ML18153A442

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Monthly Operating Repts for Aug 1997 for Surry Power Station,Units 1 & 2.W/970915 Ltr
ML18153A442
Person / Time
Site: Surry  Dominion icon.png
Issue date: 08/31/1997
From: Fanguy M, Mason D, Mccarthy J
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
97-521, NUDOCS 9709230090
Download: ML18153A442 (23)
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Text

  • VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 September 15, 1997 United States Nuclear Regulatory Commission Serial No.97-521 Attention: Document Control Desk NUJDK/MAE RO 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 The Monthly Operating Report for Surry Power Station Units 1 and 2 for the month of August 1997 is provided in the attachment.

If you have any questions or require additional information, please contact us.

Very truly yours,

] ~ h y , Manager

/~~~Li~ksing and Operations Su Attachment Commitments made by this letter: None cc: U. S. Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth Street, S. W.

Suite 23T85 Atlanta, Georgia 30303 Mr. R. A. Musser NRC Senior Resident Inspector Surry Power Station 9709230090 970831 IIIIIII IIIIIHlll lllll 111111111111111111 PDR ADOCK 05000280 *&&62F7*:

I R PDR

  • Surry Monthly Operating Report No. 97-08 Page 1 of 22 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION MONTHLY OPERATING REPORT REPORT No. 97-08 Approved:
  • TABLE OF CONTENTS

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9surry Monthly Operating Report No. 97-08 Page 2 of 22 Section 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 Summary of Operating Experience - Unit Nos. 1 and 2 ............................................................................................. 9 Facility Changes That Did Not Require NRC Approval ............................................................................................ 10 Procedure or Method of Operation Changes That Did Not Require NRC Approval ................................................. 15 Tests and Experiments That Did Not Require NRC Approval .................................................................................. 18 Chemistry Report .....................................................................................................................................................19 Fuel Handling - Unit Nos. 1 and 2 ............................................................................................................................20 Description of Periodic Test(s) Which Were Not Completed Within the Time Limits Specified in Technical Specifications ....................................................................................................................... 22

  • OPERATING DATA REPORT I

ltsurry Monthly Operating Report No. 97-08 Page 3 of 22 Docket No.: 50-280 Date: 09/1/97 Completed By: D. K. Mason Telephone: (757) 365-2459

1. Unit Name: ........................................................... . Surry Unit 1
2. Reporting Period: ................................................. . August, 1997
3. Licensed Thermal Power (MWt): ......................... . 2546
4. Nameplate Rating (Gross MWe): ......................... . 847.5
5. Design Electrical Rating (Net MWe): ................... . 788
6. Maximum Dependable Capacity (Gross MWe): ... . 840
7. Maximum Dependable Capacity (Net MWe): ....... . 801
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):

1O. Reasons For Restrictions, If Any:

This Month Year-To-Date Cumulative

11. Hours in Reporting Period 744.0 5831.0 216455.0
12. Number of Hours Reactor Was Critical 744.0 4262.1 151096.8
13. Reactor Reserve Shutdown Hours 0.0 0.0 3774.5
14. Hours Generator On-Line 744.0 4139.5 148670.5
15. Unit Reserve Shutdown Hours 0.0 0.0 3736.2
16. Gross Thermal Energy Generated (MWH) 1877848.2 10164497.5 348799941.3
17. Gross Electrical EnergyGenerated (MWH) 617542.0 3364745.0 114337563.0
18. Net Electrical Energy Generated (MWH) 595807.0 3245381.0 108839130.0
19. Unit Service Factor 100.0% 71.0% 68.7%
20. Unit Availability Factor 100.0% 71.0% 70.4%
21. Unit Capacity Factor (Using MDC Net) 100.0% 69.5% 62.8%
22. Unit Capacity Factor (Using DER Net) 101.6% 70.6% 63.8%
23. Unit Forced Outage Rate 0.0% 6.5% 15.0%
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 Start-up:
26. Unit In Test Status (Prior to Commercial Operation):

FORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION

  • OPERATING DATA REPORT I

4ltsurry Monthly Operating Report No. 97-08 Page 4 of 22 Docket No.: 50-281 Date: 09/01/97 Completed By: D. K. Mason Telephone: (757) 365-2459

1. Unit Name: ............................................................ Surry Unit 2
2. Reporting Period: ................................................. . August, 1997
3. Licensed Thermal Power (MWt): ......................... . 2546
4. Nameplate Rating (Gross MWe): ......................... . 847.5
5. Design Electrical Rating (Net MWe): ................... . 788
6. Maximum Dependable Capacity (Gross MWe): ... . 840
7. Maximum Dependable Capacity (Net MWe): ....... . 801
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:

This Month Year-To-Date Cumulative

11. Hours in Reporting Period 744.0 5831.0 213335.0
12. Number of Hours Reactor Was Critical 744.0 5767.0 148842.6
13. Reactor Reserve Shutdown Hours 0.0 0.0 328.1
14. Hours Generator On-Line 744.0 5760.1 146857.9
15. Unit Reserve Shutdown Hours 0.0 0.0 0.0
16. Gross Thermal Energy Generated (MWH) 1890618.9 14496295.2 345970552.0
17. Gross Electrical Energy Generated (MWH) 621329.0 4818019.0 113268818.0
18. Net Electrical Energy Generated (MWH) 599580.0 4656416.0 107848295.0
19. Unit Service Factor 100.0% 98.8% 68.8%
20. Unit Availability Factor 100.0% 98.8% 68.8%
21. Unit Capacity Factor (Using MDC Net) 100.6% 99.7% 63.1%
22. Unit Capacity Factor (Using DER Net) 102.3% 101.3% 64.2%
23. Unit Forced Outage Rate 0.0% 1.2% 12.1%
24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):

Refueling, October 6, 1997, 30 Days

25. . If Shut Down at End of Report Period, Estimated Date of Start-up:
26. Unit In Test Status (Prior to Commercial Operation):

FORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION

  • UNIT SHUTDOWN AND POWER REDUCTION (EQUAL To OR GREATER THAN 20%)

aurry Monthly Operating Report No. 97-08 Page 5 of 22 REPORT MONTH: August, 1997 Docket No.: 50-280 Unit Name: Surry Unit 1 Date: 09-03-97 Completed by: M. J. Fanguy Telephone: (757) 365-2155 (1) (2) (3) (4) (5)

Method Duration of LER No. System Component Cause & Corrective Action Date Type Hours Reason Shutting Code Code to Prevent Recurrence Down Rx 8/21/97 s NA B NA NA TA V Power reduction for turbine valve freedom test (1) (2) (3)

F: Forced REASON: METHOD:

S: Scheduled A - Equipment Failure (Explain) 1 - Manual B - Maintenance or Test 2 - Manual Scram C - Refueling 3 - Automatic Scram D - Regulatory Restriction 4 - Other (Explain)

E - Operator Training & Licensing Examination F - Administrative G - Operational Error (Explain)

(4) (5)

Exhibit G - Instructions for Preparation of Data Entry Sheets Exhibit 1 - Same Source for Licensee Event Report (LER) File (NU REG 0161)

__J

  • UNIT SHUTDOWN AND POWER REDUCTION (EQUAL TO OR GREATER THAN 20%)

.Surry Monthly Operating Report No. 97-08 Page 6 of 22 REPORT MONTH: August, 1997 Docket No.: 50-281 Unit Name: Surry Unit 2 Date: 09-03-97 Completed by: M. J. Fanguy Telephone: (757) 365-2155 (1) (2) (3) (4) (5)

Method Duration of LEA No. System Component Cause & Corrective Action Date Type Hours Reason Shutting Code Code to Prevent Recurrence Down Rx None During the Reporting Period (1) (2) (3)

F: Forced REASON: METHOD:

S: Scheduled A - Equipment Failure (Explain) 1 - Manual B Maintenance or Test 2 - Manual Scram C Refueling 3 - Automatic Scram D Regulatory Restriction 4 - Other (Explain)

E Operator Training & Licensing Examination F Administrative G Operational Error (Explain)

(4) (5)

Exhibit G - Instructions for Preparation of Data Entry Sheets Exhibit 1 - Same Source for Licensee Event Report (LER) File (NU REG 0161)

esurry Monthly Operating Report No. 97-08 Page 7 of 22 AVERAGE DAILY UNIT POWER LEVEL Docket No.: 50-280 Unit Name: Surry Unit 1 Date: 09-04-97 Completed by: J. D. Kilmer Telephone: (757) 365-2792 MONTH: August, 1997 Average Daily Power Level Average Daily Power Level Day (MWe - Net) Day (MWe - Net) 1 811 17 804 2 802 18 804 3 805 19 805 4 812 20 810 5 813 21 572 6 813 22 805 7 813 23 809 8 812 24 808 9 812 25 806 10 813 26 808 11 811 27 809 12 811 28 807 13 810 29 805 14 809 30 805 15 808 31 806 16 806 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 esurry Monthly Operating Report No. 97-08 Page 8 of 22 AVERAGE DAILY UNIT POWER LEVEL Docket No.: 50-281 Unit Name: Surry Unit 2 Date: 09-04-97 Completed by: John D. Kilmer Telephone: (757) 365-2792 MONTH: August, 1997 Average Daily Power Level Average Daily Power Level Day (MWe - Net) Day (MWe - Net) 1 812 17 796 2 815 18 798 3 812 19 799 4 809 20 785 5 810 21 805 6 813 22 806 7 813 23 785 8 812 24 762 9 812 25 810 10 813 26 816 11 810 27 814 12 808 28 814 13 806 29 813 14 805 30 814 15 801 31 815 16 799 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.

-Surry Monthly Operating Report No. 97-08 Page 9 of 22

SUMMARY

OF OPERATING EXPERIENCE MONTHNEAR: August, 1997 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.

UNIT ONE:

8/1/97 0000 Unit 1 begins the month at 100% / 840 MWe.

8/21 /97 0655 Start power decrease from 100% / 840 MWe to perform turbine valve freedom test 1105 Stop power decrease at 73%.

1122 Start power decrease to 50% to repair "C" intercept valve due to it failing to reopen following testing.

1401 Stopped power decrease at 51%.

2121 Start power increase after repair of "C" intercept valve 8/22/97 0142 Stop power increase at 100% / 837 MWe.

8/31/97 2400 Unit 1 finishes the month at 100% /835 MWe.

UNITTwo:

8/1/97 0000 Unit 2 starts the month at 100% / 845 MWe.

8/23/97 0711 Adjusting turbine load to maintain condenser vacuum with "C" waterbox being removed from service for cleaning.

1845 Start power increase from 96.5% I 790 MWe after "C" waterbox is returned to service.

1945 Stop power increase at 100% / 845 MWe.

8/24/97 0438 Adjusting turbine load to maintain condenser vacuum with "A" waterbox being removed from service for cleaning.

1620 Start power increase from 90% / 745 MWe after "A" waterbox is returned to service.

1756 Stop power increase at 100% / 845 MWe.

8/31/97 2400 Unit 2 finishes the month at 100% / 845 MWe.

e Surry Monthly Operating Report No. 97-08 Page 10 of 22 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTHNEAR: August, 1997 TM S-97-013 Temporary Modification 8-4-97 (Safety Evaluation No.97-096)

This Temporary Modification (TM) installs tygon tubing in place of flexible stainless steel tubing for the Emergency Service Water Pump (ESWP) flow indicators. This will allow troubleshooting of the oscillating flow indicator by visually inspecting the tubing to ensure the arrangement allows for proper venting of all air.

The currently installed tubing is 1/2 inch diameter Swagelok corrugated stainless tubing.

The flow element and tubing are normally isolated from ESWP piping by a three inch ball valve. The flow element and instrumentation are only placed in service for pump testing.

Normal operating pump discharge pressure is 13 psig with a system design pressure of 25 psig. The tygon tubing will be verified leak tight when placed in service. Should the tubing fail, the resultant leakage, a few gpm, is insignificant when compared to the pump design requirement of 14,020 gpm. Additionally the tygon tube can be isolated and will not affect the accuracy of the flow indication. Therefore an unreviewed safety question does not exist.

TM 82-97-07 Temporary Modification 8-4-97 (Safety Evaluation No.97-100)

This Temporary Modification (TM) installs a temporary blower to provide additional ventilation to Main Steam Radiation Monitor 2-MS-RM-225. The monitor has had a history of failure that has been associated with the ambient temperature in which it operates. The manufacturer's recommended normal operating range for this device is 32 to 120 °F. The contact temperature on the radiation monitor shield box is about 150 °F.

The required performance characteristics of the radiation monitor will not be altered by improving the cooling of the shield enclosure. No air flow will be diverted away from safety related components. The 60 °F minimum temperature for the Aux Feedwater Pump lubricating oil will be met by securing the blower when the ambient temperature in the Main Steam Valve House reaches 70 °F. The blower will be secured to an existing structure to prevent movement during operation and a seismic event. Therefore, an unreviewed safety question does not exist.

e Surry Monthly Operating Report No. 97-08 Page 11 of 22 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: August, 1997 JCO S1-97-002 Justification For Continued Operation 8-11-97 (Safety Evaluation No.97-091 Rev. 1)

The Justification for Continued Operation (JCO) discusses continued Unit 1 operation with a containment hydrogen concentration limit of 0.8%. The evaluation supporting this change is documented in Engineering Transmittal (ET) NAF-970185, Rev. 0. The ET contains the following limiting conditions: 1) continued operation of the containment upper dome air fan and pressurizer cubicle ventilation while at power; 2) a hydrogen sampling program which confirms bulk concentrations do not exceed the 0.8% limit. These conditions will be tracked in accordance with JCO S1-97-002.

The results of Unit 1 containment sampling data have shown that the hydrogen concentration is well mixed from the containment basement to the operating floor.

Adhering to the 0.8% hydrogen concentration limit provides high confidence that the accident analysis concentration limit will remain less than the 4% lower flammability limit.

A sampling program will be maintained to confirm that hydrogen concentrations do not exceed the 0.8% value. The initial presence of 0.8% hydrogen in the containment will have a negligible impact on the containment peak pressure, depressurization and pump NPSH analyses. The self imposed limit of 0.8% is bounded by the conservatisms assumed in the post LOCA hydrogen analysis which include the one day hydrogen recombiner start time and the 50 scfm flowrate through the recombiner. Therefore an unreviewed safety question does not exist.

JCO S1-97-005 Justification For Continued Operation 8-11-97 (Safety Evaluation No.97-099)

The Justification for Continued Operation (JCO) provides the justification of conditions and requirements for Letdown System operability. As previously identified, valve weights used in the piping seismic analyses were incorrect and non-conservative. The seismic pipe stress analysis and representative vibration analysis was performed with the revised valve weights. As documented by Engineering Transmittal, the piping stress was within code allowables, with a valve support exceeding allowables. Modifications have been made to the valve support to satisfy ASME Ill Appendix F criteria to allow continued operation until the next refueling outage in accordance with NRG Generic Letter 91-18.

This degraded and non-conforming condition has been reviewed and analyzed. The previous letdown operational restrictions and support modifications made provide reasonable assurance that the letdown piping system and supports meet the structural integrity requirements outlined in Generic Letter 91-18 Section 6.13. Therefore, an unreviewed safety question does not exist.

FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL e Surry Monthly Operating Report No. 97-08 Page 12 of 22 MONTHNEAR: August, 1997 FS95-034 Updated Final Safety Analysis Report Change 8-14-97 (Safety Evaluation 97-101)

Updated Final Safety Analysis Report (UFSAR) Change FS95-034 revises UFSAR Section 14.3.1, Steam Generator Tube Rupture, because the current wording implies that operator action can accomplish the Steam Generator Tube Rupture (SGTR) procedure steps through the Safety Injection termination and initiation of post-SGTR cooldown within 30 minutes. In most cases this is impractical, as shown by simulator drills, and is not a requirement of the currently applicable analysis.

The UFSAR update clarifies the requirements imposed on operator action for coping with a SGTR accident. No plant hardware, systems, setpoints, procedures or operational methods are being impacted. Existing abnormal and emergency procedures remain unchanged. The SGTR analysis methodology, assumptions, results and conclusions are not changed. Therefore, an unreviewed safety question does not exist.

SE 97-104 Safety Evaluation 8-14-97 Safety Evaluation 97-104 was performed to evaluate equipment that has been Out of Service for greater than 30 days. Technical Specification related Heat Trace circuits 2A1-24, 26, 40, 43 and 281-28 have been out of service greater than 30 days. These circuits are on the boron injection flow path of record.

Redundant heat trace circuits maintain the associated boric acid solution temperature greater than Technical Specification limits to prevent solidification of boric acid. This is verified by Auxiliary Building logs. A Main Control Room annunciator alarms whenever the temperature of the heat trace circuits are not within required parameters. Therefore, an unreviewed safety question does not exist.

SE 97-0051 Safety Evaluation 8-14-97 Safety Evaluation 97-0051 was performed for the Surry ISFSI Safety Analysis Report.

During fabrication of TN-32 spent fuel storage casks, carbon steel shims are inserted between the inner vessel flange and the outer gamma shield during fabrication of the casks. The shims are required because of a small circumferential gap between the flange and the gamma shield forms during the fabrication process. The use of the shims was not part of the cask design addressed in the TN-32 Topical Safety Analysis Report.

Transnuclear has evaluated the impact of the shims on the design of the TN-32 cask and found that they do not affect the strength of the weld, have no effect on the subcritical margin of stored fuel and have no affect on the TN-32 thermal analysis. Therefore, an unreviewed safety question does not exist.

FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL e Surry Monthly Operating Report No. 97-08 Page 13 of 22 MONTH/YEAR: August, 1997 FS93-033 Updated Final Safety Analysis Report Change 8-14-97 (Safety Evaluation 97-107)

Updated Final Safety Analysis Report (UFSAR) Change FS93-033 revises chapter 8.6 by deleting the paragraph which reads "A preventative maintenance program conducted by VEPCO includes the periodic testing of insulation values of circuits and equipment."

Virginia Power Engineering has determined that there is no technical basis for requiring a preventative maintenance program that periodically tests the insulation resistance of all circuits and equipment. Insulation resistance testing is a potentially destructive test which could actually have some adverse effects on the life of a component. A Licensing or Engineering basis for a preventative maintenance program which performs periodic insulation resistance testing on all circuits/equipment other than the UFSAR was not found.

The UFSAR request will clarify the UFSAR, chapter 8.6 with no procedures or other documents being affected. Therefore, an unreviewed safety question does not exist.

TM 81-97-14 Temporary Modification 8-18-97 (Safety Evaluation No.97-105)

This Temporary Modification (TM} installs a Passive Autocatalytic Recombiner to control the presence of hydrogen in the Unit 1 containment.

This activity will not affect the Updated Final Safety Analysis Report (UFSAR) hydrogen generation analysis. The post-LOCA DBA hydrogen generation analysis does not exceed the 4% flammability limit with an assumed 50 cfm electric recombiner operating. This activity does not alter the installed accident electric recombiner capacity. However, operation of a temporary recombiner could decrease the consequences of malfunctions associated with a failure of the installed electric recombiners. Therefore, an unreviewed safety question does not exist.

FS96-061 Updated Final Safety Analysis Report Change 8-21-97 (Safety Evaluation 97-109)

Updated Final Safety Analysis Report (UFSAR) FS96-061 reformats and revises Section 5.3, Containment Systems and 9.13, Auxiliary Ventilation System. The Ventilation sections of the Surry UFSAR require updating to reflect the equipment, systems and system operation changes. Where appropriate, the Ventilation Systems have been regrouped and clarification of equipment served added. The Post-TMI Technical Support Center and Local Emergency Offsite Facility systems have been added. The testing of charcoal from the safety related filters has been changed to agree with Technical Specification Change 325.

The UFSAR descriptions are being altered to reflect actual station operation and system design. Therefore, an unreviewed safety question does not exist.

I, FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL

- Surry Monthly Operating Report No. 97-08 Page 14 of 22 MONTH/YEAR: August, 1997 FS97-035 Updated Final Safety Analysis Report Change 8-21-97 (Safety Evaluation 97-110)

Updated Final Safety Analysis Report (UFSAR) Change FS97-035 revises section 11.2 Radioactive Waste Systems, because an implied 35 mph wind speed restriction on process vent downwash is in error. Downwash concerns on the process vent have been analyzed by Environmental Policy and Compliance. The assessment used an EPA recommended model for nobel gas effluent concentrations referenced in the UFSAR.

The implied 35 mph wind speed limitation was verified to be unsubstantiated. Wind speeds to 67 mph were evaluated.

Dose rates from downwash intuitively rise as ambient wind speed approaches zero. The assessment evaluated atmospheric Stability Classes A thru F. No violations or compromising or encroaching of the limits could happen. The radioactive gaseous effluent concentrations used in the analysis are bounding because they are much larger than what is released. Nominal Waste Gas Decay Tank releases involve 100 curies or less. The dispersion modeled by the downwash assessment coupled with the low concentrations results in excess margin. Therefore, an unreviewed safety question does not exist.

SE97-113 Safety Evaluation 8-25-97 Safety Evaluation 97-113 was performed to determine the acceptability of the tree heights with respect to the meteorological tower and the resulting data. The UFSAR states the tree heights are 40 to 50 feet at the primary site and 10 to 15 feet at the secondary site.

However, the tree heights are estimated to be 60 to 80 feet high. EPA guidance recommends that the measurement instruments should be placed on a tower in open terrain. Currently, Surry is not in compliance with this EPA document and the UFSAR.

The effect of different tree heights as stated in the UFSAR would result in wind speeds and atmospheric stability that are not representative of the actual conditions in the surrounding area. The calculated offsite doses from the met tower data would be conservative in the event of an accident. The met tower is not designated as a safety related component or system. Therefore, an unreviewed safety question does not exist.

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  • PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL e Surry Monthly Operating Report No. 97-08 Page 15 of 22 I

MONTH/YEAR: August, 1997 O-FCA-1.00 Fire Contingency Action Procedure (Safety Evaluation 97-093 Rev. 1) 8-11-97 (Safety Evaluation 97-093 Rev. 2) 8-14-97 Two separate areas of Appendix R non-compliance have been identified with respect to the 120 Volt Vital Bus System. The first issue deals with a fire in the main control room with no means to isolate the affected Vital Bus panel from the associated Uninterruptable Power Supplies (UPS). This postulated event could potentially affect the associated UPS and, therefore, affect the downstream Appendix R Panel. Revision 1 of the Safety Evaluation clarified the power supply for the excore flux monitor. Secondly, Chapter 9 of the Appendix R Report indicates that proper selective tripping is required for faults on Vital Bus branch circuits; a high fault current could result in the main breaker tripping simultaneously with, or in lieu of, the branch circuit breaker.

Compensatory measures implemented in the FCAs will a) disconnect the feeder conductors routed to the control room at UPSs 1A1, 1A2, 2A1, and 2A2 for a main control room fire to ensure the availability of distribution panels located in the ESGR 1 and 2 fire areas; Revision 1 included fuse replacements as necessary to ensure UPS restoration and b) restore Control Room or ESGR Vital Bus distribution panels lost as a result of fire induced hot shorts and mis-coordination between the main and branch circuit breakers as needed to accomplish safe shutdown. Revision 2 clarified that the compensatory measures can be accomplished within the Appendix R Report timeline. These compensatory measures do not adversely impact the Class 1E electrical distribution system. Therefore, an unreviewed safety question does not exist.

O-OP-HG-001 Operating Procedure 8/11/97 (Safety Evaluation 97-097)

Operating Procedure O-OP-HG-001, "Unit 1 and 2 Main Generator Gas Sampling," is being revised to include the installation of procedurally controlled temporary modifications to facilitate calibration and analysis of the main generator hydrogen gas cooling system. This activity involves the use of temporary plastic tubing connections to facilitate safe collection of main generator hydrogen gas samples for collection and analysis.

The major consequence is the potential ignition of the hydrogen gas sample flow. The potential for uncontrolled ignition is minimized by utilizing the limited flow rate capability of the existing analyzer sample tubing, routing the sample purge flow to an external vent line, and stationing an operator by the sample valve for immediate isolation. Existing fire suppression controls in the area of the activity are designed for such an event. This activity does not affect the ability of the station to achieve and maintain safe shutdown under any conditions. Therefore, an unreviewed safety question does not exist.

  • PROCEDURE OR METHOD OF OPERATION CHANGES

- Surry Monthly Operating Report No. 97-08 Page 16 of 22 THAT DID NOT REQUIRE NRC APPROVAL MONTHNEAR: August, 1997 2-NPT-RX-002 Engineering Periodic Test Procedure 8-14-97 2-0PT-RX-006 Operations Periodic Test Procedure (Safety Evaluation 97-103)

Engineering Periodic Test Procedure 2-NPT-RX-002, "Reactor Core Flux Maps," and Operations Periodic Test Procedure 2-0PT-RX-006, "Rod Position Verification Using lncore Flux Mapping System," are being revised to provide a temporary modification which provides steps for jumpering out the safety limit switch on the "E" incore detector for Unit 2 which is not functioning properly. This switch prevents use of the "E" detector for the performance of a flux map.

Technical Specifications require a flux map be performed on a monthly basis and when conditions, such as a dropped control rod, require a flux map to measure core power distribution. The incore instrumentation system does not perform any safety related functions. This activity directly affects the operation of the moveable incore instrumentation system, as it jumpers out the safety limit switch for the "E" incore detector. The operation of the incore instrumentation system is not adversely impacted as a result of this activity, as cautions will be placed in the procedure to warn the performer to monitor the "E" detector while withdrawal in case of an additional component failure. Therefore, an unreviewed safety question does not exist.

1/2-0PT-RX-001 Operations Periodic Test Procedures 8-20-97 1/2-0PT-RX-002 (Safety Evaluation 96-108) 1/2-0PT-RX-003 1/2-0PT-RX-004 These procedures perform Reactor Power Calorimetrics using various methods. This change modifies the specified tolerance between Nuclear Instrumentation System (NIS) and calorimetric power from +2%/-0% to +4%/-0%. This limits the potential for downward NIS Gain adjustments at reduced power preventing the unintentional decalibration and loss of accuracy for the NIS channels when calorimetric alignment is done at reduced power.

The Safety Analysis Report (SAR) does not describe the adjustment of the power range drawers to match the calorimetric, nor does it describe any tolerance associated with the adjustment. This activity will change the tolerance to +4%/-0% between the power range detectors and the power calorimetric. As such, the change in the tolerance ensures the NIS will continue to operate the way it is described in the SAR, as it will ensure continued overpower protection. Therefore, an unreviewed safety question does not exist.

2-MOP-DG-001 Maintenance Operating Procedure 8-21-97 (Safety Evaluation 97-111)

Maintenance Operating Procedure 2-MOP-DG-001, "Removal and Return to Service of the PDTI for Maintenance" is being revised to install an instrument air jumper by-passing the SOV for the Primary Drain Transfer Tank (POTT) Pressure Control Valve, keeping it in the open position. The valve must be maintained open to provide a flowpath for the RCS loop drains and other drains to the PDTI.

The jumper will be in place while Unit 2 is at Cold Shutdown (CSD) or Refueling Shutdown (RSD) and the Reactor Coolant System and containment are at atmospheric pressure; therefore there will be little or no pressure differential to cause gas flow from the PDTI to containment. The installation of this jumper does not affect containment isolation capability nor compliance of Technical Specification required systems. Therefore, an unreviewed safety question does not exist.

  • PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL

- Surry Monthly Operating Report No. 97-08 Page 17 of 22 MONTHNEAR: August, 1997 O-ECM-1509-03 Electrical Corrective Maintenance Procedure 8-29-97 (Safety Evaluation 97-114)

Electrical Corrective Maintenance Procedure O-ECM-1509-03, "Votes MOV Testing for Quarter Turn Valves," incorporates a One Time Only change to administratively control Component Cooling Water Heat Exchanger (CCHX) Service Water (SW) Supply MOV, 1-SW-MOV-102A, to maintain the capability to isolate the non-essential Service Water flowpaths within one hour. The one hour time limit for isolating the CCHX SW Supply are assumed as part of the design basis calculations associated with minimum initial canal level required to maintain the capability to supply adequate SW flow to the Recirculating Spray Heat Exchangers (RSHXs).

The administrative controls invoked provide acceptable contingency actions to ensure the non-essential SW isolation function can be met within the required design basis time frames assumed in calculation ME-0166. The administrative controls will be established to locally close the SW MOV. In event the MOV fails to fully close, local manual isolation of the CCHX SW supply is directed by O-AP-12.01. Therefore, an unreviewed safety question does not exist.

  • e Surry Monthly Operating Report No. 97-08 Page 18 of 22 TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTHNEAR: August, 1997 None During the Reporting Period
  • e Surry Monthly Operating Report No. 97-08 Page 19 of 22 CHEMISTRY REPORT MONTHNEAR: August, 1997 Unit No. 1 Unit No. 2 Primary Coolant Analysis Max. Min. Avg. Max. Min. Avg.

Gross Radioactivity, µCi/ml 3.31 E-1 1.73E-1 2.38E-1 2.63E-1 1.62E-1 2.04E-1 Suspended Solids, oom - - - - - -

Gross Tritium, µCi/ml 6.13E-1 5.82E-1 5.92E-1 2.05E-1 1.55E-1 1.77E-1 1131, µCi/ml 3.30E-4 2.14E-4 2.87E-4 1.56E-4 7.25E-5 1.16E-4 113111133 0.08 0.06 0.07 0.14 0.07 0.10 Hydrogen, cc/kg 31.0 26.2 28.9 29.8 25.3 27.4 Lithium, ppm 2.36 2.07 2.21 2.23 1.93 2.09 Boron - 10, ppm* 239.9 223.6 232.4 75.8 56.1 66.2 Oxygen, (DO), ppm s0.005 s0.005 s0.005 s0.005 s0.005 s0.005 Chloride, ppm 0.007 0.003 0.004 0.005 so.001 0.002 pH at 25 deqree Celsius 6.24 5.63 6.01 6.96 6.44 6.76 Boron - 10 = Total Boron x 0.196 Comments:

None

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' V

  • FUEL HANDLING e Surry Monthly Operating Report No. 97-08 Page 20 of 22 UNITS 1 &2 MONTHNEAR: August, 1997 New Fuel Number of New or Spent Shipment or Date Stored or Assemblies Assembly ANSI Initial Fuel Shipping Cask No. Received per Shipment Number Number Enrichment Cask Activity Unit 2 Batch 17 8/19/97 12 30L LM15A7 3.8950 15.00Ci Shipment 1 46L LM15AP 4.0631 15L LM159S 3.9195 47L LM15AQ 4.0643 28L LM15A5 3.9000 07L LM159J 3.9057 36L LM15AD 4.0545 14L LM159R 3.8985 16L LM159T 3.8985 04L LM159F 3.9179 23L LM15AO 3.9104 44L . LM15AM 4.0650 Unit 2 Batch 17 8/21/97 12 18L LM159V 3.8948 14.88Ci Shipment 2 31L LM15A8 3.9006 01L LM159C 3.9095 11 L LM159N 3.8992 19L LM159W 3.9047 10L LM159M 3.8989 13L LM159Q 3.8996 17L LM159U 3.8984 40L LM15AH 3.0640 27L LM15A4 3.9003 12L LM159P 3.8954 43L LM15AL 4.0657

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  • FUEL HANDLING e Surry Monthly Operating Report No. 97-08 Page 21 of 22 UNITS 1&2 MONTH/YEAR: August, 1997 New Fuel Number of New or Spent Shipment or Date Stored or Assemblies Assembly ANSI Initial Fuel Shipping Cask No. Received per Shipment Number Number Enrichment Cask Activity Unit 2 Batch 17 8/25/97 12 02L LM159D 3.9170 15.10 Ci Shipment 3 03L LM159E 3.9156 55L LM15AY 4.0556 51L LM15AU 4.0652 08L LM159K 3.9093 20L LM159X 3.9043 29L LM15A6 3.9006 32L LM15A9 3.9002 49L LM15AS 4.0638 54L LM15AX 4.05880 50L LM15AT 4.0622 52L LM15AV 4.0591 Unit 2 Batch 17 8/26/97 12 37L LM15AE 4.0602 15.14Ci Shipment 4 34L LM15AB 4.0473 26L LM15A3 3.9031 42L LM15AK 4.0714 38L LM15AF 4.0600 39L LM15AG 4.0621 22L LM159Z 3.9156 53L LM15AW 4.0653 35L LM15AC 4.0551 41L LM15AJ 4.0622 56L LM15AZ 4.0593 48L LM15AR 4.0651 J

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  • - Surry Monthly Operating Report DESCRIPTION OF PERIODIC TEST(S) WHICH WERE NOT COMPLETED No. 97-07 Page 22 of 22 WITHIN THE TIME LIMITS SPECIFIED IN TECHNICAL SPECIFICATIONS MONTH/YEAR:AUGUST, 1997 None During the Reporting Period
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