ML18153C331

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Monthly Operating Repts for Jul 1990 for Surry Power Station Units 1 & 2.W/900815 Ltr
ML18153C331
Person / Time
Site: Surry  Dominion icon.png
Issue date: 07/31/1990
From: Stewart W, Warren L
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
90-502, NUDOCS 9008210126
Download: ML18153C331 (21)
v • d • e


Text

'.l e e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 August 15, 1990 U. S. Nuclear Regulatory Commission Serial No.90-502 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 MONTHLY OPERATING REPORT Enclosed is the Monthly Operating Report for Surry Power Station Units 1 and 2 for the month of July 1990.

Very truly yours,

~~SW\-

fl.l..Q.- 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 NRG Senior Resident Inspector Surry Power Station 900A210i26 , ~ - .- - - ~

Ann~K 1

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  • POW 34-04 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION MONTHLY OPERATING REPORT REPORT IJ 90-07 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 16 Fuel Handling - Unit No. 1 17 Fuel Handling - Unit No. 2 17 Description of Periodic Test(s) Which Were Not Completed Within the Time Limits Specified in Technical Specifications 18

e OPERATING DATA REPORT DOCKET NO.: 50-280 DATE: --:0"""'8-/0~3-/9"'""0_ _ _ __

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

1. Unit Name: Surry Unit 1
2. Reporting Period: July 01-31, 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 744.0 5087.0 154343.0
12. Number of Hours Reactor Was Critical 712 .4 4796.7 97547.5
13. Reactor Reserve Shutdown Hours 0 0 3774.5
14. Hours Generator On-Line 707.6 4783.0 95606.2
15. Unit Reserve Shutdown Hours 0 0 3736.2
16. Gross Thermal Energy Generated (MWH) 1691143.0 11511993. 5 222628796.5
17. Gross Electrical Energy Generated (MWH) 555660.0 3865370.0 72410773.0
18. Net Electrical Energy Generated (MWH) 526851.0 3677287. 0 68688217.0
19. Unit Service Factor 95.1% 94% 61.9%
20. Unit Availability Factor 95.1% 94% 64.4%
21. Unit Capacity Factor (Using MDC Net) 90.7% 92.6% 57.5%
22. Unit Capacity Factor (Using DER Net) 89.9% 91. 7% 56.5%
23. Unit Forced Outage Rate 4.9% 6% 21%
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---.-A"""'CH..,...I=E.,...,.VE=D.---

INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 1

e OPERATING DATA REPORT DOCKET NO.: 50-281 DATE: 08/03/90 COMPLETED BY: -L~."""=A-.-;W....,..a-r-re_n_ _ __

TELEPHONE: (804)357-3184 x355 OPERATING STATUS NOTES

1. Unit Name: Surry Unit 2
2. Reporting Period: July 01-31, 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 744.0 5087.0 151224.0
12. Number of Hours Reactor Was Critical 744.0 4941.0 96139.6
13. Reactor Reserve Shutdown Hours 0 0 328.1
14. Hours Generator On-Line 744.0 4930.3 94579.2
15. Unit Reserve Shutdown Hours 0 0 0
16. Gross Thermal Energy Generated (MWH) 1802276.0 11797223.3 221407558.1
17. Gross Electrical Energy Generated {MWH) 588430.0 3933285.0 72013884.0
18. Net Electrical Energy Generated (MWH) 557391. 0 3739386.0 68280345.0
19. Unit Service Factor 100% 96.9% 62.5%
20. Unit Availability Factor 100% 96.9% 62.5%
21. Unit Capacity Factor (Using MDC Net) 95.9% 94.1% 57.9%
22. Unit Capacity Factor (Using DER Net) 95.1% 93.3% 57.3%
23. Unit Forced Outage Rate 0 3.1% 15.1%
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..,..,.VE=D=--

INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 2

DOCKET NO.: 50-280

-=----=,,........-=--==-----

UNIT SHUTDOWN AND POWER REDUCTION UNIT NAME: Surry Unit 1 (Equal To or Greater Than 20%) DATE: 08/03/90 COMPLETED BY: L.A. Warren REPORT MONTH: - -JULY

- -1990

TELEPHONE: 804-357-3184 x355 METHOD OF LICENSEE CAUSE &CORRECTIVE DURATION SHUTTING EVENT SYSTEM COMPONENT ACTION TO PREVENT DATE TYPE(l) (HOURS) REASON(2) DOWN REACTOR(J) REPORT# CODE(4) CODE(5) RECURRENCE 07/01/90 F 36.4 A 2 1-90-006 LO DRY At 1802 hours0.0209 days <br />0.501 hours <br />0.00298 weeks <br />6.85661e-4 months <br />, a turbine runback was initiated from

!RPI dropped rod signal t I

resulted from the trip of the 1

AI reserve station service transformer. At 1807 hours0.0209 days <br />0.502 hours <br />0.00299 weeks <br />6.875635e-4 months <br />, a manual reactor trip was initiated due to a loss of instrument air causing the 1 C1 main steam trip valve to close.

The instrument air dryer was bypassed and the instrument air pressure was restored two minutes post trip.

e (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 I - Same Source

DOCKET NO.: 50-281

--..---....-----=---

UNIT SHUTDOWN AND POWER REDUCTION UN IT NAME: Surry Un1t 2 (Equal To or Greater Than 20%) DATE: 08/03/90 COMPLETED BY: L.A. Warren REPORT MONTH: - -JULY

- -1990

TELEPHONE: 804-357-3184 x355 METHOD OF LICENSEE CAUSE &CORRECTIVE DURATION SHUTTING EVENT SYSTEM COMPONENT ACTION TO PREVENT DATE TYPE(l) (HOURS) REASON(2) DOWN REACTOR(3) REPORT# C0DE(4) CODE(5) RECURRENCE 07/06/90 S 0 B 4 TA V Ramped down to perform main turbine valve 11 freedom movement 11 test 2-PT-29.1.

e 07/29/90 S 0 B 4 TA V Ramped down to perform main turbine valve 11 freedom of movement 11 test 2-PT-29.1, also to clean waterboxes.

(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: 08/03/90 COMPLETED BY: L.A. Warren TELEPHONE:(804)357-3184 x355 MONTH: JULY 1990 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 582 17 763 2 0 18 763 3 317 19 760 4 758 20 759 5 757 21 757 6 759 22 759 7 760 23 756 8 772 24 758 9 767 25 751 10 764 26 716 11 765 27 740 12 765 28 709 13 765 29 721 14 765 30 710 15 765 31 705 16 764 NOTE:

Listed above for each day in the reporting month is the average daily unit power level in MWe-Net computed to the nearest whole megawatt.

5

AVERAGE DAILY UNIT POWER LEVEL

~------

DOCKET NO.: 50-281 UNIT NAME: Surry Unit 2 DATE: 08/03/90 COMPLETED BY: L.A. Warren TELEPHONE:(804)357-3184 x355 MONTH: JULY 1990 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 763 17 757 2 765 18 756 3 761 19 754 4 758 20 754 5 755 21 753 6 678 22 752 7 761 23 749 8 762 24 749 9 760 25 749 10 757 26 748 11 757 27 748 12 754 28 744 13 758 29 649 14 758 30 752 15 761 31 746 16 759 NOTE:

Listed above for each day in the reporting month is the average daily unit power level in MWe-Net computed to the nearest whole megawatt.

6

SUMMARY

OF OPERATING EXPERIENCE MONTH/YEAR: JULY 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 07/01/90 0000 This reporting period started with the Unit operating at 100% power, 815 MW.

1802 Turbine runback initiated due to IRPI dropped rod signal due to the trip of the 'A' reserve station service transformer; 100% power, 815 MW.

1807 Manual reactor trip due to the loss of instrument air pressure.

07/03/90 0145 Reactor critical.

0630 Unit on line and ramping up.

0715 Stopped ramp for chemistry hold; 30% power, 170 MW.

0927 Started ramp up after chemistry hold was released; 30%

power, 170 MW.

2240 Stopped ramp; 99.4% power, 795 MW.

07/26/90 0816 Started ramp down to perform 1-PT-29.1; 99% power, 785 MW.

0915 Stopped ramp; 85% power, 645 1220 Started ramp up after 1-PT-29.1 completed; 85% power, 640 MW.

1530 Stopped ramp; 99% power, 780 MW.

07/27/90 0800 Started slow power reduction; Unit placed on excess letdown making the deborating demineralizers unavailable for use.

Core is at end of life, 99% power, 780 MW.

07/28/90 2000 Stopped power reduction and started ramp up; normal letdown and deborating demineralizer placed in service; 92% power, 730 MW.

2200 Stopped ramp while flux mapping was in progress; 95.5%

power, 750 MW.

7

SUMMARY

OF OPERATING EXPERIENCE MONTH/YEAR: ~-JU_L_Y_1_9_90~~~

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 07/29/90 1356 Reduced power to 94.4%; as per Reactor Engineers, power must remain below 95% due to FQ problem.

07/31/90 0210 As per Reactor Engineers, maximum power allowed equals 100%;

maintaining 94.3% power, 745 MW while depleted deborating demineralizer is replaced.

2400 This reporting period ended with the Unit operating at 94%

power, 740 MW.

8

SUMMARY

OF OPERATING EXPERIENCE MONTH/YEAR: JULY 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 TWO 07/01/90 0000 This reporting period started with the Unit operating at 100% power, 800 MW.

07/06/90 0043 Started ramp down to clean waterboxes; 99.8% power, 795 MW.

0125 Stopped ramp; 89% power; 700 MW.

0407 Started ramp up; 89% power, 710 MW.

0455 Stopped ramp; 99.8% power, 780 MW.

0931 Started ramp down to perform 2-PT-29.1; 99.8% power, 805 MW.

1225 Stopped ramp; 74.2% power, 605 MW.

1736 Started ramp up after completion of 2-PT-29.1; 74% power, 600 MW.

1920 Stopped ramp; 99% power, 800 MW.

07/29/90 0631 Started ramp down to perform 2-PT-29.1 and clean all waterboxes; 100% power, 785 MW.

0835 Stopped ramp; 75% power, 590 MW.

1011 Started ramp down to maintain condenser vacuum while cleaning waterboxes; 74.8% power, 585 MW.

1050 Stopped ramp; 67% power, 510 MW.

1350 Started ramp up after waterbox cleaning operations and 2-PT-29.1 completed; 65% power, 500 MW.

1918 Stopped ramp; 100% power, 795 MW.

07/31/90 2400 This reporting period ended with the Unit operating at 100%

power, 785 MW.

9

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

--- 1990 DC 86-10 SERVICE WATER AND CIRCULATING WATER BUTTERFLY VALVE REPLACEMENT 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 are 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 SW to CC HX 1-SW-25,29,33,37 CC HX SW inlets The following valves have been replaced which completes 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 The replacement valves and expansion joints are one for one replacement of existing equipment and are designed to specification 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.

AC-Sl-90-0629 ADMINISTRATIVE CONTROL 07/05/90 (Safety Evaluation #90-0167)

The missile plug above the 1 A1 main steam non-return valve was removed to permit maintenance to be performed inside the steam side of Unit 2 safeguards. Administrative control was assigned to reinstall the plug in the event of severe weather warnings.

Personnel will be standing by with the tools and equipment necessary to quickly reinstall the missile shield in the event of severe weather warnings (hurricane, tornado, or severe thunderstorm watch/warning). An unreviewed safety question is not created.

10

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

- -JULY

- -1990 1/2-EWR-89-518 ENGINEERING WORK REQUEST 07 /11/90 This request removed the gravel and installed a new single ply membrane roofing on the turbine building.

This change involved repairs to the turbine building roof. An unreviewed safety question does not exist.

C-S2-90-0718 ADMINISTRATIVE CONTROL 07 /18/90 (Safety Evaluation #90-0175)

This change ensures that containment integrity can be maintained in accordance with the definition of containment integrity in Technical Specification 1.0 with the outside Instrument Air supply manual isolation valves open.

Administrative control over 2-IA-703 and 2-IA-704 in accordance with SUADM-0-026 ensures the ability to maintain containment integrity as defined in Technical Specification.

The valves will be closed upon notification by the main control room. In addition, control over the valves will ensure the ability to .maintain the containment partial pressure within the limits established by T.S. 3.8.

Therefore, an unreviewed safety question is not created.

AC-Sl-90-0727 ADMINISTRATIVE CONTROL 07/27/90 (Safety Evaluation #90-0187)

This change ensures that containment integrity can be maintained with the manual letdown leakage monitoring valve opened intermittently.

Administrative control over 1-CH-440 ensures the ability to maintain containment integrity as defined in Technical Specifications. The valve will be closed immediately upon notification by the main control room. Therefore, an unreviewed safety question is not created.

11

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

--- 1990 SPS 1, CYCLE 10 MEMO DATED 07/27/90, SPS CYCLE 10 TEMPERATURE/POWER COASTDOWN OPERATING AT REDUCED NOMINAL OPERATING PRESSURE 07/27/90 (Safety Evaluation #90-0188)

This change determined the impact on continued operation for the remainder of Surry 1, Cycle lOA at a reduced nominal operating pressure of 2135 psig. The pressure will be reduced to attempt to reseat the pressurizer safety valve 1 A1 which appears to be experiencing a small amount of reactor coolant leakage.

A review of the applicable accident analyses show that the operation of Surry Unit 1 Cycle lOA in a temperature coastdown mode at a reduced reactor coolant system operating pressure of 2135 psig is acceptable. Current accident analysis margins are conserved. Changes to protection setpoints or other compensatory measures are not required to support implementation of the pressure reduction. The same conclusions can be drawn for an extended pressure reduction of 150 psi (down to 2085 psig) provided the hi~h pressurizer pressure reactor trip setpoint is reduced by 50 psi (from 2385 psig to 2335 psig). The extension of operation into the power coast region is also covered by this evaluation in that thermal margins will continue to increase as temperature and power are reduced.

12

PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: JULY 1990 l/2-0P-4.22 OPERATING PROCEDURE 07/06/90 (Safety Evaluation #90-0169)

This change secured and restarted the spent fuel pit cooling pump. The cooling pump is secured to prevent handling cask and fuel in turbulent water flow.

Fuel pit cooling removes the heat from the pool. To secure the cooling pump will result in an increase in temperature but will allow easier handling of the cask and fuel. The heatup of the pool and additional surveillance is bounded by the UFSAR. Therefore, an unreviewed safety question does not exist.

1/2-0P-23.2.4 OPERATING PROCEDURE 07/18/90 *

(Safety Evaluation #90-0176)

This change deleted the requirements in the operating procedure for a 60 day hold up of the 1 8 1 waste gas decay tank (WGDT) under normal evolutions. The 1 8 1 WGDT is presently at approximately 100 PSIG and has been in holdup for eight days.

The WGDT is provided to accommodate the volume of gasses generated during normal operations of both reactor units and is designed under normal operation for a 30 day filling period followed by a 20 day holdup period and a 10 day bleed period.

Procedural controls normally provide additional holdup time of at least 45 days to allow decay time for the accumulated gasses. Due to the recent large amounts of gas generated by the preparations for maintenance on the gas stripper system, the 1 8 1 WGDT will be released with only eight days of holdup.

Radioactive gaseous release permits have been approved by the Health Physics Department for this release. The total activity in the tank is approximately 0.2 µCi/ml. The primary gaseous constituent being XE-133. This volume of activity is much less than that assumed in the UFSAR to be released upon rupture of the WGDT with one percent failed fuel. Further, the release rates are to be restricted to three standard cubic feet per minute (SCFM) and the release will be continually monitored and can be immediately terminated, if required.

The calculations performed at the above release rate show that the total activity released and the site boundary doses are much less than that allowed by Technical Specifications.

Therefore, an unreviewed safety question is not created.

13

e e TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: ~~JU_L_Y_l....:.9...:...90_;:__~

2-ST-213 SPECIAL TEST 08/02/89 This test verified/adjusted the setpoints of the valves listed below.

VALVE SETPOINT PLUS 1% BAND ORIFICE SIZE SV-MS-201A,B,C 1085 psig 1074-1095 psig 6 SV-MS-202A,B,C 1095 psig 1084-1105 psig R SV-MS-203A,B,C 1110 psig 1098-1121 psig R SV-MS-204A,B,C 1120 psig 1108-1131 psi g R SV-MS-205A,B,C 1135 psi g 1123-1146 psi g R This test did not create an unreviewed safety question because the main steam system is not required for core cooling at cold shutdown.

2-ST-223 SPECIAL TEST 08/22/89 This test was performed to prove the operability of each inside recirculation spray pump to perform its intended function by measuring flow through a special recirculation test loop. This test also established reference values for inside recirculation spray pump testing in accordance with ASME Section XI, Subsection

!WP.

The test was performed at cold shutdown conditions utilizing equipment installed specifically for the test. The pumps are not required to be operable below 350°F/450 psig. Following the test, the system piping was returned to a normal configuration.

The test was performed at cold shutdown and the system returned to a normal configuration following the test. An unreviewed safety question was not created.

2-ST-240 SPECIAL TEST 09/11/89 The purpose of this test was to functionally test the sequencing of loads onto Emergency Bus 2H following the injection of an engineered safeguards feature (ESF) signal. An undervoltage signal was injected five minutes after the initial ESF signal to verify that the required loads are tripped off the emergency bus and that they are re-sequenced on the emergency bus in the required order.

An unreviewed safety question was not created since one train of the main control room air bottle system was available for automatic actuation on Unit 1 safety injection; the other train was manually isolated and under administrative control to be opened within one minute of a Unit 1 safety injection. One train of air bottles is sufficient to pressurize the main control room/relay room boundary for one hour following a design basis accident.

14

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

- -1990 2-ST-249 SPECIAL TEST 09/03/89 This test was performed to measure the combined leakage of service water header isolation valves 2-SW-MOV-202A/B. This leakage value was combined with the leakage values for the circulating water valves, and other service water valves to quantify the total leakage flow for the station.

The test was performed with the unit in cold shutdown conditions and the Technical Specification Limiting Conditions of Operation were met prior to commencement of the test. No safety related equipment or equipment important to safety utilize service water supplied through the piping were isolated during the test. An unreviewed safety question was not created.

2-ST-254 SPECIAL TEST 08/02/89 This test was performed to verify that when the identified breaker is open on vital buses, DC panels, or MCC s, that the 1

valves shown to be connected to that breaker will not cycle.

The test was performed at cold shutdown conditions and the Limiting Conditions for Operation were satisfied prior to commencement of the test. The probability of accidents analyzed in the FSAR were not increased, and the possibility of new accidents was not created. Redundant equipment was operable during the test. An unreviewed safety question was not created.

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

This test simulates auxiliary feedwater flow operation and did not prevent the system from performing its intended function. An unreviewed safety question was not created.

2-ST-273 SPECIAL TEST 02/06/90 This test was performed to obtain trending data for auxiliary feedwater full flow to each steam generator for auxiliary feedwater pump 2-FW-P-2 with and without the use of auxiliary feedwater booster pumps 2-FW-P-4A and 2-FW-P-48.

This test simulates auxiliary feedwater flow operation and did not prevent the system from performing its intended function. An unreviewed safety question was not created.

15

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

Gross Radioact., µCi/ml 1.30E+O 1. 51E-1 8.78E-1 2.79E-1 1.58E-1 2.16E-1 Suspended Solids, ppm 0.0 0.0 0.0 0.0 0.0 0.0 Gross Tritium, µCi/ml 7. 71E-2 3.20E-2 6.41E-2 3.05E-1 2.42E-1 2.91E-1 Iodine-131, µCi/ml 2.91E-1 6.43E-3 2.98E-2 1.12E-3 3.35E-4 6.39E-4 Iodine-131/Iodine-133 0.19 0.14 0.16 0.17 0.06 0.10 Hydrogen, cc/kg 36.0 16.8 28.6 37.0 25.2 26.7 Lithium, ppm 2.23 0.75 0.98 2.28 2.06 2.12 Boron - 10, ppm* 64.5 0.80 11. 7 101.9 86.2 93.7 Oxygen, (DO), ppm ~0.005 ~0.005 ~0.005 ~0.005 ~0.005 ~0.005 Chloride, ppm 0.003 0.001 0.001 0.006 0.004 0.005 pH@ 25 degree Celsius 8.82 6.99 7.86 7.04 6.83 6.93

UNIT ONE: On 07/04/90 at 0810, the reactor coolant system (RCS) dissolved hydrogen concentration went out-of-spec low, with the concentration being 21.6cc/kg. Hydrogen was low due to isolation of hydrogen at the manifold. Dissolved hydrogen was back in spec at 1630 on 07/05/90.

Total out-of-spec hours: 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> and 20 minutes. On 07/28/90 at 0500, the RCS dissolved hydrogen concentration went out-of-spec low, with the concentration being 17.2cc/kg. Hydrogen was low due to letdown isolation. Dissolved hydrogen was back in spec at 2220 on 07/28/90.

Total out-of-spec hours: 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> and 20 minutes.

16

UNIT 1&2 e

FUEL HANDLING

  • DATE: JULY 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 17

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

-- 1990 NONE DURING THIS REPORTING PERIOD 18

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