ML17308A486

1989 Annual Operating Rept for St Lucie Units 1 & 2 & Steam Generator Inservice Insp Rept

ML17308A486
Person / Time
Site:	Saint Lucie
Issue date:	12/31/1989
From:	Sager D FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-90-77, NUDOCS 9003060110
Download: ML17308A486 (151)

Text

ACCELERATED DIS UTION DEMONS ON SYSTEM 4

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REGULATORY XNFORMATXON DISTRIBUTION SYSTEM (RIDS)

FACIL:50-335 St. Lucie Plant, Unit 1, Florida 50-389 St. Lucie Plant, Unit 2, Florida AUTH.NAME AUTHOR AFFILIATION SAGER,D.A.

Florida Power

& Light Co.

RECIP.NAME RECIPIENT AFFILIATION Document, Control Branch (D

NOTARIZED: NO Power

& Light Co.

Power ight Co ument Control Desk)

DOCKET 05000335 05000389 SUBZECT: Forwards "1989 Annual Operating Rept for St Units 1

2

& 1989 Steam Generator Inservice...,,"

900227 tr.

DISTRIBUTION CODE:

A047D COPIES RECEIVED:LTR Q ENCL SIZE:

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NOTES:

RECIPIENT ID CODE/NAME PD2-2 LA NORRIS,J INTERNAL: ACRS NRR/DET/ECMB 9H NUDOCS-ABSTRACT OGC/HDS2 RES/DSXR/EIB EXTERNAL: EG&G BROWN,B LPDR NSIC COPXES LTTR ENCL 1

0 1

1 6

6 1

1 1

1 1

0 1

1 1

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1 RECIPIENT ID CODE/NAME PD2-2 PD AEOD/DSP/TPAB NRR/DET/EMEB 9H OC EG FILE 0

EG&G RANSOME,C NRC PDR COPIES LTTR ENCL 5

5 1

1 1

1 1

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NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP, US TO REDUCE WASTEI CONTACT THE.DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAMEFROM DISIRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEEDt TOTAL NUMBER OF COPIES REQUIRED:

LTTR 26

,ENCL 23

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P.O. Box14000, Juno Beach, FL 33408-0420 fEBRUARY 2 V 1990 L-90-77 10 CFR 50.36 U. S. Nuclear Regulatory Commission Attn:

Document Control Desk Washington, D.C.

20555 Gentlemen:

Re:

St. Lucie Units 1 and 2

Docket Nos.

50-335 and 50-389 1989 Annual Operating and Steam Generator Inservice Ins ection Re ort.

Pursuant Operating pursuant Generator attached.

to Technical Specification Report for St. Lucie Units 1

to Technical Specification Inservice Inspection Report 6.9.1.4, the 1989 Annual and 2 is attached.

Also, 4.4.5.5, the 1989 Steam for St.

Lucie Unit 2 is Very truly yours, D. A.

er Vice P ident St. Lucie Plant DAS/WFK/gp Attachments cc:

Stewart D. Ebneter, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant 9003060iiO 85'i23i PDR ADOCK 05000335 pnt="

o<I an FpL Group company

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1989 ANNUAL OPERATING REPORT ST LUCIE UNITS 1 AND 2 AND 1989 ST.

LUCIE UNIT 2 STEAM GENERATOR INSERVICE INSPECTION REPORT

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TABLE OP CONTENTS SECTION 1

SECTION 2

SECTION 3

SECTION 4

Annual 10 CFR 50.59 Report Steam Generator In-Service Inspection Mangrove Study Personnel Exposure Summary

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SECTION 1 St. Lucie Annual 10 CFR 50.59 Report A summary of changes to the facility as described in the Final Safety Analysis Report (FSAR)

(10 CFR 50.59 (A)(1)(i)) is submitted by separate letters at the same time as the annual FSAR update for each unit. (July 22 for St. Lucie Unit 1 and April 6 for St. Lucie Unit 2).

Changes to procedures as described in the FSAR (10 CFR 50.59(A)(1)

(ii)) and tests and experiments not described in the FSAR (10 CFR 50.59(A)(l) (iii)) are attached.

10 CFR 50.59 Evaluations Temporary Changes via Jumper/Lifted Lead Requests

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10 CFR 50.59 Evaluations forTemporary Changes via Jumper/Lifted Leads Unit 1

Request Number 9-18 Component/System AGected Hydropneumatic Tank to Domestic Water Desn~ption ofChange This jumper willprevent the loss ofdomestic water system by welding in a new pipe and remove jumper line 42-SW-13.

Safety Evaluation The activity does not increase the probability ofoccurrence ofan accident

'previously identified in the FUSAR, nor increase the consequences ofan accident previously evaluated in the FUSAR.

This activity does not increase the probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the FUSAR, nor increase the consequences ofmalfunction ofequipment important to safety previously evaluated in the FUSAR.

This activity does not create the possibility ofa malfunction ofequipment important to safety ofa different type than previously evaluated in the FUSAR, nor reduce the margin ofsafety as defined in the basis for any technical specification.

Basis: The domestic water system serves no safety function since itis not required for safe shutdown nor to mitigate the consequences ofan accident.

Domestic water is not connected to any system which is a potential source ofradioactive contamination.

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10 CFR 50.59 Evaluations forTemporary Changes via Jumper/Lifted Leads Unit 1

Request Number 9-26 Component/System Affected HVE-8Aand SB, Containment Main Purge Description ofChange This jumper willprevent main purge from "shutting down when MCCs A6 and B6 are de-energized.

Safety Evaluation The activity does not increase the probability ofoccurrence ofan accident previously identified in the FUSAR, nor increase the consequences ofan accident previously evaluated in the FUSAR.

The activity does not increase the probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the FUSAR, nor increase the consequences ofmalfunction ofequipment important to safety previously evaluated in the FUSAR.

The activity does not create the possibility ofa malfunction ofequipment important to safety ofa different type than previously evaluated in the FUSAR, nor reduce the margin ofsafety as defined in the basis for any technical specification.

The FUSAR evaluates the purge assuming containment integrity. This jumper willbe installed only with the main hatch open.

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10 CFR 60.69 Evaluations forTemporary Cages via Jumper/Ufted Leads Unit 1

Request Nuxxdmr 9-28 Component/System Affected TT-1112 HB and TT-1122 HB Description ofChange While trouble-shooting noise, I&Cdiscovered the shield connected to negative (terminal 5) ofoutput, rather than connected to ground (terminal 13), as indicated on CWD 382. The manufacturer recommends connecting shield to ground. This jumper/lifted lead request is to likthe shield from terminal 5 and land iton the transmitter ground.

Safety Evaluation The disabling ofhot leg RTD TE-1122HA and cold leg RTD TE-1112CA inputs to the RPS does not represent an unreviewed safety question because ofthe followingreasons;

1) Withregard to the probability ofaccidents and equipment malfunctions addressed in the FSAR, the RTDs are not utilized in the determination of accident probability. The RPS uses an average hot leg signal (i.e., other hot leg signals are available) and a maximum cold leg signal. Considering these acts, disabling the subject RTDs would have no impact on the probabilities ofaccident or equipment malfunctions addressed in the FSAR.

2) With regard to the consequences ofaccidents and equipment malfunctions addressed in the FSAR, required safety related functions (assuming design basis single failures) willoperate as required. The TM/

LP trip related with these RTDs is not utilized for any steam generator asymmetric events. Withregard to other events which utilize the TM/LP trips, since the RPS channel averages the hot leg signal and selects the maximum cold leg signal, disabling one hot and one cold leg RTD to one RPS channel would not alter the results ofthe transient analysis for four pump operation.

Therefore, the consequences ofaccidents or equipment malfunctions would not be affected.

3) With regard to accidents and equipment malfunctions which are different than those previously evaluated in the FSAR, no new equipment is being added, and safety systms willfunction as assumed in the cycle 6 reload analyses.

Thus no new types ofaccidents or equipment malfunctions have been created.

4) Withregard to the margin ofsafety as defined in the bases for the Technical Specifications, for the reasons outlined in items 1 and 2 above,

. -.-disabling ofthe hot leg RTD TE-1122HA and cold leg RTD TE-1112CA will not result in a reduction ofthat margin.

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10 CFR 50.59 Evaluations forTemporary Changes via Jumper/Lifted Leads Unit 1

Request Number 9-29 Component/System Affected RPS Channel A delta T power Description ofChange TE-1112HA failed high, causing RPS channel Ato trip. The action will repair TE-1112HA and reland leads.

Safety Evaluation The disabling ofhot leg RTD TE-1122HA and cold leg RTD TE-1112CA inputs to the RPS does not represent an unreviewed safety question because

-'f the followingreasons:

1) Withregard to the probability ofaccidents and equipment malfunctions addressed in the FSAR, the RTDs are not utilized in the determination of accident probability. The RPS uses an average hot leg signal (i.e., other hot leg signals are available) and a maximum cold leg signal. Considering these acts, disabling the subject RTDs would have no impact on the probabilities ofaccident or equipment malfunctions addressed in the FSAR.

2) With regard to the consequences ofaccidents and equipment malfunctions addressed in the FSAR, required safety related functions (assuming design basis single failures) willoperate as required. The TM/

LP trip related with these RTDs is not utilized for any steam generator asymmetric events. With r'egard to other events which utilize the TRPLP trips, since the RPS channel averages the hot leg signal and selects the maximum cold leg signal, disabling one hot and one cold leg RTD to one RPS channel would not alter the results ofthe transient analysis for four pump operation. Therefore, the consequences ofaccidents or equipment malfunctions would not be affected.

3) With regard to accidents and equipment malfunctions which are different than those previously evaluated in the FSAR, no new equipment is being added, and safety systms willfunction as assumed in the cycle 6 reload analyses.

Thus no new types ofaccidents or equipment malfunctions have been created.

4) With regard to the margin ofsafety as defined in the bases for the Technical Specifications, for the reasons outlined in items 1 and 2 above, disabling ofthe hot leg RTD TE-1122HA and cold leg RTD TE-1112CA will not result in a reduction ofthat margin.

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10 CFR 50.59 Evaluations forTemporary Changes via Jumper/Doited Leads Unit 1

Request Number 944 Component /System Affected lAand 1B MSIVAirAccumulator Description ofChange This jumper/liRed lead request is to install an electric air compressor on a temporary basis as back-up to the MSIVair accumulator.

The compressor willbe required until a permanent solution to the problem is resolved.

Safety Evaluation The proposed temporary modification has been reviewed and was deemed to not involve an unreviewed safety question since:

1) The probability ofoccurrence or the consequences ofan accident or malfunction ofequipment important to safety previously evaluated in the safety analysis report (FSAR) willnot be increased because the proposed temporary modification willbe performed on the non-nuclear safety related instrument air system and willnot change the operability, degrade closure time, or change any other function or capability ofthe MSIV.

Therefore, the probability ofoccurrence or the consequences ofan accident are not increased.

2) The possibility for an accident or malfunction ofa different type than any evaluated previously in the safety analysis report is not created because there is no change in function or capabilities ofthe MSIVs or their control system. In addition, no new failure modes for safety related equipment have been created.

3) The margin ofsafety as defined in the bases for any Technical Specification is not reduced since there is not change in the ability ofthe MSIVs to respond as required to maintain the bases ofthe applicable technical specfications.

The implementation ofthis change does not require a change to the plant technical specifications.

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10 CFR 60.69 Evaluations forTemporary Changes via Jumper/Lifted Leads Vnit 1

E&gxest Number 9-51 Component/System Affected HVA-3A,Control Room A/C Description ofChange This lifted lead willprevent HVA-3AComp. from starting and willkeep heaters energized.

Safety Evaluation The proposed activity does not increase the probability ofoccurrence ofan accident previously identi6ed in the FUSAR, nor increase the consequences ofan accident previously evaluated in the FUSAR.

The activity does not increase the probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the FUSAR, nor increase the consequences ofmalfunction ofequipment important to safety previously evaluated in the FUSAR.

"I Basis:

9.4;1.'2 Equipment in control room is qualified for higher temperatures than ifall three chiller units were lost.

The proposed activity does not create the possibility ofa malfunction of equipment important to safety ofa di6'erent type than previously evaluated in the FUSAR.

Basis: 9.4.1.2 Control room equipment can withstand loss ofall three chiller units.

The proposed activity does not reduce the margin ofsafety as defined in the basis for any technical speci6cation.

Basis: T/S 3.7.7.1 requires two operable A/C units. Two were operable at the time ofthe jumper/lifted lead.

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10 CFR 50.59 Evaluations forTemporary Changes via Jumper/Lifted Leads Unit 2

Request Number 945 Comyonent/System Affected Aux Feedwater and AFAS Description ofChange

- This action willremove test boxes and.land liftedleads to prevent AFAS

-...'ctuation while performing OP 2-0030127 "Reactor Plant Cooldown - Hot Standby to Cold Shutdown."

Safety Evaluation The activity does not increase the probability ofoccurrence ofan accident previously identified in the FUSAR, nor increase the consequences ofan accident previously evaluated in the FUSAR.

The activity does not increase the probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the FUSAR, nor increase the consequences ofmalfunction ofequipment important to safety previously evaluated in the FUSAR.

Basis: AFAS is not required to be operable in modes 4, 5, and 6, based on T/

S 3/4 3.2.2.

The activity does not create the possibility ofa malfunction ofequipment important to safety ofa different type than previously evaluated in the FUSAR.

Basis: AFAS is not required to be operable in modes 4, 5, and 6.

The proposed activity does not reduce the margin ofsafety as defined in the basis for any technical speci6cation.

Basis: AFAS is not required to be operable in modes 4, 5, and 6.

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10 CFR50.59 Evaluations forTemporary Changes via Jumper/Lifted Leads Unit 2

Request Number 9-16 Component/System Affected, Non-essential load breaker 2A5 MCC Description ofChange This jumper is provided to allow non-essential portion of480V MCC 2A5 to be energized while the relays (B2952) that open the MCC breaker on a SIAS are being moved. This SIAS interlock to the breaker is not needed while the fuel is off-loaded.

Safety Evaluation The activity does not-increase the probability ofoccurrence ofan accident previously identified in the FUSAR, nor increase the consequences ofan accident previously evaluated in the FUSAR.

The activity does not increase the probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the FUSAR, nor increase the consequences ofmalfunction ofequipment important to safety previously evaluated in the FUSAR.

Basis: N/Awith fuel off-loaded.

-f The activity does not create the possibility ofa malfunction ofequipment important to safety ofa different type than previously evaluated in the FUSAR.

Basis: N/Awithfuel off-loaded.

The activity does not reduce the margin ofsafety as defined in the basis for any technical specification.

Basis: N/Awith fuel off-loaded.

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Request Number 9-17 Component/System Affected Non-essential load breaker 2A6 MCC Description ofChange This jumper is provided to allow non-essential portion of480V MCC 2A6 to be energized while the relays (B2952) that open the breaker to this MCC on a SIAS are being moved. This SIAS interlock to the breaker is not needed while the fuel is off-'loaded.

Safety Evaluation The activity does not increase the probability ofoccurrence ofan accident previously identified. in the FVSAR, nor increase the consequences ofan accident previously evaluated in the FUSAR.

The activity does not increase the probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the FUSAR, nor increase the consequences ofmalfunction ofequipment important to safety previously evaluated in the FUSAR.

Basis: N/Awhile fuel is off-'loaded.

The activity does not create the possibility ofa malfunction ofequipment important to safety ofa different type than previously evaluated. in the FUSAR.

Basis: N/Awhile fuel is ofF-loaded.

The activity does not reduce the margin ofsafety as defined in the basis for any technical specification.

Basis: N/Awhile fuel is ofF-loaded.

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10 CFR 60.69 Evaluations forTemporary Changes via Jumper/Uifted Leads Unit 2

Request Number 9-18 Component/SystemAffected Nonessential Load Breaker 2AS MCC Description ofChange This jumper is provided to allow non-essential portion of480V MCC 2AS to be energized while the relays that open the breaker (B2952) to the MCC on a-SIAS are being moved. This SIAS interlock to the breaker is not needed while the fuel is off-loaded.

Safety Evaluation The activity does not increase the probability ofoccurrence ofan accident previously identified in the FUSAR, nor increase the consequences ofan accident previously evaluated in the FUSAR.

The activity does not increase the probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the FUSAR, nor increase the consequences ofmalfunction ofequipment important to safety previously evaluated in the FUSAR.

Basis: N/Awhile fuel is off-loaded.

The activity does not create the possibility ofa malfunction ofequipment important to safety ofa different type than previously evaluated in the FUSAR.

Basis: N/Awhile fuel is off-loaded.

The activity does not reduce the margin ofsafety as defined in the basis for any technical specification.

Basis: N/Awhile fuel is off-loaded.

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10 CFR 60.69 Evaluations forTemporary Changes via Jumper/Mted Leads Unit 2

Request Number 9-20 Component/ System Affected Nonessential Load Breaker 2A5 MCC Description ofChange This jumper is provided to allow non-essential portions of480V MCC 2A5 to be energized while the relays (B2952) that open the MCC breaker on a SIAS are being moved. This SIAS interlock to the breaker is not needed while the fuel is off-loaded.

Safety Evaluation The activity does not increase the probability ofoccurrence ofan accident previously identified in the FUSAR, nor increase the consequences ofan accident previously evaluated in the FUSAR.

The activity does not increase the probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the FUSAR, nor increase the consequences ofmalfunction ofequipment important to safety previously evaluated in the FUSAR.

Basis: N/Awithfuel off-loaded.

The activity does not create the possibility ofa malfunction ofequipment important to safety ofa different type than previously evaluated in the FUSAR.

Basis: N/Awithfuel off-loaded.

The activity does not reduce the margin ofsafety as defined in the basis for any technical specification.

Basis: N/Awith fuel off-loaded.

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10 CFR 60.69 Evaluations forTemporary Changes via Jumper/Lifted Leads Unit 2

Component/System Affected PS-21-4A ICW'"lube water pressure annunciation Description ofChange This change willrevise PC/M 043-287 to-remove PS-21-4A and associated wiring. Lube water for the "A"ICW pump is no longer used. PC/M 043-287 removed the Qow switches, but did not address the pressure switch. With no lube water pressure, annunciator E-15 is locked in.

Safety Evaluation This proposed change does not involve an unreviewed safety question because:

1) The probability ofoccurrence or the consequences ofan accident or the malfunction ofequipment important to safety previously evaluated. in the safety analysis report has not been increased.

The previous accidents evaluated in the FSAR include loss ofone ICW pump and/or loss ofone vital bus. Both accidents result in one fullyfunctional ICW train. The probability ofoccurrence or consequences ofan accident or malfunction does not increase as a result ofthis modification since the probability of failure ofan ICW pump,has not increased and the modifications do not adversely impact the operability ofany safety related functions.

2) The possibility for an accident or malfunction ofa diFerent type than any evaluated previously in the safety analysis report is not created.

The modifications proposed herein do not create the possibility for an accident or malfunction ofa diFerent type that any previously evaluated in the FSAR. Failure modes associated withpump 2A are identical with those previously analyzed failure modes associated with the non-modifiied pumps 2B and 2C, with the exception oflubewater piping failures, the failure ofwhich cannot result in failure ofthe 2A pump. Also, no new active components are added by this modification which could adversely impact other safety related equipment or functions. Failure ofthe modified components willnot impact any other safety related equipment or functions.

3) The margin ofsafety as defined in the bases for any technical specification has not been reduced.

This modification has no adverse impact on the operability ofthe affected ICW system as addressed in the Technical Specifications.

No changes to the Technial Specifications are required and no new technical specifications are required by this modification.

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10 CFR 60.69 Evaluations forTemporary Changes via Jumper/Lifted Leads Unit 2

Eh:cgxest Number 9-37 Component/System Affected 2A ICW annunciator E-15 low lube water pressure/flow Description ofChange This jumper willclear annunciator E-15 by placing PS-21-4A back in service, or remove with supplement to PC/M 043-287.

Safety Evaluation This proposed change does not involve an unreviewed safety question because:

1) The probability ofoccurrence or the consequences ofan accident or the malfunction ofequipment important to safety previously evaluated in the safety analysis report has not been increased.

The previous accidents evaluated in the FSAR include loss ofone ICWpump and/or loss ofone vital bus. Both accidents result in one fullyfunctional ICWtrain. The probability ofoccurrence or consequences ofan accident or malfunction does not increase as a result ofthis modification since the probability of failure ofan ICW pump has not increased and the modifications do not adversely impact the operability ofany safety related functions.

2) The possibility for an accident or malfunction ofa different type than any evaluated previously in the safety analysis report is not created.

The modifications proposed herein do not create the possibility for an accident or malfunction ofa different type that any previously evaluated in the FSAR. Failure modes associated with pump 2A are identical with those previously analyzed failure modes associated with the non-modified pumps 2B and 2C, with the exception oflubewater piping failures, the failure ofwhich cannot result in failure ofthe 2A pump. Also, no new active components are added by this modification which could adversely impact other safety related equipment or functions. Failure ofthe modified components willnot impact any other safety related equipment or functions.

3) The margin ofsafety as defined in the bases for any technical specification has not been reduced.

This modification has no adverse impact on the operability ofthe affected ICW system as addressed in the Technical Specifications.

No changes to the Technial Specifications are required and no new technical specifications are required by this modification.

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10 CFR 50.59 Evaluations forTemporary Changes via Jumper/Ufted Leads Unit 2

Request Number

&47 Component/System Affected 2A2 RCP Upper Seal Cavity Press High and Low.Pressure Annunciator Description ofChange This jumper willsilence an audible alarm.

Safety Evaluation The activity does not increase the probability ofoccurrence ofan accident previously identified in the FUSAR, nor increase the consequences ofan accident previously evaluated in the FUSAR.

The activity does not increase the probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the FUSAR, nor

= increase the consequences ofmalfunction ofequipment important to safety previously evaluated in the FUSAR.

The activity does not create the possibility ofa malfunction ofequipment important to safety ofa different type than previously evaluated in the FUSAR, nor reduce the margin ofsafety as defined in the basis for any technical specification.

Basis: RCPs are not safety-related.

They are not required for the safe shutdown ofthe plant. Forced circulation is the preferred method for cooling but natural circulation provides the necessary heat removal capability when RCPs are unavailable.

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'0 CFR 5059 Evaluations forTemporary Changes viaJumperILifted Leads Unit 2

Request Number 949 ComponentSystem Affecte 2B1 RCP Reverse Rotation and Gasket Leakage Pressure Switch Description ofChange This action willflixground on 2Bl RCP and restore lifted leads.

Safety Evaluation Removing the annunciator capabilities from FS-1176 and FS-1177 would not constitute an unreviewed safety question based on the following:

(1) The probability of occurrence or the consequences of an accident or malfunction previously evaluated in the FSAR is not increased.

Both flow switches monitor oil flow in the lube oil system at the main thrust bearing bracket. In order for a reverse rotation condition to exist, the RCP would have to be stopped and restarted, and the Anti-reverse Rotation device would have to malfunction in its entirety. Ifthe RCP were to be stopped while the unit was at power, the unit would trip due to low RCS fiow(RPS trip unit). Should the pump be restarted, and the motor somehow reverse rotation, one pin in the anti-reverse rotation device is capable'f holding the pump stationary against the torque produced by the application of 100% voltage in such a reverse phase rotation. Therefore, even though the control room operators were unaware of a reverse rotation condition existing, the anti-reverse rotation device would prevent the pump from turning in the wrong direction.

(2) The possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created.

When the RCP shak stops rotating when the motor is stopped, the pins in the anti-reverse rotation device prevent the RCP from rotating in the reverse direction, even against 100% voltage applied to the motor in the reverse phase.

With annunciator capabilities lost to the operators, itis conceivable that the pump may be restarted; while damage to the motor may result, and the anti-reverse rotation device would prevent backflow through the RCP and the 2B steam generator, thus corresponding to a cooldown using less than four RCPs for RCS circulation. As the unit is designed to accommodate a natural circulation cooldown, no new accidents or malfunctions are assumed to be created.

(3) The margin of safety as defined in the basis for the Technical Specifications is not reduced.

Reverse rotation indication is not required by Tech Specs.

Removing the annunciator capabilities on PS-1170 does not constitute an unreviewed safety question based on the following:

(1) The probability ofoccurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.

PS-1170 is discussed in section 5.2.5, "Detection of Leakage Through RCS Pressure Boundary," of the FSAR. The primary indications ofRCS leakage are given as the containment sump level and containment radioactivity alarms.

As the loss of annunciator in the 2Bl RCP gasket pressure switch does not affect any ofthese primary indications, no existing analyses are affected.

(2) A possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR is not created.

With no indication of gasket leakage in the control room, the operators would still be made aware of any RCS pressure boundary leakage

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10 CFR50$ 9Evaluations forTemporary Changes viaJumperILifted Leads Unit 2

Request Number 949 Componentaystem Affecte 2B1 RCP Reverse Rotation and Gasket Leakage Pressure Switch Safety Evaluation (continued.)

'etectable by the RCS Pressure Boundary leakage detection system.

This system is capable of detecting unidentified leakage of 1.0 gpm or less within one hour. This redundant capability ensures that RCS inventory is maintained within analyzed limits.

(3) The margin of safety as defined in the basis for any technical specification is not reduced. Unit 2 TS 3.4.6.1 and 3.4.6.2 state the operability requirements for the RCS leakage detection systems and RCS leakage.

The inoperability of the RCP gasket pressure switch is not included in these technical specifications and is bounded by the capabilities of the leakage detection system.

Therefore, the margin of safety defined by the bases for the technical specifications is not affected.

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10 CFR 50.59 Evaluations Summaries of Evaluations Approved by St. Lucie Facility Review Group

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St. LucieUnit1 Potential Leaking Steam Generator Tube Plugs

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Introduction:==

During maintenance operations at St. Lucie Unit 1, visual inspection of the primary side of the tubesheet revealed a total of approximately sixteen plugs in steam generator A and five plugs in steam generator B exhibiting signs ofpotential leakage.

The plugs in question are Westinghouse mechanical plugs installed during the December 1985 refueling outage.

In addition, eleven plugs in steam generator Aand six in steam generator B installed at the same time as the plugs with apparent signs ofleakage were not inspected due to inspection fixture interference.

Safety Evaluation:

Itis concluded that an unreviewed safety question does not exist relative to the potentially leaking plugs since (1) The probability of occurrence of a design basis accident is not increased since the complete circumferential severance ofa plug resulting in a plug top release willnot rupture the parent tube.

'2)

The consequences of a previously postulated design basis accident are not made more severe since no accident mitigating equipment or systems have been altered and a steam generator severed tube plug would not be expected to produce any greater eAects than'hose analyzed for a rupture of a steam generator tube as discussed in the SAR. The mechanical expander functions as a leak limitingorifice and the maximum leakage expected through the severed plug is less than 80 gpm. This leakage is substantially less than the design basis tube rupture.

(3) The possibility of an accident of a different type than previously addressed in the SAR does not exist since a steam generator tube rupture concurrent with tube plug failure is considered not a possible failure mode.

Only tube plug leakage (within Tech Spec limits) is expected for the limitingcase of a tube fish-mouth failure.

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St. LucioUnit1 LPSIPunip 1AGland Plate Gasket Groove Modi6cation

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Introduction:==

This safety evaluation permits the removal of.022" ofmetal from the inboard face ofthe 1A-LPSI pump gland ring until the next scheduled teardown ofthe pump. Atthat time, the gland should be restored to the original design to eliminate the need to maintain unique gaskets in stores for the two pumps.

Safety Evaluation:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety evaluated in the FSAR has not been increased because:

1) the design minimum wall ofthe LPSI pump pressure boundary has not been changed.

2) the removal of.022" ofmetal from the face ofthe gland ring willnot have an adverse effect on the operation ofthe mechanical seal.

Therefore, the probability ofLPSI pump failure has not been increased and the consequences of a LPSI pump failure remain the same.

The possibility of an accident or malfunction of a different type than any previously evaluated in the SAR has not been created since this modification does not alter the operational'characteristics ofthe lALPSI pump. In addition, this change affects no other system design bases.

Therefore, no new accident or malfunction is created.

The margin of safety as defined in the basis for any Technical Specification has not been reduced because the Technical Specification requirement for two operable LPSI pumps is maintained.

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St. LucieUnit2 Increase pHLimitinthe Secondaxy System

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Introduction:==

This change willincrease the secondary system feedwater/condensate pH specification to permit-operation with a feedwater pH up to 10.0. Increasing the pH in the secondary system reduces erosion/corrosion in the secondary system and reduces the transport of corrosion products to the steam generators.

Safety Evaluation:

The change does not increase the probability of occurrence or the consequences ofan accident or malfunction of equipment important to safety previously evaluated in the safety analysis report because:

1) increasing pH in the secondary system willreduce corrosion product transport by reducing corrosion offerrous materials,

2) increasing pH in the secondary system willnot increase corrosion of other secondary materials

3) increasing pH willnot increase the likelihood of caustic stress corrosion cracking of the steam generator tubing because the chemicals used are volatile.

The change does not increase the possibility of an accident or malfunction of a different type than any previously evaluated in the SAR because:

1) only chemicals that have been previously considered by the FSAR are used to increase the pH in the secondary system, and

2) the change willnot result in other forms of corrosion attack on the steam generator tubing.

The change does not reduce the margin of safety as defined in the basis for any technical specification because:

1) increasing pH reduces corrosion in the secondary system,

2) increasing pH does not increase corrosion of the steam generator materials, and

3) the change is consistent with T.S. B 3/4 k 4.5 to operate in such a manner that the secondary coolant willbe maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes.

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St. Lucie Plant Temporary Installation ofthe Intake Canal Sonar Zone Isolation Net

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Introduction:==

'lant Change/Modification 045-986 installed an underwater intrusion detection system.

An unacceptable number ofalarms were occurring, and have been attributed to the number oflarge fish in the intake canal. In order to move these fish away from the system, the plant has asked the Land Utilization department to temporarily install a net across the intake canal downstream of '--

the detection system. After a net is installed, the detection system willbe operated for a thirty day period to verify that itis operating properly.

Safety Evaluation:

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Placing this net temporarily across the intake canal does not involve an unreviewed safety question, and the followingare the bases:

(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased.

The temporary placement of.

~ the net has no impact on the function ofthe intake canal, and therefore no impact on the plant. The net has no ability to impact any analysis previously evaluated in the SAR.

(2) The possibility for an accident or malfunction of equipment of a different type than any previously evaluated in the SAR is not created.

There is no concern during the installation phase ofthe project due to the large number ofpeople and equipment that willbe utilized. Once the net is installed, the configuration that willbe used to hold itin place leaves it single failure proof. The net willbe moored using two large fluked anchors, a 1/2" stainless steel fail safe cable attached to the intruder barrier, and three ropes anchoring the top ofthe net to the two banks.

(3) The margin of safety as defined in the basis for any Technical Specification is not reduced by this temporary change.

The change has absolutely no impact on Tech Spec 3.7.4 for Unit 1 and 2 which requires two independent loops of ICW to be available. The temporary installation ofthis net does not therefore reduce the margin of safety as defined in the basis for any Tech Spec.

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St. LucieUnit1 Increase pHLimitinthe Secondary System

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Introduction:==

This change willincrease the secondary system feedwater/condensate pH specification to permit operation with a feedwater pH up to 10.0. Increasing the pH in the secondary system reduces erosion/corrosion in the secondary system and reduces the transport of corrosion products to the steam generators.

Safety Evaluation:

The change does not increase the probability of occurrence or the consequences ofan accident or malfunction ofequipment important to safety previously evaluated in the safety analysis report because:

1) increasing pH in the secondary system willreduce corrosion product transport by reducing corrosion of ferrous materials,

2) increasing pH in the secondary system willnot increase corrosion of other secondary materials

3) increasing pH willnot increase the likelihood of caustic stress corrosion cracking of the steam generator tubing because the chemicals used are volatile.

The change does not increase the possibility of an accident or malfunction of a different type than any previously evaluated in the SAR because:

1) only chemicals that have been previously considered by the FSAR are used to increase the pH in the secondary system, and

2) the change willnot result in other forms ofcorrosion attack on the steam generator tubing.

The change does not reduce the margin of safety as defined in the basis for any technical specification because:

1) increasing pH reduces corrosion in the secondary system,

2) increasing pH does not increase corrosion of the steam generator materials, and

3) the change is consistent with T.S. B 3/4 5 4.5 to operate in such a manner that the secondary coolant willbe maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes.

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St. LucieVnit2 BoricAcidMakeup System Insulation Removal

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Introduction:==

The request to remove insulation was due to the fact that (1) Plant Change/Modification 283-288 permits the temporary de-energization ofportions of the BANsystem heat tracing and 2) the insulation was removed over a period of time by various departments for various reasons.

The removed insulation is considered to be a temporary change concurrent with the trial implementation of the Boric Acid Concentration Reduction Technical Specifications and is effective for the duration of the PSL-2 Cycle 5 and subsequent end-of-cycle refueling outage.

Safety Evaluation:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR has not been increased because:

1) the de-energization ofheat tracing for the BAMsystem piping and valves eliminates the need for insulation, and

2) the removal ofthe insulation willnot affect the seismic qualification ofthe piping because:

a) reduced weight on piping increases the natural frequency ofthe piping and therefore seismic acceleration/loads

decrease, b) reduced weight decreases loading on supports/restraints, and c) the removed insulation does not provide a support/restraint function in conjunction with any seismic support.

Therefore, the probability ofBAMsystem failure has not been increased and the consequences of the BAMsystem failure remains the same.

The possibility of an accident ofmalfunction of a different type than any evaluated previously in the SAR has not been created since this modification does not alter the operational characteristics ofthe BAMsystem.

In addition, this change affects no other system design bases because the safety requirements for the insulation are not required due to the de-energization ofheat tracing.

Therefore, no new accident or malfunction is created.

The margin of safety as defined in the basis for and Technical Specification has not been reduced because the Technical Specification requirement for BAMsystem heat tracing was eliminated as a result of the Boric Acid Concentration Reduction effort.

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1 St.LucieUnit 2 Core Shroud Indications Found During 1989 RefuelingExamination

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Introduction:==

Two anomalous indications were noted on the core shroud during the 1989 refueling outage inspection.

FPL requested a safety evaluation from Combustion Engineering to determine acceptability ofthe existing condition ofthe core shroud.

C-E completed the evaluation and determined that the Core Shroud is acceptable for continued use in the existing condition, and there are no changes to the Technical Specifications required.

Safety Evaluation:

The probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased because the indications do not-affect fitor function ofthe Core Shroud and the damage has no effect on the structural integrity of the shroud.

The design basis accidents or malfunctions are not impacted by the changes because no additional stresses are induced by the indications and there are no mechanisms present for cracking or further damage to develop.

As a result, the consequences of an accident or malfunction are not increased.

The systems important to safety are not impacted by these anomalies: therefore there are no effects on failure modes associated with the probability ofa failure ofa system important to

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The possibility for an accident or malfunction of a different type than any previously evaluated in the SAR is not created.

The noted anomalies do not result in a new unanalyzed condition because

'he fit,function, and strength ofthe core shroud are not affected, therefore the change does not impact plant operation or response.

The margin of safety as defined in the bases for any Technical Specification is not reduced.

There is no reduction on the margin of safety as defined in the bases ofthe Technical Specifications.

The Technical Specification requirements are met by the noted anomalies because.

there is no increase in the probability of exceeding a safety limit.

The change does not degrade the performance of a safety system below that determined in the SAR.

There are no changes to the consequences on any protective boundary and there is no impact on the Technical Specifications.

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St. LucieUnit2 Increased RCS LithiumProgram

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Introduction:==

The purpose ofthe pH chemistry control program is to minimize the inventory of corrosion products circulating in the RCS and reduce the amount of corrosion products activated (i.e., crud) by minimizing its deposition on fuel cladding surfaces.

The eventual release of crud from fuel surfaces and accumulation in cooler RCS regions, such as the steam generators, is the principal contributor to out-of-core radiation field buildup. The basis for operating with a pH in the 6.9 range was that the composition ofthe crud was believed to be magnetite, and accumulation ofcorrosion products on the fuel surfaces would be minimized.

Safety Evaluation:

The probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased because the effect ofoperating at the higher lithium and pH levels proposed does not result in an increase in the probability offuel failure either directly or indirectly. Corrosion failure is not probable as neither sufficient lithium concentration nor a high enough pH exists for accelerated corrosion to occur. Nor is there any secondary failure mechanisms that might increase the probability of fuel cladding failure.

Corrosion levels willnot be'reater than previously experienced, and no accelerated hydrogen pickup due to lithium has been demonstrated, consequently no weakening of the cladding or failure due to hydriding willoccur.

- e Neither the probability nor the consequences of any accident previously analyzed are increased because operation at a higher lithium concentration and pH does not result in fuel corrosion or mechanical behavior either greater than or different from previously considered in the input to any safety analysis.

The proposed increase in lithium concentration does not increase the materials'orrosion rate nor does itincrease the incident ofstress corrosion ofthe components wetted by the primary coolant or its letdown to other systems.

The slight increase in operating pH willhave no adverse effect on the design life or performance of equipment important to safety, based on ample test data.

Therefore, the increase in lithium concentration and pH does not increase the probability ofan accident or malfunction of equipment important to safety.

The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR is not increased, because neither the fuel nor its modes ofoperation willbe changed by operating at the proposed lithium and pH limits. No changes to the fuel or to the fuel operating environment, other than increasing the lithium and pH, are being proposed.

Because these changes have no impact on the design ofthe fuel or its operation, the possibility of creating a new or different type of accident than previously analyzed in the FSAR is not created.

The margin of safety as defined in the basis of any technical specification is not reduced because the limitations offuel performance include the power distribution limits which ensure that peak fuel cladding temperatures, heat rate, power density and DNB related parameters are maintained within their analyzed normal and transient conditions in the safety analyses.

There are no changes in any fuel analysis input or assumptions are required as a result of the proposed change nor are any changes to analysis methodology necessary to describe fuel rod behavior.

As no inputs, assumptions, or methods have changed, the results ofprevious safety analyses remain unchanged and the margin of safety for any technical specification is not reduced.

Allof the fuel performance parameters described previously are not adversely affected'

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St. LucieUnit2 Increased RCS LithiumProgram by the proposed increases in lithium concentration.

The limitations on the RCS chemistry ensure that corrosion ofthe components wetted by primary water is minimized and reduce the potential for RCS leakage or failure due to stress corrosion.

Increasing the lithium concentration as proposed willnot reduce the corrosion protection nor increase the potential for stress corrosion, and the structural integrity ofthe wetted components-.-

willnot be adversely affected.

The previously analyzed performance of wetted materials regarding stress levels and fatigue are not changed.

Therefore, the margin of safety as defined in-the bases of any technical specification is not reduced.

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St. LucieUnit1 Increasing RCS LithiumLevel

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Introduction:==

The purpose ofthe pH chemistry control program is to minimize the inventory of corrosion products circulating in the RCS and reduce the amount ofcorrosion products activated (i.e., crud) by minimizing its deposition on fuel cladding surfaces.

The eventual release of crud from fuel surfaces and accumulation in cooler RCS regions, such as the steam generators, is the principal contributor to out-of-core radiation field buildup. The basis for operating with a pH in the 6.9 range was that the composition ofthe crud was believed to be magnetite, and accumulation ofcorrosion products on the fuel surfaces would be minimized.

Safety Evaluation:

The proposed increase in lithium concentration does not increase corrosion nor increase the corrosion rate nor increase the incidence of stress corrosion ofthe components wetted by primary coolant or its letdown to other systems.

The resultant increases to a slightly more basic primary chemistry willhave no effect on the probability of an accident or equipment failure, since the material integrity or performance of equipment important to safety willnot be adversely affected.

The operability of the spray addition system ensures that sufficient amounts of sodium hydroxide is added to the containment spray in the event ofa LOCAto maintain a pH value of8.5 and 11.0.

The increase in lithium concentration is negligible compared to the sodium hydroxide concentration in the containment spray and containment sump solution and does not alter the resultant pH for the analyzed events.

The basis for the spray additive system and the consequences of an analyzed event is not affected.

Therefore, the proposed increase in lithium level does not increase the probability or consequences of an accident or equipment malfunction previously evaluated.

The corrosion rates willnot increase for the plant components in contact with primary coolant or letdown. The design life ofthese components is not reduced and therefore the probability ofany failure is not increased, and the operation of the plant is not affected.

The proposed increase in lithium level willnot create a malfunction or a different failure mechanism than previously evaluated.

The limitations on the RCS chemistry defined in the technical basis ensure that corrosion ofthe components wetted by primary water is minimized and reduce the potential for RCS leakage or failure due to stress corrosion.

Increasing the lithium concentration as proposed willnot reduce the corrosion protection nor increase the potential for stress corrosion of any affected components.

The increase in lithiumwillnot be detrimental to the safety margin or performance ofthe wetted components, since the structural integrity willnot be adversely affected.

Therefore, the proposed change in lithium levels willnot reduce the margin of safety as defined for any technical specification basis.

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St. LucieUnit2 DrainValveforLT-110$ LowPressure Connection, Installation ofPlug

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Introduction:==

'The drain valve for the level transmitter (LT-1103) for the pressurizer level - cold condition was leaking at the seat.

No replacement valve was available.

The proposed modification to prevent leakage is to plug the downstream side ofthe valve. A 1/2" plug 4" long willbe welded into the outlet ofthe subject valve.

Safety Evaluation:

This change does not involve an unreviewed safety question and the following are the bases for - '-

this conclusion.

(1) The probability of occurrence or the consequences or an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased because the plugging ofthe valve prevents the leakage problem but does not hinder the performance of the level transmitter.

The added weight of the valve is insignificant and willnot create a seismic concern.

(2) The possibility for an accident or malfunction of a diA'erent type than any evaluated previously in the safety analysis report is not created, since the plugging of a drain valve does not-aFect the operation of any safety equipment.

(3) The margin of safety as defined in the basis for any technical specification is not afFected by this change, as mentioned above. This change requires no change to the Unit 2 Technical Specifications.

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St. LucieUnit2 Safety Evaluation ofRegulatoxy Guide 1.63

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Introduction:==

Regulatory Guide 1.63, "Electric Penetration Assemblies in Containment Structures for Light Water Cooled Nuclear Power Plant," presents methods acceptable to the NRC staff for complying

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with General Design Criterion 50 ofAppendix Aand with Appendix B with respect to the mechanical, electrical, and test requirements for the design, qualification, construction, installation, and testing of electric penetration assemblies in containment structures oflight water cooled nuclear power plants.

Safety Evaluation:

The probability of occurrence or the consequences ofan accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased.

The identification ofthe proper mode ofoperation in the evaluation indicates that power isolation ofthe SIT isolation valves and SDC valves does not increase the probability of occurrence or the consequences of an accident.

The possibility for an accident or malfunction ofa di6'erent type than previously evaluated is not created.

The Conax test report has shown that a failure ofthe electrical penetration is not possible.

In addition, the technical specification has shown that providing power to these circuits until mode 2 is in conformance with R.G. 1.63. Therefore, the possibility for an accident or malfunction of a different type is not created.

The margin of safety as defined in the technical specification is not impacted by this safety

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The fact that some ofthese circuits do not have dual isolation devices does not violate technical specification 3.6.1.1.

Failure of the electrical penetration would not occur after a multi-phase fault.

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St. LucieUnit2 Use ofCrouse-Hinds FS/613 Boxes on Target Rock Solenoid Valves

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Introduction:==

Currently at St. Lucie, a gasketed Grouse-Hinds (C-H) FS/FD box is being used as the slice box for Target Rock model 74@ solenoid valves. The purpose ofthis evaluation is to demonstrate that this box is sufFicient to permit limitswitches to remain operational for the required ninety minutes

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Safety Evaluation:

The limitswitches willbe able to provide indication for at least the required operability time and thus accomplish their safety related function. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated-in the SAR are not increased.

The use ofthe C-H FS/FD box willnot increase the possibility of an accident or malfunction of a different type of any evaluated previously in the FSAR since the limitswitches willoperate as required.

Thus no new modes offailure are being introduced.

The margin of safety as defined in the basis for any technical specification willnot be reduced since the operation of the solenoid valves is not affected.

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St. LucieUnit2 CCWHeat Exchanger Operation withTen Percent Tube Plugging Limitation

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Introduction:==

During the St. Lucie 2 1989 refueling outage, the component cooling water heat exchanger (CCW-Hx) tubes were examined using Eddy Current Testing (ECT). The ECT results of the "B" CCW Hx indicated that 105 tubes contained credible flaws with 70% or greater through-wall penetration.

Seventy percent has been established as the maximum permissible flaw depth to preclude leakage during the next cycle ofoperation. This amount, combined with the tubes already plugged, totals 186, or about 9.3% ofthe total CCW Hxtube population of1996.

The operability of the CCW system is defined as having sufficient cooling capacity for continued--

operation of equipment during normal and accident conditions. The analysis of the Spent Fuel Pool Cooling System and the analysis for the design basis accident utilize a maximum CCW system service temperature of 100'F. The description for the Intake Cooling Water System (ICW) which services the CCW Hx is sized to ensure adequate heat removal at design inlet temperature of 95'F. However, the ICWinlet temperature has not been recorded to exceed 89'F. The purpose ofthis safety evaluation is to demonstrate that operating St. Lucie 2 with up to 10% ofthe tubes plugged in each CCW Hx with a maximum Intake Cooling Water inlet temperature of 90'F does not constitute an unreviewed safety question.

Safety Evaluation:

The probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased.

ICW services the CCW Hx, and the previously analyzed CCW Hx outlet temperatures willbe maintained for both normal and accident heat load. At the changed (lower) ICW. design inlet temperature it was determined that 10% ofthe CCW Hx tubes could be plugged. The change in ICW design inlet temperature does not create the consequences of an accident because the analyzed temperatures in the Safety Analysis willbe maintained at the beginning and during the postulated accidents.

The probability or consequence of equipment malfunction is not increased since the CCW Hx outlet temperatures willstill be maintained within design specifications for all the safety-related equipment or equipment important to safety.

The possibility for an accident or malfunction of a type different than any evaluated previously in the SAR is not created.

The lower ICW design inlet temperature provides sufficient CCW Hx margin to allow 10% ofthe tubes to be plugged. Although the CCW Hx surface area is reduced, the CCW Hx outlet temperature is maintained within the specified and analyzed values provided equipment specifications and the FSAR Since the design bases and performance ofthe CCW system are not adversely affected with 10% tubes plugged and a 90'F design ICW inlet temperature, the possibility of an accident or malfunction of a type different than that previously evaluated in the FSAR is not created.

The bases for the CCW and ICW systems operability provided in the technical specification is to ensure sufficient cooling capacity is available for continued operation of equipment during normal and accident conditions. The redundant cooling capacity of these systems, assuming a single failure, is consistent with the assumptions used in the safety analysis.

The CCW Hx was designed and built with a specific amount of safety margin (i.e., excess tube surface area) for servicing the normal and accident heat loads.

The calculation to justify a 10%

tube plugging limitwith a design seawater temperature of90'F retains that amount ofsafety margin, while demonstrating that the CCW Hx outlet temperature willbe maintained as previously analyzed in the FSAR. The redundancy of either system is not impacted by this change

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> reactor. As such, only one hot leg plug can be installed with fuel in the reactor vessel. Safety Evaluation: The change described herein does not increase the probability of occurrence or the consequences of-an accident or malfunction ofequipment important to safety previously evaluated in the safety analysis report. The installation, use and removal of a plug in only one hot leg at a time with fuel in the reactor vessel willoccur only in mode 6 when the water level above the top ofthe reactor pressure vessel flange is greater than or equal to twenty-three feet. During this time only one shutdown cooling loop is required to be operable and in operation. Since only one hot leg may be plugged at a time, this change does not affect the decay heat removal capabilities ofthe Shutdown Cooling System. The consequences of a hot leg plug seal failure in regard to loss of decay heat removal is similar to that ofa pool seal failure in that both the SG manway and the pool seal are above the horizontal centerline of the hot legs. To maintain residual heat removal capability, the water level must be above the horizontal centerline ofthe hot leg (mid loop). Potential paths ofRCS inventory loss have been considered. No inventory willbe lost out ofthe steam generator hot leg manway or the RTD thermowell nozzle holes because they are located above the centerline; RCS inventory can be lost from the drain line ifthe isolation valve is not shut. Inventory lost through this nozzle is 54 gpm and is well within the makeup capacity ofthe shutdown cooling system. In addition, three hot leg RTD nozzles were assumed to be cut out and conservatively located at the centerline. The flow rate required to equal this loss was calculated to be approximately 100 gpm. This combined leakage flow rate of 154 gpm, to maintain greater than mid-loop operation, is also well within the makeup capabilities of available makeup sources. These factors willensure that shutdown cooling willnot be lost due to air binding ofthe pump and that residual heat removal capability willbe maintained. The change does not create the possibility for an accident or malfunction ofa different type than previously evaluated in the safety analysis report. The installation, use and removal of a plug in only one hot leg at a time with fuel in the reactor vessel is bounded by the current safety analysis report. This includes the loss of shutdown cooling, refueling pool drain down due to hot leg plug seal failure, and fuel failure due to liftingheavy loads over the reactor vessel. No changes to systems or plant conditions willbe made which affect or change the previously evaluated safety analysis. These changes therefore do not create the possibility for an accident or malfunction of a different type than previously evaluated. The change does not reduce the margin of safety as defined in the basis for any technical specification. The installation, use and removal of a plug in only one hot leg at a time with fuel in the reactor vessel does not require any changes to the technical specifications. The technical specifications require that only one shutdown cooling loop be operable and in operation when in mode 6 and ifthe water level above the top ofthe reactor pressure vessel flange is greater than or equal to twenty-three feet, the bases for the technical specifications remain unchanged. These changes therefore do not reduce the margin ofsafety as defined in the basis for any technical specification. 1 sq ~ 1 ~E lI e St. LucieUnit1 Instrument AirSupply to Main Steam Isolation Valve == Introduction:== An instrument air line for MSIVI-HCV-08-1Ais a 1/4 inch copper tubing line which functions as a-bypass line to main instrument air supply line with solenoid valve 7A. The main instrument air line with solenoid valve 7A supplies air to MSIVto keep it open. Plant personnel found the subject 1/4 inch copper tubing line leaking. In order to keep the MSIVI-HCV-08-1Aopen and not lose the air pressure, plant personnel performed an interim replacement by installing a 1/4 inch stainless steel (SS) braided hose line in place ofcopper tubing. Safety Evaluation: The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in the safety analysis report willnot be increased. The possibility of an accident or malfunction ofa different type than any previously evaluated in the safety analysis report willnot be created. The margin of safety as defined in the bases for technical specifications is not reduced. The installation ofSS braided tubing in place ofcopper tubing does not create any change to the existing design from system viewpoint. The SS braided tubing line assists in keeping the MSIVI-HCV-08-1A open during a test mode. Any malfunction in this line willinitiate closure of the MSlVI-HCV-08-1A, which is the intended safety function. Thus the design change does not impair the intended safety function. Therefore, the probability of occurrence or the consequences ofan accident or malfunction of equipment important to safety as defined in the SAR is not increased. Amalfunction in the 1/4 inch tubing willnot prevent the MSIV associated with this line from performing its intended function. Therefore, no new type of accident or malfunction is created. The subject V4 inch line does not have any effect on the technical specification requirements for the MSIV. Therefore, the margin of safety as defined in the bases for technical specifications is not reduced. 'I 'I pe A e ~\\ ~ I ~ ~ St. LucieUnit2 Cycle 5 Reload Safety Evaluation == Introduction:== This reload evaluation concerns modifications made to the fuel being loaded into the core. Fuel assemblies in region G willbe modified to render them more resistant to debris-induced failure. Safety Evaluation: " The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. The St. Lucie Unit 2 Cycle 5 reload design does not change the overall configuration of the plant. The mode of operation of the plant remains unchanged. Changes to the fuel assembly design to make the assembly resistant to debris-induced fretting do not change the mechanical or thermal-hydraulic performance of the fuel assembly. The mode 6 Boron Dilution Event was reanalyzed and the results were found to meet acceptance criteria. The Small and Large Break LOCA events were re-evaluated against the Cycle 5 fuel design changes and shown to meet acceptance criteria., The Reload Safety Evaluation report demonstrates that the consequences of an accident or ~ malfunction have not been increased beyond those evaluated in the previous analyses since all transients meet current criteria. A possibility for an accident or malfunction of a different type than any previously evaluated in the safety analysis is not created. The St. Lucie Unit 2 Cycle 5 reload design does not change the overall configuration of the plant. The design of the debris resistant Region G assemblies is the same as that of other assemblies residing in the core, with the exception ofminor lower end fitting changes and the use oflonger end caps on fuel rods. Therefore, a possibility for a new accident or equipment malfunction has not been created. The margin of safety as defined in the basis for any Technical Specification is not reduced. The St. Lucie Unit 2 Cycle 5 reload design neutronics input and the resulting safety analysis has been reviewed, and in all cases the results are well within the acceptance criteria ofthe design basis. Based on FPL's independent review ofthe RSE report it can be concluded that the St. Lucie Unit 2 Cycle 5 reload design does not result in a reduction to the margin of safety relative to the Technical Specification basis for St. Lucie Unit 2 Cycle 5. ~ ~ 1 ~ i '+ Wl I ,'lt >-I 4 ~~ -k , o f Wtf St. LucieUnit2 Operation ofthe AICWHeader withPressure Transmitter PT-21~Temporarily Out OfService == Introduction:== Pressure transmitter PT-21-SA provides annunciation in the control room for low pressure in the Unit 2 "A"ICW header. There remain alternate methods for ensuring proper operation ofthe "A"- header. This situation willcontinue until a replacement can be installed or the present transmitter replaced. Safety Evaluation: The probability of occurrence or the consequences of an accident or-malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. Operation ~ ofthe "A"ICWheader with PT-21-SA temporarily out ofservice is bounded by the ICW design basis to provide a heat sink for the CCW system under design basis accident conditions, assuming a single failure coincident with a loss of offsite power. The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created. There are several alternative parameters that can be monitored to ensure proper operation ofthe "A"ICW header with PT-21-SA temporarily out of service. This does not create any new scenarios for accidents or malfunctions of equipment. The margin of safety as defined in the basis for any Technical Specification is not affected by this. change. PT-21-SA does not directly affect operability of the "A"ICWheader. The bases ofthe technical specifications is not affected as the header remains available with several alternate means of monitoring its performance. This change requires no change to the Unit 2 Technical Specifications. l, ~ ~ WHO' ~b N ~,~ ~ ~ ~ ~~v~r ~ $ R law St. Lucie Units 1 and 2 Change inAuxiliaryCoolingWater Systems Corrosion InhibitorSpecification == Introduction:== St. Lucie 1 and 2 used potassium dichromate as the corrosion inhibitor for the auxiliary cooling water systems from initial startup until the summer of 1985 when a change to molybdate chemistry was made. Since the changeover to molybdate/TTA chemistry control, the plant has reported problems with occasional "crud bursts" in the systems and an on-going problem with high suspended solids in the Unit 2 CCW system. The change willadd nitrates (200-500 ppm) to the current corrosion inhibitor solution used in the auxiliary cooling water systems. Safety Evaluation: The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FUSAR are not increased. (1) The proposed corrosion inhibitor serves the same function as that ofthe current inhibitor. (2) The proposed corrosion inhibitor in no way affects the ability of the CCW system to perform as designated. (3) The proposed corrosion inhibitor provides improved corrosion inhibition for the CCW system components to help minimize corrosion and maintain integrity of the system. Attachments F, 6, and H ofthis report document corrosion rates for sodium-based and potassium-based molybdate/nitrate corrosion inhibitor programs. When used in recommended concentrations, these programs provide as effective corrosion inhibition as the original potassium chromate chemistry. (4) Higher concentrations (500 ppm) of the corrosion inhibitor willprovide increased system corrosion protection capability. (5) Although the corrosion inhibitor is normally purchased in a premixed solution, the ability to purchase individual components to the specifications listed in the chemistry specifications book willallow the system chemistry to be adjusted to provide the maximum system protection. The possibility of an accident or malfunction ofa different type than any previously evaluated in the FUSAR willnot be created. (1) No.mechanical components are. changed or altered. The proposed corrosion inhibitor is more effective at minimizing corrosion than the inhibitor currently used. The material is non-toxic and non-hazardous in the concentrations used. (2) Ifa CCW/RCS interface leak were to occur during operation, leakage would be from the RCS into the CCW system. The CCW system is equipped with a radiation monitor to determine if leakage exists between the RCS and CCW. Ifa leak developed during shutdown, the CCW could enter the RCS. Routine sampling of RCS fluids as specified in plant chemistry procedures would identify a contaminant intrusion. The margin of safety as defined in the basis for any technical specification is not reduced. (1) The plant technical specifications do not specify a particular type of corrosion inhibitor. (2) Nitrites are a proven corrosion inhibitor, and have been shown to enhance the capability of molybdates to control corrosion in closed cooling water systems. Testing has shown that solutions containing 500 ppm of molybdates and nitrites provide optimum passivation for carbon steel components. Ifsystem corrosion is controlled at a low level the chance of equipment failure caused by corrosion willbe reduced. I 'I ~ y <<(v V 4l 'a St. LucieUnit2 Inetrievable Ch MonitorInside Containment == Introduction:== During maintenance in the Unit 2 containment building on December 27, 1989, an Oz monitor was dropped inside the secondary shield wall and cannot be retrieved with the reactor at power. The monitor measures approximately 1-1/2" x 4" x 7". Safety Evaluation: The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FUSAR are not increased. Section 6.2.2.2.3 ofthe FUSAR describes that debris generated inside containment as a result ofan accident willbe confined between the primary and secondary walls. Large debris generated here is prevented from reaching and possibly damaging the sump by the trash racks (3/4" gaps) located at the secondary shield wall openings. These trash racks would prevent the meter from reaching the containment sump located outside the secondary shield wall. The sump design provides additional protection by incorporating a 90 mil fine mesh filter screen to protect the ECCS suction from entrained particles. These screens are sized to eliminate particles too large to pass through the fuel assemblies, which is the most restrictive flow path in the system. Particle smaller than this would pass through all system components including reactor, pumps, heat exchangers, and spray nozzles. The possibility of an accident or malfunction of a different type than any evaluated previously in the FUSAR willnot be created. Operation with the Oz monitor inside containment does not involve any changes to the plant ofits design basis other than combining with debris generated inside containment as a result of an accident. -Debris generated inside containment as a result of an accident is analyzed in section 6.2.2.2.3 of the FUSAR The margin of safety as defined in the basis for any technical specification is not reduced. The Oz monitor is prevented from reaching and possibly damaging the sump by the trash racks located at the secondary shield wall openings. The sump design provides additional protection by incorporating a 90 mil fine mesh filter screen to protect the ECCS suction from entrained particles. These screens are sized to eliminate particles too large to pass through the reactor fuel assemblies which is the most restrictive flow path in the system. Particles smaller than this would pass through all system components including reactor, pumps, heat exchangers, and spray nozzles. The technical specification basis is not reduced during operation with the Oz monitor inside containment. ~ ~ i tw St. LucieUnit1 Repair of1"-M&49Downstream ofIHCV48-lAbythe Instillationofa Valve and Welded Cap == Introduction:== This change willinstall temporary modification in accordance with the attached safety evaluation and restore the line to the original design configuration during the next unit 1 refueling outage. Safety Evaluation: The proposed activity does not increase the probability of occurrence of an accident previously evaluated in the SAR since this modification does not affect any accident initiating components. This modification occurs outside the steam trestle, close to the turbine building, and cannot a6ect any equipment capable of initiating an accident. Any failure of this modification is encompassed by the analyzed main steam line break described in FSAR section 3, appendix 3C. The consequences of an accident previously evaluated in the SAR have not been increased by this, modification since it does not a6ect any equipment required to mitigate the effects of an accident. Failure ofnon-seismic (not Class I) portions of main steam or feedwater lines, the major portions ofwhich are run through the turbine building, cannot adversely affect the mitigation of the consequences ofthe postulated accidents and the capability to bring the unit to a cold shutdown condition since there is not safety related equipment located in the turbine building. The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased by this modification. The only safety related equipment that could be affected by a break in the main steam or feedwater lines are the three auxiliary feedwater pumps which are located under the trestles. This modification is outside the steam trestle, and is therefore unable to affect any safety related equipment. The consequences of a malfunction of equipment important to safety previously evaluated in the SAR have not been increased since this modification does not affect any equipment important to safety and is to be implemented in a section ofpiping which is not required to mitigate the consequences ofa postulated accident. The possibility of an accident ofa different type than any previously evaluated in the safety analysis report has not been created since this modification does not add or affect any equipment capable ofinitiating an accident. The proposed modification does not provide a new mode of normal or emergency plant operation. The possibility of a malfunction ofa different type than any previously evaluated in the safety analysis report has not been created since this temporary modification willnot inhibit or otherwise adversely affect the operation of any equipment important to safety. The proposed activity does not reduce the margin of safety as defined in the basis for any technical specification since this modification does not a6'ect any code safety valves, main steam isolation valves or any components as described in the basis for any plant technical specifications for the main steam system. 'I 1 I II'4 I A<< JV 4 ~ ~ St. LucieUnit1 MainSteam Isolation Valves ActuatorAirPressure Testing == Introduction:== As part of the instrument air upgrade modification, the air system operating pressure band was increased to provide the required MSIVpressure. Due to an apparent leakage ofinstrument air and pressure loss in the instrument air system, the required pressure could not be maintained. A temporaiy air system was installed to provide an additional air supply to maintain MSIVpressure above the alarm setpoint. Safety Evaluation: The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The MSIVs provide steam line isolation by closing for all analyzed plant accidents. The design of the MSIVcontrol system is such that closure can be assured within the technical specification time requirement of 6 seconds whether actuator air is above the alarm setpoint (greater than 102psig) or at lower pressures between 102 and 98 psig. In addition, for the SGTR case, su6icient air exists in the MSIVto facilitate closure up to 30 minutes into the accident. The possibility of an accident or malfunction of equipment ofa difFerent type than any previously evaluated in the FSAR is not created. The configuration of the main steam system and MSIVs willnot be altered as a result ofthe data gathering system. Operability ofthe system and the MSIVs with respect to accident mitigation willnot be altered. The additional Heise gages and tie-in tubing required to support the evolution willbe attached independently to each MSIV. Failure of the temporary equipment could result in failure of the associated MSIVonly. Single MSIVfailure . is already addressed by the FSAR. The margin of safety as defined by the technical specifications is not reduced. The requirement for MSIVclosure within 6 seconds willnot be reduced during the data gathering session. Previous testing has verified closure capability within the 6 second time restraint at pressures significantly lower than the expected 98 psig. I i, > V'1 I tl %!5& [, ~ ~ ~ ~ St. LucieUnit2 Leak Repair on 2AMainFeedwater Pump == Introduction:== The outer lip ofthe flange half on the nut fastener side ofthe MFWP 2A willbe drilled with from one to three 7/16 inch diameter holes to allow injection ofsealant into the annulus between the flange studs and their locating holes. Safety Evaluation: The probability of occurrence or the consequences of an accident or malfunction of equipment important to'safety previously evaluated in the FSAR willnot be increased because the proposed repair willbe performed on non-nuclear safety related MFWP 2A. The repair willnot degrade the integrity ofthe pump since the drilled injection holes, required for the repair, are suf6ciently small and are located in a sufficiently low stressed region of the flange. The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR is not created because the total failure ofthe main feedwater pump is within the spectrum'f accidents previously evaluated in the FSAR. The sealant to be utilized for the proposed repair willbe injected in such a manner and quantity that the sealant cannot be introduced into the secondary system. The margin of safety as defined in the bases for any technical specifications is not reduced since the Main Feedwater Pump 2A does not form part ofthe bases ofany technical specification. I ~< ~ 4 pa 4+ k ~ ~ S kal. i bi&a 1 I, St. Lucie Units 1 and 2 Waste Gas Drain System == Introduction:== The purpose ofthis evaluation is to justify the use ofa condensate removal system offthe refrigerant cooler at the suction of the Automatic Gas Analyzer sample pump. The implementation ofthis temporary modification involves the installation of a pump to ensure that a: 'ufficient pressure differential exists to provide proper drainage, a vacuum tank which*willact as-an accumulator, and a pressure switch to initiate drainage and prevent sample pump damage. Safety Evaluation: The probability of occurrence of an accident previously evaluated in the SAR has not increased since this modification does not affect any accident initiating components. The function of the gaseous waste management system, to monitor and control the release ofgaseous radioactive efHuents to the environment, willnot be affected by this temporary modification. The consequences of an accident previously evaluated in the SAR have not been increased by this.. modification since this modification does not affect any equipment required to mitigate the effects of an accident. This modification willin no way affect any radiation monitoring equipment which controls the release ofgaseous radioactive efHuents to the environment. The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR has no been increased since this modification willhelp to preclude sample pump failure due to water ingestion. This modification does not alter the function of any existing components, and thus does not increase the possibility oftheir failure. The consequences of a malfunction of equipment important to safety previously evaluated in the SAR have not been increased since this modification reduces the possibility of sample pump failure and the possible release ofradioactive gaseous eNuents into the RAB. The consequences of a failure are no worse than a complete failure ofthe sample pump or a sample line rupture. The possibility of an accident of a different type than any evaluated previously in the SAR has no been created since this modification does not add or affect any equipment capable ofinitiating an accident. The implementation of this temporary modification willnot affect any radiation monitoring devices which control the release ofgaseous radioactive efHuents to the environment. The possibility ofa malfunction of a different type than any evaluated previously in the SAR has not been created since this temporary modification willnot inhibit or otherwise adversely affect the operation ofthe GWMS automatic gas analyzer. The components of the modification are in compliance with the SAR requirements for the system elements. The proposed modification does not reduce the margin of safety as defined in the basis for any technical specification since the requirement ofoperability ofat least one analyzer system willbe maintained. a I 4 A St. LucieUnit1 Main Steam Isolation Valves Temporary Instrument AirSupplies == Introduction:== Due to leakage ofinstrument air and pressure drop in the instrument air system, the instrument air pressure at the supply to MSIVs is less than 102 psig as specified by PC/M 050-186. Ifinstrument 'air pressure goes lower, there may be an unintended closure of the affected MSIV. To avoid ""'nnecessary challenges to safety systems;a temporary air compressor willbe used to supplement-. instrument air to the MSIVthrough instrument air line 3/4-IA-153, in addition to compressed air bottles which are now being used. Safety Evaluation: The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR willnot be increased because the proposed temporary modification willbe performed on the non-nuclear safety related instrument air system and willnot change the operability, degrade closure time, or change any other function or ~ capability ofthe MSIV. The possibility for an accident or malfunction of a different type than any previously evaluated in the SAR is not created because there is no change in function or capabilities ofthe MSIVs or their control system. In addition, no new failure modes for the safety-related equipment have been created. The margin of safety as defined in the bases for nay technical specification is not reduced since there is no change in the ability ofthe MSIVs to respond as required to maintain the bases ofthe applicable technical specifications. ~ 1 St. LucieUnit1 Testing ofDiverse Scram System Modules == Introduction:== As part ofthe project to meet the requirements ofthe ATWS Rule, a Diverse Scram System (DSS) has been designed by Consolidated Controls Corporation (CCC) and willbe installed in the St.. 'ucie Unit 1 Engineered Safety Features Actuation System cabinets. CCC intends to conduct functional tests on prototype modules to ensure compatibility with the existing system, to check ~. ~ ~ interchangeability, and to check in-cabinet module operation. Safety Evaluation: The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR willnot be increased because one offour channels willbe considered inoperable at any time. This leaves the channel functions in a two-out of-three trip condition during the majority ofthe test. For a short period oftime that the function is one-out-of-three logic (i.e., during isolation module changeout), the condition is more ~ conservative than plant technical specification limitingconditions. The possibility for an accident or malfunction of a difFerent type than any previously evaluated in the SAR willnot be created because the trip logic conditions are those previously evaluated and are specifically stated in the plant technical specifications. In addition, since the cabinet design separates and electronically isolates measurement channels from each other and safety channels from each other, feedback afFecting other channels is unanticipated. Therefore the operation of another channel willnot be adversely affected. The margin'of safety as defined in the basis for any technical specification is not reduced since the T/S allow for plant operations in a two-out-of-three trip logic for a specified time period. C 1 I I I bg tf+ H ~ St. LucieUnit1 BoraflexPanel IntegrityAssessment Program == Introduction:== St. Lucie Unit 1 willbe conducting a Boraflex Integrity Assessment Program to determine the integrity ofthe Boraflex panels used in the design ofthe region 1 spent fuel storage racks. Certain aspects ofthe program have been provided to and approved by the NRC. These aspects include the'=" '" frequency oftesting and the number ofcells to be tested, as well as the scheduling ofthe test. Safety Evaluation: The SAR evaluates the consequences and probability ofa fuel handling accident and a cask drop accident. Since the test tool and calibration cell to be used weigh less than a fuel assembly, and since all liAingoperations willbe conducted in accordance with plant procedures, the probability ofan accident has not been increased. The liRing and handling ofthe test tool willnot occur over spent fuel, and the test tool contains only a very small (1.3 millicurie) neutron source. Therefore, the radiological consequences of a postulated dropping ofthe test tool are bounded by the previously analyzed accident scenarios since the test tool cannot damage any fuel and does not contain any fuel. Also, since the test tool and calibration cell weigh less than a fuel assembly, no damage will occur to the spent fuel pool structure. Based on this, the probability ofoccurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR have not been increased. The SAR evaluates the dropping ofequipment in the spent fuel pool. These accidents are bounded by the fuel handling accident and the cask drop accident. Since the test tool and calibration cell do not introduce any interaction potentials other than the dropping of equipment, and they weigh less than a fuel assembly, the possibility of an accident or malfunction ofa different type than any previously evaluated in the SAR has not been created. St. Lucie Unit 1 T/S 3.9.7 prohibits loads in excess of2000 lb. from being carried over spent fuel assemblies. This ensures that no more than one fuel assembly is damaged in a fuel handling accident. T/S 3.9.13 identifies the maximum load to be handled by the fuel cask crane to be twenty-five tons. This restriction ensures the structural integrity ofthe spent fuel pool in the event ofa cask drop accident. The two specifications ensure that the accident analysis provided in Chapter 15 ofthe FSAR remains valid. Section 5 of the Technical Specifications discusses the criticality restrictions for the spent fuel pool (Keff< 0.95). These restrictions ensure that there is adequate subcriticality margin at all times. The test tool neutron source is so small that the efl'ect on the multiplication factor Eeff is negligible. The test tool and calibration cell both weigh much less than the technical specification limitof2000 lb. and willnot be carried over spent fuel. The test tool willbe carried over spent fuel storage racks, but willnot be carried over the cells containing fuel. Therefore, since all equipment is less than 2000 lb., and within all design requirements of the liftingequipment, the margin of safety as defined in the basis for any technical specification has not been reduced. I' ~ ~ "+t 5<4 v C f ~ 4 >s' ( ~ ~ ~ ~ ~ St. LucieUnit2 Component CoolingWater HeatExchangers 2Aand RBExpandable Rubber Tube Plugs == Introduction:== During previous St. Lucie Unit 2 outagesJ the CCW heat exchanger tubes were examined, with the results showing the need to plug some ofthe tubes. One hundred thirteen ofthe tubes plugged used an expandable rubber type material. The Yuba plugging procedure detailed in Instruction Manual 2998-3511, however, identifies only metallics as the recommended plug material. Safety Evaluation: The use ofexpandable rubber tube plugs in place ofmetallic tube plugs in the CCW heat exchangers willnot increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR. Testing has shown that these plugs willcontinue to perform their desired function even when subjected to worst case (accident) pressures, temperatures, and chemistry conditions. The CCW heat exchangers willcontinue to perform their design function during all normal and accident conditions. The rubber plugs willnot become dislodged, even under the most adverse temperature and differential pressure conditions, and therefore there is no accident or malfunction of a type different than any evaluated previously. The bases for the CCW system operability provided in the technical specification is to ensure that sufficient cooling capacity is available for continued operation of equipment during normal and accident conditions. Use of the rubber plugs in lieu ofthe metallic plugs willnot reduce the cooling capacity ofthe CCW system, since the tube plugging criteria has not been exceeded. As such, the margin of safety has not been reduced. There are no technical specifications which are affected or changed by use ofthe expandable rubber plugs in the CCW heat exchangers. ~ ~4 ~ ~ St. LucieUnit1 Temporary Use ofTwo Spare Fuse Holders on RTGB 106 Iatroduction: Recently in RTGB 106, fuse block CCC points F55 and F56, the fuse clip holders, were damaged. This was caused by the insulating barrier between the positive fuse F55 and the negative fuse F56 'breaking offand the opposite polarities ofthe 125V DC circuit coming into contact with each other.~ " A replacement for the damaged fuse block is currently unavailable. By using two spare fuse holders on fuse block ZZ and four jumpers, the containment isolation valves FCV-25-2, 4, and 6 can be returned to service. This is only a temporary measure until replacement parts can be procured and installed. Safety Evaluation: The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased as demonstrated in the evaluation section. The temporary use ofthe spare fuse holders and jumpers does not adversely impact the accident analysis as previously evaluated in the safety analysis report. The possibility for an accident or malfunction of a different type than any evaluated previously in. the safety analysis report is not created. The fuse blocks CCC and ZZ are the same train power supplies. Therefore use ofthe spare fuse holders and jumpers does not introduce any new accident scenarios. The margin of safety as defined by the basis for any technical specification has not been reduced ~ ~ since the basis of any technical specification is not affected. The requirements of the evaluation ensure that the temporary use ofthe spare fuse holders and jumpers have no adverse interaction with other equipment. The operability requirement for this component imposed by the Technical Specifications is met. l 1 ~a ~ 4 4 ~ '~ ~ ~ ~ ~ St. LucieUnit2 Temporary InstillationofCable forRCP2A2Upper OilHeservoirLevelIndication == Introduction:== Shortly after the last Unit 2 startup, level indication for the 2A2 RCP upper oil reservoir was lost. It was determined that Cable 20112E, from the sensing probe to transmitter LT-1166, had failed. A temporary replacement cable was installed from the sensing probe to=the level transmitter. Due to.. the'insufficient length ofthe replacement cable, it could not be routed through the existing conduit. ~ Itwas therefore routed external to the conduit from the 2A2 RCP to LT-1166 located on the 43'evel of the RCB. Safety Evaluation: The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. The temporary installation of the cable can have no impact on any equipment important to safety. Failure ofthe cable cannot result in any adverse eFects on equipment important to safety. The temporary installation of the cable cannot impact any analysis previously evaluated in the safety analysis report. The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created. Failure of the cable cannot result in any adverse effects on equipment important to safety. Routing ofthe cable outside ofthe conduit does not introduce any new accident scenarios. The margin of safety as defined by the basis for any technical specification is not reduced by the temporary installation of the cable. There are no technical specifications associated with this level instrumentation. I J ~A St. LucieUnit2 Submergence ofValveV4S41, Reactor Dram Taak Pump Suction Isolation Valve == Introduction:== St. Lucie Unit 2 Non-Conformance Report 2-253, Description of Condition, indicates that valve V-6341, RDT Pump Suction Isolation Valve, is located at elevation 9 ftofthe RCB. This is below the post-LOCA flood level. The limitswitch and solenoid valve associated with this valve are mounted on it. However, all the solenoids were designed to be located above the flood level ~ elevation. Safety Evaluation: The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR are not increased since the safety function of the valve, closing on CIS, is not affected. The functionality ofthe solenoid under submerged conditions is not affected, and the limitswitches under submerged conditions willalso perform the required function by providing valve position indication for sufficient duration for operators to assess containment isolation. Leaving the solenoid and the limitswitches in the present location willnot increase the possibility of an accident or malfunction of a different type of any evaluated previously in the FSAR since valve V-6341 willclose on CIS under submergence conditions. Thus, no new modes offailure are being introduced. The margin of safety as defined in the basis for any technical specification willnot be reduc'ed since the operation, of the reactor drain system is not affected and containment integrity is maintained. ~ ~ SECTION 2 Steam Generator In-Service Inspection St. Lucie Unit 2 ~! A

• SECTION 2 STEAM GENERATOR IN-SERVICE INSPECTION Unit 2 Technical Specification 4.4.5.5 The inspection of the St.

Lucie Unit 2 Steam Generators was conducted by Florida Power & Light, Material Codes and Inspection ,Group, supplemented by FPL certified Eddy Current personnel and complemented by contractor personnel. The Eddy Current In-Service Inspection of Steam Generators was performed during the 1989 Unit. 2 refueling outage. One hundred percent of each steam generator was inspected. L-89-253 da have been i The information concerning Eddy current Testing and Tube Plugging was submitted as part of the Owner's Data Report for Inservice Inspection (Form NIS-1). This was submitted in accordance with the provisions of the ASME Code,Section XI, (IWA-6230) by FPL letter ted July 17, 1989. Selected pages of the above report ncluded in this section. , ~ jl t; Wt s Ia I A t PJ~f I Pg -E Cl APL FORM NIS-BB OWNERS'ATA REPORT FOR EDDY CURRENT EXAMINATIONRESULTS As required by the provisions of the ASME CODE RULES EDDY CURRENT EXAMINATIONRESULTS PLANT: St. Lucie Unit 5 2 EXAMINATION DATES: February 9, 1989 Thru March 1, 1989 STEAM GENERATOR TOTAL TUBES INSPECTED 8203 204' 39% 40~o 1004 52 TOTAL INDICATIONS TUBES TUBES PLUGGED AS PLUGGED PREVENTIVE THIS MAINTENANCE OUTAGE 13 TOTAL PLUGGED TUBES IN S/G 221 8256 41 162 LOCATION OF INDICATIONS r ( 204 1004 ) EAM ENERATOR U BENDS DHB to DCB 54 EGGCRATES 1 TO 7 H/L C/L 1 2 PARTIAL SUPPORTS 8 AND 9 H/L C/L TOP OF TUBE SHEET TO 5 1 EGGCRATE H/L C/L Remarks: 40 1 1 ASR = Adjacent stay rod tube TRS = Tube restriction TBP = To be plugged CERTIFICATION OF RECORD We certify that the statements in this report are correct and the tubes inspected were tested in accordance with the requirements of Section XI of the ASME Code. i> <Yl' DATE ~ Fl rid Power & Light Co. BY FLORIDA POWER & LIGHT COMPANY 700 UNIVERSE BLVD. JUNO BEACH, FLORIDA 33408 ST. LUCIE NUCLEAR PLANT P.OUTBOX 128 FT.PIERCE, FLORIDA 33454 UNIT 2 STEAM GENERATOR 2A & 2B 1989 EDDY CURRENT RESULTS CUMULATIVE REPORT 1 I I 0 ~ ~ COMPONENT

S/G g A CUMMULATIVE DISTRIBUTION

SUMMARY

ST.

LUCIE 2 02/89 Page 1 of 1 Date 04/17/89 Tame 1:58 PM Examination Dates

02/09/89 thru 03/01/89 Total Number of Tubes Inspected 8203 Total Indications Between 20% and 39%

Greater than or equal to 40%

Total Tubes Plugged as Preventive Maint Total Tubes Plugged Location Of Indications 20% to 100%

52 6

7 13 Hot Leg TSH.5 to 01H -2.1 01H -2.0 to 07H +2.0 07H +2.1 to DHB -3.1 DHB -3.0 to DCB -3.0 0

1 0

54 Cold Leg TSC -.5 to 01C -2.1 01C -2.0 to 07C +2.0 07C +2.1 to DCB -3.1 1

2 0

r 1

E ~

COMPONENT

STG IIA PLUGGABLE TUBES LIST St. Lucie 2 02/89 Page 1 of 1

Date :

04/14/89 Time :

9:53 AM

)REQ/

/

TESTED

/

ONEIEXTILEGI EXTENT I

REEL I

I I

I ROW I LINE I2

+

+

+

I126I 48I i136i 64' 94'6' 39(

73/

I 42I 76I I 135 I 81 I (50/

86/

I I

I I 47I 87I I

I I

I I

I I 941 90I J 40'6/

i 50l 96' 115 I 129I I 56I 150I

+

+

+

+

14 ITEHI C ITEH PCIAC121 14iTEH/

C iTEH iAC124 16 ITEHI C I TEH PCGY IAC116 CITEHI C ITEN PCGYIAC006 CITEHI C

ITEM PSGYIAC006 15ITEHI C

ITEN GYIAC064 6(TEH/

C JTEH GY/AC069

)TEH/

C

/TEH PSGN(AC069 6ITEHI C

ITEN GNIAC069 ITEHI C

ITEN GMIAC069

/TEH/

C lTEH PSGN/AC069 17iTEH/

C (TEH PSGN/AC056 6iTEH]

C JTEH PSGNJAC069 6ITEHI C

ITEN PSGNIAC069 10ITEHI C

ITEN PCIAC048 8ITEHI C

ITEN PSGYIAC034 I

I I

I CHI PROBE I

L

+

+

+

I 1IA-580-SF/RM I

OHT I

1IA 580 SF/RM I DHT IM 1IA-580-SF/RM I VS4 I

1IA-580-SF/RM I

OHB IN 1IA-580 SF/RN I VS3 IN 1IA.580-SF/RN I VS4 I

1IA-580-SF/RM I DCT I

1IA 580 SF/RN I DHB I

1IA.580.SF/RN I

OHT I

1IA-580-SF/RM I OCT I

1IA 58 SF/RM I

DHB IM 1IA-580-SF/RM I VS3 IM 1IA-580.SF/RN I VS3 IN 1IA-580-SF/RN I VS3 I

1IA-580-SF/RM I

TSH IN 1IA.580.SF/RN I VS3 OCATION 0.0 0.0 0.7 0.0 0.5

-0.8 0.0 0.0 0.0 0.0 0.0 1.0

-0.7

-0.7 14.5 1.0

+

+

+

+

+

+

+

+

Number of Pluggable Tubes:

13

+

+

+

t

+

+

+

+

+

+

+

+

+

+

+

I I

I I

I

/ X /Volts/Oeg/ Oataset

+

+

+

+

+

/ 45/

6.4i110/DEGRADED IPTP I 4.7I 130 I IMPERFECT I

J COAL 5.1I 95IIMPERFECT I

IPTPI 3.6I 120I IMPERFECT I

IPTPI 1 ~ 11108IINPERFECT

/

IPTPI 2.5I 1021 IMPERFECT I

JPTPi 2.4i 129'MPERFECT IPTPI 0.5I129IIMPERFECT I

IPTPI O.CI1351IKPERFECT I

IPTPI 2.8I129IINPERFECT I

iPTP/

1.3/126/IMPERFECT

/

I COI 3.2I 97IINPERFECT

[

I 48I 2.5I 90IOEGRADEO I

I CBI 1.3I 90IOEGRAOEO I

ILPII O.DI OI1-S I

I 43I 1.8I 87IBALANCE I

+

+

+

+

Number of Indications 16 Selection Criteria :

PerCent T.N.

~..:

Cg Includes AOI,NQI,OSI,OTI,DRI,LPI,TBP,TRS,PTP Codes

PONENT :

STG IIA DESCRIPTION:

20K TO 100K CUMMULATIVE EXAHIHATION REPORT St. Lucie 2 OUTAGE : 02/89 Page 1 of 2 Date :

4/ 3/89 Time : 12:33 PH I

(Row(Lin(

I"-I---I

( 86( 22(

I 921 261 I 931 271 I 981 281 I 981 301 11051 331

( 55( 37(

(126( 48(

I I

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I Extent Leg(Req(Tst

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ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

(TEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

(TEH(TEH C

ITEHITEH C

ITEHITEH C

(TEH(TEH C

ITEHITEH C

ITEH(TEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

I'TEHITEH C

(TEH(TEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

(TEH(TEH C

ITEHITEH C

(TEH(TEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

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--I---I---

PSIAC105 PSIAC106 PSGN(AC106 PSIAC107 IAC107 IAC108 GNIAC103 PC(AC121 IAC121 SSGNIAC002 PSIAC112 (AC114 IAC123 IAC123 GNIACOOS SSIAC115 IAC124 IAC124 PCGYIAC116 GHIAC116 PCGYIAC006 GH(AC006 PSGNIAC006 PSGNIAC062 IAC117 PSGYIAC006 GNIAC014 (AC006 SSGNIAC006 SSGNIAC006 SSGNIAC006 GY(AC064 GNIAC016 PSGH(AC007 GNIAC069 PCGN(AC075 PSGNIAC069 GYIAC069 PCGN(AC055 GNIAC069 PSGNIAC069 GHIAC069 PSIAC069

---

I-02/89 I

n (Volts(Oeg(Ch

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-I -------- --I----

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.SI 98IH 11 351 2.8( 97(H 1( 36(

.91112IH 11 201

.611251 11 281 1.011321 11 211 2.0(102(H 1( 28(

6.411101 11 451 1.51113(H 11 211

.511201 11 331 2.011271 11 271

.81106 IH 11 271 1.11107IH 11 241

.911301 11 231

.8(110(H 1( 27(

1.81101IH 11 321 2.811311 11 241 C.7(130(

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---I'-----------I-----I---I--I --I

I yl

'L

CUHMULATIVE EXAMINATION REPORT St. Lucia 2 OUTAGE : 02/89 OHPONENT :

S'IG ¹A

~

~

DESCRIPTION : 20X TO 100K Page 2 of 2 Date :

4/ 3/89 Time : 12:33 PH

+

+

I

( Extent I

02/89 N/A I

(Row(LIn(Leg(Req(Tst/Note(

Reel

(

Probe

(

Location (Volts(Deg(Ch

( X (Diff(

Location (Volts(Oeg(Ch

( X

(

I 'i'.-- -"--.---

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--I-"-- --I-

-I I-

"I I---I-- I---I----I

.-----.I-----I" I---I---I I 94( 90(

C ITEHITEH PSGHIAC056 IA 580-SF/RH IVS3 1.0 I

3.2( 97(H 1( 40(

I I

I I

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I 43( 91(

C ITEHITEH PSIAC069 IA-580-SF/RN (DHB

.0 I

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1( 20(

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(109( 91(

C (TEH(TEH (AC056 (A-580-SF/RH (TSC 5.0

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C (TEH(TEH (AC069 (A-580-SF/RH (DCT

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C (TEH(TEH (AC069 (A-SBO.SF/RN (DHT

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C ITEHITEH PSGNIAC069 IA-580-SF/RN IVS3

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C (TEH(TEH GN(AC069 (A 580.SF/RH (VS3

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C ITEHITEH GNIAC045 IA-580.SF/RH IVS1

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C

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C (TEH(TEH PSGN(AC040 (A-580-SF/RM (VS2

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"-I-I --"------I--------"----(----(-- I---( -I----I---------- --

I ----(---(--(-- (

r of INDICATIONS Selected from Current Outage:

58 r of TUBES Selected from Current Outage 52

4 4

~

~

~

~

0 COMPONENT

S/G 8

B CUMMULATIVE DISTRIBUTION

SUMMARY

ST.

LUCIE 2 02/89'age 1 of 1 Date 04/17/89 Time:

1: 58 PM Examination Dates

02/09/89 thru 03/01/89 Total Number of Tubes Inspected.....:

8256 Total Indications Between 204 and 39%

Greater than or equal to 404 Total Tubes Plugged as Preventive Maint Total Tubes Plugged Location Of Indications 204 to 1004 41 4

3 7

0 Hot Leg TSH.5 to 01H -2.1 01H -2.0 to 07H +2.0 07H +2.1 to DHB -3.1 DHB -3.0 to DCB -3.0 1

0 2

40 Cold Leg TSC

.5 to 01C -2.1 01C -2.0 to 07C +2.0 07C +2.1 to DCB -3.1 1

1 0

~

~

1 ~'

~ y

~

e

~

COMPOHEHT

STG ¹8 PLUGGABLE TUBES LIST St. Lucia 2 02/89 Page 1 of 1

Date 04/14/89 Time :

9:53 AH

+

+

+

+

+

+

+

+

+

+

+

+

+

+

+

I I

I IREQI I

TEBTEO I

IROMILINEIZONEIEXTILEGIEXTENT I

REEL

+

+

+

+

+

+

+

( 47(

79(

4(TEHI C

(TEH PCGY(BC008 I 46I BOI 4ITEHI C ITEH PSGYIBC008 I 56I BOI 3ITEHI C

ITEH PSGNIBC016 I 46(

82(

4(TEKI C

ITEH PCGYIBC008

( 48(

86(

6(TEN(

C ITEH PSGYIBC070

( 96(

94(

17(TEHI C

ITEH SCGYIBC057 I

I I

ITEHI C

ITEH GYIBC057 I 311 101I 6ITEHI C

ITEH SSGMIBC071

+

+

+

+

+

+

+

I I

I CHI PROBE I

I I

I I

I LOCATION I X (Volts(Oeg( Oataset I

1IA-580-SF/RH I

1(A-580-SF/RH IM 1IA-580-SF/RH 1(A-580-SF/RH I

1IA-580-SF/RH I

1IA-580 SF/RH IH 1IA-580-SF/RH I

1IA-580-SF/RH I

OHT 0.0 I

OHB 0.0

( VS3 0.9 I

OHB 0.0 I OCT 0.0 I VS3 0.5 I VS2.0.9 I OCr 0.0

+

'PTP(

3.2(128(IHPERFECT

(

(PTP(

2.6(131( IMPERFECT I

( 50(

1.5(

94(IHPERFECT I 41I 3 'I117IIMPERFECT I

I 46I 1.5I114I1.s I

IPTPI'.8('134(IMPERFECT IPTPI 2.41112IIMPERFECT I

I 46I 0.9I113IIMPERFECT I

+

+

+

+

+

+

+

+

+

+

+

+

Number of Pluggable Tubes:

Nether of Indications Selection Criteria :

PerCent T.M....:

40 Includes AOI,NQI,OSI,DTI,ORI,LPI,TBP,TRS,PTP Codes

\\ ~

~

~

PONENT :

STG ¹a DESCRlPTION:

20X TO 100X CUHMULATlVE EXAHlNATlON REPORT St. Lucie 2 OUTAGE : 02/89 Page 1 of 2 Date 4/ 3/89 Time : 12:34 PH I

(Row(Lin(L I- -I --I-( 96( 24(

I 97( 41(

I 68( 50(

I 571 511 11311 551 I 391 691 I 521 761 I 431 771 I 501 781

( 47( 79(

I 491 791 I 51(

79(

I 57( 79(

I 461 801 I 561 801 I 471 811 i 46( 82(

91 831 TI 851 I 53( 85(

I 48( 86(

I I

I

( 52( 88(

I 541 901 I 871 931 I 961 941 I

I I

I 941 961 I 961 961 I

I I

( 35( 97(

I 89( 97(

I 97( 97(

I 341 981

( 35( 99(

I 3111011 I 8911031 I 5211101 I 9611141 (129(115

(

I 94(1221 55(125(

11261 I- -I-( Extent eg(Req(Tst

- I---I---

C ITEHITEH C

ITEH(TEH C

ITEH(TEH C

ITEHITEH C

ITEHITEH C

(TEH(TEH C

(TEHITEH C

ITEHITEH C

ITEHITEK C

ITEHITEH C

(TEH(TEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

(TEH(TEH C

ITEH(TEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

(TEH(TEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEH(TEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

ITEHITEH C

(TEHITEH

--I"-I-I

/Note(

Reel

---

I-IBC107 Pc IBC110 GN(BC009 PSGNIBC010 Ps IBC124 GNIBCOOT PCGHIBC015 PSGNIBCOOS PCIBC008 PCGYIBC008 PCIBCOOS PSGNIBC016 PSGNIBC016 PCGYIBC008 PSGN(BC016 PCIBCOOS pcovlac008 I8C121 (BC070 PSGMIBC076 sslacO7O PSGYIBC070 PSGNIBC076 PSIBC077 PSGNIBC057 GvlecO57 SCGYIBC057 PsGN(aco57 SSGNIBC057 PSGHIBC057 PSGNIBC070 PSGNIBC058 IBC058 PSGNIBCOTO IBC070 SSGMIBC071 GNIBC059 GN(ec081 GNIBC036 PSGNIBC047 PSGNIBC038 PSGN(BC029 SSGNIBC039

---

I-I Probe

(

Locatio

-I I

IA-580-SF/RH IOSH IA 580-SF/RH IVS4 IA-580-SF/RH IVS3 IA-580 SF/RH IVS3 IA-580-SF/RH IOCB IA.580-SF/RH IVS3 IA-580-SF/RM IVS3 (A 580 SF/RH IDCB IA.Sso.sF/RM IDGT (A.SSO-SF/RH IOHT IA 580-SF/RH IDCT (A.SBO-SF/RH (OCB IA-580-SFIRM IDCT IA-SBO.SF/RH IOHB IA 580 SF/RH IVS3 IA 580 SF/RH IOHT IA 580-SF/RH IOHB IA-580 SF/RH IVS3 IA-580 SF/RH loca IA 580.SF/RH IOCB IA 580.SF/RH IOHB (A-seo-sF/RH (ocT (A-580-SF/RM (VS3 IA-580:SF/RH IOCB IA-580-SF/RH IVS3 IA-580-SF/RH IVS2 (A.SSO-SF/RH IVS3 IA-580-SF/RH IVS2 IA-580.SF/RH IVS2 IA-580-SF/RH IVS3 IA-580-SF/RH IOHB IA-580-SF/RM IVS4 IA-580-SF/RM ITSH IA-SSO.SF/RM (VS3 IA-580-SF/RH IOCB IA-580 SF/RH IDCT IA-580-SF/RH IVS2 IA-580.SF/RH IVS3 IA.580-SF/RM IVS2 IA-580-SF/RH IOBH (A-580-SF/RH (VS2 IA-SBO.SF/RH IVS3 IA-580.SF/RH IVS2 I

-"" -I-

.6

-.9

-1.0

.0

.9

.0

.0

.0

.0

.9

.0

.0

-.7

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.0

.0

.0

.7

.0

-.9

-.9

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-1.0

-1.0

.8

.0 21.6

-.8

~ 0

.0

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-.8

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-.7

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- 9 02/89 I

I n

(Volts(Deg(Ch

( X (Diff(

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.91 97IH 11

.711401 11

.51117IH 11

.811051M 11

.7(139(

1(

1.1(109IH 11

.81107IH 11 1.z(1391 11

.41 1361 11 3.211281 11 1.511311 11

. 711391 11 1.111331 11 2.611311 11 1.51 94IH 11 1.711401 11 3.111171 11 1.71110IH 11 1.911391 11 1.011311 11

.811371 11 1.511141 11 1.1(110(H 1(

.411371 11

.51112 IH 11 381 211 25(

37(

23(

30(

381 201 23(

311 ze(

23(

291 28(

501 20(

41(

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241 241 461 20(

211 241 24(

ze(

2.4(112(H 1(

F811341 11 1.01108IH 11

.91106 IH 11 28(

30(

(,1 71113(H 11 231 23(

34(

zi I 34(

.411381 11

.61 1031M 11

.611401 11

.91104 IH 11

~ 611371 11

~ 911131 11

~ 91110IH 11

.6(109(H 1(

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271 251 261 1.51112 IH 11 1.21124 IH 11 241

.51114IH 11 1.31113(H 11 341 26("I-"-I -'-

~ 61114(H 11

".I" I---I-

~

r I

A t

~l4 '

CUHMULATIVE EXAHIHATION REPORT St. Lucie 2 OUI'AGE : 02/89 POHEHT STG ¹B OESCRIPTION:

20K TO 100'X Page 2 of 2 gate 4/ 3/89 Time : 12:34 PH

+

J Extent

/Row/Lin/Leg/Req(Tst/Note]

Reel

(

Probe I-"I---I---I-I I

"---I ---""----

I108I136I C

ITEHITEH IBC041 IA-580-SF/RH I 12I166I C

ITEHITEH PSGHIBC028 (A.580-SF/RH I

I I

I I ---"-I------I---"-----

I 02/89 I

I

'/A I

Location

/Volts/Deg/Ch

/ X /Diff/

Location

/Volts(0eg/Ch

/

X

/

I---------.---I-----I---I---I--I"--I------""--- I-----I- - I-- I--- I I01C 31.0 I

.9I126IH I I 20I I

I I

I I

I (TSC 27.0

/

.8/135/

1/ 27/

J

(

(

I---------.I ---I--I--I"-I-"I-------------I--"I --I---I---I Naker of IHOICATIOHS Selected from Current Outage:

45 Hmher of TUBES Selected from Current Outage:

42

1

'I

FLORIDA POWER

& LIGHT COMPANY 700 UNIVERSE BLVD.

JUNO BEACH, FLORIDA 33408 ST.

LUCIE NUCLEAR PLANT P.O.BOX 128 FT.PIERCE, FLORIDA 33454 UNIT 2 STEAM GENERATOR 2A

& 2B LIST OF TOTAL TUBES PLUGGED 1989 CUMULATIVE REPORT

CUHKULATIVE EXAHIHATION REPORT St. Lucia 2 OUTAGE : 02/89 PONENT STG ¹A

VERIFIED EXISTING PLUGGED TUBES AND 02/89 PLUGGABLES DESCRIPTION Page 1 of 6 Date :

3/ 1/89 Tiae : 9:26 AH I

Probe

--I-I I Extent LinlLeglReqlTst/N I

I-- I----

181 C

ITECITEC 181 C

ITECITEC 261 C

ITECITEC 301 C

ITECITEC 361 C

ITECITEC 381 C

ITECITEC 381 C

ITECITEC 441 C

ITECITEC 471 C

ITECITEC 481 C

ITEKITEH I

C ITEHITEH 501 C

ITECITEC sol c ITEGITEc 501 C

ITECITEC 501 C

ITECITEC 511 C

ITECITEC 511 C

ITECITEC 521 C

ITECITEC 521 C

ITECITEC 531 C

ITECITEC 561 C

ITECITEC 561 C

ITECITEC 591 C

ITECITEC 591 C

ITECITEC 601 C

ITECITEC 601 C

ITECITEC 611 C

ITECITEC 621 C

ITECITEC 641 C

ITEHITEH 651 C

ITECITEC 651 C

ITECITEC 661 C

ITECITEC 661 C

ITECITEC 661 C

ITECITEC 661 C

ITEHITEH P 671 C

ITECITEC 671 C

ITECITEC 671 C

ITECITEC 681 C

ITECITEC 681 C

ITECITEC 691 C

ITECITEC 691 C

ITECITEC 691 C.ITECITEC

-- I---I---I"---

02/89 I

I I

IRoMI I---I I 601 I 661 I 741 I 681 I 941 I 361 I 441 I 441 I

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CUHHULATIVE EXAHI NATION REPORT St. Lucia 2 OUTAGE : 02/89 POMENT :

STG gA t

DESCRIPTIOM: VERIFIEO EXISTING PLUGQEO TUBES ANO 02/89 PLUGGABLES Page :

5 of 6 Oate :

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CNOR/LATIVE EXAHIHATION REPORT St. Lucia 2 OUTAGE : 02/89 POHENT :

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CUNNULATIVE EXANI NATION REPORT St. Lucie 2 OUTAGE : 02/89 PONENT :

STG BB DESCRIPTION : VERIFIED EXISTING PLUGGED TUBES AND 02/89 PLUGGABLES Page :

Date :

Time:

3of4 3/ 1/89 10:09 AN Hotel Reel I

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I CUNNULATIVE EXAMINATION REPORT St. Lucie 2 OUTAGE : 02/89 DESCRIPTION: VERIFIEO EXISTING PLUGGED TUBES AND 02/89 PLUGGABLES Page I

Date:

Time:

4of4 3/ 1/89 10:09 AN I

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I Nunber of INDICATIONS Selected from Current Outage:

163 Kueber of TUBES Selected from Current Outage:

162 VERIFI BY CE HEHRY LABIENIEC DATE

SECTION 3 St. Lucie Mangrove Study Technical Specification 4.7.6.1.2

In accordance with Technical Specification 4.7.6.1.2.,

FPL conducted the St. Lucie Mangrove Photographic Survey in June 1989.

The survey revealed that the deterioration of the mangroves east of the power plant between the intake and discharge canals remains in excess of 10%,

compared to the 1975 baseline condition.

An engineering evaluation, to determine the effects of deteriorating mangroves on the FSAR wave runup

analysis, was completed in July 1987.

This evaluation concluded that the mangroves are not required to maintain the design basis of the St.

Lucie site to protect safety-related structures and equipment from Probable Maximum Hurricane surge and erosion damage.

Therefore, the deterioration of the mangroves does not create a condition of safety significance.

Activities have been in progress to improve the vitality of the mangroves, and the coverage of the area in questions has improved by almost 6%

when compared to the 1988 condition.

In 1990 fertilizing, water pumping regimes and additional planting are expected to further improve the condition of the area.

s.

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SECTION 4 Personnel Exposure Report Technical Specification 6.9.1.5