ML18152A311
| ML18152A311 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/31/1990 |
| From: | Stewart W, Warren L VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 90-356, NUDOCS 9006200423 | |
| Download: ML18152A311 (25) | |
Text
--
---~-----~--------
e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555 Gentlemen:
June 15, 1990 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 MONTHL V OPERATING REPORT Serial No.
NO/RPC:vlh Docket Nos.
License Nos.90-356 50-280 50-281 DPR-32 DPR-37 Enclosed is the Monthly Operating Report for Surry Power Station Units 1 and 2 for the month of May 1990.
Very truly yours,
\\j~
W. L. Stewart Senior Vice President - Nuclear Enclosure cc:
U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.
Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station
- 1
~ I
~
VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION MONTHLY OPERATING REPORT REPORT fJ 90-05 APPROVED:
POW 34-04
\\
==---- ---
- l
' J.,
e TABLE OF CONTENTS SECTION Operating Data Report - Unit No. 1 Operating Data Report - Unit No. 2 Unit Shutdowns and Power Reductions - Unit No. 1 Unit Shutdowns and Power Reductions - Unit No. 2 Average Daily Unit Power Level
- Unit No. 1 Average Daily Unit Power Level
- Unit No. 2 Summary of Operating Experience - Unit No. 1 Summary of Operating Experience - Unit No. 2 Facility Changes That Did Not Require NRC Approval Procedure or Method of Operation Changes that Did Not Require NRC Approval Tests and Experiments That Did Not Require NRC Approval Chemistry Report Fuel Handling - Unit No. 1 Fuel Handling - Unit No. 2 Description of Periodic Test(s) Which Were Not Completed Within the Time Limits Specified in Technical Specifications PAGE 1
2 3
4 6
7 8
9 12 17 19 20 21 21 22
--~-1
'J e
OPERATING DATA REPORT DOCKET NO.:
50-280 DATE:
06/06/90 COMPLETED BY:
L.A. Warren TELEPHONE:
(804)357-3184 x355 OPERATING STATUS NOTES
- 1. Unit Name:
Surry Unit 1
- 2. Reporting Period: May 01-31, 1990
- 3. Licensed Thermal Power (MWt):2441
- 4. Nameplate Rating (Gross MWe):847.5
- 5. Design Electrical Rating (Net MWe):
788
- 6. Maximum Dependable Capacity (Gross MWe):
820
- 7. Maximum Dependable Capacity (Net MWe):
781
- 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. Reason For Restrictions, If Any: ----------------------
THIS MONTH YTD CUMULATIVE
- 11. Hours In Reporting Period 744.0 3623.0 152879.0
- 12. Number of Hours Reactor Was Critical 516.0 3395.0 96145. 8
- 13. Reactor Reserve Shutdown Hours 0.0 0.0 3774.5
- 14. Hours Generator On-Line 516.0 3395.0 94218.2
- 15. Unit Reserve Shutdown Hours 0.0 0.0 3736.2
- 16. Gross Thermal Energy Generated (MWH) 1228562.0 8179150.0 219295953.5
- 17. Gross Electrical Energy Generated (MWH) 413250. 0 2765625.0 71311028. 0
- 18. Net Electrical Energy Generated (MWH) 393464.0 2633948.0 67644878.0
- 19. Unit Service Factor 69.4%
93.7%
- 61. 6%
- 20. Unit Availability Factor 69.4%
93.7%
64.1%
- 21. Unit Capacity Factor (Using MDC Net) 67.7%
93.1%
57.1
- 22. Unit Capacity Factor (Using DER Net) 67.1 92.3%
56.2
- 23. Unit Forced Outage Rate 30.6 6.3%
21.2%
- 24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):
Refueling, October OS, 1990, 58 days
- 25. If Shut Down at End of Report Period Estimated Date of Startup: ---===:c=----
- 26. Unit In Test Status (Prior to Commercial Operation):
FORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 1
- I
'J.
e OPERATING DATA REPORT DOCKET NO.:
50-281 DATE:
06/06/90 COMPLETED BY:
L.A. Warren TELEPHONE: --,.(~80~4~)~3~5~7--3-1~8~4,--x~3~55---~
OPERATING STATUS NOTES
- 1. Unit Name:
Surry Unit 2
- 2. Reporting Period: May 01-31, 1990
- 3. Licensed Thermal Power (MWt):2441
- 4. Nameplate Rating (Gross MWe):847.5
- 5. Design Electrical Rating (Net MWe):
788
- 6. Maximum Dependable Capacity (Gross MWe):
820
- 7. Maximum Dependable Capacity (Net MWe):
781
- 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. Reason For Restrictions, If Any: ----------------------
THIS MONTH YTD CUMULATIVE
- 11. Hours In Reporting Period 744.0 3623.0 149759.0
- 12. Number of Hours Reactor Was Critical 643.9 3522.9 94721.5
- 13. Reactor Reserve Shutdown Hours 0.0 0.0 328.1
- 14. Hours Generator On-Line 638.3 3517.3 93166.2
- 15. Unit Reserve Shutdown Hours o.o 0.0 0.0
- 16. Gross Thermal Energy Generated (MWH) 1492017.3 8369094.8 217979429.6
- 17. Gross Electrical Energy Generated (MWH) 490685.0 2805530.0 70886129.0
- 18. Net Electrical Energy Generated (MWH) 466036.0 2669839.0 67210798.0
- 19. Unit Service Factor 85.8%
97.1%
62.2%
- 20. Unit Availability Factor 85.8%
97.1%
62.2%
- 21. Unit Capacity Factor (Using MDC Net) 80.2%
94.4%
57.6%
- 22. Unit Capacity Factor (Using DER Net) 79.5%
93.5%
57%
- 23. Unit Forced Outage Rate 14.2%
2.9%
15.3%
- 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 Startup: ---====--
- 26. Unit In Test Status (Prior to Commercial Operation):
FORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 2
F:
S:
UNIT SHUTDOWN AND POWER REDUCTION (Equal To or Greater Than 20%)
REPORT MONTH:
MAY 1990 METHOD OF LICENSEE SHUTTING EVENT DATE TifPE(l)
DURATION (HOURS)
REASON(2) DOWN REACTOR(3) REPORT#
SYSTEM CODE(4)
COMPONENT CODE(5) 05/10/90 S
05/22/90 F
(1)
Forced Scheduled 0
B 4
N/A EL FAN 228.0 A
3 1-90-005 EL XFMR.
(2)
REASON:
A - Equipment Failure (Explain)
B - Maintenance or Test C - Refueling D - Regulatory Restriction E - Operator Training & License Examination F - Administrative G - Operational Error (Explain)
H - Other (Explain)
(3)
METHOD:
1 - Manual 2 - Manual Scram.
3 - Automatic Scram.
4 - Other (Explain) 3 DOCKET NO.:
50-280 UNIT NAME:
Surry Unit I DATE:
06/06/90 CO MP LE TED BY:
L.A. Warren TELEPHONE: --=-so,e-4=-_-=3=5=7_--:3:-::1-=s-=-4-x--:3:-::5=5-CAUSE & CORRECTIVE ACTION TO PREVENT RECURRENCE Ramped unit to 69% power in order.
repair 'A' & 'C' isolated phase duct cooling fan.
A fault on the Unit 1
'A' main transformer resulted in a Unit 1 generator trip and lockout of the
'A' reserve station service transformer (RSST).
The Unit 1 generator trip immediately initiated a
turbine/reactor trip and the lockout of the 'A' RSST resulted in the deenergization of the Unit 1 'A' station service (SS) which feeds the
'A' RCP.
- 3 emergency diesel generator started and loaded as designed on the Unit lJ bus.
Unit will remain at HSD until restart.~
(4)
Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report (LER)
File (NUREG 0161)
(5)
Exhibit 1 -
Same Source
_ _J
DATE 05/02/90 05/03/90 05/15/90 (1)
F:
Forced S:
Scheduled TYPE(l) s s
s (2)
REASON:
UNIT SHUTDOWN AND POWER REDUCTION (Equal To or Greater Than 20%)
REPORT MONTH:
MAY 1990 METHOD OF LICENSEE DURATION (HOURS)
SHUTTING EVENT REASON(2) DOWN REACTOR(3) REPORT#
0 B
0 B
0 B
4 4
4 N/A N/A N/A (3)
METHOD:
SYSTEM CODE(4)
SG SG TA A - Equipment Failure (Explain)
B - Maintenance or Test 1 - Manual C - Refueling D - Regulatory Restriction E - Operator Training & License Examination F - Administrative G - Operational Error (Explain)
H - Other (Explain) 2 - Manual Scram.
3 - Automatic Scram.
4 - Other (Explain) 4
- \\
.,;. \\
DOCKET NO.:
50-281
,;* \\
UNIT NAME: --=s,...u_r_ry_U=n-1..,..* t-2,,----'--
DATE:
06/90/90 COMPLETED BY: ---L-.-A-.-W-ar_r_e_n ___ _
TELEPHONE:
804-357-3184 x355 COMPONENT CODE(5)
CAUSE & CORRECTIVE ACTION TO PREVENT RECURRENCE COND COND V
Ramped down to 80.5% -
maintain condenser vacuum while cleaning waterboxes.
Ramped down to 62.7% power over the course of 2.25 days to maintain condenser vacuum while hydro lazing and repairing tube leaks in 'D' waterbox.
Ramped to 74% power to perform 2-PT-29.1 (turbine valves freedom of movement test).
(4)
Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report (LER)
File (NUREG 0161)
(5)
Exhibit 1 - Same Source e
\\
\\
I I
I I
F:
S:
UNIT SHUTDOWN AND POWER REDUCTION (Equal To or Greater Than 20%)
REPORT MONTH:
MAY 1990 METHOD OF LICENSEE SHUTTING EVENT DATE TYPE(l)
DURATION (HOURS)
REASON(2) DOWN REACTOR(3) REPORT#
SYSTEM CODE(4)
COMPONENT CODE(5) 05/22/90 F
05/31/90 F
(1)
Forced Scheduled 101. 8 3.9 (2)
REASON:
A 2
A 2
A - Equipment Failure (Explain)
B - Maintenance or Test C - Refueling D - Regulatory Restriction 1-90-005 EL 2-90-003 SJ (3)
METHOD:
1 - Manual XFMR FCV 2 - Manual Scram.
3 - Automatic Scram.
4 - Other (JB:xplain)
E - Operator Training & License Examination F - Administrative G - Operational Error (Explain)
H - Other (Explain) 5 DOCKET NO.:
50-281
~-
UN IT NAME:
Surry Unit 2 DATE:
06/90/90 COMPLETED BY:
L.A. Warren TELEPHONE:
804-357-3184 x355';
CAUSE & CORRECTIVE ACTION TO PREVENT RECURRENCE An electrical fault on Unit 1.,
main transformer caused an a
1
\\
reactor trip on Unit 1 and a loss of :
'A' RSST.
This fault resulted in \\
erratic indications of some Unit 2
parameters.
These indications prompted the Unit 2
operator to manually trip the Unit 2 reactor.
HSD was maintained until restart.
At 2005 hrs, with Unit 2 at 100%
power, 2-FW-FCV-2478A ('A' main feed regulator valve) failed closed in automatic.
Operators attempted to gain control of the valve in manual but the valve would not open.
Operators then manually tripped the reactor and turbine.
Unit w~
remain at HSD until restart.
(4)
Exhibit G - Instructions for Preparation of Data Entry Sheets for Licensee Event Report (I.ER)
File (NUREG 0161)
(5)
Exhibit 1 - Same Source
\\
I I
I
~--
e I
')
AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.:
50-280 UNIT NAME:
Surry Unit 1 DATE:
06/06/90 COMPLETED BY:
L.A. Warren TELEPHONE:(804)357-3184 x355 MONTH:
MAY 1990 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 1 780 17 782 2
783 18 784 3
785 19 784 4
775 20 784 5
782 21 783 6
784 22 781 7
779 23 0
8 777 24 0
9 777 25 0
10 620 26 0
11 517 27 0
12 782 28 0
13 786 29 0
14 787 30 0
15 787 31 0
16 786 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.
6
~=---*-------~--- -
MONTH:
MAY 1990 AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.:
UNIT NAME:
DATE:
50-281 Surry Unit 2 06/06/90 COMPLETED BY:
L.A. Warren TELEPHONE:(804)357-3184 x355 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 1 727 17 775 2
681 18 777 3
699 19 778 4
556 20 777 5
477 21 777 6
571 22 776 7
714 23 0
8 773 24 0
9 744 25 0
10 776 26 379 11 775 27 761 12 777 28 771 13 778 29 775 14 773 30 775 15 723 31 775 16 777 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.
7
--~---
e e
SUMMARY
OF OPERATING EXPERIENCE MONTH/YEAR:
MAY 1990 Listed below in chronological sequence experiences for this month which required significant non-load related incidents.
by unit is a summary of operating load reductions or resulted in UNIT ONE 05/01/90 05/04/90 05/10/90 05/11/90 05/12/90 05/22/90 05/31/90 0000 1004 This reporting period started with the Unit operating at 100% power; 820 MW.
Started ramp down in order to perform l-PT-29.1.
1040 Stopped ramp at 91% power, 750 MW.
1307 Started ramp up after completion of l-PT-29.1.
1348 Stopped ramp; 100% power, 815 MW.
0848 Started ramp down to work on isolated phase bus duct cooling fan, 820 MW.
1115 Stopped ramp; 69% power, 530 MW.
2144 0126 1157 2400 Started ramp up after completion of work on isolated phase bus duct cooling fan; 67% power, 540 MW.
Stopped ramp; 100% power, 825 MW.
A fault on the Unit 1
'A' main transformer resulted in automatic generator trip and lockout of the 'A' reserve station service transformer (RSST).
The generator trip immediately initiated a turbine/reactor trip.
The lockout of 'A' RSST resulted in the deenergization of the Unit 1 'A' station service transformer which feeds the 'A' RCP.
The #3 emergency diesel generator started and loaded as designed on the Unit lJ bus.
This reporting period ended with the Unit stable at HSD.
8
---==---- -- -- - - ~
e
SUMMARY
OF OPERATING EXPERIENCE MONTH/YEAR:
MAY 1990 Listed below in chronological sequence experiences for this month which required significant non-load related incidents.
by unit is a summary of operating load reductions or resulted in UNIT TWO 05/01/90 05/02/90 05/03/90 05/04/90 05/05/90 0000 This reporting period started with the Unit operating at 100% power, 795 MW.
1156 Started ramp down in order to maintain condenser vacuum during waterbox cleaning operations.
1538 1923 2028 0520 1227 1305 2110 1309 1710 1437 1508 0440 0513 1354 1415 1847 Stop ramp; 86% power, 670 MW.
Started ramp up after completion of waterbox cleaning operations; 86% power, 670 MW.
Stopped ramp, 100% power, 620 MW.
Started ramp down in order to maintain condenser vacuum during waterbox cleaning operations; 100% power, 780 MW.
Stopped ramp; 80.5% power, 620 MW.
Started ramp up after completion of waterbox cleaning operations; 79.5% power, 610 MW.
Stopped ramp; 100% power, 800 MW.
Started ramp down in order to maintain condenser vacuum; 'D' waterbox removed from service for hydrolazing and tube leak repair; 100% power; 795 MW.
Stopped ramp; 78% power, 630 MW.
Started ramp down in order to maintain condenser vacuum; 78%
power, 610 MW.
Stopped ramp; 73.5% power, 580 MW.
Started ramp down in order to maintain condenser vacuum; 73.5% power, 555 MW.
Stopped ramp; 70% power, 545 MW.
Started ramp down in order to maintain condenser vacuum; 70%
power, 540 MW.
Stopped ramped; 66% power, 500 MW.
Started ramp down in order to maintain condenser vacuum; 65.5% power, 490 MW.
9
e
SUMMARY
OF OPERATING EXPERIENCE MONTH/YEAR:
MAY 1990 Listed below in chronological sequence experiences for this month which required significant non-load related incidents.
by unit is a summary of operating load reductions or resulted in UNIT TWO 05/05/90 05/06/90 05/07/90 05/09/90 05/14/90 05/15/90 1851 1125 Stopped ramp; 62.7% power, 470 MW.
Started ramp up after completion of 'D' waterbox hydrolazing and tube leak repairs; 64% power, 530 MW.
1400 Stopped ramp; 100% power, 810 MW.
1904 Started ramp down in order to maintain condenser vacuum; 2-CW-S-lD, high level traveling screen, will not rotate forwards or backwards due to high differential pressure across screen.
This is due to excessive build-up of seaweed on screens; 100% power, 815 MW.
2138 1105 1300 0115 0240 0635 0750 2201 2258 0010 0135 1027 1213 1654 1840 Stopped ramp; 84.5% power, 670 MW.
Started ramp up after completion of seaweed clean-up and subsequent screen repairs; 85% power, 680 MW.
Stopped ramp; 100% power, 810 MW.
Started ramp down in order to maintain condenser vacuum during waterbox cleaning operations; 100% power, 780 MW.
Stopped ramp; 86.5% power, 680 MW.
Started ramp up after completion of waterbox cleaning operations; 86.5% power, 710 MW.
Stopped ramp; 100% power, 810 MW.
Started ramp down in order to maintain condenser vacuum during waterbox cleaning operations; 100% power, 815 MW.
Stopped ramp; 96% power, 750 MW.
Started ramp up after completion of waterbox cleaning operations; 96% power, 750 MW.
Stopped ramp; 100% power, 825 MW.
Started ramp down in order to perform 2-PT-29.1.
Stopped ramp; 74% power, 620 MW.
Started ramp up; 2-PT-29.1 completed; 74% power, 610 MW.
Stopped ramp; 100% power, 820 MW.
10
-~------
e e
SUMMARY
OF OPERATING EXPERIENCE MONTH/YEAR:
MAY 1990 Listed below in chronological sequence experiences for this month which required significant non-load related incidents.
by unit is a summary of operating load reductions or resulted in UNIT TWO 05/22/90 05/26/90 05/27/90 05/31/90 05/31/90 1157 1030 Manual reactor trip; an electrical fault on Unit 1 'A' transformer caused an automatic reactor trip on Unit 1 main and a fault loss of 'A' reserve station service transformer.
This resulted in erratic indications of some of Unit 2 parameters.
These indications prompted the Unit 2 operator manually trip the Unit 2 reactor.
Commenced reactor startup.
1215 Reactor is critical.
1750 Unit is on-line and ramping up.
1800 Unit is at 30%
power with main feed regulator valves in automatic; ramp continues.
0124 2005 2400 Stopped ramp; 100% power, 815 MW.
Manual reactor trip; 2-FW-FCV-2478A, 'A' main feed regulator valve; failed closed in automatic.
Attempted to gain control in manual but the attempt was unsuccessful.
The operator then manually tripped the reactor.
This reporting period ended with the Unit stable at hot shutdown.
11 J
--~- --- --~*--
DCP 86-11
-~---- -----
e FACILITY-CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:
MAY 1990 SERVICE WATER AND CIRCULATING WATER BUTTERFLY VALVE REPLACEMENT UNIT 2 This design change replaced 96",
36" and 10" circulating and service water butterfly valves and expansion joints.
The replacement valves are ductile iron with the wetted portions coated with a liquid epoxy.
The following valves and expansion joints were replaced:
MOV-CW-200A,B,C,D MOV-CW-206A,B,C,D MOV-SW-201A,B MOV-SW-202A,B CW outlets CW inlets SW to BC HX SW to CC RX The replacement valves and expansion joints are one for one replacement of existing equipment and are designed to specification requirements which meet or exceed the original specification.
The equipment will operate and function identically to existing equipment.
The new valves and expansion joints do not affect or change the basis for any Technical Specification.
The replacement valves are seismically qualified.
SCAFFOLD REQUEST UNIT 2 05/01/90 (Safety Evaluation #90-0121)
The request is for the erection of a temporary scaffold to gain access to the Unit 2 'A' waterbox to support tube sheet cleaning.
The temporary scaffold is required for safe working conditions.
Installation of scaffold constructed per SUADM-ADM-07 has a high confidence level against failure and was reviewed for effects on accident analyses and equipment operability/function.
It is thus concluded that assumptions, bases and probabilities of accident analyses and equipment malfunctions are not affected.
SCAFFOLD REQUEST UNIT 2 05/01/90 (Safety Evaluation #90-0122)
The request is for the erection of a temporary scaffold to gain access to the Unit 2 'C' waterbox to support tube sheet cleaning.
The temporary scaffold is required for safe working conditions.
Installation of scaffold constructed per SUADM-ADM-07 has a high confidence level against failure and was reviewed for affects on accident analyses and equipment operability/function.
It is thus concluded that assumptions, bases and probabilities of accident analyses and equipment malfunctions are not affected.
12
It TM-S2-90-06 e
FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:
MAY 1990
~----
SCAFFOLD REQUEST UNIT 2 05/01/90 (Safety Evaluation #90-0123)
The request is for the erection of a temporary scaffold to gain access to the Unit 2 'B' waterbox to support tube sheet cleaning.
The temporary scaffold is required for safe working conditions.
Installation of scaffold constructed per SUADM-ADM-07 has a high confidence level against failure and was reviewed for effects on accident analyses and equipment operability/function.
It is thus concluded that assumptions, bases and probabilities of accident analyses and equipment malfunctions are not affected.
TEMPORARY MODIFICATION (Safety Evaluation #90-0124) 05/05/90 This modification is to restore the
'B' high level intake structure screenwash pump to service by installing a
temporary spool-piece in place of the pump's discharge expansion joint.
The installation of the temporary spool-piece on the discharge of 2-CW-P-2B will in no manner affect the operability of the intake canal level instrumentation or of the required service water flowpaths to the recirculation spray heat exchangers (RSHX),
component cooling heat exchangers (CCHX),
bearing cooling heat exchangers (BCHX),
or the charging pump service water (SW) system/main control room chiller SW system.
In addition, the copper-nickel spool-piece design is fully capable of withstanding the expected piping stresses and is resistant to general corrosion to the brackish water contained within this system.
Therefore, an unreviewed safety question is not created.
SCAFFOLD REQUEST UNIT 2 05/10/90 (Safety Evaluation #90-0126)
This request is for the erection of a temporary scaffold in the Unit 2 cable tunnel.
The scaffold will be erected and supported per SUADM-ADM-07.
This scaffold may be attached to an existing scaffold in the area.
The scaffold will be approximately 27' high from the floor at elevation 9'6".
The temporary scaffold is required for safe working conditions.
Installation of scaffold constructed per SUADM-ADM-07 has a high confidence level against failure and was reviewed for effects on accident analyses and equipment operability/function. It is thus concluded that assumptions, bases and probabilities of accident analyses and equipment malfunctions are not affected.
13
~
I I
EWR-90-222 TM-S2-90-07 TM-Sl-90-10 FACILITY CHANGES THAT DID NOT REQUIRE NRC-APPROVAL MONTH/YEAR:
MAY 1990 ENGINEERING WORK REQUEST UNITS 1&2 05/17/90 (Safety Evaluation #90-131)
This request repowered flow transmitter (FT-VS-116),
the radiation monitor for the ventilation stack, from an electrical power source connected to an emergency power electrical bus.
The flow transmitter was previously powered from a
station service electrical bus.
Flow transmitter FT-VS-116 is being repowered from a
diesel backed source in heat trace panel HTP-2B3.
This will provide a reliable power supply to the vent stack radiator monitor flow transmitter in the event of a
loss of offsite AC power.
Repowering the flow transmitter from heat trace panel HTP-2B3 will not have an impact on safety related equipment, safe shutdown systems, or require any technical specification changes.
This activity has no impact on accidents or malfunction of equipment previously evaluated in the safety analysis report.
Therefore, an unreviewed safety question does not exist.
TEMPORARY MODIFICATION (Safety Evaluation #90-0133) 05/23/90 This modification added jumpers to allow replacement of the P7XA relay and returns the system functions to normal.
This relay was replaced when the Unit was in hot shutdown.
This temporary modification does not affect the plant operation and was cleared before startup.
Therefore, an unreviewed safety question was not created.
TEMPORARY MODIFICATION (Safety Evaluation #90-0134) 05/28/90 This modification added jumpers to allow replacement of failed relays SRB-XA and P8-YA.
The replacement is required to system logic to technical temporary modification will
- Therefore, an unreviewed safety 14 return the reactor protection specification compliance.
The be cleared before startup.
question does not exist.
FS-88-71 FS 90-11 FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR: ~~MA~Y~l_9_9_0~~-
UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) CHANGE REQUEST (Safety Evaluation #90-0136) 05/29/90 The UFSAR states that an alarm is sounded in the Control Room which indicates a pipe break in the discharge piping of the recirculation spray (RS) pumps.
No alarm exists in the Control Room of this type.
This change clarifies that the operator will use variations in discharge pressure and other indications to detect a RS line break.
The proposed UFSAR change does not present an unreviewed safety question for the following reasons:
(1) the probability of an additional failure in addition to the primary failure i.e., a large break loss of coolant accident (LOCA) inside containment is very remote.
A break in the outside recirculation spray (RS) piping is not postulated because the lines are not high energy and are only used after an accident condition.
Thus the leakage assumed from the outside RS piping is not greater than that specified in the technical specifications.
(2) the probability of occurrence or the consequence of any accidents or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased as a result of the lack of a
RS low discharge pressure alarm.
Other plant indications, e.g. RS pump discharge pressure, stack radiation monitors, and sump level indication, are available to the operator.
UPDATED FINAL SAFETY ANALYSIS REPORT (FSAR) CHANGE REQUEST (Safety Evaluation #90-0135) 05/29/90 This change request is to achieve agreement between technical specifications and UFSAR on boric acid concentration for the boric acid storage tanks.
The bases for Surry Power Station (SPS) TS 3.2 supports a minimum boric acid concentration of 7.0%
by weight and an upper concentration of 8.5%
by weight.
Surry UFSAR section 9.1 describes the boric acid concentration in three different ways ie; approximately 2.5 weight%, 7.5%, and between 7.0% and 8.5% by weight.
They are at variance with each other and with the SPS TS.
This change brings the UFSAR into alignment with the SPS TS.
An unreviewed safety question does not exist, since no change in the plant chemistry is being made, merely an editorial change to achieve uniformity in the wording between the Surry UFSAR and TS.
15
.. ----=---- ------ --------
1
/
FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:
MAY 1990 AC-Sl-90-531 ADMINISTRATIVE CONTROL (Safety Evaluation #90-0141) 05/31/90 Administrative control will be maintained over one motor driven auxiliary feedwater pump when the control switch is in pull-to-lock.
The switch will be placed in auto in the event of a
safety injection signal, low-low steam generator level signal or loss of the running pump.
This administrative control will be in effect while maintenance is performed on a main feed pump and the other motor driven auxiliary feedwater pump is operating to supply necessary makeup to the steam generators.
In order to prevent auto start of an auxiliary feedwater pump the control switch will be placed in pull-to-lock, when main feedwater is removed from service.
The other auxiliary feedwater pump will be supplying water to the generators.
Therefore, this administrative control will not create an unreviewed safety question.
16
PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:
l/2-TOP-3038 TEMPORARY OPERATING PROCEDURE (Safety Evaluation #90-0125)
MAY 1990 05/08/90 This temporary procedure is for the temporary shutdown of the auxiliary ventilation system while cleaning the honeycomb flow element Ol-VS-FE-116.
Adequate precautions and guidance are provided in the controlling procedure to closely monitor charging pump temperature during the time the fans are secured.
If the charging pump temperature exceeds 100°F, the auxiliary building central system will be restarted.
The workers will leave the door closed and the access port will be clamped shut.
Both trains' auxiliary ventilation filtered exhaust systems will remain operable and in automatic during this activity.
Therefore, an unreviewed safety question is not created.
1/2-PT-15.lE PERIODIC TESTS 1/2-PT-15.lF (Safety Evaluation #90-0129) 1/2-PT-15.lG These tests verify the full feedwater (AFW) system using the alignments.
05/10/90 flow capability of the auxiliary design basis flow paths and This change will provide period verification that the auxiliary feedwater system flow path to the steam generators is available using the analyzed alignments.
This change was required to address SOER-86-1, Reliability of Power to Auxiliary Feedwater System.
Basically, this SOER recommends that the AFW pumps be periodically tested under conditions and configurations expected during any operating event demand.
Since the plant has been previously analyzed and designed to operate under these configurations, an unreviewed safety question does not exist.
17
PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:
SUADM-ADM-07 ADMINISTRATIVE PROCEDURE (Safety Evaluation #90-0127)
MAY 1990 05/10/90 This procedure was revised to simplify the approval and administration process currently utilized for scaffold requests; to eliminate/reduce the need for individual safety evaluations for scaffold; and to establish a more clearly defined statement of risk assessment for ladders and scaffold around safe shutdown equipment.
The purpose of this procedure is to increase awareness of the risks associated with ladders and scaffolds.
This evaluation considers the causes and effects for potential ladder and scaffold failure on station personnel, systems and components.
The conclusions of this evaluation are:
- 1) the probability of failure (risk) due to scaffolds and ladders is minimized; 2) the consequences of failure have no effect on the existing method of operation, plant systems or emergency actions; 3) the level of risk due to ladder and scaffold use is considered to be within acceptable bounds;
- 4) safety concerns relating to ladders and scaffold are addressed and no unreviewed safety question exists.
18
- v 2-ST-210 l/2-ST-229 TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MONTH/YEAR:
MAY 1990 SPECIAL TEST 12/27/87 This test recorded the data to be used to determine the safe minimum operating flow conditions for the auxiliary feedwater system.
Specifically, the test obtains pump vibration data as specified at various flows from full design flow (700 GPM plus recirculation) to present minimum recirculation flow (35 GPM) condition and obtains the pump and turbine driven bearing oil and cooling water temperatures at specified flows from full design flow to minimum recirculation flow condition.
In order to obtain the most accurate information possible, it was necessary that the auxiliary feedwater flow be directed to one steam generator for feedwater flow rates less than 250 GPM.
Therefore, this procedure was written with 2-FW-P-2 providing flow to steam generator 2-RC-E-lA only when operating at a reduced auxiliary feedwater flow rate.
In order to ensure the auxiliary feedwater system was available to perform its intended safety function, if required, the motor-driven pumps remained in their normal plant line-up and only the turbine-driven pump discharge valves were repositioned during portions of this test.
The technical specifications limiting conditions of operations (LCOs) were met prior to commencement of the test.
Therefore, an unreviewed safety question was not created.
SPECIAL TEST 03/02/89 This test verified, through actual performance, the design calculation assumptions for the time required to manually secure nonessential service water loads in the event of a
large break LOCA coupled with a
station blackout.
The time assumed for manual isolation was less than one hour.
This test did not constitute an unreviewed safety question nor did it require a change in technical specifications.
The non-essential service water loads are not required to mitigate the consequences of a design basis accident.
19
VIRGINIA POWER SURRY POWER STATION CHEMISTRY REPORT MONTH/YEAR:
MAY 1990 PRIMARY COOLANT UNIT NO. 1 UNIT NO. 2 ANALYSIS MAX.
MIN.
AVG.
MAX.
MIN.
AVG.
Gross Radioact., µCi/ml 8.39E-l
- 1. 92E-2 5.36E-l 3.13E-l 7.40E-3
- 1. 40E-l 9.03E-2 l.19E-l 4.45E-l
- 1. 83E-l 3.24E-l Iodine-131, µCi/ml 4.35E-l 4.75E-4 5.90E-2 l.86E-3 l.75E-4 6.54E-4 Iodine-131/Iodine-133 0.15 0.12 0.13 0.15 0.06 0.09 Hydrogen, cc/kg 29.7 25.3 27.5 32.1 25.3 27.5 Lithium, ppm 2.20 1.35
- 1. 67 2.34 2.05 2.21 Boron - 10, ppm*
113.1 26.5 57.4 216.2 116.2 139. 6 Oxygen, (DO), ppm 0.005 0.005 0.005 0.005 0.005 0.005 Chloride, ppm 0.004 0.001 0.002 0.008 0.004 0.006 pH ti 25 degree Celsius 7.18 6.60 6.99 6.82 6.42 6.66
20
UlNTf-* 1&2 NEW OR SPE!n FUEL SHIPMENT I DATE SHIPPED OR RECEIVED FUEL HANDLING NUMBER OF ASSEMBLIES ASSEMBLY ANSI PER SHIPMENT NUMBER NUMBER NONE DURING THIS REPORTING PERIOD 21 DATE:
MAY 1990 INITIAL ENRICHMENT NEW OR SPENT FUEL SHIPPDiG CASK ACTIVITY LEVEL
DESCRIPTION OF PERIODIC TEST(S) WHICH WERE NOT COMPLETED WITHIN THE TIME LIMITS SPECIFIED IN TECHNICAL SPECIFICATIONS MONTH/YEAR:
MAY 1990 NONE DURING THIS REPORTING PERIOD 22