ML18153A398

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Monthly Operating Repts for Oct 1997 for Surry Power Station Units 1 & 2.W/971113 Ltr
ML18153A398
Person / Time
Site: Surry  Dominion icon.png
Issue date: 10/31/1997
From: Christian D, Fanguy M, Mason D
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
97-644, NUDOCS 9711210252
Download: ML18153A398 (25)
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Text

e e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 November 13, 1997 United States Nuclear Regulatory Commission Serial No.97-644 Attention: Document Control Desk SPS Lic/JDK RO Washington, D.C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-:~2 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 October 1997 is provided in the attachment.

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

Very truly yours,

i. J_C{j___

D. A. Christian, Station Manager Surry Power Station 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

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I 9711210252 971031 PDR ADOCK'. 05000280

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e Surry Monthly Operating Report No. 97-10 Page 1 of 24 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION MONTHLY OPERATING REPORT REPORT No. 97-10 Approved: ~ J cC-1,,.,~-'17

~ t a t i o n Manager Date

eSurry Monthly Operating Report No. 97-10 Page 2 of24 TABLE OF CONTENTS 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. ............................................................................. 21 Chemistry Report .............................................................................................................................................. 22 Fuel Handling - Unit Nos. 1 and 2 ...................................................................................................................... 23 Description of Periodic Test(s) Which Were Not Completed Within the Time Limits Specified in Technical Specifications ................................................................................................................. 24

e Surry Monthly Operating Report No. 97-10 Page 3*of24 OPERATING DATA REPORT Docket No.: 50-280 Date: 10/1/97 Completed By: D. K. Mason Telephone: (757) 365-2459

1. Unit Name: ........................................................ . Surry Unit 1
2. Reporting Period: .............................................. . October, 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 745.0 7296.0 217920.0
12. Hours Reactor Was Critical 745.0 5727.1 152561.8
13. Reactor Reserve Shutdown Hours 0.0 0.0 3774.5
14. Hours Generator On-Line 745.0 5604.5 150135.5
15. Unit Reserve Shutdown Hours 0.0 0.0 3736.2
16. Gross Thermal Energy Generated 1893653.7 13886566.2 352522010.0
17. Gross Electrical Energy Generated (MWH) 628890.0 4597663.0 115570481.0
18. Net Electrical Energy Generated (MWH) ............... . 607153.0 4435838.0 11 0029587. 0
19. Unit Service Factor ............................................... . 100.0% 76.8% 68.9%
20. Unit Availability Factor .......................................... . 100.0% 76.8% 70.6%
21. Unit Capacity Factor (Using MDC Net) .................. . 101.7% 75.9% 64.9%
22. Unit Capacity Factor (Using DER Net) ................... . 103.4% 77.2% 64.1%
23.
  • Unit Forced Outage Rate ....................................... . 0.0% 4.8% 14.9%
24. Shutdowns Scheduled Over Next 6 Months (Type, Date, *and Duration of Each):
25. If Shut Down at End of Report Period, Estimated Date of Start-up:
26. Unit In Test Status (Prior to Commercial Operation):

FORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION

e Surry Monthly Operating Report No. 97-10 Page 4 of24 OPERATING DATA REPORT Docket No.: 50-281 Date: 10/01/97 Completed By: D. K. Mason Telephone: (757) 365-2459

1. Unit Name: ........................................................ . Surry Unit 2
2. Reporting Period: .............................................. . October, 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 745.0 7296.0 214800.0
12. Hours Reactor Was Critical 156.9 6643.9 149719.5
13. Reactor Reserve Shutdown Hours 0.0 0.0 328.1
14. Hours Generator On-Line 129.8 6609.9 147707.7
15. Unit Reserve Shutdown Hours 0.0 0.0 0.0
16. Gross Thermal Energy Generated 305138.1 16629603.9 348103860.7
17. Gross Electrical Energy Generated (MWH) 101590.0 5524769.0 113975568.0
18. Net Electrical Energy Generated (MWH) ............... . 97882.0 5338938.0 108530817.0
19. Unit Service Factor ............................................... . 17.4% 90.6% 68.8%
20. Unit Availability Factor .......................................... . 17.4% 90.6% 68.8%
21. Unit Capacity Factor (Using MDC Net) .................. . 16.4% 91.4% 64.6%
22. Unit Capacity Factor (Using DER Net) ................... . 16.7% 92.9% 64.1%
23. Unit Forced Outage Rate....................................... . 0.0% 1.1% 12.0%
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

e eSurry Monthly Operating Report No. 97-10 Page 5 of24 UNIT SHUTDOWN AND POWER REDUCTION (EQUAL To OR GREATER THAN 20%)

REPORT MONTH: OCTOBER, 1997 Docket No.: 50-280 Unit Name: Surry Unit 1 Date: 11-04-97 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 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 (NUREG 0161)

eSurry Monthly Operating Report No. 97-10 Page 6 of 24 UNIT SHUTDOWN AND POWER REDUCTION (EQUAL To OR GREATER THAN 20%)

REPORT MONTH: OCTOBER, 1997 Docket No.: 50-281 Unit Name: Surry Unit 2 Date: 11-04-97 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 10/5/97 S 614 C 1 NA NA NA Unit 2 Ramping Off-Line for Scheduled Refueling Outage (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 (NUREG 0161)

e e Surry Monthly Operating Report No. 97-10 Page 7 of24 AVERAGE DAILY UNIT POWER LEVEL Docket No.: 50-280 Unit Name: Surry Unit 1 Date: 11-01-97 Completed by: J. D. Kilmer Telephone: (757) 365-2792 MONTH: OCTOBER, 1997 Average Daily Power Level Average Daily Power Level Day (MWe- Net) Day (MWe- Net) 1 819 17 820 2 820 18 821 3 820 19 822 4 820 20 822 5 818 21 819 6 805 22 819 7 814 23 820 8 819 24 820 9 813 25 820 10 814 26 856 11 814 27 810 12 815

  • 28 796 13 816 29 818 14 783 30 816 15 794 31 817 16 819 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 e Surry Monthly Operating Report No. 97-10 Page 8 of24 AVERAGE DAILY UNIT POWER LEVEL Docket No.: 50-281 Unit Name: Surry Unit 2 Date: 11-01-97 Completed by: John D. Kilmer Telephone: (757) 365-2792 MONTH: OCTOBER, 1997 Average Daily Power Level Average Daily Power Level Day (MWe- Net) Day (MWe- Net) 1 821 17 0 2 822 18 0 3 823 19 0 4 823 20 0 5 718 21. 0 6 33 22 0 7 0 23 0 8 0 24 0 9 0 25 0 10 0 26 0 11 0 27 0 12 0 28 0 13 0 29 0 14 0 30 0 15 0 31 55 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.

eSurry Monthly Operating Report

. No. 97-10 Page 9 of24

SUMMARY

OF OPERATING EXPERIENCE MONTH/YEAR: OCTOBER, 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:

10/1/97 0000 Unit 1 starts the month at 100%/ 850 MWe.

10/14/97 1624 Start power decrease from 100%/ 825 MWe to maintain condenser vacuum after removing 1-CW-E-I D from service.

I 1652 Stop power decrease at 93%/ 750 MWe.

10/15/97 0245 Start power increase after returning 1-CW-E-1D to service.

0858 Unit 1 is at 100%/ 845 MWe.

10/31/97 2400 Unit 1 finishes the month at 100% / 845 MWe.

UNITTwo:

10/1/97 0000 Unit 2 starts the month at 100% / 850 MWe.

10/05/97 0945 Start power decrease IAW 2-GOP-2. 1 from 100% / 850 MWe for Unit 2 refueling outage.

10/06/97 0037 Unit 2 is off-line.

0057 Manual Rx trip.

10/30/97 1209 Rx is critical.

10/31/97 1451 Unit is on-line.

2400 Unit 2 finishes the month at 29.5% / 210 MWe.

eSurry Monthly Operating Report No. 97-10 Page 10 of 24 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: OCTOBER, 1997 TM S1-97-016 Temporary Modification 10-2-97 (Safety Evaluation No.97-129)

This Temporary Modification (TM) was performed to remove the valve actuator for suction valve 1-GW-PCV-127 and install a blank flange on the body to allow trouble shooting of the Waste Gas Diaphragm Compressor 1-GW-C-I A.

The installation of the blank flange will meet system design requirements for pressure.

The flange will be leak tested following installation. If leakage were to occur, it would be detected by area radiation monitors. 1-GW-PCV-12_7 is normally operated open with the 1-GW-C-1A compressor running, therefore the compressor will not be subject to adverse operation. Normal alignment from the compressor to the inservice Waste Gas Decay Tank will be utilized. The installation of the blank flange will not increase the probability or consequences of an accident in the Safety Analysis Report and will not create the possibility of an accident of a different type. Therefore, an unreviewed safety question does not exist.

FS97-040 Updated Final Safety Analysis Report Change 10-2-97 (Safety Evaluation No.97-130)

Updated Final Safety Analysis Report (UFSAR) Change FS97-040 revises UFSAR Sections 5.4 and 6.2 to address the removal of concrete floor plugs and pressurizer cubicle roof slabs and to address the relaxation of containment RTD recalibration/recertification schedule.

The removal of concrete floor plugs and pressurizer roof slabs and the relaxation of containment RTD recalibration/recertification schedules on the containment air partial pressure (CV System) and containment bulk temperature (LM System) parameters will have a slight impact on the containment peak pressure and a minimal impact on the repressurization time, the sub atmospheric peak pressure and the pump NPSH. The acceptance criteria for these parameters continue to be met as shown by VP Calculation SM-1116 Rev 0, "Evaluation of Reduction in Containment Heat Sink and Extended RTD Replacement Schedule on Surry Containment Analysis." Therefore, an unreviewed safety question does not exist.

TM S1-97-15 Temporary Modification 9-24-96 (Safety Evacuation No.96-095 Rev 6)

Temporary modification S1-97-15 routes the Unit 1 Pressurizer Relief Tank vapor space through the sample line to the Gas Purge Line in the Sample Sink. The continuous vent will then be processed through the Overhead Gas System to the Waste Gas Decay Tanks (WGDT).

Overall radioactive gas releases will be lower from this processing method because the

  • radioactive gaseous waste will now decay in the WGDT instead of being routinely released. This process involves an improvement in the processing of gaseous waste and properly utilizes the existing system for processing gas. Therefore, an unreviewed safety question does not exist.

eSurry Monthly Operating Report No. 97-10 Page 11 of24 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: OCTOBER, 1997 TM S1-97-017 Temporary Modification 10-4-97 (Safety Evaluation No.97-133)

Temporary Modification (TM) S1-97-017 installed electrical jumpers to ensure continuity of the circuit during SVIXB Unit 1 Train 'B' Reactor Protection relay replacement.

The affected train will be out of service during this evolution unless directed otherwise by Operations

  • and the train will be tested following the replacement of the relay and removal of the jumper to verify operability of the circuit. .Train 'A' Reactor Protection will not be affected during the activity and compliance with Technical Specifications will be maintained. Therefore, an unreviewed safely question does not exist.

SE 97-135 Safety Evaluation 10-6-97 Safety Evaluation 97-135 was performed to evaluate the 1997 Unit 2 Refueling Outage Schedule.

The evaluation concluded that the refueling outage schedule is acceptable based on a review of (a) the capability to satisfy Cold Shutdown (CSD) and Refueling Shutdown (RSD) critical safety functions for Unit 2 and (b) the effects of Unit 2 outage activities on critical safety functions for Unit 1. Therefore, an unreviewed safety question does not exist.

TM S1-97-018 Temporary Modification 10-7-97 (Safety Evaluation No.97-136)

This Temporary Modification (TM) is required during the time the Solenoid Operated Valve to the selector valve isolating the CO2 from the cardox tank is replaced with a manual valve. The manual valve will allow operation of the selector main valve and provide CO2 to the hydrogen flats for use in purging the turbine generator~

The CO2 supply for the turbine generator will be provided by the temporary jumper and Administrative Control will be established to monitor tank level. The administrative limit will be set at 80%. During generator purging the temporary valve and the Low Pressure CO2 tank gauges will be continuously monitored by an operator in contact with the control room. The operator will isolate the turbine generator purge line if either we drop below the administrative limit of 80% or in the event of a fire, the system operates as described in the Safety Analysis Report. Therefore, an unreviewed safety question does not exist.

eSurry Monthly Operating Report No. 97-10 Page 12 of24 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: OCTOBER, 1997 SE 97-0061 Safety Evaluation 10-9-97 This safety evaluation reviews five TN-32 cask lids that have dimensional deviations.

None of these lids are currently in use at Surry. The lids are nominally 4.5 inches thick with a tolerance of .+/-. 0.030 inches. The maximum deviation in lid thickness from the nominal dimension is 0.061 inches (a lid thickness of 4.439 inches). This safety evaluation is applicable for lid deviations up to 0.071 inches less than the nominal lid thickness (a lid thickness of 4.429 inches).

No performance characteristics of the cask will be affected. Lid stress will remain below allowable limits and no structural performance criterion is challenged. Also, the use of these lids will not affect any performance characteristic associated with the thermal design or spent fuel criticality. The lid surface dose rate may increase slightly, but will remain well below the lid dose rates assumed in the ISFSI SAR and will not exceed the Technical Specifications limit for lid dose rates. Therefore, an unreviewed safety question does not exist.

TM S2-97-08 Temporary Modification 10-11-97 (Safety Evaluation No.97-138)

This Temporary Modification installs jumpers in the Unit 2 Train " 8 11 Hi CLS to allow the replacement of relay 1812.

The replacement of failed relay 1812 is a like for like replacement with no change to the design of the system or to the design of any existing separation between protection and control instrument systems. The work will be performed in train '8' only and train 'A' will be unaffected by this work. All work will be performed with the unit at cold shutdown.

Therefore, an unreviewed safety question does not exist.

JCO S1-97-002 Justification For Continued Operation 10-20-97 (Safety Evaluation No.97-091 Rev. 2)

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.

Safety Evaluation 97-091 was revised due to changing activity levels in the RCS due to failed fuel. The level of activity is well below Technical Specification limits. 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 conservatism 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.

e FACILITY CHANGES THAT DID Nor REQUIRE NRC APPROVAL

-Surry Monthly Operating Report No. 97-10 Page 13 of24 MONTH/YEAR: OCTOBER, 1997 TM S2-97-07 Temporary Modification 10-25-97 (Safety Evaluation No.97-100 Rev. 1)

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 characteristic 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. The *TM will be removed as new radiation monitors are installed or when ambient temperatures in the Main Steam Valve House are such that additional cooling is no longer required.

Therefore, an unreviewed safety question does not exist.

SE 97-146 Safety Evaluation 10-26-97 This safety evaluation address the acceptability of returning Unit 2 'C' Reactor Coolant Pump (RCP) to operation at the end of the current refueling outage with bolt #16 unloaded. The RCP sustained damage in 3 of 24 casing bolt holes. One of the three holes is unusable (#16) and the other two have reduced thread engagement (#20 and

  1. 23). The remaining 21 bolts have no reported thread damage.

The described damage has been determined acceptable from an analytical perspective.

23 of 24 bolts with a preload stretch in the 0.020 to 0.024 inch range provide the necessary loads for this pressure boundary joint and meet the ASME Code requirements in the UFSAR. The proposed change will not affect the ability of the RCP to provide forced coolant flow to the core or affect the integrity of the RCS. Therefore, an unreviewed safety question does not exist.

TM S2-97-10 Temporary Modification 10-28-97 (Safety Evaluation No.97-146)

This Temporary Modification connects a Dranetz Signal Monitor to the 2-CH-FC-2113 Flow Controller Components to monitor the input power leads and interconnecting leads for power supply disturbances.

Connecting the Dranetz Signal Monitor will not affect the operation of the Blender Control Circuits. The test equipment is designed with isolation circuitry such that the test equipment will not induce signals or add additional load to the monitored circuit. The Blender System will operate as designed and expected during installation, testing and removal of the test equipment. Required Technical Specification boration flow paths are not affected. Therefore, an unreviewed safety question does not exist. '

e Surry Monthly Operating Report No. 97-10 Page 14 of24 FACILITY CHANGES THAT DID Nor REQUIRE NRC APPROVAL MONTH/YEAR: OCTOBER, 1997 TM 82-97-11 Temporary Modification 10-29-97 (Safety Evaluation No.97-147)

This Temporary Modification installs jumpers in the Unit 2 Train "A" Reactor Protection system to allow the replacement of relay P8-YA.

The jumpers will allow the replacement of failed relay PB-YA Train "A" of the Reactor Protection system. The jumpers keep the daisy chain for the R2 and R1 wires intact to the Train "A" logic circuit and keep the Reactor Low Flow trip logics intact. They allow Train "A" of the reactor protection circuit to function as designed. Train "B" will not be worked and will also function as required. Therefore, an unreviewed safety question does not exist.

SE 97-0071 Safety Evaluation 10-30-97 This safety evaluation determines the effects, if any, of the localized reduction in neutron shield material on the TN-32 Topical Safety Analysis Report, the Surry ISFSI SAR, or the ISFSI Technical Specifications.

The localized reductions in the neutron shielding thickness occur in the vicinity of the upper trunnions. The neutron source strength in this area, of the cask is significantly reduced due to the reduction in fuel burnup at the top of the active fuel. Transnuclear evaluated the dose rates in the regions of the neutron shield deviations and determined that the dose rates remain below those reported in the TN-32 TSAR. Therefore, an unreviewed safety question does not exist.

I

e Surry Monthly Operating Report No. 97-10 Page 15 of24 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: OCTOBER, 1997 2-TOP4080 Temporary Operating Procedure 10-1-97 2-TOP-4081 2-AP-10.1 OTO Abnormal Procedure 2-AP-10.2 OTO (Safety Evaluation No.97-128) 2-AP-10.3 OTO 2-AP-10.4 OTO Temporary Operating Procedures 2-TOP-4080, "UPS 2A-1 and 2A-2 Operation To Support Implementation Of DCP 94-018," and 2-TOP-4081, "UPS 28-1 and 28-2 Operation To Support Implementation Of DCP 94-018," Abnormal Procedures 2-AP-10.1, "Loss of Vital Bus I", 2-AP-0.2, Loss of Vital Bus II", 2-AP-10.3, "Loss of Vital Bus Ill", 2-AP-10.4, "Loss of Vital Bus IV" are being revised to include the installations associated with DCP 94.018 which modifies Vital Bus and UPS equipment. These modifications are being made to .

enhance the ability of the system to cope with hot shorts due to an Appendix R fire.

There is no equipment being cross tied at an operational mode that is required to be separate or redundant. Compensatory measures are taken in the 2-TOP-4080 and 4081 procedures to prevent spurious actuation's of protective equipment. Additionally, because*

the vital buses at Surry do not normally have cross tie capability, the issues in GL 91-11 must be addressed. However, the mode of operation (Cold Shutdown or Empty Vessel) preclude most requirements. It was determined that a requirement does exist to maintain two 120 volt AC vital buses energized from their associated inverters connected to their respective AC buses during Cold Shutdown and Refueling conditions. This activity cross ties only two of the four vital buses on each unit at a time. In addition, none of the inverters will be separated from their respective DC buses during this evolution. Therefore, an unreviewed safety question does not exist.

1(2)-MOP-EP-204 Maintenance Operating Procedure 10-3-97 1(2)-MOP-EP-205 (Safety Evaluation No.94-190 Rev 1) 1(2)-MOP-EP-208 1(2) -MOP-EP-207 Maintenance Operating Procedures 1(2)-MOP-EP-204, "Removing 4160V Bus 1(2)H, 480V Buses 1(2)H and 1(2)H-1, and 480V MCC 1(2)H1-1 and 1(2)H1-2 From Service,"

1(2)-MOP-EP-205, "Returning 4180V Bus 1(2)H, 480V Buses 1(2)H and 1(2)H-1, and 480V MCC 1(2)H1-1 and 1(2)H1-2 to Service," 1(2) -MOP-EP-206, "Removing 4180V Bus 1(2)J, 480V Buses 1(2)J and 1(2)J1, and 480V MCC 1(2)J1-1 and 1(2)J1-2 From Service," and 1(2) -MOP-EP-207, "Returning 4180V Bus 1(2)J, 480V Buses 1(2)J and 1(2)J1, and 480V MCC 1(2)J1-1 and 1(2)J1-2 to Service," were revised to provide detailed instructions for lifting of battery leads when the DC Bus Tie is closed. When the battery leads are reconnected, a jumper is used to prevent electrical sparking near the batteries.

The procedures include the use of an electrical jumper to equalize the open circuit battery terminal voltage with the DC bus voltage when reconnecting the battery leads. This and other procedural controls minimize the potential of a dual battery short circuit and enhance the overall safety of the subject activity. The electrical load of the subject equipment is minimal and does not represent a significant addition to the distribution system. The .

affected equipment will be verified operable following the installation and removal of the TMs. The activity will only be performed when the affected unit is de-fueled. Therefore, an unreviewed safety question does not exist.

e Surry Monthly Operating Report No. 97-10 Page 16 of24 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTHNEAR: OCTOBER, 1997 1(2)-MOP-EP-204 Maintenance Operating Procedure 10-3-97 1(2)-MOP-EP-205 (Safety Evaluation No.94-189 Rev 1)

Maintenance Operating Procedures 1(2)-MOP-EP-204, "Removing 4160V Bus 1(2)H, 480V Buses 1(2)H and 1(2)H-1, and 480V MCC 1(2)H1-1 and 1(2)H1-2 From Service,"

1(2)-MOP-EP-205, "Returning 4160V Bus 1(2)H, 480V Buses 1(2)H and 1(2)H-1, and 480V MCC 1(2)H1-1 and 1(2)H1-2 to Service," were revised to provide detailed instructions for implementing temporary modifications (TM) to provide alternate power supplies for the radiation monitoring and fire detection equipment during the associated bus outages.

The TM provides temporary power to the Radiation Monitor and Fire Detection Equipment to maintain the equipment operable during a bus outage. The equipment will be removed and returned to service by procedure and will not affect the opposite unit. Therefore, an unreviewed safety question does not exist.

1/2-MOP-FW-004 Maintenance Operating Procedure 10-3-97 (Safety Evaluation No.97-132)

Maintenance Operating Procedures 1(2)-MOP-EP-204, "Removal and Return to Service of MFW Pumps 1(2) -FW-P-1 A and 1(2)-FW-P-I B" was revised to provide enhanced cooling to support main feed pump seal maintenance by vacuum dragging to the main condenser.

The procedure change verifies vacuum drag is established following jumper installation and ensures system restoration following the completion of repairs. The procedure does not affect the operating Main Feedwater Pump. Monitoring of condenser vacuum and secondary dissolved 02 will not be affected. The vacuum drag lineup will be secured prior to significantly affecting Main Condenser Vacuum or Dissolved 02. The procedure change does not involve safety related equipment. Therefore, an unreviewed safety question does not exist.

TSl-019 Technical Specification Interpretation 10-06-97 (Safety Evaluation No.97-134)

Technical Specification Interpretation TSl-019 was developed to clarify the Technical Specification clock entry requirements for an inoperable Emergency Power Supply on the opposite unit for the purpose of satisfying the Auxiliary Feedwater cross-connect requirements. (Re: Technical Specification 3.16. B).

  • The TSI does not impact any system's ability to perform its design function and does not reduce the Technical Specifications margin of safety, Therefore, an unreviewed safety question does not exist.
  • e Surry Monthly Operating Report No. 97-10 Page 17 of24 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: OCTOBER, 1997 1/2-0P-FH-001 Operating Procedure 10-9-97 (Safety Evaluation No.97-137)

Operating Procedure 1(2)-0P-FH-001, "Controlling Procedure for Refueling" was revised to change the sequence of loading the core. By completing the onload using alternative sequences the risk of fuel assembly damage due to interactions with other fuel assemblies

  • and with the fuel assembly loading guide is reduced.

The use of an alternative core loading sequence only changes the order in which fuel assemblies are loaded into the core. The only plant evolutions involved in loading the core are transfer of the fuel from the spent fuel Pool to the containment and handling the assemblies. No equipment other than the fuel handling equipment is directly involved. All operating procedures and Technical Specification Limiting Conditions for refueling will be met. Therefore, an unreviewed safety question does not exist.

1/2-0PT-Sl-022 Operations Periodic Test Procedure 10-14-97 (Safety Evaluation No.96-045, Rev. 3)

Operations Periodic Test Procedure, 1(2)-0PT-Sl-022, "SI Accumulator Discharge Check Valve Test with Reactor Head Removed," was revised to provide instructions for verifying that the Safety Injection System accumulator discharge check valves are free to open by discharging the accumulators into an open, de-fueled Reactor Coolant System (RCS).

The Unit will be at Refueling Shutdown and de-fueled. Controls are in place to avoid .

injection of nitrogen from the accumulators into the RCS. Either containment purge will be in operation or the containment hatch properly secured with at least one personnel hatch closed. To mitigate the potential radiological consequences if nitrogen is injected into the reactor vessel and ultimately into the containment atmosphere, the equipment and systems will be operated within design limits during the performance of this test. A Westinghouse lnfogram, No. 92-02, identified the potential of dislodging core specimen access plugs during the inadvertent discharge of an accumulator with the reactor head removed.

Although the reactor head will be removed, either the upper and lower internals will be in place, which is sufficient to prevent the dislodging of the plugs, or the upper internals will

  • be removed with the specimen access plugs removed and stored during the SI Accumulator Discharge Check Valve Test. Therefore, an unreviewed safety question does not exist.

O-ECM-0103-02 Electrical Corrective Maintenance Procedure 10-16-97 2-TOP-4080 Temporary Operating Procedure (Safety Evaluation No.97-139 Electrical Corrective Maintenance Procedure O-ECM-0103-02, "Station and Black Battery UPS System Maintenance* and Temporary Operating Procedure 2-TOP-4080, "UPS 2A-1 and 2A-2 Operation To Support Implementation of DCP 94-018," were revised to connect a dummy load to the UPS for troubleshooting, periodic maintenance, and testing.

Since the dummy load will be isolated from other loads on the output side of either the battery charger or the vital bus terminals, it will not alter the performance of any equipment on the output of the battery charger or the vital bus terminals. Since actual load imposed on the battery charger or inverter by the dummy load is procedurally limited to the design loads, the operation of the emergency bus and battery are not altered. Installation and removal of the dummy load will be done in accordance with these procedures. Therefore, an unreviewed safety question does not exist.

e Surry Monthly Operating Report *

. No. 97-10 Page 18 of24 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MoNTHNEAR: OCTOBER, 1997 FS96-033 Updated Final Safety Analysis Report Change 10-16-97 (Safely Evaluation No.97-142)

Updated Final Safety Analysis Report (UFSAR) Change FS96-033 revises UFSAR section 11.3.5, Environmental Survey Program, to reflect current industry standards.

This change is consistent with the Design Basis to sample the environs. The environmental survey program is not safety related. Therefore, an unreviewed safety question does not exist.

FS96-034 Updated Final Safety Analysis Report Change 10-16-97 (Safety Evaluation No.97-143)

Updated Final Safety Analysis Report (UFSAR) Change FS96-034 revised UFSAR section 7.2 because the section states that no tools are needed and no wires are disconnected when testing an analog protection channel. Contrary to this, the RCS loop temperatures need a decade box to simulate RCS temperature in order to calibrate the loop. The temperature sensor must be simulated to the loop with a decade box because the instrument design didn't provide test jacks in parallel to the protective circuit. The protection provided by these channels is the Overpower/Overtemperature DT.

Loop calibrations are performed when the RTD protection circuit is not needed. The calibrations do not defeat protective circuits that are needed for the plant condition. The design secures the protective circuitry and the procedures ensure correct reconnects.

Therefore, an unreviewed safety question does not exist.

1-MOP-Rl-001 Maintenance Operating Procedure 10-17-97 2-TMOP-Rl-3041 Temporary Maintenance Operating Procedure (Safety Evaluation No.96-147 Rev. 2)

Maintenance Operating Procedure 1-MOP-Rl-001, "Removal and Return to Service of Annunciator Panels 1A through 1E or 1F through 1K," and Temporary Maintenance Operating Procedure 2-TMOP-Rl-3041, "Removal and Return to Service of Annunciator Panels 2F through 2K With The Reactor De-fueled" provide instructions for increased monitoring and compensatory measures required while the affected annunciator panels are removed from service for maintenance.

Compensatory actions will be implemented in accordance with O-AP-10.13, "Loss of Main Control Room Annunicators," to monitor the affected systems and components. There is no unreviewed safety question raised by this evolution, and the radiological consequences

  • question does not exist.

lrry Monthly Operating Report No. 97-10 Page 19 of 24 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID Nor REQUIRE NRC APPROVAL MONTH/YEAR: OCTOBER, 1997 1/2-MOP-EP-30 Maintenance Operating Procedure 1/2-MOP-EP-31 (Safety Evaluation No.96-047 Rev 1) 10-19-97 l/2-MOP-EP-204 (Safety Evaluation No.96-047 Rev 2) 10-22-97.

l/2-MOP-EP-205 l/2-MOP-EP-206 Maintenance Operating Procedures l/2-tw3P-EP-30, "A Main Station Battery l/2-MOP-EP-207 Remove/Return to Service," 1/2-MOP-EP-31, "A Main Station Battery Remove/Return to Service," l/2-MOP-EP-204, "Remove "H" Emergency Bus From Service," l/2-MOP-EP-205, "Remove "J" Emergency Bus From Service," l/2-MOP-EP-206, "Return "H" Emergency Bus To Service," 1/2-MOP-EP-207, "Return "J. Emergency Bus To Service," were revised to perform Main Station Battery (MSB) charges at a high rate during refueling outages. The MSB must be determinate and charged for approximately 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />. Both MSBS cannot be charged within the de-fueled reactor vessel window.

The performance characteristics of systems dependent on a DC or Vital Bus power sources will remain unchanged as a result of this activity. All indication and controls will function normally throughout this activity. The jumper used to connect the MSB leads is considered a Temporary Modification (TM). However, the TM will only be used momentarily while the battery leads are being connected. The return to service procedure is adequate to ensure proper installation, removal and testing of the TM. Therefore, an unreviewed safety question does not exist.

1/2-0SP-TM-003 Operations Surveillance Procedure 10-20-97 .

(Safely Evaluation No.97-144)

Operations Surveillance Procedure 1/2-0SP-TM-003, "Functional Check of Turbine Valves and Limit Switch Operation," were revised to add a procedurally controlled Temporary Modification (TM). The turbine must be latched to open the valves, however, it is unlikely that the trips will be clear during the refueling period in which the test is to be performed.

The fuses for the turbine trip solenoids will be pulled to clear the trips and allow latching of the turbine.

Fuse removal will be used to selectively isolate inputs to the turbine generator protection system. The procedure adequately identifies the fuses to be pulled and the procedure also requires reinstallation of the fuses prior to test conclusion. The systems involved are not safety related and interface with the reactor trip circuitry is not required with the unit at.

refueling shutdown. The procedure is performed below 10% reactor power (P7) so that there will be no effect on the reactor. Therefore, an unreviewed safety question does not exist.

e* Surry Monthly Operating Report No. 97-10 Page 20 of24 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: OCTOBER, 1997 O-FCA-1.00 Fire Contingency Action Procedure 10-27-97 1-FCA-3.00 Operations Surveillance Procedure 1-FCA-4.00 Abnormal Procedure 1-FCA-6.00 (Safety Evaluation 97-093 Rev 3) 1-FCA-16.00 O-OSP-FP-011 This Safety Evaluation revision eliminates Unit 2 from non compliance with Appendix R.

O-AP-4S.OO Two separate areas of Appendix R non-compliance have been identified with respect to the JCO C-97-003 120 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 Uninterruptible Power

  • Supplies (UPS). This postulated event could potentially affect the associated UPS and, therefore, affect the downstream Appendix R Panel which could affect the Remote Monitoring Panel, the Emergency Diesel Generator Isolation Panels and Vital Bus Panels feeding Appendix R communications equipment. 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. All Vital Bus Panels are potentially affected by this issue. Compensatory measures implemented in the FCAS will a) disconnect the feeder conductor routed to the control room at UPSs 1A1 and 1A2 for a main control room fire to ensure the availability of distribution panels located in the ESGR fire area 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. These compensatory measures do not adversely impact the Class 1E electrical distribution system. Therefore, an unreviewed safety question does not exist.

e Surry Monthly Operating Report No. 97-10 Page 21 of24 TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: OCTOBER, 1997 None During the Reporting Period

e Surry Monthly Operating Report No. 97-10 Page 22 of24 CHEMISTRY REPORT MONTH/YEAR: OCTOBER, 1997 Unit No. 1 Unit No. 2 Primary Coolant Analysis Max. Min. Avg. Max. Min. Avg.

Gross Radioactivity, µCi/ml 3.96E-1 1.94E-1 2.68E-1 1.93E-1 4.80E-4 2.88E-2 Suspended Solids, ppm - - - 0.250  :$;0.010 0.046 Gross Tritium, µCi/ml 6.36E-1 5.94E-1 6.21E-1 1.69E-1 2.18E-2 9.54E-2 1131, µCi/ml 1.36E-2 3.18E-4 3.25E-4 1.14E-4 9.52E-5 1.06E-4 113111133 1.25 0.07 0.38 0.08 0.06 0.08 Hvdroaen, cc/ka 31.5 30.0 30.9 29.2 2.5 14.1 Lithium, ppm 2.34 2.05 2.19 3.53 0.11 1.03 Boron - 10, ppm* 208.7 193.3 200.9 508.8 10.2 281.7 Oxygen, (DO), ppm  :$;0.005  :$;0.005  :$;0.005 6.0  :$;0.005 3.2 Chloride, ppm  :$;0.05 0.003 0.007  :$;0.05 0.002 0.005 pH at 25 degree Celsius 6.85 6.15 6.53 7.27 4.85 5J4

None

e Surry Monthly Operating Report No. 97-10 Page 23 of 24 FUEL HANDLING UNITS 1 & 2 MONTH/YEAR: OCTOBER, 1997 New Fuel Number of New or Spent Shipment or Date Stored Assemblies Assembly ANSI Initial Fuel Shipping Cask No. or Received per Shipment Number Number Enrichment Cask Activity None During the Reporting Period

eSurry Monthly Operating Report No. 97-10 Page 24 of 24 DESCRIPTION OF PERIODIC TEST(S) WHICH WERE NOT COMPLETED WITHIN THE TIME LIMITS SPECIFIED IN TECHNICAL SPECIFICATIONS MONTHNEAR: OCTOBER, 1997 None During the Reporting Period

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