0 18 766 3 577 19 760 4 770 20 765 5 776 21 763 6 775 22 757 7 773 23 771 8 773 24 771 9 772 25 770 10 770 26 769 11 771 27 766 12 774 28 763 13 773 29 760 14 773 30 763 15 775 31 16 774 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

SUMMARY

OF OPERATING EXPERIENCE MONTH/YEAR: __J_U_N_E_1_99_0_ __

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 06/01/90 0000 This reporting period started with the Unit stable at hot shutdown.

06/02/90 0645 Reactor critical.

1537 Unit on line.

06/03/90 0635 Unit at 100% power, 805 MW.

06/23/90 0715 Started slow ramp down to maintain condenser vacuum while 1

D waterbox was removed from service for trash grate 1

cleaning at the high level intake structure; 100% power, 780 MW.

1530 Started ramp up after 1 D1 waterbox was returned to service; 92% power, 710 MW.

1610 Stopped ramp; 100% power, 810 MW.

06/24/90 0819 Started ramp down; 1 D1 waterbox out of service for trash grate cleaning; 100% power, 780 MW.

1457 Started ramp up; 1 D1 waterbox returned to service and 1 C1 waterbox removed from service; 93.5% power, 750 MW.

1535 Stopped ramp; 97% power, 760 MW.

1738 Started ramp up; 1 C1 waterbox returned to service; 97.5%

power, 780 MW.

1753 Stopped ramp; 100% power, 805 MW.

06/25/90 1440 Started ramp down; 1 C1 waterbox removed from service for trash grate cleaning; 100% power, 780 MW.

2032 Started ramp up; 1 C1 waterbox returned to service; 96%

power, 770 MW.

2048 Stopped ramp; 100% power, 815 MW.

7

e e

SUMMARY

OF OPERATING EXPERIENCE MONTH/YEAR: - -JUNE---1990 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 06/26/90 1347 Started ramp down; removed 'B' waterbox from service for trash grate cleaning; 100% power, 775 MW.

1642 Started ramp up; 'B' waterbox returned to service; 84%

power, 655 MW.

1810 Stopped ramp; 100% power, 815 MW.

06/27/90 1330 Started ramp down; 'A' waterbox removed from service for trash grate cleaning; 100% power, 790 MW.

1430 Stopped ramp; 94% power, 730 MW.

2000 Started ramp up; 'A' waterbox returned to service; 94%

power, 735 MW.

2045 Stopped ramp; 100% power, 810 MW.

06/28/90 0844 Started ramp down; 'B' waterbox removed from service and prepared for performance of l-PT-29.1; 100% power, 790 MW.

0931 Stopped ramp; 94% power, 730 MW.

1106 Started ramp up; l-PT-29.l completed satisfactorily; 94%

power, 730 MW.

1345 Started ramp down to maintain condenser vacuum while 'B' waterbox remained out of service; 98% power, 760 MW.

1610 Started ramp up; 'B' waterbox returned to service; 90%

power, 700 MW.

1703 Stopped ramp; 100% power, 805 MW.

06/30/90 2400 This reporting period ended with the Unit operating at 100%

power, 815 MW.

8

SU"'1ARY OF OPERATING EXPERIENCE MONTH/YEAR: ~~J_U_N_E_1_99_0~~~

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 06/01/90 0000 This reporting period started with the Unit stable at hot shutdown.

06/02/90 2154 Reactor critical.

06/03/90 0259 Unit on line.

1104 Unit at 100% power, 820 MW.

06/30/90 2400 This reporting period ended with the Unit operating at 100%

power, 800 MW.

9

e FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: JUNE 1990 2-EWR-89-684A ENGINEERING WORK REQUEST 06/04/90 This change evaluated the possible addition of a small quantity of foreign material (flexitallic stainless steel ribbon) into the reactor coolant system due to previous maintenance activities on valve l-SI-79.

The issues considered by the possible addition of 304 stainless steel ribbon (flexitallic gasket material) are potential fuel damage and reactor coolant system leakage. An unreviewed safety question does not exist because any accident or malfunction is bounded by the UFSAR, Chapter 14 - Safety Analysis.

1-EWR-89-307 ENGINEERING WORK REQUEST 06/13/90 A spool piece was installed in the 1-RC-95 location.

The installation of a spool piece for 1-RC-95 will not change any accident analysis presented in the UFSAR. The valve does not perform any safety related function since the new spool piece had the same design characteristics as the original valve and did not affect associated equipment. Therefore, an unreviewed safety question does not exist.

FS-89-53 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) CHANGE REQUEST (Safety Evaluation #90-0150) 06/14/90 This UFSAR change is to make table 9.9.1 consistent with actual plant configuration.

An unreviewed safety question is not created by this UFSAR change since the table is being changed to clarify the automatic operation of the circulating water (CW) and service water (SW) motor operated valves (MOVs).

EWR-89-765 ENGINEERING WORK REQUEST UNITS 1&2 06/22/90 This change involved a permanent installation for the underground diesel fuel oil tanks level indicators, adjacent valves and tubing.

Temporary level indicators for the underground fuel oil tanks were installed pe.r temporary modification Sl-87-126. This EWR provided instructions for a permanent installation of the fuel oil tanks level indicators (1-FE-LI-lOOA, B), adjoining valves and tubing. An unreviewed safety question did not exist as a result of this installation, due to compatibility of material, safety related pressure boundary equipment, and no change in system function.

10

e FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: ~-J_U_N_E_1_9_90~~

JC0-90-1-001 JUSTIFICATION FOR CONTINUED OPERATION 06/22/90 JC0-90-2-001 (Safety Evaluation #90-0156)

Justification for continued operation of the turbine driven auxiliary feedwater (AFW) pump with over pressurization concerns resulting from a turbine governor malfunction with throttle discharge.

The combination of conservatisms in the design of the piping valves and fittings, and the procedural changes which have been implemented will ensure the integrity of the AFW pump discharge piping system. Therefore, the 10CFR50.59 review concluded that an unreviewed safety question did not exist.

TM-S2-90-08 TEMPORARY MODIFICATION 06/26/90 (Safety Evaluation #90-0157)

This temporary modification is to remove and replace test push button FC-474-TA from train 1 A1 safety injection circuitry. A jumper will be installed to maintain continuity on the Jl and Jl22 lines affected. Relay FC-474-XA will actuate causing one half of the required logic for high steam flow. This is equivalent to pushing the test push button.

This change places relay FC-474-XA in its tripped or safe condition. This condition would be present if channel 3 (three) steam flow on 1 A1 steam generator sensed high flow. Therefore, the safety injection signal resulting from high steam flow with low reactor coolant system (RCS) average temperature or low steam line pressure is not degraded by the change. A jumper will maintain continuity on the 1 A1 train Jl and Jl22 lines. If the jumper were to become dislodged it would disable 1 A1 train steam dump auto closure at 543° RCS average temperature and 1 A1 train header to line delta P alarms and computer inputs. Other indications of these parameters are not affected. Operator control of the steam dumps is not affected. 1 8 1 train of reactor protection/safeguards is not affected. Therefore, this change does not constitute an unreviewed safety question.

11

e FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: - -JUNE

- - 1990 FS-90-21 UPDATED FINAL SAFETY ANALYSIS REPORT CHANGE REQUEST 06/28/90 (Safety Evaluation #90-161)

This UFSAR change specifies that startup core power distribution measurements are made at power levels of 30% to 100%.

The first power distribution measurement will be taken at a power level between 0% and 30%. This should reduce the uncertainties in measuring the power distribution. NE Technical Report No. 499 documents that this change will have an insignificant impact on the limiting Condition II event, rod withdrawal at power. The revised startup physics testing program continues to provide adequate assurance that reload core characteristics are bounded by those assumed in the safety analysis. Since the existing accident analysis remains bounding, an unreviewed safety question is not created by this change.

FS-90-24 UPDATED FINAL SAFETY ANALYSIS REPORT CHANGE REQUEST 06/28/90 (Safety Evaluation #90-162)

This UFSAR change was initiated to replace the word 11 double 11 with the phrase "less than or equal to double" to reflect the more conservative setpoint for the control rod deviation alarm.

Additionally, this change will ensure that any future setpoint changes will not affect the correctness of the text of the UFSAR.

The description of the operation of the control rod deviation alarm will remain unchanged. The performance characteristics of the alarm (i.e. alarm setpoints) will be enveloped to ensure compliance with the actual station setpoints as specified in the precaution limitations and setpoint (PLS) document. Any future changes to the alarm setpoints will be accomplished via applicable station procedures ensuring that no unreviewed safety question will exist. The current setpoint is less than double the "at rest" setpoint and is therefore more conservative. As a result, no unreviewed safety question exists. Plant operation and design remains unchanged as a result of this UFSAR change.

12

PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: JUNE 1990 l/2-TOP-3044 TEMPORARY OPERATING PROCEDURE (TOP) 06/28/90 This TOP provides instructions for shutdown of the auxiliary ventilation system so that 1-VS-F-8A/88 discharge ductwork can be repaired.

Adequate precautions and guidance are provided in the controlling procedure to closely monitor the charging pump temperatures during the short period of time the auxiliary building general area exhaust fans are secured. If the charging pump temperatures exceed 180°, the auxiliary ventilation system will be restarted and the flexible joint repair stopped. Both trains of the auxiliary ventilation system filtered exhaust system will remain operable and in automatic during this activity. In the event that the filtered exhaust fans start. The gases blown through the opening created by the temporary removal of the flexible joints will be prefiltered by HEPA and charcoal filters and the release will be contained within the auxiliary building. The duration of the removal of the flexible joints will be short.

Therefore, an unreviewed safety question is not created.

13

TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: - -JUNE

- - 1990 1-ST-212 SPECIAL TEST 07 /11/89 This test was to verify/adjust setpoint of main steam safety valves on C steam generator and perform blowdown and seat 1 1 leakage tests.

This test did not create an unreviewed safety question because the main steam system is not required for core cooling at cold shutdown.

l/2-ST-231 SPECIAL TEST 06/21/89 The purpose of this test was to obtain baseline performance data for emergency service water pumps, to perform a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> performance test of 1-SW-P-lA subsequent to refurbishment, and to obtain operational data for use in resolving the SSFI issues associated with the service water house temperatures.

The pump was operated in a normal manner and met the technical specification requirements. Therefore, an unreviewed safety question was not created.

l/2-ST-235 SPECIAL TEST 06/16/89 The purpose of this test was to provide baseline data on the performance of control room chiller equipment and determine condenser performance and cleaning requirements.

This test did not constitute an unreviewed safety question, since the special test only installed temporary test instrumentation for data collection purposes.

1-ST-250 SPECIAL TEST 06/20/89 The purpose of this test was to obtain flow and temperature data from temporary instrumentation by monitoring 1-CH-P-2C seal parameters during boric acid transfer pump operation in order to help determine what is causing the excessive pump seal failures.

This test did not constitute an unreviewed safety question or require a change to technical specification. The special test data is applicable to pump seal parameters on 1-CH-P-2A, B, C and D.

14

e e TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: ~~JU_N_E_l_9_90~~

1-ST-253

  • SPECIAL TEST 06/24/89 The purpose of this test was to stroke test selected motor operated valves (MOV) in the charging and safety injection system under differential pressure. The selected MOVs are CH pump discharge (normal header), CH pump recirculation, and HHSI to cold leg. Also, this test was to stroke test the charging pump discharge MOVs under the 11 piston-effect 11
  • This test did not constitute an unreviewed safety question, nor a change to technical specifications. The technical specification limiting conditions of operation were met prior to commencement of the test.

l/2-ST-255 SPECIAL TEST 06/21/89 The purpose of this test was to obtain baseline data to determine the service water flow through each component cooling water heat exchanger (CCW-HX). This test obtained data necessary to calibrate the annubar installed in the service water inlet piping to each component cooling water heat exchanger.

This test did not constitute an unreviewed safety question nor require a change in technical specifications. The components were operated within their design specifications.

l/2-ST-260 SPECIAL TEST 06/29/89 This test was to verify that when a specified breaker was opened for various components on the vital buses, DC busses, or MCCs, the fans and dampers and the other components connected to that breaker will not operate other than to go to a fail safe condition.

This test verified design basis power supply integrity for safety related components with the unit at a shutdown condition, the safety effects of the breaker and manipulations were reviewed; an unreviewed safety question does not exist.

15

TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: ~~JU_N_E_1_9_90~~

l/2-ST-261 SPECIAL TEST 06/23/89 The purpose of this test was to verify that MS-PCV-102A, Band MS-PCV-202A, Bare able to cycle a designed number of times using the bottled nitrogen source upon loss of the normal instrument air supply. Also, thfs test was to verify that check valves, I-IA-947 and 2-IA-587 which separate the backup bottled nitrogen supply from the normal instrument air supply for MS-PCV-102A, B respectively are able to fully close and hold pressure when experiencing a slow bleed down loss of the normal instrument air supply.

The components were operated as designed. The technical specifications limiting conditions of operation (LCO) were met prior to commencement of test. Therefore, an unreviewed safety question was not created.

1-ST-263 SPECIAL TEST 10/10/89 This test was to obtain trending data for auxiliary feedwater full flow to each steam generator from auxiliary feedwater pumps 1-FW-P-3A, 1-FW-P-38 and 1-FW-P-2 with and without the use of the auxiliary feedwater booster pumps 1-FW-P-4A and 1-FW-P-48.

This test simulates actual auxiliary feedwater operation; an unreviewed safety question is not created.

l/2-ST-268 SPECIAL TEST 06/25/89 This test was to verify that when an identified breaker is open, the component(s) supplied by that breaker will not energize.

This test verified design basis power supply integrity for safety related components. An unreviewed safety question was not created.

1-ST-270 SPECIAL TEST 06/23/89 This test was to individually test an auxiliary feedwater MOV under differential pressure conditions. A MOVATS signature will be taken on the MOV as it is cycled (closed - opened - closed) while an auxiliary feedwater pump is running.

This test did not constitute an unreviewed safety question, nor require a change to technical specifications. The technical specification limiting conditions of operations were met prior to commencement of the test.

16

TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: - -JUNE

--- 1990 1-ST-271 SPECIAL TEST 06/23/89 This test was to stroke test the following valves under differential pressure condition: LSHI to HHSI MOVs, CH pump 8 1 1 suction MOV, and normal CH isolation MDV.

This test did not constitute an unreviewed safety question, nor require a change to technical specifications. The technical specification limiting conditions of operations were met prior to commencement of the test.

1-ST-272 SPECIAL TEST 06/30/89 This test was to verify that when a specified breaker is opened, that the pump receiving its electric power from the breaker will not start. This will verify that pump power supply is as identified and that there are no cross ties to other power supplies.

This test verified design basis power supply integrity for safety related components. Therefore, an unreviewed safety question is not created.

1-ST-273 SPECIAL TEST 06/19/90 This test was to obtain trending data for auxiliary feedwater full flow to each steam generator from auxiliary feedwater pump 1-FW-P-2 with and without the use of the auxiliary feedwater booster pumps 1-FW-P-4A and 1-FW-P-48.

This test simulates actual auxiliary feedwater operation. It does not prevent the system from performing its intended function. An unreviewed safety question does not exist.

l/2-ST-280 SPECIAL TEST 06/28/90 (Safety Evaluation #90-0051)

The purpose of this test was to obtain flow and pressure data by operating the safety related filter exhaust system in a variety of configurations. This will assist in the preparation of an assessment report evaluating the auxiliary ventilation system.

I This test did not constitute an unreviewed safety question since no modification to the controls or the equipment is involved.

The system will remain capable of responding to safety signals for starting and stopping of the fans and positioning of the dampers.

17

VIRGINIA POWER SURRY POWER STATION CHEMISTRY REPORT MONTH/YEAR: JUNE 1990 PRIMARY COOLANT UNIT NO. 1 UNIT NO. 2 ANALYSIS MAX. MIN. AVG. MAX. MIN. AVG.

Gross Radioact., µCi/ml 9.57E-1 1.67E-2 7.33E-1 2.25E-1 1.36E-2 1.73E-1 Suspended Solids, ppm 0.0 0.0 0.0 0.0 0.0 0.0 Gross Tritium, µCi/ml 9.00E-2 2.68E-2 5.57E-2 2.62E-1 1. 28E-1 1. 98E-1 Iodine-131, µCi/ml 9.56E-3 1. 94E-4 6.45E-3 3.12E-3 3.93E-4 7.75E-4 Iodine-131/Iodine-133 0.17 0.09 0.14 0.18 0.07 0.10 Hydrogen, cc/ kg 32.2 22.6 28.0 34.4 25.1 28.9 Lithium, ppm 2.36 1.02 1.37 2.33 2.01 2.16 Boron - 10, ppm* 117 12 34 201 100 119

~0.005 ~0.005 ~0.005  ::!:Q.005 i?Q,005 ~0.005 Oxygen, (DO), ppm 0.004 ~0.001 0.001 0.007 0.004 0.006 Chloride, ppm pH@ 25 degree Celsius 7.50 6.61 7.25 6.94 6.53 6.60

UNIT ONE Hydrogen concentration was out-of-spec (low) from 06/03/90 at 1320 hrs. until 2240 hrs. The hydrogen was 22.6 cc/kg, the limit is 25 cc/kg.

18 L

e UNIT 1&2 FUEL HANDLING DATE: JUNE 1990 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 19

DESCRIPTION OF PERIODIC TEST(S) WHICH WERE NOT COMPLETED WITHIN THE TIME LIMITS SPECIFIED IN TECHNICAL SPECIFICATIONS MONTH/YEAR: ~~JU_N_E_1_9_90~~-

NONE DURING THIS REPORTING PERIOD 20

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