ML18152A552

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Monthly Operating Repts for Apr 1997 for Surry Power Station Units 1 & 2.W/970513 Ltr
ML18152A552
Person / Time
Site: Surry  Dominion icon.png
Issue date: 04/30/1997
From: Fanguy M, Kilmer J, Mason D
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
97-299, NUDOCS 9705210309
Download: ML18152A552 (20)
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Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 May 13, 1997 United States Nuclear Regulatory Commission Serial No.97-299 Attention: Document Control Desk NUJDK/GDM 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 Enclosed is the Monthly Operating Report for Surry Power Station Units 1 and 2 for the month of April 1997.

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 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 Surry Monthly Operating Report No. 97-04 Page 1 of 19 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION MONTHLY OPERATING REPORT

~*\

REPORT No. 97-04 Approved:

Station Manager

  • TABLE OF CONTENTS
  • Surry Monthly Operating Report No. 97-04 Page2of19 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 No. 1 ......................................................................................................... 9 Summary of Operating Experience - Unit No. 2 ......................................................................................................... 9 Facility Changes That Did Not Require NRG Approval ............................................................................................ 10 Procedure or Method of Operation Changes That Did Not Require NRG Approval. ...... ;......................................... 14 Tests and Experiments That Did Not Require NRG Approval .................................................................................. 16 Chemistry Report ........................................................................................,............................................................ 17 Fuel Handling - Unit No. 1 ..........................................................................'. ............................................................. 18 Fuel Handling - Unit No. 2 ........................................................................................................................................ 18 Description of Periodic Test(s) Which Were Not Completed Within the Time Limits
  • Specified in Technical Specifications ........................................... :,......... :............... :... :.............. :.... :.*....................... 19
  • OPERATING DATA REPORT
  • Surry Monthly Operating Report No. 97-04 Page 3 of 19 Docket No.: 50-280 Date: 05/01/97 Completed By: D. K. Mason Telephone: (804) 365-2459
1. Unit Name: .......................................................... . Surry Unit 1
2. Reporting Period: ............................................... .. April, 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 YTD Cumulative

11. Hours In Reporting Period ............................... 719.0 2879.0 213503.0
12. Number of Hours Reactor Was Critical ........... . 0.0
  • 1300.8 148135.5
13. Reactor Reserve Shutdown Hours .................. 0.0 0.0 3774.5
14. Hours Generator On-Line ................................ 0.0 * .1278.3 145809.3
15.
  • Unit Reserve Shutdown Hours ........................ 0.0 0.0 3736.2
16. Gross Thermal Energy Generated (MWH) ...... 0.0 * *3190453.6 341825897.4
  • 17. Gross Electrical Energy Generated (MWH) ..... *o.o 1065255.0 112038073.0
18. Net Electrical Energy Generated (MWH) ......... 0.0 1029190.0' 106622939.0
19. Unit Service Factor .......................................... *0.0% 44.4% 68.3%
20. Unit Availability Factor ..................................... '0.0% .44.4% 70.0%
21. Unit Capacity Factor (Using MDC Net) ............ 0.0% 44.6% 64.2%
22. Unit Capacity Factor (Using DER Net) ............ 0.0% 45.4% 63.4%
23. Unit Forced Outage Rate ................................ 0.0% *
  • 18.3% 15.2%
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: 5/5/97
26. Unit In Test Status (Prior to Commercial Operation):

FORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION

Surry Monthly Operating Report No. 97-04 Page 4 of 19 OPERATING DATA REPORT Docket No.: 50-281 Date: 05/01/97 Completed By: D. K. Mason Telephone: (804) 365-2459

1. Unit Name: .......................................................... . Surry Unit 2
2. Reporting Period: ............................................... .. April, 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 YTD Cumulative

11. Hours In Reporting Period ......... :..................... 719.0 2879.0 210383.0
12. Number of Hours Reactor Was Critical ........... 719.0 2815,0 . 145890.6
13. Reactor Reserve Shutdown Hours .................. 0.0 0.0 328.1
14. Hours Generator On-Line .......... :..................... 719.0 2808,1 . 143905.9
15. Unit Reserve Shutdown Hours .. , ..................... . 0.0 0.0 0.0
16. Gross.Thermal Energy Generated (MWH) ...... 1779183.6 _ : 7082546.5 338556803.3
17. Gross Electrical Energy Generated (MWH) ..... 596045.0 . 2376705.0 * .110827504.0
18. Net Electrical Energy Generated (MWH) ......... 576216.0 2300468:0* 105492347.0
19. Unit Service Factor. ......................................... 100.0% ,* . '97.5% 68.4%
20. Unit Availability Factor. .................................... 100.0% 97.5% 68.4%
21. Unit Capacity Factor (Using MDC Net) ............ 100.1% 99.8% 64.2%
22. Unit Capacity Factor (Using DER Net) ............ 101.7% 101.4% 63.6%
23. Unit Forced Outage Rate ................................ 0.0%
  • 2.5% 12.3%

24.. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):

Refueling, October 4, 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

  • UNIT SHUTDOWN AND POWER REDUCTION
  • Surry Monthly Operating Report No. 97-04 Page 5 of 19 (EQUAL To OR GREATER THAN 20%)

REPORT MONTH: April, 1997 Docket No.: 50-280 Unit Name: Surry Unit 1 Date: 05-05-97 Completed by: M. J. Fanguy Telephone: (804) 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 NA s 719 C NA NA NA NA Unit 1 is in RFO.

. (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)

  • UNIT SHUTDOWN AND POWER REDUCTION
  • Surry Monthly Operating Report No. 97-04 Page 6 of 19 (EQUAL To OR GREATER THAN 20%)

REPORT MONTH: April, 1997 Docket No.: 50-281 Unit Name: Surry Unit 2 Date: 05-05-97 Completed by: M. J. Fanguy Telephone: {804) 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 4/23/97 s 48 B NA NA SJ p Reduction in power to perform maintenance on 2-FW-P-1 A and 2-FW-P-1 B.

(1) ..

(2) (3)

F: Forced REASON: METHOD:

S: Scheduled A Equipment Failure (Explai~) .'. 1

  • Manual

C Refueling. 3 *

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)
  • AVERAGE DAILY UNIT POWER LEVEL
  • Surry Monthly Operating Report No. 97-04 Page 7 of 19 Docket No.: 50-280 Unit Name: Surry Unit 1 Date: 05-05-97 Completed by: J. D. Kilmer Telephone: (804) 365-2792 MONTH: April, 1997 Average Daily Power Level Average Daily Power Level Day (MWe - Net) Day (MWe - Net) 1 0 17 0 2 0 18 0 3 0 19 0 4 0 20 0 5 0 21 0 6 0 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 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.
  • AVERAGE DAILY UNIT POWER LEVEL
  • Surry Monthly Operating Report No. 97-04 Page 8 of 19 Docket No.: 50-281 Unit Name: Surry Unit 2 Date: 05-05-97 Completed by: John D. Kilmer Telephone: (804) 365-2792 MONTH: April, 1997 Average Daily Power Level Average Daily Power Level Day (MWe - Net) Day (MWe - Net) 1 831 17 820 2 831 18 821 3 831 19 821 4 833 20 823 5 832 21 823 6 831 22 824 7 830 23 486 8 831 24 433 9 831 25 820 10 832 26 824 11 831 27 824 12 832 28 820 13 826 29 820 14 823 30 818 15 821 16 820 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. 97-04 Page 9 of 19

SUMMARY

OF OPERATING EXPERIENCE MONTHNEAR: April, 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:

04/01/97 0000 Unit 1 starts month in Refueling Shutdown.

04/28/97 1226 Reactor is critical.

04/29/97 0743 Manual Rx trip due to unresolved problems with rod control.

04/30/97 2400 Unit 1 finishes the month at HSD.

UNIT TWO 04/01/97 0000 Unit 2 starts the month at 100% / 857 MWe.

04/23/97 0216 Commence power decrease to 55% for FW pump repairs.

0544 Stop power decrease at 53% / 440 MWe.

04/24/97 2139 Commence power increase from 54% / 445 MWe.

04/25/97 0237 Stop power increase, Unit is stable at 100% / 860 MWe.

04/30/97 2400 Unit 2 finishes the month at 100% / 855 MWe.

FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL Surry Monthly Operating Report No. 97-04 Page 10 of 19 MONTHNEAR: April, 1997 TM 81-97-006 Temporary Modification 4-2-97 (Safety Evaluation No.97-054)

This Temporary Modification (TM) installs a jumper on the Fuel Transfer tube gate valve limit switch. This will allow the operation of the fuel transfer carriage with the limit switch actuator removed. The limit switch's only function is to provide an indication that the valve is fully open and to provide an interlock to prevent fuel transfer cart movement if the valve is not fully open.

Operations will verify the valve is fully opened by counting the number of turns. This does not require a change in the method of operation of the fuel transfer system. Therefore it will not create a different type of accident than previously evaluated in the Safety Analysis Report (SAR) nor increase the probability of an accident. The consequences of an accident with the fuel transfer system is bounded by the fuel handling accident in the containment or fuel handling accident in the fuel building which was previously evaluated in the SAR. Therefore, an unreviewed safety question does not exist.

DCP 96-046 Design Change Package 4-4-97 (Safety Evaluation 96-011)

Design Change Package 96-046, "Steam Generator Channel Head Drain Line Isolation for Surry Unit 1," was worked due to the potential for additional leakage from the steam generator channel head drain lines. The lines were cut and capped at the steam generator connection. The modification removed the existing lines and plugged the remaining steam generator socket connection.

This modification will not increase the probability or consequences of an accident as the system was modified consistent with the system design standards and

  • codes.

Additionally, the failure of a steam generator plugged connection is bounded by previous accident analysis. Therefore, an unreviewed safety question does not exist.

FS97-011 Updated Final Safety Analysis Report Change 4-7-97 (Safety Evaluation 97-055)

Updated Final Safety Analysis Report Change FS 97-11 revised Section 11.2.4.1.2, "Expended Filter Cartridge Handling Operations," to update the UFSAR to account for current approved methodology for handling and storing expended radioactive filters.

The NRCs Technical Position on Waste Form (Rev. 1) states that encapsulation of the filter cartridge in a solidification binder or the use of a high integrity container (HIC) are acceptable options for providing stability as required by 10CFR61. HICs typically provide the most economic disposal option. HICs are approved for burial of radioactive waste under the guidance of the South Carolina Radioactive Material License. Therefore, an unreviewed safety question does not exist.

FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL

  • Surry Monthly Operating Report No. 97-04 Page 11 of 19 MONTHNEAR: April, 1997 SE 97-056 Safety Evaluation 4-7-97 The Component Cooling (CC) Surge Tank relief valve, 1-CC-RV-122, lift setpoint is scheduled for testing. To test the relief valve it must be removed from the CC Surge Tank. The CC System is required to remain in operation when either unit is above 350°F/450 psig.

To minimize the consequences of the surge tank relief valve being removed when the system is required, a tested spare relief valve will be on hand to be installed as soon as the surge tank relief valve is removed. Instrumentation and alarms are in place to alert an operator of any problem associated with the CC Surge Tank. Operations can isolate the CC Surge Tank upon such notification. The CC System is not used for accident mitigation. Therefore, an unreviewed safety question does not exist.

DCP 96-050 Design Change Package 4-8-97 (Safety Evaluation 97-021)

Design Change Package 96-050, "Service Water (SW) Installation of Vent Lines for Component Cooling Heat Exchangers Piping," installed vent lines upstream of 1-SW-MOV-102A/B. This was done to eliminate air in the SW lines which accumulates from the divers' breathing apparatus during the SW line inspection.

The vent valves are safety related and seismically qualified. The installation of the vent valves did not change the operation or ability of the Service Water System to perform its safety function nor did it reduce the margin of safety as defined in the basis of the Technical Specification. Therefore, an unreviewed safety question does not exist.

DCP 97-009 Design Change Package 4-9-97 (Safety Evaluation 97-039)

Design Change Package 97-009, "Reactor Coolant Pump Modification (1-RC-P-1A),"

installed jacking bolts to facilitate the alignment of the motor stand assembly with the pump. Additionally, the floating ring seal housing had its outside diameter machined to a diameter to provide sufficient reassembly clearance without compromising the function of the backup ring/0-ring combination on the lower floating ring seal housings.

These modifications will not affect pump operation, performance or integrity. Therefore, an unreviewed safety question does not exist.

DCP 95-013 Design Change Package 4-10-97 (Safety Evaluation 96-128)

Design Change Package 95-013, "Safety Injection System Vent Additions," installed additional system high point vents to improve the venting of gasses that have the potential to gas bind or cause pressure transients in the charging/High Head Safety Injection or Low Head Safety Injection pumps.

The vent valves are safety related and seismically qualified. The installation of the vent valves did not change the operation or ability of the Safety Injection System to perform its safety function nor did it reduce the margin of safety as defined in the basis of the Technical Specification. Therefore, an unreviewed safety question does not exist.

FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL

! Monthly Operating Report No. 97-04 Page 12 of 19 MONTHNEAR: April, 1997 DCP 96-004 Design Change Package 4-12-97 (Safety Evaluation 96-042 Rev. 1)

Design Change Package 96-004, "RC Pressurizer Insulation Replacement," installed new insulation on the pressurizer below the 55'-6" elevation to the 24'-9 1/2" elevation.

The heat loss of the new insulation is slightly less than the original design and will not adversely affect operation of the pressurizer or increase containment temperatures. The insulation system is seismically supported. The new insulation system will not increase the potential for containment sump blockage following a LOCA. Therefore, an unreviewed safety question does not exist.

TM 81-97-007 Temporary Modification 4-13-97 (Safety Evaluation No.97-059)

This Temporary Modification {TM) is required to clear a failed PC Card slot for annunciator F-G-3 " FW HTR 6A HI-HI LEVEL." The TM abandons annunciator F-G-3 and moves the alarm to location F-H-3" FW HTR 68 HI-HI LEVEL." This window will now alarm with a HI-HI signal from the 6A or 68 6th point heater.

There is no change to any protection circuit nor is there any change to the isolation between the protection and control systems. The TM will allow the "F" window first out alarms to function as designed with no loss of any alarms. Therefore, an unreviewed safety question does not exist.

SE 97-066 Safety Evaluation 4-16-97 The pressurizer vent solenoid valves were noted to be leaking by at approximately 20 drops per minute. To isolate this leakage path, isolation valve 1-RC-133 has been closed. This defeats the function of these valves and isolates the Pressurizer vapor space sample line and also defeats the function of the Pressurizer vent valve flowpath.

This condition is not a concern since the valves are not addressed as: accident mitigation equipment in Technical Specification {TS) 3.7, nor are the valves required to be operable per TS 3.1.A. The valves are utilized only during implementation of 1/2-FR-H.1, " Loss of Hear Sink," or 1/2-FR-C.1, " Response to Inadequate Core Cooling." These conditions are beyond the design basis of the plant. Other vent flowpaths, which are required to be operable per TS 3.1.A, exist to provide this function (Pressurizer PORVs/Reactor Head Vents). Therefore, an unreviewed safety question does not exist.

TM 81-97-002 Temporary Modification 4-24-97 (Safety Evaluation No.97-062)

This Temporary Modification {TM) installs a jumper to provide enhanced cooling for the out of service Unit 2 Main Feedwater (MFW) Pump 1A during seal maintenance.

The probability and consequences of a Loss of Feedwater, Loss of Condenser Vacuum, and Excessive Dissolved Oxygen are not increased. The Loss of Feedwater event is not affected by vacuum drag on an out-of-service MFW Pump. A reduction in condenser vacuum or unacceptable increase in dissolved oxygen will be terminated by securing the vacuum drag lineup. The possibility of a new or different type of accident or malfunction is not created because the TM creates no new failure modes or accident precursors. The margin of safety as described in the Technical Specification Bases is not affected.

Therefore, an unreviewed safety question does not exist.

FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL t Monthly Operating Report No. 97-04 Page 13 of 19 MONTHNEAR: April, 1997 FS97-021 Updated Final Safety Analysis Report Change 4-24-97 (Safety Evaluation 97-063)

Updated Final Safety Analysis Report Change FS97-021 revised Section 10.3.1.2," Main Steam System Description." It was determined during a review of this section that the qualifications for the Appendix R solenoid operated valves associated with the Main Steam trip valves were incorrectly stated as meeting the requirements of 10CFR50.49.

The valves are not safety related and are not required to operate in a harsh environment, therefore 10CFR50.49 does not apply.

This UFSAR change was implemented to accurately reflect the environmental qualification of the subject valves. The change did not involve any physical changes to plant equipment. Therefore, an unreviewed safety question does not exist.

DCP 90-029 Design Change Package 4-29-97 (Safety Evaluation 90-183)

Design Change Package 90-029, "Installation of Lifting Anchors for Removable Concrete Plugs in Safeguards," installed Maxi-Bolt coupling anchors to replace the existing deteriorated anchors to provide a safer means of lifting the concrete plugs.

This Design Change Package is consistent with the original station design criteria.

Therefore, an unreviewed safety question does not exist.

FS96-051 Updated Final Safety Analysis Report Change 4-30-97 (Safety Evaluation 97-065)

Updated Final Safety Analysis Report Change FS96-051 revised Section 15.5, "Specific Containment Structural Designs," to establish the test method of containment electrical penetrations and upgrade the description of the penetrations, to account for the various manufacturers' designs .

.* This change provides clarification of the containment electrical penetration seal design by manufacture type and description of the test method for electrical penetrations to demonstrate compliance with 10 CFR50 Appendix J. The equipment operation or system, as described in the Safety Analysis Report, will not be altered by the editorial changes presented by the UFSAR Change Request. Each containment penetration will continue to function to mitigate the consequences of a Design .Basis Accident or other station accidents. Therefore, an unreviewed safety question does not exist.

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

' r r y Monthly Operating Report No. 97-04 Page 14 of 19 MONTHNEAR: April, 1997 1-0P-CH-013 Rev. 2 Operating Procedure 4-7-97 (Safety Evaluation 97-045 Rev. 1)

Operating Procedure 1-0P-CH-013, "Removal From and Return To Service of eves Deborating Demineralizers," was temporarily changed to provide instructions for loading demineralizer 1-CH-l-3B with mixed bed resin to remove sulfates from the Unit 1 primary side. This demineralizer's normal function is to de-borate the RCS at end of life.

This demineralizer is compatible for this function in that it satisfies the temperature, pressure, and flow requirements of the system. The same precautions will be used as those for placing a fresh mixed bed ion exchanger in service. All other ion exchange vessels that contain resin will be isolated during the cleanup. Periodic sampling of the RCS will be performed to verify effectiveness of the cleanup and verify that RCS boron concentration is not affected by this evolution. This evolution will be performed with the unit at Cold Shutdown/Refueling Shutdown conditions and up to and including Hot Shutdown.

Therefore, an unreviewed safety question does not exist.

1/2-NSP-RX-005 Engineering Surveillance Procedure 4-11-97 (Safety Evaluation 97-058)

Engineering Surveillance Procedure 1/2-NSP-RX-005, "RPI Calibration Data Collection,"

was revised to allow the movement of one bank of control rods at a time but allowing a total of three banks of control rods to be out of the core at the same time. This was done to be consistent with the time requirement to reach Rod Position Indication (RPI) equilibrium per Westinghouse guidelines.

Technical Specification 3.12 requires RPls to be operable from rod withdrawal to critical through power operations. The testing phase of this activity will be performed at hot shutdown, where no Limiting Condition ,for Operation applies. The same section requires operable. control rods. This procedure will not take any control rods out of service. It will withdraw rods, but they will be trippable at all times. Additionally, a shutdown margin

. calculation has been performed and .verifies.the core will remain adequately subcritical at the most limiting rodded configuration. Therefore, an unreviewed safety question does not exist.

1-NPT-RX-014 Engineering Surveillance Procedure 4-11-97 (Safety Evaluation 97-057)

Engineering Surveillance Procedure 1/2-NSP-RX-014, "Hot Rod Drops By Banks," had a One Time Only Change submitted to collect data to determine the feasibility of using the RPI Primary coils' output versus the Secondary coils' output. This would determine the feasibility of a corrective maintenance method for Individual Rod Position Indication (IRPI) spiking, and ultimately to modify the circuit to reduce erratic IRPI indication.

The initial conditions for procedure 1/2-NSP-RX-014 require that the reactor be adequately subcritical at the most limiting rodded configuration. In addition, the procedure's Precautions and Limitations section prohibits dilution during the procedure's performance.

Technical Specification 3.12 requires RPls to be operable from rod withdrawal to critical through power operations. This activity will be performed at hot shutdown, where no Limiting Condition for Operation applies. Therefore, an unreviewed safety question does not exist.

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

~rry Monthly Operating Report No. 97-04 Page 15 of 19 MONTHNEAR: April, 1997 1-EMT-1508-11 Electrical Maintenance Test 4-15-97 1-EMT-1508-12 (Safety Evaluation 97-060)

Electrical Maintenance Test 1-EMT-1508-11/12, "1-SI-MOV-1855B/C Interlock Test," installs a procedurally controlled Temporary Modification (TM) to clear the currently illuminated 1A-D4, "SI Valve Out of Position," annunciator. The annunciator is lit because the valves are in the closed position due to the plant being shutdown. The TM will allow the verification of the annunciator function for valves 1-SI-MOV-1855B/C. The TM is installed and removed in accordance with 1-EMT-1508-11 and 1-EMT-1508-12 while the plant is shutdown and safety injection is not required.

The activity will not affect any previously analyzed accidents or equipment malfunctions.

The activity will be performed in accordance with a procedurally controlled TM and the unit's reactor operator will have valve position indication for the accumulator discharge valves, which are currently closed and de-energized. Therefore, an unreviewed safety question does not exist.

IMP-C-EH-31 Instrument Maintenance Procedure 4-28-97 (Safety Evaluation 97-064)

Instrument Maintenance procedure IMP-C-EH-31, "AEH Controller Checkout," was revised to incorporate the installation of a procedurally controlled temporary modification. The modification prevents the main turbine gear from disengaging when required by plant conditions and allows input power to be applied to the secondary power supplies for.

calibration because the main generator permanent magnet generator (PMG) supply is unavailable.

The potential for a reactor. trip is .precluded by the procedure's initial .conditions requiring the turbine generator to be off line. Failure of the jumper installation such that the associated circuits are grounded will have no significant effect. The EHC power supply is isolated from the semi-vital bus by an individual circuit breaker in the EHC cabinet. Grounding the jumper associated with the turning gear interlock would result in no damage to the EHC relay card.

This circuit is also isolated from the semi-vital bus power supply by an individual circuit breaker. Should failure of the activity occur, operation of the plant would be bounded by current accident analyses. Therefore, an unreviewed safety question does not exist.

  • l Monthly Operating Report No. 97-04 Page 16 of 19 TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTHNEAR: April, 1997 None During the Reporting Period
  • l Monthly Operating Report No. 97-04 Page 17 of 19 CHEMISTRY REPORT MONTHNEAR: April, 1997 Unit No. 1 Unit No. 2 Primary Coolant Analysis Max. Min. Avg. Max. Min. Avq.

Gross Radioactivity, µCi/ml 9.95E-2 7.84E-4 1.31E-2 2.64E-1 1.05E-1 1.96E-1 Suspended Solids, ppm 0.250 £0.010 0.051 - - -

Gross Tritium, µCi/ml 1.24E-2 1.24E-2 1.24E2 4.99E-1 3.88E-1 4.55E-1 1131, µCi/ml 3.97E-5 6.49e-6 1.81E-5 2.86E-4 2.60E-5 7.35E-5 113111133 - - - 0.11 0.05 0.09 Hydroqen, cc/kq 28.8 2.3 15.6 32.5 27.5 30.2 Lithium, oom 3.63 <0.1 0.78 2.34 2.06 2.21 Boron - 10, ppm* 508.4 190.1 431.9 161.9 132.1 142:2 Oxvqen, (DO), ppm 7.0 £0.005 2.3 * £0.005 £0.005 £0.005 Chloride, oom 0.040 <0.001 0.003 0.004 0.002 0.003 pH at 25 degree Celsius 5.78 .4,10 * -4:6fr 6.58 6.31 6.44

None

  • FUEL HANDLING I Monthly Operating Report No. 97-04 Page 18 of 19 UNITS 1&2 MONTHNEAR: April, 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 No Fuel Received or Stored During the Reporting Period

Surry Monthly Operating Report No. 97-04 Page 19 of 19 DESCRIPTION OF PERIODIC TEST(S) WHICH WERE Nor COMPLETED WITHIN THE TIME LIMITS SPECIFIED IN TECHNICAL SPECIFICATIONS MONTH/YEAR: APRIL, 1997 None During the Reporting Period

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