ML18152A404

Monthly Operating Repts for Dec 1996 for Surry Power Station Units 1 & 2.W/970110 Ltr

ML18152A404
Person / Time
Site:	Surry
Issue date:	12/31/1996
From:	Fanguy M, Mason M, Sarver S VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
97-020, 97-20, NUDOCS 9701220318
Download: ML18152A404 (23)

Text

-*~ l -*----. . . t.., VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 January 10, 1997 United States 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.97-020 NURPC Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Enclosed is the Monthly Operating Report for Surry Power Station Units 1 and 2 for the month of December 1996. If you have any questions or require additional information, please contact us. Very truly yours, S. P. Sarver, Acting Manager Nuclear Licensing and Operations Support Enclosure cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N.W. Suite 2900 Atlanta, Georgia 30323 Mr. R. A. Musser NRC Senior Resident Inspector Surry Power Station .,-----9701220318 961231 ' PD R ADOCK 05000280 ~---~-I I R PDR l 0~)on "-*-u39

-VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION MONTHLY OPERATING REPORT REPORT No. 96-12 Approved:

~S)G_{l Station Manager TABLE OF CONTENTS Section tl'8urry-Monthly Operating Report No. 96-12 Page 2 of 22 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 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 ..................................................................................

18 Chemistry Report .....................................................................................................................................................

19 Fuel Handling -Unit No. 1 ........................................................................................................................................

20 Fuel Handling -Unit No. 2 ........................................................................................................................................

20 Description of Periodic Test(s) Which Were Not Completed Within the Time Limits Specified in Technical Specifications

.......................................................................................................................

22 OPERATING DATA REPORT -Surry Monthly Operating Report Docket No.: Date: 50-280 01/01/97 No. 96-12 Page 3 of 22 Completed By: D. K. Mason (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 December, 1996 2546 847.5 788 840 801 Telephone:

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 YTD Cumulative

11. Hours In Reporting Period ...............................

744.0 8784.0 210624.0 12. Number of Hours Reactor Was Critical ...........

744.0 8784.0 146834.7 13. Reactor Reserve Shutdown Hours ..................

0.0 0.0 3774.5 14. Hours Generator On-Line ................................

744.0 8784.0 144531.0 15. Unit Reserve Shutdown Hours ........................

0.0 0.0 3736.2 16. Gross Thermal Energy Generated (MWH) ...... 1892818.6 22237651.3 338635443.8

17. Gross Electrical Energy Generated (MWH) ..... 633235.0 7395635.0 110972818.0

18. Net Electrical Energy Generated (MWH) .........

610120.0 7137776.0 105593749.0

19. Unit Service Factor ..........................................

100.0% 100.0% 68.6% 20. Unit Availability Factor. ....................................

100.0% 100.0% 70.4% 21. Unit Capacity Factor (Using MDC Net) ............

102.4% 101.4% 64.5% 22. Unit Capacity Factor (Using DER Net) ............

104.1% 103.1% 63.6% 23. Unit Forced Outage Rate ................................

0.0% 0.0% 15.2% 24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each): Refueling, March 6, 1997, 37 Days 25. If Shut Down at End of Report Period, Estimated Date of Start-up:

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

FORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION

'-* OPERATING DATA REPORT -Surry Monthly Operating Report Docket No.: Date: 50-281 01-01-97 No. 96-12 Page 4 of 22 Completed By: D. K. Mason {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 2 December, 1996 2546 847.5 788 840 801 Telephone:

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 YTD Cumulative

11. Hours In Reporting Period ...............................

744.0 8784.0 207504.0 12. Number of Hours Reactor Was Critical ...........

502.5 7572.7 143075.6 13. Reactor Reserve Shutdown Hours ..................

0.0 0.0 328.1 14. Hours Generator On-Line ................................

493.6 7541.7 141097.8 15. Unit Reserve Shutdown Hours ........................

0.0 0.0 0.0 16. Gross Thermal Energy Generated (MWH) ...... 1238394.8 18939202.8 331474256.8

17. Gross Electrical Energy Generated (MWH) ..... 415495.0 6295155.0 108450799.0

18. Net Electrical Energy Generated (MWH) .........

401148.0 6081464.0 103191879.0

19. Unit Service Factor ..........................................

66.3% 85.9% 68.0% 20. Unit Availability Factor .....................................

66.3% 85.9% 68.0% 21. Unit Capacity Factor (Using MDC Net) ............

67.3% 86.4% 63.7% 22. Unit Capacity Factor (Using DER Net) ............

68.4% 87.9% 63.1% 23. Unit Forced Outage Rate ................................

0.0% 1.7% 12.5% 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:

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

FORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION

-Surry Monthly Operating Report No. 96-12 Page 5 of 22 UNIT SHUTDOWN AND POWER REDUCTION (1) (2) (EQUAL To OR GREATER THAN 20%) REPORT MONTH: December, 1996 (3) Method (4) (5) Docket No.: 50-280 Unit Name: Surry Unit 1 Date: 01-02-97 Completed by: M. J. Fanguy Telephone:

(804) 365-2155 Duration of LER . System _ Component Cause & Corrective Action to Date Type Hours Reason Shutting No. Code Code Prevent Recurrence (1) F: Forced S: Scheduled (4) (2) REASON: Down Rx None During the Reporting Period A -Equipment Failure (Explain)

B Maintenance or Test C Refueling D Regulatory 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 (LER) File (NU REG 0161) (3) METHOD: 1 -Manual 2 -Manual Scram 3 -Automatic Scram 4 -Other (Explain)

(5) Exhibit 1 -Same Source (1) --(2) Surry-Monthly Operating Report No. 96-12 Page 6 of 22 UNIT SHUTDOWN AND POWER REDUCTION

{EQUAL To OR GREATER THAN 20%} REPORT MONTH: December, 1996 (3) Method (4) Docket No.: 50-281 Unit Name: Surry Unit 2 Date: 01-02-97 Completed by: M. J. Fanguy Telephone:

(804) 365-2155 (5) Duration of LER No. System Component Cause & Corrective Action to Date Type Hours Reason Shutting Code Code Prevent Recurrence 12/12/96 s 250.4 (1) F: Forced S: Scheduled B (2) REASON: Down Rx 3 96-006 A Equipment Failure (Explain)

B Maintenance or Test C Refueling D Regulatory Restriction SJ 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 (NU REG 0161) NA Scheduled maintenance outage to repair the letdown line (had automatic Rx trip from 11 % power due to steam flow/feed flow mismatch with low level on "A" SG). (3) METHOD: 1 -Manual 2 -Manual Scram 3 -Automatic Scram 4 -Other (Explain)

(5) Exhibit 1 -Same Source

'* --. Surry Monthly Operating Report No. 96-12 Page 7 of 22 AVERAGE DAILY UNIT POWER LEVEL Docket No.: 50-280 Unit Name: Surry Unit 1 Date: 01-05-97 Completed by: J. D. Kilmer Telephone:

(804) 365-2792 MONTH: December, 1996 Average Daily Power Level Average Daily Power Level Day (MWe -Net) Day (MWe -Net) 824 17 821 2 824 18 819 3 813 19 819 4 821 20 813 5 822 21 802 6 826 22 799 7 824 23 815 8 824 24 820 9 821 25 823 10 824 26 820 11 823 27 822 12 825 28 823 13 820 29 823 14 820 30 824 15 822 31 824 16 822 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.

_____J J ~urry Monthly Operating Report No. 96-12 Page 8 of 22 AVERAGE DAILY UNIT POWER LEVEL MONTH: December, 1996 Average Daily Power Level Day 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 INSTRUCTIONS (MWe -Net) 822 821 824 826 827 827 828 828 827 827 828 801 5 0 0 0 Day 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 Docket No.: 50-281 Unit Name: Surry Unit 2 Date: 01-05-97 Completed by: John D. Kilmer Telephone:

(804) 365-2792 Average Daily Power Level (MWe -Net) 0 0 0 0 0 0 203 818 825 826 826 825 825 825 825 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.

SUMMARY

OF OPERATING EXPERIENCE MONTHNEAR:

December, 1996 .. urry Monthly Operating Report No. 96-12 Page 9 of 22 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: 12/01/96 12/03/96 12/31/96 UNIT Two: 12/01/96 12/03/96 12/12/96 12/13/96 12/23/96 12/24/96 12/31/96 0000 0925 The reporting period began with the unit operating at 100% power, 850 MWe. Start power reduction to change out E/P controller for 1-FW-FCV-1478.

1017 Stopped power reduction at 90%, 770 MWe. 1206 Started power increase.

1345 Stopped power increase at 100%, 850 MWe. 2400 0000 1538 2102 0235 0255 The reporting period ended with the unit operating at 100% power, 855 MWe. The reporting period began with the unit operating at 99% power, 850 MWe. Unit returned to 100% power, Computer calorimetric operable after failed Feedwater RTD replaced.

Started power reduction in preparation for planned outage to repair the letdown line. Auto Reactor Trip from 11 % power due to steam flow/feed flow mismatch with low steam generator water level in the "A" steam generator.

Commence Reactor Startup. 0348 Reactor critical.

1243 Unit on line, start power increase.

1548 2400 Stopped power increase at 100%, 858 MWe. The reporting period ended with the unit operating at 100% power, 855 MWe.

FS 96-44 TM S2-96-28 FS 96-47 FS 96-49 -urry Monthly Operating Report No. 96-12 Page 10 of 22 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

December, 1996 Updated Final Safety Analysis Report Change (Safety Evaluation 96-150) 12-02-96 Section 14.3.2 of the UFSAR discusses the rupture of a main steam pipe. Questions have been raised about what is meant by the term "main steam lines." This change adds information to clarify what is meant by the term "main steam lines." Since this change is clarification of existing information only, no change to the meaning or intent of the UFSAR will be made. Therefore, an unreviewed safety question does not exist. Temporary Modification 12-3-96 (Safety Evaluation No.95-005 Rev.1) Temporary Modification (TM) S2-96-28 installed a replacement resistance temperature device (RTD) in the local temperature indicator thermowell to provide the Unit 2 Steam Generator (SG) "8" inlet feedwater temperature.

The local temperature indicator thermowell is of the same design as the main RTD thermowell, except for length. An evaluation of this difference concluded, based on a comparison of previous temperature data, that a reliable and accurate indication of the SG "B" inlet feedwater temperature can be obtained.

This modification did not reduce the Technical Specifications margin of safety. Therefore, an unreviewed safety question does not exist. -Updated Final Safety Analysis Report Change (Safety Evaluation 96-154) 12-5-96 Currently, a note in UFSAR Table 6.2-2 and Section 6.2.2.1.1 states "any of the listed motor operated valves will actuate a combination light and alarm when one or more SIS valves are out of position." This does not accurately reflect actual plant control room indication, since not all of the SIS valves listed will actuate the "SIS Valves Out Of Position" alarm. A review of the SI Design Basis Document and the applicable General Design Criteria for control room indication did not indicate any specific requirement for all SIS valves listed in Table 6.2-2 or those implied in Section 6.2.2.1.1 to actuate an alarm when out of position.

Sufficient monitoring exists to ensure that the SI valves are in their correct positions.

Therefore, an unreviewed safety question does not exist. Updated Final Safety Analysis Report Change (Safety Evaluation 96-152) 12-5-96 Due to limited space, a change to the UFSAR is required to allow storage of QA records offsite at an approved facility.

Records will continue to be stored in accordance with the applicable ANSI commitments.

This change is administrative in nature with no modification to plant systems or components.

Therefore, an unreviewed safety question does not exist.

FS 96-41 FS 96-54 DCP 95-023 SE 96-0021 .,,urry Monthly Operating Report No. 96-12 Page 11 of 22 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/VEAR:

December, 1996 Updated Final Safety Analysis Report Change (Safety Evaluation 96-153) 12-5-96 This UFSAR change moves the stated numerical value of the auxiliary feedwater flow for the licensing basis Loss Of Normal Feedwater

{LONF) event from Section 10.3.5.3 to Section 14.2.11.1.3 where the other analysis assumptions are located. The statement in Section 10.3.5.3 will be changed from "a minimum of 500 gpm of auxiliary feedwater flow to "adequate auxiliary feedwater flow." This change is administrative in nature with no modification to plant systems or components.

Therefore, an unreviewed safety question does not exist. Updated Final Safety Analysis Report Change (Safety Evaluation 96-155) 12-5-96 This UFSAR change updates Section 3.2.1 to be consistent with the current plant configuration.

When the plant configuration was changed in 1984, Section 6.2 Safety Injection System and Section 14.3.2 accident analysis were updated but Section 3.2.1 was overlooked.

This UFSAR change updates Section 3.2.1 for consistency and has no physical impact on any plant system or component.

Therefore, an unreviewed safety question does not exist. Design Change Package * * (Safety Evaluation 95-129) 12-5-96 Design Change Package 95-023 replaced existing fire pump 1-FP-P-3 with a pump which meets the specified design requirements and provides the required pressure and flow capacity.

Therefore, an unreviewed safety question does not exist. Safety Evaluation 12-11-96 (Safety Evaluation 96-0021) The NRC has approved the TN-32 Dry Storage Cask Topical Safety Analysis Report and a Surry ISFSI Technical Specification amendment for the use of this cask. This safety evaluation is performed to incorporate the use of the TN-32 storage cask into the Surry ISFSI Safety Analysis Report, Environmental Report and License Application.

The design and operation of the TN-32 cask is similar to that of the four cask designs already approved for use at the Surry ISFSI. There is no reduction in the margin of safety as defined in the basis for any Technical Specification.

The current design basis for the Surry ISFSI was evaluated and remains bounding.

Therefore, an unreviewed safety question does not exist.

,_, SE 96-0031 SE 96-0041 SE 96-161 e Surry Monthly Operating Report No. 96-12 Page 12 of 22 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

December, 1996 Safety Evaluation 12-11-96 (Safety Evaluation 96-0031) This safety evaluation is performed to allow the revision of the ISFSI Safety Analysis Report for the storage of burnable poison rod assemblies and thimble plugging devices with the spent fuel in dry storage casks at the Surry ISFSI. The NRC issued a draft Standard Review Plan for Dry Cask Storage Systems which indicated that the:storage of "non-fuel core components" or "control assemblies" in dry storage casks is allowed, if they are described in the cask Topical Safety Analysis Report and included in the structural and shielding analyses.

The casks have been analyzed and it has been determined that the margin of safety has not been reduced by the storage of these core components since the cask and basket stresses remain below the allowable values during a cask end drop or tipover event. Additionally, the existing surface dose rate limits in Technical Specification 3.3 still apply -to ensure compliance with the dose rate limits for the ISFSI perimeter fence and nearest resident.

Therefore, an unreviewed safety question does not exist. Safety Evaluation 12-11-96 (Safety Evaluation 96-0041) This safety evaluation assesses the as-built conditions of ISFSI Surry Pad 2 to determine whether this pad meets the criteria established by the NRC in their Safety Evaluation Report (SER) for cask drops or tipover of the Transnuclear TN-32 cask. A review of the pad design criteria was performed.

The results of the review determined that the pad meets the criteria established by the NRC SER. This change is administrative in nature with no modification to plant systems or components.

Therefore, an unreviewed safety question does not exist. Safety Evaluation 12-11-96 (Safety Evaluation 96-161) This safety evaluation assesses the update to the Small Break Loss of Coolant Accident (SBLOCA) analysis which is being incorporated into the licensing analysis basis for Surry Units 1 and 2. The reanalysis of the SBLOCA was performed using Westinghouse LOCA-ECCS SBLOCA Evaluation Model. The analytical techniques are in full compliance with 1 OCFR50, Appendix K. The evaluation model is an approved methodology in the COLR list. This reanalysis used conservative assumptions with respect to existing limits and plant capabilities.

The analysis results show that the emergency core cooling system will meet the acceptance criteria in 1 OCFR50.46.

Therefore, an unreviewed safety question does not exist.

. . " FS 96-59 FS 96-53 SE 96-158 e .

• Surry Monthly Operating Report No. 96-12 Page 13 of 22 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTHIYEAR:

December, 1996 Updated Final Safety Analysis Report Change (Safety Evaluation 96-158) 12-12-96 Currently, UFSAR Sections 4.1.2.6 and 4.2.7.1 reference humidity indication as a method of monitoring containment coolant leakage. In 1978, the containment humidity monitoring instrumentation was abandoned in place. Primary coolant system leakage is most readily detected by increased makeup requirements and by c secondary coolant system leak. Therefore, an unreviewed safety question does not exist. Updated Final Safety Analysis Report Change (Safety Evaluation 96-157) 12-12-96 Currently, UFSAR Section 4.3.4.2 contains a statement "Two additional bottles charged to 2200 psig are connected to the system but isolated as spares" may lead to misinterpretation of the requirements associated with maintaining ready spares. A revision is needed to clarify current operating practices associated with the pressurizer PORV backup air supply system. The change will state "The two bottles that are not aligned to the manifold are initially fully charged and used as installed spare bottle capacity.

The intent is to reduce the time required to reestablish bottle pressure when a low pressure alarm occurs with the unit at load. Instead of changing out bottles, the spare bottles need only be valved in." The intent of this change is to minimize the radiation exposure personnel would receive changing depleted bottles during power operations.

The backup air bottle pressure limits are set by the low temperature overpressure mitigating system, based on the assumption that no operator action is required for ten minutes. The air supply should be capable of providing 100 cycles. Two high pressure bottles are normally valved in with each bottle capable of 115 cycles at 1000 psig. The low pressure backup air annunicator alarms at 1000 psig. Therefore, an unreviewed safety question does not exist. Safety Evaluation 12-16-96 (Safety Evaluation No.96-159) This safety evaluation assesses the impact of the "B" Circulating Water (CW) Pump being out of service for greater than 30 days. The purpose of the pump is to provide the water required to condense the turbine exhaust steam and supply water to the Service Water System by transferring water from the James River into a common intake canal for Units 1 and 2. There are four CW pumps for each unit. The minimum requirement per Technical Specification (TS) 3.14 to support plant operation is two CW pumps. The TS requirement is satisfied and adequate canal level is maintained by the remaining operable CW pumps. Therefore, an unreviewed safety question does not exist.

TM 82-96-030 FS 96-55 FS 96-58 .,,urry Monthly Operating Report No. 96-12 Page 14 of 22 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

December, 1996 Temporary Modification 12-17-96 (Safety Evaluation No.96-113 Rev. 1) This Temporary Modification (TM) adds spray shields to protect the RCP flange studs from boric acid spray and reduce the amount of boric acid spray being drawn into the RCP motor. Additionally, this TM will provide remote monitoring of the RCP flange from the Main Control Room by the installation of a camera and lights in the "C" RCP Loop Room. The spray shields will not affect the original configuration of the Reactor Coolant System or the loading on the primary loop piping supports.

The shields will be secured in order to prevent them from becoming missiles during postulated earthquakes or high energy line breaks. The camera has no ties to existing station protection or control systems. The installation of equipment by this TM will not create an unreviewed safety question as described in 1 OCFR50.59.

Therefore, an unreviewed safety question does not exist. Updated Final Safety Analysis Report Change (Safety Evaluation 96-164) 12-19-96 UFSAR Section 9.9.1.3 is being revised to delete some details on specific times when actions would be taken prior to the arrival of a hurricane on site and the specified hurricane wind speed. UFSAR Section 9.10.4.18 will be revised to reflect the Circulating Water (CW) isolation valves as safety-related.

This Section incorrectly identifies the valves as non safety-related.

The details removed from Section 9.9.1.3 are not vital to the hurricane actions. Important actions as described in Technical Report 0032 will remain in the UFSAR. The CW isolation valves that were described as non safety-related were installed and maintained as safety-related.

These changes are administrative in nature with no modification to plant systems or components.

Therefore, an unreviewed safety question does not exist. Updated Final Safety Analysis Report Change (Safety Evaluation 96-163) 12-19-96 Currently, UFSAR Section 10.3.7.3 contains a statement indicating the non-safety related turbine generator DC bearing oil pump is required to function during a LOCA or loss of station power. Additionally, Section 10.3.7.4 contains a statement that the DC bearing oil pump will be tested on a monthly basis. Changes to the UFSAR Section 10.3.7.3 will state the DC bearing oil pump is designed to function during a loss of station power. Section 10.3.7.4 will change the frequency of DC oil pump testing from "monthly" to "periodically." Although the turbine generator DC bearing oil pump is designed to operate during a loss of station power, there is no design basis assumption that requires it to be operable during a LOCA or loss of station power. The testing frequency is governed by industry good practice and insurance carrier requirements.

There are no Technical Specification or design basis assumption requirements for the monthly testing of the pump. Therefore, an unreviewed safety question does not exist.

. . ' SE 96-167 SE 96-0051 ~urry Monthly Operating Report No. 96-12 Page 15 of 22 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

December, 1996 Safety Evaluation 12-20-96 (Safety Evaluation No.96-167) This safety evaluation assesses continued Unit 2 power operation with a single Unit 2 RCP~1 C flange bolt preload stretch below the minimum required by the vendor. The flange bolts provide a structural connection between the flange and the pump casing and provide adequate preload and compression to the gasketed joint. The bolt has full thread engagement which meets the structural requirements.

However, with the bolt preload below the required minimum some leakage may occur at the casing joint. A temporary modification was installed to monitor the flange area using a camera. The probability of a rapid propagation type failure is not credible based on an evaluation performed by the vendor. The RC System pressure boundary will continue to accommodate the temperatures and pressures attained under all expected modes of station operations or anticipated transients, as well as maintain integrity without catastrophic rupture. Therefore, an unreviewed safety question does not exist. Safety Evaluation 12-20-96 (Safety Evaluation No. 96-0051) This safety evaluation assesses the continued storage of TN-32 Cask No.1 on the ISFSI pad with a pinhole leak in a weld. The weld connects the outer shell to the body of the cask. The cask is to remain on the ISFSI pad until appropriate procedures can be developed for the repair of the pinhole leak. The function of the weld is to attach the outer skin to the cask body. The outer skin encloses the resin boxes which provide the primary neutron shielding for the cask. The presence of the pinhole could potentially allow the contents of the atmosphere inside the resin chamber to escape as the cask reaches equilibrium temperature.

The gases given off by the resin during heatup could include carbon dioxide, carbon monoxide and other hydrocarbons.

Gas samples taken from the leakage site determined that no radioisotopes were present and that the primary constituents of the sample were nitrogen and oxygen in the percentages that would indicate that the leaking gas was air. No helium was detected and a smear taken of the area was not contaminated.

The release of a small amount of these gases will not result in any detrimental environmental or radiological conditions.

Therefore, an unreviewed safety question does not exist.

O-ICM-EH-CAB-001 1 (2)-NSP-SD-001 SE 96-160 ~urry Monthly Operating Report No. 96-12 Page 16 of 22 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTHNEAR:

December, 1996 Instrument Corrective Maintenance (Safety Evaluation No.96-156) 12-12-96 Instrument Corrective Maintenance Procedure O-ICM-EH-CAB-001, "EHC System Diagnostic Checks," was revised to provide instructions for a procedurally controlled temporary test lead. This will allow the collection of test data using a recorder during the Unit 2 ramp down to investigate a possible signal problem with the governor valve control system. Should the governor valves begin to open, the valve position limiting circuit would act to limit the power increase.

Should the valve limiter fail, the resultant load increase is still bounded by the current safety analyses.

The physical limit on steam flow through all governor valves wide open (105%) is bounded by the analyzed step load increase safety analysis limit (116%). Should a step load decrease occur, the plant response would be within the current design requirements and the reactor protection system would function as required.

Therefore, an unreviewed safety question does not exist. Engineering Surveillance Procedure (Safety Evaluation No.96-162) 12-16-96 Engineering Surveillance Procedure 1 (2)-NSP-SD-001, "Optimum Feedwater Heater Level Test," provides instructions for determining/calculating the optimum water level in the shell side of feedwater heaters 1A, 1 B, 3A, 3B, 4A, 4B, 5A, 5B. This will maximize the efficiency of the feedwater heaters, increasing overall plant efficiency.

This test will affect the efficiency of the feedwater heaters, but will not affect the operation of the heaters. The heaters will operate as required, but with different setpoints for liquid level. Therefore, an unreviewed safety question does not exist. Safety Evaluation 12-16-96 (Safety Evaluation No.96-160) Safety evaluation 96-160 was performed to assess the acceptability of continued usage of the procedure O-OSP-RM-002 to check source the Component Cooling Radiation Monitors in accordance with Technical Specification requirements, although the methodology differs from that stated in the UFSAR. This condition will continue until June 1997 when the Main Control Room Radiation Monitoring Cabinet portion of the Design Change is installed.

The evaluation establishes the acceptability of the current method of source checking the Component Cooling Water Radiation Monitors and does not affect any safety related equipment.

Therefore, an unreviewed safety question does not exist.

.. L O-IMP-BR-001 O-AP-23.01

-~urry Monthly Operating Report No. 96-12 Page 17 of 22 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

December, 1996 Instrument Maintenance Procedure (Safety Evaluation No.96-166) 12-19-96 Instrument Maintenance Procedure O-IMP-BR-001, "Monitoring of Boron Recovery Tank During Reactor Coolant System Drain Down Evolutions," provides instructions to connect a calibrated 250 ohm resistor in series in the Boron Recovery Tank (BRT) level instrumentation loop. This will allow a chart recorder to be connected across the resistor to allow accurate monitoring of RCS inventory during RCS draining evolutions.

The system is passive and its characteristics will be unchanged.

The "Channel Operation and Accuracy" Section of the affected tank's level calibration procedure will be completed once the resistor is installed.

The BRT high and low level alarms, which are not safety related, will be unaffected by this activity.

Therefore, an unreviewed safety question does not exist. Abnormal Procedure 12-19-96 (Safety Evaluation No.96-165) Abnormal Procedure O-AP-23.01, "Rapid RCS Cooldown," was revised to modify the current practice of borating to Cold Shutdown boron prior to initiating a cooldown and blocking Safety Injection to a method of concurrent borations and cooldown.

In addition, this change increases the current administrative Reactor Coolant System (RCS) cooldown rate from 50 °F/hr to 75 °F/hr while above the Low Temperature Overpressure Mitigation System {L TOPs) enabling temperature of 350 °F. Below the L TOPs enabling temperature of 350 °F, the current administrative cooldown rate of 50 °F/hr is still .. applicable.

The proposed changes do not alter the design or principles of operation of the CVCS or any associated safety related system. Therefore, the probability of the malfunction of equipment important to safety is not increased, and no new or unique equipment malfunction scenarios are created. Since the CVCS is not relied on for accident mitigation, the consequences of the CVCS malfunction has not increased.

The exception is when the plant is aligned to Safety Injection and the Charging Pumps are considered High Head Safety Injection Pumps. Inadvertent boron dilution by addition of primary grade water to the RCS is addressed in the UFSAR and this analysis remains bounding.

The severity of an inadvertent dilution with the proposed changes would be less severe than the bounding analysis.

Therefore, an unreviewed safety question does not exist.

.. .. j L ~urry Monthly Operating Report No. 96-12 Page 18 of 22 TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:

December, 1996 None During the Reporting Period

.,.. I, CHEMISTRY REPORT MONTHNEAR:

December, 1996 Unit No. 1 Primarv Coolant Analysis Max. Min. Avg. Gross Radioactivity, µCi/ml 9.55E-1 4.97E-2 7.21 E-1 Suspended Solids, ppm sO.D10 so.010 so.010 Gross Tritium, µCi/ml 3.30E-1 2.38E-1 2.85E-1 1131, µCi/ml 1.15E-2 8.07E-3 9.43E-3 113111133 0.41 0.23 0.29 Hydrogen, cc/kg 40.1 31.2 36.6 Lithium, oom 1.67 1.24 1.53 Boron -10, ppm* 47.4 28.4 37.6 Oxygen, (DO), ppm s0.005 s0.005 s0.005 Chloride, ppm 0.004 <0.001 0.002 pH at 25 deqree Celsius 7.30 6.78 6.98

• Boron -10 = Total Boron x 0.196 Comments:

None ~urry Monthly Operating Report No. 96-12 Page 19 of 22 Unit No. 2 Max. Min. Avg. 4.10E-1 2.27E-3 1.44E-1 2.5 so.010 0.389 6.32E-1 1.73E-1 4.00E-1 9.36E-5 1.15E-5 4.65E-5 0.11 0.06 0.08 31 1.7 16.6 3.69 0.10 1.43 346.7 192.1 289.6 6.0 s0.005 0.18 0.008 0.002 0.004 6.81 4.79 5.61

' '. .... , . I. New or Spent Fuel Shipment Date Stored or Number Received Spent Fuel Cask TN-32-01 12/18/96 e FUEL HANDLING UNITS 1 & 2 MONTH/YEAR:

December, 1996 Number of Assemblies Assembly ANSI per Shipment Number Number 32 N40 NOANSI 6P2 LM05YN N44 NOANSI N46 NOANSI N33 NOANSI D02 LM007P D25 LM007D D39 LM0075 2E8 LMODEP N34 NOANSI D31 LM0084 D05 LM008E D14 LM007U D41 LMOOCA D50 LM0073 N42 NOANSI N38 NOANSI D06 LM008F D49 LM007G D43 LM007J V17 LM041K 6P4 LM05YR N39 NOANSI 2E7 LMODFH D33 LM008K D46 LMOOCD ~urry Monthly Operating Report No. 96-12 Page 20 of 22 New or Spent Initial Fuel Shipping Enrichment Cask Activity 2.5560 3.6070 2.5560 2.5560 2.5560 3.3250 3.3250 3.3250 2.5933 2.5560 3.3250 3.3250 3.3250 3.3250 3.3250 2.5560 2.5590 3.3250 3.3250 3.3250 2.9060 3.6070 2.5560 3.6028 3.3250 3.3250

• • ' ' . "*. J New or Spent Fuel Shipment Date Stored or Number Received FUEL HANDLING UNITS 1 & 2 MONTHNEAR:

December, 1996 Number of Assemblies Assembly ANSI per Shipment Number Number V19 LM042H N45 NOANSI N41 NOANSI N43 NOANSI 6P8 LM009PA N47 NOANSI . R Surry Monthly Operating eport Initial Enrichment 2.9060 2.5560 2.5560 2.5560 3.6070 2.5560 No. 96-12 Page 21 of 22 New or Spent Fuel Shipping Cask Activity

' .... tturry Monthly Operating Report No. 96-12 Page 22 of 22 DESCRIPTION OF PERIODIC TEST(S) WHICH WERE NOT COMPLETED WITHIN THE TIME LIMITS SPECIFIED IN TECHNICAL SPECIFICATIONS MONTH/YEAR:

December, 1996 None During the Reporting Period