ML20024F820
| ML20024F820 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 11/30/1990 |
| From: | Stewart W, Warren L VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 90-760, NUDOCS 9012210164 | |
| Download: ML20024F820 (21) | |
Text
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- 4 VIHOINIA Ei.ncrurc Ann Powen COMPANY H icii st ox n, VikOINI A 20201 i December 17,1990 U. S. Nuclear Regulatory Commission Serial No.90-760
- Attention: Document Control Desk NO/RPC:vlh 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 OPER ATING REPORT Enclosed is the Monthly. Operating Report for Surry Power Station Units 1 and 2 for the month of November 1990.
Very truly yours, 5
W; L. Stewart Senior Vice President - Nuclear
' Enclosure cc: Ui S. Nuclear Regulatory Commission Region ll 101 Marietta Street, N. W.
Suite 2900 Atlanta, Georgia 30323 :
?
Mr. W. E. Holland -
- NRC Senior. Resident inspector-Surry Power Station i
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"90122'0164 1 901130 f e - Ll>
d POW 34-04 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER 11TATION HOFTHLY OPERATINC 8tEPORT REPORT i 90-11
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APPROVED: g2 .
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_TABLn OF CONTENTS SECTION PAGE Operating Data Report - Unit No. I 1 Operating Data Report - Unit No. 2 2 Unit Shutdowns and Power Reductions - Unit No. 1 3 Unit Shutdowns and Power Reductions - Unit No. 2 4 Average Daily Unit Power Level - Unit No. 1 5 Average Daily Unit Power Level - Unit No. 2 6 Summary of Operating Experience - Unit No. 1 7-Summary of Operating Experience - Unit No. 2 7 Facility Changes That Did Not Require NRC Approval 8 Procedure or Method of Operation Changes that Did Not Require NRC Approval 14 i
- Tests and Experiments That Did Not Require NRC Approval 15 Chemistry Report. 16 Fuel Handling - Unit No. 1 17 i
Fuel Handling - Unit No. 2 17 Description of Periodic Test (s) Which Were Not Completed Within
- the Time Limita Specified in' Technical Specifications 18
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1 OPERATING DATA REPORT I
l DOCKET NO.: 50-280 DATE: 12/06/90 COMPLETED BY L.A. Warren TELEPil0NE: (804)357-3184 x355 OPERATING STATUS NOTES
.1. Unit Name Surry Unit 1
- 2. Reporting Periods _Nov._01-30, 1990
- 3. Licensed Thermal Power (MWt):2441
- 4. Nameplate Rating (Cross MWe):847.5
- 5. Design Electrical Rating (Net FNe): 788
- 6. Maximum Dependable Capacity (Gross MWe):_ 820
.7. Maximum Dependable Capacity (Net MWe): 781 8.-If Changes Occur in Capacity Ratings (Items Number a Through 7) Since Last Report, Give Reasons: _
- 9. Power Level To Which Restricted, If Any (Net MWe):
- 10. Reason For Restrictions, If Any: __,.
TilIS MONTH YTD CUMULATIVE
- 11. Ilours.In Reporting Period 720.0 8016.0 157272.0
- 12. Number of flours Reactor Was Critien1 0 6387.2 99138.0
- 13. Reactor-Reservo-Shutdown Hours 0 0 ~
3774.5
- 14. llours Generator On-Line 0 ~
6373.5 97196.7
- 15. Unit Reserve Shutdown .lioura 0 0 _
3736.2
- 16. Cross Thermal Energy Generated (MWil) 0 14488795.5 '225605598.5
'17. Gross Electrical _ Energy Generated (MWR) 0 4826105.0 73371508.0
- 18. Net Electrical Energy Generated (MWil) 0 4577151.0 _69588081.0 i 19J . Unit Service Factor 0 79.5% 61.8%.
- 20. Unit Availability. Factor 0 79.5% 64.2%
- 21. Unit Capacity Factor (Using MDC Net) 0 73.lf 57.1%
'22. Unit Capacity Factor (Using DER Net) 0 72.5% 56.2%
- 23. Unit Forced Outage Rate' O 4.6% 20.7%
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- 24. Shutdowtis- Scheduled Over Next 6 Months (Type', Date, and Duration of Each):
Refueling / Snubber Outage started 10/06/90 ,
25'. If. Shut'Down at End of Report Period Estimated Date of Startup: 12-18-90
- 26. Unit In Test Status (Prior to Commercial Operation): FORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICIIT ,
COMMERCIAL OPERATION 1
l OPERATING DATA REPORT 1 DOCKET No.: _50-281 DATEt 12/06/90 COMPLETED BY: L.A. Warren TELEP110NE: _ (804)357-3184 x355
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_0PERA_ TING STATUS NOTES
- 1. UMt Nat.e: Surry Unit 2
- 2. Reporting Periods _Nov. 01-30 1990
- 3. Licensed Thermal Power (MWt): 2441
- 4. Nameplate Rating (Grosa MWe)t_847.5
- 5. Design E1cetrical Rating (Net MWe): 788
- 6. Maximum Dependable Capacity (Gross MWe):_ 820
- 7. Maximum Dependable Capacfty (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 Anyt T}{IS MONTil YTD CUMULATIVE
- 11. flours In Reporting Period 720.0 8016.0 ~
154152.0
- 12. Number'of Hours Reactor Was Critical 320.8 7229.7 98428.3
- 13. Reactor Reserve Shutdown lloura ~
0 0 328.1
- 14. Ilours Generator On-Line 313.4 7204.8 96853.7
- 15. Unit Reserve Shutdown lloura 0 0 0
- 16. Gross Thermal Energy Generated (MWil) 607751.0 17073012.1 226683346.9
- 17. Cross Electrical Energy Generated (MWl!) 195615.0- 5662815.0 73743414.0
- 18. Net Electrical Energy Generated (MWil) 183884.0 _ 5378459.0 69919418.0
- 19. Unit Service Factor 43.5% 89.9% 62.8%
- 20. Unit Availability Factor 43.5 89.9% 62.8%
l- 21. Unit Capacity Factor (Using MDC Net) 32.7 85.9% 58.2%
- l. 22., Unit Capacity Factor (Using DER Net) 32.4% 85.1% 57.6%
- 23. Unit-Forced Outage Rate 56.5% 10.1% 15.3%.
'24.- Shutdowns Scheduled Over Next 6 Months (Type Date, and Duration of Each):
l p Refueling /4-5-91/60 dayn -
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- 25. If Shut Down at End of Report Period Ettimated Date of Startup:
- 26. Unit in Test Status. (Prior to Co.maercial Operation):
FORECAST ACHJEVED I
- l. INITIAL CRITICA'.ITY l INI11AL ELECTRICITY COMMERCIAL OPERATION _
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DOCKET No.: 50-280 -
l l UNIT SHUTDOWN AND POWER. REDUCTION UNIT NAME: SURRY LMIT ONE ~i (Equal To or Greater Than 20%) DATE: 12/06/90 COMPLETED BY: L.A. Warren !
- REPORT MONTH
- N0VD!BER " 1990 TELEPHONE: 804-357-3184 x355 METHOD OF LICENSEE .
i DURATION. SHUTTING EVENT SYSTEM COMP 0 MENT CAUSE & CDRRECTIVE ACTION TU !
f DATE TYPE (1) '(HOURS) = REASON (2) DOWN REACTOR (3) REPORTI CODE (4) CODE (5) PREVENT RECURRENCE
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- 11/01/90 S 720.0 C I N/A N/A N/A Unit is in a Refueling Outage.. j i t 6
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- . (1) (2) (3) (4) I 3
F: Forecd REASON:~ METHOD: }
, S: Scheduled A - Equipment Failure (Explain) 1 - Manual Exhibit G - Instructions for [
B - Maintenance or Test 2 - Manual Scram. Preparation of Data Entry Sheets !
C - Refueling 3 - Automatic Scram. for Licensee Event Report (LER) f D - Regulatory Restriction 4 - Other (Explain) File (NUREG 0161) i E_.- Operator Training & License Examination [
. F'- Administrative (5) '
l C - Operational Error (Explain)
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DOCKET No.: 50-281 -
UNIT SHUTDOWN AND POWER REDUCTION UNIT NAME: SURRY UNIT TWO (Equal To or Greater Than 20%) DATE: 12/06/90 COMPLETED BY: L.A. Warren REPORT MOltrII: NOVMBER 1990 TELEPHONE: 804-357-3184 x355 METHOD OF LICENSEF DURATION SHUTTING EVENT SYSTEM COMPONENT CAUSE & CORRECTIVE ACTION TO DATE TYPE (1) (HOURS) REASON (2) DOWN REACTOR (3) REPORTF CODE (4) CODE (5) PREVENT RECURRENCE 11/01/90 F 406.6 D 4 90-014-00 B1 HX Unit was shutdown as a result of concerns related to Icv !
service water flow through the .
recirculation spray heat '
exchangers (RSHX). These concerns were addressed as a result of a Unit One flow test '
of its RSHXs. !
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i (1) (2) (3) (4)
F: Forced REASON: METHOD:
S: Scheduled A - Equipment Failure (Explain) 1 - Manual Exhibit C - Instructions for !'
B - Maintenance or Test 2 - Manual Scram. Preparation of Data Entry Sheets C - Refueling 3 - Automatic Scram. for Licensee Event Report (LER)
D - Regulatory Restriction 4 - Other (Explain) File (NUREC 0161)
E - Operator Training & License Examination F - Administrative (5)
C - Operational Error (Explain)
II - Other (Explain) 4 Exhibit 1 - Same Source
AVERACE DATLY UNIT POWER LEVEL DOCKET NO.: 50-280 UNIT NAME: SURRY UNIT ONE DATE: 12/06/90 COMPLETED BY: L.A. Warren TELEPHONE:(804)357-3184 x355 I
MONTH ' NOVEMBER 1990 l
1 DAY AVERACE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 0 17 0 2 0 18 , p ___ _
3 0 19 , )_
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 31 0 16 0 l
INSTRUCTIONS On this format, Ifst the average daily unit power level in MWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.
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,f AVERAGE DAILY UNIT POWER LEVEL l
DOCKET No.: 50-281 l UNIT NAME: Surry Unit 2 )
DATE: 12/06/90 COMPLETED BY: L.A. Warren TELEPHONE:(804)357-3184 x355 MONTH: NOVEMBER 1990 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Not) (MWe-Net) 1 0 17 0 2 0 18 192 3 0 19 266 4 _
0 20 364 5 0- 21 555 6 0 22 696 7 0 23 698 .
8 0 24 699 9 0 25 696 10 0 26 700 _
' ll 0 27 700 3
12 0 28 699 13- 0 29 700 14 0 30 700 15 0 31-16 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.
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SUMMARY
OF OPERATING EXPERIENCE HONTII/ YEAR: NOVEMBER 1990 Listed below in 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 11/01/90 0000 This reporting period started with the Unit at_ Refueling Shutdown.
11/30/90 2400_ This reporting period ended with the Unit at Cold Shutdown.
ITNIT TWO 11/01/90 0000 This reporting period started with the Unit at Cold Shutdown due to -concerns regarding service water flow through the recirculation spray heat exchangers.
11/12/90 0451 Unit at hot shutdown.
11/17/90 1249 Commenced reactor startup with control rod M-12 withdrawn two steps (inoperable).
1510 Reactor Critical.
1736 Reactor at 2% power.
2239 Unit on line, 11/18/90 0005 lloiding for Chemistry; 30% power, 240 MWe.
0533 Started ramp up; 30% power, 240 MWe.
11/21/90 0110 Unit at 90% power, 740 MWe (limit-due to inoperable control rod),
t j 11/30/90 2400 This reporting period ended with the' Unit operating at 90%
p power. 740 MWe.
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v FACII.ITY CHANGES TilAT DTD NOT REQUIRE NRC APPROVAL MONTH / YEAR: NOVFMBER 1990 l
SAFETY ANALYSIS 11/02/90 (Safety Evaluation #90-0261)
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The request was for the erection of temporary scaffolds for ;
performing various testing, maintenance and construction l activities. 1 The temporarv scaffold is required for. safe working condition.
Installation of scaffold constructed per SUADM-ADM-07 has a high confidence level against failure and was reviewed for effects on 1 accident analyses and equipment operability / function. It is thus concluded that assumptions, bases and probabilities of accident analyses and equipment malfunctions are not affected.
TEMP 0RARY MODIFICATIONS (Safety Evaluation #90-0264) 11/07/90 TM-SI-90-039 Mark #01-CW-MOV-1000 & 01-SW-MOV-105C TM-SI-90-040 Mark #01-CW-MOV-100B & 01-SW-MOV-105C TM-SI-90-041 Mark #01-SW-MOV-105C & 01-CS-MOV-100D TM-SI-90-042 Mark #01-CW-MOV-1000 & Ol-SW-MOV-105C This change was requested to provide temporary power for certain valves during performance of ST-290, Recirculation Spray lleat Exchanger (RSHX) Service Water (SW) Flow Teat. The normal power supply to the valves was out of service for scheduled breaker and relay preventive maintenance.
This modification will cross-tie the motor control centera from the "H" and "J" emergency busses via temporary power feeds from spare breakers located in MCCs IJI-l and IJI-2. This modification will ta) performed utilizing NUS-2030 as a guideline and utilizing safe electrical practices in accordance 'with the accident prevention manual. . The cabling shall tai protected to the maximum extent possibic to' limit-the potential for _ fires.
Both units will be in cold shutdown or refueling shutdown during this modification. Therefore, the technical specification LCOs for emergency power supplies, service water supplies and nonessential service-water isolation are not applicable. This modification will only be installed on a temporary basis (approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) and shall be insediately removed upon ,
completion of_ scheduled testing. Electricians shall be standing '
by the affected motor control centers upon initial stroking of the valves to observe the breakers for unusual conditions. The valves shall not be declared operable until completion of stroke testing -in accordance with PT-25.1 or 25.2 as applicable.
Finally, this activity in no manner affects the controls, interlocho, or indications of the affected valves. Therefore, an unrevievrd safety question does not exist. ;
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FACILITY' CIIANGES TilAT DID NOT REQUIRE NRC APPROVAL HONTil/ YEAR: NOVEMilER 1990 SPS UNIT 1 - CYCLE 11 11/08/90 TSafety Evaluation #90-0266)
This change evaluates cold shutdown operations for Surry Unit 1 Cycle 11 refueling operations and cold shutdown conditions prior to completion of analyses for Cycle 11 power operation.
Technical Report NE-806 presents a discussion on the analyses and evaluations reviewed to support the conclusion that the Surry 1 Cycle 11 reload core can be safely placed in refueling and cold shutdown conditions. This safety evaluation supplements the evaluation in Technical Report NE-806. No unreviewed safety questions were identified.
JCO-90-1-004 JUSTIFICATION FOR CONTINUED OPERATION 11/09/90 (Safety Evaluation #90-269)
An evaluation was performed of compensatory measures taken to justify operation of the 'C' reactor coolant pump motor without an oil collection system as required by 10CFR50, Appendix R.
The new motor was ordered as a duplicate of our_ existing motors, flowever, the mounting arrangement for the oil co11cetion system was unique. Specifically, the bearing oil piping is rotationally oriented approximately 90 degrees different than the mounting arrangement.
The interim measures being taken are to install temporary barriers to contain an oil fire associated with the 'C' Reactor Coolant Pump motor-to the pump cubicle and to provide additional monitoring of the ptunp cubicle. These measures coupled with certain operating procedure changes will provide adequate fire-stops and early warning of problems in the pump cubicle containing the effects of a fire from other safe shutdown equipment in Containment. .With these measures in place we are confident that the intent of 10CFR50 Appendix R will be met ' f oi Cycle 11 on RCP 'C' and an unreviewed safety question is not created, r An exemption from these requiremants for Surry 1 Cycle 11 was h subsequently requested in a letter to the NRC -dated November 14, 1990. The NRC granted the exemption on December 6,1990, 9
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FACILITY CHANGES THAT D1D NOT REQUIPJ!, NRC APPROVAL _
M( NTil/ YEAR: NOVDiBER 1990 DR-SI-90-1421 DEVIATION rep 0RT 11/09/90 (Safety Evaluation # 90-0268)
This evaluation was performed to evaluate certain indications observed during the eddy current testing on Surry Unit I steam genetators in support of a Unit 2 startup following a forced outage.
Steam generator tube testing in Unit 1, during shutdown and refueling in the fall of 1990, has revealed eddy current indications. To date, it is undetermined if these indications represent true degradation in the tubes. The purpose of this safety evaluation is to address the issue of operability of steam generators in Unit 2 based on the Unit 1 eddy current results to date, kestinghouse, the Surry NSSS vendor and contractor for the eddy current testing, was consulted. The Westinghouse evaluation supports resumption of safe operation of Unit 2. Therefore, an unreviewed safety question was not created.
TH-SI-90-46 TEMPORARY MODIFICATION 11/15/90 (Safety Evaluation #90-0273)
This temporary modification covers the installation of a
, temporary modification to permit routine draining of the Containment sumps during removal of TV-DA-100B. TV-DA-100B must be removed for valve maintenance and overhaul following in11ure of the valve to pass 10 CFR$0 Appendix J 1eakage testing.
Failure of the spoolpiece is highly unlikely as the -material used for the spoolpiece will meet the applicable design criteria Ilowever,
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of the piping system in which it will be installed.
should leakage / failure of the spoolpiece occur, the sump . pump (s) can'be stopped and the redundant trip valve, TV-DA-100A, closed to stop the. leakage. Any leakage that could occur will be contained within the Auxiliary Building and'be drained- to- the c Aux 1.11ary Building sump. Further, the unit will be in CSD or RSD during installation of this modification and the j umpers l must be removed prior to the reactor coolant system exceeding 200;degrecc. The installation of this modification must be tracked utilir.ing OD-lG to satisfy refueling- containment integrity requirement of Technical Specification 3.10 and/or containment closure requirements pursuant to Generic Letter 88-17. If refueling-integrity must be set per Op-4.1, the redundart trip valve TV-DA-100A must be operable or tagged closed. Therefore, an unreviewed safety question is not
-created.
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FACILITY CilANGES TilAT DID NOT REQUIRE NRC APPROVAL MONTil/ YEAR: NOVEMBER 1990 SAFETY EVALUATION FOR SURRY 2 CYCLE 10 RESTART WITH INOPERABLE CONTROL ROD M-12 11/16/90 This evaluation reviewed operation of Surry 2 Cycle 10 through 15,700 mwd /T (EOC) with control rod M-12 inserted in the core incorporating a 90% maximum power administrative limit and revised control rod insertion limits.
The evaluation covered UFSAR transients for rod withdrawal, dropped and/or misa11gned rod, rod ejection, loss of :oolant accident (LOCA), main steam line break (MSLB) and lockeu rotor.
Calculations have been performed for these accidents, where appropriate, and we have determined that the current licensing basis remains bounding. No other tranients were identified as potentially affected by the inoperabic rod. Therefore, we conclude that operating with control assembly M-12 inserted does not result in an unreviewed safety question as defined in 10CFRSO. 59 and does not require a change to 'he Technical Specification.
TM-S2-90-018 TEMPORARY MODIFICATION 11/16/90 (Safety Evaluation #90-0276)
Control rod M-12 is inoperable and remains stuck in nearly a fully inserted position. To allow operation above 70% power (as allowed by the safety analyses for operation with control rod M-12 inoperable), the rod bottom bistable and runback signals will be defeated. With control rod M-12 inoperabic, the lift coil disconnect- will' be Icft open to prevent control rod M-12 withdrawal or insertion.
l Operation with control rod M-12 .in a nearly fully inserted
- position and inoperable has been- justified in a separate analysis- and is allowed by Technical Specification 3.12.C provided additional remedial actions are performed. Since control rod M-12 will remain inserted, the IRPI rod drop signal L and runback for this rod is unnecessary and will be defeated.
Assurance that the-control room operators (CRO) are aware of the control rod position and the defeat of the rod bottom runbacks will be provided through completion of the CR0 turnover checkoff list which shall be annotated to reflect the implementation of the above activities. Affected procedures are being changed to ref1cet the control rod status and actions implemented that will l require an immediate reactor trip in the event of an l uncontrollable. reactivity excursion and/or an additional rod drop incident. The IRPI rod drop circuits for the remaining l control rods will remain functional. Defeat of control rod M-12 lift coil will prevent rod motion which is consistent with~the analyses used to support operation of the unit with an inoperable control rod. The rod remains capable of performing its safety function of falling into the core. Therefore, an unreviewed safety question is not created.
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FACILITY CHANCES TilAT DID NOT REQUIRE NRC APPROVAL MON'111/ YEAR: NOVEMBER 1990 10CFR50 APPENDIX R REPORT 11/20/90 (Safety Evaluation #90-281)
This evaluation is being performed to assess the 1990 Update of the SPS Appendix R Report. It incorporates design changes and engineering work requests completed in 1988 and 1989.
The 1990 Update of the SPS Appendix R Report incorporates.
changes and plant modifications (DCPs and EWRs) made since revision 5 of the report dated April 1987 up through mid 1990.
A number of engineering evaluations' in Chapter 10 have been revised to reflect more specific detail of actual plant conditions. Additionally, five new evaluations have been added.
Finally, a number of editorini and administrative changes were:
made to enhance the accuracy and usability of the Appendix R Report. A separate safety evaluation was performed for each DCP/CWR. This change merely updates text to reflect plant conditions. An unreviewed safety question is not created.
1-TOP-3061 TEMPORARY
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OPERATING PROCEDURE 11/21/90 (Safety Evaluation #90-282)
An evaluation of a temporary procedure was performed for filling of isolated reactor coolant system loops via the RCS with the Unit in CSD.
The isolated loop boron concentration will be verified as equal to or greater than RCS (RHR) boron concentration and at least 5%
shut down margin (SDM) is required by the procedure. The ,
isolated loop will be filled via the RCS with. RCS makeup.
j provided from the refueling water storage tank (RWST). During the filling sequences the affected loop will be observed for leakage. Any leakage that results will drain to the containment sump and/or the primary drain transfer tank (PDTT). Standpipe level. of at Icast 18 feet is required and provisions to provide-at least one makeup flowpath to the core-(low /high head safety injection with. an available. pump) ensures an adequate water supply to prevent loss of RCS inventory and potential loss of RHR. Inadvertent . dilution potential is minimized by sampling the RCS and_the isolated loop to ensure that_the isolated' loop boron concentration is equal to or greater than the RCS Boron.
The loop will be filled from unisolated portions of the RCS with kCS makeup .provided from- the RWST. Finally, a RCS dilution would be very slow and would not create the potential to'cause-a loss of SDM that has not'been previously evaluated. Therefore, i an unreviewed safety question does not exist.
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FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTH / YEAR: NOVEMBER 1990 1-EWR-89-377 ENGINEERING WORK REOUEST 11/28/90 This request is to replace the existing Auxiliary Feedwater System 3" motor operated globe valves and 1/3 hp motor operators with new " notor operated globe valves with 2/3 hp motor operators. The valver, will have flow control trim and larger limitorque operators. The MOVs will be replaced to eliminate seat Icakage and provide motor operators which are sir.ed for postulated system conditions. The replacement valves and motor operators are seismically and environmentally qualified.
i The Surry Technical Specifications and Updated Final Safety Analysis Report (UFSAR) were reviewed to determine if this modification required a change to the Technical Specifications or involved an unreviewed safety question as defined in 10 CFR50.59. The Technical Specifications and Accident Analysis are not affected by this valve replacement.
SPS 1-CYCLE-Il SHUTDOWN HARGIN DATA FOR ECP CONDITIONS AND ROD OPERABILITY TESTS 11/29/90 This report evaluated operation of the Surry Unit 1, Cycle 11 core.
Technical report NE-809 presents a discussion of the accident and other analyses and evaluations which support the conclusion that the Surry Unit 1 Cycle 11 reload core can be safely operated to a burnup limit of 14,100 MWD /MTU. The safecy evaluation supplements the evaluation of Technical Koport NE-809 and concludes that an unreviewed safety question does not exist.
SERVICE WATER FLOWS TO RECIRCULATION SPRAY HEAT EXCHANCER SPS, UNIT t- 11/30/90 This change is. to ensure that design basis criteria for i
recirculation spray systems 6.c met for operation of Unit 1 until June 1, 199).- Prior .to this date, a RSHX/SW system inspection / treatment program for hydroids should be developed l and approved.
Testing 1 of the Unit 1 RSHX revealed lower than expected SW flowrates to the RSHXs on 10/14/90 and 10/22/90. An extensive cleaning program resulted. The SW flowpath to the RSHX was l The system was retested. Based on placed in partial wet layup.
l- the results and analyses of the RSHX services flow test, these
! measures and projected service water temperatures ensure that l the system design basis is satisfied through June 1, 1991.
Therefore, an unreviewed safety question is not created.
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e PROCEDURE OR METHOD OF OPERATION CHANGES THAT DID NOT REQUIRE NRC APPROVAL MONTil/ YEAR: NOVEHEER 1990 NONE DURING THIS REPORTING PERIOD
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TESTS AND EXPERIMENTS TilAT DID NOT REQUIRE NRC APPROVAI.
MONTil/ YEAR: NOVEMBER 1990 1-PT-18.4 PERIODIC TEST 11/02/90 (Safety Evaluation #90-0260)
A new test was performed to prove proper operation of check valves 1-SI-130 and 1-SI-147.
The test proves that check valves 1-S1-130 and.1-SI-147 will open fully under design operating conditions. With full residual heat removal flow, these valves are required to-open and properly backstop. The test will be performed with the core offloaded and system operating parameters will not be exceeded.
Therefore, an unreviewed safety question does not exist.
1-OPT-ZZ-001 OPERATIONS PERIODIC TEST-(Safety Evaluation #90-0279) 11/20/90 1-0PT-ZZ-003 OPERATIONS PERIODIC TEST (Safety Evaluation #90-0280) 11/20/90 This test was required to comply with UFSAR Section 1.4 tesn requirements for testing of certain engineered safeguards features.
The Periodic Test will test existing safety systems one train at a time. The unit will be within Technical Specifications and
'have - two coolants loops consisting of one reactor coolant loop and the one residual heat removal pump .not being tested. No changes to systems are required, jumpers and lifted leads for the test require a temporary modification with double o verification. Retesting is performed by the periodic test when the modification is removed. An unreviewed safety question does not exist.'
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VIRGINIA POWER SURRY POWER STATION _
CilEMISTRY REPORT HONTil/ YEAR: NOVEMBER 1990 PRIMARY COOLANT UNIT NO. 1 UNIT NO. 2 ANALYSIS MAX. MIN. AVG. MAX. MIN. AVG.
Gross Radioact., pC1/ml 5.74E2 1.70E-3 6.34E-3 1.48E-1 2.54E-3 3.74E-2 Suspended Solids, ppm 0.0 0.0 0.0 0.0 0.0 0.0 Gross Tritium, pCi/ml 4.94E-2 2.61E-2 3.78E-2 lodine-131, uC1/ml 1.75E-4 5.75E-6 4.31E_-5 2.83E-3 8.96E-6 4.52E-4 Iodine-131/ Iodine-!33 0.15 0.07 0.10 Hydrogen, cc/kg 29.9 5.9 20.3 Lithium, ppm 2.44 0.93 2.01 Boron - 10, ppm
- 421.8 410.2 415.5 261.1 45.1 159.4 4.0 10.005 3.0 0.005 0.005 0.005
.gygen, (DO),' ppm Chloride, ppm 0.012 0.002 0.006 0.008 0.002 0.005 ,
,pH @ 25 degree Celsius 5.64 4.66 5.01 7.31 5.04 6.52
- Boron - 10 = Total Baron x 0.196 REMARKS UNIT TWO-Hydrogen out of specification on 11/30/90 at 0820. Ilydrogen concentration was 19.2 cc/kg while the acceptance criteria is >25 cc/kg. Hydrogen returned to specification at'1235.on 11/30/90.
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UNIT 142 FUEL llANDLING DATE: NOVEMBER 1990 NEW OR DATE NUMBER OF NEW OR SPENT SPENT FUEL SIIIPPED ASSLnBLIES ASSEMBLY ANSI INITIAL FUEL SilIPPING SilIPMENT i OR RECEIVED PER SilIPMENT NUMBER NUMBER ENRICHMENT CASK ACTIVITY LEVEL NONE DURING TilIS REPORTING PERIOD l
1 L
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DESCRIPTION OF PERIODIC TEST (S) WilICil WERE NOT COMPLETED WITilIN Tile TIME LIMITS SPECIFIED IN TECllNICAL SPECIFICATIONS MONTil/ YEAR: NOVEMBER 1990 NONE DURING THIS REPORTING PERIOD
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