ML17228B043

1994 Annual Operating Rept.

ML17228B043
Person / Time
Site:	Saint Lucie
Issue date:	12/31/1994
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17228B042	List: ML17228B041 ML17228B043
References
NUDOCS 9503030282
Download: ML17228B043 (369)

Text

OPERATING REPORT 9503030282 950224 PDR ADOCK 05000335 R PDR

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Florida Power 5 Light Company, P.O. Box 128, Fort Pierce, FL 34954.0120 February 24, 1995 L-95-46 10 CFR 50.36 APL U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Re: St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 9 ua 0 e a '~eitt:

Pursuant to Technical Specification 6.9.1.4, this letter transmits the 1994 Annual Operating Report for St. Lucie Units 1 and 2. The report covers activities on a calendar year basis for both units.

Section 1 Man-REM Report required by Technical Specification 6.9.1.5.

Section 2 Mangrove Photographic Survey Results required by Technical Specification 4.7.6.1.2.

Section 3 Primary chemistry information required by Technical Specification 6.9.1.5.

Section 4 10 CFR 50.59 Report of Facility Reviev Group Approved Safety Evaluations. This attachment includes a brief description and safety evaluation summary for changes to activities described in the Updated Final Safety Analysis Report (UFSAR) that are stand alone Engineering Safety Evaluations; that support changes to procedures; or that support jumpers and lifted leads for the calendar year 1994.

Should there be any questions on this information, please contact us ~

Very truly yours, D. A. ger Vice r sident St. Lucie Plant DAS/GRM cc: Stevart D. Ebneter, Regional Administrator, Region Senior Resident Inspector, USNRC, St. Lucie Plant II, USNRC Enclosure an FPL Group company

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 994 e ort, SECTION 1 TECHNICAL SPECIFICATION 6.9.1.5 ST. LUCIE UNITS 1 AI'6) 2 1994 1VRM-REM REPORT

OR I RAD I AT ION EXPOSURE DAPOWE MONITORING R8L I GHT S CONTROL SYSTEM - REMACS 1994 MAN-REM RT STANDARD FORMAT FOR REPORT ING NUMBER OF PERSONNEL S MAN-REM FOR WORI JOB FUNCTIONS NUMBER OF PERSONNEL 100 MREM TOTAL MAN / REM WORK AND JOB FUNCTION STATION UT IL ITY CONTRACT ST AT ION UTILITY CONTRACT REACTOR OPERATIONS 8 SURV ENGINEER ING 0 0 1 000.000 OOO. I O OPS 153 HEALTH PHYSI CS 20 0 16 OO5.3ZO 000OIQ 005.135 MAINTENANCE 0 0 1 000.301 OOO.OIo'00.533 001.048 ar ERAT IONS 31 1 1 007.966 002.367 SUPERVISOR 0 0 0 ooo.nlz ooo.aoo ooo.ooo ROUTINE MAINTENANCE ENGI NEER ING 0 2 oeo.ooo 001. 529 001.633 HEALTH PHYSICS 33 0 100 011 .54Z 000.007 037 . 037 MA IN TE NANCE 210 ZZZ 066 ~ 1 ZT 001.781 064 .9Z5 OPERAT IONS 13 8 49 005.Q76 003. 471 014.873 SUPERV I SOR 0 0 o OOO.O15 PQO. POQ ooo.ooo INSERV ICE INSPECTION ENGINEERING 0 0 000.000 000.586 002.779 HEALTH PHYSICS 0 0 n OOO.O51 ooo.ooo oov.ooo MAINTENANCE 5 2 002.380 000.480 008 ~ I 25 OPERAT IONS 2 2 000.'911 000 '40 004 .036 SUPERV ISOR 0 o 0 000.000 000.000 ooo.ooo SPECIAL MAINTENANCE ENGINEER ING 0 I 2 ooo.ooo 000 ~ 41 2 000.550 HEALTH PHYSI CS 3 0 I OOI.OZO 000. 000 000.217 MAINTENANCE 59 154, 018.452 000 ~ 21 1 062 ~ 619 OPERATIONS 0 5 86 000 273 001. 307 051. 221 SUPE RV ISOR 0 0 o ooo.ooo QQQ. QOP ooo.ooo WASTE PROCESSING ENGINEERING 0 0 0 000.000 ooo.ooo noo.ooo HEALTH PHYSICS 9 0 37 003.144 ooo.ooo 010 ~ 746 MA INTENANCE n 0 7 000.494 OQO.OOO 001 .931 OP ERAT IONS 0 0 13 000.106 000 000 006.901 SUPERVISOR 0 0 0 000.000 000.000 000.000 REFUELING ENGI NEER ING 0 1 2 000.000 000.769 000 2+~1 HEALTH PHYSICS 6 0 2 001 603 000.002 001.596 MA INTENANCE 128 2 14 050.147 000.737 005.736 OP ERAT 1ONS 16 1 11 005.358 000.612 003.739 SUPERV I SOR 0 0 0 000.005 000.000 000.000 TOTALS ENGI NFER ING 0 8 17 non.noo 003. 416 005.366 HEALTH PHYSICS 57 0 150 022.68n Ql)n ~ Q19 054.731 MAINTENANCE 269 8 376 137 '01 nn3.".".5 144.384 OP ERAT IONS 67 13 135 019.690 006. 363 083.137 SUFERV I SOR 0 0 0 nnn.o32 non.noo 0017 ~ nl)l)

GRAND TOTALS 393 678 180.303 013. 0 3 287 6 18

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St. Lucie Units

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1 and 2 Docket Nos. 50-335 and 50-389 1994 ua

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Re ort SEXTION 2 TECHMCAL SPECIFICATION 4.7.6.1.2 ST. LUCIE UNITS 1 AND 2 M.MGROVE PHOTOGlVG'HIC SURVEY RESULTS Based on the evaluation of the false color infrared photograph taken on July 8, 1994, the condition of the mangrove trees situated between the intake and discharge canals (Impoundment 8E) continue to show additional plant growth throughout most of the 50 acre impoundment. There is approximately a nine (9) percent" increase from last year's (1993) aerial evaluation. This increases the mangrove coverage to sixty-one (61) percent. The coverage in impoundment 8E is still below the 1975 baseline condition, however, with the continuation of the current management efforts, the mangroves should continue to improve in their health and vitality as w'ell as provide additional plant growth to the area.

St. Lucie Units

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1 and 2 Docket Nos. 50-335 ' and 50-389

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199 e o SECTION 3 TECEPGCAL SPECIFICATION 6.9.1.5 AZG) 3.4.8 CHEMISTRY RESULTS Zn accordance vith-.Technical Specification 6.9.1.5, the primary coolant specific activity did. not exceed the limits of Technical Specification 3.4.8.

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St. Lucie Units

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1 and 2 Docket Nos. 50-335 and 50-389

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1994 An u e a

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Re ort SEC'I'ION 4 10 CFR 50.59 FRG APPROVED SAFETY EVALUATIONS FOR CALENDAR YEAR 1994

Safety Evaluations reportable pursuant to 10 CFR 50.59 for St. Lucie Units 1 S 2 Sumber Revision Title JPN-PSL-SEMJ-90-039 Installation of Blind Flange on Outlet of Purge Exhaust Valve FCV-25-6 ZPN-PSL-SEFJ-92-009 St. Lucie Plant Control Element Assembly Operational Life Determination ZPN-PSL-SECJ-93-002 1-2 Specif ication SPEC-C-013; Installation Guidelines for Miscellaneous Non-System Related Items on Existing Structures JPN-PSL-SEMS-93-010 0-1 Installation of Hydro Plugs into the ICI Penetrations for Locations J7 and G18 JPN-PSL-SECJ-93-011 Safety Evaluation for Specification SPEC-C-005, Component Mounting and Supports ZPN-PSL-SECZ-93-012 Safety Evaluation for Specification SPEC-C>>019, Tubing and Tubing Supports JPN-PSL-SECJ-93-014 Safety Evaluation for Specification SPEC-C-017 Small Bore Piping Supports ZPN-PSL-SEFZ-93-014 0-2 Safety Evaluation of Spent Fuel Pool (SFP) Coupon Surveillance Program ZPN-PSL-SEFZ-93-024 Loss of Feedwater Transient with Corrected Steam Generator Inventory Error ZPN-PSL-SENP-93-035 Evaluation of Inventory Loss for the Refueling Water Tank ZPN-PSL-SEMP-94-001 Temporary Installation of Strain Measurement'Devices on the Pressurizer Relief Valve Discharge Piping JPN-PSL-SEFJ-94-002 Evaluation of'educed Shutdown Cooling Flow Rate for St. Lucie Unit 2 Refueling Outage

Safoty Evaluations reportable pursuant to 10 CPR 50.59 for St. Lucie Units 1 & 2 Revision Ti.tie JPN-PSL-SEIS-94-005 Operation of the Wide Range Containment Level Monitoring Channels L-07-13A & L-07-13B with Inoperable Sensors ZPN-PSL-SENP-94-005 Shutdown Cooling Suction Valve Interlock Design ZPN-PSL-SEMS-94-008 Gasket Leak Repair for Shutdown Cooling= Return Isolation Valve V3480 JPN-PSL-SENS-94-010 Evaluation for Alternate ECCS Valve Alignment to Repair Line 3/4-SI-121 JPN-PSL-SEMS-94-011 Pressurizer Spray Bypass Valve V1454 Needle Tip Failure ZPN-PSL-SEMS-94-013 2-3 Freeze Seal Application for V3480 on the 1A Shutdown Cooling Return Line JPN-PSL-SENS-94-. 015 Safety Evaluation for Service Water System Modifications JPN-PSL-SENP-94-017 Disabling the Steam Dump and Bypass Control System Quick Open Feature for Load Reduction JPN-PSL-SENS-94-018 Safety Evaluation for Hypochlorite System Modifications JPN-PSL-SENP-94-019 0-1 Alternative Valve Position for Spray Header Isolation Valve I-FCV-07-1B JPN-PSL-SEFJ-94-021 '0 RTD Response Time Limit Increase From 8.0 Seconds to 14.0 Seconds JPN-PSL-SENP-94-021 0-1 Removing the Automatic Control Function for I-TCV-14-4A JPN-PSL-SENS-94-025 Safety Evaluation for Fuel Handling'Equipment UFSAR Discrepancies JPN-PSL-SEMS-94-028 Installation of a Blind Flange on the Inlet of Containment Purge Valve FCV-25-1 ZPN-PSL-SENP-94-029 0 Shutdown Operations Criteria for Reduced Inventory and Draining the Reactor Coolant System

Safety Evaluations reportable pursuant to 10 CPR SO.SQ for St. Lucie Units 1 0 2 Sumher Revision Title JPN-PSL-SENP-94-037 SIT Discharge/Loop Check Valve Stroke Test JPN-PSL-SENP-94-039 Jumper/Lifted Lead for PDIS-2216 JPN-PSL-SEFJ-94-040 Removal of TE-1122CD Input from Channel "D" of the RPS for PSL 1 JPN-PSL-SENP-94-043 Safety Evaluation Temporary Removal of the ICW Pump Missile Shield JPN-PSL-SENP-94-044 Safety Evaluation for the use of Devoe Devran 140 Epoxy Compound and Kansai Biox as a Coating System for the St. Lucie Unit 1 Intake Structure JPN-PSL-SENP-94-047 SIT Discharge/Loop Check Valve Stroke Test JPN-PSL-SEMP-94-050 '0 Temporary Alterations to the Refueling Water Tank JPN-PSL-SENP-94-065 Containment Air Conditioning for Refueling Outage JPN-PSL-SEEP-94-066 Safety Evaluation for Operation of Three Charging Pumps JPN-PSL-SEMP-94-076 Increase of Engineered Safeguards Suction Piping Design Pressure PROCEDURE SAFETY EVALUATIONS HP-74 0 Personal Dosimetry Procedure 2-LOI-0-65 Licpxid Waste Management System Procedure 1-ONP-01.05 Shutdown cooling Off-Normal Procedure 2-LOI-T-88 Fuel Pool Purification System Temporary Procedure JUMPER AND LIFTED LEAD SAFETY EVALUATIONS 2-94-007 Upender Vertical Circuit Limit Switch (2LS-BV) 1-94-019 Nitroen to Sodium Hydroxide (NaOH)

Tank 1-94-020 Circulating Water Pumps

Safety Evaluations reportable pursuant ta 10 CFR 50.59 far St. Lucie Units 1 f 2 Number Revision Title 1-94-030 Temporary Power Panel 1-94-031 Temporary Power Panel 2-94-039 Circulating Water Pump Seal Water 1-94-046 See ZPN-PSL-SEFZ-94-040 above 2-94-052 Pressurizer Heater Banks 1-94-054 Temperature indicator Substitution

Unit: JPN-PSL-SEMJ-90-039

Title:

Installation of Blind Flange on Outlet of Purge Exhaust Valve FCV-25-6 This temporary change will allow the installation of a blind flange outside of the containment purge discharge isolation valve 2-FCV-25-6. This blind flange will act as a containment isolation device replacing 2-FCV-25-5 which has recently failed its LLRT.

This containment isolation system provides the means of isolating fluid systems that pass through containment penetrations such that any radioactivity that may be released to the containment atmosphere following a postulated Design Basis Accident (DBA) is confined. As such this temporary alteration performs a safety related .function and this report and its associated modification are considered to be safety related.

This change does not affect the function of the 48" containment purge exhaust system during plant-power operations as the system is not used in modes 1, 2, 3 and 4.

Revision 5 deletes reference to the fuel cycle, to allow temporary installation of a blind flange as a containment isolation device in the future and adds references to the evaluation. This revision also deletes the requirement for closure of the penetration bleed-off line (reference valve I-V25208), because the Unit 2 UFSAR Section 6.2.3.3.1 and Figure 9.4-9 require that the containment penetrations for the Containment Purge System (P-10 and P-11) each contain an open 3/4" bleed-off line in to the shield building annulus.

This change does not involve an unreviewed safety question or require a change to the Technical Specification.

SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety if analysis report may be created; or (iii) the margin of safety as defined in the bases for any Technical Specification is reduced.

The Containment Zsolation System provides the means of isolating fluid systems that pass through containment penetrations so that any radioactivity that may be released into the containment atmosphere following a postulated Design Basis Accident (DBA) is confined. The containment isolation valves, the penetration and the piping are designated Seismic Category Z and designed to ASME Code, Section ZZZ and Quality Group B requirements.

Based on the above description, this report and its associated modifications are considered Nuclear Safety Related. This report does not involve an unreviewed safety question, and the following are bases for this conclusion:

'he 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'containment purge exhaust system is neither required to operate during normal plant operation nor after a Design Basis Accident. However, it does perform a containment isolation function. The installation of a temporary blind flange on the exhaust side of valve FCV-25-6 provides the second isolation boundary. Extending the containment isolation boundary into the HVAC room does not affect the environmental qualification of equipment or decrease the potential for off-site dose release during a DBA. The piping associated with valve FCV-25-6 and Penetration P-10, the weld between the closure plate and shield building anchor plant ring and penetration sleeve were evaluated for the additional seismic and dead weight loads of the newly designed flange and existing valve FCV-25-6 associated piping. They were found to be adequate for the additional loads. Application of sealant in the valve packing of valves FCV-'25-5 and 6 will not adversely affect the normal function of the valves and will enhance the ability of the valves to perform its design function.

The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created because the containment purge exhaust system is neither required to function following a postulated Design Basis Accident nor is it required for the operational "design of any system. The blind flange performs the passive function of containment isolation and does not adversely impact any safety related equipment.

iii) The margin of safety as defined in the bases for any Technical Specification is not reduc'ed since the blind flange provides the second isolation boundary. This is consistent with the requirements of the Technical Specifications.

This evaluation constitutes, per 10CFR50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question or requires a change to the technical specifications, thus prior NRC approval for this temporary modification is not required. The temporary blind flange on the outlet of Purge Valve FCV-25-6 is acceptable in that it replaces the isolation function of valve FCV-25-5. The enhancement of valve packing on valves FCV-25-5 and 6 provide additional assurance for the leak tightness. This change does not affect the function of the 48" containment purge exhaust system during the plant-power operation because the system is not used in modes 1, 2, 3, and 4.

Unit: 1 0 2 JPN-PSL-SEFJ-92-009

Title:

St. Lucie Plant Control Element Assembly Operational Life Determination The UFSARs for St. Lucie Units 1 and 2 imply that the operational life of a Control Element Assembly (CEA) is ten years. The vendor the CEAs (Asea Brown Boveri/Combustion who manufactured Engineering) clarified that this limit is ten calendar years and corresponds to approximately eight effective full power years.

This limit is based on a different design CEA than that currently used at St. Lucie. The purpose of this evaluation is to utilize CEA inspection data taken specific to both St. Lucie Units 1 and 2 to determine the operational life of the CEAs being used.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed activity increase the probability of an accident previously evaluated in the UFSARs?

The life limits determined in this evaluation for the PSL-1 and PSL-2 CEAs do not change the overall configuration of either plant. The mode of operation of the plants remains unchanged. No equipment important to safety is affected.

Therefore, the probability of occurrence of an accident previously evaluated in either of the UFSARs is not changed.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSARs?

The life limits presented in this evaluation for the PSL-1 and PSL-2 CEAs have been determined to ensure that both the CEAs currently being used in each reactor and all future replacement CEAs of the same design are capable of performing their intended function over those stated lives. This intended function is to fully insert upon receiving a reactor scram signal. Implementation of the replacement schedules developed in this evaluation will ensure that all CEAs are discharged prior to exceeding the stated design life limits.

As such, the consequences of an accident previously evaluated in either of the UFSARs that are mitigated by a reactor scram are not increased.

' Does the proposed activity increase the 'probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSARs?

Modifying the design life of the CEAs does not impact any other equipment important to safety, nor interfere with the function of any other equipment important to safety. The life limits presented in this evaluation for the PSL-1 and PSL-2 CEAs have been determined to ensure that both the CEAs currently being used in each reactor and all future replacement CEAs of the same design are capable of performing their intended function over those stated lives.

Implementation of the replacement schedules developed in this safety evaluation will ensure that all CEAs are discharged prior to exceeding the stated design life limits. The probability of a CEA malfunction is not increased. Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in either of the UFSARs is not changed.

~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSARs?

Modifying the design life of the CEAs does not require interaction with any equipment important to safety, or prevent any functions of other equipment important to safety.

The life limits presented in this evaluation for the PSL-1 and PSL-2 CEAs have been determined to ensure that both the CEAs currently being used in each reactor and all future replacement CEAs of the same design are capable of performing their intended function over those stated lives.

Implementation of the replacement schedules'developed in this evaluation will ensure that all CEAs are discharged prior to exceeding the stated design life limits. The CEAs will perform their intended function over their design life given a malfunction of other equipment important to safety.

Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in either of the UFSARs are not increased.

5~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR's?

Modifying the design life of the CEAs does not require interaction with any equipment important to safety, or interfere with any functions of other equipment important to safety. The life limits determined in this evaluation for the PSL-1 and PSL-2 CEAs do not change the normal operation of

either of the plants. These life limits have been determined used in each to ensure that both the CEAs currently being reactor and all future replacement CEAs of the same overdesign are capable of performing their intended function those stated lives. Therefore, the possibility of an accidenteither of a different type other than those previously evaluated in of the UFSARs is not created.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSARs?

Modifying the design life of the CEAs does not introduce any new physical interactions with equipment important to safety.

The life limits presented in this evaluation for the PSL-1 and PSL-2 CEAs have been 'determined to ensure that both the CEAs currently being used in each reactor and all future replacement CEAs of the same design are capable of performing their intended function over those developed stated lives.

Implementation of the replacement schedules in this evaluation will ensure that all CEAs are discharged prior to exceeding the stated design life limits. Therefore, the possibility of a malfunct:ion of equipment important to safety of a different type of accident other than those previously evaluated in either of the UFSARs is not created.

7 ~ Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The life limits determined in this evaluation for the PSL-1 and PSL-2 CEAs ensure that both the CEAs currently being used in each reactor and future replacement CEAs of the same design are capable of performing their intended function. As such, the margin of safety defined in the basis for any Technical Specification is not reduced.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewedTechnical safety question, does not require a change to plant Specifications and does not adversely affect plant operation or safety.

Unit: 1 & 2 JPN-PSL-SECJ-93-002

Title:

Specification SPEC-C-013; Installation Guidelines for Miscellaneous Non-System Related Items on Existing Structures Nuclear Engineering Specification SPEC-C-013, entitled "Installation Guidelines for Miscellaneous Non-System Related Items on Existing Structures," has been developed to provide generic guidance for the installation of storage racks, fire extinguisher, storage cabinets, and other miscellaneous items, which are not part of a plant system. These guidelines can be used directly for those installations covered by the support details provided in Appendix B of the specification, or in conjunction with additional guidance provided by Engineering via the "Request for Specification Clarification or Change" form provided in Appendix C.

This evaluation will provide the basis for the acceptability of using the Specification in the maintenance process, in lieu of the current practice which requires that a Plant Change or Modification (PC/M) package be issued and implemented for such cases.

also demonstrate that the Specification meets all technical and It will licensing requirements for St. Lucie Units 1 & 2.

This safety evaluation that the use of the Specification will meet all technical concludes and licensing requirements and will have no adverse impact on plant operations. It is also concluded that the use of the Specification will not compromise the safety and licensing bases for St. Lucie Units 1 & 2. As such, an unreviewed safety question does not exist and a change to the Technical Specifications is not required.

Revision 1 of this Safety Evaluation incorporates the following:

The materials section of this evaluation has been rewritten to denote that Appendix A of the specification provides restrictions only for those details in Appendix B. All other installations require an Engineering evaluation.

Reference to a Civil Calculation has been changed to "latest revision" to allow changes in the corresponding calculation without requiring a revision to the Safety Evaluation when affected.

it is not The UFSAR and Technical Specification amendments have been updated.

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References to specification sections have been updated to support revisions to the specification format. A section has been added to identify affected documents.

Revision 2 of this Safety Evaluation incorporates the revision 2 of the Spec No. SPEC-C-013 which clarifies the definition of Miscellaneous Item in Sect 1.4.1 of the specification and adds a Figure to Appendix B of Specification.

The revisions summarized above do not affect the conclusions of this Safety Evaluation. This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the safety analysis report?

Implementation of the guidelines and support details delineated within the Specification will result in installations that meet all UFSAR requirements for the applicable installation safety classification. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR has not been increased.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the safety analysis report?

As evaluated in the design bases section of this evaluation, installations of miscellaneous items performed in accordance with the Specification will not adversely affect the UFSAR

. accident analysis. Therefore, the consequences of an accident previously evaluated in the UFSAR have not been increased.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety'analysis report?

The installation guidelines and support details delineated within the Specification meet all UFSAR requirements for the applicable installation safety classification. This will assure that adverse interaction with equipment important to 12

safety is precluded. Therefore, the probability of a malfunction of equipment important to safety previously evaluated in the UFSAR has not increased.

4. Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the safety analysis report.

The installation guidelines and support details delineated within the Specification meet all UFSAR requirements for the applicable installation safety classification. This will assure that adverse seismic interaction with equipment important to safety is precluded. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased.

5~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the safety analysis report?

The installation of miscellaneous, non-system related, items in accordance with the Specification will preclude adverse interaction with existing equipment important to safety.

Therefore, the proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the safety analysis report.

Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the UFSAR?

The installation of miscellaneous, non-system related, items in accordance with the Specification will assure that equipment important to safety is not adversely affected.

Therefore, the proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the UFSAR.

70 Does the proposed. activity reduce the margin of safety as defined in the basis for any Technical Specification?

Implementation of the Specification will not impact the Technical Specifications in any way. The use of the installation guidelines and support details delineated within the Specification will result in installations that meet all UFSAR requirements for the applicable installation safety classification. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

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The foregoing constitutes the determination, per 10 CFR 50-59(b),

that the subject activity does not involve an unrevieved safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

Unit: 1 JPN-PSL-SEMS-93-010

Title:

Installation of Hydro Plugs into the ICI Penetrations for Locations J7 and G18 This safety evaluation documents the acceptability of installing hydro plugs in the reactor head instrument flange penetrations for the location J7 and G18 incore detectors. These two detectors are not being installed because difficulties were experienced while attempting to insert the detectors into the conduits in the reactor head. The hydro plugs are similar to the normal instrument seal plugs, except that the hydro plugs are solid. The ICI (incore instrumentation) rhodium detectors do not perform a safety related function per UFSAR Section 4.2.2.2.8; however, they perform a significant monitoring function, assuring that operation remains within the requirements of Technical Specification. The ICI's are designed to fulfil their required function when at least 754 are available. The ICI detector assemblies do perform a safety related pressure boundary function, and house a Core Exit Thermocouple (CET), which performs a safety related function per UFSAR Section 7.5.4.2. Revision 0 of this e'valuation is applicable to St. Lucie Unit 1 operating cycle 12 only because the plugs were to be replaced with functional ICI assemblies during the next refueling outage.

The installation of the hydro plugs will functionally abandon detector locations J7 and G18. Technical Specification 3.3.3.2 requires at least 754 of all incore detector locations to be operable and a minimum of two quadrant symmetric incore detector locations 'er core quadrant to be operable. Although the installation of the hydro plugs will reduce the number of available incore detectors, it will not reduce the number below that required by the technical specifications. The hydro plugs will perform a pressure boundary function identical to the existing detectors and are therefore designed in accordance with ASME Class 1 requirements. Thus, the integrity of the reactor coolant system will not be adversely affected.

The installation of the hydro plugs will result in the loss of two CETs; however, the available number of CETs and the inputs to the Qualified Safety Parameter Display System (QSPDS) will not be reduced to a number below that required by Technical Specification 3.3.3.8. This evaluation concludes that the proposed configuration does not represent an unreviewed safety question and has no impact on plant safety or operations. A review of the technical specifications and the Safety Analysis Report has shown that there are no technical specification changes involved. Revision 0 of this evaluation was valid through the end of Cycle 12 operation.

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Revision 1 to this safety evaluation extends the requirement to replace the hydro plugs with functional ICI assemblies to the refueling outage following cycle 14 operation. This will coincide with the planned replacement of the ICI's. This evaluation concludes that the proposed configuration does not represent an unreviewed safety question and has no impact on plant safety or operations. A review of the technical specifications and the Safety Analysis Report has shown that there are no technical specification changes involved. The analyses and conclusions of the original evaluation have been reviewed and remain valid through the end of Cycle 14 operation.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an

1. Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased since the proposed replacement does not adversely affect any accident initiating components. The ICI detectors do not perform any active functions necessary for the safe shutdown of the plant and the proposed replacement does not create any new unmitigated failure modes for any equipment or systems capable of initiating an accident. The hydro plugs will perform a pressure boundary function identical to the existing detectors and are designed in accordance with ASME Class 1 requirements to ensure reliability. The removal of the ICI detectors and the subsequent introduction of water moderator will present an inconsequential effect on the local power distribution within the fuel assemblies at locations J7 and G18. The removal of this ICI detector will not affect the ability of the ICI system to perform its intended function of measuring the core power distribution.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased since the proposed replacement does not create a new path for'ncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function. The ICI detectors are not required to perform any active safety related functions and the proposed replacement does not adversely impact any equipment which is

required to perform a safety related function or initiate actuation of any safety systems. The hydro plugs are dimensionally equivalent to the normal detector assemblies.

The hydro plugs will perform a pressure boundary function identical to the existing detectors and are designed in accordance with ASME Class 1 requirements. The installation of the hydro plugs will result in the loss of two CETs; the J7 CET'is located in quadrant 3 and provides a signal to channel B of the {}ualified Safety Parameter Display, System ({}SPDS) and

~

the G18 CET is located in quadrant 1 and provides a signal to

{}SPDS channel A. However, the available number of CETs will not be reduced below that which is required by Technical Specification 3.3.3.8.

Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased. No new unmitigated failure modes for any equipment important to safety are introduced by the proposed replacement and no new components or equipment are introduced that could adversely interact with any equipment important to safety. The hydro plugs perform a pressure boundary function identical to the existing detectors and are designed in accordance with ASME Class 1 requirements.

Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased since the proposed replacement does not create a new path for uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function.

The proposed replacement will not adversely impact any equipment which is required to perform an active safety related function or to initiate actuation of any safety systems.

Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The possibility of an accident of a different type than any evaluated previously in the UFSAR has not been created since the proposed replacement does not add or adversely affect any 17

equipment capable of initiating an accident. The proposed replacement does not present any new paths for the loss of reactor coolant system inventory since the hydro plugs are dimensionally equivalent to the normal detector. The hydro plugs will perform a pressure boundary function identical to the existing detectors and are designed in accordance with ASME Class 1 requirements. There are no new unmitigated failure modes for the hydro plugs. In addition, the detector assemblies are passive measurement devices, so their removal could not result in the initiation of an accident of a different type. The removal of the ICI detector and subsequent introduction of water moderator will present an inconsequential effect on the local power distribution within the fuel assemblies at locations J7 and G18. The removal of these ICI detectors will not affect the ability of the ICI system to perform its intended function of measuring the core power distribution.

Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the UFSAR?

The possibility of a malfunction of a different type than any evaluated previously in the UFSAR has not been created since the proposed replacement will not inhibit or otherwise adversely affect the operation of any equipment important to safety. The ICI detectors are passive measurement devices and are not required to perform an active safety related function or activate any safety related systems. The physical interfaces of the detector assembly are not affected by the proposed configuration. The hydro plugs will perform a pressure boundary function identical to the existing detectors and are designed in accordance with ASME Class 1 requirements.

There are no new unmitigated failure modes for the hydro plugs. The installation of the hydro plugs will result in the loss of two CETsg however, the two CETs are located in different quadrants and supply signals to different QSPDS channels. The available number of CETs will not be reduced below that which is required by Technical Specification 3.3.3.8. In addition, these hydro plugs do not create any new modes of operation for any safety related equipment.

Does the proposed activity reduce the margin of safety as defined in the basis for any technical specification?

The proposed activity does not reduce the margin of safety as defined in the basis for any technical specification since the removal of the incore detectors at locations J7 and G18 will not adversely impact the minimum number of incore detectors required for operation as defined in Technical Specification 18

3.3.3.2. The hydro plugs will perform a pressure boundary function identical to the existing detectors and are designed in accordance with ASME Class 1 requirements. Thus, the integrity of the reactor coolant system will not be adversely affected. The installation of the hydro plugs will result in the loss of two CETs; the J7 CET is located in quadrant 3 and provides a signal to QSPDS channel B and the G18 CET is located in quadrant 1 and provides a signal to QSPDS channel A. .Since the two CETs are located in different quadrants and supply signals to different QSPDS channels, the available number of CETs will not be reduced below that which is required by Technical Specification 3.3.3.8.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and 'does not adversely affect plant operation or safety.

Unit: 1 Si 2 JPN-PSL-SECJ-93-011

Title:

Safety Evaluation for Specification SPEC-C-005, Component Mounting and Supports Nuclear Engineering Specification SPEC-C-005, entitled "Specification for Component Mounting and Supports, St Lucie Units 1 and 2, and Turkey Point Units 3 and 4", has been developed to provide generic component mounting instructions and support details. These mounting instructions and support details can be used in conjunction with design output documents presently used in the procurement/maintenance process (e.g.,'rocurement Technical Evaluation, Item Equivalency Evaluation, Plant cwork Order) to install replacement components weighing less than 50 pounds.

This evaluation will provide the basis for the acceptability of using the Specification in'the maintenance process, in lieu of the current practice which requires that a Plant Change or Modification (PCM) package be issued and implemented for such cases. It also demonstrate that the Specification meets all technical and will licensing requirements for thh St Lucie Units 1 and 2.

This safety evaluation concludes that the use of the Specification will meet all technical and licensing requirements and will have no adverse impacts on plant operations. It is also concluded that the use of the Specification will not compromise the safety and licensing bases for St Lucie Units 1 and 2. As such, an unreviewed safety question does not exist and a change to the Technical Specifications is not required.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine if the subject activity constitutes an unreviewed safety question:

Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the safety analysis report?

The use of mounting instructions and support details delineated within the Specification will result in installations that meet all UFSAR requirements for the applicable installation safety classification. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR has not been increased.

20

0'oes the proposed activity increase the consequences of an accident previously evaluated in the safety analysis report?

As evaluated in the design basis and analysis section of this evaluation, all mounting or support modifications performed in accordance with the Specification will not adversely affect the UFSAR accident analysis. Therefore, the consequences of an accident previously evaluated in the UFSAR have not been increased.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report?

The mounting instructions and support details delineated within the Specification meet all UFSAR requirements for the applicable installation safety classification. This will assure that equipment important to safety is properly supported under all design basis loading conditions, and that adverse seismic interaction with equipment important to safety is precluded. Therefore the probability of a malfunction of equipment important to safety previously evaluated in the UFSAR has not increased.

Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the safety analysis report?

The mounting instructions and support details delineated within the Specification meet all UFSAR requirements for the applicable installation safety classification. This will assure that equipment important to safety is properly supported under all design basis loading conditions, and that adverse seismic interaction with equipment important to safety is precluded. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the safety analysis report?

Installation of replacement supports and remounting of equipment in accordance with the Specification will assure that equipment important to safety is appropriately supported under all design basis loading conditions. It will also preclude adverse seismic interaction with existing equipment important to safety. Therefore, the possibility of an accident of a different type than any previously evaluated in the safety analysis report has not been created.

21

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the safety analysis report'P Installation of replacement supports and remounting of equipment in accordance with the Specification will assure that equipment important to safety is appropriately supported under all design basis loading conditions. It will also preclude adverse seismic interaction with existing equipment important to safety. Therefore, the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the UFSAR has not been created.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The use of the Specification will not impact the Technical Specifications in any way. The use of mounting instructions and support details delineated within the Specification will result in installations that meet all UFSAR requirements for the applicable installation safety classification. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

22

Unit: 1 0 2 JPN-PSL-SECJ-93-012

Title:

Safety Evaluation for Specification SPEC-C-019, Tubing and Tubing Supports Nuclear Engineering Specification SPEC-C-019, entitled Specification for Tubing and Tubing Supports, St. Lucie Units 1 and 2", has been developed to provide generic installation instructions and support details for Safety Related, Quality Related and Non-Nuclear tubing. These installation instructions and support details can be used as specified within Engineering output design documents such as the Plant Change or Modification (PC/M). Additionally, maintenance repair/replacement activities may be performed on tubing supports using the specified standard supports for those installations directly covered by the specification criteria or in conjunction with additional guidance provided by Engineering via the "Request for Specification Clarification or Change" form provided in Appendix A.

This evaluation will provide the basis for the acceptability of using the specification in the maintenance process, in lieu of the current practice which requires that a PC/M package be issued and implemented for such cases.

~

~ ~ ~ ~

~ It will also demonstrate that the

~

specification meets technical and licensing requirements for the

~ ~ ~

St. Lucie Units 1 and 2.

~ ~

~

will meet technical and licensingthatrequirements This safety evaluation concludes

~

the use of the specification

~ ~

and will have no adverse impacts on plant operations. It is also concluded that the use of the specification will not compromise the safety and licensing bases for St. Lucie Units 1 and 2. As such, an unreviewed safety question does not exist and a change to the Technical Specifications is not required.

SAFETY EVALUATION Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the safety analysis report?

The use of installation instructions and support details delineated within the specification will result. in installations that meet UFSAR requirements for the applicable installation safety classification. Therefore, the probability of occurrence'of an accident previously evaluated in the UFSAR has not been increased.

23,

2. Does the proposed activity increase the consequences of an accident previously evaluated in the safety analysis report?

As evaluated in the design basis and analysis section of this evaluation, all support modifications performed in accordance with the specification will not adversely affect the UFSAR accident analysis. Therefore, the consequences of an accident previously evaluated in the UFSAR have not been increased.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report?

The installation instructions and support details delineated within the specification meet UFSAR requirements for the applicable installation safety classification. This will ensure that equipment important to safety is properly supported under all design basis loading conditions, and that adverse seismic interaction with equipment important to safety is precluded. Therefore the probability of a malfunction of equipment important to safety previously evaluated in the UFSAR has not increased.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the safety analysis report?

The installation instructions and support details delineated within the specification meet UFSAR requirements for the applicable installation safety classification. This will ensure that equipment important to safety is properly supported under all design basis loading conditions, and that adverse seismic interaction with equipment important to safety is precluded. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased.

5~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the safety analysis report'?

Installation of replacement supports in accordance with the specification will assure that equipment important to safety is appropriately supported under all design basis loading conditions. It will also preclude adverse seismic interaction with existing equipment important to safety. Therefore, the possibility of an accident of a different type than any previously evaluated in the safety analysis report has not been created.

24

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the safety analysis report?

Installation of replacement supports in accordance with the specification will assure that equipment important to safety is appropriately supported under all design basis loading conditions. It will also preclude adverse seismic interaction with existing equipment important to safety. Therefore, the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the UFSAR has not been created.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The use of the specification will not impact the Technical Specifications in any way. The use of installation instructions and support details delineated within the specification will result in installations that meet UFSAR requirements for the applicable installation safety classification. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

~

that the subject activity does not involve an unreviewed safety

~ ~ ~

question, does not require a change to plant Technical

~ ~

Specifications and does not adversely affect plant operation or

~ ~

safety.

25

Unit: 1 E 2 JPN-PSL-SECS-93-014

Title:

Safety Evaluation for Specification SPEC-C-017 Small Bore Piping supports Nuclear Engineering Specification SPEC-C-017, entitled "Procurement, Fabrication and Installation of Small Bore Pipe Supports," has been developed to provide generic installation instructions and support details for Safety Related and Quality Related small bore piping systems. These installation instructions and support details can be used as specified within Engineering output design documents such as the Plant Change or Modification (PC/M). Additionally, maintenance repair/replacement activities may be performed on small bore pipe supports using the specified standard supports in conjunction with additional guidance provided by Engineering via the "Maintenance Request Approval" (MRA) form provided in Appendix B.

This evaluation will provide the basis for the acceptability of using the specification in the maintenance process in lieu of the current practice which requires that a PC/M package be issued and implemented for such cases. It will also demonstrate that the specification meets all technical and licensing requirements for St. Lucie Units 1 E 2.

This safety evaluation concludes that the use of the specification will meet all technical and licensing requirements and will have no adverse impact on plant operations; It is also concluded that the use of the specification will not compromise the safety and licensing bases for St. Lucie Units 1 and 2. As such, an unreviewed safety question does not exist and a change to the Technical Specifications is not required.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine if the subject activity constitutes an unreviewed safety question:

Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the safety analysis report?

The use of installation instructions and support details delineated within the specification will result in installations that meet all UFSAR requirements for the applicable installation safety classification. Therefore, the 26

probability of occurrence of an accident previously evaluated in the UFSAR has not been increased.

2 the proposed activity increase the consequences of an accident previously evaluated in the safety analysis report?

As evaluated in the design basis and analysis section of this evaluation, all support modifications performed in accordance with the specification will not adversely affect the UFSAR accident analysis. Therefore, the consequences of an accident previously evaluated in the UFSAR have not been increased.

'oes 3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report?

The installation instructions and support details delineated within the specification meet all UFSAR requirements for the applicable installation safety classification. This will assure that equipment important to safety is properly supported under all design basis loading conditions, and that adverse seismic interaction with equipment important to safety is precluded. Therefore the probability of a malfunction of equipment important to safety previously evaluated 'in the UFSAR has not increased.

Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the safety analysis report?

The installation instructions and support details delineated within the specification meet all UFSAR requirements for the applicable installation safety classification. This will assure that equipment important to safety is properly supported under all design basis loading conditions, and that adverse seismic interaction with equipment important to safety is precluded. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased.

5. Does .the proposed activity create the possibility of an accident of a different type that any previously evaluated in the safety analysis report?

Installation of replacement supports in accordance with the specification will assure that equipment important to safety is appropriately supported under all design basis loading conditions. It will also preclude adverse seismic interaction with existing equipment important to safety. Therefore, the possibility of an accident of a different type than any 27

previously evaluated in the safety analysis report has not been created.

6. Does the proposed activi'ty create the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the UFSAR?

Installation of replacement supports in accordance with the specification will assure that equipment important to safety is appropriately supported under all design basis loading conditions.

with existing It will also equipment preclude important adverse to seismic safety.

interaction Therefore, the possibility of.a malfunction of equipment important to safety of a different type other than any previously evaluated in the UFSAR has not been created.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The use of the specification will not impact the Technical Specifications in any way. The use of installation instructions and support details delineated within the specification will result in installations that meet all UFSAR requirements for the applicable installation safety classification. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

28

I Unit: 1 JPN-PSL-SEFJ-93-014

Title:

Safety Evaluation of Spent Fuel Pool (SFP) Coupon Surveillance Program High density fuel storage racks were installed into the spent fuel pool (SFP) of St. Lucie Unit 1 in 1988. The design of the high density racks includes use of the strong neutron absorption material, Boraflex, to maintain the SFP in a subcritical condition.

Boraflex panels were installed. between spent fuel assemblies in the SFP racks to control the neutron multiplication factor. To monitor the Boraflex panel in-service performance, a surveillance program, "SFP Boraflex Coupon Surveillance," has been implemented to study Boraflex degradation mechanisms with time.

A technical review of the SFP Boraflex coupon surveillance program at St. Lucie Unit 1 was performed and completed. Review of the results has indicated that the coupon surveillance program has little merit with respect to practical applications to the in-service performance of Boraflex panels in racks. The St. Lucie Unit 1 ongoing surveillance program will be suspended, once the new improved program is implemented.

An enhanced Boraflex panel verification program consisting of a technique for predicting Boraflex gamma dose and performing periodic blackness testing has been developed to replace the existing in-service Boraflex coupon surveillance program. Details of the program are described in this evaluation.

Revision 1 of this Safety Evaluation was added to specifically address the removal of the coupon surveillance program and provided an unreviewed safety question determination. In addition, some minor wording changes have been added to clarify information associated with the coupon surveillance program. Nevertheless, the previous conclusion and recommendations are still valid for this revision.

Revision 2 of this Safety Evaluation has been made to extend the blackness test performance date to 1995. The original conclusion and recommendations are still valid for revision 2.

This evaluation concludes that the activity as described above does ~

not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

i' 29

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine if the subject activity constitutes an unreviewed safety question:

Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

'I The verification program contains the use of the blackness test which was previously approved by the NRC staff. Based on the past test at PSL1, fuel movements in the SFP are needed to make room for testing. However, this is a routine operation.

The proposed blackness testing is not an accident initiator and, therefore, the probability of occurrence of an accident evaluated in the UFSAR is not. increased.

The purpose of the coupon program which will be eliminated is to provide surveillance of the Boraflex used in the SFP. The coupon program is a passive surveillance program and has no function which can impact the probability of occurrence of an accident previously evaluated in the UFSAR. Therefore, the removal of the coupon surveillance program will not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

2 0 Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The proposed activity can involve two independent accidents that can result in radiological activity releases. The first accident is a fuel handling accident previously evaluated in the UFSAR. The postulated accident due to fuel mishandling can result in fission product releases. The blackness test operation does not affect the fission product release mechanism during the accident. The second accident is a drop of the source container on top of the storage cells. Review of the fuel assembly drop analysis has shown that the falling source container would not damage the stored fuel due to the small weight of the container in water. The worst case from a drop of the source container would be radioactivity releases in water from the ~Cf source. But, the level of the source radioactivity release is low and is bounded by the fuel handling accident. Therefore, the proposed activity will not increase the consequences of an accident previously evaluated in the UFSAR.

The coupon program which will be eliminated is used to provide surveillance of the Boraflex used in the SFP. The coupon program is a passive surveillance and has no impact on the 30

consequences of any accident. Therefore, the removal of the coupon program will not increase the consequences of an accident previously evaluated in the UFSAR.

Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The. blackness test includes a movable stainless steel container which contains a neutron source and four BF~

detectors. During the test operation, the container will axially traverse from the storage cell top to bottom or vice versa. Since the size of the 'container is smaller than the storage cell opening and the'raversing speed of the container is slow in the SFP water, the container traversing speed is well controlled and will scrape neither the steel cover plate (for Region 2) nor the thick cell wall in region 1. The Boraflex panel is protected by the cover plate and cell wall and thus will not be damaged by the container during the testing. Therefore, this proposed test will not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

The coupon program which will be eliminated is used to provide surveillance of the Boraflex used in the SFP. The surveillance program does not have a functional relationship with any systems, structures, and components. Thus, the elimination of the surveillance program will not create any new failure modes. Therefore, the probability of occurrence of any equipment malfunction important to safety previously evaluated in the UFSAR will not increase.

Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The blackness test neither alters the dimensions of the storage cell nor changes the chemical composition of the

'Boraflex. The criticality control parameters are not affected. Therefore, the proposed test activity will not increase the consequence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

The coupon program which will be eliminated is used to provide surveillance of the Boraflex used in the SFP. Removal of the coupon program will not impact any equipment important to safety since the coupon program has,no interaction with any structure, system or component important to safety. None of the results of any transient analysis reported in the UFSAR are changed because of the coupon surveillance removal.

31

Therefore, the removal of the coupon surveillance program will not increase the consequences of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

E Based on experience gained from the blackness testing conducted at Turkey Point Unit 3 SFP in December 1993, the test requires a minimum three (3) emptied cells distance between neutron source and the BF~ detectors in all direction.

During the test, the ~Cf neutron source, considered as a point source, is positioned at the center cell and is surrounded by the Boraflex panels in racks and a thick layer of borated water. The distance between the source and the nearest fuel is at least 27 inches. Such physical configuration with a low intensity of neutron source can not sustain the K-eff value greater than 0.95 (SFP criticality already considered in UFSAR). As such, the proposed test, will not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

The removal of the coupon surveillance program does not change the operation, function or design bases of any structures, systems or components important to safety as described in the UFSAR. No new failure modes or limiting single failures have been created as a result of the coupon surveillance removal.

Therefore, there is no possibility of an accident of a different type than any previously evaluated in the UFSAR.

Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR?

The only physical function of the Boraflex panels installed between two consecutive storage cells is to control and to maintain a K-eff equal to or less than 0.95. Performance of the blackness test does not change the physical function of Boraflex panels in racks. Therefore, the proposed test will not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR.

The removal of the coupon surveillance program produces no new hazard that can be postulated to cause a malfunction of equipment important to safety different from those previously analyzed in the UFSAR. Thus, there is no possibility of a 32

malfunction af equipment important to safety of a different type than any previously evaluated in the UFSAR.

7 ~ Does the proposed activity reduce the margin of safety as defined in the basis of any Technical Specification?

The removal of the coupon surveillance program will not impact the margin of safety as defined in the basis of the Technical Specifications since the blackness test will be used as the mechanism for monitoring Boraflex degradation. The coupon has been used to determine that the Boraflex in the 'rogram SFP has not degraded below the analysis assumptions used in the criticality evaluation. The initiation af the blackness test as a replacement for the coupon surveillance will improve our ability to track Boraflex degradation and therefore, better ensure that the SFP design basis will be maintained.

The requirement for maintaining a SFP K-eff below 0.95 is not modified by this change. Therefore, the removal of the coupon surveillance program will not impact the margin of safety as defined in the bases for the Technical Specifications.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety questian, does not require a change to plant Technical Specifications and does not adversely affect plant aperation or safety.

,33

Unit: 1 JPN-PSL-SEFZ-93-024

Title:

Loss of Feedwater Transient with Corrected Steam Generator Inventory Error During a review of the current licensing basis analyses performed to support the on-going steam generator replacement project, a discrepancy between the results obtained by Babcock and Wilcox International Limited (BWIL) and those given in the UFSAR for the Loss of Feedwater event (LOFW) was identified. Upon review, determined that the correct initial steam generator (SG) mass it was value was approximately 18,000 ibm lower than what was used in the St.

Lucie Unit 1 safety analysis (non-conservative). This mass error impacts the UFSAR Chapter 15 LOF event and the auxiliary feedwater section of UFSAR Chapter 10. I Seimens Power Corporation and FPL performed a review to determine if a significant safety defect existed in 'he SPC analyses performed and reported in the Loss-of-Feedwater Transient With Reduced Steam Generator Low Level Trip Setpoint, ANF-89-113.

was concluded that substantial safety defect did not exist and that It the reporting requirements of 10CFR21 were not applicable. The errors were corrected and a re-analyses was performed using more realistic input assumptions. The corrected results meet all acceptance criteria and are supported by the Technical Specification limits.

This evaluation concludes that the activity as described above does 10CFR50.59, Specifications does not require a change and does not adversely

'o not represent an unreviewed safety question as defined in affect plant Technical plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The re-analyses of the Chapter '10 and Chapter 15 LOFW Transient with the corrected SG inventory do not change the overall configuration of the plant. The mode of operation of the plan remains unchanged. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR report is not increased.

34

0'oes the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The re-analyses of the Chapter 10 and Chapter 15 LOFW transient with the corrected SG inventory do not affect any structures systems or components that function to mitigate the consequences of an accident by containing or detecting the release of radioactivity.

Based on the review of the identified SG inventory errors for all Chapter 15 events, the Chapter 10'LOFW Transient With Off-site Power Available and the Chapter 15 LOFW transient without initiation of AFW events were determined to require reanalysis. The results of these analyses demonstrate that the conclusions reached to support the reduce low SG level setpoint remain valid, and that no safety limits are violated.

Therefore, the consequences of an accident previously evaluate in the UFSAR are not increased.

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

This activity does not create any spatial or functional interaction with any structure, system or component important to safety. Specifically, the re-analysis of the LOFW Transient with the corrected SG inventory error does not have a functional relationship with any systems, structures and components nor does it create any new failure modes.

Therefore, the probability of occurrence of any equipment malfunction important to safety previously evaluated in the UFSAR will not increase.

4 ~ Does the proposed activity increase the consequences of malfunction of equipment important to safety previously evaluated in the UFSAR?

The re-analyses of the LOFW Transients with the corrected SG inventory error do not create any spatial or functional interaction with any structure, system or component important to safety. Chapter 15 of the UFSAR describes postulated plan

'transients which could occur as a result of equipment failures. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR are not increased.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

35

The re-analyses of the LOFW Transients with the corrected SG inventory error do not change the operation, function or design bases of any structures, systems or component important to safety as described in the UFSAR. No new failure result modes or of this limiting single failures have been created created a

result of the re-analysis. No new hazaxds are as a LOFW Transient re-analysis that can be postulated to cause an accident different from those previously analyzed in the UFSAR.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The re-analyses of the LOFW Transients with the corrected SG inventory error do not create any spatial or functional interaction with any structure, system or component important to safety. No new hazards are created that can be postulated to cause a malfunction of equipment important to safety different from those previously analyzed in the UFSAR. Thus, the possibility of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased.

71 Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification' The re-analyses of the LOFW Transients with the corrected SG inventory error do not change the design bases, functions, or operations of any safety-related equipment and do not adversely affect any other safety-related structures, systems or components. The Technical Specification requirements and bases applicable to the LOFW event. with the corrected mass error and more realistic assumptions are not affected.

Therefore, this mass error does not reduce the margin of safety as defined in the basis for the Technical Specifications.

The re-analysis of the LOFW Transient with the corrected SG inventory error does not impact safe operation of the plant, and does not constitute an unreviewed safety question.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical ~

Specifications and does not adversely affect plant operation or safety.

36

Unit: JPN-PSL-SENP-93-035

Title:

Evaluation of Inventory Loss for the Refueling Water Tank PURPOSE The purpose of Revisions 0 and 1 of this evaluation is to:

1. Determine the functionality of the Unit 1 Refueling Water Tank (RWT) given the presence of a leak through the bottom of the tank.

2. Determine a maximum allowable leakage rate.

3. Evaluate functionality of the RWT during visual inspections and repairs requiring tank entry.

4. Provide recommendations for temporary non-code repair.

The purpose of Revision 2 of this evaluation is to:

Provide direction for a permanent ASME Code acceptable repair to the bottom of the Refueling Water Tank.

2.~ Provide direction for activities required to support a root

~ ~ ~ ~ ~ ~

cause evaluation.

~

~

BACKGROUND The Refueling Water Tank provides a source of borated primary grade water for refueling, reactor coolant makeup, and accident conditions. The RWT is an ASME" Class 2 structure, erected in accordance with ANSI B96.1-1967. Following construction of the RWT in 1974, Fort Pierce city water was added to the tank; this city water remained in the tank for approximately four months. This water aggressively attacked the aluminum surface and caused pits on the bottom and side plates. A cleaning process was performed, followed by inspections and evaluations of the pitting; recommended at that time that no further corrective actions be it was performed, and that the tank should be filled with borated primary water.

A steady loss of RWT inventory was observed for several weeks in June-and July of 1993. The rate of the loss measured at that time was approximately 2 gpm. The piping systems connected to the RWT and the above ground exterior surfaces of the tank were inspected and were eliminated as a source of inventory loss. Individual isolation of all lines penetrating the tank was performed, with no resultant decrease in tank leakage rates. Samples from underground test wells located near the RWT indicated both tritium and boron.

37

1 Based on this evidence, was through the 'bottom it of was concluded that the inventory loss the tank. Although the exact nature of the leak due to a was unknown single at small the hole time, it was considered to be likely (resulting from pitting or similar mechanism), a series of small holes, or a small separation in a weld joint.

In July, 1993, an acoustic emissions (AE) analysis was performed using externally mounted equipment and a transducer mounted on a mini-rover submarine. From this analysis, a single leak approximately 3/16 inch in diameter was located, in an area on the east side of the tank. Approximately 1000 ultrasonic thickness readings were taken in areas identified through. AE analysis as probable leak locations, as well as at random locations over the entire tank bottom. No general wall thinning of the tank bottom was identified.

A review of Generic Letter No 90-05 was performed to determine the basis for allowing non-code repairs to code class components. It was determined that any non-code repair to Class 1 or 2 piping must be performed in accordance with an engineered specification which would provide a boundary that would be of equal structural'strength as the original design basis of the component.

The repair recommendation was to perform an immediate non-code repair by using an epoxy coating to adhere an aluminum plate to the tank bottom. A code relief request was prepared requesting NRC approval for the non-code repair. This relief request was submitted to the NRC, which also included Revision 1 of this evaluation as a basis for the temporary repair; in the relief request, FPL committed to providing a code acceptable repair during the Fall 1994 refueling outage. The NRC in accordance with 10 CFR 50.55a(g)(6)(i) granted the requested relief and accepted the temporary repair until the Fall 1994 refueling outage, at which time the RWT bottom plate is required to be repaired or replaced in accordance with the provisions of the ASME Code. The Code also requires a root cause evaluation to be performed to ensure that the repair process addresses the cause of the failure.

In accordance with 10, CFR 50.59, the responses to the below listed questions serve to determine if the subject activity constitutes an unreviewed safety question:

Does the proposed activity increase the probability of occurrence of an accident previously- evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR will not increase because the recommended repair process does not initiate an accident or affect any accident scenario. The erection of scaffolding 38

\

will be performed in accordance erected with Administrative Procedure No. 0010724. All scaffolding in accordance with the guidelines of the procedure is capable of withstanding seismic loads. For this reason, the erection of the scaffolding will have no adverse effect on any safety related equipment. The code repair will be performed in accordance with the applicable requirements of the ASME Code (Section XI) and ASME B96.1 (in accordance with the original design specification).

Therefore,'for the reasons discussed above, this repair will not increase the probability of occurrence of any accident previously evaluated in the UFSAR. N 2~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

All scaffolding will be erected in accordance with the provisions of Administrative Procedure No. 0010724, and will therefore be capable of withstanding seismic loads, thus ensuring that there will be no adverse interactions with safety related items. Additionally, precautions have been specified to be taken during decontamination operations to ensure that the divers or equipment introduced into the RWT do not adversely affect the operation or safety related functions of the tank. Therefore, the scaffolding and decontamination operations to be performed while Unit 1 is on line will not affect the ability of the RWT to supply borated water in the event of a SIAS. The permanent code repair will be performed when the RWT is out of service and not being relied upon to provide borated water for any safety-related function as described in the UFSAR. The code repair will be performed in accordance with the applicable requirements of the ASME Code (Section XX) and ASME B96.1. For the reasons discussed above,

'this repair will not, increase the consequences of any accident previously evaluated in the UFSAR.

3 ~ Does the proposed activity increase the probability of an occurrence of a'alfunction of equipment important to safety previously evaluated in the UFSAR?

All scaffolding will be erected in accordance with the provisions of Administrative Procedure No. 0010724, and will therefore be capable of withstanding seismic loads, thus ensuring that there will be no adverse interactions with safety related items. Precautions have been specified to be taken during decontamination operations to ensure that the divers or equipment introduced into the RWT do not adversely affect the operation or safety related functions of the tank.

Therefore, the scaffolding and decontamination operations to be performed while Unit 1 is on line will not affect the ability of the RWT to supply borated water in the event of a 39

SIAS. The permanent code repair will be performed when the RWT is out- of service; the repair will be performed in accordance with the applicable requirements of the ASME Code (Section XI) and ASME B96.1, and will therefore create no adverse interaction with any structure, system, or component important to safety. The inspections and examinations to support the root cause analysis will be performed on the section of the bottom plate removed from the tank bottom, and will therefore have no effectof onoccurrence any safety related items.

of any equipment Theiefore, the probability malfunction important to safety previously evaluated in the UFSAR will not increase.

4~ Does the proposed activity increase the cohsequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The permanent code repair (and associated operations related to scaffolding and decontamination) will not prevent the RWT or any other safety related equipment from performing its safety related functions. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The permanent code repair involves only a repair to the RWT bottom to prevent further leakage of borated water. This repair does not change the operation, function or design bases of any structure, system or component, important to safety as described in the UFSAR. Precautions have been specified for diving operations related to decontamination. Scaffolding will be erected in accordance with procedure, and will therefore be capable of withstanding seismic loads, thus ensuring that there will be no adverse interactions with safety related items. Therefore, no new hazards are created that can be postulated to cause an accident different from those previously analyzed in the UFSAR, and there is no possibility that 'an accident may be created that is different from any already evaluated in the UFSAR.

6 Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The permanent code repair does not change the operation, function, or design basis of any structure, system or component important to safety as described in the UFSAR. The 40

code repair will be performed in accordance 1

with the applicable requirements of the ASME Code (Section XI) and ASME B96.1, and will therefore create no adverse interaction with important to safety.

any structure, system, or component Precautions have been specified for diving operations related to decontamination. Scaffolding will be erected in accordance with procedure, and will therefore be capable of withstanding seismic loads, thus ensuring that there will be no adverse interactions with safety related items. This repair, as evaluated above and in Section VI, does not create any new malfunction of equipment important to safety. Therefore, the possibility of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

This repair does not change the design bases, functions or operations of any safety related equipment and does not adversely affect any other safety related structures, systems or components. The code repair will be performed in accordance with the applicable requirements of the ASME Code (Section XI) and ASME B96.1, and will therefore create no adverse interaction with any safety related items. A new filler plate, along with- additional bearing material (as necessary) will be added to provide uniform support for the new plate which will be added for the code repair. The Technical Specification requirements and bases for RWT inventory and boron concentration are not affected by this repair. Therefore, this activity does not reduce the margin of safety as defined in the bases for the Technical Specifications.

The permanent code repair to the RWT does not impact safe operation of the plant, does not constitute an unreviewed safety question and does not require a change to the Technical Specifications.

Unit: JPN-PSL-SEMP-94-001

Title:

Temporary Installation of Strain Measurement Devices on the Pressurizer Relief Valve Discharge Piping This safety evaluation will allow the temporary installation of strain measuring devices (i.e., strain gages) on the three pressurizer relief valve discharge pipe lines. Strain measurements will permit the determination of moments valves.

imposed by the discharge piping on the pressurizer safety relief The installation of strain gages is classified as Non-Nuclear Safety Related because: 1) The strain measurements will not be used for safety related functions; 2) The pressurizer relief valve discharge Lines are Non-Nuclear Safety (i.e., not Safety Class 1, 2 or 3),

designed in accordance with the ANSI B31.1 Code; and 3) The strain measuring devices are non-intrusive and will not impact the operation or function of any Nuclear Safety Related Systems,

'tructures or Components.

The safety evaluation addresses the potential impact of the the guidelines set forth in 10 CFR 50.59. 's installation of strain gages has on plant operation and responds to long as the requirements of this evaluation are followed, there are no adverse effects on plant operation or safety. The results of the safety evaluation show that no unreviewed safety question exists and that no Technical Specification changes are required. Therefore, prior NRC approval is not required pursuant to 10 CFR 50.59.

The strain gages and thermocouples, excluding the data acquisition equipment ("Strain Indicator VISHAY P-3500" and "OMEGA Engineering Multipoint Digital Thermometer Model 2166A"), will remain in place for Cycle 8 and shall be removed prior to Mode 2 for Cycle 9. This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed change increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed change does not affect any equipment whose malfunction is postulated in the UFSAR to initiate an accident 42

or prevent an accident from occurring. No physical modifications have been performed to the RCS or connected systems except to attach the strain gage shims and Omega clips to the pressurizer relief valve discharge piping using micro-spot welding generating energy output less than 50 watt-seconds. The piping is not Safety Class 1, 2 or 3.

Additionally, no failure modes have been identified that could initiate an accident previously evaluated in the UFSAR. As such, the probability of occurrence of an accident previously evaluated in the UFSAR has not increased.

2 ~ Does the proposed change increase the consequences of an accident previously evaluated in the UFSAR?

The proposed change does not diminish in any way the ability of the pressurizer relief valve discharge piping or any other safety system to perform its intended function. There is no interaction with any safety related equipment. As such, the proposed change does not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed change increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed change is to install strain measuring devices which do not interfere with the operation of any system.

There is no interaction with any safety related equipment.

The strain gage an/or thermocouples (including temporary field routed cables) will not pass through the sump water screens the assembly becomes loose and carried to the containment if sump. As such, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased by the proposed change.

4 ~ Does the proposed change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The installation, operation or failure of the strain measuring devices will neither impact operation of any system nor cause any adverse affect to any safety related equipment.

Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR are not increased by the proposed change.

5. Does the proposed change create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

43

As discussed in the failure modes and effects analysis section of this evaluation, the proposed change does not introduce any new failure modes to or impact safety related equipment in any way. The proposed change will not change or impact the requirements of the design bases of the safety related systems as described in the UFSAR. There are no postulated failure modes which could be considered accident initiating.

Therefore, the possibility of an accident of a different type than any previously evaluated in the UFSAR is not created by this modification.

6~ Does the proposed change create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed change does not interact spatially or functionally with any structure, system or component important to safety other than the attachment of strain gages to the pressurizer safety relief valve discharge piping. No new failure modes are created that can be postulated to interact with any equipment important to safety different than those previously analyzed in the UFSAR. Therefore, the possibility of a malfunction of equipment important to safety which is of a different type than any previously evaluated in the UFSAR is not created by the proposed change.

7 ~ Does the proposed change reduce the margin of safety as defined in the basis for any Technical Specification?

The Technical Specification requirements and Technical Specification Bases are not affected by the proposed change.

The proposed change does not affect any plant Technical Specification requirement. The proposed change maintains the level of protection previously evaluated in the UFSAR.

Therefore, the margin of safety as defined in the bases for any Technical Specification is not reduced by this modification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

44

Unit: 2 JPN-PSL-SEFJ-94-002

Title:

Evaluation of Reduced Shutdown Cooling Flow Rate For St.

Lucie Unit 2 Cycle 8 Refueling Outage This evaluation addresses the single point injection operation of the shutdown cooling (SDC) system in Mode 5, for St. Lucie Unit 2, during the 1994 outage. For the maintenance activity on the low pressure safety injection (LPSI) header lines, one of the four safety injection lines will be isolated. If a failure of one SDC train is assumed,, this configuration will leave one SDC loop in operation with reduced flow from one LPSI pump injecting through one safety injection line. The St. Lucie Unit 2 Technical Specifications (T.S.) 3.4.1.4.1 and 3.4.1.4.2 require that two SDC loops be operable in Mode 5 (at least one loop be in operation) when the reactor coolant loops are not filled or when the water level in steam generators is less than 104 indicated narrow range level. The evaluation performed here verifies the adequacy of one train with single point injection to. satisfy the Technical Specifications bases requirements associated with decay heat removal, boron dilution and boron stratification, under the initial conditions and assumptions specified in referenced letters, except that the operational mode is for Mode 5.

The T.S. 3/4.4.1 bases requirements 'ensure that i) sufficient cooling is available to remove decay heat, and ii) sufficient coolant mixing and circulation is maintained to minimize the effects of boron concentration reduction and prevent boron stratification.

The evaluation shows that in the proposed configuration the maximum reactor coolant system (RCS) temperature is 134.3 F if the maintenance is done 7 days after the reactor shutdown, and is close to 140 F for the 5 days after shutdown case. These, temperatures are well below the maximum Mode 5 temperature of 200'F. Additionally, boron stratification is shown not, to 'be a concern and sufficient time is available for the operator to terminate any boron dilution event. It is shown that there is no unreviewed safety question associated with the proposed maintenance activity.

This safety evaluation is revised to delete the assumption that nozzle dams would 'e installed. This assumption is not a requirement to ensure the validity of the safety analyses. A review was performed and concluded that there is no effect on the conclusions of this evaluation. This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to 45

plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION below listed In accordance with 10 CFR 50.59, the responses to theconstitutes questions serve to determine if the subject activity an unreviewed safety question:

Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed activity removes part of the shutdown cooling (SDC) system for ma'intenance on low pressure safety injection (LPSI) header lines under specified conditions. The only UFSAR event related to the proposed activity is the Boron Dilution event. However, the proposed activity does not lead to an increase in the frequency of such a transient.

Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR is not increased.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The proposed activity will reduce the SDC flow during the maintenance on the LPSI header lines. This flow reduction has been shown to be acceptable for decay heat removal, boron mixing and prevention of boron stratification, and meet T.S.

bases requirements. Therefore, the consequences of an accident previously evaluated in the UFSAR are not increased.

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The only occurrences of malfunction of equipment important to safety previously evaluated in the UFSAR, related to the proposed activity, are the failure of one train of SDC system and the inadvertent injection via charging pumps causing boron dilution. The probability of malfunction of these equipment is not increased by the activity proposed. Additionally, the proposed activity has no effect on any other equipment or system configuration, nor does modes.

it create any new failure Therefore, the probability of occurrence of any equipment malfunction important to safety previously evaluated in the UFSAR will not increase.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated .in the UFSAR?

46

The malfunction of equipment important to safety previously evaluated in the UFSAR is not changed by the proposed activity. The consequences of reduced flow on the decay heat removal requirements and the boron dilution event have been shown in this evaluation not to increase. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR are not increased.

Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The only change due to the proposed activity is the operation of SDC with single point injection. This configuration has been evaluated and shown to meet all the T.S. bases requirements. There are no new systems or system interactions involved important to safety as described in the UFSAR.

Therefore, the proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

Does the proposed activity create the possibility of a malfunction of equipment. important to safety of a different type than any previously evaluated in the UFSAR?

The proposed activity is related to the SDC system configuration. The malfunction of the SDC system has been evaluated and shown to meet the Technical Specifications and UFSAR bases requirements. There are no changes to any other systems or equipment important to safety. Thus, the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR is not increased.

Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The bases for Technical Specification 3.4.1.4.1 and 3.4.1.4.2, decay heat removal, and boron concentration reduction and boron stratification. The decay heat removal requirements have been shown to require that sufficient decay heat removal capacity is available, and sufficient coolant circulation is maintained to minimize the effects of a boron dilution incident and prevent boron stratification. These requirements have been shown to be satisfied under. the conditions specified in this evaluation. Therefore, the proposed activity does not reduce the margin of safety as defined in the bases for the Technical Specifications.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

48

Unit: ~

2 JPN-PSL-SEIS-94-005

Title:

~

Operation of the Wide Range Containment Level Monitoring Channels L-07-13A & L-07-13B With Inoperable Sensors This Safety Evaluation allows the continued use of the containment level monitoring instruments channels L-07-13A and L-07-13B by providing a means of compensating for a failed sensor in each channel. This evaluation is necessary as sensor 9 of LE-07-13B is no longer operational. Additionally, a non-conformance report identified that sensor ll of LE-07-13A had failed in November of 1990. An engineering evaluation evaluated sensor 11 and determined that operation with one failed sensor is acceptable. Sensor 11 (LE-07-13A) has since been replaced. Therefore, this evaluation will allow the proper operation of channels L-07-13A and L-07-13B with one failed sensor in one or both channels. This evaluation will also allow continued operation in the future by considering the containment level monitoring instruments channels L-07-13A and L<<07-13B operational with one failed sensor in one or both channels while waiting for replacement parts.

This evaluation documents the acceptability of the level transmitter's circuitry alteration. The modification will not adversely affect the operation or the existing qualification of the containment level system.

~

This temporary alteration will have no impact on plant safety or

~

operation. This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an 1~ Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of.,an accident previously evaluated in the UFSAR has not been increased. The failure of this system is not considered an initiating event in any accident scenario. The wide range containment water level monitoring loops are utilized solely for post accident monitoring purposes.

49

C

2. Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased by this reconfiguration. The wide range containment level loops will continue to monitor the water level in containment during an analyzed accident.

Each channel will continue to provide post accident monitoring capabilities with the exception of decreased resolution between a failed sensor and the next highest sensor. The lack of indication at the failed sensor will not adversely impact any operator actions associated with accident mitigation as no actions or decision points are anticipated to occur based solely on containment water level.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased with this circuitry reconfiguration. The channel still provides monitoring of containment water level durihg an analyzed accident. The design/procurement requirements imposed on the components to be installed are equivalent to that of the level monitors. No new system interactions are being introduced, only a resistor is being added in the level transmitter. This condition does not result in an increase in probability of a malfunction.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased. Because the channel will continue to monitor post accident containment water level as described in 2. above, the consequences of a malfunction have not been changed.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The possibility of an accident of .a different type than any evaluated previously in the UFSAR has not been created. The modified instrument loop provides only monitoring capability of wide range containment level during an analyzed accident and will operate as described in item 2. above.

50

t 6. Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any evaluated previously in the UFSAR?

The possibility of a malfunction of a different type than any evaluated previously in the UFSAR has not been created as this condition does not introduce any new failure modes to the post accident containment level monitoring system.

E

7. Does the proposed activity reduce the margin of safety as defined in the basis for any technical specifications?

The proposed activity does not reduce the margin of safety as defined in the basis for any technical specifications as the reconfigured channel will continue to provide the necessary monitoring function of post accident containment water level as required by the Plant Technical Specifications.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

51

Unit: JPN-PSL-SENP-94-005

Title:

Shutdown Cooling Suction Valve Interlock Design ABSTRACT t

This safety evaluation provides justification for use of an alternative design in the shutdown cooling system (SDCS) of St Lucie Unit 1. The current licensed design for shutdown cooling per the UFSAR includes diverse, redundant and independent pressure sensors feeding the suction side isolation valve permissive interlocks. The proposed design provides for redundant and independent. pressure sensors, however, the requirement for diversity is eliminated since safety.

it does not improve reliability or This safety evaluation also demonstrates that safe plant operation was not affected by the elimination of diversity in the operation of pressure sensors since their replacement in 1984. This conclusion is substantiated by both evaluation and analysis which shows that neither safety nor reliability of the system is adversely affected. Note that PCM 001-182 upgraded this instrumentation to meet the requirements of 10 CFR 50.49 (i.e.,

environmental qualification of electrical equipment).

An UFSAR change package is provided in this evaluation to affect the update to the design features for shutdown cooling. Since the design of a safety related system is being changed, this safety evaluation is classified as safety. related.

This design change neither involves unreviewed safety questions nor requires 'changes to Technical Specifications. Therefore, this design change and revision to the UFSAR may be performed without prior NRC approval. Note that there are no configuration changes to the plant required by this safety evaluation.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the following evaluation serves to determine whether use of an alternative design in the shutdown cooling system of St Lucie Unit 1 constitutes an unreviewed safety question:

Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The pressure transmitters PT-1103 and PT-1104 are used for the open permissive interlock (OPI) function of the SDCS suction valves. Although not accident initiators, the pressurizer pressure instrumentation was evaluated for the effect on the 52

OPI and determined that the increase in frequency for a ISLOCA (ISLOCA frequency is governed by a catastrophic failure of both series isolation valves) due to elimination of diversity betveen the transmitters is negligible. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR is not increased.

Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR are not increased because the elimination of diversity in the pressure sensing instrumentation does not affect the ability of the SDCS to perform its design basis function. The OPI function is unaffected by the elimination of diversity in the instrumentation. The OPI function neither degrades nor prevents. actions used to mitigate UFSAR accidents. Therefore, the removal of diversity in the pressure sensing instrumentation does not increase the consequences of an accident previously evaluated in the UFSAR.

Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

Since reliability remains essentially unchanged as a result of elimination of diversity in the pressure sensing instrumentation, there is no increase in the probability of an occurrence of a malfunction of equipment. Therefore, the probability of occurrence of equipment malfunction important to safety previously evaluated in the UFSAR is not increased.

Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The elimination of diversity in the pressure sensing instrumentation does not affect the ability of the SDCS to perform its design basis function. The OPI function is unaffected by the elimination of diversity in the instrumentation. The OPI function neither degrades nor prevents action used to mitigate consequences of UFSAR accidents. Therefore, the removal of diversity in the pressure sensing instrumentation does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR.

Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

53

The ISLOCA, as the, postulated accident involved with the elimination of diversity in the pressure sensing instrumentation for the SDCS suction valves is an analyzed event in the UFSAR. Therefore, new types of accidents are not created that are different from any already evaluated in the UFSAR.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

Reliability remains essentially unchanged as a result of elimination of diversity in the pressure sensing instrumentation. Therefore, the elimination'f diversity of the pressure sensing instrumentation does not increase the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The setpoints for the SDCS OPI,'SITS, and PORVs - selected to the low temperature mode of operation are unaffected by the removal of diversity in the pressurizer pressure transmitters, and therefore, the proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

54

Unit:~

1 JPN-PSL-SEMS-94-008

Title:

~

Gasket Leak Repair for Shutdown Cooling Return Isolation Valve V3480 This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely'ffect plant operation or safety.

This evaluation demonstrates the acceptability of the repair of a body-to-bonnet gasket leak on safety related valve V3480 through the use of a seal clamping device and sealant. The leaking valve gasket shall be replaced at the next forced outage of sufficient duration or the next refueling outage, therefore, this evaluation is applicable to operating cycle 12 only.

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an

1. Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased. This repair method restores the integrity of the valve body to bonnet flange gasket, and will not hinder the operation of any equipment capable of initiating an accident. The use of this repair method will not overstress or adversely affect valve V3480 or the 1A shutdown cooling return piping. The functionality of any equipment important to safety is not affected by this repair method since. the seal clamp performs an identical function as the leaking gasket; the bolt loadings and pressure boundary materials are not adversely affected by the injection of sealant. Also, contaminants from the sealant will not leach into the reactor coolant system medium in 55

sufficient quantities to impact any other wetted materials as described in this document.

2~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased 'y the application of this repair activity. This repair will not adversely affect any equipment which is required for accident mitigation or safe shutdown. This repair activity creates no new paths for the uncontrolled release of radioactivity in the event of a postulated accident and does not adversely affect any radiation monitoring equipment.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased. The application of this seal clamp and sealant will restore the integrity of the valve body to bonnet gasket. The functionality of any equipment important to safety is not affected by this repair activity.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased. The consequences of the failure of the injection seal is the same as the failure of the gasket, which would result in a loss of system fluid. No new unevaluated system leakage paths or possible paths for an uncontrolled radioactive release are created by the implementation of these repair methods. These repairs will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function, as detailed in this document.

t 5~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The possibility of an accident of a different type than any evaluated previously in the safety analysis report has not been created. The proposed repair activity does not provide a new method of normal or emergency component or system 56

operation and does not provide any new component failure modes. Zn addition, no new plant hardware other than the seal clamp and fittings, as previously described, is added by this repair activity. Thus, no new accident initiators are introduced through these repairs.

6. Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the UFSAR? "

possibility of a malfunction of a different type than any

'he evaluated previously in the safety analysis report has not been created. The failure of valve V3480 to performed has been previously evaluated in the UFSAR. Additionally, leakage of sealant into the RCS is prevented by procedural compliance, the use of thermosetting sealant compounds, and the method and location of injection. Also, contaminants from the sealant will not leach into the affected system medium in quantities sufficient to impact any other wetted materials and the sealant chemistry is compatible with the affected system chemistry requirements. 'I

7. Does the proposed activity reduce the margin of safety as defined in the basis for any technical specification?

This repair activity does not reduce the margin of safety as defined in the basis for any technical specifications.

Chemistry limits are not altered and no other change is proposed to the plant design, modes of operation or assumptions in the bases for the Technical Specifications or Safety Analysis.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

57

Unit: JPN-PSL-SENS-94-010

Title:

Evaluation for Alternate ECCS Valve Alignment to Repair Line,3/4-SI-121 St Lucie Unit 2 was in Mode 3 during a plant return,to power startup sequence following the cycle 8 refueling outage of an approximately two month duration. A non-conformance report was written to document leakage on an instrument line off of the 2B1 Emergency Core Cooling System,(ECCS) injection line. Repair of this line will require the draining of adjacent piping, rendering the 2B1 injection line inoperable. The drained portion of piping is on the upstream side of the 2B1 injection line check valve (isolation to the RCS) back to the ECCS pumps'A & B HPSI and B LPSI) 2Bl loop isolation valves and the 2B1 Safety Injection Tank (SIT) isolation valve. The loop 2B1 ECCS injection isolation valves will be closed with control power removed (valves V3634, HCV-3635, HCV-3636 & HCV-3637) to prevent flow through this line.

This evaluation restricts plant operation to Mode 3 with pressurizer pressure below 1750 psia or Mode 4. Operation in Mode 5 with the 2B1 injection loop isolated is addressed in a safety evaluation.

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an 1 ~ Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

k The proposed activity does not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

This activity provides an alternate valve alignment for repair of a line in the 2B1 ECCS injection line. The valve alignment is acceptable for the plant mode of operation identified in this safety evaluation. There are no components being installed by this evaluation and no interactions with any components that would be considered as accident initiators.

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I The repair does not introduce any condition considered as accident initiators. This alignment is required to effect repair and return the ECCS loop to its original design. As such, the probability of occurrence of an accident previously evaluated in the UFSAR is not increased.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased by this alternate valve alignment or repair. Valve V3237 provides an RCS pressure boundary function. The ECCS injection function is maintained for the mode of operation in which the line repair is effected.'ince the ECCS function mitigates the consequences of accidents, the proposed activity does not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased. The alternate valve alignment maintains pressure boundary to allow repair of the line. There are no components being installed by this evaluation and no interactions with any components important to safety. The ECCS function is maintained acceptable by the proposed valve alignment. Only the ECCS injection line affected by the repair is inoperable. The remaining ECCS injection lines are unaffected and meet the technical specification requirements for ECCS operability. Therefore, the probability of occurrence of a. malfunction of equipment important to safety is not increased.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased. Since the ECCS function mitigates the consequences and remains unaffected by the alternate valve alignment to effect repair of the line, the consequences of a malfunction of equipment important to safety has not increased.

5. Does the proposed activity create 'the possibility of an accident of a different type than any previously evaluated in the UFSAR?

59

r The possibility of an accident of a different type than any evaluated previously in the safety analysis report has not been created. There are no components being installed by this evaluation and no interactions with any components important to safety. Thus, no new accident initiators are introduced through this repair.

6. Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the UFSAR?

The possibility of a malfunction of a different type than any evaluated previously in the safety analysis report has not been created. There are no components being installed by this evaluation and no adverse interactions with any components important to safety. Thus, no new accident initiators are introduced through the alternate valve alignment and line repair.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any technical specification?

The inoperable ECCS injection line as a result of the alternate valve alignment is acceptable for the plant mode of operation in which the line repair is performed. There are no changes proposed to the plant design, modes of operation or assumptions in the basis for the Technical Specifications or Safety Analysis. Therefore, the alternate valve alignment to effect the line repair does not reduce the margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not. adversely affect plant operation or safety.

60

Unit: JPN-PSL-SEMS-94-011

Title:

Pressurizer Spray Bypass Valve V1454 Needle Tip Failure This 10CFR50.59 evaluation addresses the acceptability of V1454 for USE AS XS with a captured broken disc. V1454 is a 3/4" manual flow control valve (needle valve) which maintains minimal bypass flow around pressurizer spray valve PCV-1100F. NCR 178-293-3037M identified the tip of the needle disc to be broken off and wedged into the valve body seat. Multiple attempts to remove the broken tip through the valve bonnet have failed due to the tip being wedged tightly into the seat. V1454 is a Quality Group A valve, designed in accordance with ASME B&PV Code Section III, Class 1 requirements, and is located within the Reactor, Coolant System pressure boundary. Therefore, this evaluation is classified as Safety Related.

V1454 has been reassembled with the original disc/stem assembly and the broken tip in the valve body seat. V1454 will be left in the closed position, under a plant operational clearance, to assure the tip is captured between the valve seat and the upper disc/stem assembly. Maintaining the valve closed provides assurance the tip will not become free to move about the Reactor Coolant System as a loose part.

There are two pressurizer spray valves, PCV-1100E & F, and each spray valve has an associated spray bypass valve; V1453 and V1454, respectively. V1454 and V1453 were replaced during the current refueling outage with Velan valves to replace the obsolete AiResearch valves.

The spray valves and spray bypass valves are designed to allow a small continuous flow, diverted from the RCS cold leg loop Bl and B2, to bypass the normally shut spray valves. The continuous flow maintains the spray lines and pressurizer surge line warm relative to the pressurizer in order to prevent thermal shock during plant transients requiring pressurizer spray.

V1454 shall be replaced or repaired during the next forced outage of sufficient duration or the next refueling outage; therefore, this evaluation is applicable to operating cycle 8 only.

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION below listed In accordance with 10 CFR 50.59, the responses to theconstitutes questions serve to determine if the subject activity an unreviewed safety question:

Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed activity does not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

This evaluation provides for plant operation with valve V1454, a normally throttled open valve, to operate in a closed position. The function of V1454 is to provide for a small amount of RCS flow through the spray lines (bypassed around spray valve PCV-1100F) to protect the spray and surge lines from the effects of a thermal shock during plant transients.

Sufficient continuous flow exist to minimize the differential temperature between the spray line and the pressurizer.

Operation of PCV-1100F is not impacted by the position of V1454, thus the probability of occurrence of a pressure transient requiring mitigation by the code safety valves is not increased. A failure modes and effects analysis has concluded that there is no credible failure mode of this configuration that could lead to the initiation of an analyzed accident.

2~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an. accident previously evaluated in the UFSAR have not been increased by this valve alignment. Valve V1454 is part of the pressurizer spray system which is relied upon for control of pressurizer pressure during normal plant transients. The pressurizer spray system is not relied upon for accident mitigation. Overpressure protection is provided by the pressurizer code safety valves. There is no impact to the operation and performance of any accident mitigating equipment.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a. malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased. Valve V1454 will be maintained in the closed position. There are no components being installed by this evaluation and no component interactions being introduced. A failure modes and effects 62

analysis has concluded that there is no credible failure which could result in the failed disc of V1454 traveling downstream to the spray line or any other downstream components.

Sufficient continuous flow exists to minimize the differential temperature between the spray line and the pressurizer.

Operation of spray valve PCV-1100F is not impacted.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important "to safety previously evaluated'in the UFSAR?

The consequences of a malfunction of equipment important to safety previously 'valuated in the UFSAR have not been increased.. Valve V1454 is part of the piessurizer spray system which is relied upon for control of pressurizer pressure during normal plant transients. The pressurizer spray system is not relied upon for accident mitigation.

There is no impact to the operation or performance of any equipment relied upon for accident mitigation.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The possibility of an accident of a different type than any evaluated previously in the safety analysis report has not been created. There are no components being installed by this evaluation and no new interactions being introduced.

Pressurizer pressure control via PCV-1100E & F will not be affected. A failure modes and effects analysis has concluded that there is no credible failure which could create an accident of a different type than any previously evaluated in the UFSAR.

6. Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the UFSAR?

The possibility of a malfunction of a different type than any evaluated previously in the safety analysis report has not been created. There are no components being installed by this evaluation and no adverse interactions with any components important to safety. A failure modes and effects analysis has concluded that there is no credible failure which could result in the failed disc of V1454 traveling downstream to the spray line or any other downstream. component. Sufficient continuous flow exists to minimize the differential temperature between the spray line and the pressurizer.

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P

7. Does the proposed activity reduce the margin of safety as defined in the basis for any technical specification?

The margin of safety as defined in the basis for any Technical Specification has not been reduced. There are no Technical Specification Limiting Conditions of Operation or Surveillance Requirements related to the pressurizer spray system.

Design Features specification, Component Cyclic or Transient Limits, identifies cyclic or transient limits for the pressurizer spray system and provides a method for calculating the pressurizer spray nozzle cumulative usage factor. This specification requires a lifetime accounting of the number of spray cycles where the temperature difference between the pressurizer water and spray water is greater than 200'F.

Operation with V1454 in the closed position does not impact the use of this table. The cumulative usage factor calculation and the inherent safety margins of the cyclic and transient limits remain unchanged.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

64

Unit: JPN-PSL-SEMS-94-013

Title:

Freeze Seal Application for V3480 on the 1A Shutdown Cooling Return Line The purpose of this Safety Evaluation is to evaluate the safety significance of applying a freeze seal to the reactor coolant system 1A hot leg shutdown cooling return line while refueling operations are in progress. This activity will render the "A" shutdown cooling loop out of service, and therefore inoperable.

Since the freeze seal is classified as not nuclear safety (NNS),

the two concerns to be evaluated are 1) the loss of the remaining operable shutdown cooling loop due to either an active or passive failure, and 2) the loss of freeze seal integrity.

V3480 is the primary motor operated valve for the loop 1A shutdown cooling return line. The valve and piping are {}uality Group A, Seismic Class I, and designed to the Class 1 requirements of USAS B31.7. The valve is not isolatable from the 1A hot leg and is the first of two isolation valves in series that isolates the reactor coolant system (RCS) from the low-pressure shutdown cooling system.

V3480 is also relied upon to maintain the pressure boundary when the RCS is open to the refueling pool during refueling operations.

The purpose of the freeze seal is to maintain this pressure boundary function during refueling operations so that V3480 may be repaired or replaced. This activity is consistent with the use of other non-permanent equipment used during St. Lucie outages in similar operating conditions (i.e., the reactor cavity seal ring, the steam generator nozzle dams, and the RCS hot and cold leg plugs).

This evaluation establishes that the application of a freeze seal on the 1A shutdown cooling return line does not present a nuclear safety concern since its failure would not cause a loss of the inservice shutdown cooling loop or adversely impact refueling operations. Although the freeze seal is an NNS device, the St.

Lucie freeze seal application procedure (GMP-10) provides adequate measures to ensure a low probability for a freeze seal failure.

This evaluation also states additional requirements to provide added assurance that the likelihood of uncontrolled refueling pool leakage remains improbable.

Revision 1 is issued to specify the requirements for maintaining the RCS pressure boundary intact and to specify the lowest elevation at which the pipe may be cut. The conclusions of the safety evaluation remain unchanged. Revision 2 is issued to clarify requirements for the upper guide structure (UGS) concurrent with the freeze seal application, clarify ECCS lift 65

~ I parameters, and revise the evaluation to reference the new plant procedure on freeze seal application (GMP-10). The conclusions of the safety evaluation remain unchanged.

Revision 3 is issued to address contingency plans forwhich the radiographic access port in the pup piece upstream of V3480, was added per PCM 082-194M, Rev. 1. The conclusions of the safety evaluation remain unchanged.

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and .does not adversely affect plant operation or safety.

SAFETY EVALUATION

/

In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine if the subject activity constitutes an unreviewed safety question:

1~ Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased since the proposed activity does not adversely affect any accident-initiating components. This evaluation establishes that a loss of reactor coolant system integrity due to a catastrophic freeze seal failure is not considered a credible event. The evaluation establishes that the freeze seal is a reliable pressure boundary for the given plant conditions and the actions required section of this evaluation provides required actions to minimize the likelihood of uncontrolled leakage.

2~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

i The consequences of an accident previously evaluated in the UFSAR have not been increased since the proposed activity does not adversely affect any equipment which is required for accident mitigation. Since the reactor head will be removed, a postulated freeze seal failure would not create a new path for uncontrolled radioactive releases and would not adversely affect any radiation monitoring equipment. Also, a freeze seal failure would not affect the capability of V3481 to function as a containment isolation valve while V3480 is either open or removed from the system, nor would it affect the operation of the inservice shutdown cooling loop because 66

a freeze seal failure would not result in the draining of the reactor coolant system hot legs.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased. Although the freeze seal is not considered a safety related device, the freeze seal is a reliable pressure boundary for the plant conditions. Thus, no new unmitigated failure modes for any equipmen't important to safety are introduced by the proposed activity and no new components or equipment are introduced that could adversely interact with any equipment important to safety.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety. previously evaluated in the UFSAR?

The consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased since the proposed activity does not create a new path for uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a,containment isolation function.

Furthermore, the freeze seal will not adversely impact any equipment which is required to perform an active safety related function or to initiate actuation of any safety systems. Even in the unlikely event of a catastrophic freeze seal failure, the operability of the operating shutdown cooling train- would not be impacted.

5~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The possibility of an accident of a different type than any evaluated previously in the UFSAR has not been created since the proposed activity does not add or adversely affect any equipment that could reasonably be capable of initiating an accident. The installation or postulated failure of the freeze seal would not present any credible new paths for the loss of refueling pool and reactor coolant system inventory while irradiated fuel will be transported above the reactor vessel. This conclusion is based on an evaluation of the times required for fuel pool drainage and the fuel transit time above the reactor vessel, as well as the improbability of a catastrophic freeze seal failure. A freeze seal is a 67

I Unit: 1 & 2 JPN-PSL-SENS-94-015

Title:

Safety Evaluation for Service Water System Modifications ABSTRACT The Service Water (SW) system for St. Lucie Units 1 & 2 (sometimes referred to as the potable and sanitary water system) has been identified as a system requiring increased maintenance activities.

A review of system functions and interactions has concluded that SW system piping and components located outside of the reactor auxiliary buildings, fuel handling buildings, diesel generator buildings and component cooling water areas are classified as Not Nuclear Safety, have no interactions with equipment important to safety and are not required to be considered within the scope of the FPL Quality Assurance Program. As such, modifications to those portions of the SW system are planned to be performed outside of the Plant Change/Modification (PC/M) process. Marked-up drawings reflecting the configuration of the system will be provided by the plant to engineering for review and as-building after modifications are complete.

This evaluation does not apply to the entire SW system. An attachment to the evaluation provides a detailed description of those portions of the SW system to which this evaluation applies.

The SW system is a part of System Number 15, which consists of the Fire Protection, Demineralized Water, Service Water and Primary Makeup Water systems. This evaluation does not include any portions of the Fire Protection, Demineralized Water and Primary Makeup Water systems (typically indicated on plant piping and instrumentation drawings as FP, DW & PMW).

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has 'not been increased. The SW system is not assumed to cause any analyzed accident in the UFSAR.

69

0 r ~

reliable pressure boundary for the plant conditions and this evaluation provides required actions to minimize the likelihood of uncontrolled leakage. Furthermore, Off-Normal Operating Procedure 1-0120031 (Excessive Reactor Coolant System Leakage) provides the required actions for the scenario in which a loss of inventory occurs while refueling operations are in progress.

6 Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the UFSAR?

The possibility of a malfunction of a different type than. any evaluated previously in the UFSAR has not been created since the proposed activity does not add or adversely affect any equipment that has a reasonable possibility of malfunctioning.

The installation or postulated failure of the freeze seal does not present any credible new paths for the loss of refueling pool and reactor coolant system inventory while refueling operations are in progress. Also, the freeze seal will not inhibit or otherwise adversely affect the operation of any equipment important to safety.

70 Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed activity does not reduce the margin of safety as defined in the basis for any. Technical Specification since the proposed activity is intended to maintain the RCS pressure boundary to ensure that the refueling cavity water level is maintained in accordance with the Technical Specifications.

Technical Specifications 3.9.8.1 and 3.9.10 both require a minimum of 23 feet of water above the irradiated fuel assemblies when only one shutdown cooling loop is operable or when core alterations are in progress. This evaluation establishes that the likelihood of uncontrolled leakage as a result of a freeze seal failure is improbable; therefore, the proposed activity does not reduce any Technical Specification margins of safety.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does,not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

68

The equipment within the SW .system boundaries does not initiate design basis accidents. Furthermore, any modifications made to the SW system as a result of this evaluation could not initiate a design basis accident. The boundaries of the SW system within the scope of this evaluation have been selected to ensure that there is no possible interaction with equipment important to safety.

Piping located within the Reactor Auxiliary Buildings, Diesel Generator Buildings and Fuel Handling Buildings has been excluded from this evaluation.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences. of an accident previously evaluated in the UFSAR have not been increased. The SW system provides water for various non-safety uses throughout the site such as washdown stations, emergency eyewash stations, human consumption, sinks, toilets and numerous hose connections.

The SW system is not relied upon in any way to provide for accident mitigation. Modifications to the SW system that are in accordance with the guidance in this safety evaluation will not affect any systems relied upon for accident mitigation.

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased. The SW system does not contain nuclear safety related equipment nor does it interact with any equipment important to safety. Modifications to the SW system that are in accordance with the guidance. contained in this safety evaluation will not increase the challenges to or the likelihood of failure of equipment important to safety.

4 Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR. The SW system, classified as Not Nuclear Safety, does not interface. with any system required for accident mitigation. This evaluation allows modifications to those portions of the SW system that are not in the vicinity of any accident mitigating equipment (i.e., potential system interactions with equipment important to safety are 70

avoided). Thus, the consequences of any failure or malfunction of equipment important to safety are'not changed.

1 5 ~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the UPSAR. The SW system is a non-safety system that does not interface with safety related equipment. This evaluation is limited to those portions of the SW system that are not in the vicinity of safety related equipment; 'therefore, potential interactions with safety related equipment are not possible.

There are no credible failure modes associated with potential modifications to the SW system (including flooding and physical interactions) that could create the possibility of a nuclear accident different than any previously evaluated.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than any. previously evaluated in the UFSAR. The SW system is a non-safety system that does not interface with safety related equipment. As noted above, there are no credible failure modes associated with potential modifications to the SW system (including flooding and physical interactions) that could impact safety related equipment.

7 ~ Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The SW system is not discussed in any Technical Specification or Technical Specification basis. Modifications to the SW system that are in accordance with the guidance provided in this safety evaluation can not adversely affect the margin of safety to any Technical Specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the -subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

71

evaluated. The proposed modification is bounded by the UFSAR.

Therefore, the proposed activity does not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important, to safety previously evaluated in the UFSAR?

The proposed modification is previously evaluated in the UFSAR. Failure of the SBCS does not result in the malfunction of any safety related equipment. Therefore, the proposed activity does not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The SBCS does not affect any equipment required to mitigate the consequences of an accident, nor does it affect any other equipment important to safety. The proposed modification is bounded by the UFSAR. Therefore, the proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR.

5~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed modification does not adversely interact with any components important to safety. No new accident initiators are introduced through the proposed modification.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed modification does not adversely interact with any components important to safety. The proposed modification will disable the automatic quick open feature of the SBCS for, load reduction. Loss of this automatic feature is evaluated in the UFSAR. Therefore, the proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR.

7 Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

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There are no limiting conditions of operation associated with the SBCS. The proposed activity does not change assumptions in the basis for the Technical Specifications. Acceptance in the SER is based on an evaluation of the failure modes of this system. Therefore, the proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

74

Unit: 1 & 2 JPN-PSL-SENS-94-018

Title:

Safety Evaluation for Hypochlorite System Modifications The Hypochlorite (CL) system for St. Lucie Units 1 & 2 has been identified as a system requiring increased maintenance activity.

A review of system functions and interactions has concluded the CL system (as bounded in an attachment provided in the evaluation) is not required to be considered within the scope of the FPL {}uality Assurance Program since it is classified as Not Nuclear Safety and has no interactions with equipment important to safety. As such, modifications to the CL system are planned to be performed outside of the Plant Change/Modification (PC/M) process. Marked-up drawings reflecting the configuration of the system will be provided by the plant to engineering for as-building after modifications are complete. Because of interaction concerns, this evaluation does not include the individual injection lines located within the Unit 1 & 2 intake cooling water (ICW) pump bays. The boundaries for this evaluation are detailed in an attachment.

The CL system is a non-safety system common to St. Lucie Units 1 &

2 that produces a sodium hypochlorite solution via electrolytic decomposition of filtered seawater. The hypochlorite solution is periodically injected into the suction side of the intake cooling water (ICW) and circulating water (CW) pumps for the control of biological fouling.

This evaluation concludes that the activity as described above does not represent an ,unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased. The CL system is not assumed to cause any analyzed accident in the UFSAR.

Furthermore, any modifications made to the CL system as a result of this evaluation could not initiate a design basis 75

accident. The boundaries of the CL system within the scope of this evaluation have been selected to ensure that there is no possible interaction with equipment important to safety.

Piping located within the Unit 1 t 2 ICW pump intake bays has been excluded from this evaluation.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of'an accident previously evaluated in the UFSAR have not been increased. The CL system .injects a hypochlorite solution for the control of biofouling in the ICW and CW systems of both units. The CL system is not relied upon in any way to provide for accident mitigation. The ability of the Unit 1 & 2 ZCW systems to perform their safety function is not affected by the CL system. Although the CL system helps to maintain ICW system heat exchanger surfaces clean, the capability of the heat exchangers to perform their design function is not dependant on the CL system.

Modifications to the CL system, in accordance with the guidance and scope of this safety evaluation, will not affect any systems relied upon for accident mitigation.

3 ~ Does the proposed activity increase the'robability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

'he probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased. The CL system does not contain nuclear safety related equipment nor does it functionally interact with any equipment important to safety. Potential failures of the CL system would have no effect on the operation or reliability of equipment important to safety. Modifications to the CL system that are in accordance with the guidance contained in this safety evaluation will not increase the challenges to or the likelihood of failure of equipment important to safety.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR. The CL system is classified as Not Nuclear Safety and does not interface with any system required for accident mitigation. This evaluation allows modifications to those portions of the CL system that are not in the vicinity of any accident mitigating equipment (i.e., potential system interactions with equipment important to safety are 76

avoided). Thus, the consequences of any failure or malfunction of'equipment important to safety are not changed.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR. The CL system is a non-safety system that does not interface with safety related equipment. This evaluation is limited to those portions of the CL system that are not in the vicinity of safety related equipment; therefore, potential interactions with safety related equipment are not possible.

There are no credible failure modes associated with potential modifications to the CL system (including flooding and physical interactions) that could create the possibility of a nuclear accident different than any previously evaluated.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR. The CL system is a non-safety system that does not interface with safety related equipment. As noted above, there are no credible failure modes associated with potential modifications to the CL system (including flooding and physical interactions) that could impact safety related equipment.

7 ~ Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The CL system is not discussed in any Technical Specification or Technical Specification basis. Modifications to the CL system that are in accordance with the guidance provided in this safety evaluation can not adversely affect the margin of safety to any Technical Specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely .affect plant operation or safety.

77

Unit: 2 JPN-PSL-SENP-94-019

Title:

Alternative Valve Position for Spray Header Isolation Valve I-FCV-07-1B The purpose of this safety evaluation is to demonstrate that the containment spray headez isolation valve (I-FCV-07-1B) is capable of performing its design functions while aligned in the open position. I-FCV-07-1B is normally closed during power operation.

This change is expected to remain in 'effect until the next refueling outage.

Valve I-FCV-07-1B is zequired to open on a containment spray actuation signal for accident mitigation. The valve is air operated, normally closed. Its fail safe position is open either on loss of air or power. The valve also performs a containment isolation function for the B train located outside containment.

However, this valve is not required by technical specifications to close for containment isolation nor does isolation actuation signal.

it receive a containment This safety evaluation involves an assessment of the effects on a safety related system, and therefore, is classified as safety related.

This evaluation concludes that operation of the plant with I-FCV-07-1B maintained in the open position during power operation does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

Revision 1 extends the period that I-FCV-07-1B may be maintained in the open position until the next refueling outage. This revision also establishes valve lineup requirements for surveillance testing associated with the containment spray line while in this configuration. This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an 78

~ ~

Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

UFSAR includes analysis for the effects related to an inadvertent actuation of containment spray during normal plant operation leading to containment differential pressure. The assumptions and conclusions for this analysis remain unchanged by the proposed valve alignment. Although the valve remains open, the probability of occurrence for this event is not considered to have increased since compensatory measures are being implemented to reduce the 'probability of inadvertent spray actuation. Therefore, the proposed activity does not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

20 Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The proposed valve alignment does not increase the consequences of an accident previously evaluated in the UFSAR since the valve is in its fail safe position andTherefore, the system is capable of performing its intended functions.

of the proposed activity does not increase the consequences an accident previously evaluated in the UFSAR.

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed valve alignment does not adversely affect the function of equipment important to safety and the containment spray system is capable of performing its intended design functions. The probability of occurrence of a malfunction of equipment important to safety has been reduced by eliminating one failure mode (failure of valve I-FCV-07-1B to open).

Therefore, the proposed activity does not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

4~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed valve alignment does not adversely affect the containment spray system and its required function to mitigate the consequences of an accident. Equipment important to safety is not adversely affected by the valve alignment.

Therefore, the proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR.

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s. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed valve alignment to the containment spray system was determined not to adversely interact with any other components important to safety. This alignment does not prevent containment spray system from performing its intended functions. New accident initiators are not introduced through this alignment. Therefore, the proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

6. Does the proposed activity create the possibility. of a malfunction of equipment important to safety of a different type than any previously evaluated 'in the UFSAR?

New accident initiators are not introduced through the proposed valve alignment. The containment spray system design, function, and method of performing the function has neither changed nor created a new failure mode. This alignment is bounded by the UFSAR. Therefore, this activity does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed activity does not change assumptions in the basis for any of the Technical Specifications. Therefore, the proposed activity does not reduce the. margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

80

Unit: 1 JPN-PSL-SEFJ-94-021

Title:

RTD Response Time Limit Increase From 8.0 Seconds to 14.0 Seconds The evaluation performed here supports an increase in the maximum allowable Resistance Temperature Detector (RTD) response time for St. Lucie Unit 1 from 8 seconds to 14 seconds. St. Lucie Unit 1 uses RTDs which are Weed and Rosemount types. During the Loop Current Step Response (LCSR) testing of the RTDs in the past few years, the response times of these RTDs have been in the range of 3 to 8 seconds. The Weed RTDs have averaged response times less than 5 seconds, whereas the Rosemount RTDs have typically shown response times between 6 and 8 seconds.

Due to difficulties associated with the removal and installation of RTDs, the RTD response time has been of interest to Florida Power

& Light (FPL) Company and to the nuclear industry in general.

Relaxation of the RTD response time will allow greater operational flexibility, that would prevent a sound RTD from being replaced when it is otherwise acceptable for use.

An evaluation is conducted to assess the impact of increased RTD response time limit on the safety analysis. The Basis to Technical Specification (TS) for the Thermal Margin/Low Pressure (TM/LP) trip specifies an allowance of 30 psia to compensate for the associated time delays. The pressure bias factor of 30 psia bounds the present RTD delay. time of 8 seconds. This bias term has been re-evaluated for the RTD response time of 14 seconds. The new bias term, calculated to be 42 psia, has been accounted for in the current safety analyses which, therefore, remain unaffected.

It should be noted that a similar change to increase the RTD response time to 14 seconds was approved by the NRC in 1991 for St.

Lucie Unit 2 . Relocation of tables of instrument response time limits to the Final Safety Analysis Report (UFSAR) has recently been approved by the NRC for the St. Lucie Units 1 and 2, in response to FPL's request for a license amendment. The proposed change to the RTD response time limit is in accordance with 10 CFR 50.59 and Enclosure 1 to Generic Letter (GL) 93-08 ).

It has also been shown by this safety evaluation .that this activity neither constitutes an unreviewed safety question nor requires changes to the Technical Specifications. Therefore, prior NRC approval for implementation of these changes is not required.

81

t SAFETY EVALUATION In accordance with 10 CFR Part 50.59, the following evaluation serves to determine whether this modification constitutes an unreviewed safety question:

Does the proposed change increase the probability of occurrence of an accident previously evaluated in the UFSAR?

(

The proposed change affects only the slow CEA Withdrawal events analyzed in the UFSAR. The initiation of these events is independent of the RTD response time. The probability of occurrence of such events is, thus, unaffected by the proposed activity. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR is not increased.

2~ Does the proposed change increase the consequences of an accident previously evaluated in the UFSAR?

The proposed change will allow longer RTD response time, which will affect the pressure bias term used in determining the TM/LP margin. The new higher pressure bias term is used in the current slow CEA Withdrawal events to compensate for the increased RTD delay time. Thus the consequences of these events remain unchanged. Therefore, the consequences of an accident previously evaluated in the UFSAR are not increased.

O. Does the proposed change increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed change allows an increase in the response time limit (from 8 seconds to 14 seconds) to be met during surveillance testing of the RTDs. This change has no effect on the malfunction of any equipment or system important to safety as evaluated in the UFSAR, nor does it create any new failure modes. Therefore, the probability of occurrence of any equipment malfunction important to safety previously evaluated in the UFSAR will not increase.

4 Does the proposed change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed change does not adversely affect the performance of any safety related equipment.'herefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR are not increased.

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5. Does the proposed change create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed change is an increase in the RTD response time limit and a compensating change in the pressure bias Noterm.

There are no changes to the plant configuration. new systems or system interactions are involved that adversely affect equipment or systems important to safety. Therefore, the proposed change does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

6. Does the proposed change create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The change proposed allows longer RTD response times and does not adversely affect any safety related equipment.

Additionally, there are no changes to any system configuration or equipment important to safety. Thus, the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR is not increased.

7~ Does the proposed change reduce the margin of safety as defined in the basis for any Technical Specification?

The Basis to Technical Specification for TM/LP trip specifies an allowance of 30 psia to account for time delays. The value for this pressure allowance changes to 42 psia corresponding to the proposed RTD response time .of 14.0 seconds. The increase in the pressure bias term from 30 psia to 42 psia, used in the current safety analysis, compensates for the effects of proposed increase in the RTD response time. The margin of safety is, thus, not reduced. The NRC has approved the relocation of the response time tables from the TS to the UFSAR. Therefore, the proposed change does not reduce the margin of safety as defined in the bases for the Technical Specifications.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely. affect plant operation or safety.

83

Unit: JPN-PSL-SENP-94-021

Title:

Removing the Automatic Control Function for I-TCV-14-4B ABSTRACT During the performance of In-Service Testing stroke time testing, I-TCV-14-4B, a temperature control valve in the St. Lucie Unit 2 Intake Cooling Water (ICW) system failed to demonstrate repeatable stroke time results. Nonconformance Report (NCR) f2-612 was generated as a result of this event.

The purpose of this safety evaluation is to demonstrate that the 2B ICW system is capable of performing its intended safety function while I-TCV-14-4B is clamped in a predetermined condition. Valve I-TCV-14-4B is automatically controlled to maintain component cooling water (CCW) temperatures during power operation. Valve I-TCV-14-4B will be clamped such that it is in a locked position to provide sufficient flow through the 2B CCW heat exchanger during accident conditions.

Valve I-TCV-14-4B is required to be open during all operational modes. During Design Basis Accident conditions, I-TCV-14-4B opens further to permit increased ICW flow through the 2B CCW heat exchanger. I-TCV-14-4B performs the safety function of re-positioning from the Normal operating throttle position (Design Flow = 8,250 gpm) to Emergency throttle'osition (Design Flow =

14,500 gpm).

This safety evaluation provides an assessment of the effects on a safety related system, and therefore, is classified as safety related. This evaluation concludes that operation of the plant with I-TCV-14-4B maintained in the clamped open throttle position, during power operation, does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

evisio A clarification is provided to address changes in ICW system performance (i.e. fouling, pump performance changes, etc) before the next scheduled outage. This is addressed by addition of a curve of ICW accident flowrate requirements through a CCW heat exchanger vs ICW inlet temperatures, which can be used to demonstrate operability of the ICW system at accident flowrates below 14,500 gpm. This revision has no effect on the conclusions of the safety evaluation.

84

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine if the subject activity constitutes an unreviewed safety question:

Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of an accident occurring which will challenge the valve's safety function has not changed with the new configuration of the valve. With respect to the description and FMEA in the UFSAR for valve I-TCV-14-4B, the valve in the clamped position will increase the availability of flow to the CCW train. The clamp itself is designed to withstand a design basis seismic event. Therefore, the proposed activity does not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The proposed configuration does not adversely affect the safety function of the valve. One failure mode (ie, valve failure in the closed position) has been removed. With the valve in its clamped position, the consequences of providing one less train of cooling water is decreased. The clamp itself is designed to withstand a design basis seismic event.

Therefore, the proposed activity does not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed valve alignment does not change the valve's designed safety function. The valve is required to be open to admit flow through the CCW heat exchanger. The clamped throttle position ensures the valve would perform this function. The clamp itself is designed to Seismic Category I requirements and therefore maintains the position of the valve. The proposed activity does not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

85

4 ~

The proposed valve alignment does not affect the CCW system operation, performance or safety function, nor does the new condition affect the operation, performance or safety function of the intake cooling water system. Equipment and systems important to safety will function in the same manner in the new condition. The consequences of a malfunction in the new condition are equal to those of the previous condition.

Therefore, the proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR.

Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

New accident initiators are not introduced through this configuration. Therefore, the proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

New accident initiators are not introduced through the proposed valve alignment. Failure of the valve is currently considered in the UFSAR, the clamp design reduces the probability of failure as described in the UFSAR. This alignment is bounded by the UFSAR. Therefore, this activity does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR.

Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed activity does not change'ssumptions used as the basis for any of the Technical Specifications. Therefore, the proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.

The proposed configuration does not affect safe operation of the Intake Cooling Water System and the Component Cooling Water system. In addition, the proposed configuration does not constitute an unreviewed safety question and does not require a change to the Technical Specifications. Therefore, implementation of the proposed configuration does not require prior NRC approval.

86

I~

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely, affect plant operation or safety.

87

Unit: 1 JPN-PSL-SENS-94-025

Title:

Safety Evaluation for Fuel Handling Equipment UFSAR Discrepancies FPL Quality Assurance Department Audit No. QSL-OPS-94-24 identified several minor discrepancies between existing plant procedures and various parts of the St. Lucie Unit 1 UFSAR. The discrepancies noted in the QA audit all pertain to fuel handling equipment. This evaluation provides a review and resolution to several of the discrepancies and provides an UFSAR Change Package.

This safety evaluation demonstrates that the UFSAR changes provided in an UFSAR Change Package do not adversely affect plant safety, security or operation. It has also been shown by this safety evaluation that this activity neither constitutes an unreviewed safety question nor requires changes to the Technical Specifications. Therefore, prior NRC approval for implementation of these changes is not required.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an 1 ~ Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased. The Fuel Handling Accident, UFSAR section 15.4.3, is the only relevant analyzed accident. Neither the parking location of the spent fuel handling machine nor the use of the CEA handling tool can be postulated to result in an increase in the probability of occurrence of a fuel handling accident. The UFSAR changes pertaining to the use of a dummy fuel assembly and the PT of fuel handling 'rapples are considered editorial clarifications.

1 2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR'?

The consequences of an accident previously evaluated in the UFSAR have not been increased. The consequences of the UFSAR fuel handling accident are not affected by this evaluation 88

since the design and operation of relevant accident mitigation systems are not impacted in any way.

3 Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety

~

previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased. This evaluation analyzes the potential failure modes and effects'f the proposedinteractions changes and concludes that no new failure modes or system are introduced. The design of the spent fuel handling machine, including its ability to he seismically stable, has not been changed. The CEA handling tool has been engineered by the NSSS vendor for its intended application. The UFSAR changes pertaining to the use of a dummy fuel assembly and the PT of fuel handling grapples are considered editorial clarifications.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR. The changes provided via this evaluation are all associated with fuel handling equipment.

There is no impact to any UFSAR accident analysis assumptions or to the operation of any system required for accident mitigation.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR. This evaluation analyzes the potential failure modes and effects of the proposed changes and concludes that no new failure modes or system interactions are introduced.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR. As stated 89

above, there are no new failure modes or system interactions as a result of the changes provided via this evaluation.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

There is no impact'o any Technical Specification Limiting Condition for Operation, Surveillance or Bases as a result of this evaluation.

The foregoing constitutes the determination, per. 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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Unit: JPN-PSL-SEMS-94-028

Title:

Installation of a Blind Flange on the Inlet of Containment Purge Valve FCV-25-1 Containment isolation for penetration P-11 is normally accomplished by closing FCV-25-2 and FCV-25-3. FCV-25-3 has exhibited leakage during the LLRT. This temporary change will allow the installation of a blind flange on the inlet of containment purge supply valve FCV-25-1. This blind flange will act as a containment isolation device replacing FCV-25-3.

The containment isolation system provides the means to isolate such that fluid systems that pass through containment penetrationscontainment any radioactivity that may be released to the atmosphere following a postulated Design Basis Accident (DBA) is confined. As such this temporary alteration performs a safety related function and this evaluation and its associated modifications are considered to be safety related.

This change does not affect the function of the 48" containment purge supply system during plant-power operations as the system is not used in modes 1, 2, 3 and 4. It has also been shown by this safety evaluation that this activity neither constitutes an unreviewed safety question nor requires changes to the Technical Specifications. Therefore, prior NRC approval for implementation of these changes is not required.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine if the subject activity constitutes an unreviewed safety question:

Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased since the proposed activity does not adversely affect any accident-initiating components. This evaluation establishes that the installation of the blind flange replaces the function of a closed FCV-25-3. Additionally, this valve is not required to open for any safety related requirements. The analysis of effects on safety section of this evaluation establishes that the blind flange is a reliable pressure boundary for the given plant conditions.

Does the proposed activity increase the consequences of an accident previously evaluated in the. UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased since the proposed activity does not adversely affect any equipment which is required for accident mitigation. Since the blind flange is one of two isolation's for P-11, its failure would not create a new path for uncontrolled radioactive releases and would not adversely affect any radiation monitoring equipment. Also, the blind flange failure would not affect the capability of FCV-25-2 to function as a containment isolation valve.

3 0 Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased. The gasket and valve packing seal of FCV-25-1 may become a new leak path. An LLRT will be performed on this new boundary. The packing of FCV-25-1 or FCV-25-2 may have a sealant injected to prevent leakage. Thus, no new unmitigated failure modes for any equipment important to safety are introduced by the proposed activity and no new components or equipment are introduced that could adversely interact with any equipment important to safety.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased since the proposed activity does not create a new path for uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function.

Furthermore, the blind flange will not adversely impact any equipment which is required to perform an active safety related function or to initiate actuation of any safety systems. Potential gasket and packing leaks from FCV-25-1 or FCV-25-2 will be tested and prevented, as required, with the use of the sealant PRI-201N.

5~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

92

The possibility of an accident of a different type than any evaluated previously in the UFSAR has not been created since the proposed activity does not add or adversely affect any equipment that could reasonably be capable of initiating an accident. The installation or postulated failure of the blind flange would not present any credible new .paths for the loss of containment atmosphere following a DBA. The blind flange is a reliable pressure boundary for the plant conditions.

6 Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the UFSAR?

The possibility of a malfunction of a different type than any evaluated previously in the UFSAR has n'ot been created since the proposed activity does not add or adversely affect any equipment that has a reasonable possibility of malfunctioning.

The installation or postulated failure of the blind flange does not present any credible new paths, for the containment

,atmosphere following a DBA. Also, the blind flange will not inhibit or otherwise adversely affect the operation of any equipment important to safety. h

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed activity does not reduce'the margin of safety as defined in the basis for any Technical Specification since the proposed activity is intended to maintain the containment boundary in accordance with the Technical Specifications.

Technical Specification 3/4.6.1.1 provide the requirements to ensure containment integrity is maintained. This evaluation establishes that the likelihood of uncontrolled leakage as a result of a blind flange failure is improbable; therefore, the proposed activity does not reduce any Technical Specification margins of safety.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

93

Unit: 1 & 2 JPN-PSL-SEMS-94-029

Title:

Shutdown Operations Criteria for Reduced Inventory and Draining the Reactor Coolant System The purpose of this evaluation is to demonstrate the acceptability of shutdown operations given the following proposed changes:

A) The criteria for reduced inventory will now be defined as 3 feet below the reactor vessel flange.

B) The criteria for draining, the RCS after shutdown will now be limited by the time to incover the core.

These proposed changes will bring St. Lucie Plant more in-line will NRC and industry guidelines on shutdown operations and will provide more flexibility for refueling outages without compromising plant safety. Implementation of these changes effectively amends previous submittals to the NRC on shutdown operations, however, such changes are allowed under 10 CFR 50.59 as outlined in NRC correspondence on the same subject.

'This safety evaluation involves an assessment of changes to

~ ~

shutdown operations, ~

and therefore, is classified as safety related. ~

This evaluation concludes that the proposed changes to operation of

~ ~

the 'plant during shutdown neither involve an unreviewed safety question nor require a change to plant Technical Specifications, as defined in 10CFR50.59, and do not adversely affect plant operation or safety. Therefore, prior NRC approval is not required for implementation. t SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions criteria (A serve

& B) to determine for shutdown if the operations proposed changes to the constitutes an unreviewed safety question:

Do the proposed changes increase the probability of occurrence of an accident previously evaluated..in the UFSAR?'he probability of ,occurrence of an accident previously evaluated in the UFSAR has not been increased. The proposed changes do not affect any accidents discussed in the UFSAR.

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2 ~ Do the proposed changes increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased. This evaluation does not affect any of the accidents discussed in the UFSAR and therefore does not increase any of the consequences of the accidents discussed in the UFSAR.

3~ Do the proposed changes increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunct'ion of equipment important to safety previously evaluated in the UFSAR has not been increased by reducing the RCS water inventory to 3 feet below the reactor vessel flange because this water level is above the mid-nozzle of the hot leg. The probability of occurrence would not be affected until the RCS water level reached the mid-nozzle elevation. The change in the criteria for containment closure does not change any probability of occurrence of a malfunction of equipment previously evaluated in the UFSAR.

4 Do the proposed changes increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

Adopting the Reduced Inventory definition established by the NRC in Generic Letter 88-17 does not increase the consequences of a malfunction of equipment important to safety since the loss of shutdown cooling event was previously evaluated and the results of that evaluation remain valid. The proposed change in the containment closure criteria provides for containment closure prior to uncovering the core following a loss of shutdown cooling event. Therefore, the consequences from the previously evaluated loss of shutdown cooling event remain valid.

5. Do the proposed changes create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed change to the criteria for Reduced Inventory will not create a different type of accident than those found in the UFSAR. The proposed change is above mid-nozzle. The plant can operate safely in Mode 5 at mid-nozzle. The proposed change to the containment closure criteria provides for containment closure prior to uncovering the core.

95

Therefore, no different type of accident is created and the accident assumptions found in the UFSAR remain valid.

6. Do the proposed changes create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed change to the criteria for Reduced Inventory provides adequate margin to prevent the loss of shutdown cooling due to draining the RCS below mid-nozzle. The loss of shutdown cooling is evaluated in the UFSAR and this change does not alter the assumptions for that evaluation. The proposed change to the containment closure criteria does not affect equipment important to safety as analyzed in the UFSAR.

This criteria is based on the ability to provide containment closure prior to uncovering the core and fission product release. Therefore a different type of accident is not created by this change.

7. Do the proposed changes reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed changes do not affect any Technical Specification nor do they reduce any margins of safety defined by the Technical Specifications. The proposed changes incorporate the NRC accepted criteria and definitions.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

96

I Unit: JPN-PSL-SENP-94-037

Title:

SIT Discharge/Loop Check Valve Stroke Test This safety evaluation demonstrates the acceptability of performing a full stroke test of the Safety Injection Tank (SIT) discharge/loop check valves. The proposed test will be used to address NRC requirements for SIT check valves testing delineated in Generic Letter 89-04.

The test is to be performed during refueling with the reactor vessel head and upper internals removed, and the refueling cavity flooded. The test consists of establishing a reduced nitrogen pressure in the SIT to be used as a motive force to discharge a portion of the SIT inventory through the subject check valves. The test is initiated by opening the SIT discharge line motor operated valve and observing via acoustic and magnaflux monitoring that the check valves fully open.

Calculations have been performed to demonstrate that sufficient velocities can be achieved and maintained to fully open the subject check valves. This evaluation addresses the various potential adverse effects on the plant including those on the reactor internals and steam generator nozzle dams resulting from the proposed flow test.

Since this safety evaluation assesses the effects of the proposed test on safety related components, it is classified as safety related.

This evaluation demonstrates that the proposed test neither involves an unreviewed safety question nor requires a change to plant Technical Specifications, and does not adversely affect plant operation or safety. Therefore, it is concluded that the proposed test may be performed without prior NRC approval.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the following questions serve to determine whether the proposed test constitutes an unreviewed safety question:

Does the proposed test increase the probability of occurrence of an accident previously evaluated in the UFSAR?

Performance of this test will not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

Specifically, the test will not increase the probability of a 97

loss of shutdown cooling and will not cause an upset of fuel assemblies. The maximum fluid velocity that will occur in the vessel, as a result of this test, is on the order of 1/2 ft/sec, and therefore, is not capable of adversely impacting the fuel or CEAs.

2 ~ Does the proposed test increase the consequences of an accident previously evaluated in the UFSAR?

This test will be conducted in Mode 6 with the reactor vessel head removed. This test does not impact any of the assumptions for refueling accidents evaluated in the UFSAR and will not increase the consequences of any of these analyses.

3 ~ Does the proposed test increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The shutdown cooling system will be secured in the tested loop. There are no potential failure modes resulting =from this test that would adversely impact shutdown cooling equipment. The additional pressure that the nozzle dams may be subjected to as a result of this test have been analyzed and determined not to affect nuclear safety. Therefore, this test will not increase the probability of a malfunction of equipment important to safety.

4 ~ Does the proposed test increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The consequences of malfunction of equipment important to safety are unaffected by this test. The consequences of a loss of shutdown cooling or a failure of the steam generator nozzle dams are the same regardless of this testing.

5. Does the proposed test create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The possibility of introducing nitrogen into the RCS as a result of this test has been evaluated. The initial test parameters have been chosen such that after the nitrogen gas has fully expanded there will still be liquid inventory remaining in the SIT. By conducting the test within the bounds. of these initial parameters, the possibility of injecting nitrogen into the RCS has been precluded. None of these effects will result in an accident of a different type than previously evaluated in the UFSAR.

98

6. Does the proposed test create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

This test does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR since the failure modes of the nozzle dams have been previously evaluated.

7. Does the proposed test reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed test does not change assumptions in the basis for any of the Technical Specifications. Therefore, the proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

99

1 Unit: ~

~ 2 JPN-PSL-SENP-94-039

Title:

~

Jumper/Lifter Lead for PDIS-2216 The purpose of this safety evaluation is to demonstrate the acceptability of applying a jumper to the pressure differential switch (PDIS-2216) in the chemical & volume control system (CVCS) letdown line. This jumper will remain in effect until the switch is repaired or replaced.

The switch functions to sense high differential pressure across the regenerative heat exchanger, which is indicative of high flow from a downstream line break outside containment, initiating a signal to close isolation valve V-2516. A letdown line break is postulated in the UFSAR as resulting from a seismic event. The jumper will defeat closure on high differential pressure. However, letdown isolation via V-2515 still occurs from temperature element TE-2221 located immediately downstream of the regenerative heat exchanger.

This safety evaluation involves an assessment of safety related systems, and therefore, is classified as safety related.

This evaluation concludes that operation of the plant with the

~ ~

proposed jumper on PDIS-2216 does not represent an unreviewed

~

safety question as defined in 10CFR50.59, does not require a change

~ ~

~

to plant Technical Specifications and does not adversely affect

~ ~ ~

plant operation or safety. Therefore, prior NRC approval is not

~

required for implementation.

~

SAFETY EVALUATION In accordance with 10CFR50.59, the responses to the following questions serve to determine whether the proposed jumper constitutes an unreviewed safety question:

1. Does the proposed jumper increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR will not increase because this jumper does not affect any equipment whose malfunction is postulated in the UFSAR to initiate an accident.

2 ~ Does the proposed jumper increase the consequences of an accident previously evaluated in the UFSAR?

100

The consequences of an accident previously evaluated in the UFSAR will not increase because this jumper does not adversely affect valve closure from a SZAS and CIAS.

3 Does the proposed jumper increase the probability of an occurrence of a malfunction of equipment important to safety

~

previously evaluated in the UFSAR?

The proposed jumper does not adversely affect the function of equipment important to safety. Therefore, the proposed jumper does not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

4 ~ Does the proposed jumper increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed jumper does not adversely affect equipment required to mitigate the consequences of an accident. Nithout automatic isolation via differential pressure, letdown isolation still occurs from temperature element (TE-2221) located immediately downstream of the regenerative heat exchanger. Therefore, the proposed jumper does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR.

5. Does the proposed jumper create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

New accident initiators are not introduced through the proposed jumper. Therefore, the proposed jumper does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

6 ~ Does the proposed jumper create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed jumper was determined not to create any new failure modes. Therefore, this jumper does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR 7~ Does the proposed jumper reduce. the margin of safety as defined in the basis for any Technical Specification?

The proposed jumper does not change assumptions in the basis for any of the Technical Specifications. Therefore, the proposed jumper does not reduce the margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),,

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

102

Unit: JPN-PSL-SEFZ-94-04 0

Title:

Removal of TE-1122CD Input from Channel "D" of the RPS for PSL 1 During power ascension following the refueling outage for St. Lucie Unit 1 Cycle 13, erratic cold leg temperature indications were obtained from one of the two temperature measurement channels which provide input to Channel "D" of the reactor protection system (RPS). This safety evaluation assessed the effect on UFSAR analyses of temporarily removing the faulty channel from operation for the remaining portion of Cycle 13.

This safety evaluation assessed the effects on safety of removing a faulty channel from input to RPS, as such this safety evaluation is classified as Safety Related.

This evaluation concludes that the proposed removal of the TE-1122CD input to Channel D of the RPS for the duration of St. Lucie Unit 1 Cycle 13 does not represent an unreviewed safety question nor does Therefore it it require a change to the Technical Specifications.

is concluded that the proposed temporary change may be performed without prior NRC approval.

SAFETY EVALUATION In accordance with 10 CFR Part 50.59, the following evaluation serves to determine whether this modification constitutes an unreviewed safety question:

Does the proposed change increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed change affects one of two cold leg temperature indications to RPS Channel "D". These instrument channels do not initiate any of the events analyzed in the UFSAR. The probability of occurrence of such events is, thus, unaffected by the proposed activity. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR is not increased.

2 ~ Does the proposed change increase the consequences of an accident previously evaluated in the UFSAR?

This evaluation demonstrated that the proposed change will not affect the functions which are credited in mitigating 103

consequences of accidents. Therefore, the consequences of an accident previously evaluated in, the UFSAR are not increased.

3 ~ Does the proposed change increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed change removes one of the two cold leg temperature indications to Channel "D" of the RPS. This change has no effect on the malfunction of any equipment or system important to safety as evaluated in the UFSAR, nor does it create any new failure modes. Therefore, the probability of occurrence of any equipment malfunction important to safety previously evaluated in the UFSAR will not increase.

4 ~ Does the proposed change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed change does not adversely affect the performance of any safety related equipment which functions to mitigate consequences of accidents. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR are not increased.

'oes the proposed change create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed change involves removal of one of the two cold leg temperature measurement channels which input Channel "D" in the RPS cabinet. There are no changes to the plant configuration and/or RPS functions. No new systems or system interactions are involved that adversely affect equipment or systems important to safety. Therefore, the proposed change does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

6~ Does the proposed change create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The change proposed allows operation with one of the two cold leg temperature measurement channels which input Channel "D" of the RPS removed. No new hazards are created as a result of the proposed change. This change does not adversely affect any safety related equipment. Additionally, there are no changes to any system configuration or equipment important to safety. Thus, the possibility of a malfunction of equipment 104

important to safety of a different type than any previously evaluated in the UFSAR is not increased.

7. Does the proposed change reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed change does not affect the UFSAR conclusions nor does it change the margin of safety as defined in the based for the Technical Specifications. Therefore, the proposed change does not reduce the margin of safety as defined in the bases for the Technical Specifications.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unrevieved safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

105

t Unit:

Title:

2 V

JPN-PSL-SENP-94-043 Safety Evaluation Temporary Removal of the Missile Shield ICW Pump This safety evaluation demonstrates the acceptability of plant operation while a section of the intake cooling water (ICW) pump missile shield roof is removed temporarily to perform maintenance activities on an out of service pump. The function of the missile shield is to protect the ICW pumps from missiles during a hurricane/tornado.

This safety evaluation documents the design intent of the ICW pump missile shield with respect to maintenance access. It concludes that removal of missile shield roof sections for maintenance during plant operation is consistent with the original design intent of shields. Furthermore, the risk of tornado missiles is negligible for the short period of time the roof section is not in place. As an additional precaution the missile shield roof sections are re-installed in the event of a threatening hurricane when the risk of damage from tornadoes is the greatest.

Since this safety evaluation assesses the effects on missile protection of the safety related ICW pumps, this evaluation is classified as safety related.

~

This safety evaluation demonstrates

~

that the temporary configuration of the ICW missile shield neither involves an

~ ~

unreviewed safety question nor requires a change to plant Technical Specifications, and does not adversely affect plant operation or safety. Therefore, it is concluded that the proposed temporary shield configuration may be implemented without prior NRC approval.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the following questions serve to determine whether the proposed temporary missile shield configuration constitutes an unreviewed safety question:

Does the proposed temporary change increase the probability of occurrence of an accident previously evaluated in the UFSAR?

Tornadoes/hurricanes do not initiate design basis accidents.

Therefore, the proposed temporary change does not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

106

2 ~ Does the proposed temporary change increase the consequences of an accident previously evaluated in the UFSAR?

Tornadoes/hurricanes are not postulated to occur simultaneously with design basis accidents. The performance of the operating pumps will not be adversely impacted.

Therefore, the consequences of an accident previously evaluated in the UFSAR are not increased by the proposed temporary missile shield configuration.

3 ~ Does the proposed temporary change increase the probability of an occurrence .of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The failure modes and effects analyses of the ICW system as described in the UFSAR are not changed by the temporary ICW missile shield roof configuration. The risk from missiles is negligible for the short period of time the roof section is not in place. In addition, the missile shield will be re-installed in the event of a threatening hurricane when the risk of damage from missiles is the greatest. Therefore, the proposed temporary change does not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

4~ Does the proposed temporary change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The risk from missiles is negligible for the short period of time the roof section is not in place. The performance of the operating pumps will not be adversely impacted. Therefore, the proposed temporary change does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR.

5. Does the proposed temporary change create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

Tornadoes/hurricanes do not initiate design basis accidents.

The failure modes and effects analyses of ICW as described in the UFSAR are not changed by the temporary ICW missile shield roof configuration. Therefore, the proposed temporary change does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

6. Does the proposed temporary change create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

107

The failure modes and effects analyses of the ICW system as described in the UFSAR are not changed by the temporary ICW missile shield roof configuration. The risk from missiles is negligible- for the short period of time the roof section is not in place. In addition, the missile shield will be re-installed in the event of a threatening hurricane when the risk of damage from missiles is the greatest. Therefore, the proposed temporary change does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR.

7. Does the proposed temporary change reduce the margin of safety as defined in the basis for any Technical Specification?

The temporary missile shield configuration does not require a change to Technical Specifications. The bases indicates that the ICW system must provide sufficient cooling water to vital.,

components, assuming a single failure, consistent with-assumptions used in the safety analysis. Two ICW loops remain operable and accident analyses single failure assumptions are not affected by the temporary missile shield configuration.

The risk from missiles is negligible for the short period of time the roof section is not in place. In addition, the missile shield will be re-installed in the event of a threatening hurricane when the risk of damage from missiles is the greatest. Therefore, the margin of safety as defined in the basis for technical specifications is not reduced by the temporary ICW missile shield roof configuration.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

108

Unit: JPN-PSL-SENP-94-044

Title:

Safety Evaluation for the use of Devoe Devran 140 Epoxy Compound and Kansai Biox as a Coating System for the St.

Lucie Unit 1 Intake Structure The purpose of this evaluation is to demonstrate the acceptability of applying coating systems to the St. Lucie Unit 1 intake well walls (including the safety related Intake Cooling Water (ICW) wells) consisting of Devoe Devran 140 Epoxy Compound and Kansai Biox for concrete surfaces and Amerlock 400 with a Kansai Biox overcoat for steel surfaces.

The purpose of coating the intake structure wells is to limit the growth of marine organisms within the intake structure. Reducing marine growth on the intake structure will improve heat exchanger performance by reducing fouling and blockages. -

Currently, a hypochlorite solution and Clamtrol is injected into the sea water to control the marine growth. The use of this coating system may reduce the amount of chemicals injected into the sea water, which will have a positive effect on the environment.

To demonstrate the acceptability of using this coating system at St. Lucie, an intake well was cleaned and prepared in accordance with the instructions provided by Specification CN-2.27, "Furnishing and Application of Service Level II and Balance-of-Plant Protective Coatings." A Coating Data Sheet (Attachment Al to CN-2.27) was prepared by the FPL Coating Specialist and the coating system was applied to the concrete walls of intake well 2B2 in the spring of 1994. In addition, a steel plate was coated with Amerlock 400 and Kansai Biox was applied as an overcoat.

Acceptable adhesion test results were obtained with no cohesion failures noted. Industry data and the data collected from the tests at St. Lucie have indicated that the coating. systems do not have a failure mode which results in the delamination of large sheets of epoxy which could cause ICW system blockages, that the coating systems maintain acceptable adhesion characteristics and reduce the growth of marine organisms. These results indicate that the surface preparation and the use of these coating systems at St.

Lucie is acceptable.

This safety evaluation involves an assessment of changes to safety related ICW intake wells and is therefore classified as safety related.

This evaluation concludes that the proposed coating of the intake wells neither involves an unreviewed safety question, as defined in 10 CFR 50.59(a)(2), nor recpxires a change to plant Technical 109

Specifications, and does not adversely affect plant operation or

~

safety. Therefore, in accordance with 10 CFR 50.59, prior NRC

~

approval is not required for implementation.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine if the proposed coating of the intake structure concrete walls with Devoe Devran 140 Epoxy Compound and Kansai Biox and the steel surface with Amerlock 400 and Eansai Biox as coating systems constitutes an unreviewed safety question as defined in 10 CFR 50.59(a) (2):

1. Does the proposed coating of the intake structure surfaces increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The delamination of the coating systems cannot initiate an accident. As discussed in the failure modes and effects section of this evaluation, the industry data and the data from the St. Lucie demonstration well confirms the performance of the coating systems to adhere to the properly prepared concrete and steel surfaces of the intake structure. The test well also confirms that the coating does not fail in large layers. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR has not been increased and the proposed change does not affect any accidents discussed in the UFSAR.

2 ~ Does the proposed coating of the intake structure surfaces increase the consequences of an accident previously evaluated in the UFSAR?

The demonstration well confirms that the coating systems will adhere to the intake structure walls and steel surfaces, and therefore will not clog the ICW strainers or degrade CCW heat exchanger performance. Therefore, the proposed coating of the intake structure surfaces with the coating systems will not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed coating of the intake structure surfaces increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The coating systems will be applied in accordance with Specification CN-2.27 and a new Coating Data Sheet. This specification was used to coat a demonstration well at St.

Lucie to confirm the adhesion properties of the coating and 110

verify the preparation requirements. The demonstration well confirms that the coating performs as the industry data indicates, has excellent adhesion to the intake structure surfaces and does not fail in large heat pieces which could clog performance.

the ZCW strainers or degrade CCW exchanger structure Therefore, the proposed coating of the intake surfaces with these coating systems will not change the probability of occurrence of a malfunction of equipment previously evaluated in the UFSAR.

Does the proposed coating of the intake structure surfaces increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

Since the demonstration well confirms that the coating systems will adhere to the intake structure surfaces and therefore will not clog the ICWlosses strainers, the consequences from the of the intake cooling water system previously evaluated remain valid. Therefore, the proposed coating of the intake structure surfaces does not increase the consequences of a malfunction of equipment important to safety. previously evaluated in the UFSAR.

Does the proposed coating of the intake structure surfaces create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The UFSAR evaluates the'ailure of an ZCW pump or the intake strainer which are the only credible failure modes which this change could create. Evaluation of the hypothetical simultaneous failure of the coating systems on both trains of ICW intake wells during an accident sequence concludes that due to the excellent adhesion properties of this coating system, this is not a credible event. Therefore, no different type of accident is created and the accident assumptions found in the UFSAR remain valid.

Does the proposed coating of the intake structure surfaces create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

One failure mode involves the hypothetical simultaneous failure of the coating system on both trains of ICW intake well walls during an accident sequence and it was that due to the excellent adhesion properties of this coatinga concluded system this was not a credible event. The malfunction of single ICW pump or ZCW strainer is evaluated in the UFSAR and the results remain valid. Therefore, the proposed coating of the intake structure surfaces does"not create the possibility 111

of a malfunction of equipment important to safety of, a different type than any previously evaluated in the UFSAR.

7. Does the proposed coating of the intake structure surfaces reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed change does not affect any Technical Specification nor does it reduce any margins of safety defined by the Technical Specifications. The proposed change will improve heat exchanger performance by reducing marine growth in the intake structure wells.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

112

Unit: JPN-PSL-SENP-94-047

Title:

SIT Discharge/Loop Check Valve Stroke Test ABSTRACT This safety evaluation demonstrates the acceptability of performing a full strokecheck test of the Safety Injection Tank ,(SIT) valves. The proposed test will be used to discharge/loop address NRC requirements for SIT check valves testing delineated in Generic Letter 89-04.

The test is to be performed during refueling with the reactor vessel head and upper internals removed, 'and the refueling cavity flooded. The test consists of establishing a reduced nitrogen pressure in the SIT to be used as a motive force to discharge a portion of the SIT inventory through the subject check valves. The test is initiated by opening the SIT discharge line motor operated valve and observing via acoustic and magnaflux monitoring that the check valves fully open.

Calculations have been performed to demonstrate that sufficient velocities can be achieved and.maintained to fully open the subject check valves. This evaluation addresses the various potential adverse effects on the plant including those on the reactor internals and steam generator nozzle dams resulting from the proposed flow test.

Since this safety evaluation assesses the effects of the proposed test on related.

safety related components, it is classified as safety This evaluation demonstrates that the proposed test neither involves an unreviewed safety question nor requires a change to plant Technical Specifications, and does not adversely affect plant operation or safety. Therefore, test may be performed without it prior is concluded that the proposed NRC approval.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the following questions serve to determine whether the proposed test constitutes an unreviewed safety question:

Does the proposed test increase the probability of occurrence of an accident previously evaluated in the UFSAR?

Performance of this test will not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

Specifically, the test will not increase the probability of a 113

loss of shutdown cooling and will not cause an upset of fuel assemblies. The maximum fluid velocity that will occur in the vessel, as a result of this test, is on the order of 1/2 ft/sec, and therefore, is not capable of adversely impacting the fuel or CEAs.

2 ~ Does the proposed test increase the consequences of an accident previously evaluated in the UFSAR?

This test will be conducted in Mode 6 with the reactor vessel head removed. This test does not impact any of the assumptions for refueling accidents evaluated in the UFSAR and will not increase the consequences of any of these analyses.

3 ~ Does the proposed test increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The shutdown cooling system will be secured in the tested loop. There are no potential failure modes resulting from this test that would adversely impact shutdown cooling equipment. The additional pressure that the nozzle dams may be subjected to as a result of this test have been analyzed and determined not to affect nuclear safety. Therefore, this test will not increase the probability of a malfunction of equipment important to safety.

Does the proposed test increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The consequences of malfunction of equipment important to safety are unaffected by this test. The consequences of a loss of shutdown cooling or a failure of the steam generator nozzle dams are the same regardless of this testing.

5. Does the proposed test create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The possibility of introducing nitrogen into the RCS as a result of this test has been evaluated. The initial test parameters have been chosen such that after the nitrogen gas has fully expanded there will still be liquid inventory remaining in the SIT. By conducting the test within the bounds of these initial parameters, the possibility of injecting nitrogen into .the RCS has been precluded.

114

Potential effects on safety are discussed in this evaluation.

None of these'ffects will result in an theaccident of a different type than, previously evaluated in UFSAR.

6. Does the proposed test create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

This test does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR since the failure modes of the nozzle dams have been previously evaluated.

7. Does the proposed test reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed test does not. change assumptions in the basis for any of the Technical Specifications. Therefore, the proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

Unit:

~

JPN-PSL-SEMP-94-050

Title:

~

~ Temporary Alterations to the Refueling Water Tank During the upcoming St. Lucie Unit 1 1994 Fall Refueling Outage, repairs are to be made to the Refueling Water Tank (RWT), per engineering evaluation JPN-PSL-SENP-93-035. The RWT must be completely drained in order to implement the required repairs. The removal of the RWT inventory can be expedited by the implementation of certain temporary changes. These changes are as follows:

1) The first proposed alteration is the addition of temporary piping inside the RWT. The proposed piping arrangement would serve to extend the fuel pool line connection to within several inches of the tank bottom. This would allow the use of the Fuel Pool Purification pump to drain the tank below the elevation of the fuel pool line connection.

2) The second temporary change would be the connection of a temporary valve and piping arrangement to the RWT drain line.

The exterior drain connection has a valve and blind flange.

The blind flange would be removed and a temporary valve and piping arrangement would be attached to permit the stroking of valve I-V-07-106'o verify its operation. This drain line will be used during the Unit 1 outage to pump a portion of the RWT inventory to the Waste Management System in the Reactor Auxiliary Building (RAB).

It is the intent to implement these temporary changes while St.

Lucie Unit 1 maintains power operation. Neither of the proposed changes require a permanent change to the facility. The 50.59 safety evaluation will address the acceptability of implementing these temporary changes during operation. The make-up water supply to the Spent Fuel Pool is not affected by the addition of the temporary piping. With the exception of cycling I-V-07-106, the modification to the drain line will not be put into service until the unit is shutdown. The tank boundary will not be altered, unless valve I-V-07-106 is found to leak in the closed position.

Engineering Evaluation JPN-SENP-93-035 provides the 50.59 safety evaluation for the RWT repair.

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety. Therefore, prior NRC approval is not required.

116

SAFETY EVALUATION In accordance with 10 CFR 50.59, the following evaluation serves to determine whether this modification constitutes an unreviewed safety question or requires a change to the Technical Specifications:

1 ~ Does the proposed change increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed temporary changes do not affect any equipment whose malfunction is postulated in the UFSAR to initiate an accident or prevent an accident from occurring. The proposed change maintains the Refueling Water Tank's ability to perform its intended functions. As such, the probability of occurrence of an accident previously evaluated in the UFSAR has not increased.

2 ~ Does 'the proposed change increase the consequences of an accident previously evaluated in the UFSAR?

The proposed temporary changes do not diminish in any way the ability of the Refueling Water Tank to perform its intended function to mitigate the consequences of an accident previously evaluated in the UFSAR. The inventory of borated water required by the Emergency Core Cooling Systems is maintained. As such, the proposed change does not increase the consequences of an accident previously evaluated in the UFSAR.

3~ Does the proposed change increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed temporary changes maintain the quality level and the level of protection previously established for the Refueling Water Tank. The proposed temporary changes do not affect the boundary integrity of the tank. As such, the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased by the proposed change.

4 ~ Does the proposed change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The Refueling Water Tank has a passive function to maintain the required inventory of borated water for the Emergency Core Cooling Systems. The proposed temporary changes do not affect the level of borated water in the Refueling Water Tank. The 117

flow path from the Refueling Water Tank to the Emergency Core Cooling Systems is not,affected by the proposed changes.

Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased by the proposed change.

5. Does the proposed change create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

As discussed in failure modes and effects analysis section of this evaluation, the proposed temporary changes do not introduce any new failure modes. The proposed temporary changes are intended to maintain the requ'irements of the design bases of the Refueling Water Tank as described in the UFSAR. Therefore, the possibility of an accident of a different type than any previously evaluated in the UFSAR is not created by this modification.

6. Does the proposed change create the possibility of a malfunction of equipment important to .safety of a different type than any previously evaluated in the UFSAR?

I The proposed temporary changes do not interact spatially or functionally with any structure, system or component important to safety other than the Refueling Water Tank. No new failure modes are created for the temporary proposed changes that can be postulated to cause a malfunction of equipment important to safety different than those previously analyzed in the UFSAR.

Therefore, the possibility of a malfunction of equipment important to safety which is of a different type than any previously evaluated in the UFSAR is not created by the proposed change.

7~ Does the proposed change reduce the margin of safety as defined in the basis for any Technical Specification?

The Technical Specification requirements and Technical Specification Bases are not affected-by the proposed temporary changes. The proposed temporary changes do not affect any plant Technical Specification requirement. The proposed change maintains the quality and level of protection previously evaluated in the UFSAR. Therefore, the margin of safety as defined in the bases for any Technical Specification is not reduced by this modification.

The foregoing constitutes the .determination, per 10 CFR 50.59(b),

that the subject activity does not. involve an unreviewed safety question, does not require a change to plant Technical 118

Specifications and does not adversely affect plant operation or safety.

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Unit: JPN-PSL-SENP-94-065

Title:

Containment Air Conditioning for Refueling Outage ABSTRACT This safety evaluation demonstrates the acceptability of temporarily modifying the component cooling water (CCW) system during the 1994 refueling outage to provide air conditioning inside containment via a closed loop chilled water system. This temporary change will control the temperature and humidity inside containment during the refueling outage. This system utilizes two 200 ton self-contained chilled water units located outdoors between the reactor auxiliary building (RAB) and the fuel handling building to supply chilled water to the coils in the containment fan coolers.

The containment fan cooler fans move air across the coils cooling the air while the heat is carried away by the water. Temporary hoses will be used to transport the chilled water into the RAB pipe penetration area where connection via flanged spool pieces is made into the existing CCW supply/return piping for the containment fan coolers. The connection is made by removing the normally locked open/throttled CCW supply/return isolation valves and installing the spool pieces. The temporary change will be made to a single train (one or two containment fan coolers) at a time, allowing the other train to be available for reduced inventory evolutions. The spool pieces are blind flanged such that the CCW system upstream of these connections are unaffected by the chilled water and remain operable. Temporary cables will be used to power the chilled water units from an offsite power source and would therefore not impact any safety related or non-safety related plant power supply.

There are no permanent modifications/configurations affecting operation of the CCW system, containment fan coolers, or the electrical power supply. All system alignments are temporary. The system, equipment and piping will be restored to its normal condition at the end of the refueling outage, prior to entering Mode 4.

This safety evaluation documents the temporary design of the CCW system to provide containment air conditioning during the refueling outage. Although the temporary CCW configuration will be operated at temperatures lower than normal, this, safety evaluation concludes that the temporary CCW configuration utilizing chilled water and operation of the containment fan coolers is consistent with the design intent of the CCW system and does not have an adverse affect on plant safety, security, or operation.

Since this safety evaluation assesses the effects on containment fan cooler operation utilizing chilled water for containment air 120

conditioning, supplied through CCW essential headers A and B, this evaluation is classified as safety related.

This safety evaluation demonstrates that the temporary configuration of the CCW system to supply chilled water through the fan cooler supply/return header for containment air conditioning during the refueling outage neither involves an unreviewed safety question, requires change to plant Technical Specifications, and a

does not adversely affect plant operation or safety. Therefore, is concluded that the proposed temporary CCW configuration to it provide containment air conditioning during the refueling outage may be implemented without prior NRC approval.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the following questions serve to determine whether the proposed temporary CCW configuration constitutes an unreviewed safety question:

1. Does the proposed temporary change increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of, occurrence of an accident previously evaluated in the UFSAR will not increase because neither implementation of the modification to the CCW system nor operation of the containment fan coolers in the temporary CCW configuration affect any equipment postulated in the UFSAR to initiate an accident or prevent an accident from occurring.

Therefore, the proposed temporary change does not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

2. =

Does the proposed temporary change increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR will not increase because neither implementation of the modification to the CCW system nor operation of the containment fan coolers in the temporary CCW configuration affect any structures, systems or components that functions to mitigate the consequences of an accident, to contain or detect the release of radioactivity or to provide post-accident shielding. The containment fan coolers and CCW flow for containment fan coolers are not required while in Mode 5 or 6.

The temporary CCW configuration will be restored to its normal condition at the end of the refueling outage, prior to entering Mode 4.

s. Does the proposed temporary change increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated .in the UFSAR?

The failure modes and effects analyses of the CCW system as described in the UFSAR are not changed by the temporary CCW configuration. The portion of CCW system affected by this temporary configuration is isolated from the remainder of the CCW.system. During the refueling outage, this temporary CCW configuration is considered not to interact functionally with any structure, system or component important to safety.

Additionally, the design of the temporary system is within the design envelop of CCW system 'and does not adversely affect the CCW system. Therefore, the proposed temporary change does not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

4 ~ Does the proposed temporary change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The failure modes and effects analyses of the CCW system as described in the UFSAR are not changed by the temporary CCW configuration. This temporary CCW configuration is considered not to interact spatially or functionally with any structures, systems or components that functions to mitigate the consequences of an accident, to contain or detect the release of radioactivity or to provide post-accident shielding.

Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR is not changed.

Does the proposed temporary change create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The failure modes and effects analyses of the CCW system as described in the UFSAR are not changed by the temporary CCW configuration. The temporary CCW configuration provides-chilled water flow to the containment fan coolers under conditions within the design of the CCW system. Operation in the temporary configuration does not adversely affect the system. The CCW system will be restored to its original design prior to entering'ode 4 following the refueling outage. No new hazards are created that can be postulated to cause an accident different than those previously analyzed in the UFSAR. Therefore, there is no possibility that an accident may be created that is different from one already evaluated in the UFSAR.

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6. Does the proposed temporary change create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The failure modes and effects analyses of the CCW system as described in the UFSAR are not changed by the temporary CCW configuration.. The temporary CCW configuration provides chilled water flow to the containment fan coolers under conditions within the design of the CCW system. Operation in the temporary configuration does not adversely affect the system. The CCW system will be restored to its original design prior to entering Mode 4 following the refueling outage. Interaction due to condensation runoff was addressed in the analyses and effects on safety section with recommendations to assure that there will be no adverse interaction. The temporary configuration is considered not to create any new hazards which can be postulated to cause a malfunction of equipment important to safety different than those previously analyzed in the UFSAR. Therefore, the proposed temporary change does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR.

7~ Does the proposed temporary change reduce the margin of safety as defined in the basis for any Technical Specification?

The Technical Specification requirements and bases applicable to this temporary CCW configuration are not affected by this temporary change. The equipment operated due to this temporary configuration is not required by the Technical Specification while temporary containment cooling is in service (i.e., Modes 5, and 6).. The CCW system will be restored to its original design prior to entering Mode 4 following the refueling outage. Therefore, the temporary CCW configuration and operation of the containment fan coolers while in this 'emporary configuration does not reduce the margin of safety as defined in the bases for the Technical Specifications.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question; does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

123

Unit: JPN-PSL-SEEP-'94-066

Title:

Safety Evaluation for Operation of Three Charging Pumps ABSTRACT The purpose of this safety evaluation is to demonstrate acceptability of the concurrent operation of three charging pumps.

This safety evaluation revises the UFSAR EDG loading table (Table 8.3-2) to indicate the auto-start of 2 charging pumps on one EDG and the UFSAR description of the charging pumps control logic.

Operation of 'the charging system ,is with two charging pumps operating; one running continuously, one in automatic standby, and one in off. The charging pump in off condition will not respond to a SIAS start; however, the charging pump running continuously and the charging pump in automatic standby will respond to a SIAS .

start. This results in loading only one charging pump on each emergency diesel generator'n the event of a LOOP or coincident LOOP/LOCA. The present. emergency diesel generator loading analysis accounts for loading of one charging pump in the first load block after diesel generator breaker closure.

During plant operations which require maximum RCS purification or inventory makeup, the CVCS may be operated with all three charging pumps running concurrently. This could result in the loading of two charging pumps on one diesel, generator in the event of a LOOP or LOOP/LOCA.

This safety evaluation involves assessment of the charging and safety related onsite power distribution systems and is therefore classified as safety related. Operation of three charging pumps will not exceed the capability of the emergency diesel, generators in the event of a loss of offsite power or loss of offsite power coincident with a LOCA and is within the design parameters of the CVCS. There is no unreviewed safety question as defined in 10CFRS0.59, no changes are required to plant Technical Specifications, and safe plant operation is not adversely affected.

SAFETY EVALUATION Zn accordance with'0CFR50.59, the responses to the following questions serve to determine whether the operation of three charging pumps constitutes an unreviewed. safety question:

1. Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

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increase in the first block loading is within the I'he capability of the emergency diesel generator. The LTOP analysis, charging system stress and fatigue analysis, the UFSAR Chapter 15 .analysis and RCP seal injection are not adversely affected. Therefore, it does not probability of occurrence of an accident since this could not increase the initiate an accident previously evaluated in the UFSAR.

2. Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The operation of three charging pumps has been shown to have no adverse effect on the safety functioning of the CVCS system or the EDQs and their ability to mitigate the effects of an accident have not changed. Therefore, the consequences of an accident previously evaluated in the UFSAR has not been increased.

3. Does the proposed activity increase the probability of occurrence- of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The increased EDG loading due to the operation of three charging pumps has been found to be within the design capacity of the EDG. Zn addition, other components in the CVCS are not affected by running three charging pumps. Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased.

4. . Does the proposed activity increase the consequences of a malfunction of equipment important, to safety previously evaluated in the UFSAR?

The increased EDG loading has been found to be within the design capacity of the EDG. -

The additional flow from operating three charging pumps is within the design parameters of the CVCS components. Therefore, there is no increase to the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The increased EDG loading has been found to be within the design capacity of the EDG. The additional flow from operating three charging pumps is within the design parameters of the CVCS components. No new failure modes have been created. Therefore, no additional possibilities have been 125

I created for an accident of a different type than any previously evaluated in the UFSAR.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR?

The increased EDG loading has been found to be within the design capacity of the EDG. The additional flow from operating three charging pumps is within the design parameters of the CVCS components. Therefore, the possibility of a malfunction of, equipment important to safety of a different type than previously evaluated in the UFSAR has not been created.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical SpecificationP The maximum loading of the EDG is maintained at less than the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> maximum loading, in accordance with Technical Specifications. The function and components of the CVCS are not affected by the operation of three charging pumps.

Therefore, the margin of safety, as defined in the basis for the Technical Specifications, has not been reduced.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

126

Unit: JPN-PSL-SEMP-94-076

Title:

Increase of Engineered Safeguards Suction Piping Design Pressure After successful performance of a motor operated valve differential pressure test as required by NRC Generic Letter 89-10, water was observed on the floor of the pipe tunnel in the Reactor Auxiliary Building. The source of the water was determined to be from relief valve, I>>SR-07-1A, located on the 1A Engineered Safeguards suction piping.

An examination of the alignment used for the test revealed a flow path from the discharge of the 1B Containment Spray pump to the suction of the 1A Containment Spray pump through a common header in the Sodium Hydroxide (NaOH) Spray Additive system. With the 1B Containment Spray pump operating and all the A ECCS pumps secured, the A train suction piping was pressurized through the Spray Additive system common header. As part of the investigation of the event, the 1B Low Pressure Safety Injection pump was aligned to the B Containment Spray header and discharged through the B Shutdown Cooling heat exchanger. The maximum pressure observed at the 1A Low Pressure Safety Injection pump discharge header was 80 psig.

Relief valve I-SR-07-1A was determined to discharge. Plant personnel documented the event on St. Lucie Action Request (STAR) 1-94100259, which was assigned to Nuclear Engineering for disposition.

An Initial Assessment of Operability was performed to respond to the concerns addressed by the STAR. A calculation determined that the components whose design pressure had'been exceeded were in fact capable of withstanding considerably higher pressures. The suction piping and components, therefore, did not suffer any damage as a result of the event. However, with regard to the issue of the system design, it was concluded that a design basis scenario exists which could result in lifting the relief valve; a containment spray actuation signal (CSAS) and a loss of off-site power (LOOP) coincident with one Emergency Diesel Generator (EDG) failing to operate. The relief valve could potentially open and release containment sump inventory in excess of the Engineered Safeguards equipment external leakage rate of 2 liter per hour, UFSAR 15.4.1.7 and 15.4.1.8. This could result in a condition outside of the design basis of the Engineered Safeguards systems. Based on the Initial Assessment of Operability, plant management Commission made a one hour non-emergency notification to the Nuclear Regulatory in accordance with 10 CFR 50.72.

127

This safety evaluation was prepared to evaluate the acceptability of higher pressures in the Engineered Safeguards suction piping in order to disable relief valves I-SR-07-1A and I-SR-07-1B for this cycle while above. Mode 4. This interim measure is being implemented to preclude the possibility of an unwanted release.

The proposed change does not require a permanent change to the facility. The safety evaluation addresses the acceptability of implementing the proposed change during operation. This safety evaluation involves Engineered Safeguards systems and is therefore classified as safety related.

Based on the evaluation herein, it has been determined that an unreviewed safety question does not exist and the plant Technical Specifications are not affected. Therefore, prior notification of the NRC is not required.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the fo'llowing evaluation serves to determine whether this modification constitutes an unreviewed safety question or requires a change to the Technical Specifications:

1 ~ Does the proposed change increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed change does not affect any equipment whose malfunction is postulated in the UFSAR to initiate an accident or prevent an accident from occurring. As such, the probability of occurrence of an accident previously evaluated in the UFSAR has not increased.

2 ~ Does the proposed change increase the consequences of an accident previously evaluated in the UFSAR?

The proposed change does not affect the design function of any equipment designed to mitigate the consequences of an accident previously evaluated in the UFSAR. No new failure modes are being introduced and the design margin of equipment important to safety is not being decreased. As such, the proposed change does not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed change increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The component affected by this change is the Engineered Safeguards suction piping. The proposed change maintains the quality level and the level of protection previously 128

established for the Engineered Safeguards suction piping. The design allowable pressure rating of the piping and associated components's above the maximum system pressure resulting from this change. Although the new higher pressure is above the original tested pressure, the design margin between allowable stresses and ultimate capacity is not being decreased. The proposed change, therefore, does not affect the pressure boundary integrity of the Engineered Safeguards suction piping. As such, the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased by the proposed change.

4 ~ Does the proposed change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The malfunction evaluated in the UFSAR is the complete or partial failure of one train of Engineered Safeguards to perform its function. The proposed change does not in any way affect the ability of the redundant train of Engineered Safeguards to perform its function to inject and recirculate borated water. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased by the proposed change.

5. Does the proposed change create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed change does not introduce any new failure modes.

The proposed change serves to reduce the possibility of a component failure. Therefore, the possibility of an accident of a different type than any previously evaluated in the UFSAR is not created by this proposed change.

6. Does the proposed change create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed change does not interact spatially or functionally with any structure, system or component important to safety other than the Engineered Safeguards suction piping.

No new failure modes are created by the proposed change that can be postulated to cause a .malfunction of equipment important to safety different than those previously analyzed in the UFSAR. Therefore, the possibility of a malfunction of equipment important to safety which is of a different type than any previously evaluated in the UFSAR is not created by the proposed change.

129

PROCEDURE! 2-LOZ-0-65 Unit: 2 DESCRIPTION OF THE CHANGE This change involves the, transfer of borated water from a PSL2 Boric Acid Makeup Tank (BAM Tk) to the 2A Holdup Tank (2A HUT).

This will provide a back-up borated water supply for PSL1 during the repairs to the Refueling Water Tank (RWT). This change will be controlled via a procedure (LOI) and will.be in effect only during the outage time when the RWT is under repairs.

SAFETY EVALUATION

SUMMARY

The probability of occurrence or the consequences 'of an accident or malfunction previously evaluated has not increased, since there is no additional likelihood of an evaluated accident occurring due to storage of borated water in the HUT, since this is the design function of the HUT.

The possibility of an accident or malfunction of a different type than any evaluated previously has not been created, since the jumper to be used is of the, same design requirements as the piping where it is connected and the procedure calls for an operator to be stationed at the jumper during this evolution.

The margin of safety as defined in the basis for any technical

~ ~ ~

specification is not reduced, since this evolution is bounded by

~ ~ ~ ~ ~ ~ ~

the LCOs associated with Boric Acid Injection flowpaths and

~ ~ ~ ~ ~

reactivity

~ ~

it control.'ased on the above, is concluded that the change in question does not constitute an unreviewed safety question or a'hange to the Technical Specifications.

131

PROCEDURE< HP-7i REVISION 0 Unit: 1 6 2 DESCRIPTION OF THE CHANGE This change involves a change to a procedure as outlined in Chapter 12 of the UFSAR. Specifically, the change involves a revision to the procedural requirements, which are outlined in the UFSAR Chapter 12, used in the issuing of personal dosimetry. Currently in the UFSAR, self-reading dosimeters are called out for as the instruments to be provided for the purpose of external exposure monitoring. This change involves using digital alarming dosimeters.

SAFETY EVALUATION

SUMMARY

The new type dosimeters have a wider range and can be read more accurately than the self-reading dosimeters. The use of the digital dosimeters enhances the ability of an individual in monitor their personal external exposure.

Therefore, this change in exposure monitoring does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated; does not create the possibility for an accident or malfunction of a different type than any previously evaluated; and does not reduce the margin of safety as defined in the basis for any technical specification.

Based on the above, it is concluded that the change in question does not constitute an unreviewed safety question or a change to the Technical Specifications.

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PROCEDURES 1-ONP 01 05 REVISION 0 Unit: 1 DESCRIPTION OF THE CHANGE The proposed change involves using a Containment Spray (CS) pump for decay heat removal during an outage. This alignment would only be used as a contingency plan, should a Low Pressure Safety Injection (LPSI) pump be unavailable. This proposed decay heat removal alignment can only be used during Mode 6, and, with at least 23 feet of water above the top of the irradiated fuel assemblies. If less than 23 feet of water exist above the fuel, the action statements of,the Technical Specifications apply. The change would require installation of a flange on the LPSI suction header and the removal of the internals of check valve'V07000. An ISLT should be conducted subsequent to installation of the flange and removal of check valve internals to insure that a leakage path has not been created. This would allow the use of the lA CS pump for decay heat removal. The configuration of the plant when this change is allowed would be covered under LCO 3.9.8.1, hence the limitation for this change shall require greater than 23 feet water level above the fuel assemblies, and one SDC train operable and in operation.

SAFETY EVALUATION

SUMMARY

The probability of occurrence or the consequences of an accident or malfunction previously evaluated has not been increased, since no credit is taken for this change as an "operable" SDC train, and since this change has been evaluated against the design criteria for the SDC system. In addition, the CS pump has similar design attributes as the LPSI pump and the differences in configuration as relating to piping, elevation and materials, have been evaluated.

The loss of decay heat removal capability has already been evaluated under the system LCO.

The possibility of an accident or malfunction -of a different type than any evaluated previously has not been created, since the alternate decay heat removal system will be verified as not creating any leakage paths as a result of this change. Also, since the actions required by the Technical Specifications are being taken (i.e. Containment'integrity established) there is no increase in any off-site dose releases previously analyzed. Furthermore, since this configuration cannot be used subsequent to a LOCA, there can be no likelihood of fission products being introduced to the RAB by this change.

The margin of safety as defined in the basis for any technical specification is not reduced, since this change is bounded by the associated with the plant conditions and system LCO associated with it.

Based on the above, it is concluded that the change in question does not constitute an unreviewed safety question or a change to the Technical Specifications.

133

PROCEDUREs 2-LOZ-T-88 Unit: 2 DESCRIPTION OF THE CHANGE This change involves the use of a "roughing" filter for the Fuel Pool Purification System. This change will be controlled via a procedure (LOI) and will be in effect for about 6 weeks. The need for this change involves the desire to remove particulate matter from the fuel pool at rates larger that currently available with the existing filter. This will allow the capability of processing larger volumes of water through the filter while minimizing the amount of times the filter element needs to be replaced.

SAFETY EVALUATION

SUMMARY

The probability of occurrence or the consequences of an accident or malfunction previously evaluated has not increased, since there is no additional likelihood of an evaluated accident occurring due to the addition of the "roughing" filter.

The possibility of an accident or malfunction of a different type than any evaluated previously has not been created, since the filter to be used is of the same design requirements as the previous filter and the system configuration will be controlled by the LOI.

The margin of safety as defined in the basis for any technical specification is not reduced, since there are no technical specifications affected by this change.

Based on the above, it is concluded that the change in question does not constitute an unreviewed safety question or a change to the Technical Specifications.

134

g t Jumper/Lifted Lead g 2-94-007 Unit: 2 Component and Systems Affected:

Upender Vertical Circuit Limit Switch (2LS-BV)

Reason for the Request:

To remove contact 2CR-BV which isolates the refueling transfer upender [refueling side only] hydraulic positioning cylinder from the pump when the upender has reached the upper limit switch. The jumper is installed with a ganged switch to the existing upender control switch so that it is only in the circuit wh'en the switch is taken to vertical.

Safety Analysis:

This J/LL does not increase the probability of occurrence of an accident postulated in the UFSAR. The removal of this interlock and the accident scenarios interlocks that will remain it prevents are ensured by other functional.

This J/LL does not increase the consequences of an accident previously evaluated in the UFSAR. The design basis fuel handling accident described in the UFSAR section 15.7.4.1.2 is significantly conservative with respect to the circumstances of this jumper.

This J/LL does not increase the probability of an occurrence of a malfunction of equipment important to safety. The remaining interlocks ensure the equipment is properly protected from damage.

This J/LL does not increase the consequences of a malfunction of equipment important to safety. No failure of the upender hydraulic system will induce a condition different from normal upender use.

This J/LL does not create the possibility of an accident different than described in the UFSAR. The fuel transfer machine and the refueling machine do not interact with plant systems other than refueling equipment. Therefore, the postulated accident in UFSAR 15.7.4.1.2 is bounding.'his J/LL does not increase the possibility of malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR. Failure of the hydraulic controls is not considered more likely than those failures already postulated.

This J/LL does not reduce the margin of safety as defined in the basis for any technical specification. The jumper affects refueling equipment systems only and does not affect the basis for any technical specification.

135

JUMPER/LIFTED LEAD gi 94 019 Unit:

Component and Systems Affected:

Nitrogen to the Sodium Hydroxide(NaOH) Tank Reason for the Request:

This J/LL is to install a temporary nitrogen supply to the NaOH tank during the time that the Reactor Auxiliary Building(RAB) nitrogen header is out of service for modifications. The temporary supply consists of a nitrogen bottle and regulator'.

The purpose of the nitrogen supply is to provide a cover gas to inert the tank and to prevent the NaOH from reacting with the atmosphere. The cover gas serves no safety function and is not designed as safety related. NPSH for the eductors is provided via atmosphere. There are two vacuum breakers on the top of the NaOH tank that are not affected by this J/LL. The jumper was installed and removed within 2 days.

Safety Analysis:

The nitrogen cover gas is not safety related. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR is not increased.

UFSAR page 6.2-111 states that the nitrogen cover gas is not required for NaOH injection. Therefore, the consequences of an accident previously evaluated in the UFSAR are not increased.

The probability of occurrence of a malfunction of equipment important to safety is not increased. Chemistry periodically monitors the NaOH concentration. Also, the limited time exposure to air should not significantly affect NaOH concentration temporary nitrogen is lost.

if the 136

JUMPER/LIFTED LEAD g 1 94 20 Unit: 1 Component and Systems Affected:

Circulating Water Pumps Reason for the Request:

The plant service water system supplies lube water - to the Circulating Water Pumps(CWP) . The design is shown on dwg 8770-G-082 and 8770-G-084 sheet 2. To perform maintenance on the service water piping, a backup water supply is needed for the CWP. The plant fire water system was used as the backup. A manifold connected to each CWP lube water supply was fed from fire hydrant FH-9 located at the intake area. The fire water system has two pumps with a rated capacity of 2500 gpm. (see UFSAR section 3.1.3) Hose station HH-2-9 is operable for fire protection of equipment at the intake structure.

Safety Analysis:

The effects of the alternate lube oil arrangement for the CWP is limited to the fire protection system. This system is not, postulated to in the UFSAR as an initiator of an accident.

Therefore, this jumper will not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

The consequences of an accident previously evaluated in the UFSAR are not increased because the alternate arrangement does not affect the ability of the fire system to perform is designed function. A hose station in the area of the intake will be available to protect equipment in the event of a fire.

The fire protection system is designed with isolation valves that allow operators to isolate a rupture in the fire protection system without disabling the remainder of the system. In the event that the manifold fails, it could be isolated. Therefore, the probability of occurrence of equipment malfunction important to safety is not increased.

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\I Jumper/Lifted Lead g 1-94-030 Units 1 Component and Systems Affected!

Temporary Power Panel fed from Intake 480V MCC 1A3 Reason for tho Requests To provide power from 480V MCC 1A3 to a temporary power panel to support miscellaneous equipment for Unit 1 intake work.

Saf sty Analysis:

The jumper does not affect 1A3 LC protection or other electrical distribution protection for non-essential power from the startup transformer The response of the onsite electorial distribution system is not altered and therefore the consequences of an accident are not increased.

Protection of non-essential ties to safety related electrical distribution are not affected. No safety related loads exist on the 1A3 MCC.

This additional load on the 1A3 MCC does not interact significantly with the safety related systems and since the response of the onsite electrical distribution system from a fault is not altered.

Therefore there is no increase in the consequences of a malfunction of equipment important to safety as evaluated in the UFSAR.

The response of the onsite electrical system protection is not affected by this jumper, therefore it does not create the possibility of an accident of a different type than previously evaluated in the UFSAR.,

There is no reduction of margin as defined in the basis of any technical specification because 1A3 MCC is part of the non-essential onsite power distribution system.

138

Jumper/Lifted Lead g 1-94-031'omponent and Systems Affected'emporary Power Panel fed from Intake 480V MCC 1B3 Reason for the Request:

To provide power from 480V MCC 1B3 to a temporary power panel to support miscellaneous equipment for Unit 1 intake work.

Safety Analysis:

The jumper does not affect 1B3 LC protection or other electrical distribution protection for non-essential power from the startup transformer The response of the onsite electorial distribution system is not altered and therefore the consequences of an accident are not increased.

P rotection of non-essential ties to safety related electrical distribution are not affected. No safety. related loads exist on the 1B3 MCC.

This additional load on the 1B3 MCC does not interact significantly with the safety related systems and since the response of the onsite electrical distribution system from a fault is not altered.

Therefore there is no increase in the consequences of a malfunction of equipment important to safety as evaluated in the UFSAR.

The response of the onsite electrical.,system -protection is not affected by this jumper, therefore possibility of an accident of a different type than previously it does not create the evaluated in the UFSAR.

There is no reduction of margin as defined in the basis of any technical specification because 1B3 MCC is part of the non-essential onsite power 'distribution system.

139

Jumper/Lifted Lead g 2-94-039 Unite 2 Component and Systems Affected:

Fire Header and Circulating Water Pump Seal Water Reason for the Request:

Provide a backup supply of seal water from the fire header to the circulating water pumps while completing work on the service water system.

Saf ety Analysis:

This J/LL does not increase the probability of occurrence of an accident postulated in the UFSAR. The fire protection system is no postulated to cause an accident, therefore this jumper will not affect accident probability.

This J/LL does not increase the consequences of an accident previously evaluated in the UFSAR. This alternate lineup will not impact the fire protection system from performing its design basis function.

This J/LL does not increase the probability of an occurrence of a malfunction of equipment important to safety. The fire protection system is designed with multiple isolation valves which will be isolated in the event of a manifold rupture.

This J/LL does not increase the consequences of a malfunction of equipment important to safety. The alternate arrangement does not create new failure modes because the system is designed with isolation valves.

This J/LL does not create the possibility of an accident different than described in the UFSAR. New failures are not created as a function of this jumper.

This J/LL does not increase the possibility of malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR. The fire protection system is equipped with isolation capability to ensure the entire system is not lost due to a single rupture point.

This J/LL does not reduce the margin of safety as defined in the basis for any technical specification. The fire protection system is not a technical specification system.

140

Jumper/Lifted Lead.t 2-94-052 Units 2 Component and Systems Affected<

Pressurizer Heater Banks B-1 and B-2 Reason for the Requests Pressurizer Heater Bank B-1 circuit g4 has a ground on it. With B-1 and B-4, the banks that can be loaded on the EDG, a heater from bank B-2 will be'jumpered to the B-1 bank.

Safety Analysis:

This J/LL does not increase the probability of occurrence of an accident postulated in the UFSAR. The heater banks are not postulated to cause and accident as described in the UFSAR.

This J/LL does not increase the consequences of an accident previously evaluated in the UFSAR. The pressurizer heaters. are not mentioned in the UFSAR analysis section 15.1.

This J/LL does not increase the probability of an occurrence of a malfunction of equipment important to safety. Section 8.3 of the UFSAR states the power panels form the first level of protection for a short circuit at the heaters. Therefore, a short would not endanger the safety'related buses.

This J/LL does not increase the consequences of a malfunction of equipment important to safety. A short circuit is protected against by the power panels which are the first, level of protection.

This J/LL does not create the possibility of an accident different than described in the UFSAR. No new hazards are created as a function of this jumper.

This J/LL does not increase the possibility of malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR. The jumper does not affect the control circuitry nor the class 1E to non-safety related interface, therefore, no new hazards are created.

This J/LL does not reduce the margin of safety as defined in the basis for any technical specification. The jumper restores the 150 KW to the B-1 heater bank.

141

I Jumper/Lifted Lead g 1-94-054 Unit! 1 Component and Systems Affected!

Temporary Substitution of Temperature Indicators for containment Temperature Recorder TR-25-1 Reason for the Requests The purpose of the Jumper is to provide a temporary means of monitoring containment temperature in compliance with Technical Specification surveillance requirement 4.6.1.5. Normally this is done with TR-25-1 which is temporarily out of service. The temperature inputs from thermocouples TE-25-3, TE-25-5 and TE-25-7 are functioning normally such that they may be used with another indicating source to provide accurate monitoring of this parameter.

Safety Evaluation:

This jumper does interface with any components or equipment other than for the purposes of monitoring. Non of the equipment involved is evaluated for possible malfunctions in the UFSAR.

The jumper does not affect any structures utilized in mitigating accident consequences as evaluated in the UFSAR.

The jumper does not affect any structures utilized in mitigating accident consequences as evaluated in the UFSAR.

The change does not affect the assumptions of containment temperatures credited in the UFSAR.

There is no possibility of a new malfunction of equipment important to safety 142

St. Lucie Units

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1 and 2 Docket Nos. 50-335 and 50-389

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99 ua 0 erat e ort SECTION 1 TECHNICAL SPECIFICATION 6.9.1.5 ST. LUCIE Ul&TS 1 AIM2 1994 MAN-REM REPORT

QQQ F L 0 R I D A P Q R 8 L I G H T RADI AT ION EXPOSURE MONITORING SS CONTROL SYSTEM - REMACS 1994 MAN-REM PORT STANDARD FORMAT FOR REPORTING NUMBER OF PERSONNEL S: MAN-REM FOR WORK JOB FUNCTIONS NUMBER OF PERSONNEL > 100 NREM TOTAL MAN / REM WORK AND J OB F UNCT ION STATION UTILITY CONTRACT STAT ION UTILITY CONTRACT REACTOR OP ERAT IONS 8 SURV ENGINEERING 0 1 000 000 000. 120 000. 153 HEALTH PHYSICS 20 16 005.320 000. 010 005.135 MAINTENANCE 0 1 000.301 000. 01 6 OO1 .O48 OPERATIONS 31 1 007.966 000 533 002 367 SUPERVISOR 0 0 OOO.OIZ 000.000 ooo.ooo ROUT INE MA IN TENANCE ENGINEERING 0 2 8 000 000 001 529 001 633 HEALTH PHYSI CS 33 0 100 011 .542 OOO.OOT 037 . 037 MAINTENANCE Z10 222 066 1 27

~ 001.781 064 .925 13 8 49 OOZED'76 005 003 471 014.873 OPERAT IONS SUPERVISOR 0 0 o OOO.O15 oooiooo 000.QQQ INSERV ICE INSPECTION ENGINEERING 0 000.000 000. 586 002.779 HEALTH PHYSI CS 0 0 000.051 000. 000 ooo.ooo MAINTENANCE 2 26 380 000. 480 008.125 OPERATIONS 2 5 000.911 ooo.44o 004.036 SUPE RV I SOR 0 0 ooo.ooo ooo.ooo ooo.ooo SPECIAL MAINTENANCE ENGINEERING 0 1 2 ooo.ooo 000 '12 ooo.55o HEALTH PHYSI CS 3 0 1 001.0ZQ 000.0QQ 000.217 MAINTENANCE 59 0 154 018 '5Z 000 ~ 21 1 062 6 19

~

OPERATIONS 0 5 86 000 273 001. 307 051 .221 SUPERV ISOR 0 Q Q ooo.ooo ooo.ooo ooo.ooo WASTE PROCESSING ENGINEERING 0 0 000.000 000. 000 000.000 HEALTH PHYSICS 0 37 003 144 000.000 010.746 MAINTENANCE 0 7 000.494 000.000 001 .931 OPERAT IONS 0 000 '06 000.000 006.901 SUPERVISOR 0 0 000.000 000.000 ooo.ooo REFUEL ING ENGINEERING 0 1 2 000 000 000.769 000.251 HEALTH PHYSICS 6 0 2 001 .603 ooo.ooz 001 .596 MAINTENANCE 128 2 14 050.147 000.737 005 . 736 OPERATIONS 16 1 11 005.358 000 612 003. 739 SUPERVISOR 0 0 0 000.005 000. 000 000.000 TOTALS ENG I NEER ING 0 8 17 000.000 003.416 005.366 HEALTH PHYSICS 57 0 150 022.680 000.019 054.731 MAINTENANCE 269 8 376 137.901 003.225 144.384 OPERAT IONS 67 13 135 O19 69O 006.363 083 ~ 137 SUPERVISOR 0 0 0 000 032 000.000 OOO.OOO GRAND TOTALS 393 678 1 80. 303 013. QZ3 287. 618

St. Lucie Units

~ 1 and 2 Docket Nos. 50-335 and 50-389

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1994 A nua erat Re ort SECTION 2 TECKNICALSPECIFICATION 4.7.6.1.2 ST. LUCIE Ul'Kl'S 1 &6) 2 MANGROVE PHOTOGRAPHIC SURVEY RESULTS Based on the evaluation of the false color infrared photograph taken on July 8, 1994, the condition of the mangrove trees situated between the intake and discharge canals (Impoundment 8E) continue to show additional plant growth throughout most of the 50 acre impoundment. There is approximately a nine (9) percent increase from last year's (1993) aerial evaluation. This increases the mangrove coverage to sixty-one (61) percent. The coverage in impoundment 8E is still below the 1975 baseline condition, however, with the continuation of the current management efforts, the mangroves should continue to improve in their health and vitality as well as provide additional plant growth to the area.

St. Lucie Units 1 and 2

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~ ~

Docket Nos. 50-335 and 50-389 A

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a 0 er t'~

o SECTION 3 TECHMCAI SPECIFICATION 6.9.1.5 AZ6) 3.4.S CHEMISTRY RESVI TS In accordance with Technical Specification 6.9.1.5, the primary coolant specific activity did not exceed the limits of Technical Specification 3.4.8.

St. Lucie Units

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~ ~

1 and 2 Docket Nos. 50-335 and 50-389

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994 A u 0 e a in Re ort SECTION 4 10 CFR 50.59 FRG APPROVED SAX<ETY EVALUATIONS FOR CALENDAR YEAR 1994

Safety Evaluations reportable pursuant to 10 CPR 50.59 for St. Lucie Units 1 f 2 Number Revision Title JPN-PSL-SEMJ-90-039 Installation of Blind Flange on Outlet of Purge Exhaust Valve FCV-25-6 PN-PSL-SEFJ-92-009 St. Lucie Plant Control Element Assembly Operational Life Determination PN-PSL-SECJ-93-002 1-2 Specification SPEC-C-013; Installation Guidelines for Miscellaneous Non-System Related Items on Existing Structures PN-PSL-SEMS-93-010 0-1 Installation of Hydro Plugs into the ICI Penetrations for Locations J7 and G18 PN-PSL-SECJ-93-011 Safety Evaluation for Specification SPEC-C-005, Component Mounting and Supports PN-PSL-SECJ-93-012 Safety Evaluation for Specification SPEC-C-019, Tubing and Tubing Supports PN-PSL-SECJ-93-014 Safety Evaluation for Specification SPEC-C-017 Small Bore Piping Supports PN-PSL-SEFJ-93-014 0-2 Safety Evaluation of Spent Fuel Pool (SFP) Coupon Surveillance Program PN-PSL-SEFJ-93-024 0 Loss of Feedwater Transient with Corrected Steam Generator Inventory Error PN-PSL-SENP-93-035 Evaluation of Inventory Loss for the Refueling Water Tank PN-PSL-SEMP-94-001 Temporary Installation of Strain Measurement Devices on the Pressurizer Relief Valve Discharge Piping PN-PSL-SEFJ-94-002 Evaluation of Reduced Shutdown Cooling Flow Rate for St. Lucie Unit 2 Refueling Outage 0

Safety Evaluations reportable pursuant to 10 CFR 50.59 for St. Lucie Units 1 f 2 Number Revision Title JPN-PSL-SEIS-94-005 Operation of the Wide Range Containment Level Monitoring Channels L-07-13A & L-07-13B with Inoperable Sensors JPN-PSL-SENP-94-005 Shutdown Cooling Suction Valve Interlock Design JPN-PSL-SEMS-94-008 Gasket Leak Repair for Shutdown Cooling Return Isolation Valve V3480 JPN-PSL-SENS-94-010 Evaluation for Alternate ECCS Valve Alignment to Repair Line 3/4-SI-121 JPN-PSL-SEMS-94-011 Pressurizer Spray Bypass Valve V1454 Needle Tip Failure JPN-PSL-SEMS-94-013 2-3 Freeze Seal Application for V3480 on the 1A Shutdown Cooling Return Line JPN-PSL-SENS-94-015 Safety Evaluation for Service Water System Modifications JPN-PSL-SENP-94-017 Disabling the Steam Dump and Bypass Control System Quick Open Feature for Load Reduction JPN-PSL-SENS-94-018 Safety Evaluation for Hypochlorite System Modifications JPN-PSL-SENP-94-019 0-1 Alternative Valve Position for Spray Header Isolation Valve I-FCV-07-1B JPN-PSL-SEFJ-94-021 RTD Response Time Limit Increase From 8.0 Seconds to 14.0 Seconds JPN-PSL-SENP-94-021 0-1 Removing the Automatic Control Function for I-TCV-14-4A JPN-PSL-SENS-94-025 Safety Evaluation for Fuel Handling Equipment UFSAR Discrepancies JPN-PSL-SEMS-94-028 Installation of a Blind Flange on the Inlet of Containment Purge Valve FCV-25-1 JPN-PSL-SENP-94-029 Shutdown Operations Criteria for Reduced Inventory and Draining the Reactor Coolant System

Safety Evaluations reportable pursuant to 10 CPR 50.59 for St. Lucie Units 1 f 2 Number Revision Title JPN-PSL-SENP-94-037 SIT Discharge/Loop Check Valve Stroke Test JPN-PSL-SENP-94-039 Jumper/Lifted Lead for PDIS-2216 JPN-PSL-SEFJ-94-040 Removal of TE-1122CD Input from Channel "D" of the RPS for PSL 1 JPN-PSL-SENP-94-043 Safety Evaluation Temporary Removal of the ICW Pump Missile Shield JPN-PSL-SENP-94-044 Safety Evaluation for the use of Devoe Devran 140 Epoxy Compound and Kansai Biox as a Coating System for the St. Lucie Unit 1 Intake Structure JPN-PSL-SENP-94-047 SIT Discharge/Loop Check Valve Stroke Test JPN-PSL-SEMP-94-050 Temporary Alterations to the Refueling Water Tank JPN-PSL-SENP-94-065 Containment Air Conditioning for Refueling Outage JPN-PSL-SEEP-94-066 Safety Evaluation for Operation of Three Charging Pumps JPN-PSL-SEMP-94-076 Increase of Engineered Safeguards Suction Piping Design Pressure PROCEDURE SAFETY EVALUATIONS HP-74 0 Personal Dosimetry Procedure 2-LOI-0-65 Liquid Waste Management System Procedure 1-ONP-01 05 F Shutdown cooling Off-Normal Procedure 2-LOI-T-88 Fuel Pool Purification System Temporary Procedure JUMPER AND LIFTED LEAD SAFETY EVALUATIONS 2-94-007 Upender Vertical Circuit Limit Switch (2LS-BV) 1-94-019 Nitroen to Sodium Hydroxide (NaOH)

Tank 1-94-020 Circulating Water Pumps

Safety Evaluations reportable pursuant to 10 CFR 50.59 for St. Lucie Units 1 f 2 Number Revision Title 1-94-030 Temporary Power Panel 1-94-031 Temporary Power Panel 2-94-039 Circulating Water Pump Seal Water 1-94-046 See JPN-PSL-SEFZ-94-040 above 2-94-052 Pressurizer Heater Banks 1-94-054 Temperature Indicator Substitution

Unit: JPN-PSL-SEKT-90-039

Title:

Installation of Blind Flange on Outlet of Purge Exhaust Valve FCV-25-6 ABSTRACT This temporary change will allow the installation of a blind flange outside of the containment purge discharge isolation valve 2-FCV-25-6. This blind flange will act as a containment isolation device replacing 2-FCV-25-5 which has recently failed its LLRT.

This containment isolation system provides the means of isolating fluid systems that pass through containment penetrations such that any radioactivity that may be released to the containment atmosphere following a postulated Design Basis Accident (DBA) is confined. As such this temporary alteration performs a safety related function and this report and its associated modification are considered to be safety related.

This change does not affect the function of the 48" containment purge exhaust system during plant-power operations as the system is not used in modes 1, 2, 3 and 4.

Revision 5 deletes reference to the fuel cycle, to allow temporary installation of a blind flange as a containment isolation device in the future and adds references to the evaluation. This revision also deletes the requirement for closure of the penetration bleed-off line (reference valve I-V25208), because the Unit 2 UFSAR Section 6.2.3.3.1 and Figure 9.4-9 require that the containment penetrations for the Containment Purge System (P-10 and P-11) each contain an open 3/4" bleed-off line in to the shield building annulus.

This change does not involve an unreviewed safety question or require a change to the Technical Specification.

SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.

The Containment Isolation System provides the means of isolating fluid systems that pass through containment penetrations so that any radioactivity that may be released into the containment atmosphere following a postulated Design Basis Accident (DBA) is confined. The containment isolation valves, the penetration and the piping are designated Seismic Category I and designed to ASME Code, Section III and Quality Group B requirements.

Based on the above description, this report and its associated modifications are considered Nuclear Safety Related. This report does not involve an unreviewed safety question, and the following are bases for this conclusion:

) 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 containment purge exhaust system is neither required to operate during normal plant operation nor after a Design Basis Accident. However, it does perform a containment isolation function. The installation of a temporary blind flange on the exhaust side of valve FCV-25-6 provides the second isolation boundary. Extending the containment isolation boundary into the HVAC room does not affect the environmental qualification of equipment or decrease the potential for off-site dose release during a DBA. The piping associated with valve FCV-25-6 and Penetration P-10, the weld between the closure plate and shield building anchor plant ring and penetration sleeve were evaluated for the additional seismic and dead weight loads of the newly designed flange and existing valve FCV-25-6 associated piping. They were found to be adequate for the additional loads. Application of sealant in the valve packing of valves FCV-25-5 and 6 will not adversely affect the, normal function of the valves and will enhance the ability of the valves to perform its design function.

ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created because the containment purge exhaust system is neither required to function following a postulated Design Basis Accident nor is it required for the operational design of any system. The blind flange performs the passive function of containment isolation and does not adversely impact any safety related equipment.

iii) The margin of safety as defined in the bases Specification is not reduced since the for any Technical blind flange provides the second isolation boundary. This is consistent with the requirements of the Technical Specifications.

This evaluation constitutes, per 10CFR50.59(b) the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question or requires a change to the

'echnical specifications, thus prior NRC approval for this temporary modification is not required. The temporary blind flange on the outlet of Purge Valve FCV-25-6 is acceptable in that it replaces the isolation function of valve FCV-25-5. The enhancement of valve packing on valves FCV-25-5 and 6 provide additional assurance for the leak tightness. This change does not affect the function of the 48" containment purge exhaust system during the plant-power operation because the system is not used in modes 1, 2, 3, and 4.

Unit: 1 & 2 JPN-PSL-SEFZ-92-009

Title:

St. Lucie Plant Control Element Assembly Operational Life Determination ABSTRACT The UFSARs for St. Lucie Units 1 and 2 imply that the operational life of a Control Element Assembly (CEA) is ten years. The vendor who manufactured the CEAs (Asea Brown Boveri/Combustion Engineering) clarified that this limit is ten calendar years and corresponds to approximately eight effective full power years.

This limit is based on a different design CEA than that currently used at St. Lucie. The purpose of this evaluation is to utilize CEA inspection data taken specific to both St. Lucie Units 1 and 2 to determine the operational life of the CEAs being used.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an 1 ~ Does the proposed activity increase the probability of an accident previously evaluated in the UFSARs?

The life limits determined in this evaluation for the PSL-1 and PSL-2 CEAs do not change the overall configuration of either plant. The mode of operation of the plants remains unchanged. No equipment important to safety is affected.

Therefore, the probability of occurrence of an accident previously evaluated in either of, the .UFSARs is not changed.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSARs?

The life limits presented in this evaluation for the PSL-1 and PSL-2 CEAs have been determined to ensure that both the CEAs currently being used in each reactor and all future replacement CEAs of the same design are capable of performing their intended function over those stated lives. This intended function is to fully insert upon receiving a reactor scram signal. Implementation of the replacement schedules developed in this evaluation will ensure that all CEAs are discharged prior to exceeding the stated design life limits.

As such, the consequences of an accident previously evaluated in either of the UFSARs .that are mitigated by a reactor scram are not increased.

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSARs?

Modifying the design life of the CEAs does not impact any other equipment important to safety, nor interfere with the function of any other equipment important to safety. The life limits presented in this evaluation for the PSL-1 and PSL-2 CEAs have been determined to ensure that both the CEAs currently being used in each reactor and all future replacement CEAs of the same design are capable of performing their intended function over those stated lives.

Implementation of the replacement schedules developed in this safety evaluation will ensure that all CEAs are discharged prior to exceeding the stated design life limits. The probability of a CEA malfunction is not increased. Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in either of the UFSARs is not changed.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSARs?

Modifying the design life of the CEAs does not require interaction with any equipment important to safety, or prevent any functions of other equipment important to safety.

The life limits presented in this evaluation for the PSL-1 and PSL-2 CEAs have been determined to ensure that both the CEAs currently being used in each reactor and all future replacement CEAs of the same design are capable of performing their intended function over those stated lives.

Implementation of the replacement schedules developed in this evaluation will ensure that all CEAs are discharged prior to exceeding the stated design life limits. The CEAs will perform their intended function over their design life given a malfunction of other equipment important to safety.

Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in either of the UFSARs are not increased.

5~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR's?

Modifying the design life of the CEAs does not require interaction with any equipment important to safety, or interfere with any functions of other equipment important to safety. The life limits determined in this evaluation for the PSL-1 and PSL-2 CEAs do not change the normal operation of

f either of the plants. These'ife limits have been determined to ensure that both the CEAs currently being used in each reactor and all future replacement CEAs of the same design are capable of performing their intended function over those stated lives. Therefore, the possibility of an accident of a different type other than those previously evaluated in either of the UFSARs is not created.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSARs?

Modifying the design life of the CEAs does not introduce any new physical interactions with equipment important to safety.

The life limits presented in this evaluation for the PSL-1 and PSL-2 CEAs have been determined to ensure that both the CEAs currently being used in each reactor and all future replacement CEAs of the same design are capable of performing their intended function over those stated lives.

Implementation of the replacement schedules developed in this evaluation will ensure that all CEAs are discharged prior to exceeding the stated design life limits. Therefore, the possibility of a malfunction of equipment important to safety of a different type of accident other than those previously evaluated in either of the UFSARs is not created.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification'P The life limits determined in this evaluation for the PSL-1 and PSL-2 CEAs ensure that both the CEAs currently being used in each reactor and future replacement CEAs of the same design are capable of performing their intended function. As such, the margin of safety defined in the basis for any Technical Specification is not reduced.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

Unit: 1 & 2 ZPN-PSL-SECJ-93-002

Title:

Specification SPEC-C-013; Installation Guidelines for Miscellaneous Non-System Related Items on Existing Structures ABSTRACT Nuclear Engineering Specification SPEC-C-013, entitled "Installation Guidelines for Miscellaneous Non-System Related Items on Existing Structures," has been developed to provide generic guidance for the installation of storage racks, fire extinguisher, storage cabinets, and other miscellaneous items, which are not part of a plant system. These guidelines can be used directly for those installations covered by the support details provided in Appendix B of the specification, or in conjunction with additional guidance provided by Engineering via the "Request for Specification Clarification or Change" form provided in Appendix C.

This evaluation will provide the basis for the acceptability of using the Specification in the maintenance process, in lieu of the current practice which requires that a Plant Change or Modification (PC/M) package be issued and implemented for such cases.

also demonstrate that the Specification meets all technical and It will licensing requirements for St. Lucie Units 1 & 2.

This safety evaluation concludes that the use of the Specification will meet all technical and licensing requirements and will have no adverse impact on plant operations. It is also concluded that the use of the Specification will not compromise the safety and licensing bases for St. Lucie Units 1 & 2. As such, an unreviewed safety question does not exist and a change to the Technical Specifications is not required.

Revision 1 of this Safety Evaluation incorporates the following:

The materials section of this evaluation has been rewritten to denote that Appendix A of the specification provides restrictions only for those details in Appendix B. All other installations require an Engineering evaluation.

Reference to a Civil Calculation has been changed to "latest revision" to allow changes in the corresponding calculation without requiring a revision to the Safety Evaluation when affected.

it is not The UFSAR and Technical Specification amendments have been updated.

11

References to specification sections have been updated to support revisions to the specification format. A section has been added to identify affected documents.

Revision 2 of this Safety Evaluation incorporates the revision 2 of the Spec No. SPEC-C-013 which clarifies the definition of Miscellaneous Item in Sect 1.4.1 of the specification and adds a Figure to Appendix B of Specification.

The revisions summarized above do not affect the conclusions of this Safety Evaluation. This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an 1 ~ Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the safety analysis report?

Implementation of the guidelines and support details delineated within the Specification will result in installations that meet all UFSAR requirements for the applicable installation safety classification. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR has not been increased.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the safety analysis report?

As evaluated in the design bases section of this evaluation, installations of miscellaneous items performed in accordance with the Specification will not adversely affect the UFSAR accident analysis. Therefore, the consequences of an accident previously evaluated in the UFSAR have not been increased.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety, analysis report?

The installation guidelines and support details delineated within the Specification meet all UFSAR requirements for the applicable installation safety classification. This will assure that adverse interaction with equipment important to 12

safety is precluded. Therefore, the probability of a malfunction of equipment important to safety previously evaluated in the UFSAR has not increased.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the safety analysis report.

The installation guidelines and support details delineated within the Specification meet all UFSAR requirements for the applicable installation safety classification. This will assure that adverse seismic interaction with equipment important to safety is precluded. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the safety analysis report?

The installation of miscellaneous, non-system related, items in accordance with the Specification will preclude adverse interaction with existing equipment important to safety.

Therefore, the proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the safety analysis report.

Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the UFSAR?

The installation of miscellaneous, non-system related, items in accordance with the Specification will assure that equipment important to safety is not adversely affected.

Therefore, the proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the UFSAR.

7~ Does the proposed activity reduce the margin of safety as defined in the, basis for any Technical Specification?

Implementation of the Specification will not impact the Technical Specifications in any way.

installation guidelines and support, details delineated ofwithin The use the, the Specification will result in installations that meet all UFSAR requirements for. the applicable installation safety classification. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

13

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and 'does not adversely affect plant operation or safety.

14

Unit: 1 JPN-PSL-SEMS-93-010

Title:

Installation of Hydro Plugs into the ICI Penetrations for Locations J7 and G18 ABSTRACT This safety evaluation documents the acceptability of installing hydro plugs in the reactor head instrument flange penetrations for the location J7 and G18 incore detectors. These two detectors are not being installed because difficulties were experienced while attempting to insert the detectors into the conduits in the reactor head. The hydro plugs are similar to the normal instrument seal plugs, except that the hydro plugs are solid. The ICI (incore instrumentation) rhodium detectors do not perform a safety related function per UFSAR Section 4.2.2.2.8; however, they perform a significant monitoring function, assuring that operation remains within the requirements of Technical Specification. The ICI s are designed to fulfil their required function when at least 75% are available. The ICI detector assemblies do perform a safety related pressure boundary function, and house a Core Exit Thermocouple (CET), which performs a safety related function per UFSAR Section 7.5.4.2. Revision 0 of this evaluation is applicable to St. Lucie Unit 1 operating cycle 12 only because the plugs were to be replaced with functional ICI assemblies during the next refueling outage.

The installation of the hydro plugs will functionally abandon detector locations J7 and G18. Technical Specification 3.3.3.2 requires at least 75% of all incore detector locations to be operable and a minimum of two quadrant symmetric incore detector locations per core quadrant to be operable. Although the installation of the hydro plugs will reduce the number of available incore detectors, it will not reduce the number below that required by the technical specifications. The hydro plugs will perform a pressure boundary function identical to the existing detectors and are therefore designed in accordance with ASME Class 1 requirements. Thus, the integrity of the reactor coolant system will not be adversely affected.

The installation of the hydro plugs will result in the loss of two CETs; however, the available number of CETs and the inputs to the Qualified Safety Parameter Display System (QSPDS) will not be reduced to a number below that required by Technical Specification 3.3.3.8. This evaluation concludes that."the proposed configuration does not represent an unreviewed safety question and has no impact on plant safety or operations. A review of the technical specifications and the Safety Analysis Report has shown that there are no technical specification changes involved. Revision 0 of this evaluation was valid through the end of Cycle 12 operation.

15

( ~

Revision 1 to this safety evaluation extends the requirement to replace the hydro plugs with functional ICI assemblies to the refueling outage following cycle 14 operation. This will coincide

~

with the planned replacement of the ICI's.

~

This evaluation concludes that the proposed configuration does not represent an unreviewed safety question and has no impact on plant safety or operations. A review of the technical specifications and the Safety Analysis Report has shown that there are no technical specification changes involved. The analyses and conclusions of the original evaluation have been reviewed and remain valid through the end of Cycle 14 operation.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an 1~ Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased since the proposed replacement does not adversely affect any accident initiating components. The ICI detectors do not perform any active functions necessary for the safe shutdown of the plant and the proposed replacement does not create any new unmitigated failure modes for any equipment or systems capable of initiating an accident. The hydro plugs will perform a pressure boundary function identical to the existing detectors and are designed in accordance with ASME Class 1 requirements to ensure reliability. The removal of the ICI detectors and the subsequent introduction of water moderator will present an inconsequential effect on the local power distribution within the fuel assemblies at locations J7 and G18. The removal of this ICI detector will not affect the ability of the ICI system to perform its intended function of measuring the core power distribution.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased since the proposed replacement does not create a new path for.. uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function. The ICI detectors are not required to perform any active safety related functions and the proposed replacement does not adversely impact any equipment which is 16

required to perform a safety related function or initiate actuation of any safety systems. The hydro plugs are dimensionally equivalent to the normal detector assemblies.

The hydro plugs will perform a pressure boundary function identical to the existing detectors and are designed in accordance with ASME Class 1 requirements. The installation of the hydro plugs will result in the loss of two CETs; the J7 CET is located in quadrant 3 and provides a signal to channel B of the Qualified Safety Parameter Display System (QSPDS) and the G18 CET is located in quadrant 1 and provides a signal to QSPDS channel A. However, the available number of CETs will not be reduced below that which is required by Technical Specification 3.3.3.8.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased. No new unmitigated failure modes for any equipment important to safety are introduced by the proposed replacement and no new components or equipment are introduced that could adversely interact with any equipment important to safety. The hydro plugs perform a pressure boundary function identical to the existing detectors and are designed in accordance with ASME Class 1 requirements.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased since the proposed replacement does not create a new path for uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function.

The proposed replacement will not adversely impact any equipment which is required to perform an active safety related function or to initiate actuation of any safety systems.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The possibility of an accident of a different type than any evaluated previously in the UFSAR has not been created since the proposed replacement does not add or adversely affect any 17

tI equipment capable of initiating an accident. The proposed replacement does not present any new paths for the loss of reactor coolant system inventory since the hydro plugs are dimensionally equivalent to the normal detector. The hydro plugs will perform a pressure boundary function identical to the existing detectors and are designed in accordance with ASME Class 1 requirements. There are no new unmitigated failure modes for the hydro plugs. In addition, the detector assemblies are passive measurement devices, so their removal could not result in the initiation of an accident of a different type. The removal of the ICI detector and subsequent introduction of water moderator will present an inconsequential effect on the local power distribution within the fuel assemblies at locations J7 and G18. The removal of these ICI detectors will not affect the ability of the ICI system to perform its intended function of measuring the core power distribution.

Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the UFSAR?

The possibility of a malfunction of a different type than any evaluated previously in the UFSAR has not been created since the proposed replacement will not inhibit or otherwise adversely affect the operation of any equipment important to safety. The ICI detectors are passive measurement devices and are not required to perform an active safety related function or activate any safety related systems. The physical interfaces of the detector assembly are not affected by the proposed configuration. The hydro plugs will perform a pressure boundary function identical to the existing detectors and are designed in accordance with ASME Class 1 requirements.

There are no new unmitigated failure modes for the hydro plugs. The installation of the hydro plugs will result in the loss of two CETs; however, the two CETs are located in different quadrants and supply signals to different QSPDS channels. The available number of CETs will not be reduced below that which is required by Technical Specification 3.3.3.8. In addition, these hydro plugs do not create any new modes of operation for any safety related equipment.

Does the proposed activity reduce the margin of safety as defined in the basis for any technical specification?

The proposed activity does not reduce the margin of safety as defined in the basis for any technical specification since the removal of the incore detectors at locations J7 and G18 will not adversely impact the minimum number of incore detectors required for operation as defined in Technical Specification 18

~ 1 3.3.3.2. The hydro plugs will perform a pressure boundary function identical to the existing detectors and are designed in accordance with ASME Class 1 requirements. Thus, the integrity of the reactor coolant system will not be adversely affected. The installation of the hydro plugs will result in the loss of two CETs; the J7 CET is located in quadrant 3 and provides a signal to QSPDS channel B and the G18 CET is located in quadrant 1 and provides a signal to QSPDS channel A. Since the two CETs are located in different quadrants and supply signals to different QSPDS channels, the available number of CETs will not be reduced below that which is required by Technical Specification 3.3.3.8.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

19

Unit:~

1 6 2 JPN-PSL-SECJ-93-011

Title:

~

~ Safety Evaluation for Specification SPEC-C-005, Component Mounting and Supports ABSTRACT Nuclear Engineering Specification SPEC-C-005, entitled "Specification for Component Mounting and Supports, St Lucie Units 1 and 2, and Turkey Point Units 3 and 4", has been developed to provide generic component mounting instructions and support details. These mounting instructions and support details can be used in conjunction with design output documents presently used in the procurement/maintenance process (e.g., Procurement Technical Evaluation, Item Equivalency Evaluation, Plant Work Order) to install replacement components weighing less than 50 pounds.

This evaluation will provide the basis for the acceptability of using the Specification in the maintenance process, in lieu of the current practice which requires that a Plant Change or Modification (PCM) package be issued and implemented for such cases.

also demonstrate that the Specification meets all technical and It will licensing requirements for the St Lucie Units 1 and 2.

This safety evaluation concludes that the use of the Specification will meet all technical and licensing requirements and will have no adverse impacts on plant operations. It is also concluded that the use of the Specification will not compromise the safety and, licensing bases for St Lucie Units 1 and 2. As such, an unreviewed safety question does not exist and a change to the Technical Specifications is not required.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the safety analysis report?

The use of mounting instructions and .support details delineated within the Specification will result in installations that meet all UFSAR requirements for the applicable installation safety classification. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR has not been increased.

20

Does the proposed activity increase the consequences of an accident previously evaluated in the safety analysis report?

As evaluated in the design basis and analysis section of this evaluation, all mounting or support modifications performed in accordance with the Specification will not adversely affect the UFSAR accident analysis. Therefore, the consequences of an accident previously evaluated in the UFSAR have not been increased.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report?

The mounting instructions and support details delineated within the Specification meet all UFSAR requirements for the applicable installation safety classification. This will assure that equipment important to safety is properly supported under all design basis loading conditions, and that adverse seismic interaction with equipment important to safety is precluded. Therefore the probability of a malfunction of equipment important to safety previously evaluated in the UFSAR has not increased.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the safety analysis report?

The mounting instructions and support details delineated within the Specification meet all UFSAR requirements for the applicable ins'tallation safety classification. This will assure that equipment important to safety is properly supported under all design basis loading conditions, and that adverse seismic interaction with equipment important to safety is precluded. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the safety analysis report?

Installation of replacement supports and remounting of equipment in accordance with the Specification will assure that equipment important to safety. is appropriately supported under all design basis loading conditions. It will also preclude adverse seismic interaction with existing equipment important to safety. Therefore, the possibility of an accident of a different type than any previously evaluated in the safety analysis report has not been created.

21

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the safety analysis report'P Installation of replacement supports and remounting of equipment in accordance with the Specification will assure that equipment important to safety is appropriately supported under all design basis loading conditions.

preclude adverse seismic interaction with It existing will also equipment important to safety. Therefore,, the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the UFSAR has not been created.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The use of the Specification will not impact the Technical Specifications in any way. The use of mounting instructions and support details delineated within the Specification will result in installations that meet all UFSAR requirements for the applicable installation safety classification. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

22

~ ~

Unit:~ 1 & 2 JPN-PSL-SECJ-93-012

Title:

~

~ Safety Evaluation for Specification SPEC-C-019, Tubing and Tubing Supports ABSTRACT Nuclear Engineering Specification SPEC-C-019, entitled "Specification for Tubing and Tubing Supports, St. Lucie Units 1 and 2", has been developed to provide generic installation instructions and support details for Safety Related, Quality Related and Non-Nuclear tubing. These installation instructions and support details can be used as specified within Engineering output design documents such as the Plant Change or Modification (PC/M). Additionally, maintenance repair/replacement activities may be performed on tubing supports using the specified standard supports for those installations directly covered by the specification criteria or in conjunction with additional guidance provided by Engineering via the "Request for Specification Clarification or Change" form provided in Appendix A.

This evaluation will provide the basis for the acceptability of using the specification in the maintenance process, in lieu of the current practice which requires that a PC/M package be issued and implemented for such cases. It will also demonstrate that the specification meets technical and licensing requirements for the St. Lucie Units 1 and 2.

This safety evaluation concludes that the use of the will meet technical and licensing requirements and specification will have no adverse impacts on plant operations. It is also concluded that the use of the specification will not compromise the safety and licensing bases for St. Lucie Units 1 and 2. As such, an unreviewed safety question does not exist and a change to the Technical Specifications is not required.

SAFETY EVALUATION Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the safety analysis report?

The use of installation instructions and support details delineated within the specification will result in installations that meet UFSAR requirements for the applicable installation safety classification. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR has not been increased.

23

Does the proposed activity increase the consequences of an accident previously evaluated in the safety analysis report?

As evaluated in the design basis and analysis section of this evaluation, all support modifications performed in accordance with the specification will not adversely affect the UFSAR accident analysis. Therefore, the consequences of an accident previously evaluated in the UFSAR have not been increased.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report?

The installation instructions and support details delineated within the specification meet UFSAR requirements for the applicable installation safety classification. This will ensure that equipment important to safety is properly supported under all design basis loading conditions, and that adverse seismic interaction with equipment important to safety is precluded. Therefore the probability of a malfunction of equipment important to safety previously evaluated in the UFSAR has not increased.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to 'safety previously evaluated in the safety analysis report?

The installation instructions and support details delineated within the specification meet UFSAR requirements for the applicable installation safety classification. This will ensure that equipment important to safety is properly supported under all design basis loading conditions, and that adverse seismic interaction with equipment important to safety is precluded. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the safety analysis report?

Installation of replacement supports in accordance with the specification will assure that equipment important to safety is appropriately supported under all design basis loading conditions. It will also preclude adverse seismic interaction with existing equipment important to safety. Therefore, the possibility of an accident of a different type than any previously evaluated in the safety analysis report has not been created.

24

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the safety analysis report?

Installation of replacement supports in accordance with the specification will assure that equipment important to safety is appropriately supported under all design basis loading conditions. It will also preclude adverse seismic interaction with existing equipment important to safety. Therefore, the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the UFSAR has not been created.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The use of the specification will not impact the Technical Specifications in any way. The use of installation instructions and support details delineated within the specification will result in installations that meet UFSAR requirements for the applicable installation, safety classification. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

The foregoing constitutes the determination, per 10 CFR that the subject activity does not involve an unreviewed50.59(b),

safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

25

Unit:

~

1 & 2 JPN-PSL-SECJ-93-014

Title:

~ Safety Evaluation for Specification SPEC-C-017 Small Bore Piping supports ABSTRACT Nuclear Engineering Specification SPEC-C-017, entitled "Procurement, Fabrication and Installation of Small Bore Pipe Supports," has been developed to provide generic .installation instructions and support details for Safety Related and Quality Related small bore piping systems. These installation instructions and support details can be used as specified within Engineering output design documents such as the Plant Change or Modification (PC/M). Additionally, maintenance repair/replacement activities may be performed on small bore pipe supports using the specified standard supports in conjunction with additional guidance provided by Engineering via the "Maintenance Request Approval" (MRA) form provided in Appendix B.

This evaluation will provide the basis for the acceptability of using the specification in the maintenance process in lieu of the current practice which requires that a PC/M package be issued and implemented for such cases. It will also demonstrate that the specification meets all technical and licensing requirements for St. Lucie Units 1 & 2.

This safety evaluation concludes that the use of the specification will meet all technical and licensing requirements and will have no adverse impact on plant operations. It is also concluded that the use of the specification will not compromise the safety and licensing bases for .St. Lucie Units 1 and 2. As such, an unreviewed safety question does not exist and a change to the Technical Specifications is not required.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an 1 ~ Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the safety analysis report?

The use of installation instructions and support details delineated within the specification will result in installations that meet all UFSAR requirements for the applicable installation safety classification. Therefore, the

\

26

probability of occurrence of an accident previously evaluated in the UFSAR has not been increased.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the safety analysis report'?

As evaluated in the design basis and analysis section of this evaluation, all support modifications performed in accordance with the specification will not adversely affect the UFSAR accident analysis. Therefore, the consequences of an accident previously evaluated in the UFSAR have not been increased.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report?

The installation instructions and support details delineated within the specification meet all UFSAR requirements for the applicable installation safety classification. This will assure that equipment important to safety is properly supported under all design basis loading conditions, and that adverse seismic interaction with equipment important to safety is precluded. Therefore the probability of a malfunction of equipment important to safety previously evaluated in the UFSAR has not increased.

4. Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the safety analysis report?

The installation instructions and support details delineated within the specification meet all UFSAR requirements for the applicable installation safety classification. This will assure that equipment important to safety is properly supported under all design basis loading conditions, and that adverse seismic interaction with equipment important to safety is precluded. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased.

5. Does the proposed activity create the possibility of an accident of a different type that any previously evaluated in the safety analysis report?

Installation of replacement supports in accordance with the specification will assure that equipment important to safety is appropriately supported under all design basis loading conditions. It will also preclude adverse seismic interaction with existing equipment important to safety. Therefore, the possibility of an accident of a different type than any 27

previously evaluated in the safety analysis report has not been created.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the UFSAR?

Installation of replacement supports in accordance with the specification will assure that equipment important to safety is appropriately supported under all design basis loading conditions. It will also preclude adverse seismic interaction with existing equipment important to safety. Therefore, the possibility of.a malfunction of equipment important to safety of a different type other than any previously evaluated in the UFSAR has not been created.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The use of the specification will not impact the Technical Specifications in any way. The use of installation instructions and support details delineated within the specification will result in installations that meet all UFSAR requirements for the applicable installation safety classification. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

28

0 Unit: 1 -JPN-PSL-SEFJ-93-014

Title:

Safety Evaluation of Spent Fuel Pool (SFP) Coupon Surveillance Program High density fuel storage racks were installed into the spent fuel pool (SFP) of St. Lucie Unit 1 in 1988. The design of the high density racks includes use of the strong neutron absorption material, Boraflex, to'maintain the SFP in a subcritical condition.

Boraflex panels were installed between spent fuel assemblies in the SFP racks to control the neutron multiplication factor. To monitor the Boraflex panel in-service performance, a surveillance program, "SFP Boraflex Coupon Surveillance," has been implemented to study Boraflex degradation mechanisms with time.

A technical review of the SFP Boraflex coupon surveillance program at St. Lucie Unit 1 was performed and completed. Review of the results has indicated that the coupon surveillance program has little merit with respect to practical applications to the in-service performance of Boraflex panels in racks. The St. Lucie Unit 1 ongoing surveillance program will be suspended, once the new improved program is implemented.

An enhanced Boraflex panel verification program consisting of a technique for predicting Boraflex gamma dose and performing periodic blackness testing has been developed to replace the existing in-service Boraflex coupon surveillance program. Details of the program are described in this evaluation.

Revision 1 of this Safety Evaluation was added to specifically address the removal of the coupon surveillance program and provided an unreviewed safety question determination. In addition, some minor wording changes have been added to clarify information associated with the coupon surveillance program. Nevertheless, the previous conclusion and recommendations are still valid for this revision.

Revision 2 of this Safety Evaluation has been made to extend the blackness test performance date to 1995. The original conclusion and reco'mmendations are still valid for revision 2.

This evaluation concludes that the activity as described above does not represent an unreviewed safety ..question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

29

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine if the subject activity constitutes an unreviewed safety question:

Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The verification program contains the use of the blackness test which was previously approved by the NRC staff. Based on the past test at PSL1, fuel movements in the SFP are needed to make room for testing. However, this is a routine operation.

The proposed blackness testing is not an accident initiator and, therefore, the probability of occurrence of an accident evaluated in the UFSAR is not increased.

The purpose of the coupon program which will be eliminated is to provide surveillance of the Boraflex used in the SFP. The coupon program is a passive surveillance program and has no function which can impact the probability of occurrence of an accident previously evaluated in the UFSAR. Therefore, the removal of the coupon surveillance program will not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The proposed activity can involve two independent accidents that can result in radiological activity releases. The first accident is a fuel handling accident previously evaluated in the UFSAR. The postulated accident'ue to fuel mishandling can result in fission product releases. The blackness test operation does not affect the fission product release mechanism during the accident. The second accident is a drop of the source container on top of the storage cells. Review of the fuel assembly drop analysis has shown that the falling source container would not damage the stored fuel due to the small weight of the container in water. The worst case from a drop of the source container would be radioactivity releases in water from the ~'Cf source. But, the level of the source radioactivity release is low and is bounded by the fuel handling accident. Therefore, the proposed activity will not increase the consequences of an accident previously evaluated in the UFSAR.

The coupon program which will be eliminated is used to provide surveillance of the Boraflex used in the SFP. The coupon program is a passive surveillance and has no impact on the 30

consequences of any accident. Therefore, the removal of the coupon program will not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR'P The blackness test includes a movable stainless steel container which contains a neutron source and four BF~

detectors. During the test operation, the container will axially traverse from the storage cell top to bottom or vice versa. Since the size of the container is smaller than the storage cell opening and the traversing speed of the container is slow in the SFP water, the container traversing speed is well controlled and will scrape neither the steel cover plate (for Region 2) nor the thick cell wall in region 1. The Boraflex panel is protected by the cover plate and cell wall and thus will not be damaged by the container during the testing. Therefore, this proposed test will not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

The coupon program which will be eliminated is used to provide surveillance of the Boraflex used in the SFP. The surveillance program does not have a functional relationship with any systems, structures, and components. Thus, the elimination of the surveillance program will not create any new failure modes. Therefore, the probability of occurrence of any equipment malfunction important to safety previously evaluated in the UFSAR will not increase.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The blackness test neither alters the dimensions of the storage cell nor changes the chemical composition of the Boraflex. The criticality control parameters are not affected. Therefore, the proposed test activity will not increase the consequence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

The coupon program which will be eliminated is used to provide surveillance of the Boraflex used in the SFP. Removal of the coupon program will not impact any equipment important to safety since the coupon program has no interaction with any structure, system or component important to safety. None of the results "of any transient analysis reported in the UFSAR are changed because of the coupon surveillance removal.

31

Therefore, the removal of the coupon surveillance program will not increase the consequences of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR'?

Based on experience gained from the blackness testing conducted at Turkey Point Unit 3 SFP in December 1993, the test requires a minimum three (3) emptied cells distance between neutron source and the'F, detectors in all direction.

During the test, the ~'Cf neutron source, considered as a point source, is positioned at the center cell and is surrounded by the Boraflex panels in racks and a thick layer of borated water. The distance between the source and the nearest fuel is at least 27 inches. Such physical configuration with a low intensity of neutron source can not sustain the K-eff value greater than 0.95 (SFP criticality already considered in UFSAR). As such, the proposed test will not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

The removal of the coupon surveillance program does not change the operation, function or design bases of any structures, systems or components important to safety as described in the UFSAR. No new failure modes or limiting single failures have been created as a result of the coupon surveillance removal.

Therefore, there is no possibility of an accident of a different type than any previously evaluated in the UFSAR.

Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR7 The only physical function of the Boraflex panels installed between two consecutive storage cells is to control and to maintain a K-eff equal to or less than 0.95. Performance of the blackness test does not change the physical function of Boraflex panels in racks. Therefore, the proposed test will not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR.

The removal of the coupon surveillance program produces no new hazard that can be postulated to cause a malfunction of equipment important to safety different from those previously analyzed in the UFSAR. Thus, there is no possibility of a 32

malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR.

7. Does the proposed activity reduce the margin of safety as defined in the basis of any Technical Specification?

The removal of the coupon surveillance program will not impact the margin of safety as defined in the basis of the Technical Specifications since the blackness test will be used as the mechanism for monitoring Boraflex degradation. The coupon program has been used to determine that the Boraflex in the SFP has not degraded below the analysis assumptions used in the criticality evaluation. The initiation of the blackness test, as a replacement for the coupon surveillance will improve our ability to track Boraflex degradation and therefore, better ensure that the SFP design basis will be maintained.

The requirement for maintaining a SFP K-eff below 0.95 is not modified by this change. Therefore, the removal of the coupon surveillance program will not impact the margin of safety as defined in the bases for the Technical Specifications.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that, the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

,33

Unit: ~ 1 JPN-PSL-SEFJ-93-024

Title:

~

~ Loss of Feedwater Transient with Corrected Steam Generator Inventory Error ABSTRACT During a review of the current licensing basis analyses performed to support the on-going steam generator replacement project, a discrepancy between the results obtained by Babcock and Wilcox International Limited (BWIL) and those given in the UFSAR for the Loss of Feedwater event (LOFW) was identified. Upon review, it was determined that the correct initial steam generator (SG) mass value was approximately 18,000 ibm lower than what was used in the St.

Lucie Unit 1 safety analysis (non-conservative). This mass error impacts the UFSAR Chapter 15 LOF event and the auxiliary feedwater section of UFSAR Chapter 10.

Seimens Power Corporation and FPL performed a review to determine if a significant safety defect existed in the SPC performed and reported in the Loss-of-Feedwater Transient With analyses Reduced Steam Generator Low Level Trip Setpoint, ANF-89-113.

was concluded that substantial safety defect did not exist and It that the reporting requirements of 10CFR21 were not applicable. The errors were corrected and a re-analyses was performed using more realistic input assumptions.

~ ~

~ The corrected results meet all acceptance criteria and are supported by the Technical

~ ~

Specification limits.~

~

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and .does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The re-analyses of the Chapter ..10 and Chapter 15 LOFW Transient with the corrected SG inventory do not change the overall configuration of the plant. The mode of operation of the plan remains unchanged. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR report is not increased.

34

~ ~

Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The re-analyses of the Chapter 10 and Chapter 15 LOFW transient with the corrected SG inventory do not affect any structures systems or components that function to mitigate the consequences of an accident by containing or detecting the release of radioactivity.

Based on the review of the identified SG inventory errors for all Chapter 15 events, the Chapter 10 LOFW Transient With Off-site Power Available and the Chapter 15 LOFW transient without initiation of AFW events were determined to require reanalysis. The results of these analyses demonstrate that the conclusions reached to support the reduce low SG level setpoint remain valid, and that no safety limits are violated.

Therefor'e, the consequences of an accident previously evaluate in the UFSAR are not increased.

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

This activity does not create any spatial or functional interaction with any structure, system or component important to safety. Specifically, the re-analysis of the LOFW Transient with the corrected SG inventory error does not have a functional relationship with any systems, structures and components nor does it create any new failure modes.

Therefore, the probability of occurrence of any equipment malfunction important to safety previously evaluated in the UFSAR will not increase.

4 ~ Does the proposed activity increase the consequences of malfunction of equipment important to safety previously evaluated in the UFSAR?

The re-analyses of the LOFW Transients with the corrected SG inventory error do not create any spatial or functional interaction with any structure, system or component important to safety. Chapter 15 of the UFSAR describes postulated plan transients which could occur as a result of equipment failures. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR are not increased.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

35

The re-analyses of the LOFW Transients with the corrected SG inventory error do not change the operation, function or design bases of any structures, systems or component important to safety as described in the UFSAR. No new failure modes or limiting single failures have been created a result of this re-analysis. No new hazards are created as a result of the LOFW Transient re-analysis that can be postulated to cause an accident different from those previously analyzed in the UFSAR.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The re-analyses of the LOFW Transients with the corrected SG inventory error do not create any spatial or functional interaction with any structure, system or component important to safety. No new hazards are created that can be postulated to cause a malfunction of equipment important to safety different from those previously analyzed in the UFSAR. Thus, the possibility of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased.

7~ Does the proposed activity reduce the margin of safety as defined in the. basis for any Technical Specification?

The re-analyses of the LOFW Transients with the corrected SG inventory error do not change the design bases, functions, or operations of any safety-related equipment and do not adversely affect any other safety-related structures, systems or components. The Technical Specification requirements and bases applicable to the LOFW event with the corrected mass error and more realistic assumptions are not affected.

Therefore, this mass error does not reduce the margin of safety as defined in the basis for the Technical Specifications.

The re-analysis of the LOFW Transient with the corrected SG inventory error does not impact safe operation of the plant, and does not constitute an unreviewed safety question.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely .affect plant operation or safety.

36

Unit: JPN-PSL-SENP-93-035

Title:

Evaluation of Inventory Loss for the Refueling Water Tank PURPOSE The purpose of Revisions 0 and 1 of thi.s evaluation is to:

1. Determine the functionality of the Unit 1 Refueling Water Tank (RWT) given the presence of a leak through the bottom of the tank.

2. Determine a maximum allowable leakage rate.

3. Evaluate functionality of the RWT during visual inspections and repairs requiring tank entry.

4. Provide recommendations for temporary non-code repair.

The purpose of Revision 2 of this evaluation is to:

1. Provide direction for a permanent ASME Code acceptable repair to the bottom of the Refueling Water Tank.

2. Provide direction for activities required to support a root cause evaluation.

BACKGROUND The Refueling Water Tank provides a source of borated primary grade water for refueling, reactor coolant makeup, and accident conditions. The RWT is an ASME Class 2 structure, erected in accordance with ANSI B96.1-1967. Following construction of the RWT in 1974, Fort Pierce city water was added to the tank; this city water remained in the tank for approximately four months. This water aggressively attacked the aluminum surface and caused pits on the bottom and side plates. A cleaning process was performed, followed by inspections and evaluations of the pitting; recommended at that time that no further corrective actions be it was performed, and that the tank should be filled with borated primary water.

A steady loss of RWT inventory was observed for several weeks in June and July of 1993. The rate of the loss measured at that time was approximately 2 gpm. The piping systems connected to the RWT and the above ground exterior surfaces of the tank were inspected and were eliminated as a source of inventory loss. Individual isolation of all lines penetrating the tank was performed, with no resultant decrease in tank leakage rates. Samples from underground test wells located near the RWT indicated both tritium and boron.

37

\

Based on this evidence, it was concluded that the inventory loss was through the. bottom of the tank. Although the exact nature of the leak was unknown at the time, it was considered to be likely due to a single small hole (resulting from pitting or similar mechanism), a series of small holes, or a small separation in a weld joint.

In. July, 1993, an acoustic emissions (AE) analysis was performed using externally mounted equipment and a transducer mounted on a mini-rover submarine. From this analysis, a single leak approximately 3/16 inch in diameter was located in an area on the east side of the tank. Approximately 1000 ultrasonic thickness readings were taken in areas identified through AE analysis as probable leak locations, as well as at random locations over the entire tank bottom. No general wall thinning of the tank bottom was identified.

A review of Generic Letter No 90-05 was performed to determine the basis for allowing non-code repairs to code class components. It was determined that any non-code repair to Class 1 or 2 piping must be performed in accordance with an engineered specification which would provide a boundary that would be of equal structural strength as the original design basis of the component.

The repair recommendation was to perform an immediate non-code repair by using an epoxy coating to adhere an aluminum plate to the tank bottom. A code relief request was prepared requesting NRC approval for the non-code repair. This relief request was submitted to the NRC, which also included Revision 1 of this evaluation as a basis for the temporary repair; in the relief request, FPL committed to providing a code acceptable repair during the Fall 1994 refueling outage. The NRC in accordance with 10 CFR 50.55a(g)(6)(i) granted the requested relief and accepted the temporary repair until the Fall 1994 refueling outage, at which time the RWT bottom plate is required to be repaired or replaced in accordance with the provisions of the ASME Code. The Code also requires a root cause evaluation to be performed to ensure that the repair process addresses the cause of the failure.

In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine if the subject activity constitutes an unreviewed safety question:

Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR will not increase because the recommended repair process does not initiate an accident or affect any accident scenario. The erection of scaffolding 38

will be performed in accordance with Administrative Procedure No. 0010724. All scaffolding erected in accordance with the guidelines of the procedure is capable of withstanding seismic loads. For this reason, the erection of the scaffolding will have no adverse effect on any safety related equipment. The code repair will be performed in accordance with the applicable requirements of the ASME Code (Section XI) and ASME B96.'1 (in accordance with the original design specification).

Therefore, for the reasons discussed above, this repair will not increase the probability of occurrence of any accident previously evaluated in the UFSAR.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

All scaffolding will be erected in accordance with the provisions of Administrative Procedure No. 0010724, and will therefore be capable of withstanding seismic loads, thus ensuring that there will be no adverse interactions with safety related items. Additionally, precautions have been specified to be taken during decontamination operations to ensure that the divers or equipment introduced into the RWT do not adversely affect the operation or safety related functions of the tank. Therefore, the scaffolding and decontamination operations to be performed while Unit 1 is on line will not affect the ability of the RWT to supply borated water in the event of a SIAS. The permanent code repair will be performed when the RWT is out of service and not being relied upon to provide borated water for any safety-related function as described in the UFSAR. The code repair will be performed in accordance with the applicable requirements of the ASME Code (Section XI) and ASME B96.1. For the reasons discussed above, this repair will not increase the consequences of any accident previously evaluated in the UFSAR.

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

All scaffolding will be erected in accordance with the provisions of Administrative Procedure No. 0010724, and will therefore be capable of withstanding seismic loads, thus ensuring that there will be no adverse interactions with safety related items. Precautions have been specified to be taken during decontamination operations to ensure that the divers or equipment introduced into the RWT do not adversely affect the operation or safety related functions of the tank.

Therefore, the scaffolding and decontamination operations to be performed while Unit 1 is on line will not affect the ability of the RWT to supply borated water in the event of a 39

4 SIAS. The permanent code repair will be performed when the RWT is out of service; the repair will be performed in accordance with the applicable requirements of the ASME Code (Section XI) and ASME B96.1, and will therefore create no adverse interaction with any structure, system, or component important to safety. The inspections and examinations to support the root cause analysis will be performed on the section of the bottom plate removed from the tank bottom, and will therefore have no effect on any safety related items.

Therefore, the probability of occurrence of any equipment malfunction important to safety previously evaluated in the UFSAR will not increase.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The permanent code repair (and associated operations related to scaffolding and decontamination) will not prevent the RWT or any other safety related equipment from performing its safety related functions. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The permanent code repair involves only a repair to the RWT bottom to prevent further leakage of borated water. This repair does not change the operation, function or design bases of any structure, system or component important to safety as described in the UFSAR. Precautions have been specified for diving operations related to decontamination.

will be erected in accordance with procedure, Scaffolding and will therefore be capable of withstanding seismic loads, thus ensuring that there will be no adverse interactions with safety related items. Therefore, no new hazards are created that can be postulated to cause an accident different from those previously analyzed in the UFSAR, and there is no possibility that an accident may be created that is different from any already evaluated in the UFSAR.,

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The permanent code repair does not change the operation, function, or design basis of any structure, system or component important to safety as described in the UFSAR. The 40

code repair will be performed in accordance with the applicable requirements of the ASME Code (Section XI) and ASME B96.1, and will therefore create no adverse interaction with any structure, system, or component important to safety.

Precautions have been specified for diving operations related to decontamination. Scaffolding will be erected in accordance with procedure, and will therefore be capable of withstanding seismic loads, thus ensuring that there will be no adverse interactions with safety related items. This repair, as evaluated above and in Section VI, does not create any new malfunction of equipment important to safety. Therefore, the possibility of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

This repair does not change the design bases, functions or operations of any safety related equipment and does not adversely affect any other safety related structures, systems or components. The code repair will be performed in accordance with the applicable requirements of the ASME Code (Section XI) and ASME B96.1, and will therefore create no adverse interaction with any safety related items.

filler plate, along with additional bearing materialA new (as necessary) will be added to provide uniform support for the new plate which will be added for the code repair. The Technical Specification requirements and bases for RWT inventory and boron concentration are not affected by this repair. Therefore, this activity does not reduce the margin of safety as defined in the bases for the Technical Specifications.

The permanent code repair to the RWT does not impact safe operation of the plant, does not constitute an unreviewed safety question and does not require a change to the Technical Specifications.

41

Unit: 2 JPN-PSL-SEMP-94-001

Title:

Temporary Installation of Strain Measurement Devices on the Pressurizer Relief Valve Discharge Piping ABSTRACT k

This safety evaluation will allow the temporary installation of strain measuring devices (i.e., strain gages) on the three pressurizer relief valve discharge pipe lines. Strain measurements will permit the determination of moments imposed by the discharge piping on the pressurizer safety relief valves. The installation of strain gages is classified as Non-Nuclear Safety Related because: 1) The strain measurements will not be used for safety related functions; 2) The pressurizer relief valve discharge Lines are Non-Nuclear Safety (i.e., not Safety Class 1, 2 or 3),

designed in accordance with the ANSI B31.1 Code; and 3) The strain measuring devices are non-intrusive and will not impact the operation or function of any Nuclear Safety Related Systems, Structures or Components.

The safety evaluation addresses the potential impact of the installation of strain gages has on plant operation and responds to the guidelines set forth in 10 CFR 50.59. As long as the requirements of this evaluation are followed, there are no adverse effects on plant operation or safety. The results of the safety evaluation show that no unreviewed safety question exists and that no Technical Specification changes are required. Therefore, prior NRC approval is not required pursuant to 10 CFR 50.59.

The strain gages and thermocouples, excluding the data acquisition equipment ("Strain Indicator VISHAY P-3500" and "OMEGA Engineering Multipoint Digital Thermometer Model 2166A"), will remain in place Cycle 8 and shall be removed prior to Mode 2 for Cycle 9. This 'or evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an

1. Does the proposed change increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed change does not affect any equipment whose malfunction is postulated in the UFSAR to initiate an accident 42

or prevent an accident from occurring. No physical modifications have been performed to the RCS or connected systems except to attach the strain gage shims and Omega clips to the pressurizer relief valve discharge piping using micro-spot welding generating energy output less than 50 watt-seconds. The piping is not Safety Class 1, 2 or 3.

Additionally, no failure modes have been identified that could initiate an accident previously evaluated in the UFSAR. As such, the probability of occurrence of an accident previously evaluated in the UFSAR has not increased.

2 ~ Does the proposed change increase the consequences of an accident previously evaluated in the UFSAR?

The proposed change does not diminish in any way the ability of the pressurizer relief valve discharge piping or any other safety system to perform its intended function. There is no interaction with any safety related equipment. As such, the proposed change does not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed change increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed change is to install strain measuring devices e which do not interfere with the operation of any system.

There is no interaction with any safety related equipment.

The strain gage an/or thermocouples (including temporary field routed cables) 'will not pass through the sump water screens if the assembly becomes loose and carried to the containment sump. As such, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased by the proposed change.

4 ~ Does the proposed change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The installation, operation or failure of the strain measuring devices will neither impact operation of system nor cause any'dverse affect to any safety any related equipment.

Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR are not increased by the proposed change.

5. Does the proposed change create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

43

As discussed in the failure modes and effects analysis section of this evaluation, the proposed change does not introduce any new failure modes to or impact safety related equipment in any way. The proposed change will not change or impact the requirements of the design bases of the safety related systems as described in the UFSAR. There are no postulated failure modes which could be considered accident initiating.

Therefore, the possibility of an accident of a different type than any previously evaluated in the UFSAR is not created by this modification.

6. Does the proposed change create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed change does not interact spatially or functionally with any structure, system or component important to safety other than the attachment of strain gages to the pressurizer safety relief valve discharge piping. No new failure modes are created that can be postulated to interact with any equipment important to safety different than those previously analyzed in the UFSAR. Therefore, the possibility of a malfunction of equipment important to safety which is of a different type than any previously evaluated in the UFSAR is not created by the proposed change.

7. Does the proposed change reduce the margin of safety as defined in the basis for any Technical Specification?

The Technical Specification requirements and Technical Specification Bases are not affected by the proposed change.

The proposed change does not affect any plant Technical Specification requirement. The proposed change maintains the level of protection previously evaluated in the UFSAR.

Therefore, the margin of safety as defined in the bases for any Technical Specification is not reduced by this modification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

44

Unit: JPN-PSL-SEFJ-94-002

Title:

Evaluation of Reduced Shutdown Cooling Flow Rate For St.

Lucie Unit 2 Cycle 8 Refueling Outage ABSTRACT This evaluation addresses the single point injection operation of the shutdown cooling (SDC) system in Mode 5, for St. Lucie Unit 2, during the 1994 outage. For the maintenance activity on the low pressure safety injection (LPSI) header lines, one of the four safety injection lines will be isolated. If a failure of one SDC train is assumed, this configuration will leave one SDC loop in operation'ith reduced flow from one LPSI pump injecting through one safety injection line. The St. Lucie Unit 2 Technical Specifications (T.S.) 3.4.1.4.1 and 3.4.1.4.2 require that two SDC loops be operable in Mode 5 (at least one loop be in operation) when the reactor coolant loops are not filled or when the water level in steam generators is less than 104 indicated narrow range level. The evaluation performed here verifies the adequacy of one train with single point injection to satisfy the Technical Specifications bases requirements associated with decay heat removal, boron dilution and boron stratification, under the initial conditions and assumptions specified in referenced letters, except that the operational mode is for Mode 5.

The T.S. 3/4.4.1 bases requirements 'ensure that i) sufficient cooling is available to remove decay heat, and ii) sufficient coolant mixing and circulation is maintained to minimize the effects of boron concentration reduction and prevent boron stratification.

The evaluation shows that in the proposed configuration the maximum reactor coolant system (RCS) temperature is 134.3 F if the maintenance is done 7 days after the reactor shutdown, and is close to 140 F for the 5 days after shutdown case. These temperatures are well below the maximum Mode 5 temperature of 200 F. Additionally, boron stratification is shown not to 'be a concern and sufficient time is available for the operator to terminate any boron dilution event. It is shown that there is no unreviewed safety question associated with the proposed maintenance activity.

This safety evaluation is revised to delete the assumption that nozzle dams would 'e installed. This assumption is not a requirement to ensure the validity of'he safety analyses. A review was performed and concluded that there is no effect on the conclusions of this evaluation. This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to 45

plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed activity removes part, of the shutdown cooling (SDC) system for maintenance on low pressure safety injection (LPSI) header lines under specified conditions. The only UFSAR event related to the proposed activity is the Boron Dilution event. However, the proposed activity does not lead to an increase in the frequency of such a transient.

Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR is not increased.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The proposed activity will reduce the SDC flow during the maintenance on the LPSI header lines. This flow reduction has been shown to be acceptable for decay heat removal, boron mixing and prevention of boron stratification, and meet T.S.

bases requirements. Therefore, the consequences of an accident previously evaluated in the UFSAR are not increased.-

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The only occurrences of malfunction of equipment important to safety previously evaluated in the UFSAR, related to the proposed activity, are the failure of one train of SDC system and the inadvertent injection via charging pumps causing boron dilution. The probability of malfunction of these equipment is not increased by the activity proposed. Additionally, the proposed activity has no effect on any other equipment or system configuration, nor does modes.

it create any new failure Therefore, the probability of occurrence of any equipment malfunction important to safety previously evaluated in the UFSAR will not increase.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

46

The malfunction of equipment important to safety previously evaluated in the UFSAR is not changed by the proposed activity. The consequences of reduced flow on the decay heat removal requirements and the boron dilution event have been shown in this evaluation not to increase. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR are not increased.

Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSARP The only change due to the proposed activity is the operation of SDC with single point injection. This configuration has been evaluated and shown to meet all the T.S. bases requirements. There are no new systems or system interactions involved important to safety as described in the UFSAR.

Therefore, the proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR'?

The proposed activity is related to the SDC system configuration. The malfunction of the SDC system has been evaluated and shown to meet the Technical Specifications and UFSAR bases requirements. There are no changes to any other systems or equipment important to safety. Thus, the possibility of a malfunction of equipment important to safety of a different type than any previously. evaluated in the UFSAR is not increased.

Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The bases for Technical Specification 3.4.1.4.1 and 3.4.1.4.2, decay heat removal, and boron concentration reduction and boron stratification. The decay heat removal requirements have been shown to require that sufficient decay heat removal capacity is available, and sufficient coolant circulation is maintained to minimize the effects of a boron dilution incident and prevent boron stratification. These requirements have been shown to be satisfied under the conditions specified in this evaluation. Therefore, the proposed activity does not reduce the margin of safety as defined in the bases for the Technical Specifications.

47

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

48

Unit: 2 JPN-PSL-SEIS-94-005

Title:

Operation of the Wide Range Containment Level Monitoring Channels L-07-13A 6 L-07-13B With Inoperable Sensors ABSTRACT This Safety Evaluation allows the continued use of the containment level monitoring instruments channels L-07-13A and L-07-13B by providing a means of compensating for a failed sensor in each channel. This evaluation is necessary as sensor 9 of LE-07-13B is no longer operational. Additionally, a non-conformance report identified that sensor 11 of LE-07-13A had failed in November of 1990. An engineering evaluation evaluated sensor 11 and determined that operation with one failed sensor is acceptable. Sensor 11 (LE-07-13A) has since been replaced. Therefore, this evaluation will allow the proper operation of channels L-07-13A and L-07-13B with one failed sensor in one or both channels. This evaluation will also allow continued operation in the future by considering the containment level monitoring instruments channels L-07-13A and L-07-13B operational with one failed sensor in one or both channels while waiting for replacement parts.

This evaluation documents the acceptability of the level transmitter's circuitry alteration. The modification will not adversely affect the operation or the existing qualification of the containment level system.

This temporary alteration will have no impact on plant safety or operation. This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased. The failure of this system is not considered an initiating event in any accident scenario. The wide range containment water level monitoring loops are utilized solely for post accident monitoring purposes.

49

Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased by this reconfiguration. The wide range containment level loops will continue to monitor the water level in containment during an analyzed accident.

Each channel will continue to provide post accident monitoring capabilities with the exception of decreased resolution between a failed sensor and the next highest sensor. The lack of indication at the failed sensor will not adversely impact any operator actions associated with accident mitigation as no actions or decision points are anticipated to occur based solely on containment water level.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased with this circuitry reconfiguration. The channel still provides monitoring of containment water level during an analyzed accident. The design/procurement requirements imposed on the components to be installed are equivalent to that of the level monitors. No new system interactions are being introduced, only a resistor is being added in the level transmitter. This condition does not result in an increase in probability of a malfunction.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important .to safety previously evaluated in the UFSAR?

The consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased. Because the channel will continue to monitor post accident containment water level as described in 2. above, the consequences of a malfunction have not been changed.

5~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The possibility of an accident of.a different type than any evaluated previously in the UFSAR has not been created. The modified instrument loop provides only monitoring capability of wide range containment level during an analyzed accident and will operate as described in item 2. above.

50

I e

~

6. Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any evaluated previously in the UFSAR?

The possibility of a malfunction of a different type than any evaluated previously in the UFSAR has not been created as this condition does not introduce any new failure modes to the post accident containment level monitoring system.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any technical specifications?

The proposed activity does not reduce the margin of safety as defined in the basis for any technical specifications as the reconfigured channel will continue to provide the necessary monitoring function of post accident containment water level as required by the Plant Technical Specifications.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

51

Unit: 1 JPN-PSL-SENP-94-005

Title:

Shutdown Cooling Suction Valve Interlock Design ABSTRACT This safety evaluation provides justification for use of an alternative design in the shutdown cooling system (SDCS) of St Lucie Unit 1. The current licensed design for shutdown cooling per the UFSAR includes diverse, redundant and independent pressure sensors feeding the suction side isolation valve permissive interlocks. The proposed design provides for redundant and independent pressure sensors, however, the requirement for diversity is eliminated since safety.

it does not improve reliability or This safety evaluation also demonstrates that safe plant operation was not affected by the elimination of diversity in the operation of pressure sensors since their replacement in 1984. This conclusion is substantiated by both evaluation and analysis which shows that .neither safety nor reliability of the system is adversely affected. Note that PCM 001-182 upgraded this instrumentation to meet the requirements of 10 CFR 50.49 (i.e.,

environmental qualification of electrical equipment).

An UFSAR change package is provided in this evaluation to affect the update to the design features for shutdown cooling. Since the design of a safety related system is being changed, this safety evaluation is classified as safety related.

This design change neither involves unreviewed safety questions nor requires changes to Technical Specifications. Therefore, this design change and revision to the UFSAR may be performed without prior NRC approval. Note that there are no configuration changes to the plant required by this safety evaluation.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the following evaluation serves to determine whether use of an alternative design in the shutdown cooling system of St Lucie Unit 1 constitutes an unreviewed safety question:

1. Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The pressure transmitters PT-1103 and PT-1104 are used for the open permissive interlock (OPI) function of the SDCS suction valves. Although not accident initiators, the pressurizer pressure instrumentation was evaluated for the effect on the 52

OPI and determined that the increase in frequency for a ISLOCA (ISLOCA frequency is governed by a catastrophic failure of both series isolation valves) due to elimination of diversity between the transmitters is negligible. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR is not increased.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR are not increased because the elimination of diversity in the pressure sensing instrumentation does not affect the ability of the SDCS to perform its design basis function. The OPI function is unaffected by the elimination of diversity in the instrumentation. The OPI function neither degrades nor prevents actions used to mitigate UFSAR accidents. Therefore, the removal of diversity in the pressure sensing instrumentation does not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed activity increase the probability of 'an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

Since reliability remains essentially unchanged as a result of elimination of diversity in the pressure -

sensing instrumentation, there is no increase in the probability of an occurrence of a malfunction of equipment. Therefore, the probability of occurrence of equipment malfunction important to safety previously evaluated in the UFSAR is not increased.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The elimination of diversity in the pressure sensing instrumentation does not affect the ability of the SDCS to perform its design basis function. The OPI function is unaffected by the elimination of diversity in the instrumentation. The OPI function neither degrades nor prevents action used to mitigate consequences of UFSAR accidents. Therefore, the removal of diversity in the pressure sensing instrumentation does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR.

5~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

53

The ISLOCA as the postulated accident involved with the elimination of diversity in the pressure sensing instrumentation for the SDCS suction valves is an analyzed event in the UFSAR. Therefore, new types of accidents are not created that are different from any already evaluated in the UFSAR.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

Reliability remains essentially unchanged as a result of elimination of diversity in the pressure sensing instrumentation. Therefore, the elimination of diversity of the pressure sensing instrumentation does not increase the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The setpoints for the SDCS OPI, SITS, and PORVs - selected to the low temperature mode of operation are unaffected by the removal of diversity in the pressurizer pressure transmitters, and therefore, the proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

54

Unit: JPN-PSL-SEHS-94-008

Title:

Gasket Leak Repair for Shutdown Cooling Return Isolation Valve V3480 ABSTRACT This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

This evaluation demonstrates the acceptability of the repair of a body-to-bonnet gasket leak on safety related valve V3480 through the use of a seal clamping device and sealant. The leaking valve gasket shall be replaced at the next forced outage of sufficient duration or the next refueling outage, therefore, this evaluation is applicable to operating cycle 12 only.

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased. This repair method restores the integrity of the valve body to bonnet flange gasket, and will not hinder the operation of any equipment capable of initiating an accident. The use of this repair method will not overstress or adversely affect valve V3480 or the 1A shutdown cooling return piping. The functionality of any equipment important to safety is not affected by this repair method since the seal clamp performs an identical function as the leaking gasket; the bolt loadings and pressure boundary materials are not adversely affected by the injection of sealant. Also, contaminants from the sealant will not leach into the reactor coolant system medium in 55

sufficient quantities to impact any other wetted materials as described in this document.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased 'y the application of this repair activity. This repair will not adversely affect any equipment which is required for accident mitigation or safe shutdown. This repair activity creates no new paths for the uncontrolled release of radioactivity in the event of a postulated accident and does not adversely affect any radiation monitoring equipment.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased. The application of this seal clamp and sealant will restore the integrity of the valve body to bonnet gasket. The functionality of any equipment important to safety is not affected by this repair activity.

Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased. The consequences of the failure of the injection seal is the same as the failure of the gasket, which would result in a loss of system fluid. No new unevaluated system leakage paths or possible paths for an uncontrolled radioactive release are created by the implementation of'these repair methods. These repairs will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function, as detailed in this document.

5~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The possibility of an accident of a different type than any evaluated previously in the safety analysis report has not been created. The proposed repair activity does not provide a new method of normal or emergency component or system 56

operation and does not provide any new component failure modes. In addition, no new plant hardware other than the seal clamp and fittings, as previously described, is added by this repair activity. Thus, no new accident initiators are introduced through these repairs.

6. Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the UFSAR?

The possibility of a malfunction of a different type than any evaluated previously in the safety analysis report has not been created. The failure of valve V3480 to performed has been previously evaluated in the UFSAR. Additionally, leakage of sealant into the RCS is prevented by procedural complianceg the use of thermosetting sealant compounds, and the method and location of injection. Also, contaminants from the sealant will not leach into the affected system medium in quantities sufficient to impact any other wetted materials and the sealant chemistry is compatible with the affected system chemistry requirements.

7 ~ Does the proposed activity reduce the margin of safety as defined in the basis for any technical specification?

This repair activity does not reduce the margin of safety as defined in the basis for any technical specifications.

Chemistry limits are not altered and no other change is proposed to the plant design, modes of operation or assumptions in the bases for the Technical Specifications or Safety Analysis.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

57

Unit: 2 JPN.-PSL-SENS-94-010

Title:

Evaluation for Alternate ECCS Valve Alignment to Repair Line 3/4-SI-121 ABSTRACT St Lucie Unit 2 was in Mode 3 during a plant return to power startup sequence following the cycle 8 refueling outage of an approximately two month duration. A non-conformance report was written to document leakage on an instrument line off of the 2B1 Emergency Core Cooling System (ECCS) injection line. Repair of this line will require the draining of adjacent piping, rendering the 2B1 injection line inoperable. The drained portion of piping is on the upstream side of the 2B1 injection line check valve (isolation to the RCS) back to the ECCS pumps'A & B HPSI and B LPSI) 2B1 loop isolation valves and the 2B1 Safety Injection Tank (SIT) isolation valve. The loop 2B1 ECCS injection isolation valves will be closed with control power removed (valves V3634, HCV-3635, HCV-3636 & HCV-3637) to prevent flow through this line.

This evaluation restricts plant operation to Mode 3 with pressurizer'pressure below 1750 psia or Mode 4. Operation in Mode 5 with the 2B1 injection loop isolated is addressed in a safety evaluation.

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed activity does not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

This activity provides an alternate valve alignment for repair of a line in the 2Bl ECCS injection line. The valve alignment is acceptable for the plant mode of operation identified in this safety evaluation. There are no components being installed by this evaluation and no interactions with any components that would be considered as accident initiators.

58

The repair does not introduce any condition considered to as effect accident initiators. This alignment is required repair and return the ECCS loop to its original design. As such, the probability of occurrence of an accident previously evaluated in the UFSAR is not increased.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased by this alternate valve alignment or repair. Valve V3237 provides an RCS pressure boundary function. The ECCS injection function is maintained for the mode of operation in which the line repair is effected. Since the ECCS function mitigates the consequences of accidents, the proposed activity does not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased. The alternate valve alignment maintains pressure boundary to allow repair of the line. There are no components being installed by this evaluation and no interactions with any components important to safety. The ECCS function is maintained acceptable by the proposed valve alignment. Only the ECCS injection line affected by the repair is inoperable. The remaining ECCS injection lines are unaffected and meet the technical specification requirements for ECCS operability. Therefore, the probability of occurrence of a malfunction of equipment important to safety is not increased.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased. Since the ECCS function mitigates the consequences and remains unaffected by the alternate valve alignment to effect repair of the line, the consequences of a malfunction of equipment important to safety has not increased.

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

59

~ P The possibility of an accident of a different type than any evaluated previously in the safety analysis report has not been created. There are no components being installed by this evaluation and no interactions with any components important to safety. Thus, no new accident initiators are introduced through this repair.

6. Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the UFSAR?

The possibility of a malfunction of a different type than any evaluated previously in the safety analysis report has not been created. There are no components being installed by this evaluation and no adverse interactions with any components important, to safety. Thus, no new accident initiators are introduced through the alternate valve alignment and line repair.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any technical specification?

The inoperable ECCS injection line as a result of the alternate valve alignment is acceptable for the plant mode of operation in which the line repair is performed. There are no changes proposed to the plant design, modes of operation or assumptions in the basis for the Technical Specifications or Safety Analysis. Therefore, the alternate valve alignment to effect the line repair does not reduce the margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not. adversely affect plant operation or safety.

60

Unit: 2 JPN-PSL-SEMS-94-011

Title:

Pressurizer Spray Bypass Valve V1454 Needle Tip Failure This 10CFR50.59 evaluation addresses the acceptability of V1454 for USE AS IS with a captured broken disc. V1454 is a 3/4" manual flow control valve (needle valve) which maintains minimal bypass flow around pressurizer spray valve PCV-1100F. NCR 178-293-3037M identified the tip of the needle disc to be broken off and wedged into the valve body seat. Multiple attempts to remove the broken tip through the valve bonnet have failed due to the tip being wedged tightly into the seat. V1454 is a Quality Group A valve, designed in accordance with ASME B&PV Code Section III, Class 1 requirements, and is located within the Reactor Coolant System pressure boundary. Therefore, this evaluation is classified as Safety Related.

V1454 has been reassembled with the original disc/stem assembly and the broken tip in the valve body seat. V1454 will be left in the closed position, under a plant operational clearance, to assure the tip is captured between the valve seat and the upper disc/stem assembly. Maintaining the closed provides assurance the tip will not become free to movevalveabout the Reactor Coolant System as a loose part.

There are two pressurizer spray valves, PCV-1100E & F, and each spray valve has an associated spray bypass valve; V1453 and V1454, respectively. V1454 and V1453 were replaced during the current refueling outage with Velan valves to replace the obsolete AiResearch valves.

The spray valves and spray bypass valves are designed to allow a small continuous flow, diverted from the RCS cold leg loop Bl and B2, to bypass the normally shut spray valves. The continuous flow maintains the spray lines and pressurizer surge line warm relative to the pressurizer in order to prevent thermal shock during plant transients requiring pressurizer spray.

V1454 shall be replaced or repaired during the next forced outage of sufficient duration or the next refueling therefore, this evaluation is applicable to operating cycleoutage;8 only.

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an 1 ~ Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed activity does not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

This evaluation provides for plant operation with valve V1454, a normally throttled open valve, to operate in a closed position. The function of V1454 is to provide for a small amount of RCS flow through the spray lines (bypassed around spray valve PCV-1100F) to protect the spray and surge lines from the effects of a thermal shock during plant transients.

Sufficient continuous flow exist to minimize the differential temperature between the spray line and the pressurizer.

Operation of PCV-1100F is not impacted by the position of V1454, thus the probability of occurrence of a pressure transient requiring mitigation by the code, safety valves is not increased. A failure modes and effects analysis has concluded that there is no credible failure mode of this configuration that could lead to the initiation of an analyzed accident.

Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased by this valve alignment. Valve V1454 is part of the pressurizer spray system which is relied upon for control of pressurizer pressure during normal plant transients. The pressurizer spray system is not relied upon for accident mitigation. Overpressure protection is provided by the pressurizer code safety valves. There is no impact to the operation and performance of any accident mitigating equipment.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a, malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased. Valve V1454 will be maintained in the closed position. There are no components being installed by this evaluation and no component interactions being introduced. A failure modes and effects 62

analysis has concluded that there is no credible failure which could result in the failed disc of V1454 traveling downstream to the spray line or any other downstream components.

Sufficient continuous flow exists to minimize the differential temperature between the spray line and the pressurizer.

Operation of spray valve PCV-1100F is not impacted.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously, evaluated in the UFSAR?

The consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased. Valve V1454 is part of the pressurizer spray system which is relied upon for control of pressurizer pressure during normal plant transients. The pressurizer spray system is not relied upon for accident mitigation.

There is no impact to the operation or performance of any equipment relied upon for accident mitigation.

5~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The possibility of an accident of a different type than any evaluated previously in the safety analysis report has not been created. There are no components being installed by this evaluation and no new interactions being introduced.

Pressurizer pressure control via PCV-1100E & F will not be affected. A failure modes and effects analysis has concluded that there is no credible failure which could create an accident of a different type than any previously evaluated in the UFSAR.

6. Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the UFSAR?

The possibility of a malfunction of a different type than any evaluated previously in the safety analysis report has not been created. There are no components being installed by this evaluation and no adverse interactions with any components important to safety. A failure modes and effects analysis has concluded that there is no credible failure which could result in the failed disc of V1454 traveling downstream to the spray line or any other downstream. component. Sufficient continuous flow exists to minimize the differential temperature between the spray line and the pressurizer.

63

7. Does the proposed activity reduce the margin of safety as defined in the basis for any technical specification?

The margin of safety as defined in the basis for any Technical Specification has not been reduced. There are no Technical Specification Limiting Conditions of Operation or Surveillance Requirements related to the pressurizer spray system.

Design Features specification, Component Cyclic or Transient Limits, identifies cyclic or transient limits for the pressurizer spray system and provides a method for calculating the pressurizer spray nozzle cumulative usage factor. This specification requires a lifetime accounting of the number of spray cycles where the temperature difference between the pressurizer water and spray water is greater than 200'F.

Operation with V1454 in the closed position does not impact the use of this table. The cumulative, usage factor calculation and the inherent safety margins of the cyclic and transient limits remain unchanged.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

64

l Unit: ZPN-PSL-SEMS-94-013

Title:

Freeze Seal Application for V3480 on the 1A Shutdown Cooling Return Line ABSTRACT The purpose of this Safety Evaluation is to evaluate the safety significance of applying a freeze seal to the reactor coolant system 1A hot leg shutdown cooling return line while refueling operations are in progress. This activity will render the "A" shutdown cooling loop out of service, and therefore inoperable.

Since the freeze seal is classified as not nuclear safety (NNS),

the two concerns to be evaluated are 1) the loss of the remaining operable shutdown cooling loop due to either an active or passive failure, and 2) the loss of freeze seal integrity.

V3480 is the primary motor operated valve for the loop 1A shutdown cooling return line. The valve and piping are {}uality Group A, Seismic Class I, and designed to the Class 1 requirements of USAS B31.7. The valve is not isolatable from the 1A hot leg and is the first of two isolation valves in series that isolates the reactor coolant system (RCS) from the low-pressure shutdown cooling system.

V3480 is also relied upon to maintain the pressure boundary when the RCS is open to the refueling pool during refueling operations.

The purpose of the freeze seal is to maintain this pressure boundary function during refueling operations so that V3480 may be repaired or replaced. This activity is consistent with the use of other non-permanent equipment used during St. Lucie outages in similar operating conditions (i.e., the reactor cavity seal ring, the steam generator nozzle dams, and the RCS hot and cold leg plugs).

This evaluation establishes that the application of a freeze seal on the 1A shutdown cooling return line does not present a nuclear safety concern since its failure would not cause a loss of the inservice shutdown cooling loop or adversely impact refueling operations. Although the freeze seal is an NNS device, the St.

Lucie freeze seal application procedure (GMP-10) provides adequate measures to ensure a low probability for a freeze seal failure.

This evaluation also states additional requirements to provide added assurance that the likelihood of uncontrolled refueling pool leakage remains improbable.

Revision 1 is issued to specify the requirements for maintaining the RCS pressure boundary intact and to specify the lowest elevation at which the pipe may be cut. The conclusions of the safety evaluation remain unchanged. Revision 2 is issued to clarify requirements for the upper guide structure (UGS) concurrent with the freeze seal application, clarify ECCS lift 65

parameters, and revise the evaluation to reference the new plant procedure on freeze seal application (GMP-10). The conclusions of the safety evaluation remain unchanged.

Revision 3 is issued to address contingency plans for the radiographic access port in the pup piece upstream of V3480, which was added per PCM 082-194M, Rev. 1 ~ The conclusions of the safety evaluation remain unchanged.

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and .does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an

1. Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has .not been increased since the proposed activity does not adversely affect any accident-initiating components. This evaluation establishes that a loss of reactor coolant system integrity due to a catastrophic freeze seal failure is not considered a credible event. The evaluation establishes that the freeze seal is a reliable pressure boundary for the given plant conditions, and the actions required section of this evaluation provides required actions to minimize the likelihood of uncontrolled leakage.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased since the proposed activity does not adversely affect any equipment which is required for accident mitigation. Since the reactor head will be removed, a postulated freeze, seal failure would not create a new path for uncontrolled radioactive releases and would not adversely affect any radiation monitoring equipment. Also, a freeze seal failure would not affect the capability of V3481 to function as a containment isolation valve while V3480 is either open or removed from the system, nor would it affect the operation of the inservice shutdown cooling loop because 66

a freeze seal failure would not result in the draining of the reactor coolant system hot legs.

Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased. Although the freeze seal is not considered a safety related device, the freeze seal is a reliable pressure boundary for the plant conditions. Thus, no new unmitigated fa'ilure modes for any equipment important to safety are introduced by the proposed activity and no new components or equipment are introduced that could adversely interact with any equipment important to safety.

Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased since the proposed activity does not create a new path for uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function.

Furthermore, the freeze seal will not adversely impact any equipment which is required to perform an active safety related function or to initiate actuation of any safety systems. Even in the unlikely event of a catastrophic freeze seal failure, the operability of the operating shutdown cooling train would not be impacted.

Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The possibility of an accident of a different type than any evaluated previously in the UFSAR has not been created since the proposed activity does not add or adversely affect any equipment that could reasonably be capable of initiating an accident. The installation or postulated failure of the freeze seal would not present any credible. new paths for the loss of refueling pool and reactor, coolant system inventory while irradiated fuel will be transported above the reactor vessel. This conclusion is based on an evaluation of the times required for fuel pool drainage and the fuel transit time above the reactor vessel, as well as the improbability of a catastrophic freeze seal failure. A freeze seal is a 67

j reliable pressure boundary for the plant conditions and this evaluation provides required actions to minimize the likelihood of uncontrolled leakage. Furthermore, Off-Normal Operating Procedure 1-0120031 (Excessive Reactor Coolant System Leakage) provides the required actions for the scenario in which a loss of inventory occurs while refueling operations are in progress.

6. Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the UFSAR7 The possibility of a malfunction of a different type than any evaluated previously in the UFSAR has not been created since the proposed activity does,,not add or adversely affect any equipment that has a reasonable possibility of installation or postulated failure of the freeze seal does malfunctioning.'he not present any credible new paths for the loss of refueling pool and reactor coolant system inventory while refueling operations are in progress. Also, the freeze seal will not inhibit or otherwise adversely 'affect the operation of any equipment important to safety.

7~ Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed activity does not reduce the margin of safety as defined in the basis for any Technical Specification since the proposed activity is intended to maintain the RCS pressure boundary to ensure'hat the refueling cavity water level is maintained in accordance with the Technical Specifications.

Technical Specifications 3.9.8.1 and 3.9.10 both require a minimum of 23 feet of water above the irradiated fuel assemblies when only one shutdown, cooling loop is operable or when core alterations are in progress. This evaluation establishes that the likelihood of uncontrolled leakage as a result of a freeze seal'failure is improbable; therefore, the proposed activity does not reduce any Technical Specification margins of safety.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

68

Unit: 1 & 2 JPN-PSL-SENS-94-015

Title:

Safety Evaluation for Service Water System Modifications ABSTRACT The Service Water (SW) system for St. Lucie Units 1 & 2 (sometimes referred to as the potable and sanitary water system) has been identified as a system requiring increased maintenance activities.

A review of system functions and interactions has concluded that SW system piping and components located outside of the reactor auxiliary buildings, fuel handling buildings, diesel generator buildings and component cooling water areas are classified as Not Nuclear Safety, have no interactions with equipment important to safety and are not required to be considered within the scope of the FPL Quality Assurance Program. As such, modifications to those portions of the SW system are planned to be performed outside of the Plant Change/Modification (PC/M) process. Marked-up drawings reflecting the configuration of the system will be provided by the plant to engineering for review and as-building after modifications are complete.

This evaluation does not apply to the entire SW system. An attachment to the evaluation provides a detailed description of those portions of the SW system to which this evaluation applies.

The SW system is a part of System Number 15, which consists of the Fire Protection, Demineralized Water, Service Water and Primary Makeup Water systems. This evaluation does not include any portions of the Fire Protection, Demineralized Water and Primary Makeup Water systems (typically indicated on plant piping and instrumentation drawings as FP, DW & PMW).

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an 1 ~ Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has 'not been increased. The SW system is not assumed to cause any analyzed accident in the UFSAR.

69

0 The equipment within the SW system boundaries does not initiate design basis accidents. Furthermore, any this modifications made to the SW system as a result of evaluation could not initiate a design basis accident. The boundaries of the SW system within the scope of this evaluation have been selected to ensure that there is no possible interaction with equipment important to safety.

Piping located within the Reactor Auxiliary Buildings, Diesel Generator Buildings and Fuel Handling Buildings has been excluded from this evaluation.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased. The SW system provides water for, various non-safety uses throughout the site such as washdown stations, emergency eyewash stations, human consumption, sinks, toilets and numerous hose connections.

The SW system is not relied upon in any way to provide for accident mitigation. Modifications to the SW system that are in accordance with the guidance in this safety evaluation will not affect any systems relied upon for accident mitigation.

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased. The SW system does not contain nuclear safety related equipment nor does it interact with any equipment important to safety. Modifications to the SW system that are in accordance with the guidance contained in this safety evaluation will not increase the challenges to or the likelihood of failure of equipment important to safety.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR. The SW system, classified as Not Nuclear Safety, does'ot interface, with any system required for accident mitigation. This evaluation allows modifications to those portions of the SW system that are not in the vicinity of any accident mitigating equipment (i.e., potential system interactions with equipment important to safety are 70

0

~ ~

avoided). Thus, the consequences of any failure or malfunction of equipment important to safety are not changed.

5 ~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR. The SW system is a non-safety system that does not interface with safety related equipment. This evaluation is limited to those portions of the SW system that are not in the vicinity of safety related equipment; therefore, potential interactions with safety related equipment are not possible.

There are no credible failure modes associated with potential modifications to the SW system (including flooding and physical interactions) that could create the possibility of a nuclear accident different than any previously evaluated.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than any. previously evaluated in the UFSAR. The SW system is a non-safety system that does not interface with safety related equipment. As noted above, there are no credible failure modes associated with potential modifications to the SW system (including flooding and physical interactions) that could impact safety related equipment.

Does the, proposed activity reduce the margin of safety as 7 ~

defined in the basis for any Technical Specification?

The SW system is not discussed in any Technical Specification or Technical Specification basis. Modifications to the SW system that are in accordance with the guidance provided in this safety evaluation can not adversely affect the margin of safety to any Technical Specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

71

Unit: 2 JPN-PSL-SENP-94-017

Title:

Disabling the Steam Dump and Bypass Control System Quick Open Feature for Load Reduction ABSTRACT The St Lucie Unit 2 Steam Dump and Bypass Control System (SBCS) provides a means of manually controlling reactor coolant temperature during plant startup and for removing NSSS stored energy during periods when a turbine load rejection occurs.

Additionally, the system can automatically accommodate a load rejection of up to 45 percent reactor power without opening the pressurizer safety valves, main steam safety valves or causing a reactor trip when the condenser is available. This safety evaluation provides an assessment of disabling the quick open feature of the SBCS for load reduction while retaining the other control functions. The modification is necessary due to spurious actuation of quick opening signals to the SBCS valves. The SBCS is classified as not nuclear safety and performs no safety function and is not required for plant shutdown following an accident.

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an 1 ~ Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed modification does not affect any safety related components or interact with any equipment that would be considered as accident initiators. Therefore, the proposed activity does not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The SBCS is not required or essential for the safety of the plant. In addition, the system does not perform any function to mitigate the consequences of an accident previously 72

evaluated. The proposed modification is bounded by the UFSAR.

Therefore, the proposed activity does not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed modification is previously evaluated in the UFSAR. Failure of the SBCS does not result in the malfunction of any safety related equipment. Therefore, the proposed activity does not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The SBCS does not affect any equipment required to mitigate the consequences of an accident, nor does it affect any other equipment important to safety. The proposed modification is bounded by the UFSAR. Therefore, the proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR.

Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed modification does not adversely interact with any components important to safety. No new accident initiators are introduced through the proposed modification.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed modification does not adversely interact with any components important to safety. The proposed modification will disable the automatic quick open feature of the SBCS for load reduction. Loss of this automatic feature is evaluated in the UFSAR. Therefore, the proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR.

7~ Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

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There are no limiting conditions of operation associated with the SBCS. The proposed activity does not change assumptions in the basis for the Technical Specifications. Acceptance in the SER is based on an evaluation of the failure modes of this system. Therefore, the proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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Unit: 1 & 2 JPN-PSL-SENS-94-018

Title:

Safety Evaluation for Hypochlorite System Modifications The Hypochlorite (CL) system for St. Lucie Units 1 & 2 has been identified as a system requiring increased maintenance activity.

A review of system functions and interactions has concluded the CL system (as bounded in an attachment provided in the evaluation) is not required to be considered within the scope of the FPL Quality Assurance Program since it is classified as Not Nuclear Safety and has no interactions with equipment important to safety. As such, modifications to the CL system are planned to be performed outside of the Plant Change/Modification (PC/M) process. Marked-up drawings reflecting the configuration of the system will be provided by the plant to engineering for as-building after modifications are complete. Because of interaction concerns, this evaluation does not include the individual injection lines located within the Unit 1 & 2 intake cooling water (ICW) pump bays. The boundaries for this evaluation are detailed in an attachment.

The CL system is a non-safety system common to St. Lucie Units 1 &

2 that produces a sodium hypochlorite solution via electrolytic decomposition of filtered seawater. The hypochlorite solution is periodically injected into the suction side of the intake cooling water (ICW) and circulating water (CW) pumps for the control of biological fouling.

This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change. to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased. The CL system is not assumed to cause any analyzed accident in the UFSAR.

Furthermore, any modifications made to the CL system as a result of this evaluation could not initiate a design basis 75

accident. The boundaries of the CL system within the scope of this evaluation have been selected to ensure that there is no possible interaction with equipment important to safety.

Piping located within the Unit 1 & 2 ICW pump intake bays has been excluded from this evalua'tion.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased. The CL system injects a hypochlorite solution for the control of biofouling in the ICW and CW systems of both units. The CL system is not relied upon in any way to provide for accident mitigation. The ability of the Unit 1 & 2 ICW systems to perform their safety function is not affected by the CL system. Although the CL system helps to maintain ICW system heat exchanger surfaces clean, the capability of the heat exchangers to perform their design function is not dependant on the CL system.

Modifications to the CL system, in accordance with the guidance and scope of this safety evaluation, will not affect any systems relied upon for accident mitigation.

~ 3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important 'to safety previously evaluated in the UFSAR7 The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased. The CL system does not contain nuclear safety related equipment nor does it functionally interact with any equipment important to safety. Potential failures of the CL system would have no effect on the operation or reliability of equipment important to safety. Modifications to the CL system that are in accordance with the guidance contained in this safety evaluation will not increase. the challenges to or the likelihood of failure of equipment important to safety.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR. The CL system is classified as Not Nuclear Safety and does not interface with any system required for accident mitigation; This evaluation allows modifications to those portions of the CL system that are not in the vicinity of any accident mitigating equipment (i.e., potential system interactions with equipment important to safety are 76

avoided). Thus, the consequences of any failure or malfunction of'equipment important to safety are not changed.

5 ~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR. The CL system is a non-safety system that does not interface with safety related equipment. This evaluation is limited to those portions of the CL system that are not in the vicinity of safety related equipment; therefore, potential interactions with safety related equipment are not possible.

There are no credible failure modes associated with potential modifications to the CL system (including flooding and physical interactions) that could create the possibility of a nuclear accident different than any previously evaluated.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR. The CL system is a non-safety system that does not interface with safety related equipment. As noted above, there are no credible failure modes associated with potential modifications to the CL system (including flooding and physical interactions) that could impact safety related equipment.

7 ~ Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The CL system is not discussed in any Technical Specification or Technical Specification basis. Modifications to the CL system that are in accordance with the guidance provided in this safety evaluation can not adversely affect the margin of safety to any Technical Specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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Unit: 2 JPN-PSL-SENP-94-019

Title:

Alternative Valve Position for Spray Header Isolation Valve I-FCV-07-1B ABSTRACT The purpose of this safety evaluation is to demonstrate that the containment spray header isolation valve (I-FCV-07-1B) is capable of performing its design functions while aligned in the open position. I-FCV-07-1B is normally closed during power operation.

This change is expected to remain in effect until the next refueling outage.

Valve I-FCV-07-1B is required to open on a containment spray

~

actuation signal for accident mitigation. The valve is air operated, normally closed. Its fail safe position is open either on loss of air or power. The valve also performs a containment isolation function for the B train located outside containment.

However, this valve is not required by technical specifications to close for containment isolation nor does it receive a containment isolation actuation signal.

This safety evaluation involves an assessment of the effects on a safety related system, and therefore, is classified as safety related.

This evaluation concludes that operation of the plant with I-FCV-07-1B maintained in the open position during power operation does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a 'change to plant Technical Specifications and does not adversely affect plant operation or safety.

Revision 1 extends the period that I-FCV-07-1B may be maintained in the open position until the next refueling outage. This revision also establishes valve lineup requirements for surveillance testing associated with the containment spray line while in this configuration. This evaluation concludes that the activity as described above does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an 78

Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

UFSAR includes analysis for the effects related to an inadvertent actuation of containment spray during normal plant operation leading to containment differential pressure. The assumptions and conclusions for this analysis remain unchanged

'by the proposed valve alignment. Although the valve remains open, the probability of occurrence for this event is not considered to have increased since compensatory measures are being implemented to reduce the 'probability of inadvertent spray actuation. Therefore, the proposed activity does not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

2 ~ Does the proposed activity increase, the consequences of an accident previously evaluated in the UFSAR?

The proposed valve alignment does not increase the consequences of an accident previously evaluated in the UFSAR since the valve is in its fail safe position and the system is capable of performing its intended functions. Therefore, the proposed activity does not increase the consequences of an accident previously evaluated in the UFSAR.

Does the proposed activity increase the probability of an 0 occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed valve alignment does not adversely affect the function of equipment important to safety and the containment.

spray system .is capable of performing its intended design functions. The probability of occurrence of a malfunction of equipment important to safety has been reduced by eliminating one failure mode (failure of valve I-FCV-07-1B to open).

Therefore, the proposed activity does not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed valve alignment does not adversely affect the containment spray system and its required function to mitigate the consequences of an accident. Equipment important to safety is not adversely affected by the valve alignment.

Therefore, the proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR.

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P 5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed valve alignment to the containment spray system was determined not to adversely interact with any other components important to safety. This alignment does not prevent containment spray system from performing its intended functions. New accident initiators are not introduced through this alignment. Therefore, the proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

New accident initiators are not introduced through the proposed valve alignment. The containment spray system design, function, and method of performing the function has neither changed nor created a new failure mode. 'his alignment is bounded by the UFSAR. Therefore, this activity does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed activity does not change assumptions in the basis for any of the Technical Specifications. Therefore, the proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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0 1 ~

Unit: 1 JPN-PSL-SEFJ-94-021

Title:

RTD Response Time Limit Increase From 8.0 Seconds to 14.0 Seconds ABSTRACT The evaluation performed here supports an increase in the maximum allowable Resistance Temperature Detector (RTD) response time for St. Lucie Unit 1 from 8 seconds to 14 seconds. St. Lucie Unit 1 uses RTDs which are Weed and Rosemount types. During the Loop Current Step Response (LCSR) testing of the RTDs in the past few years, the response times of these RTDs have been in the range of 3 to 8 seconds. The Weed RTDs have. averaged response times less than 5 seconds, whereas the Rosemount RTDs have typically shown response times between 6 and 8 seconds.

Due to difficulties associated with the removal and installation of RTDs, the RTD response time has been of interest to Florida Power

& Light (FPL) Company and to the nuclear industry in general.

Relaxation of the RTD response time will allow greater operational flexibility, that would prevent a sound RTD from being replaced when it is otherwise acceptable for use.

An evaluation is conducted to assess the impact of increased RTD response time limit on the safety analysis. The Basis to Technical Specification (TS) for the Thermal Margin/Low Pressure (TM/LP) trip specifies an allowance of 30 psia to compensate for the associated time delays. The pressure bias factor of 30 psia bounds the present RTD delay time of 8 seconds. This bias term has been re-evaluated for the RTD response time of 14 seconds. The new bias term, calculated to be 42 psia, has been accounted for in the current safety analyses which, therefore, remain unaffected.

It should be noted that a similar change to increase the RTD response time to 14 seconds was approved by the NRC in 1991 for St.

Lucie Unit 2 . Relocation of tables of instrument response time limits to the Final Safety Analysis Report (UFSAR) has recently been approved by the NRC for the St. Lucie Units 1 and 2, in response to FPL's request for a license amendment. The proposed change to the RTD response time limit is in accordance with 10 CFR 50.59 and Enclosure 1 to Generic Letter (GL) 93-08 ).

It has also been shown by this safety evaluation. that this activity neither constitutes an unreviewed safety question nor requires changes to the Technical Specifications. Therefore, prior NRC approval for implementation of these changes is not required.

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SAFETY EVALUATION In accordance with 10 CFR Part 50.59, the following evaluation serves to determine whether this modification constitutes an unreviewed safety question:

Does the proposed change increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed change affects only the slow CEA Withdrawal events analyzed in the UFSAR. The initiation of these events is independent of the RTD response time. The probability of occurrence of such events is, thus, unaffected by the proposed activity. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR is not increased.

2 ~ Does the proposed change increase the consequences of an accident previously evaluated in the UFSAR?

The proposed change will allow longer RTD response time, which will affect the pressure bias term used in determining the TM/LP margin. The new higher pressure bias term is used in the current slow CEA Withdrawal events to compensate for the increased RTD delay time. Thus the consequences of these events remain unchanged. Therefore, the consequences of an accident previously evaluated in the UFSAR are not increased.

Does the proposed change increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed change allows an increase in the response time limit (from 8 seconds to 14 seconds) to be met during surveillance testing of the RTDs. This change has no effect on the malfunction of any equipment or system important to safety as evaluated in the UFSAR, nor does it create any new failure modes.'herefore, the probability of occurrence of any equipment malfunction important to safety previously evaluated in the UFSAR will not increase.

4 ~ Does the proposed change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed change does not adversely affect the performance of any safety related equipment. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR are not increased.

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s. Does the proposed change create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed change is an increase in the RTD response time limit and a compensating change in the pressure bias term.

There are no changes to the plant configuration. No new systems or system interactions are involved that adversely affect equipment or systems important to safety. Therefore, the proposed change does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

6 ~ Does the proposed change create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The change proposed allows longer RTD response times and does not adversely affect any safety related equipment.

Additionally, there are no changes to any system configuration or equipment important to safety. Thus, the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR is not increased.

Does the proposed change reduce the margin of safety as defined in the basis for any Technical Specification?

The Basis to Technical Specification for TM/LP trip specifies an allowance of 30 psia to account for time delays. The value for this pressure allowance changes to 42 psia corresponding to the proposed RTD response time of 14.0 seconds. The increase in the pressure bias term from 30 psia to 42 psia, used in the current safety analysis, compensates for the effects of proposed increase in the RTD response time. The margin of safety is, thus, not reduced. The NRC has approved the relocation of the response time tables from the TS to the UFSAR. Therefore, the proposed change does not reduce the margin of safety as defined in the bases for the Technical Specifications.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely .affect plant operation or safety.

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Unit: 2 JPN-PSL-SENP-94-021

Title:

Removing the Automatic Control Function for I-TCV-14-4B During the performance of In-Service Testing stroke time testing, I-TCV-14-4B, a temperature control valve in the St. Lucie Unit 2 Intake Cooling Water (ICW) system failed to demonstrate repeatable stroke time results. Nonconformance Report (NCR) g2-612 was generated as a result of this event.

The purpose of this safety evaluation is to demonstrate that the 2B ICW system is capable of performing its intended safety function while I-TCV-14-4B is clamped in a predetermined condition. Valve I-TCV-14-4B is automatically controlled to maintain component cooling water (CCW) temperatures during power operation. Valve I-TCV-14-4B will be clamped such that it is in a locked position to provide sufficient flow through the 2B CCW heat exchanger during accident conditions.

Valve I-TCV-14-4B is required to be open during all operational modes. During Design Basis Accident conditions, I-TCV-14-4B opens further to permit increased ICW flow through the 2B CCW heat exchanger. I-TCV-14-4B performs the safety function of re-positioning from the Normal operating throttle position (Design Flow = 8,250 gpm) to Emergency throttle'osition (Design Flow =

14,500 gpm).

This safety evaluation provides an assessment of the effects on a safety related system, and therefore, is classified as safety related. This evaluation concludes that operation of the plant with I-TCV-14-4B maintained in the clamped open throttle position/

during power operation, does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

ev's'o A clarification is provided to address changes in ICW system performance (i.e. fouling, pump performance changes, etc) before the next scheduled outage. This is addressed by addition of a curve of ICW accident flowrate requirements through a CCW heat exchanger vs ICW inlet temperatures,. which can be used to demonstrate operability of the ICW system at accident flowrates below 14,500 gpm. This revision has no effect on the conclusions of the safety evaluation.

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SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine if the subject activity constitutes an unreviewed safety question:

Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of an accident occurring which will challenge the valve's safety function has not changed with the new configuration of the valve. With respect to the description and FMEA in the UFSAR for valve I-TCV-14-4B, the valve in the clamped position will increase the availability of flow to the CCW train. The clamp itself is designed to withstand a design basis seismic event. Therefore, the proposed activity does not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The proposed configuration does not adversely affect the safety function of the valve. One failure mode (ie, valve failure in the closed position) has been removed. With the valve in its clamped position, the consequences of providing one less train of cooling water is decreased. The clamp itself is designed to withstand a design basis seismic event.

Therefore, the proposed activity does not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed activity increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed valve alignment does not change the valve's designed safety function. The valve is required to be open to admit flow through the CCW heat exchanger. The clamped throttle position ensures the valve would perform this function. The clamp itself is designed to Seismic Category I requirements and therefore maintains the position of the valve. The proposed activity does not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to, safety previously evaluated in the UFSAR?

85

The proposed valve alignment does not affect the CCW system operation, performance or safety function, nor does the new condition affect the operation, performance or safety function of the intake cooling water system. Equipment and systems important to safety will function in the same manner in the new condition. The consequences of a malfunction in the new condition are equal to those of the previous condition.

Therefore, the proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR.

Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR7 New accident initiators are not introduced through this configuration. Therefore, the proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

New accident initiators are not introduced through the proposed valve alignment. Failure of the valve is currently considered in the UFSAR, the clamp design reduces the probability of fail'ure as described in the UFSAR. This alignment is bounded by the UFSAR. Therefore, this activity does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR.

Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed activity does not change assumptions used as the basis for any of the Technical Specifications. Therefore, the proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.

The proposed configuration does not affect safe operation of the Intake Cooling Water System and the Component Cooling Water system. In addition, the proposed configuration does not constitute an unreviewed safety question and does not require a change to the Technical Specifications. Therefore, implementation of the proposed configuration does not require prior NRC approval.

86

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

87

e t Unit:

Title:

ZPN-PSL-SENS-94-025 Safety Evaluation Discrepancies for Fuel Handling Equipment UFSAR FPL {}ualityAssurance Department Audit No. {}SL-OPS-94-24 identified several minor discrepancies between existing plant procedures and various parts of the St. Lucie Unit 1 UFSAR. The discrepancies noted in the {}A audit all pertain to fuel handling equipment. This evaluation provides a review and resolution to several of the discrepancies and provides an UFSAR Change Package.

This safety evaluation demonstrates that the UFSAR changes provided in an UFSAR Change Package do not adversely affect plant safety, security or operation. It has also been shown by this safety evaluation that this activity neither constitutes an unreviewed safety question nor requires changes to the Technical Specifications. Therefore, prior NRC approval for implementation of these changes is not required.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an 1 ~ Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased. The Fuel Handling Accident, UFSAR section 15.4.3, is the only relevant analyzed accident. Neither the parking location of the spent fuel handling machine nor the use of the CEA handling tool can be postulated to result in an increase in the probability of occurrence of a fuel handling accident. The UFSAR changes pertaining to the use of a dummy fuel assembly and the PT of fuel handling grapples are considered editorial clarifications.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased. The consequences of the UFSAR fuel handling accident are not affected by this evaluation 88

since the design and operation of relevant accident mitigation systems are not impacted in any way.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased. This evaluation analyzes the potential failure modes and effects of the proposed changes and concludes that no new failure modes or system interactions are introduced. The design of the spent fuel handling machine, including its ability to be seismically stable, has not been changed. The CEA handling tool has been engineered by the NSSS vendor for its intended application. The UFSAR changes pertaining to the use of a dummy fuel assembly and the PT of fuel handling grapples are considered editorial clarifications.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR. The changes provided via this evaluation are all associated with fuel handling equipment.

There is no impact to any UFSAR accident analysis assumptions or to the operation of any system required for accident mitigation.

5. Does the proposed .activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR. This evaluation analyzes the potential failure modes and effects of the proposed changes and concludes that no new failure modes or system interactions are introduced.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated, in the UFSAR?

The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR. As stated 89

above, there are no new failure modes or system interactions as a result of the changes provided via this evaluation.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

There is no impact'o any Technical Specification Limiting Condition for Operation, Surveillance or Bases as a result of this evaluation.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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Unit: JPN-PSL-SEMS-94-028

Title:

Installation of a Blind Flange on the Inlet of Containment Purge Valve,FCV-25-1 ABSTRACT Containment isolation for penetration P-11 is normally accomplished by closing FCV-25-2 and FCV-25-3. FCV-25-3 has exhibited leakage during the LLRT. This temporary change will allow the installation of a blind flange on the inlet of containment purge supply valve FCV-25-1. This blind flange will act as a containment isolation device replacing FCV-25-3.

The containment isolation system provides the means to isolate fluid systems that pass through containment penetrations such that any radioactivity that may be released to the containment atmosphere following a postulated Design Basis Accident (DBA) is confined. As such this temporary alteration performs a safety related function and this evaluation and its associated modifications are considered to be safety related.

This change does not affect the function of the 48" containment purge supply system during plant-power operations-as the system is not used in modes 1, 2, 3 and 4,. It has also been shown by this safety evaluation that this activity neither constitutes an unreviewed safety question nor requires changes to the Technical Specifications. Therefore, prior NRC approval for implementation of these changes is not required.

SAFETX EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine unreviewed safety question:

if the subject activity constitutes an Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR7 The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased since the proposed activity does not adversely affect any accident-initiating components. This evaluation establishes that the installation of the blind flange replaces the function of a closed FCV-25-3. Additionally, this valve is not required to open for any safety related requirements. The analysis of effects on safety section of this evaluation establishes that the blind flange is a reliable pressure boundary for the given plant conditions.

91

'oes 0 the proposed activity increase accident previously evaluated in the the consequences UFSAR?

of The consequences of an accident previously evaluated in the an UFSAR have not been increased since the proposed activity does not adversely affect any equipment which is required for accident mitigation. Since the blind flange is one of two isolation's for P-11, its failure would not create a new path for uncontrolled radioactive releases and would not adversely affect any radiation monitoring equipment. Also, the blind flange failure would not affect the capability of FCV-25-2 to function as a containment isolation valve.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased. The gasket and valve packing seal of FCV-25-1 may become a new leak path. An LLRT will be performed on this new boundary. The packing of FCV-25-1 or FCV-25-2 may have a sealant injected to prevent leakage. Thus, no new unmitigated failure modes for any equipment important to safety are introduced by the proposed activity and no new components or

, equipment are introduced that could adversely interact with any equipment important to safety.

Does the proposed activity increase the consequences malfunction of equipment important to safety previously evaluated in the UFSAR?

of a The consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR have not been increased since the proposed activity does not create a new path for uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function.

Furthermore, the blind flange will not adversely impact any equipment which is required to perform an active safety related function or to initiate actuation of any safety systems.. Potential gasket and packing leaks from FCV-25-1 or FCV-25-2 will be tested and prevented, as required, with the use of the sealant PRI-201N.

5~ Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

92

0

~ +

The possibility of an accident of a different type than any evaluated previously in the UFSAR has not been created since the proposed activity does not add or adversely affect any equipment that could reasonably be capable of initiating an accident. The installation or postulated failure of the blind flange would not present any credible new paths for the loss of containment atmosphere following a DBA. The blind flange is a reliable pressure boundary for the plant conditions.

6 ~ Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the UFSAR?

The possibility of a malfunction of a different type than any evaluated previously in the UFSAR has not been created since the proposed activity does not add or adversely affect any equipment that has a reasonable possibility of malfunctioning.

The installation or postulated failure of the blind flange does not present any credible new paths for the containment atmosphere following a DBA. Also, the blind flange will not inhibit or otherwise adversely affect the operation of any equipment important to safety.

7~ Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specificationi The proposed activity does not reduce the margin of safety as defined in the basis for any Technical Specification since the proposed activity is intended to maintain the containment boundary in accordance with the Technical Specifications.

Technical Specification 3/4.6.1.1 provide the requirements to ensure containment integrity is maintained. This evaluation establishes that the likelihood of uncontrolled leakage as a result of a blind flange failure is improbable; therefore, the proposed activity does not reduce any Technical Specification margins of safety.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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11 Unit: 1 & 2 JPN-PSL-SEMS-94-029

Title:

Shutdown Operations Criteria for Reduced Inventory and Draining the Reactor Coolant System The purpose of this evaluation is to demonstrate the acceptability of shutdown operations given the following proposed changes:

A) The criteria for reduced inventory will now be defined as 3 feet below the reactor vessel flange.

B) The criteria for draining the RCS after shutdown will now be limited by the time to incover the core.

These proposed changes will bring St. Lucie Plant more in-line will NRC and industry guidelines on shutdown operations and will provide more flexibility for refueling outages without compromising plant safety. Implementation of these changes effectively amends previous submittals to the NRC on shutdown operations, however, such changes are allowed under 10 CFR 50.59 as outlined in NRC correspondence on the same sub)ect.

This safety evaluation involves an assessment of changes to shutdown operations, and therefore, is classified as safety related.

This evaluation concludes that the proposed changes to operation of the plant during shutdown neither involve an unreviewed safety question nor require a change to plant Technical Specifications, as defined in 10CFR50.59, and do not adversely affect plant operation or safety. Therefore, prior NRC approval is not required for implementation.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine if the proposed changes to the criteria (A & B) for shutdown operations constitutes an unreviewed safety question:

Do the proposed changes increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased. The proposed changes do not affect any accidents discussed in the UFSAR.

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Do the proposed changes increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR have not been increased. This evaluation does not affect any of the accidents discussed in the UFSAR and therefore does not increase any of the consequences of the accidents discussed in the UFSAR.

3 ~ Do the proposed changes increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in,the UFSAR has not been increased by reducing the RCS water inventory to 3 feet below the reactor vessel flange because this water level is above the mid-nozzle of the hot leg. The probability of occurrence would not be affected until the RCS water level reached the mid-nozzle elevation. The change in the criteria for containment closure does not change any probability of occurrence of a malfunction of equipment previously evaluated in the UFSAR.

4 ~ Do the proposed changes increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

Adopting the Reduced Inventory definition established by the NRC in Generic Letter 88-17 does not increase the consequences of a malfunction of equipment important to safety since the loss of shutdown cooling event was previously evaluated and the results of that evaluation remain valid. The proposed change in the containment closure criteria provides for containment closure prior to uncovering the core following a loss of shutdown cooling event. Therefore, the consequences from the previously evaluated loss of shutdown cooling event remain valid.

5. Do the proposed changes create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed change to the criteria for Reduced Inventory will not create a different type of accident than those found in the UFSAR. The proposed change is above mid-nozzle. The plant can operate safely in Mode 5 at mid-nozzle. The proposed change to the containment closure criteria provides for containment closure prior to uncovering the core.

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Therefore, no different type of accident is created and the accident assumptions found in the UFSAR remain valid.

6. Do the proposed changes create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed change to the criteria for Reduced Inventory provides adequate margin to prevent the loss of shutdown cooling due to draining the RCS below mid-nozzle. The loss of shutdown cooling is evaluated in the UFSAR and this change does not alter the assumptions for that evaluation. The proposed change to the containment closure criteria does not affect equipment important to safety as analyzed in the UFSAR.

This criteria is based on the ability to provide containment closure prior to uncovering the core and fission product release. Therefore a different type of accident is not created by this change.

7. Do the proposed changes reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed changes do not affect any Technical Specification nor do they reduce any margins of safety defined by the Technical Specifications. The proposed changes incorporate the NRC accepted criteria and definitions.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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Unit: 2 ZPN-PSL-SENP-94-037

Title:

SIT Discharge/Loop Check Valve Stroke Test This safety evaluation demonstrates the acceptability of performing a full stroke test of the Safety Injection test will Tank (SIT) be used to discharge/loop check valves. The proposed address NRC requirements for SIT check valves testing delineated in Generic Letter 89-04.

The test is to be performed during refueling with the reactor vessel head and upper internals removed, and the refueling cavity flooded. The test consists of establishing a reduced nitrogen pressure in the SIT to be used as a motive force to discharge a portion of the SIT inventory through the subject check valves. The test is initiated by opening the SIT discharge line motor operated valve and observing via acoustic and magnaflux monitoring that the check valves fully open.

Calculations have been performed to demonstrate that sufficient velocities can be achieved and maintained to fully open the subject check valves. This evaluation addresses the various potential adverse effects on the plant including those on the reactor internals and steam generator nozzle dams resulting from the proposed flow test.

Since this safety evaluation assesses the effects of the proposed test on safety related components, it is classified as safety related.

This evaluation demonstrates that, the proposed test neither involves an unreviewed safety question nor requires a change to plant Technical Specifications, and does not adversely affect plant operation or safety. Therefore, it is concluded that the proposed test may be performed without prior NRC approval.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the following questions serve to determine whether the proposed test constitutes an unreviewed safety question:

Does the proposed test increase the. probability of occurrence of an accident previously evaluated in the UFSAR?

Performance of this test will not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

Specifically, the test will not increase the probability of a 97

loss of shutdown cooling and will not cause an upset of fuel assemblies. The maximum fluid velocity that will occur in the vessel, as a result of this test, is on the order of 1/2 ft/sec, and therefore, is not capable of adversely impacting the fuel or CEAs.

2 ~ Does the proposed test increase the consequences of an accident previously evaluated in the UFSAR?

This test will be conducted in Mode 6 with the reactor vessel head removed. This test does not impact any of the assumptions for refueling accidents evaluated in the UFSAR and will not increase the consequences of any of these analyses.

3 ~ Does the proposed test increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The shutdown cooling system will be secured in the tested loop. There are no potential failure modes resulting from this test that would adversely impact shutdown cooling equipment. The additional pressure that the nozzle dams may be subjected to as a result of this test have been analyzed and determined not to affect nuclear safety. Therefore, this test will not increase the probability of a malfunction of equipment important to safety.

4 ~ Does the proposed test increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The consequences of malfunction of equipment important to safety are unaffected by this test. The consequences of a loss of shutdown cooling or a failure of the steam generator nozzle dams are the same regardless of this testing.

5. Does the proposed test create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The possibility of introducing nitrogen into the RCS as a result of this test has been evaluated. The initial test parameters have been chosen such that after the nitrogen gas has fully expanded there will still be liquid inventory remaining in the SIT. By conducting the test within the bounds. of these initial parameters, the possibility of injecting nitrogen into the RCS has been precluded. None of these effects will result in an accident of a different type than previously evaluated in the UFSAR.

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6. Does the proposed test create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

This test does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR since the failure modes of the nozzle dams have been previously evaluated.

7. Does the proposed test reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed test does not change assumptions in the basis for any of the Technical Specifications. Therefore, the proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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Unit:

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~ 2 JPN-PSL-SENP-94-039

Title:

~ Jumper/Lifter Lead for PDIS-2216 ABSTRACT The purpose of this safety evaluation is to demonstrate the acceptability of applying a jumper to the pressure differential switch (PDIS-2216) in the chemical 6 volume control system (CVCS) letdown line. This jumper will remain in effect, until the switch is repaired or replaced.

The switch functions to sense high differential pressure across the regenerative heat exchanger, which is indicative of high flow from a downstream line break outside containment, initiating a signal to close isolation valve V-2516. A letdown line break is postulated in the UFSAR as resulting from a seismic event. The jumper will defeat closure on high differential pressure. However, letdown isolation via-V-2515 still occurs from temperature element TE-2221 located immediately downstream of the regenerative heat exchanger.

This safety evaluation involves an assessment of safety related systems, and therefore, is classified as safety related.

This eyaluation concludes that operation of the plant with the proposed jumper on PDIS-2216 does not represent an unreviewed safety question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety. Therefore, prior NRC approval is not required for implementation.

SAFETY EVALUATION In accordance with 10CFR50.59, the responses to the following questions serve to determine whether the proposed jumper constitutes an unreviewed safety question:

Does the proposed jumper increase the probability of occurrence of an accident previously evaluated in the UFSAR7 The probability of occurrence of an accident previously evaluated in the UFSAR will not increase because this jumper does not affect any equipment whose malfunction is postulated in the UFSAR to initiate an accident.

2 ~ Does the proposed jumper increase the consequences of an accident previously evaluated in the UFSAR?

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The consequences of an accident previously evaluated in the UFSAR will not increase because this jumper does not adversely affect valve closure from a SIAS and CIAS.

3 ~ Does the proposed jumper increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed jumper does not adversely affect the function of equipment important to safety. Therefore, the proposed jumper does not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

4 ~ Does the proposed jumper increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed jumper does not adversely affect equipment required to mitigate the consequences of an accident. Without automatic isolation via differential pressure, letdown isolation still occurs from temperature element (TE-2221) located immediately downstream of the regenerative heat exchanger. Therefore, the proposed jumper does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR.

Does the proposed jumper create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

New accident initiators are not introduced through the proposed jumper. Therefore, the proposed jumper does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

6. Does the proposed jumper create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed jumper was determined not to create any new failure modes. Therefore, this jumper does not create the possibility of a malfunction of equipment important to safety of a different type than any previously. evaluated in the UFSAR.

7~ Does the proposed jumper reduce. the margin of safety as defined in the basis for any Technical Specification?

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proposed jumper does not change assumptions in the basis I'he for any of the Technical Specifications. Therefore, the proposed jumper does not reduce the margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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e Unit: 1 JPN-PSL-SEFJ-94-040

Title:

Removal of TE-1122CD Input from Channel "D" of the RPS for PSL 1 ABSTRACT During power ascension following the refueling outage for St. Lucie Unit 1 Cycle 13, erratic cold leg temperature indications were obtained from one of the two temperature measurement channels which provide input to Channel "D" of the reactor protection system (RPS). This safety evaluation assessed the effect on UFSAR analyses of temporarily removing the faulty channel from operation for the remaining portion of Cycle 13.

This safety evaluation assessed the effects on safety of removing a faulty channel from input to RPS, as such this safety evaluation is classified as Safety Related.

This evaluation concludes that the proposed removal of the TE-1122CD input to Channel D of the RPS for the duration of St. Lucie Unit 1 Cycle 13 does not represent an unreviewed safety question nor does Therefore it isrequire it a change to the Technical Specifications.

concluded that the proposed temporary change may be performed without prior NRC approval.

SAFETY EVALUATION In accordance with 10 CFR Part 50.59, the following evaluation serves to determine whether this modification constitutes an unreviewed safety question:

1~ Does the proposed change increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed change affects one of two cold leg temperature indications to RPS Channel "D". These, instrument channels do not initiate any of the events analyzed in the UFSAR. The probability of occurrence of such events is, thus, unaffected by the proposed activity. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR is not increased.

2 ~ Does the proposed change increase the consequences of an accident previously evaluated in the UFSAR?

This evaluation demonstrated that the proposed change will not affect the functions which are credited in mitigating 103

consequences of accidents. Therefore, the consequences of an accident previously evaluated in the UFSAR are not increased.

3. Does the proposed change increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed change removes one of the two cold leg temperature indications to Channel "D" of the RPS. This change has no effect on the malfunction of any equipment or system important to safety as evaluated in the UFSAR, nor does it create any new failure modes. Therefore, the probability of occurrence of any equipment malfunction important to safety previously evaluated in the UFSAR will not increase.

4 ~ Does the proposed change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed change does not adversely affect the performance of any safety related equipment which functions to mitigate consequences of accidents. Therefore,. the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR are not increased.

Does the proposed change create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The proposed change involves removal of one of the two cold leg temperature measurement channels which input Channel "D" in the RPS cabinet. There are no .changes to the plant configuration and/or RPS functions. No new systems or system interactions are involved that adversely affect equipment or systems important to safety. Therefore, the proposed change does not create the possibility of an accident of a different type than any previously evaluated in the UFSAR.

6. Does the proposed change create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The change proposed allows operation with one of the two cold leg temperature measurement channels which input Channel "D" of the RPS removed. No new hazards, are created as a result of the proposed change. This change does not adversely affect any safety related equipment. Additionally, there are no changes to any system configuration or equipment important to safety. Thus, the possibility of a malfunction of equipment e

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important to safety of a different type than any previously evaluated in the UFSAR is not increased.

7. Does the proposed change reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed change does not affect the UFSAR conclusions nor does it change the margin of safety as defined in the based for the Technical Specifications. Therefore, the proposed change does not reduce the margin of safety as defined in the bases for the Technical Specifications.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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Unit: JPN-PSL-SENP-94-043

Title:

Safety Evaluation Temporary Removal of the ICW Pump Missile Shield ABSTRACT This safety evaluation demonstrates the acceptability of plant operation while a section of the intake cooling water (ICW) pump missile shield roof is removed temporarily to perform maintenance activities on an out of service pump. The function of the missile shield is to protect the ICW pumps from missiles during a hurricane/tornado.

This safety evaluation documents the design intent of the ICW pump missile shield with respect to maintenance access. It concludes that removal of missile shield roof sections for maintenance during plant operation is consistent with the original design intent of shields. Furthermore, the risk of tornado missiles is negligible for the short period of time the roof section is not in place. As an additional precaution the missile shield roof sections are re-installed in the event of a threatening hurricane when the risk of damage from tornadoes is the greatest.

Since this safety evaluation assesses the effects on missile protection of the safety related ICW pumps, this evaluation is classified as safety related.

This safety evaluation demonstrates that the temporary configuration of the ICW missile shield neither involves an unreviewed safety question nor requires a change to plant Technical Specifications, and does not adversely affect plant operation or safety. Therefore, it is concluded that the proposed temporary shield configuration may be implemented without prior NRC approval.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the following questions serve to determine whether the proposed temporary missile shield configuration constitutes an unreviewed safety question:

Does the proposed temporary change increase the probability of occurrence of an accident previously evaluated in the UFSAR7

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Tornadoes/hurricanes do not initiate design basis accidents.

Therefore, the proposed temporary change does not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

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2. Does the proposed temporary change increase the consequences of an accident previously evaluated in the UFSAR?

Tornadoes/hurricanes are not postulated to occur simultaneously with design basis accidents. The performance of the operating pumps will not be adversely impacted.

Therefore, the consequences of an accident previously evaluated in the UFSAR are not increased by the proposed temporary missile shield configuration.

3 ~ Does the proposed temporary change increase the probability of an occurrence .of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The failure modes and effects analyses of the ICW system as described in the UFSAR are not changed by the temporary ICW missile shield roof configuration. The risk from missiles is negligible for the short period of time the roof section is not in place. In addition, the missile shield will be re-installed in the event of a threatening hurricane when the risk of damage from missiles is the greatest. Therefore, the proposed temporary change does not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

Does the proposed temporary change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The risk from missiles is negligible for the short period of time the roof section is not in place. The performance of the operating pumps will not be adversely impacted. Therefore; the proposed temporary change does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR.

5. .Does the proposed temporary change create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

Tornadoes/hurricanes do not initiate design basis accidents.

The failure modes and effects analyses of ICW as described in the UFSAR are not changed by the temporary ICW missile shield roof configuration. Therefore, the proposed temporary change does not create the possibility of..an accident of a different type than any previously evaluated in the UFSAR.

6 ~ Does the proposed temporary change create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

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The failure modes and effects analyses of the ICW system as described in the UFSAR are not changed by the temporary ICW missile shield roof configuration. The risk from missiles is negligible for the short period of time the roof section is not in place. In addition, the missile shield will be re-installed in the event of a threatening hurricane when the risk of damage from missiles is the greatest. Therefore, the proposed temporary change does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR.

7. Does the proposed temporary change reduce the margin of safety as defined in the basis for any Technical Specification?

The temporary missile shield configuration does not require a change to Technical Specifications. The bases indicates that the ICW system must provide sufficient cooling water to vital components, assuming a single failure, consistent with assumptions used in the safety analysis. Two ICW loops remain operable and accident analyses single failure assumptions are not affected by the temporary missile shield configuration.

The risk from missiles is negligible for the short period of time the roof section is not in place. In addition, the missile shield will be re-installed in the event of a threatening hurricane when the risk of damage from missiles is the greatest. Therefore, the margin of safety as defined in the basis for technical specifications is not reduced by the temporary ICW missile shield roof configuration.

The foregoing constitutes the determination, per 10 CFR that the subject activity does not involve an unreviewed50.59(b),

safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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Unit: 1 JPN-PSL-SENP-94-044

Title:

Safety Evaluation for the use of Devoe Devran 140 Epoxy Compound and Kansai Biox as a Coating System for the St.

Lucie Unit 1 Intake Structure The purpose of this evaluation is to demonstrate the acceptability of applying coating systems to the St. Lucie Unit 1 intake well walls (including the safety related Intake Cooling Water (ICW) wells) consisting of Devoe Devran 140 Epoxy Compound and Kansai Biox for concrete surfaces and Amerlock 400 with a Kansai Biox overcoat for steel surfaces.

The purpose of coating the intake structure wells is to limit the growth of marine organisms within the intake structure. Reducing marine growth on the intake structure will improve heat exchanger performance by reducing fouling and blockages. Currently, a hypochlorite solution and Clamtrol is injected into the sea water to control the marine growth. The use of this coating system may reduce the amount of chemicals injected into the sea water, which will have a positive effect on the environment.

To demonstrate the acceptability of using this coating system at St. Lucie, an intake well was cleaned and prepared in accordance with the instructions provided by Specification CN-2.27, "Furnishing and Application of Service Level II and Balance-of-Plant Protective Coatings.". A Coating Data Sheet (Attachment A1 to CN-2.27) was prepared by the FPL Coating Specialist and the coating system was applied to the concrete walls of intake well 2B2 in the spring of 1994. In addition, a steel plate was coated with Amerlock 400 and Kansai Biox was applied as an overcoat.

Acceptable adhesion test results were. obtained with no cohesion failures noted. Industry data and the data collected from the tests at St. Lucie have indicated that the coating systems do not have a failure mode which results in the delamination of large sheets of epoxy which could cause ICW system blockages, that the coating systems maintain acceptable adhesion characteristics and reduce the growth of marine organisms. These results indicate that the surface preparation and the use of these coating systems at St.

Lucie is acceptable.

This safety evaluation involves an assessment of changes to safety related ICW intake wells and is therefore classified as safety related.

This evaluation concludes that the proposed coating of the intake wells neither involves an unreviewed safety question, as defined in 10 CFR 50.59(a)(2), nor requires a change to plant Technical 109

Specifications, and does not adversely affect plant operation or safety. Therefore, in accordance with 10 CFR 50.59, prior NRC approval is not required for implementation.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the below listed questions serve to determine if the proposed coating of the intake structure concrete walls with Devoe Devran 140 Epoxy Compound and Kansai Biox and the steel surface with Amerlock 400 and Kansai Biox as coating systems constitutes an unreviewed safety question as defined in 10 CFR 50.59(a)(2):

1 ~ Does the proposed coating of the intake structure surfaces increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The delamination of the coating systems cannot initiate an accident. As discussed in the failure modes and effects section of this evaluation, the industry data and the data from the St. Lucie demonstration well confirms the performance of the coating systems to adhere to the properly prepared concrete and steel surfaces of the intake structure. The test well also confirms that the coating does not fail in large layers. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR has not been increased and the proposed change does not affect any accidents discussed in the UFSAR.

2 ~ Does the proposed coating of the intake structure surfaces increase the consequences of an accident previously evaluated in the UFSAR?

The demonstration well confirms that the coating systems will adhere to the intake structure walls and steel surfaces, and therefore will not clog the ICW strainers or degrade CCW heat exchanger performance. Therefore,'the proposed coating of the intake structure surfaces with the coating systems will not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed coating of the intake structure surfaces increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The coating systems will be applied in accordance with Specification CN-2.27 and a new Coating Data Sheet. This specification was used to coat a demonstration well at St.

Lucie to confirm the adhesion properties of the coating and 110

verify the preparation requirements. The demonstration well confirms that the coating performs as the industry data indicates, has excellent adhesion to the intake structure surfaces and does not fail in large pieces which could clog the ICW strainers or degrade CCW heat exchanger performance.

Therefore, the proposed coating of the intake structure surfaces with these coating systems will not change the probability of occurrence of a malfunction of equipment previously evaluated in the UFSAR.

4 ~ Does the proposed coating of the intake structure surfaces increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

Since the demonstration well confirms that the coating systems will adhere to the intake structure surfaces and therefore will not clog the ICW strainers, the consequences from the previously evaluated losses of the intake cooling water system remain valid. Therefore, the proposed coating of the intake structure surfaces does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR.

5. Does the proposed coating of the intake structure surfaces create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The UFSAR evaluates the'ailure of an ICW pump or the intake strainer which are the only credible failure modes which this change could create. Evaluation of the hypothetical simultaneous failure of the coating systems on both trains of ICW intake wells during an accident sequence concludes that due to the excellent adhesion properties of this coating system, this is not a credible event. Therefore, no different type of accident is created and the accident assumptions found in the UFSAR remain valid.

6. Does the proposed coating of the intake structure surfaces create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

One failure mode involves the hypothetical simultaneous failure of the coating system on both trains of ICW intake well walls during an accident sequence and it was concluded that due to the excellent adhesion properties of this coating system this was not a credible event. The malfunction of a single ICW pump or ICW strainer is evaluated in the UFSAR and the results remain valid. Therefore, the proposed coating of the intake structure surfaces does not create the possibility 111

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of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR.

7. Does the proposed coating of the intake structure surfaces reduce the margin of .safety as defined in the basis for any Technical Specification' The proposed change does not affect any Technical Specification nor does it reduce any margins of safety defined by the Technical Specifications. The proposed change will improve heat exchanger performance by reducing marine growth in the intake structure wells.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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Unit: 1 JPN-PSL-SENP-94-047

Title:

SIT Discharge/Loop Check Valve Stroke Test ABSTRACT This safety evaluation demonstrates the acceptability of performing a full stroke test of the Safety Injection Tank (SIT) discharge/loop check valves. The proposed test will be used to address NRC requirements for SIT check valves testing delineated in Generic Letter 89-04.

The test is to be performed during refueling with the reactor vessel head and upper internals removed, and the refueling cavity flooded. The test consists of establishing a reduced nitrogen pressure in the SIT to be used as a motive force to discharge a portion of the SIT inventory through the subject check valves. The test is initiated by opening the SIT discharge line motor operated valve and observing via acoustic and magnaflux monitoring that the check valves fully open.

Calculations have been performed to demonstrate that sufficient velocities can be achieved and maintained to fully open the subject check valves. This evaluation addresses the various potential adverse effects on the plant including those on the reactor internals and steam generator nozzle dams resulting from the proposed flow test.

this safety evaluation Since test on related.

assesses safety related components, ittheiseffects of the proposed classified as safety This evaluation demonstrates that the proposed test neither involves an unreviewed safety question nor requires a change to plant Technical Specifications, and does not adversely affect plant operation or safety. Therefore, it is concluded that the proposed test may be performed without prior NRC approval.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the following questions serve to determine whether the proposed test constitutes an unreviewed safety question:

Does the proposed test increase the. probability of occurrence of an accident previously evaluated in the UFSAR?

Performance of this test will not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

Specifically, the test will not increase the probability of a 113

0 loss of shutdown cooling and will not cause an upset of fuel assemblies. The maximum fluid velocity that will occur in the vessel, as a result of this test, is on the order of 1/2 ft/sec, and therefore, is not capable of adversely impacting the fuel or CEAs.

2 ~ Does the proposed test increase the consequences of an accident previously evaluated in the UFSAR?

This test will be conducted in Mode 6 with the reactor vessel head removed. This test does not impact any of the assumptions for refueling accidents evaluated in the UFSAR and will not increase the consequences of any of these analyses.

3 ~ Does the proposed test increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

cooling system will be secured in the tested If The shutdown loop. There are no potential failure modes resulting from this test that would adversely impact shutdown cooling equipment. The additional pressure that the nozzle dams may be subjected to as a result of this test have been analyzed and determined not to affect nuclear safety. Therefore, this test will not increase the probability of a malfunction of equipment important to safety.

Does the proposed test increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The consequences of malfunction of equipment important to safety are unaffected by this test. The consequences of a loss of shutdown cooling 'or a failure of the steam generator nozzle dams are the same regardless of this testing.

5. Does the proposed test create the possibility of an accident of a different type than any previously evaluated in the UF SAR'?

The possibility of introducing nitrogen into the RCS as a result of this test has been evaluated. The initial test parameters have been chosen such that after the nitrogen gas has fully expanded there will still be liquid inventory remaining in the SIT. By conducting the test within the bounds of these 'nitial parameters, the possibility of injecting nitrogen into the RCS has been precluded.

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Potential effects on safety are discussed in this evaluation.

None of these'ffects will result in an accident of a different type than .previously evaluated in the UFSAR.

6. Does the proposed test create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR7 This test does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR since the failure modes of the nozzle dams have been previously evaluated.

7. Does the proposed test, reduce the margin of safety as defined in the basis for any Technical Specification?

The proposed test does not change assumptions in the basis for any of the Technical Specifications. Therefore, the proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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Unit:

~

JPN-PSL-SEMP-94-050

Title:

~

~ Temporary Alterations to the Refueling Water Tank ABSTRACT During the upcoming St. Lucie Unit 1 1994 Fall Refueling Outage, repairs are to be made to the Refueling Water Tank (RWT), per engineering evaluation JPN-PSL-SENP-93-035. The RWT must be completely drained in order to implement the required repairs. The removal of the RWT inventory can be expedited by the implementation of certain temporary changes. These changes are as follows:

1) The first proposed alteration is the addition of temporary piping inside the RWT. The proposed piping arrangement would serve to extend the fuel pool line connection to within several inches of the tank bottom. This would allow the use of the Fuel Pool Purification pump to drain the tank below the elevation of the fuel pool line connection.

2) The second temporary change would be the connection of a temporary valve and piping arrangement to the RWT drain line.

The exterior drain connection has a valve and blind flange.

The blind flange would be removed and a temporary valve and piping arrangement would be attached to permit the stroking of valve I-V-07,-106 to verify its operation. This drain line will be. used during the Unit 1 outage to pump a portion of the RWT inventory to the Waste Management System in the Reactor Auxiliary Building (RAB).

It is the intent to implement these temporary changes 'while St.

Lucie Unit 1 maintains power operation. Neither of the proposed changes require a permanent change to the facility. The 50.59 safety evaluation will address the acceptability of implementing these temporary changes during operation. The make-up water supply to the Spent Fuel Pool is not affected by the addition of the temporary piping. With the exception of cycling I-V-07-106, the modification to the drain line will not be put into service until the unit is shutdown. The tank boundary will not be altered, unless valve I-V-07-106 is found to leak in the closed position.

Engineering Evaluation JPN-SENP-93-035 provides the 50.59 safety evaluation for the RWT repair.

This evaluation concludes that the activity as described above does not represent an unreviewed safety . question as defined in 10CFR50.59, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety. Therefore, prior NRC approval is not required.

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SAFETY EVALUATION In accordance with 10 CFR 50.59, the following evaluation serves to determine whether this modification constitutes an unreviewed safety question or requires a change to the Technical Specifications:

1 ~ Does the proposed change increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed temporary changes do not affect any equipment whose malfunction is postulated in the UFSAR to initiate an accident or prevent an accident from occurring. The proposed change maintains the Refueling Water Tank's ability to perform its intended functions. As such, the probability of occurrence of an accident previously evaluated in the UFSAR has not increased.

2 ~ Does the proposed change increase the consequences of an accident previously evaluated in the UFSAR?

The proposed temporary changes do not diminish in any way the ability of the Refueling Water Tank to perform its intended function to mitigate the consequences of an accident previously evaluated in the UFSAR. The inventory of borated water required by the Emergency Core Cooling Systems is maintained. As such, the proposed change does not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed change increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The proposed temporary changes maintain the quality level and the level of protection previously established for the Refueling Water Tank. The proposed temporary changes do not affect the boundary integrity of the tank. As such, the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased by the proposed change.

4 ~ Does the proposed change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The Refueling Water Tank has a passive function to maintain the required inventory of borated water for the Emergency Core Cooling Systems. The proposed temporary changes do not affect the level of borated water in the Refueling Water Tank. The 117

ap flow path from the Refueling Water Tank to the Emergency Core Cooling Systems is not affected by the proposed changes.

Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased by the proposed change.

5. Does the proposed change create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

As discussed in failure modes and effects analysis section of this evaluation, the proposed temporary changes do not introduce any new failure modes. The proposed temporary changes are intended to maintain the requirements of the design bases of the Refueling Water Tank as described in the UFSAR. Therefore,. the possibility of an accident of a different type than any previously evaluated in the UFSAR is not created by this modification.

6. Does the proposed change create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The proposed temporary changes do not interact spatially or functionally with any structure, system or component important to safety other than the Refueling Water Tank. No new failure modes are created for the temporary proposed changes that can be postulated to cause a malfunction of equipment important to

'safety different than those previously analyzed in the UFSAR.

Therefore, the possibility of a malfunction of equipment important to safety which is of a different type than any previously evaluated in the UFSAR is not created by the proposed change.

7~ Does the proposed change reduce the margin of safety as defined in the basis for any Technical Specification?

The Technical Specification requirements and Technical Specification Bases are not affected by the proposed temporary changes. The proposed temporary changes do not affect any plant Technical Specification requirement. The proposed change maintains the quality and level of protection previously evaluated in the UFSAR. Therefore, the margin of safety as defined in the bases for any Technical Specification is not reduced-by this modification.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical 118

Specifications and does not adversely affect plant operation or safety.

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Unit: JPN-PSL-SENP-94-065

Title:

Containment Air Conditioning for Refueling Outage ABSTRACT This safety evaluation demonstrates the acceptability of temporarily modifying the component cooling water (CCW) system during the 1994 refueling outage to provide air conditioning inside containment via a closed loop chilled water system. This temporary change will control the temperature and humidity inside containment during the refueling outage. This system utilizes two 200 ton self-contained chilled water units located outdoors between the reactor auxiliary building (RAB) and the fuel handling building to supply chilled water to the coils in the containment fan coolers.

The containment fan cooler fans move air across the coils cooling the air while the heat is carried away by the water. Temporary hoses will be used to transport the chilled water into the RAB pipe penetration area where connection via flanged spool pieces is made into the existing CCW supply/return piping for the containment fan coolers. The connection is made by removing the normally locked open/throttled CCW supply/return isolation valves and installing the spool pieces. The temporary change will be made to a single train (one or two containment fan coolers) at a time, allowing the other train to be available for reduced inventory evolutions. The spool pieces are blind flanged such that the CCW system upstream of these connections are unaffected by the chilled water and remain operable. Temporary cables will be used to power the chilled water units from an offsite power source and would therefore not impact any safety related or non-safety related plant power supply.

There are no permanent modifications/configurations affecting operation of the CCW system, containment fan coolers, or the electrical power supply. All system alignments are temporary. The system, equipment and piping will be restored to its normal condition at the end of the refueling outage, prior to entering Mode 4.

This safety evaluation documents the temporary design of the CCW system to provide containment air conditioning during the refueling outage. Although the temporary CCW configuration will be operated at temperatures lower than normal, this safety evaluation concludes that the temporary CCW configuration utilizing chilled water and operation of the containment fan coolers is consistent with the design intent of the CCW system and does.not have a'n adverse affect on plant safety, security, or operation.

Since this safety evaluation assesses the effects on containment fan cooler operation utilizing chilled water for containment air I

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conditioning, supplied through CCW essential headers A and B, this

~ ~ ~

evaluation is classified as safety related.

~

This safety evaluation demonstrates that the temporary configuration of the CCW system to supply chilled water through the fan cooler supply/return header for containment air conditioning during the refueling outage neither involves an unreviewed safety question, requires a change to plant Technical Specifications, and does not adversely affect plant operation or safety. Therefore, is concluded that the proposed temporary CCW configuration to it provide containment air conditioning during the refueling outage may be implemented without prior NRC approval.

SAFETY EVALUATION In accordance with 10 CFR 50.59, the responses to the following questions serve to determine whether the proposed temporary CCW configuration constitutes an unreviewed safety question:

1. Does the proposed temporary change increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The probability of occurrence of an accident previously evaluated in the UFSAR will not increase because neither implementation of the modification to the CCW system nor operation of the containment fan coolers in the temporary CCW configuration affect any equipment postulated in the UFSAR to initiate an accident or prevent an accident from occurring.

Therefore, the proposed temporary change does not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

2 ~ Does the proposed temporary change 'increase the consequences of an accident previously evaluated in the UFSAR?

The consequences of an accident previously evaluated in the UFSAR will not increase because neither implementation of the modification to the CCW system nor operation of the containment fan coolers in the temporary CCW configuration affect any structures, systems or components that functions to mitigate the consequences of an accident, to contain or detect the release of radioactivity or to provide post-accident shielding. The containment fan coolers and CCW flow for containment fan coolers are not required while in Mode 5 or 6.

The temporary CCW configuration will be restored to its normal condition at the end of the refueling outage, prior to entering Mode 4.

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e 0

3. Does the proposed temporary change increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The failure modes and effects analyses of the CCW system as described in the UFSAR are not changed by the temporary CCW configuration. The portion of CCW system affected by this temporary configuration is isolated from the remainder of the CCW system. During the refueling outage, this temporary CCW configuration is considered not to interact functionally with any structure, system or component important to safety.

Additionally, the design of the temporary system is within the design envelop of CCW system 'and does not adversely affect the CCW system. Therefore, the proposed temporary change does not increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR.

4 Does the proposed temporary change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The failure modes and effects analyses of the CCW system as described in the UFSAR are not changed by the temporary CCW configuration. This temporary CCW configuration is considered not to interact spatially or functionally with any structures, systems or components that functions to mitigate the consequences of an accident, to contain or detect the release of radioactivity or to provide post-accident shielding.

Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR is not changed.

5~ Does the proposed temporary change create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

The failure modes and effects analyses of the CCW system as described in the UFSAR are not changed by the temporary CCW configuration. The temporary CCW configuration provides chilled water flow to the containment fan coolers under conditions within the design of the CCW system. Operation in the temporary configuration does not adversely affect the system. The CCW system will be restored to its original design prior to entering'ode 4, following the refueling outage. No new hazards are created that can be postulated to cause an accident different than those previously analyzed in the UFSAR. Therefore, there is no possibility that an -

accident may be created that is different from one already evaluated in the UFSAR.

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4

s. Does the proposed temporary change create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR?

The failure modes and effects analyses of the CCW system as described in the UFSAR are not changed by the temporary CCW configuration. The temporary CCW configuration provides chilled water flow to the containment fan coolers under conditions within the design of the CCW system. Operation in the temporary configuration does not adversely affect the system. The CCW system will be restored to its original design prior to entering Mode 4 following the refueling outage. Interaction due to condensation runoff was addressed in the analyses and effects on safety section with recommendations to assure that there will be no adverse interaction. The temporary configuration is considered not to create any new hazards which can be postulated to cause a malfunction of equipment important to safety different than those previously analyzed in the UFSAR. Therefore, the proposed temporary change does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR.

7~ Does the proposed temporary change reduce the margin of safety as defined in the basis for any Technical Specification?

The Technical Specification requirements and bases applicable to this temporary CCW configuration are not affected by this temporary change. The equipment operated due to this temporary configuration is not required by the Technical Specification while temporary containment cooling is in service (i.e., Modes 5, and 6). The CCW system will be restored to its original design prior to entering Mode 4 following the refueling outage. Therefore, the temporary CCW configuration and operation of the containment fan coolers while in this'emporary configuration does not reduce the margin of safety as defined in the bases for the Technical Specifications.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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Unit: 2 JPN-PSL-SEEP-94-066

Title:

Safety Evaluation for Operation of Three Charging Pumps ABSTRACT The purpose of this safety evaluation is to demonstrate acceptability of the concurrent operation of three charging pumps.

This safety evaluation revises the UFSAR EDG loading table (Table 8.3-2) to indicate the auto-start of 2 charging pumps on one EDG and the UFSAR description of the charging pumps control logic.

Operation of the charging system is with two charging pumps operating; one running continuously, one in automatic standby, and one in off. The charging pump in off condition will not respond to a SIAS start; however, the charging pump running continuously and the charging pump in automatic standby will respond to a SIAS start. This results in loading only one charging pump on each emergency diesel generator in the event of a LOOP or coincident LOOP/LOCA. The present emergency diesel generator loading analysis accounts for loading of one charging pump in the first load block after diesel generator breaker closure.

During plant operations which require maximum RCS purification or inventory makeup, the CVCS may be operated with all three charging pumps running concurrently. This could result in the loading of two charging pumps on one diesel generator in the event of a LOOP or LOOP/LOCA.

This safety evaluation involves assessment of the charging and safety related onsite power distribution systems and is therefore classified as safety related. Operation of three charging pumps will not exceed the capability of the emergency diesel generators in the event of a loss of offsite power or loss of offsite power coincident with a LOCA and is within the design parameters of the CVCS. There is no unreviewed safety question as defined in 10CFR50.59, no changes are required to plant Technical Specifications, and safe plant operation is not adversely affected.

SAFETY EVALUATION In accordance with'0CFR50.59, the responses to the following questions serve to determine whether the operation of three charging pumps constitutes an unreviewed safety question:

1. Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the UFSAR?

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0 The increase in the first block loading is within the capability of the emergency diesel generator. The LTOP analysis, charging system stress and fatigue analysis, the UFSAR Chapter 15 analysis and RCP seal injection are not adversely affected. Therefore, it does not increase the probability of occurrence of an accident since this could not initiate an accident previously evaluated in the UFSAR.

2 ~ Does the proposed activity increase the consequences of an accident previously evaluated in the UFSAR?

The operation of three charging pumps has been shown to have no adverse effect on the safety functioning of the CVCS system or the EDGs and their ability to mitigate the effects of an accident have not changed. Therefore, the consequences of an accident previously evaluated in the UFSAR has not been increased.

3 ~ Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The increased EDG loading due to the operation of three charging pumps has been found to be within the design capacity of the EDG. In addition, other components in the CVCS are not affected by running three charging pumps. Therefore, the probability of occurrence of a malfunction of equipment important to safety. previously evaluated in the UFSAR has not been increased.

4 ~ Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR?

~

increased EDG loading has been found to be within the

'he design capacity of the EDG". The additional flow from operating three charging pumps is within the design parameters of the CVCS components. Therefore, there is no increase to the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the UFSAR?

'I g The increased EDG loading has been found to be within the design capacity of the EDG. The additional flow from operating three charging pumps is within the design parameters of the CVCS components. No new failure modes have been created. Therefore, no additional possibilities have been 125

r created for an accident of a different type than any previously evaluated in the UFSAR.

6. Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different .

type than previously evaluated in the UFSAR?

The increased EDG loading has been found to be within the design capacity of the EDG. The additional flow from operating three charging pumps is within the design parameters of the CVCS components. Therefore, the possibility of a malfunction of. equipment important to safety of a different type than previously evaluated in the UFSAR has not been created.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

The maximum loading of the EDG is maintained at less than the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> maximum loading, in accordance with Technical Specifications. The function and components of the CVCS are not affected by the operation of three charging pumps.

Therefore, the margin of safety, as defined in the basis for the Technical Specifications, has not been reduced.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

126

e Unit:

Title:

Pressure 1 JPN-PSL-SEMP-94-076 Increase of Engineered Safeguards Suction Piping Design ABSTRACT After successful performance of a motor operated valve differential pressure test as required by NRC Generic Letter 89-10, water was observed on the floor of the pipe tunnel in the Reactor Auxiliary Building. The source of the water was determined to be from relief valve, I-SR-07-1A, located on the 1A Engineered Safeguards suction piping.

An examination of the alignment used for the test revealed a flow path from the discharge of the 1B Containment Spray pump to the suction of the 1A Containment Spray pump through a common header in the Sodium Hydroxide (NaOH) Spray Additive system. With the 1B Containment Spray pump operating and all the A ECCS pumps secured, the A train suction piping was pressurized through the Spray Additive system common header. As part of the investigation of the event, the 1B Low Pressure Safety Injection pump was aligned to the B Containment Spray header and discharged through the B Shutdown Cooling heat exchanger. The maximum pressure observed at the 1A Low Pressure Safety Injection pump discharge header was 80 psig.

Relief valve I-SR-07-1A was determined to discharge. Plant personnel documented the event on St. Lucie Action Request (STAR) 1-94100259, which was assigned to Nuclear Engineering for disposition.

An Initial Assessment of Operability was performed to respond to the concerns addressed by the STAR. A calculation determined that the components whose design pressure had been exceeded were in fact capable of withstanding considerably higher pressures. The suction piping and components, therefore, did not suffer any damage as a result of the event. However, with regard to the issue of the system design, it was concluded that a design basis scenario exists which could result in lifting the relief valve; a containment spray actuation signal (CSAS) and a loss of off-site power (LOOP) coincident with one Emergency Diesel Generator (EDG) failing to operate. The relief valve could potentially open and release containment sump inventory in excess of the Engineered Safeguards equipment external leakage rate of 2 liter per hour, UFSAR 15.4.1.7 and 15.4.1.8. This could result in a condition outside of the design basis of the Engineered Safeguards systems. on the Initial Assessment of Operability, plant management madeBased a one hour non-emergency notification to the Nuclear Regulatory Commission in accordance with 10 CFR 50.72.

127

This safety evaluation was prepared to evaluate the acceptability of higher pressures in the Engineered Safeguards suction piping in order to disable relief valves I-SR-07-1A and I-SR-07-1B for this cycle while above. Mode 4. This interim measure is being implemented to preclude the possibility of an unwanted release.

The proposed change does not require a permanent change to the facility. The safety evaluation addresses the acceptability of implementing the proposed change during operation. This safety evaluation involves Engineered Safeguards systems and is therefore classified as safety related.

Based on the evaluation herein, it has been determined that an unreviewed safety question does not exist and the plant Technical Specifications are not affected. Therefore, prior notification of the NRC is not required.

SAFETY EVALUATION In accordance with 10 CFR 50.S9, the following evaluation serves to determine whether this modification constitutes an unreviewed safety question or requires a change to the Technical Specifications:

1. Does the ,proposed change increase the probability of occurrence of an accident previously evaluated in the UFSAR?

The proposed change does not affect any equipment whose malfunction is postulated in the UFSAR to initiate an accident or prevent an accident from occurring. As such, the probability of occurrence of an accident previously evaluated in the UFSAR has not increased.

2. Does the proposed change increase the consequences of an accident previously evaluated in the UFSAR?

The proposed change does not affect the design function of any equipment designed to mitigate the consequences of an accident previously evaluated in the UFSAR. No new failure modes are being introduced and the design margin of equipment important to safety is not. being decreased. As such, the proposed change does not increase the consequences of an accident previously evaluated in the UFSAR.

3 ~ Does the proposed change increase the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR?

The component affected by this change is the Engineered Safeguards suction piping. The proposed change maintains the quality level and the level of protection previously 128

established for the Engineered Safeguards suction piping. The design allowable pressure rating of the piping and associated components is above the maximum system pressure resulting from this change. Although the new higher pressure is above the original tested pressure, the design margin between allowable stresses and ultimate capacity is not being decreased. The proposed change, therefore, does not affect the pressure boundary integrity of the Engineered Safeguards suction piping. As such, the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased by the proposed change.

4 ~ Does the proposed change increase the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR7 The malfunction evaluated in the UFSAR is the complete or partial failure of one train of Engineered Safeguards to perform its function. The proposed change does not in any way affect the ability of the redundant train of Engineered Safeguards to perform its function to inject and recirculate borated water. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased by the proposed change.

5. Does the proposed change create the possibility of an accident of a different type than any previously evaluated in the UF SAR'P The proposed change does not introduce any new failure modes.

The proposed change serves to reduce the possibility of a component failure. Therefore, the possibility of an accident of a different type than any previously evaluated in the UFSAR is not created by this proposed change.

6~ Does the proposed change create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR'P The proposed change does not interact spatially or functionally with any structure, system or component important to safety other than the Engineered Safeguards suction piping.

No new failure modes are created by the proposed change that can be postulated to cause a malfunction of equipment important to safety different than those previously analyzed in the UFSAR. Therefore, the possibility of a malfunction of equipment important to safety which is of a different type than any previously evaluated in the UFSAR is not created by the proposed change.

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7.~ Does the proposed change reduce the margin of safety as defined in the basis for any Technical Specification?

~ ~

The Technical Specification requirements and Technical Specification Bases are not affected by the proposed change.

The proposed change does not affect any plant Technical Specification requirement. The proposed change maintains the quality and level of protection previously evaluated in the UFSAR. Therefore, the margin of safety as defined in the bases for any Technical Specification is not reduced by this proposed change.

The foregoing constitutes the determination, per 10 CFR 50.59(b),

that the subject activity does not involve an unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety.

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PROCEDURE! 2-LOI-0-65 Unit: 2 DESCRIPTION OF THE CHANGE This change involves the transfer of borated water from a PSL2 Boric Acid Makeup Tank (BAM Tk) to the 2A Holdup Tank (2A HUT).

This will provide a back-up borated water supply for PSL1 during the repairs to the Refueling Water Tank (RWT). This change will be controlled via a procedure (LOI) and will be in effect only during the outage time when the RWT is under repairs.

SAFETY EVALUATION

SUMMARY

The probability of occurrence or the consequences of an accident or malfunction previously evaluated has not increased, since there is no additional likelihood of an evaluated accident occurring due to storage of borated water in the HUT, since this is the design function of the HUT.

The possibility of an accident or malfunction of a different type than any evaluated previously has not been created, since the jumper to be used is of the same design requirements as the piping where it is connected and the procedure calls for an operator to be stationed at the jumper during this evolution.

The margin of safety as defined in the basis for any technical specification is not reduced, since this evolution is bounded by the LCOs associated with Boric Acid Injection flowpaths and reactivity control.

Based on the above, it is concluded that the change in question does not constitute an unreviewed safety question or a change to the Technical Specifications.

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P ROCEDURE HP-74 REVISION 0 Unit: 1 & 2 DESCRIPTION OF THE CHANGE This change involves a change to a procedure as outlined in Chapter 12 of the UFSAR. Specifically, the change involves a revision to the procedural requirements, which are outlined in the UFSAR Chapter 12, used in the issuing of personal dosimetry. Currently in the UFSAR, self-reading dosimeters are called out for as the instruments to be provided for the purpose of external exposure monitoring. This change involves using digital alarming dosimeters.

SAFETY EVALUATION

SUMMARY

The new type dosimeters have a wider range and can be read more accurately than the self-reading dosimeters. The use of the digital dosimeters enhances the ability of an individual in monitor their personal external exposure.

Therefore, this change in exposure monitoring does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated; does not create the possibility for an accident or malfunction of a different type than any previously evaluated; and does not reduce the margin of safety as defined zn the basxs for any technical specxfzcatxon.

~ ~

Based on the above, it is concluded that the change in question does not constitute an unreviewed safety question or a change to the Technical Specifications.

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PROCEDURE'-ONP-01 05 REVISION 0 Unit: 1 DESCRIPTION OF THE CHANGE The proposed change involves using a Containment Spray (CS) pump for decay heat removal during an outage. This alignment would only be used as a contingency plan, should a Low Pressure Safety Injection (LPSI) pump be unavailable. This proposed decay heat removal alignment can only be used during Mode 6, and, with at least 23 feet of water above the top of the irradiated fuel assemblies. If less than 23 feet of water exist above the fuel, the action statements of the Technical Specifications apply. The change would require installation of a flange on the LPSI suction header and the removal of the internals of check valve V07000. An ISLT should be conducted subsequent to installation of the flange and removal of 'check valve internals to insure that a leakage path has not been created. This would allow the use of the 1A CS pump for decay heat removal. The configuration of the plant when this change is allowed would be covered under LCO 3.9.8.1, hence the limitation for this change shall require greater than 23 feet water level above the fuel assemblies, and one SDC train operable and in operation.

SAFETY EVALUATION

SUMMARY

The probability of occurrence or the consequences of an accident or malfunction previously evaluated has not been increased, since no credit is taken for this change as an "operable" SDC train, and since this change has been evaluated against the design criteria for the SDC system. In addition, the CS pump has similar design attributes as the LPSI pump and the differences in configuration as relating to piping, elevation and materials have been evaluated.

The loss of decay heat removal capability has already been evaluated under the system LCO.

The possibility of an accident or malfunction of a different type than any evaluated previously has not been created, since the alternate decay heat removal system will be verified as not creating any leakage paths as a result of this change. Also, since the actions required by the Technical Specifications are being taken (i.e. Containment integrity established) there is no increase in any off-site dose releases previously analyzed. Furthermore, since this configuration cannot be used subsequent to a LOCA, there can be no likelihood of fission products being introduced to the RAB by this change.

The margin of safety as defined in the basis for any technical specification is not reduced, since this change is bounded by the associated with the plant conditions and system LCO associated with it.

Based on the above, it is concluded that the change in question does not constitute an unreviewed safety question or a change to the Technical Specifications.

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PROCEDURE 2 LOZ T 88 Unit: 2 DESCRIPTION OF THE CHANGE This change involves the use of a "roughing" filter for the Fuel Pool Purification System. This change will be controlled via a procedure (LOI) and will be in effect for about 6 weeks. The need for this change involves the desire to remove particulate matter from the fuel pool at rates larger that currently available with the existing filter. This will allow the capability of processing larger volumes of water through the filter while minimizing the amount of times the filter element needs to be replaced.

SAFETY EVALUATION

SUMMARY

The probability of occurrence or the consequences of an accident or malfunction previously evaluated has not increased, since there is no additional likelihood of an evaluated accident occurring due to the addition of the "roughing" filter.

The possibility of an accident or malfunction of a different type than any evaluated previously has not been created, since the filter to be used is of the same design requirements as the previous filter and the system configuration will be controlled by the LOI.

The margin of safety as defined in the basis for any technical specification is not reduced, since there are no technical specifications affected by this change.

Based on the above, it is concluded that the change in question does not constitute an unreviewed safety question or a change to the Technical Specifications.

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Jumper/Li.fted Lead g 2-94-007 Units 2 Component and Systems Affecteds Upender Vertical Circuit Limit Switch (2LS-BV)

Reason for the Request!

To remove contact 2CR-BV which isolates the refueling transfer upender [refueling side only] hydraulic positioning cylinder from the pump when the upender has reached the upper limit switch. The jumper is installed with a ganged switch to the existing upender control switch so that it is only in the circuit when the switch is taken to vertical.

Safety Analysis:

This J/LL does not increase the probability of occurrence of an accident postulated in the UFSAR. The removal of this interlock and the accident scenarios interlocks that it prevents are ensured by other will remain functional.

This J/LL does not increase the consequences of an accident previously evaluated in the UFSAR. The design basis fuel handling accident described in the UFSAR section 15.7.4.1.2 is significantly conservative with respect to the circumstances of this jumper.

This J/LL does not increase the probability of an occurrence of a malfunction of equipment important to safety. The remaining interlocks ensure the equipment is properly protected from damage.

This J/LL does not increase the consequences of a malfunction of equipment important to safety. No failure of the upender hydraulic system will induce a condition different from normal upender use.

This J/LL does not create the possibility of an accident different than described in the UFSAR. The fuel transfer machine and the refueling machine do not interact with plant systems other than refueling equipment. Therefore, the postulated accident in UFSAR 15.7.4.1.2 is bounding.

This J/LL does not increase the possibility of malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR. Failure of the. hydraulic controls is not considered more likely than those failures already postulated.

This J/LL does not reduce the margin of safety as defined in the basis for any technical specification. The jumper affects refueling equipment systems only and does not affect the basis for any technical specification.

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JUMPER/LIFTED LEAD g1-94-019 Unit:

Component and Systems Affected:

Nitrogen to the Sodium Hydroxide(NaOH) Tank Reason for the Request:

This J/LL is to install a temporary nitrogen supply to the NaOH tank during the time that the Reactor Auxiliary Building(RAB) nitrogen header is out of service for modifications. The temporary supply consists of a nitrogen bottle and regulator.

The purpose of the nitrogen supply is to provide a cover gas to inert the tank and. to prevent the NaOH from reacting with the atmosphere. The cover gas serves no safety function and is not designed as safety related. NPSH for the eductors is provided yia atmosphere. There are two vacuum breakers on the top of the NaOH tank that are not affected by this J/LL. The jumper was installed and removed within 2 days.

Safety Analysis:

The nitrogen cover gas is not safety related. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR is not increased.

UFSAR page 6.2-111 states that the nitrogen cover gas is not required for NaOH injection. Therefore, the consequences of an accident previously evaluated in the UFSAR are not increased.

The probability of occurrence of a malfunction of equipment important to safety is not increased. Chemistry periodically monitors the NaOH concentration. Also, the limited time exposure to air should not significantly affect NaOH concentration temporary nitrogen is lost.

if the 136

IMPER/LIFTED LEAD g 1 94 20 Unit:

Component and Systems Affected:

Circulating Water Pumps Reason for the Request:

The plant service water system supplies lube water to the Circulating Water Pumps(CWP). The design is shown on dwg 8770-G-082 and 8770-G-084 sheet 2. To perform maintenance on the service water piping, a backup water supply is needed for the CWP. The plant fire water system was used as the backup. A manifold connected to each CWP lube water supply was'ed from fire hydrant FH-9 located at the intake area. The fire water system has two pumps with a rated capacity of 2500 gpm. (see UFSAR section 3.1.3) Hose station HH-2-9 is operable for fire protection of equipment at the intake structure.

Safety Analysis:

The effects of the alternate lube oil arrangement for the CWP is limited to the fire protection system. This system is not postulated to in the UFSAR as an initiator of an accident.

Therefore, this jumper will not increase the probability of occurrence of an accident previously evaluated in the UFSAR.

The consequences of an accident previously evaluated in the UFSAR are not increased because the alternate arrangement does not affect the ability of the fire system to perform is designed function. A hose station in the area of the intake will be available to protect equipment in the event of a fire.

The fire protection system is designed with isolation valves that allow operators to isolate a rupture in the fire protection system without disabling the remainder of the system. Zn the event that the manifold fails, it could be isolated. Therefore, the probability of occurrence of equipment malfunction important to safety is not increased.

137

Jumper/Lifted Lead g 1-94-030 Unit! 1 Component and Systems Affected:

Temporary Power Panel fed from Intake 480V MCC 1A3 Reason for the Request:

To provide power from 480V MCC 1A3 to a temporary power panel to

, support miscellaneous equipment for Unit 1 intake work.

Safety Analysis:

The jumper does not affect 1A3 LC protection or other electrical distribution protection for non-essential power from the startup transformer The response of the onsite electorial distribution system is not altered and therefore the consequences of an accident are not increased.

Protection of non-essential ties to safety related electrical distribution are not affected. No safety related loads exist on the 1A3 MCC.

This additional load on the 1A3 MCC does not interact significantly with the safety related systems and since the response of the onsite electrical distribution system from a fault is not altered.

Therefore there is no increase in the consequences of a malfunction of equipment important to safety as evaluated in the UFSAR.

The response of the onsite electrical system protection is not affected by this jumper, therefore it does not create the possibility of an accident of a different type than previously evaluated in the UFSAR.

There is no reduction of margin as defined in the basis of any technical specification because 1A3 MCC is part of the non-essential onsite power distribution system.

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Jumper/Lifted Lead g 1-94-031 Unit: 1 Component and Systems Affected:

Temporary Power Panel fed from Intake 480V MCC 1B3 Reason for the Request:

To provide power from 480V MCC 1B3 to a temporary power panel to support miscellaneous equipment for Unit 1 intake work.

Safety Analysis:

The jumper does not affect 1B3 LC protection or other electrical distribution protection for non-essential power from the startup transformer The response of the onsite electorial distribution system is not altered and therefore the consequences of an accident are not increased.

Protection of non-essential ties to safety related electrical distribution are not affected. No safety related loads exist on the 1B3 MCC.

This additional load on the 1B3 MCC does not interact significantly with the safety related systems and since the response of the onsite electrical distribution system from a fault is not altered.

Therefore'there is no increase in the consequences of a malfunction of equipment important to safety as evaluated in the UFSAR.

The response of the onsite electrical system protection is not affected by this jumper, therefore it does not create the possibility of an accident of a different type than previously evaluated in the UFSAR.

There is no reduction of margin as defined in the basis of any technical specification because 1B3 MCC is part of the non-essential onsite power distribution system.

139

Jumper/Lifted Lead g 2-9i-039 Unit! 2 Component and Systems Affected:

Fire Header and Circulating Water Pump Seal Water Reason for the Requests Provide a backup supply of seal water from the fire header to the circulating water pumps while completing work on the service water system.

Safety Analysis:

This J/LL does not increase the probability of occurrence of an accident postulated in the UFSAR. The fire protection system is no postulated to cause an accident, therefore this jumper will not affect accident probability.

This J/LL does not increase the consequences of an accident previously evaluated in the UFSAR. This alternate lineup will not impact the fire protection system from performing its design basis function.

This J/LL does not increase the probability of an occurrence of a malfunction of equipment. important to safety. The fire protection is designed with multiple isolation valves which will be 'ystem isolated in the event of a manifold rupture.

This J/LL does not increase the consequences of a malfunction of equipment important to safety. The alternate arrangement does not create new failure modes because the system is designed with isolation valves.

This J/LL does not create the possibility of an accident different than described in the UFSAR. New failures are not created as a function of this jumper.

This J/LL does not increase the possibility of malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR. The fire protection system is equipped with isolation capability to ensure the entire system is not lost due to a single rupture point.

This J/LL does not reduce the margin of safety as defined in the basis for any technical specification. The fire protection system is not a technical specification system.

140

Jumper/Lifted Lead.g 2-94-052 Units 2 Component and Systems Affecteds Pressurizer Heater Banks B-1 and B-2 Reason for the Requests Pressurizer Heater Bank B-1 circuit g4 has a ground on it. With B-1 and B-4, the banks that can be loaded on the EDG, a heater from bank B-2 will be jumpered to the B-1 bank.

Safety Analysis:

This J/LL does not increase the probability of occurrence of an accident postulated in the UFSAR. The heater banks are not postulated to cause and accident as described in the UFSAR.

This J/LL does not increase the consequences of an accident previously evaluated in the UFSAR. The pressurizer heaters are not mentioned in the UFSAR analysis section 15.1.

This J/LL does not increase the probability of an occurrence of a malfunction of equipment important to safety. Section 8.3 of the UFSAR states the power panels form the first level of protection for a short circuit at the heaters. Therefore, a short would not endanger the safety'related buses.

This J/LL does not increase the consequences of a malfunction of equipment important to safety. A short circuit is protected against by the power panels which are the first level of protection.

This J/LL does not create the possibility of an accident different than described in the UFSAR. No new hazards are created as a function of this jumper.

This J/LL does not increase the possibility of malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR. The jumper does not affect the control circuitry nor the class 1E to non-safety related interface, therefore, no new hazards are created.

This J/LL does not reduce the margin of safety as defined in the basis for any technical specification. The jumper restores the 150 KW to the B-1 heater bank.

141

Jumper/Lifted Lead g 1-Qi-054 Units 1 Component and Systems Affected<

Temporary Substitution of Temperature Indicators for containment Temperature Recorder TR-25-1 Reason for the Request:

The purpose of the Jumper is to provide a temporary means of monitoring containment temperature in compliance with Technical Specification surveillance requirement 4.6.1.5. Normally this is done with TR-25-1 which is temporarily out of service. The temperature inputs from thermocouples TE-25-3, TE-25-5 and TE-25-7 are functioning normally such that they may be used with another indicating source to provide accurate monitoring of this parameter.

Safety Evaluation:

This jumper does interface with any components or equipment other than for the purposes of monitoring. Non of the equipment involved is evaluated for possible malfunctions in the UFSAR.

The jumper does not affect any structures utilized in mitigating accident consequences as evaluated in the UFSAR.

The jumper does not affect any structures utilized in mitigating accident consequences as evaluated in the UFSAR.

The change does not affect the assumptions of containment temperatures credited in the UFSAR.

There is no possibility of a new malfunction of equipment important to safety 142

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