ML18152A456

Monthly Operating Repts for Nov 1993 for Surry Power Station Unit 1 & 2.W/931210 Ltr

ML18152A456
Person / Time
Site:	Surry
Issue date:	11/30/1993
From:	Bowling M, Mason D VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
93-764, NUDOCS 9312200020
Download: ML18152A456 (22)

Text

e e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 December 10, 1993 U. S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, D. C. 20555 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 MONTHLY OPERATING REPORT Serial No. NO/RPC:vlh Docket Nos. License Nos.93-764 50-280 50-281 DPR-32 DPR-37 Enclosed is the Monthly Operating Report for Surry Power Station Units 1 and 2 for the month of November 1993. Very truly yours, /f1L_;J~ M. L. Bowling, Manager Nuclear Licensing

& Programs Enclosure cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W. Suite 2900 Atlanta, Georgia 30323 Mr. M. W. Branch NRC Senior Resident Inspector Surry Power Station ---------i-,J 9312200020 931130 '. i PDR ADOCK 05000280 / *:*1

• R __ -----__ ___ __ PDR *---___ lj

' f VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION MONTHLY OPERATING REPORT REPORT NO. 93-11 Approved:

TABLE OF CONTENTS Section esurry Monthly Operating Report No. 93-11 Page 2of 21 Page Operating Data Report -Unit No. 1 ...*....****......*.**...............****.***.......*...*...............**...**....*...........**........

3 Operating Data Report -Unit No. 2 .......................................*...***.......***....*.....................*......................

4 Unit Shutdowns and Power Reductions

-Unit No. 1 .......*.................*****....**...*...............................*............

5 Unit Shutdowns and Power Reductions

-Unit No. 2 ...*.**......................................*....**....*...........................

6 Average Daily Unit Power Level -Unit No. 1 **.........*.*****..*.............****............................**.........................*

7 Average Daily Unit Power Level -Unit No. 2 ................................*.*.......**...............................**.................

8 Sum niary of Operating Experience

-Unit No. 1 ...........*******.*..........*.****......**...**..**...*.............**...***...........

9 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 ......................*..***.*.*...........

16 Tests and Experiments That Did Not Require NRC Approval **.........********....*.**............................***............

17 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

......****......*******.*..*...........******.......***...*............*....*..**..**....*.........

21

-Surry Monthly Operating Report No. 93-11 Page3of21 OPERATING DATA REPORT Docket No.: Date: Completed By: 50-280 12-03-93 D. Mason Telephone:

(804) 365-2459 1. Unit Name: .................................................. . 2. Reporting Period: ........................................

.. 3. Licensed Thermal Power (MWt): ...................... . 4. Nameplate Rating (Gross MWe): ...................... . 5. Design Electrical Rating (Net MWe): ................

.. 6. Maximum Dependable Capacity (Gross MWe): .. .. 7. Maximum Dependable Capacity (Net MWe): ...... .. Surry Unit 1 November, 1993 2441 847.5 788 820 781 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. Reasons For Restrictions, If Any: 11. Hours In Reporting Period .............

..........

.. 12. Number of Hours Reactor Was Critical ......... . 13. Reactor Reserve Shutdown Hours .............. . 14. Hours Generator On-Line .......................... . 15. Unit Reserve Shutdown Hours .................... . 16. Gross Thermal Energy Generated (MWH) ..... . 17. Gross Electrical Energy Generated (MWH) ... . 18. Net Electrical Energy Generated (MWH) ...... .. 19. Unit Service Factor .................................. . 20. Unit Availability Factor .............................. . 21. Unit Capacity Factor (Using MDC Net) .......... . 22. Unit Capacity Factor (Using DER Net) .......... . 23. Unit Forced Outage Rate ........................... . This Month 720.0 . 720.0 0 720.0 0 1613872.5 546345.0 523920.0 100.0o/o 100.0o/o 93.2% 92.3% 0.0% YID 8016.0 7688.2 0 7659.6 0 18243389.9 6078120.0 5809506.0 95.6% 95.6% 92.8% 92.0o/o 1.6% 24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each): Refueling (1 O Year ISi), January 21, 1994, 64 Days 25. If Shut Down at End of Report Period, Estimated Date of Start-up:

26. Unit In Test Status (Prior to Commercial Operation):

Cumulative 183576.0 123063.2 3774.5 120935.0 3736.2 281862669.0 92096373.0 87407366.0 65.9% 67.9% 61.4% 60.4% 17.5% FORECAST INITIAL CRITICALITY ACHIEVED INITIAL ELECTRICITY COMMERCIAL OPERATION OPERATING DATA REPORT -Surry Monthly Operating Report No. 93-11 Page 4of 21 Docket No.: 50-281 12-03-93 D. Mason Date: Completed By: Telephone:

(804) 365-2459 1. Unit Name: .................................................. . 2. Reporting Period: ............... , ......................... . 3. Licensed Thermal Power (MWt): ...................... . 4. Nameplate Rating (Gross MWe): ...................... . 5. Design Electrical Rating (Net MWe): ................. . 6. Maximum Dependable Capacity (Gross MWe): ... . 7. Maximum Dependable Capacity (Net MWe): ....... . Surry Unit2 November, 1993 2441 847.5 788 820 781 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): 1 O. Reasons For Restrictions, If Any: This Month 11. Hours In Reporting Period ......................... . 12. Number of Hours Reactor Was Critical.

........ . 13. Reactor Reserve Shutdown Hours .............. . 14. Hours Generator On-Line .......................... . 15. Unit Reserve Shutdown Hours .................... . 16. Gross Thermal Energy Generated (MWH) ..... . 17. Gross Electrical Energy Generated (MWH) ... . 18. Net Electrical Energy Generated (MWH) ....... . 19. Unit Service Factor .................................. . 20. Unit Availability Factor .............................. . 21. Unit Capacity Factor (Using MDC Net) .......... . 22. Unit Capacity Factor (Using DER Net) .......... . 23. Unit Forced Outage Rate ........................... . 720.0 356.3 0 356.3 0 836286.8 279895.0 269148.0 49.5% 49.5% 47.9% 47.4% 6.3% YID 8016.0 5650.9 0 5554.5 0 12643799.3 4177975.0 3990461.0 69.3% 69.3% 63.7% 63.2% 10.6% 24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each): None Cumulative 180456.0 119337.8 328.1 117485.5 0 273974873.1 89373879.0 84780874.0 65.1% 65.1% 60.3% 59.6% 14.2% 25. If Shut Down at End of Report Period, Estimated Date of Start-up:

December 1, 1993 26. Unit In Test Status (Prior to Commercial Operation):

FORECAST INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION ACHIEVED (1) -Surry Monthly Operating Report No. 93-11 Page Sot 21 UNIT SHUTDOWN AND POWER REDUCTION (EQUAL To OR GREATER THAN 20%) REPORT MONTH: November, 1993 (2) (3) (4) (5) Docket No.: 50-280 Unit Name: Surry Unit 1 Date: 12-01-93 Completed by: Craig Olsen Telephone:

(804) 365-2155 Duration Method of Reason Shutting Down Rx LEA No. System Component Cause & Corrective Action to Date Type Hours Code Code Prevent Recurrence (1) F: Forced S: Scheduled (4) (2) REASON: None during this reporting period. A -Equipment Failure (Explain)

B Maintenance or Test C Refueling D Ragulatory Restriction E Operator Training & Licensing Examination F Administrative G Operational Error {Explain)

Exhibit G -Instructions for Preparation of Data Entry Sheets for Licensee Event Report {LEA) File {NUREG 0161) (3) METHOD: 1 -Manual 2 Manual Scram. 3 -Automatic Scram. 4 -Other {Explain)

(5) Exhibit 1 -Same Source.

-Surry Monthly Operating Report No. 93-11 Page 6of 21 UNIT SHUTDOWN AND POWER REDUCTION (EQUAL To OR GREATER THAN 20%) REPORT MONTH: November, 1993 Docket No.: 50-281 Unit Name: Surry Unit 2 Date: 12-01-93 Completed by: Craig Olsen TeleQhone:

(804} 365-2155 (1) (2) (3) (4) (5) Method Duration of LER System Component Cause & Corrective Action to Date Type Hours 931115 F 24 931116 s 339.7 (1) F: Forced S: Scheduled Reason A B (2) REASON: Shutting Down Rx 3 4 No. 2-93-006-00 NIA A -Equipment Failure (Explain)

B Maintenance or Test C Refueling D Regulatory Restriction Code Code JB BKR JB SG E Operator Training & Licensing Examination F Administrative G Operational Error (Explain)

(4) Exhibit G -Instructions for Preparation of Data Entry Sheets for Licensee Event Report (LER) File (NUREG 0161) Prevent Recurrence Automatic reactor trip occurred on low steam generator level when the main feedwater regulating valves failed closed due to a loss of electrical power to the solenoid operated trip valves. Breaker was replaced.

Unit shutdown that occurred on 11 /15/93 was continued to perform steam generator maintenance.

(3) METHOD: 1 -Manual 2 -Manual Scram. 3 -Automatic Scram. 4 -Other (Explain)

(5) Exhibit 1 -Same Source.

esurry Monthly Operating Report No. 93-11 Page 7of 21 AVERAGE DAILY UNIT POWER LEVEL Docket No.: 50-280 Unit Name: Surry Unit 1 Date: 12-03-93 Completed by: Pat Kessler Telephone:

365-2790 MONTH: November, 1993 Average Daily Power Level Average Daily Power Level Day (MWe-Net) Day (MWe-Net) 1 793 17 724 2 790 18 722 3 790 19 713 4 789 20 711 5 779 21 699 6 772 22 694 7 765 23 691 8 764 24 690 9 762 25 680 10 756 26 675 11 751 27 672 12 741 28 665 13 736 29 661 14 738 30 656 15 729 16 724 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.

e Surry Monthly Operating Report No. 93-11 Page 8of 21 AVERAGE DAILY UNIT POWER LEVEL Docket No.: 50-281 Unit Name: Surry Unit 2 Date: 12-03-93 Completed by: Pat Kessler Telephone:

365-2790 MONTH: November, 1993 Average Daily Power Level Average Daily Power Level Day (MWe-Net) Day (MWe-Net) 1 762 17 0 2 762 18 0 3 761 19 0 4 752 20 0 5 765 21 0 6 765 22 0 7 766 23 0 8 761 24 0 9 758 25 0 10 754 26 0 11 746 27 0 12 741 28 0 13 748 29 0 14 745 30 0 15 630 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.

I I. ' e Surry Monthly Operating Report No. 93-11 Page 9of 17

SUMMARY

OF OPERATING EXPERIENCE MONTH/YEAR:

November, 1993 The following 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.

UNrr ONE: 11/01/93 11/04/93 11/30/93 UN[T Two.:. 11/01/93 11/10/93 11/10/93 11/15/93 11/30/93 0000 1039 2400 0000 1214 1228 2019 2400 The reporting period began with the Unit operating at 100% power, 820 MWe. Started power coastdown (due to fuel depletion) in order to maintain RCS temperature at setpoint.

The reporting period ended with the Unit operating at 83% power, 693 MWe, in a power coastdown.

The reporting period began with the Unit operating at 98.5% power, 790 MWe, due to Steam Generator "C" level oscillations.

Started power reduction to mitigate Steam Generator "C" level oscillations.

Stopped power reduction at 95%, 780 MWe. Automatic reactor trip occurred on low steam generator level when the main feedwater regulating valves failed closed due to a loss of electrical power to the solenoid operated trip valves. The reporting period ended with the Unit preparing to startup.

JCO C-92-002 EWR 88-002 EWR 88-048 JCO C-93-004 e Surry Monthly Operating Report No. 93-11 Page 10 of 21 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

November, 1993 Justification For Continued Operation (Safety Evaluation No.92-075, Revision 1) 11-01-93 Justification For Continued Operation C-92-002 assessed the acceptability of manual operator action (through annunciator response procedures) in lieu of automatic equipment functions to ensure the operation of radiation monitoring systems in the event of a loss of power. The assessment concluded that the actions required to mitigate a loss of power to the radiation monitoring system can be satisfactorily accomplished through the actions contained in the annunciator response procedures.

Therefore, an unreviewed safety question does not exist. Engineering Work Request 11-02-93 Engineering Work Request 88-002 replaced the electrical cables to Unit 1 reactor coolant pump 1-RC-P-1A and 1-RC-P-1C motors to improve the reliability of the power supply. This nonsafety-related modification improved the reliability of the subject pumps and did not adversely affect any safety-related equipment, system, or the margin of safety. Therefore, an unreviewed safety question does not exist. Engineering Work Request 11-02-93 Engineering Work Request 88-048 replaced the electrical cables to Unit 2 reactor coolant pump 2-RC-P-1A, 2-RC-P-1 B, and 2-RC-P-1C motors to improve the reliability of the power supply. This nonsafety-related modification improved the reliability of the subject pumps and did not adversely affect any safety-related equipment, system, or the margin of safety. Therefore, an unreviewed safety question does not exist. Justification for Continued Operation (Safety Evaluation No.93-197, Revision 1) 11-05-93 Justification for Continued Operation C-93-004 assessed continued operation with through wall leakage of certain Charging Pump Service Water system valves. The assessment concluded, based on the design rating of the valves, the small amount of leakage, and a seismic and structural integrity evaluation, that the subject system remained capable of performing its intended design function.

The system will be walked down on a weekly basis to ensure any further degradation is promptly identified.

Therefore, an unreviewed safety question does not exist.

AC 81-93-1108 EWR 90-253 EWR 88-576 TM S2-93-44 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

November, 1993 Administrative Control (Safety Evaluation No.93-204) -Surry Monthly Operating Report No. 93-11 Page 11 of 21 11-08-93 Administrative control of Unit 1 Ventilation system valve, 1-VS-251, was established to provide an alternate (for the Unit 2 Bearing Cooling Water system) make-up water source from the Central Chilled Water system to the Control and Relay Room Chiller system. This change was instituted to facilitate the installation of the Mechanical Equipment Room No. 5 make-up line to the Unit 2 Bearing Cooling Water system. The administrative controls limited the opening of 1-VS-251 and required that the valve be immediately closed upon a loss of water or seismic event. These controls ensured that adequate chilled water volume was maintained in the . Control and Relay Room Chiller system and did not affect the system's performance.

Therefore, an unreviewed safety question did not exist. Engineering Work Request (Safety Evaluation No.91-050) 11-10-93 Engineering Work Request 90-253 installed a pre-engineered structure to enclose contaminated materials storage pad "D". The modification decreased the potential for personnel exposure to contaminated materials and improved the reliability of contaminated material control. The structure is supplied power from an off-site source and is physically separated from plant operating systems. This change did not adversely impact any system or the margin of safety. Therefore, an unreviewed safety question does not exist. Engineering Work Request 11-15-93 Engineering Work Request 88-576 replaced three emergency lighting units with three different types of emergency lighting units for testing purposes.

The test units were installed to facilitate the identification of failure mechanisms and suitable replacement units. The modification will potentially improve the reliability of 10 CFR 50, Appendix R required lighting.

The test units are self contained and do not adversely affect any safety-related equipment, system, or the margin of safety. Therefore, an unreviewed safety question does not exist. Temporary Modification (Safety Evaluation No.93-206) . 11-15-93 Temporary Modification (TM) S2-93-44 installed a coupling and nipple in place of the leaking Unit 2 Condensate Polishing system sample line isolation valve, 2-CP-250, until a replacement can be procured.

The TM was leak tested prior to use and allowed resin regeneration operations to be continued.

The Condensate Polishing system is nonsafety-related and its operation was not affected.

Therefore, an unreviewed safety question did not exist.

EWA 90-333 FS92-118 EWR 90-272 FS 93-23 e Surry Monthly Operating Report No. 93-11 Page 12of 21 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

November, 1993 Engineering Work Request (Safety Evaluation No.90-262) 11-16-93 Engineering Work Request 90-333 implemented the necessary modifications to enable the Unit 2 recirculating spray heat exchanger (RSHX) service water supply piping to be maintained in a chemically treated wet lay-up condition.

The modifications were made to the nonsafety-related portion of the Service Water (SW) system in order to eliminate the potential for plugging the RSHX tubes with marine growth upon operation of the Recirculating Spray system. The changes improve the reliability of the SW system supply to the RSHXs during a design basis accident.

Therefore, an unreviewed safety question does not exist. UFSAR Change 11-16-93 (Safety Evaluation 93-207) Updated Final Safety Analysis Report Change 92-118 revised Sections 10.3.3, "Turbine Generator", and 14.2, "Core and Coolant Boundary Protection Analysis", to delete the discussion of the automatic turbine startup function (which was never installed) and correct the description of the turbine overspeed protection system and the turbine overspeed design value. These changes were administrative in nature and made to accurately reflect the current plant configuration and to correct the turbine overspeed design value. No procedures or plant equipment were affected and no physical modifications were involved.

Therefore, an unreviewed safety question does not exist. Engineering Work Request (Safety Evaluation No.90-216) 11-18-93 Engineering Work Request 90-272 implemented minor modifications (i.e., spring pack, gear, and motor replacements) to miscellaneous Unit 1 and 2 motor operated valves (MOVs). The modifications were made to improve the reliability and capability of the subject MOVs to overcome maximum differential system pressure.

The changes did not exceed the allowable torque ratings or other design limitations of the valves. Therefore, an unreviewed safety question does not exist. UFSAR Change 11-18-93 (Safety Evaluation 93-211) Updated Final Safety Analysis Report Change 93-23 revised Section 8.5, "Emergency Power System", to clarify the section and remove excessive detail. This change was administrative and editorial in nature. No procedures or plant equipment were affected and no physical modifications were involved.

Therefore, an unreviewed safety question does not exist.

SE 93-214 DCP 93-79-3 SE 93-216 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

November, 1993 esurry Monthly Operating Report No. 93-11 Page 13of 21 Operational Quality Assurance Program Topical Report 11-22-93 (Safety Evaluation No.93-212) Operational Quality Assurance Program Topical Report Sections 17.2.1.2, **"Management of Operational Quality Assurance" and 17.2.18, "Audits", were revised to reflect a transfer in the responsibility for the internal audit program from the Manager Quality Assurance (Corporate) to the Manager Quality Assurance (Station) and to clarify the difference between internal and external audits. The changes were of an administrative or editorial nature. The changes do not reflect or affect changes to the plant and do not impact the margin of safety. Therefore, an unreviewed safety question does not exist. Safety Evaluation 11-23-93 Safety Evaluation 93-214 was performed to evaluate the use of the TIP/CECOR computer code package for flux map analysis in lieu of the currently used INCORE code. This change was made in order to accurately analyze flux maps for Unit 1 Cycle 13, which will utilize axially zoned flux suppression inserts for control of vessel neutron fluence. The TIP/CECOR computer code package was evaluated and determined to be an equivalent replacement for the INCORE code. This change does not affect plant operation, existing accident analyses, or the margin of safety. Therefore, an unreviewed safety question does not exist. Design Change Package (Safety Evaluation No.93-217) 11-24-93 Design Change Package 93-79-3 replaced 125 volt, 15 amp rated DC breaker Nos. 14 and 16 in the 2-EPD-DB-DC2-1 panel and Nos. 14 and 15 in the DB-DC2-2 panel with like 20 amp rated breakers.

The change in tf1e current rating of the replacement breakers was analyzed and determined to be acceptable.

The modification was performed with the Unit at cold shutdown when the 125 volt DC electrical system was not required to be operable.

Therefore, an unreviewed safety question does not exist. Safety Evaluation 11-24-93 Safety Evaluation 93-216 was performed to evaluate the use of Westinghouse Corporation's Pressure Pulse Cleaning Method to remove sludge deposits from the Unit 1 and 2 steam generators (SG). The Units will be at cold shutdown when the cleaning is performed.

Evaluations of this cleaning method were performed by Virginia Power and Westinghouse Nuclear Safety. The evaluations concluded that the integrity of the SGs is not affected by pressure pulse cleaning.

Therefore, an unreviewed safety question does not exist.

SE 93-218 SE 93-219 -e Surry Monthly Operating Report No. 93-11 Page 14of 21 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

November, 1993 10 CFR 50, Appendix R Report (Safety Evaluation 93-215) 11-24-93 The 10 CFR 50, Appendix R Report was revised to incorporate completed plant modifications that affect the Appendix R program and additional/revised engineering evaluations.

• *

• The plant modifications were evaluated independently, with respect to Appendix R compliance, prior to implementation as part of the design change process. The additional engineering evaluations were performed in accordance with Generic Letter 86-10. The revised engineering evaluations assess the adequacy of some current plant configurations.

These changes document commitments which form part of the licensing basis for fire protection for Units 1 and 2. No physical changes to the plant were made. Therefore an unreviewed safety question does not exist. Safety Evaluation 11-28-93 Safety Evaluation 93-218 was performed to evaluate the operation of Unit 2 steam generators (SG) with tube support plate broached quatrefoil holes that are partially blocked by sludge deposits.

This condition was identified by Westinghouse personnel during an inspection of Unit 2 SG "C" following pressure pulse cleaning and sludge lancing operations.

Attempts to remove the deposits were only partially successful.

Evaluations of this condition were performed by Virginia Power and Westinghouse Nuclear Safety. The evaluations concluded that the existing licensing basis analysis results and conclusions remain bounding for Unit 2 operation for all events except the loss of normal feedwater.

A reassessment of a loss of feedwater event confirmed that the applicable acceptance criteria will continue to be met for operation with this condition.

Therefore, an unreviewed safety question does not exist. Safety Evaluation 11-28-93 Safety Evaluation 93-219 was performed to evaluate the presence of five small foreign objects in the secondary side of Unit 2 steam generator (SG) "C". The objects were identified by Westinghouse personnel during a foreign objects search and retrieval (FOSAR) following pressure pulse cleaning and sludge lancing operations on the SG. Attempts to remove the objects were unsuccessful.

An evaluation of .Unit 2 operation with the SG containing the foreign objects was performed by Westinghouse Nuclear Safety. The evaluation concluded that the presence of the objects is not expected to have an adverse effect on the primary pressure boundary integrity of the SG during the remainder of the current fuel cycle. Therefore, an unreviewed safety question does not exist.

SE 93-221 FS 93-35 -Surry Monthly Operating Report No. 93-11 Page 15of 21 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

November, 1993 Safety Evaluation 11-29-93 Safety Evaluation 93-221 was performed, as a result of a review of Information Notice 91-40, to evaluate the potential for unmonitored, uncontrolled radioactivity releases to the environment from the Auxiliary Steam system drains and vents (from the auxiliary building).

The evaluation concluded that contamination from a leaking pipe or spill to the floor drains would be required for a release to the discharge canal to occur. In the event of such a leak or spill, minimal circulating/service water system flow would dilute contamination levels to well below the maximum release levels. Periodic sampling is performed to monitor system contamination levels to ensure a release can be adequately diluted. Therefore, an unll'eviewed safety question does not exist. UFSAR Change 11-30-93 (Safety Evaluation 93-223) Updated Final Safety Analysis Report Change 93-35 revised Figure 9.11-1, "Well Water System", to reflect the removal of the Water Supply and Treatment system hypochlorite storage tank, 1-WT-TK-10, and pump, 1-WT-P-20.

This change was administrative in nature and made to accurately reflect the current plant configuration.

No physical changes were made to the station. Therefore, an unreviewed safety question does not exist.

1-FDTP-90-08-3-1 1-IPT-FW-FCV-001 2-IPT-FW-FCV-001 1-IPT-FW-HCV-155 2-IPT-FW-HCV-255 1-GOP-2.4 2-GOP-2.4 e e Surry Monthly Operating Report No. 93-11 Page 16 of 21 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

November, 1993 Final Design Test Procedure (Safety Evaluation No.93-208) 11-17-93 Final Design Test Procedure 1-FDTP-90-08-3-1 was developed to provide instructions for performing a functional test and verification of the Mechanical Equipment Room No. 5 control room chillers following the completion of Design Change Package 90-08. The test will not impact the existing control room chillers or any other system. The new MER No. 5 chillers are not required for and do not impact any accident analyses.

Therefore, an unreviewed safety question does not exist. Instrument Periodic Test Procedures (Safety Evaluation No.93-213) 11-23-93 Instrument Periodic Test Procedures 1-IPT-FW-FCV-001 and 2-IPT-FW-FCV-001, "Stroke Time Testing of Main Feedwater Regulating Valves" and HCV-155 and 2-IPT-FW-HCV-255, "Stroke Time Testing of Main Feedwater Bypass Valves" were developed to perform stroke testing of the main and bypass feedwater regulating valves to assure compliance with the lnservice Test Program requirements.

The procedures will be performed when the Units are at cold shutdown or refueling shutdown during which the feedwater isolation function of the subject valves is not required.

Therefore, an unreviewed safety question does not exist. General Operating Procedures (Safety Evaluation No.93-224) 11-30-93 General Operating Procedures 1-GOP-2.4 and 2-GOP-2.4, "Unit Shutdown, RCS Cooldown From HSD to 345 °F -350 °F", were revised to provide instructions for installing electrical jumpers to defeat the delta-T and valve position interlocks of the Reactor Coolant system loop stop valves. This procedurally controlled temporary modification enables the subject valves to be removed from their backseats following a reactor trip in order to prevent potential valve damage that may occur during the cooldown process. The procedural controls ensure that the valves will not be closed. Valve movement is limited to moving from the backseat.

Double verification of jumper installation/removal and post maintenance testing will be performed.

Therefore, an unreviewed safety question does not exist.

2-ST-306 i____ __ **---* e Surry Monthly Operating Report No. 93-11 Page 17of 21 TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

November, 1993 Special Test 11-09-93 (Safety Evaluation 93-196) Special Test 2-ST-306, "Steam Generator C Level Control Program Adjustment", was performed on 10-20-93 to determine if increasing the water level in the Unit 2 Steam Generator (SG) "C" would dampen level. oscillations that were being experienced above 95% power. The test allowed the level to be increased from 44% to a maximum of 53%. The test was terminated prior to completion due to unacceptable SG level oscillations that occurred at the 49% level during a ramp up to 100% power. The upper program water level control setpoint for the SG was returned to the pretest value. This test did not affect the reliability or integrity of any equipment affecting the safety analyses.

Therefore, an unreviewed safety question did not exist.

Primarv Coolant Analvsis Gross Radioactivitv, uCi/ml Susoended Solids oom Gross Tritium, uCi/ml 1131, uCi/ml 1131/1133 Hvdroaen.

cc/ka Lithium, oom Boron -10, oom* Oxygen, (DO), ppm Chloride, ppm oH at 25 dearee Celsius * . Boron-10 = Total Boron x 0.196 Comments:

None CHEMISTRY REPORT MONTH/YEAR:

November, 1993 Unit No. 1 Max. Min. Ava. 4.82E-1 2.52E-1 3.33E-1 S0.1 < 0.1 < 0.1 5.25E-2 2.03E-2 3.49E-2 2.09E-3 1.20E-3 1.67E-3 0.18 0.09 0.14 39.8 30.1 36.5 0.82 0.62 0.76 1.8 0.2 0.3 .s,0.005 S0.005 .s,0.005 .s,0.05 S0.001 0.022 9.28 8.04 8.99 e Surry Monthly Operating Report No. 93-11 Page 18of 21 Unit No. 2 Max. Min. Ava. 2.23E-1 9.10E-4 6.10E-2 < 0.1 < 0.1 < 0.1 6.34E-1 5.90E-1 6.12E-1 1.39E-4 2.82E-5 8.97E-5 0.16 0.07 0.12 38.9 17.5 30.5 2.33 1.61 2.02 378.5 219.3 309.3 .s,0.005 S0.005 .s,0.005 .s,0.05 0.007 0.026 6.42 5.76 6.12 New or Spent Fuel Shipment Date Stored or Number Received Unit 1 Batch 15 Shipment 1 11/16/93 Unit 1 Batch 15 Shipment 2 11/18/93 FUEL HANDLING UNITS 1 & 2 MONTH/YEAR:

November, 1993 Number of Assemblies Assembly ANSI per Shipment Number Number 12 OK1 LMOYK9 OK3 LMOYKB OK9 LMOYKH 1KO LMOYKJ 1K2 LMOYKL 1KB LMOYKS 2KO LMOYKU 2K1 LMOYKV 2K2 LMOYKW 2K3 LMOYKX 2K4 LMOYKY 2K7 LMOYL1 12 OK4 LMOYKC OKS LMOYKD OKS LMOYKE OK7 LMOYKF OKS LMOYKG 1K1 LMOYKK 1K3 LMOYKM 1K4 LMOYKN 1K5 LMOYKP 1K6 LMOYKQ 8 Surry Monthly Operating Report No. 93-11 Page 19of 21 New or Spent Initial Fuel Shipping Enrichment Cask Activity 3.81790 3.81890 3.81960 3.81880 3.81670 3.81480 3.81850 3.81790 3.81360 3.80810 3.81770 3.82040 3.81820 3.81900 3.82080 3.81840 3.81970 3.81850 3.81790 3.81570 3.81110 3.81600

... e FUEL HANDLING UNITS 1 & 2 MONTH/YEAR:

November, 1993 New or Spent Fuel Shipment Date Stored or Number Received Unit 1 Batch 15 Shipment2 11/18/93 Unit 1 Batch 15 Shipment 3 . 11/30/93 Number of Assemblies per Shipment 12 12 Assembly Number 1K7 1K9 3K2 3K3 3K4 3K5 3K6 3K8 3K9 4K1 4K2 5K5 5K6 5K7 ANSI Number LMOYKR LMOYKT LMOYL6 LMOYL7 LMOYL8 LMOYL9 LMOYLA LMOYLC LMOYLD LMOYLF LMOYLG LMOYLV LMOYLW LMOYLX e Surry Monthly Operating Report No. 93-11 Page 20of 21 Initial Enrichment 3.81900 3.81170 3.81380 3.99330 3.98990 3.99520 3.99100 3.99640 3.99470 3.99260 3.99070 4.00420 3.98640 3.99430 New or Spent Fuel Shipping Cask Activity e e Surry Monthly Operating Report No. 93-11 Page 21 of 21 DESCRIPTION OF PERIODIC TEST(S) WHICH WERE NOT COMPLETED WITHIN THE TIME LIMITS SPECIFIED IN TECHNICAL SPECIFICATIONS MONTH/YEAR:

November, 1993 None during this reporting period. l