ML18152A405
| ML18152A405 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/31/1997 |
| From: | Fanguy M, Mason D, Sarver S VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 97-226, NUDOCS 9704220029 | |
| Download: ML18152A405 (23) | |
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e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 April 14, 1997 United States ~uclear 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.97-226 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 March 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 101 Marietta Street, N.W.
Suite 2900 Atlanta, Georgia 30323
- Mr. R. A. Musser NRC Senior Resident Inspector Surry Power Station
-1
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11111111111111111 IIIII IIIIII IIIII IIIIIII Q
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VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION MONTHLY OPERATING REPORT REPORT No. 97-03 Approved:
j\\ !CO- -~'r-~13,--=r-J Station Manager Date
e TABLE OF CONTENTS Section
~urry Monthly Operating Report No. 97-03 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................................................. 14 Tests and Experiments That Did Not Require NRC Approval.................................................................................. 19 Chemistry Report..................................................................................................................................................... 20 Fuel Handling - Unit No. 1........................................................................................................................................ 21 Fuel Handling - Unit No. 2........................................................................................................................................ 21 Description of Periodic Test(s) Which Were Not Completed Within the Time Limits Specified in Technical Specifications....................................................................................................................... 22
OPERATING DATA REPORT
- 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 March, 1997 2546 847.5 788' 840 801 e,,urry Monthly Operating Report No. 97-03 Page 3 of 22 Docket No.:
Date:
Completed By:
Telephone:
50-280 04/03/97 D. K. Mason (804) 365-2459
- 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 2160.0 212784.0
- 12.
Number of Hours Reactor Was Critical...... :.....
148.8 1300.8 148135.5
- 13.
Reactor Reserve Shutdown Hours..................
0.0 0.0 3774.5
- 14.
Hours Generator On-Line................................
147.8 1278.3 145809.3
- 15.
Unit Reserve Shutdown Hours........................
0.0 0.0 3736.2
- 16.
Gross Thermal Energy Generated (MWH)......
355258.6 3190453.6 341825897.4
- 17.
Gross Electrical Energy Generated (MWH).....
118290.0 1065255.0 112038073.0
- 18.
Net Electrical Energy Generated (MWH).........
113965.0 1029190.0 106622939.0
- 19.
Unit Service Factor..........................................
19.9%
59.2%
68.5%
- 20.
Unit Availability Factor.....................................
19.9%
59.2%
70.3%
- 21.
Unit Capacity Factor (Using MDC Net)............
19.1%
59.5%
64.5%
- 22.
Unit Capacity Factor (Using DER Net)............
19.4%
60.5%
63.6%
- 23.
Unit Forced Outage Rate.................................
0.0%
18.3%
15.2%
- 24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):
Refueling, March 6, 1997, 41 Days
- 25. If Shut Down at End of Report Period, Estimated Date of Start-up:
- 26. Unit In Test Status (Prior to Commercial Operation):
INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION FORECAST 04/16/97 ACHIEVED
e OPERATING DATA REPORT
- 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 March, 1997 2546 847.5 788 840 801
.,,urry Monthly Operating Report No. 97-03 Page 4 of 22 Docket No.:
Date:
Completed By:
Telephone:
50-281 04-03-97 D. K. Mason
{804) 365-2459
- 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 2160.0 209664.0
- 12.
Number of Hours Reactor Was Critical............
744.0 2096.0 145171.6
- 13.
Reactor Reserve Shutdown Hours..................
0.0 0.0 328.1
- 14.
Hours Generator On-Line................................
744.0 2089.1 143186.9
- 15.
Unit Reserve Shutdown Hours........................
0.0 0.0 0.0
- 16.
Gross Thermal Energy Generated (MWH)......
1893918.5 5303362.9 336777619.7
- 17.
Gross Electrical Energy Generated (MWH).....
636690.0 1780660.0 110231459.0
- 18.
Net Electrical Energy Generated (MWH).........
617663.0 1724252.0 104916131.0
- 19.
Unit Service Factor..........................................
100.0%
96.7%
68.3%
- 20.
Unit Availability Factor.....................................
100.0%
96.7%
68.3%
- 21.
Unit Capacity Factor (Using MDC Net)............
103.6%
99.7%
64.1%
- 22.
Unit Capacity Factor (Using DER Net)............
105.4%
101.3%
63.5%
- 23.
Unit Forced Outage Rate.................................
0.0%
3.3%
12.4%
- 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
(1)
Date Type 3f7/97 s
(1)
F:
Forced S:
Scheduled (4) e UNIT SHUTDOWN AND POWER REDUCTION (EQUAL To OR GREATER THAN 20%)
REPORT MONTH: March, 1997 tlurry Monthly Operating Report No. 97-03 Page 5 of 22 Docket No.:
50-280 Unit Name:
Surry Unit 1 Date:
04-01-97 Completed by:
M. J. Fanguy
- Telephone:
(804) 365-2155 (2)
(3)
(4)
(5)
Method Duration of LER No.
System Component Cause & Corrective Action.
Hours 596.2 Reason C
(2)
REASON:
Shutting Down Rx 2
NA A
Equipment Failure (Explain)
B Maintenance or Test C
Refueling D
Regulatory Restriction Code NA E
Operator Training & Licensing Examination F
Administrative G
Operational Error (Explain)
Code to Prevent Recurrence NA Manual reactor trip in accordance with 1-GOP-2.3 (3)
METHOD:
1 -
Manual 2 -
Manual Scram 3 -
Automatic Scram 4 -
Other (Explain)
(5)
Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report {LER) File (NUREG 0161)
Exhibit 1 - Same Source
(1)
Date Type NA NA (1)
F:
Forced S:
Scheduled (4)
(2)
UNIT SHUTDOWN AND POWER REDUCTION (EQUAL To OR GREATER THAN 20%}
REPORT MONTH: March, 1997
~urry Monthly Operating Report No. 97-03 Page 6 of 22 Docket No.:
50-281 Unit Name:
Surry Unit 2 Date:
04-01-97 Completed by:
M. J. Fanguy Telephone:
(804) 365-2155 (3)
(4)
(5)
Method Duration of LER No.
System Component Cause & Corrective Action Hours NA Reason NA (2)
REASON:
Shutting Down Rx NA NA A -
Equipment Failure (Explain)
B Maintenance or Test C
Refueling D
Regulatory Restriction Code NA E
Operator Training & Licensing Examination F
Administrative G
Operational Error (Explain)
Code to Prevent Recurrence NA (3)
METHOD:
1 -
Manual NA 2 -
Manual Scram 3 -
Automatic Scram 4 -
Other (Explain)
(5)
Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report (LER) File (NUREG 0161)
Exhibit 1 - Same Source
MONTH:
March, 1997 Day 2
3 4
5 6
7 8
9 10 11 12 13 14 15 16 INSTRUCTIONS lturry Monthly Operating Report No. 97-03 Page 7 of 22 AVERAGE DAILY UNIT POWER LEVEL Average Daily Power Level (MWe - Net) 809 799 780 784 775 760 18 0
0 0
0 0
0 0
0 0
Day 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 Docket No.:
50-280 Unit Name:
Surry Unit 1 Date:
04-05-97 Completed by: J. D. Kilmer Telephone:
(804) 365-2792 Average Daily Power Level (MWe - Net) 0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 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.
MONTH:
March, 1997 Day 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 INSTRUCTIONS e
tlurry Monthly Operating Report No. 97-03 Page 8 of 22 AVERAGE DAILY UNIT POWER LEVEL Average Daily Power Level (MWe-Net)
Day 819 17 825 18 828 19 828 20 828 21 827 22 827 23 827 24 828 25 830 26 832 27 831 28 832 29 832 30 833 31 833 Docket No.: 50-281 Unit Name:
Surry Unit 2 Date: 04-05-97 Completed by:
John D. Kilmer Telephone:
(804) 365-2792 Average Daily Power Level (MWe - Net) 833 833 833 833 833 832 832 832 831 831 832 831 830 831 830 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
SUMMARY
OF OPERATING EXPERIENCE MONTH/YEAR: March, 1997
~urry Monthly Operating Report No. 97-03 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:
03/01/97 03/06/97 03/07/97 UNITTWO 03/01/97 03/31/97 0000 2205 0349 Unit 1 starts the month at 98.5%, 842 MWe, in coastdown for refueling outage.
Commence ramp down from 93.5 %, 800 MWe.
Unit 1 is off-line.
0445 Manual reactor trip in accordance with 1-GOP-2.3.
0000 2400 Unit 2 starts the month at 100% / 855 MWe.
Unit 2 finishes the month at 100% / 857 MWe.
FS 97-03 SE 97-031 TM 81-97-001 e
lturry Monthly Operating Report No. 97-03 Page 10 of 22 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/VEAR: March, 1997 Updated Final Safety Analysis Report Change (Safety Evaluation 97-012) 2-3-97 Updated Final Safety Analysis Report Change FS 97-03 revised Section 9.1.2.6.23, "Chemical and Volume Control System," to correct a statement that "the normal blender flowpath and the boric acid transfer pumps in which the boric acid solutions are rotated on at least a weekly basis... " There is no routine rotation of flowpaths. The flowpaths, and pumps are aligned as required for Technical Specification compliance, Preventative Maintenance, Corrective Maintenance, and ASME Section XI inservice testing...This testing satisfies the Technical Specification required testing.
This change was made to accurately reflect the current plant condition. None of the changes to the UFSAR increases the probability of occurrence or increases the consequence of an accident. Therefore, an unreviewed safety question does not exist.
Safety Evaluation 3-3-97 Safety Evaluation 97-031 was performed to evaluate the 1997 Unit 1 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 1 and (b) the effects of Unit 1 outage activities on critical safety functions for Unit 2. Therefore, an unreviewed safety question does not exist.
Temporary Modification (Safety Evaluation No.97-035) 3-6-97 This Temporary Modification (TM) is required during the time the Main Board Racks six and eight are being upgraded to the Foxboro I/A instrument system. The TM will maintain indication in the Main Control Room for the Containment Sump Level and the Containment Instrument Air Pressure System.
The TM will be installed in the secondary non-safety instrument racks, with the unit at Cold Shutdown. There is no change to the existing design or installation of the station isolation between primary and secondary instrument systems and, therefore, no feedback into any primary plant instrument system. It has been verified that the TM will not alter the operation of the Residual Heat Removal System.
Therefore, an unreviewed safety question does not exist.
~urry Monthly Operating Report No. 97-03 Page 11 of 22 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: March, 1997 Safety Evaluation 3-7-97 Safety Evaluation 97-037 was performed to evaluate the closure of valve 1-VS-606 so the Unit 2 Containment Air Compressor Cubicle dampers 2-VS-AOD-205 A&B can be refurbished.
During maintenance of the 2-VS-AOD-205 A&B dampers, the Unit 1 Safeguard Building Ventilation will require manual alignment in the event of a Unit 1 Safety Injection actuation.
Temporary Modification (TM) S1-96-09 installed electrical jumpers and switches to provide operators with the ability to reposition motor-operated dampers 1-VS-MOD-100A[B] and 2-VS-MOD-200A[B] in the event of a reduction of air pressure below the 65 psig lock-out point.
Emergency Operating Procedures 1-E-O, "Reactor Trip or Safety Injection," and 1-ES-1.1, "SI Termination," were revised to identify the subject switches and their appropriate positions.
In the unlikely event of a loss of air pressure prior to a safety injection, the TM allows the dampers to be correctly positioned. Therefore, an unreviewed safety question does not exist.
Design Change Package (Safety Evaluation 93-183) 3-10-97 Design Change Package 92-092, "Excore Shroud Dosimetry," installed dosimetry in two excore detector shrouds located within the neutron shield tank. This installed neutron dosimetry will permit data to be obtained to improve the accuracy of the reactor vessel fast neutron fluence calculations and allow full credit to be taken for the actual flux reduction achieved by flux suppression inserts.
The Excore Instrumentation System monitors the power level in the reactor to initiate protective action, alarms and interlocks via the Reactor Protection System when unsafe nuclear conditions are approached. The installation of the dosimetry into the excore shrouds will not prevent the excore instrumentation from performing its intended functions. The excore shroud dosimetry hardware is designated non-safety related with special quality requirements. The dosimetry is not required for safe shutdown of the unit and failure of the hardware will not damage safety related equipment. Therefore, an unreviewed safety question does not exist.
Updated Final Safety Analysis Report Change (Safety Evaluation 97-042) 3-13-97 Updated Final Safety Analysis Report Change 97-17 revises Section 9.A, "High-Density Spent Fuel Storage Rack Design," and Section 14.4.1, "Fuel Handling Accidents" to change the height fuel is handled above the fuel storage racks. Currently the UFSAR indicates this height as 24 inches verses the actual height of 42 inches. The difference in height resulted from changes made to the handling tool and differences in load cell length that have been implemented over the past several years.
The margin of safety of any part of the Technical Specifications as described in the bases section is not reduced as a result of these changes to the UFSAR. The drop of a fuel assembly onto the fuel storage racks from a height of 42 inches versus 24 inches are bounded by the analysis of a drop of the transfer canal gate and will not change the center-to-center spacing of the stored fuel assemblies.
The consequences of a fuel handling accident in the fuel building remain unchanged. The makeup capability to the pool exceeds the expected leak rate from the test channels in the event the spent fuel pool liner is punctured. Therefore, an unreviewed safety question does not exist.
DCP 96-002 TM 81-97-003 DCP 95-030 4'urry Monthly Operating Report No. 97-03 Page 12 of 22 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: March, 1997 Design Change Package (Safety Evaluation 96-119 Rev. 1) 3-17-97 Design Change Package 96-002, "In-Mast Sipping Modification," (IMS) installs the in-mast sipping hardware to increase the efficiency and accuracy of the detection of failed fuel assemblies.
The displacement of fluid inside the crane mast does not increase the level of exposure to the manipulator crane operator. The additional weight added to the manipulator crane by the Design Change Package has been seismically analyzed by the vendor and has demonstrated that the manipulator crane and IMS hardware components are structurally adequate to withstand the seismic dead weight loads. The IMS System does not affect any safety related equipment, or change the function/operation of the manipulator crane.
Therefore, an unreviewed safety question does not exist.
Temporary Modification (Safety Evaluation No.97-050) 3-24-97 During the ongoing Unit 1 outage the single point failure relays for train 'A' and 'B' HI-CLS are to be replaced. To accomplish this both trains of HI-CLS fuses will be pulled and the Safety Injection (SI) actuation from HI-CLS will be blocked by tag out. Operations has requested that a Temporary Modification (TM) be installed to block the closing of 1-TV-RM-100A, B, and C. This TM will block the HI-CLS signal to these valves and allow the gas/particulate radiation monitors to function during the time the relays are being replaced.
Work will be performed with the unit at cold shutdown when the automatic closure of trip valves is not required.
During the time the relays are being replaced, the temporary modification will allow the trip valves to remain open with the HI-CLS system tagged out.
Therefore, an unreviewed safety question does not exist.
Design Change Package (Safety Evaluation 95-120) 3-26-97 Design Change Package 96-002, "Replacing of Motor Operators," replaces the SMB-000 motor operators on 1-S1-MOV-1860A/B and 2-SI-MOV-2860A/B with SMB-00 motor operators to provide a greater safety margin in the operators' capability to perform their design function.
This modification is essentially a like-for-like replacement with the exception of the increase in operator capacity and valve stroke time. The increase in valve stroke time has been previously analyzed and does not adversely affect the sequencing for a plant shift to Recirculation Mode Transfer.
Previously analyzed accidents and equipment malfunctions concerning these MOVs are not affected nor is there an increase in the probability of occurrence or the consequences of the accidents/equipment malfunctions.
Therefore, an unreviewed safety question does not exist.
~urry Monthly Operating Report No. 97-03 Page 13 of 22 FACILITY CHANGES THAT DID Nor REQUIRE NRC APPROVAL MONTH/YEAR: March, 1997 Safety Evaluation 3-28-97 A Safety Evaluation has been performed to determine whether an unreviewed safety question will result from the refueling and operation of Surry Unit 1 Cycle 15.
The Cycle 15 core design parameters were calculated and compared to the existing safety analysis assumptions. These parameters were shown to be either 1) explicitly bounded or 2) accommodated by existing safety analysis margin and/or conservatism.
Operation of the reload core in accordance with the Technical Specifications will not violate the design basis of plant safety equipment.
Thus, the probabilities and consequences of analyzed accidents and equipment malfunctions are not changed by the reload. Therefore, an unreviewed safety question does not exist.
Design Change Package (Safety Evaluation 96-076 Rev. 1) 3-29-97 Design Change Package 96-020, "Inadequate Core Cooling Monitor Firmware Modifications," revises the configuration data to allow the Inadequate Core Cooling Monitor {ICCM) to operate with two or more operable thermocouples in a quadrant in order to maintain a good quality average.
The modification replaces two system electronic EPROMs and CPUs with a new CPU containing revised firmware for each train.
This modification was needed since a failure of a thermocouple in some quadrants of the core would result in a ICCM failure and entry into a seven day Limiting Condition of Operation to Hot Shutdown.
This change will not affect the measurements, calculations, or setpoints for the ICCM.
The Safety Analysis Report and Technical Specifications only require two operable thermocouples per quadrant per train. Since the Technical Specification requirements and design criteria continue to be satisfied, the margin of safety is not reduced.
Therefore, an unreviewed safety question does not exist.
ROP-1.87 O-TMOP-3040 O-MOP-RM-001 1 (2)-0SP-SW-007 1 (2)-0SP-SW-008 e
e Surry Monthly Operating Report No. 97-03 Page 14 of 22 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: March, 1997 Radwaste Operating Procedure (Safety Evaluation 97-014) 2-3-97 Radwaste Operating Procedure ROP-1.87, "Transferring Liquid Waste from Evaporator Bottoms Tank to a Strong Tight Container," was revised to install a temporary hose. This hose will allow the transfer of liquid waste from the evaporator bottoms tank to a strong tight container at the Surry Radwaste Facility (SRF).
This will be a procedurally controlled temporary modification.
None of the components are safety related. Pressure and temperature rating of the hose exceeds the operating parameters of the evaporator bottoms tank pump.
The hose construction material is compatible with the pumped fluid such that the fluid will not degrade the hose. The SRF has provisions to assure that any liquid spill is contained within the building, limiting the environmental exposure. Any gaseous effluent will be captured by the SRF ventilation system. The margin of safety as stated in the Technical Specifications is not affected by this temporary modification. Therefore, an unreviewed safety question does not exist.
Temporary Maintenance Operating Procedure Maintenance Operating Procedure (Safety Evaluation No.97-032) 3/3/97 Temporary Maintenance Operating Procedure O-TMOP-3040, "Unblocking Ventilation Vent Rm Nozzle," and/or Maintenance Operating Procedure O-MOP-RM-001, "Unblocking Ventilation Vent Rm Nozzle," will install a purge air supply to the downstream side of the Ventilation Vent lsokinetic Nozzle at the input of the Ventilation Vent Monitor for the purpose of flushing the Vent Vent multi ported isokinetic nozzle. The reverse flushing of the nozzle is needed to remove the clogging that is preventing the drawing of a normal representative sample from the Ventilation Vent 2 effluent discharges.
Work will be completed within 45 minutes to satisfy the one hour clock specified in the ODCM.
In the unlikely event that the allowed outage time is exceeded, the required compensatory actions stated in the OCDM will be implemented. Therefore, an unreviewed safety question does not exist.
Operations Surveillance Procedure (Safety Evaluation No 96-048 Rev. 1) 3-3-97 1 (2)-0SP-SW-007 and 1 (2)-0SP-SW-008, "Service Water Flow Test of Recirculating Spray Heat Exchangers," initiate service water flow to the Recirculating Spray Heat Exchangers (RSHXs) to collect performance data and verify operability of the system. To collect this data, wires will be connected to the Service Water (SW) flow instruments at the RSHXs discharge. Additionally, these procedures require administrative control of the SW isolation valves to ensure canal inventory is maintained.
These procedures will be performed with the Reactor Coolant System conditions less than 350°F and 450 psig when the operability of the Recirculating Spray System is not required.
Plant operators will maintain control of the test by administrative control of the SW isolation valves to ensure canal inventory is maintained for the opposite unit should an accident occur. Therefore, an unreviewed safety question does not exist.
1 [2]-PT-8.1 1 [2]-PT-8.2 1 [2]-IPT-RP-AFW-001 1 (2)-0P-RS-007 1 (2)-0P-RS-008 1 (2)-MOP-EP-204 1 (2)-MOP-EP-205 e Surry Monthly Operating Report No. 97-03 Page 15 of 22 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: March, 1997 Instrument Periodic Test Procedures (Safety Evaluation No.93-075, Revision 3) 3-6-97 Instrument Periodic Test Procedures 1 [2]-PT-8.1, "Reactor Protection System Logic (for Normal Operations)," and 1[2]-PT-8.2, "Reactor Protection Logic," were revised and Instrument Periodic Test Procedure 1 [2]-IPT-RP-AFW-001, "Under Voltage and Low Low Steam Generator Logic Start of the Steam Driven Auxiliary Feed Pump," was developed to incorporate administrative control steps for the steam generator (SG) blowdown permissive key switches to ensure proper operation of the SG blowdown trip valves.
This change places the SG blowdown permissive key switches under administrative control when in the permissive mode during monthly logic testing. The procedural steps direct an operator to place the key switches in the normal position in the event of an auxiliary feedwater system automatic start signal.
This action ensures operability of the SG blowdown trip valves. Therefore, an unreviewed safety question does not exist.
Operating Procedure 3-6-97 (Safety Evaluation No.97-034)
Operating Procedures 1 (2)-0P-RS-007 I 1 (2)-0P-RS-008, "Simultaneous Slowdown of Recirculating Spray Heat Exchanger Service Water Piping 1-RS-E-1 A & 1 D / 1-RS-E-1 B &
1 C," perform a procedurally controlled temporary modification to apply air pressure to the heat exchangers to enhance the draining of the heat exchangers.
Each procedure uses air pressure to assist in draining the heat exchangers. The amount of air pressure will be kept below the lowest pressure class component in the portion being drained. Additionally, these procedures will be performed with the Reactor Coolant System conditions less than 350°F and 450 psig when the operability of the Recirculating Spray System is not required. Therefore, an unreviewed safety question does not exist.
Maintenance Operating Procedure (Safety Evaluation No.97-036) 3-6-97 Maintenance Operating Procedures 1 (2)-MOP-EP-204 / 1 (2)-MOP-EP-205, "Removing /
Returning 4160 Bus 1 H, 480V Bus 1 H and 1 H-1, and 480V MCC 1 H-1 and 1 H-2," install and remove temporary modifications which provide power from a like source to the Radiation Monitoring Equipment and the Fire Detection Equipment from the opposite unit's power supplies during bus outages.
The power sources that will be supplying power to the Radiation Monitoring Equipment and the Fire Detection Equipment are adequately sized to supply both units. The procedures that direct the bus outages have steps to verify that the Radiation Monitoring Equipment and the Fire Detection Equipment are operable after installation and removal of the temporary modification. Therefore, an unreviewed safety question does not exist.
-~I
1 (2)-0P-VS-001 1-MOP-VS-16 1-MOP-VS-17 1-EMP-P-RT-23 1-EMP-P-RT-47 VPAP-2201 e
e Surry Monthly Operating Report No. 97-03 Page 16 of 22 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: March, 1997 Operating Procedure Maintenance Operating Procedure (Safety Evaluation No.97-040) 3/12/97 To establish proper control of airborne isotope migration during refueling outages, the containment purge must be operated in modes not currently addressed in the Surry UFSAR.
A mechanical jumper is installed between containment purge and auxiliary building general exhaust filter 1-VS-FL-14. The jumper is installed to minimize the refueling impact on safety related filters by allowing containment purge to flow through a non-safety charcoal filter except when fuel is being handled inside containment.
This lineup does not impact the ability of the safety-related filters to respond to a unit's safety injection signal. Additionally, any gasses released by a failed jumper would remain within the Auxiliary Building and be monitored prior to release. Therefore, an unreviewed safety question does not exist.
Corrective Electrical Maintenance Procedure (Safety Evaluation No.97-046) 3-13-97 Corrective Electrical Maintenance Procedures 1-EMP-P-RT-23 /
1-EMP-P-RT-47, "Protective Relay Maintenance for Circuit Breaker 15J8/15D1 Normal Feed/Reserve Supply to Bus 1 J/1 D," were revised to provide instructions for the installation and removal of procedurally controlled Temporary Modifications (TM).
These procedurally controlled TMs allow circuit breaker control interlocks to be functionally tested while maintaining reliable and desired electric plant lineup. The emergency power supplies are not affected by the installation and removal of these TMs. These procedures are performed at cold shutdown and the TMs are installed and removed by qualified individuals using double verification. Therefore, an unreviewed safety question does not exist.
Administrative Procedure (Safety Evaluation 97-043) 3-13-97 Administrative Procedure VPAP-2201, "Nuclear Plant Chemistry Program," was revised to be consistent with the latest EPRI guidance.
EPRI PWR Primary Water Chemistry Guidelines include the hydrogen values of 15 and 5 cc/kg in action levels 2 and 3.
This revision does not change the existing normal operating range for dissolved hydrogen, but will allow operation of the RCS with hydrogen concentrations >5 cc/kg during off-normal conditions without reactor shutdown. No significant chemistry changes are expected >5 cc/kg. When operation is ~ 15 cc/kg, station chemistry will increase RCS monitoring to ensure chemistry remains unaffected. Because RCS chemistry will not change significantly, the probability and/or consequences of accidents will not be increased nor will the likelihood of new unreviewed accidents be increased. Therefore, an unreviewed safety question does not exist.
1-0P-CH-013 Rev. 2 1-MOP-Rl-001 O-MOP-Rl-001 e
e Surry Monthly Operating Report No. 97-03 Page 17 of 22 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: March, 1997 Operating Procedure 3/14/97 (Safety Evaluation No.97-045)
Operating Procedure 1-0P-CH-013, "Removal From and Return To Service of eves Deborating Demineralizers," contains instructions to load demineralizer 1-CH-1-38 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 RCS boron concentration is not affected by this evolution. This evolution will be performed with the unit at Cold Shutdown or Refueling Shutdown conditions. Therefore, an unreviewed safety question does not exist.
Maintenance Operating Procedure (Safety Evaluation No.96-147 Rev. 1) 3/20/97 Maintenance Operating Procedure 1-MOP-Rl-001, "Removal and Return to Service of Annunciator Panels 1 A through 1 E or 1 F through 1 K," provides instructions for increased monitoring and compensatory measures required while the affected annunciator panels are removed from service for maintenance.
Work will be performed in the non-safety related logic panel or locally in the annunciator panels and will not affect any protection circuit or its isolation. Compensatory actions will be implemented in accordance with 1-MOP-Rl-001 to monitor the affected systems and components. There is no unreviewed safety question raised by this evolution, and the radiological consequences of previously analyzed accidents are not increased. Therefore, an unreviewed safety question does not exist.
Maintenance Operating Procedure (Safety Evaluation No.97-049) 3/20/97 Maintenance Operating Procedure O-MOP-Rl-001, "Removal and Return to Service of The VS Annunciator Panels," provides instructions for increased monitoring and compensatory measures required while the affected annunciator panels are removed from service for maintenance.
Work will be performed in the non-safety related logic panel or locally in the annunciator panels and will not affect any protection circuit or its isolation. Compensatory actions will be implemented in accordance with O-MOP-Rl-001 to monitor the affected systems and components. There is no unreviewed safety question raised by this evolution, and the radiological consequences of previously analyzed accidents are not increased. Therefore, an unreviewed safety question does not exist.
AC 81-97-03 1-MOP-DG-001 e
e Surry Monthly Operating Report No. 97-03 Page 18 of 22 PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: March, 1997 Administrative Control (Safety Evaluation No.97-047) 3-19-97 Administrative control of the Unit 1 Component Cooling Heat Exchanger (CCHX) Service Water (SW) outlet manual isolation valves and the Unit 1 Turbine Building SW supply manual isolation valve was established while SW system motor operated valves (MOV) 1-SW-MOV-102A and 1-SW-MOV-102B were de-energized in the open position.
The administrative controls were established to maintain the capability to isolate the non-essential SW flowpaths affected by the de-energized SW MOVs.
The subject manual isolation valves were controlled by a licensed operator who, if required, was capable of closing the valves within the time limit assumed as part of the design bases to ensure that intake canal inventory was maintained. Therefore, an unreviewed safety question does not exist.
Maintenance Operating Procedure (Safety Evaluation No.97-051) 3/28/97 Maintenance Operating Procedure 1-MOP-DG-001, "Removal and Return to Service of the PDTI for Maintenance," provides instructions and limiting conditions for a Temporary Modification (TM) which would fail open 1-DG-PCV-100. This valve must be opened to provide a flowpath for the RCS loop drains and other drains to the Primary Drains Transfer Tank(PDTI).
The Primary Drains System is not addressed in any Technical Specification (TS) or Basis.
The TM does not increase the probability or consequence of a previously analyzed accident, nor does it increase the probability of an accident or malfunction of a different type. The margins of safety as defined in TS or the basis are not reduced. Therefore, an unreviewed safety question does not exist.
e llurry Monthly Operating Report No. 97-03 Page 19 of 22 TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: March, 1997 None During the Reporting Period
Primary Coolant Analysis Gross Radioactivity, µCi/ml Suspended Solids, ppm Gross Tritium, µCi/ml '
1131, µCi/ml 113111133 Hvdroqen, cc/kq Lithium, ppm Boron - 10, oom*
Oxygen, (DO}, ppm Chloride, ppm pH at 25 deqree Celsius Boron - 10 = Total Boron x 0.196 Comments:
None CHEMISTRY REPORT MONTH/YEAR: March, 1997 Unit No. 1 Max.
Min.
Avq.
2.03E+O 1.10E-4 1.88E-1 0.250
- C::0.010 0.054 1.14E-2 1.14E-2 1.14E2 7.50E-1 4.15E-5 8.49E-2 0.38 0.32 0.34 33.3 1.7 15.8 0.86 0.48 0.77 489.8 0.33 396.6 5.5
- C::0.005 2.6 0.018
<0.001 0.006 8.41 4.09 4.79
~urry Monthly Operating Report No. 97-03 Page 20 of 22 Unit No. 2 Max.
Min.
Avg.
2.63E-1 1.16E-1 2.01 E-1
- C::0.010
- C::0.010
- C::0.010 4.00E-1 1.26E-1 3.11E-1 8.65E-5 1.14E-5 5.60E-5 0.19 0.06 0.10 29.6 28.5 28.9 2.35 2.07 2.21 163.7 148.6 156.2
- C::0.005
- C::0.005
- C::0.005 0.005
- C::0.001 0.003 6.56 6.18 6.38
New Fuel Shipment or Cask No.
FUEL HANDLING UNITS 1 & 2 MONTH/YEAR: March, 1997 Number of Date Stored or Assemblies Received per Shipment Assembly Number ANSI Number 4'8urry Monthly Operating Report No. 97-03 Page 21 of 22 Initial Enrichment New or Spent Fuel Shipping Cask Activity No Fuel Received or Stored During the Reporting Period J
.~
~urry Monthly Operating Report No. 97-03 Page 22 of 22 DESCRIPTION OF.PERIODIC TEST(S) WHICH WERE NOT COMPLETED WITHIN THE TIME LIMITS SPECIFIED IN TECHNICAL SPECIFICATIONS MONTH/YEAR: MARCH, 1997 None During the Reporting Period