ML17241A358

Changes,Tests & Experiments Made as Allowed by 10CFR50.59 for Period of 970526-981209.

ML17241A358
Person / Time
Site:	Saint Lucie
Issue date:	12/09/1998
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17241A357	List: ML17241A356 ML17241A358
References
NUDOCS 9906110137
Download: ML17241A358 (107)

Text

<1 PJ St. Lucie Unit 2 Docket No. 50-389 L-99-128 Enclosure ST. LUCIE UNIT 2 DOCKET NUMBER 50-389 CHANGESI TESTS AND EXPERIMENTS MADE AS ALLOWED BY 10 CFR 50.59 FOR THE PERIOD OF MAY 2 6 I 1 997 THROUGH DECEMBER '9 I 1 998 9906ii0i37 990607 05000389 PDR ADQCK R PDR

INTRODUCTION This report is submitted in accordance with 10 CFR 50.59 (b),

which requires that:

i) changes in the facility as described in the SAR; ii) changes in procedures as described in the SAR; and iii) tests and experiments not described in the SAR which are conducted without prior Commission approval be reported to the Commission in accordance with 10 CFR 50.59(b)(2) and 50.71(e)(4). ,This report is intended to meet this requirement for the period of May 26, 1997, through December 9, 1998. Note that, where practical, summaries from more recent 10 CFR 50.59 evaluations have also been included in this report.

This report is divided into three (3) sections: the first, changes to the facility as described in the Updated Final Safety Analysis Report (UFSAR) performed by a Plant Change/Modification (PC/M); the second, changes to the facility or procedures as described in the UFSAR not performed by a PC/M and tests and experiments not described in the UFSAR; and the third, a summary of any fuel reload safety evaluations.

Each of the documents summarized i;n Sections 1, 2 and 3 includes a 10 CFR 50.59 safety evaluation that evaluated the specific change(s). Each of these safety evaluations concluded that the change does not represent an unreviewed safety question nor require a change to the plant technical specifications; therefore, prior NRC approval was not required for implementation.

TABLE OF CONTENTS SECTION 1 PLANT CHANGE/MODIFICATIONS PAGE 89042 INSTRUMENT UPGRADE INSTALLATION OF CONDUIT SEALS 95035 DEBRIS FILTER AND CONTINUOUS TUBE CLEANING SYSTEM 10 INSTALLATION PHASE 1 96149 THERMO-LAG WALL MODZFICATIONS 96155 ADDZTION OF EXCESS FLOW ISOLATION VALVE IN THE 12 H2 PIPE IN THE RAB 97017 SET POINT CHANGES FOR INSTRUMENT AIR COMPRESSORS 13 2A AND 2B 97031 GL 96-06 THERMAL PRESSURIZATION RELIEF VALVES 98021 REACTOR COOLANT PUMP MECHANICAL SU SEAL REPLACEMENT 15 WITH N-9000 SEAL 98031 APPENDIX R CABLE REROUTES 98033 REPLACEMENT OF ISOLATION VALVES FOR CONTAINMENT PRESSURE TRANSMITTERS 98034 ULTRA-FINE MEDIA FOR CVCS PURIFICATION FILTERS 18 2A AND 2B 98035 FHB VENTILATION SYSTEM PDS-25-17A&B SETPOINT CHANGE 19 98038 PRESSURIZER HEATER REPLACEMENT 20 98046 TELEPHONE SYSTEM UPGRADE 21 98130 APPENDIX R SAFE SHUTDOWN ANALYSIS 22

SECTION 2 SAFETY EVALUATIONS PAGE SEMS-90-052 GENERZC USE OF SEALANT INJECTION SENP-94-029 SHUTDOWN OPERATIONS CRITERIA FOR REDUCED INVENTORY 25 AND DRAINING THE REACTOR COOLANT SYSTEM SENP-94-043 TEMPORARY REMOVAL OF THE ICW PUMP MISSILE SHIELD SEMP-95-004 REDUCED PRESSURIZER HEATER CAPACITY 27 SEMS-96-052 ISOLATION OF CHARGING PUMP RECIRCULATION LINES 28 SECP-96-059 ST. LUCIE COUNTY SOUTH HUTCHINSON ISLAND 29 WASTEWATER TREATMENT FACILITY AND WASTEWATER COLLECTION/RECLAIMED WATER DISTRIBUTION SYSTEMS SEMS-97-013 IN SITU HYDROSTATIC TESTING OF STEAM GENERATOR 30 TUBE FLAWS SEMS-97-017 TEMPERATURE CONTROL VALVE FOR THE HYDROGEN COOLERS 31 (TCV-13-15) MANUAL OPERATION SENS-97-038 SENS-97-041 SENS-97-044 PRESSURIZER CODE SAFETY VALVE MODIFICATION PARTIAL STROKE TESTING OF SZT DISCHARGE VALVES CONTAINMENT SUMP SCREENS

'332 SEMS-97-055 POST ACCIDENT SAMPLING SYSTEM FSAR RECONCILIATION 35 SEMS-97-066 ADDITION OF HYDROGEN PEROXIDE TO THE RCS DURING SDC 36 SECS-97-069 SITE SANITARY SYSTEM MODIFICATIONS 37 SEES-97-077 BYPASS OF THE "REGULINER" VOLTAGE REGULATOR CIRCUIT 38 IN THE NON-SAFETY-RELATED VITAL SUPS INVERTER SECS-97-078 TEMPORARY LEAD SHIELDING INSTALLATION CRITERIA AND 39 RESTRICTIONS SEMS-97-079 INSTALLATION OF A SECONDARY SA HEADER AND TIE-IN TO 40 THE SA SYSTEM IN THE TGB SENS-97-081 OPERATION OF THE MSIV BYPASS VALVES DURING PLANT WARM-UP SEMS-97-097 FIRE PROTECTION SYSTEM DOCUMENTATION UPDATE SENS-97-098 USE OF THE FIRE MAIN TO SUPPLY CIRCULATING WATER PUMP BACKUP LUBE WATER AND THE CONDENSER TUBE CLEANZNG SYSTEM SENS-97-101 ALTERNATE METHODS FOR MONITORING REACTOR VESSEL INVENTORY SEIS-97-108 REACTOR PROTECTION SYSTEM LOGIC MATRIX TEST STATUS 45 LIGHTS

SECTZON 2 SAFETY EVALUATIONS (Continued) PAGE SEMS-98-005 FSAR DISCREPANCIES ON ECCS PUMP NPSH SEF J-98-005 REEVALUATION OF ST. LUCIE UNZT 2 CONTROL ELEMENT ASSEMBLIES LIFETIME BASED ON TWO CEA INSPECTION RESULTS SENS-98-.007 IMPLEMENTATION OF ZCW/CCW OPERATIONAL TEMPERATURE CURVES SENS-98-011 ONSITE STORAGE OF RADIOACTIVE MATERIALS IN A REMOTE RCA SEMS-98-012 ADDITZON OF DIMETHYLAMINE TO THE SECONDARY SYSTEM 50 SEFJ-98-013 AMENDMENT TO CORE OPERATING LIMITS REPORT 51 SEMS-98-017 BIOCIDE TREATMENT OF CLOSED COOLING WATER SYSTEMS 52 SEIS-98-020 PROTECTION SYSTEM TEST CONFIGURATION WITH A 53 POSTULATED SINGLE FAILURE CONCERN OF THE CONTAINMENT PRESSURE TRANSMITTER INPUT SIGNALS SEIS-98-021 UNIT 1 & 2: RSPT SIGNAL RESTORER EVALUATION SEMS-98-024 CONTROL ROOM PAINTING 55 SESJ-98-026 'NALYSIS OF CAPSULE 263 FROM THE REACTOR VESSEL 56 IRRADIATION SURVEILLANCE PROGRAM SENS-98-027 SHARED SYSTEMS AND INTERCONNECTIONS 57 BETWEEN UNIT 1 AND UNIT 2 SENS-98-028 DUMMY FUEL ASSEMBLY USE - UFSAR CHANGES 58 SENS-98-030 IMPLEMENTATION OF AN EXPERIMENT TO DETERMINE 59 NEW OPERATIONAL SETTINGS FOR FHB VENTILATION SENS-98-038 TEST OF P-10 & P-11 CONTAINMENT PENETRATIONS WITH 60 OPEN SHIELD BUILDING DOOR ACCESS PORT SENS-98-051 TEMPORARY INSTALLATION OF A D1AGNOSTIC SUBROUTINE 61 FOR DIGITAL DATA PROCESSING SYSTEM (DDPS)

SENS-98-056 "'FSAR UPDATE FOR GRID STABILITY ANALYSIS 62 SENS-98-059 SPENT RESIN TANK DEWATERING OPERATIONS 63 UFSAR CHANGES SENS-98-062 1998 FSAR REVIEW FINDINGS REQUIRING CHANGES OR CLARIFICATZONS TO THE FSARs ZN ACCORDANCE WITH 10 CFR 50.59 SECS-98-064 OPERATION OF SPENT FUEL CASK CRANE "RESTRICTED ZONE" 65 ADJACENT TO FUEL HANDLING BUILDING

1 l 0

SECTION 2 SAFETY EVALUATIONS (Continued) PAGE SENS-98-065 RADIATION PROTECTION UFSAR CHANGES SENS-98-066 INSTRUMENT RANGE AND ACCURACY IN UFSAR 67 SENS-98-068 LIQUID WASTE MANAGEMENT SYSTEM - UFSAR CHANGES 68 SENS-98-072 WARMING, STARTING AND OPERATING THE SDC SYSTEM 69 SENS-98-073 USE OF SUPPLEMENTAL GAGES TO MONITOR THE CCW 70 HEAT EXCHANGER INLET STRAINER DIFFERENTIAL PRESSURE SEMS-98-074 OPERATION AND TESTING OF THE AFW PUMP SUCTION 71 CROSS-TIE BETWEEN UNIT 1 AND UNIT 2 AT THE UNIT 2 CST SENS-98-076 UFSAR CHANGES TO FUEL POOL SYSTEM PERFORMANCE 72 MONITORING DESCRIPTIONS SENS-98-083 SOLID WASTE MANAGEMENT SYSTEM UFSAR CHANGES 73 SENS-98-085 FLASH TANK OPERATION UFSAR CHANGES SEMS-98-087 INCREASED SDC PURIFICATION FLOW 75 SENS-98-089 INTAKE COOLING WATER STRAZNER UFSAR CHANGES 76 SENS-98-090 USE OF PRIMARY SYSTEM IN-LINE SAMPLE ANALYSIS 77 FOR DISSOLVED GASES SENS-98-091 USE OF CVCS PURIFICATION/DEBORATZNG ION EXCHANGERS 78 UFSAR CHANGES SEIS-98-093 INSTALLATION OF JUMPERS ACROSS VARIOUS ISOLATION 79 CABZNET CZRCUITS DURZNG MODES 5 AND 6 SENS-98-094 LETDOWN RADIATION MONITOR UFSAR CHANGES 80 SEES-98-097 TEMPORARY POWER CONNECTIONS TO P.A. SYSTEM PWR. 81 DIST. CAB. DURING MODES 5 AND 6 SENS-98-098 UPPER GUIDE STRUCTURE LIFT RIG USE WITHOUT 82 POSITIONING BUSHING SEIS-98-099 INSTALLATION OF JUMPERS ACROSS SB/MB/MD ISOLATION 83 CABINET CIRCUITS DURING MODES 5 AND 6 SEMS-98-101 UNZT 2 STEAM GENERATOR SECONDARY SIDE FOREIGN OBJECTS 84 SENS-98-104 B" TRAIN ESF TESTING WITH THE "B" ICW HEADER 85 OUT OF SERVICE SEMS-98-119 OPERATION WITH MOTOR OPERATED VALVE V3523 OPEN 86 SENS-99-004 PROCESS SAMPLING SYSTEM UFSAR CHANGES 87 SENS-99-005 CHANGES TO REACTOR COOLANT, AND REACTOR MAKEUP 88 WATER CHEMISTRY SPECIFZCATIONS

SECTION 3 RELOAD SAFETY EVALUATIONS PAGE 98016 ST. LUCIE UNIT 2 CYCLE 11 RELOAD 90

SECTION 1 PLANT CHANGE / MODIFICATIONS

PLANT CHANGE/MODIFICATION 89042 INSTRUMENT UPGRADE INSTALLATION OF CONDUIT SEALS

~SUIMtlR This modification provided for the replacement of Conax conduit seals with NAMCO conduit seals for various plant instruments (transmitters, solenoid valves and limit switches), including safety related devices, since some seals require replacement and Conax seals are no longer available. The replacement seals are qualified for use under seismic, LOCA and HELBA conditions and have been environmentally qualified through vendor testing. The replacement seal design also incorporates a quick disconnect feature for ease of maintenance.

9

PLANT CHANGE/MODIFICATION 95035 DEBRIS FILTER AND CONTINUOUS TUBE CLEANING SYSTEM INSTALLATION PHASE 1

~Summa A 1994 plant study concluded that performance of the circulating water system and main condenser could be improved with a debris filter system (DFS) and a condenser was tube cleaning system (CTCS).

contracted to provide and Taproggee America Corporation (TAC) install the major system components, including the modification of the circulating water piping system.

The entire project is implemented via this PC/M and PC/M 95036 (two PC/Ms were used in order to differentiate the implementation scopes of the contract with TAC).

A DFS was installed on each of the condenser inlet lines in order to reduce macro fouling of the condensers. Any debris is periodically backflushed from the filters and is directed to the discharge side of the condensers. A CTCS was al'so installed.

The CTCS design utilizes cleaning balls, which are injected into the condenser inlet, pass through the condenser tubes and are captured at the condenser outlet for reuse.

This modification included consideration of material compatability, electrical power requirements, syst: em interfaces (e.g , service water, instrument air, electrical), potential impact on the plant security plan, cathodic protection, structural considerations, and circulating water system flow requirements. As such, the design bases of the circulating water system were preserved with the installation of the DFS and CTCS.

10

PLANT CHANGE/MODIFICATION 96149 THERMO-LAG WALL MODIFICATIONS This PC/M implemented modifications and upgrades to several Thermo-Lag walls in the reactor auxiliary building (RAB).

Additionally, modifications were also made to mechanical and cable tray penetrations to restore them to qualified configurations. These changes are related to the plant's commitments associated with Generic Letter (GL) 92-08.

The NRC, via GL 92-08 and related correspondence, identified that Thermo-Lag 330-1 fire barrier materials have failed to meet their intended ratings during fire endurance testing. Testing of Thermo-Lag walls similar to those at St. Lucie showed that the walls failed to meet the temperature requirements for a 3-hour fire barrier. St. Lucie evaluation PSL-FPER-96-002 evaluated these walls per the guidance of GL 86-10 and determined they were adequate to withstand the fire hazards, provided certain modifications are implemented. These modifications generally pertain to the reinforcement of the Thermo-Lag panels at the seams and additional attachment of the Thermo-Lag panel to the structural frames. A plant walkdown also identified penetration modifications that were required. The penetration issues included improper annulus materials, lack of proper end board on cable tray penetrations and lack of properly configured penetration seals for cable trays. This PC/M remedied the above concerns.

11

PLANT CHANGE/MODIFICATION 96155 ADDITION OF EXCESS FLOW ISOLATION VALVE IN THE H2 PIPE IN THE RAB This modification added an excess flow isolation valve to the 1-inch, non-safety hydrogen line that enters the reactor auxiliary building (RAB) through the west wall at the -0.5 ft. elevation.

The new valve limits hydrogen flow to 35 scfm and is required to ensure that a break in the hydrogen line will not result in exceeding a 2% hydrogen concentration in the RAB. The maximum hydrogen flow was determined in calculation PSL-2FSL-96-025. The new valve and its upstream piping (up to the wall penetration) have been seismically supported.

This modification also added a restricting orifice in the hydrogen line immediately downstream of the volume control tank hydrogen regulator. This orifice is intended to limit system flow to a value that will provide adequate service to the VCT without activating the excess flow valve.

None of the above changes alter the'unction or operation of the affected systems.

12

PLANT CHANGE/MODIFICATION 97017 SET POINT CHANGES FOR INSTRUMENT AIR COMPRESSORS 2A AND 2B

~SUIAtllR The instrument air (IA) system is a non-safety system that provides a reliable supply of clean, oil-free compressed air for use by various pneumatically operated components. Per the original design, the set points for the 2A and 2B compressors were 92/98 psig for the full load/unload configuration and 88/96 psig for the auto start/reset configuration. The set points for the 2C and 2D compressors are 110/119 psig for load/unload and 105/113 psig for auto start/reset. This modification revised the 2A and 2B compressor set points to 110/115 psig and 100/107 psig.

Prior to this modification, the new setpoints were temporarily installed via procedure 2-LOI-T-80, which demonstrated improved system performance with the new values.

IA compressors 2C and 2D provide the principle source of compressed air during normal plant operations and compressors 2A and 2B normally remain off, but available for use under abnormal operating conditions. Raising the set points improves the system operating margin and improves the process of surveillance testing since the original set points were too low to allow the compressors to load until the air receiver pressure decays to a low set point (92 psig). The increased discharge pressure and outlet air temperatures are within the capability of the compressors and the IA system. This modification has no adverse impact on cooling water flow requirements.

13

PLANT CHANGE/MODIFICATION 97031 GL 96-06 THERMAL PRESSURIZATION RELIEF VALVES

~SUIIUIIS Generic Letter 96-06 identified the potential for thermally induced pressurization of lines penetrating containment due to design basis accident heating of water-filled, isolated sections of piping. Of specific concern was the potential for damage .to

.the containment boundary. A plant review identified several locations that warranted design enhancement.

This modification installed thermal relief valves to the service water, primary make-up water, component cooling water, containment spray and chemical a volume control systems. A total of eight thermal relief valves were added to seven penetrations.

The valves were installed inside containment'o as not to create additional containment leakage paths. The set points for the new valves were established to provide adequate piping protection and to preclude inadvertent relief valve actuation during normal system operation. A relief valve capacity of 5 gpm was selected since rate.

it accommodates the maximum expected thermal expansion 14

PLANT CHANGE/MODIFICATION 98021 REACTOR COOLANT PUMP MECHANICAL SU SEAL REPLACEMENT WITH N-9000 SEAL

~Summa The previous reactor coolant pump (RCP) seals were multi-stage mechanical seal cartridges with two component stationary and rotating face assemblies and rotating springs. These seals had some undesirable responses to pressure and temperature transients, which limit seal life. Additionally, the potential for a seal to pop open, resulting in unmitigated. seal leakage, has been identified by the NRC as a serious concern, especially during station blackout conditions when make-up provisions are affected.

This modification replaced the existing seals with new N-9000 seals, which will provide increased service life and improved seal reliability during both normal and abnormal plant operations. The N-9000 seal design has been developed and tested by the original pump and seal manufacturer in response to the above NRC concerns. The new seals are designed to be completely interchangeable with the seals they replace. This modification also provided some modifications to the seal bleed-off lines to ensure that bleed-off flow can be isolated in the event that seal cooling is lost. Isolation of seal bleed-off within 30 minutes of a loss of seal cooling is required in order to minimize the thermal gradient across the seal and improve seal performance during the event.

15

PLANT CHANGE/MODIFICATION 98031 APPENDIX R CABLE REROUTES In preparation for the 1998 NRC Fire Protection Functional Inspection (FPFI), an original FPL licensing position (previously accepted by the NRC) regarding the credibility of multiple cable hot shorts resulting in spurious equipment energization was determined to be unacceptable based on recent NRC audits and correspondence. As such, a review of the existing safe shutdown analysis was performed and several potential multiple cable hot short situations were identified.

This modification provided for the rerouting of cables associated with shutdown cooling system and reactor head gas vent system valves. The affected cables are located within the reactor auxiliary building. The rerouted cables, generally in dedicated conduits, will assure compliance with Appendix R requirements by eliminating the potential for fire induced cable failures that could result in high/low pressure interface concerns. The new cable routing has no affect on the original functions or design basis of the affected valves.

16

PLANT CHANGE/MODIFICATION 98033 REPLACEMENT OF ISOLATION VALVES FOR CONTAINMENT PRESSURE TRANSMITTERS The UFSAR review project identified a design weakness associated with the normally open, fail closed, solenoid operated isolation valves for the safety related containment pressure instruments.

Specifically, with one of the four channels of containment pressure in bypass, a single electrical failure could result in the isolation of two of the remaining three channels, thus potentially preventing the fulfillment of a safety function since containment pressure is used as an input to a reactor trip and several engineered safeguards functions. This condition was documented in plant Condition Report 98-0512 and in LER 98-003.

This modification replaced the subject valves with normally open, fail open valves such that a power loss would not prevent This the associated instruments from sensing containment pressure.

change was determined to be consistent with the provisions of Regulatory Guide 1.11 (related to instrument lines penetrating containment).

17

PLANT CHANGE/MODIFICATION 98034 ULTRA-FINE MEDIA FOR CVCS PURIFICATION FILTERS 2A AND 2B Studies have indicated that a large percentage of radioactivity within the shutdown system can be attributed to particles of 1 micron or less. Out of core dose rates and radiation levels increase as a result of the introduction of particulates into the reactor coolant and chemical volume & control systems. This modification provided for the use of various sized filter media in one or both of the chemical volume and control system (CVCS) purification filters.

The original plant design called for a 98% (weight) retention of synthetic crud particles 2 microns and larger. Unit 2 has previously used 1.0 micron filters in accordance with safety evaluation JPN-PSL-SEMS-96-016. This modification allowed the use of higher efficiency filters to remove particulates down to approximately 0.1 microns. The use of ultra-fine filter media, an industry trend, has no adverse impact on system 'operation, although filter replacement frequency may be affected. Filter differential pressure is monitored and annunciated such that filter replacement can be initiated as appropriate.

18

PLANT CHANGE/MODIFICATION 98035 FHB VENTILATION SYSTEM PDS-25-17A & B SETPOINT CHANGE

~summa In the event of a fuel handling accident, the shield building ventilation system (SBVS) provides emergency ventilation for the fuel handling building (FHB). Make-up dampers FCV-25-11(12)" open to provide an outside air intake that supplements the flow from the inleakage to assure that total flow through the SBVS filter train is maintained at the required rate. Damper D-23(24) j.s also throttled to maintain flow through the filter train.

Differential pressure switches PDS-25-17A & B contain two separate contacts; one contact opens FCV-25-11(12) when the FHB pressure is -1 in. wg. The second contact allows D-23 (24) to be throttled when the FHB pressure is -2 in. wg. Testing-1 has demonstrated that FHB pressure can not be maintained below in.

wg. when FCV-25-11(12) is open and that FCV-25-11(12) oscillates open and closed as the pressure oscillates around -1 in. wg. The pressure never reaches -2 in. wg.; therefore, D-23(24) never throttles the flow to 'he required value. The pressure requirement for the FHB is -0.125 in. wg.

This modification increases the set points of PDS-25-17 A & B to values'hat are above the operating pressure and below the required pressure in the FHB. This will allow the system to operate in accordance with the design basis by establishing the required negative pressure and throttling flow to the SBVS filter at the design value. The set points are changed as follows: the HI set point (opens FCV-25-22 (12)) is changed from -1 to -0.25 in. wg; and the LOW set point (throttles D-23(24)) is changed from -2 to -0.5 in. wg. The new set points will open FCV 11(12) and throttle D-23(24) after the required negative pressure has been established in the FHB. This change has no effect on SBVS operation in response to a LOCA event.

Note Prior to this modification, testing of the new set points was successfully performed per 2-LOI-SCE-22 (see 10 CFR 50.59 safety eval'uation PSL-ENG-SENS-98-030).

19

PLANT CHANGE/MODIFICATION 98038 PRESSURIZER HEATER REPLACEMENT

~summa Eight pressurizer heaters were identified as requiring replacement during the Cycle 11 refueling outage. As a result of heater failures, jumpers have been installed as documented in PC/Ms 96168 and 95223. This PC/M restores the original pressurizer heater wiring configuration, provides for the abandonment of a stuck heater in the event it becomes necessary during maintenance, and incorporates the conclusions of safety evaluation JPN-PSL-SEMP-95-004, Rev. 4 into the pressurizer vendor manual (evaluation permits plant operation with limited reduced heater capacity).

20

PLANT CHANGE/MODIFICATION 9804 6 TELEPHONE SYSTEM UPGRADE

~Summa The plant communication system provides both intra-plant and offsite communications during normal and emergency conditions.

The telephone system is one component of the plant communication system, which includes the plant paging system, sound powered phone system, plant radios and the emergency notification system (ENS). Replacement of the System 85 AT&T plant telephone system with the NORTEL Meridian system was recommended by the FPL telecommunications group in order to standardize 'orporate telecommunication equipment throughout the FPL system and for reasons associated with Year 2000 compatibility.

This modification does the following: replaces the System 85 AT&T PBX system in the north service building (NSB) with a new NORTEL Meridian PBX telephone system in the south service building (SSB); replaces a System 85 AT&T remote module with a NORTEL Meridian remote module in building D-13; installs a new 120/208 V power panel in the new NSB telecommunication room to provide power to the plant telephone system equipment through a series power line conditioner; reroutes power supply cables and conduit; removes an existing power panel from the NSB; provides for powering of a backup A/C system for the new NSB telecommunications room; and deletes the UFSAR discussion of the HF/ALE communication system, which has been discontinued by the NRC. The telephone system is considered non-safety; however, the NSB remote module and the backup A/C equipment is automatically loaded onto the emergency diesel generator (EDG). As such, this modification included an analysis of its effects on EDG loading.

, 21

PLANT CHANGE/MODIFICATION 98130 APPENDIX R SAFE SHUTDOWN ANALYSIS Condition Report 97-2288 identified the need to validate the Unit 2 Appendix R Safe Shutdown Analysis (SSA). The validation of the SSA and associated evaluations and calculations are documented in engineering evaluation PSL-ENG-SEMS-98-067. This PC/M documents the completion of the Unit 2 Appendix R SSA validation effort as indicated in PSL-ENG-SEMS-98-067. The purpose of this PC/M is to provide the documents necessary to as-build the validated SSA, Essential Equipment List and the on-line Cable and Raceway System, as well as to provide the 10 CFR 50.59 evaluation necessary to support plant procedure changes to incorporate the manual actions specified in the referenced engineering evaluation.

22

SECTION 2 SAFETY EVALUATIONS 23

0 SAFETY EVALUATION JPN-PSL-SEMS-90-052 REVISION 1 UNITS 1 6 2 GENERIC USE OF SEALANT INJECTION This evaluation was provided as a vehicle to allow temporary repairs of gasket,and packing leaks on safety systems and quality related systems (such as feedwater) or components with the use of sealant injection. The affected components/systems will be replaced or repaired in accordance with time constraints delineated in the applicable Condition Report dispositions or other Engineering outputs. Three basic types of repairs were evaluated, including: 1) capnuts or sealing nuts with or without wire wrap,') drill and tap with gland plugs and shutoff adapters, and 3) slotted studs and capnuts with or without wire wrap. Regulatory and industry issues concerning the use of seal injection to repair leaks as identified by NRC IN 97-74, INFO SER 5-97, and NRC Part 9900 Technical Guidance were addressed in the evaluation.

24

SAFETY EVALUATION JPN-PSL-SENP-94-029 REVISIONS 1 & 2 SHUTDOWN OPERATIONS CRITERIA FOR REDUCED INVENTORY AND DRAINING THE REACTOR COOLANT SYSTEM The purpose of this evaluation was to discuss the plant conditions required to ensure the acceptability of shutdown operations given the following conditions:

A) The criterion for reduced inventory is defined as 3 feet below the reactor vessel flange.

B)The criteria for draining the RCS after shutdown is limited by the time to core uncovering and by the available RCS hot side vent path area.

These conditions, combined with the use of updated decay heat generation rates appropriate for current operating cycle lengths, provide a basis for shutdown operations with reduced inventory or at mid-loop conditions without degrading plant safety margins.

Implementation of the above conditions amends previous submittals to the NRC on shutdown operations; such changes are allowed under 10 CFR 50.59 as outlined in NRC correspondence on this subject.

25

0 SAFETY EVALUATION JPN-PSL-SENP-94-043 REVISION 1 TEMPORARY REMOVAL OF THE ICW PUMP MISSILE SHIELD

~Summa This safety evaluation demonstrates the acceptability of removing a section of ICW pump missile shield roof and/or ventilation fan missile shield "removable hood" section(s) over a single out of service pump, during power operations to allow maintenance on the pump. The ICW pumps are contained within a steel reinforced concrete missile barrier with a roof composed of 1-3/4 inch thick steel sections. Two ventilation fans serving the intake structure are located on the roof. Each fan's missile shield has a top "removable hood" section that extends outward from the support framing. Because of this overhang it may be necessary to remove this top section from one or both fans in order to provide access to an out of service ICW pump.

With a roof section and/or ventilation section(s) removed, the operating pumps are not fully protected against missiles.

However, this risk from missiles is negligible for the short period of time the shielding is not in place. In addition, missile shielding will be, re-installed in the event of a hurricane watch or warning when the risk of damage from missiles is the greatest.

26

SAFE TY EVALUATION JPN-PSL-SEMP-95-004 REVISION 4 REDUCED PRESSURIZER HEATER CAPACITY This evaluation documents the acceptability of removing up to a total of six pressurizer heaters from service if needed; one proportional heater and five backup heaters for a total of 300 kW. This would leave a heater capacity of 1200 kW. The capability to safely shutdown the plant under natural circulation conditions is not adversely affected since the backup heaters required to provide pressure control under a loss of offsite power will be maintained at or above the technical specification limit of 150 kW per heater bank.

Revision 4 of the evaluation increases the number of pressurizer heaters, which may be out of service from six to nine. This revision also evaluates operation with failed pressurizer heaters until the SL2-11 refueling outage.

27

SAFETY EVALUATION JPN-PSL-SEMS-96-052 REVISION 3 ISOLATION OF CHARGING PUMP RECIRCULATION LINES

~SUIMIIR This safety evaluation addresses the acceptability of manual isolation of the Unit 2 charging pump recirculation lines during all modes of operation until the refueling outage in the second quarter of 1997. The deviation from normal operating alignments is desired to preclude a loss of charging should the discharge check valve of a charging pump fail open when the pump, is stopped. Such an event occurred previously in 1996 when the 2C charging pump was secured for maintenance. Manual isolation of the recirculation line is considered to be a pro-active measure to ensure charging system reliability.

Revision 3 extended the applicability of the evaluation to August 2, 1997 since corrective actions for the charging pump check valve issue were not complete at the time.

28

SAFETY EVALUATION JPN-PSL-SECP-96-059 REVISION 3 UNITS 1 & 2 ST. LUCIE COUNTY SOUTH HUTCHINSON ISLAND WASTEWATER TREATMENT FACILITY AND WASTEWATER COLLECTION/RECLAIMED WATER DISTRIBUTION SYSTEMS To provide treatment for the increasing quantity of domestic wastewater generated on Hutchinson Island, St. Lucie County has constructed a new Wastewater Treatment Plant (WWTP) on Hutchinson Island. This facility is located approximately 2 miles south of the power plant. During periods of high flow and/or rainy weather, excess reclaimed water will be discharged through an outfall to the St. Lucie Plant discharge canal.

Three piping lines run from the WWTP to the'St. Lucie Plant an 8-inch force main for wastewater collection, an 8-inch reclaimed water line, and an 18-inch reclaimed water outfall line. The pipes are installed under the intake and discharge canals via subaqueous crossings, utilizing directional boring.

This evaluation addresses the activities related to the construction of the wastewater collection/reclaimed water systems to be performed on FPL p'roperty (subaqueous crossing of pipes under the intake and discharge canals, and construction of the reclaimed water outfall line along the discharge canal dikes).

This evaluation also addresses the discharge of excess reclaimed water at an outfall point at the discharge canal (including consideration of PSL plant operations, quality of discharged water, and environmental concerns). Consideration of the storage of hazardous materials at the WWTP, and transport of these materials to the WWTP, were also addresse'd in this evaluation.

29

~

SAFETY EVALUATION PSL-ENG-SEMS-97-013 REVIS ION 1 IN-SITU HYDROSTATIC TESTING OF STEAM GENERATOR TUBE FLAWS

~Summa In-situ hydrostatic test of flawed steam generator (S/G) tubing is performed when necessary to demonstrate that adequate structural margin exists. This evaluation develops and justifies the test parameters and demonstrates that the testing can be performed in accordance with the requirements of 10 CFR 50.59.

Outer Diameter Stress Corrosion Cracking is the prevalent form of degradation in Combustion Engineering S/Gs. Currently, when eddy current testing confirms axial or circumferential indications, the affected tubes are plugged. Only mechanical wear type degradation is sized and repaired in accordance with the technical specification limit of 40% through-wall.

One approach to verify adequate tube integrity is to demonstrate via pressure testing that degraded tubes (with axial or circumferential indications) can sustain the pressure requirements of Reg. Guide 1.121 "Basis for Plugging Degraded PWR Steam Generator Tubes," without bursting. FPL may demonstrate the adequacy of tube integrity by testing tubes with pluggable flaws to the pressure requirements of =the Reg. Guide. A secondary objective of in-situ pressure testing is to acquire leak rate data at normal operating differential pressure and main steam line break differential pressures:

Revision 1 of this evaluation removes the correction factor added to the test pressures to justify a longer run time between cycles 9 and 10, and updates the evaluation for conditions applicable to cycle 11 operation. Since cycle 11 operation time is scheduled to be less than cycle 10, no pressure correction is needed. A change in the plant restrictions was also made to account for in-situ testing being performed with fuel in the reactor vessel as opposed to being defueled.

30

I SAFETY EVALUATION PSL-ENG-SEMS-97-017 REVISION 0 TEMPERATURE CONTROL VALVE FOR THE HYDROGEN COOLERS (TCV-13-15)

MANUAL OPERATION

~Summa This evaluation demonstrates the acceptability of operating with the Turbine Cooling Water System (TCWS) Temperature Control Valve for the Hydrogen Coolers (TCV-13-15) in manual mode. The function of this valve and the TCWS is not safety related. The valve maintains the cold gas temperature .within the hydrogen coolers at an acceptable temperature. Placing the valve in manual mode allows hydrogen cold gas temperature in the turbine generator to trend with the injection temperature profile without operator actions. Per discussions with the turbine vendor and the FPL turbine specialist, long term operation with the valve in manual will not increase the rate of machine degradation as long as the hydrogen cold gas temperature is maintained within the recommended operating band. Temporary operation (8 months) with TCV-13-15 bypassed and flow manually controlled via valve SB13300 has demonstrated that operating the system in manual is acceptable.

31

SAFETY EVALUATION PSL-ENG-SENS-97-038 REVISION 1 PRESSURIZER CODE SAFETY VALVE MODIFICATION S~ary This evaluation clarifies that although the pressurizer code safety valves installed via PC/M 96139 were specified to have a blowdown of 4%, this value is considered a maximum value and the valves were tested and accepted with actual blowdown values of less than 4%. A 4% blowdown is acceptable since: (1) it meets the requirements of Section III, Class 1, 1974 edition of the ASME Code; (2) it is consistent with the current accident analysis ground rules; and (3) it is within the requirements of the Combustion Engineering Project Specification (i.e., less than 15%) .

Plant Change/Modification (PC/M) 96139, Pressurizer Code Safety Valve Replacement Forged Body Design, provided for the replacement of the pressurizer code safety valves. This evaluation confirms that the change provided by the PC/M did not result in an Unreviewed Safety Question and has no impact on plant .safety or operation.

32

SAFETY EVALUATION PSL-ENG-SENS-97-041 REVISION 0 UNITS 1 & 2 PARTIAL STROKE TESTING OF SIT DISCHARGE VALVES This safety evaluation demonstrates the acceptability of performing a partial stroke test of the safety injection tank (SIT) discharge check valves. Check valve full-stroke, partial flow testing has previously been performed at St. Lucie Units 1 and 2 and has been 'emonstrated

. successful in addressing NRC requirements for SIT check valve testing as delineated in Generic Letter 89-04. Check valve partial stroke tests are performed following valve maintenance activities.

The check valve partial stroke test can be performed in any Mode such that the SITs are not required and/or meet the minimum SITs operable requirement per Technical'Specifications; however, fuel movement shall be suspended during the discharge of the SITs. A check valve partial stroke test can be aligned to discharge to any of the following:

1. SIT to refueling cavity

2. SIT to pressurizer

3. SIT to refueling water tank (RWT)

These tests may be performed utilizing nitrogen pressure in the SIT as a motive force to discharge a portion of the SIT inventory through the discharge check valves, or with the SIT vented depending on the method selected for the test and SIT operability requirements. The test is initiated by opening the SIT discharge line motor operated valve and/or controlling discharge flow by throttling the "SI loop check valve leakage" valve depending on the selected method. Satisfactory performance of the test is by observing a 1% or greater level decrease in the tested SIT. As a SIT drains, the level and pressure are closely monitored to ensure Technical Specification operating limits are not exceeded.

Safety evaluations and calculations have previously been performed to demonstrate and address safety of 'full-stroke check valve testing. Check valve partial stroke tests are bounded by the conclusions addressed in full stroke valve tests.

33

SAFETY EVALUATION PSL-ENG-SENS-97-044 REVISIONS 0 & 1 CONTAINMENT SUMP SCREENS The containment sump screens act as a barrier to prevent debris from entering the emergency core cooling system (ECCS) and containment spray system (CSS). Per the UFSAR and the original NRC Safety Evaluation Report, these screens were designed in accordance with NRC Regulatory Guide 1.82, Revision 0. This evaluation documents the design and licensing requirements for the sump screens and clarifies the design bases, including a discussion on the acceptability of gaps in the screen.

'I St. Lucie Condition Report 99-1766 identified a number of tolerance issues with the containment sump screens wherein certain mesh openings were larger than the 0.0081 square inch

.value provided in the UFSAR. As a result, this evaluation was revised to provide the basis for why a eoolable core geometry is assured in a post-accident, re-circulation flow environment, even with slightly larger openings in the containment sump screen mesh. This evaluation also revises the UFSAR to state that the required mesh size of 0.0081 square inches (90 mils on a side) is a nominal value, to allow for tolerances.

h 34

SAFETY EVALUATION JPN-PSL-SEMS-97-055 e REVISION 0 POST ACCIDENT SAMPLING SYSTEM FSAR RECONCILIATION This Evaluation provides justification for UFSAR reconciliation of the post-accident sampling system (PASS). Since its initial installation, changes were made to this system and associated procedures. The changes were made based on regulatory requirements; however, the UFSAR was not updated to reflect the configuration changes. This evaluation incorporates the UFSAR changes to reflect plant configuration and procedures implemented for the PASS for various changes. The evaluation identifies the current design operational requirements and evaluates the system's conformance to these requirements. The changes to the system as described in 'the UFSAR include:

. a) removal of the requirement for measurement of the dissolved oxygen and pH in the liquid sample; and b) the removal of component details and values such as flows from the UFSAR when they are not regulatory requirements.

35

SAFETY EVALUATION PSL-ENG-SEMS-97-066 REVISIONS 0 & 1 UNITS 1 & 2 ADDITION OF HYDROGEN PEROXIDE TO THE RCS DURING SDC

~SUIMIIR This evaluation provides justification to add hydrogen peroxide to the RCS during shutdown cooling (SDC) system operations to oxygenate the RCS and facilitate crud burst/removal for both units. In order to comply with RCS chemistry requirements in the UFSAR and Technical Specifications, the peroxide shall be added with an RCS average temperature of less than or equal to 200'F and a hydrogen concentration of less than 5 cc/Kg.

36

SAFETY EVALUATION PSL-ENG-SECS-97-069 REVISION 0 UNITS 1 & 2 SITE SANITARY SYSTEM MODIFICATIONS

~Summa The St. Lucie Sanitary System has been identified as a system requiring periodic modifications and maintenance. A review of system functions, requirements and interactions concluded that this system 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 potential interactions with equipment important to safety. As such, this evaluation allows any modifications to this system to be performed outside of the formal Plant/Change Modification process. Drawings of the system will be maintained by Engineering.

The site sanitary system is a Not Nuclear Safety system that processes sanitary waste from the north side of the plant site, including the toilet and shower facilities located within the power block. The system consists of a sanitary treatment facility powered from plant 480-volt motor control center MCC-1B4.

37

SAFETY EVALUATION PSL-ENG-SEES-97-077 REVISION 0 BYPASS OF THE "REGULINER" VOLTAGE REGULATOR CIRCUI T IN THE NON-SAFETY-RELATED VITAL SUPS INVERTER The purpose of this safety evaluation is to document that an unreviewed safety question does not exist for Temporary System Alteration (TSA) ¹2-97-015. TSA ,¹2-97-015 temporarily bypasses the "Reguliner" voltage regulator circuit used in the alternate source feed to the nonessential (non-Class 1E / non safety-related) vital SUPS inverter.

The non safety-related vital SUPS alternate power source was lost on 9/19/97 due to a fire at one of the two voltage regulating transformers used in the "Reguliner" portion of the non-safety vital SUPS. Replacement of the burned transformer and the adjacent wiring with the inverter energized and Unit 2 operating on-line presents an unacceptable risk both to personnel safety and to a unit trip. In addition, the "Reguliner" components are no longer manufactured and a modification would be required in order to replace the entire alternate source regulating circuit or specific portions of the "Reguliner" circuit. The alternate power source to the non-safety vital SUPS wil'1 be restored without the voltage regulating portion of the circuit (Reguliner) in service to r'educe the probability of a unit trip should the inverter fail.

The non-safety vital SUPS, which will have its "Reguliner" bypassed under TSA ¹2-97015, powers non-safety plant instrumentation, control and communication circuits. This equipment does not perform a nuclear safety function.

38

SAFETY EVALUATION PSL-ENG-SECS-97-078 REVISIONS 0 & 1 UNITS 1 & 2 TEMPORARY LEAD SHIELDING INSTALLATION CRITERIA AND RESTRICTIONS This safety evaluation was developed to address the criteria and restrictions for the installation of temporary radiation shielding for: (1) systems and components which must remain operable/in-service; or (2) systems and equipment which are not required to be operable/in-service, but could pose a seismic Class II over I hazard for systems important to safety which are required to be operable/in-service. This temporary shielding may be required to implement ALMA requirements or meet NRC radiation exposure criteria limitations. Lead shielding may be in the form of lead bricks, lead blankets, lead sheeting, lead wool or other materials including steel and concrete blocks. These barriers can be supported from permanent or temporary plant structures, components, or be applied directly to piping systems.

The intent of the generic criteria and restrictions established in this evaluation are to ensure that the installation and use of temporary plant shielding will have no impact on plant operations or nuclear safety.

39

SAFETY EVALUATION PSL-ENG-SEMS-97-079 REVISION 0 UNITS 1 & 2 INSTALLATION OF A SECONDARY SA HEADER AND TIE-IN TO THE SA SYSTEM IN THE TGB S~ary This safety evaluation supports Temporary System Alteration (TSA)

N2-97-016 for installation of: (1) a branch connection and isolation valve in the Unit 1 to Unit 2 station air (SA) cross-tie line; (2) piping headers to various portions of the Unit 1 to Unit 2 turbine generator buildings (TGBs); (3) a tie-in to the construction air system header located in the Unit 2 condenser pit; and (4) tie in to the Unit 1 TGB ring header. Installation of the above piping and components will provide adequate breathing air during plant outages and provide the necessary air in the TGB for operation of pneumatic tools and equipment used for plant maintenance. In addition, the system will provide a back-up source of air during normal plant operation. The station air system is classified as Not Nuclear Safety.

40

SAFETY EVALUATION PSL-ENG-SENS-97-081 REVISION 1 OPERATION OF THE MSIV BYPASS VALVES DURING PLANT WARM-UP Main steam isolation valve (MSIV) bypass valves MV-08-1A a 1B are normally closed, motor-operated valves that are opened to warm-up and pressurize downstream piping in preparation for power operations. Condition Reports were issued to document problems of these valves in meeting their established design required closing thrust with flow under the seat. Because of these problems, the valves were considered inoperable when in the open position since the valves'bility to perform their safety function (automatically close upon receipt of a main steam isolation signal) was indeterminate. 'As such, the referenced Condition Reports required the valves to be maintained in the closed position for the duration of Cycle 10 in order to ensure valve operability.

This evaluation justifies the acceptability of opening either bypass valve for plant warm-up and pressurization, provided certain procedural measures are established to accommodate the ability to close against reverse flow. This is because the valves cannot be assured of automatically closing against steam flow (flow under the seat). In addition, the IST program testing frequency for these valves is being adjusted to address this operational restriction.

Note The corrective action to modify the valves to meet the design basis requirements was subsequently implemented via PC/M 98014.

41

SAFETY EVALUATION PSL-ENG-SEMS-97-097 REVISION 0 FIRE PROTECTION SYSTEM DOCUMENTATION UPDATE This evaluation addresses the fire detection instrumentation discrepancies between UFSAR Table 9.5A-13 "Fire Detection Instruments", Administrative Procedure 1800022 "Fire Protection Plan" and the installed plant condition. Based on the review performed by Engineering, no field changes were necessary; however, update of UFSAR, procedures and design documentation was found to,be warranted. This evaluation provided justification to update the following documentation:

UFSAR Table 9.5A-13 to show the number of detectors required and the number of detectors for each zone with a 50%

operability criterion applied to each detector type.

2. Update Appendix A of Administrative Procedure 1800022 to show the number of detectors required along with the total number of detectors for each zone with a 50% operability criteria.

3. Update the data sheets for I&C Procedure 1800051 "Semi-Annual Testing of the Smoke Detectors" to show Reactor Auxiliary Building West elevation at -0.5 feet for zones 1A

& 1B; RAB East elevation at -0.5 ft for zones 2A & 2B; and add zone 3A detector numbers 24 & 25 to the "Detector Numbers to be Tested" block.

42

.0 SAFETY EVALUATION PSL-ENG-SENS-97-098 REVISION 0 UNITS 1 & 2 USE OF THE FIRE MAIN TO SUPPLY CIRCULATING WATER PUMP BACKUP LUBE WATER AND THE CONDENSER TUBE CLEANING SYSTEM Because maintenance is periodically required on the service water system it becomes necessary to remove the system from service.

Although not required for nuclear safety, the service water system provides water to the circulating water (CW) pumps'ackup lube water system and to the condenser tube cleaning system (CTCS) for the mechanical seals of the ball retrieval pumps and the debris filter housing seals. Since these non-safety uses of service water are important to plant operation, interruption of

, service water flow is undesirable.

This evaluation allows the use of the fire main system to provide water for the specific use of the CW pumps'ackup lube water system and for the CTCS during periods of service water system unavailability., Temporary supply manifolds shall be connected to local fire hydrants via wye connectors and fire hose. During this time the fire main, which is normally pressurized via the domestic water pumps and associated h'ydropneumatic tank, will be pressurized via a temporary pump. The temporary pump suction and discharge connections shall be connected portable fire pump connections provided in system piping. Use of a temporary pump in this manner is consi'stent with system design.

This evaluation is limited to application on either Unit 1 or Unit 2 at any given time. The maximum water demand placed on the fire main will be approximately 54 gpm. The threat of this proposed water demand to impact the dedicated water volume for fire mitigation has been evaluated and was determined not to be credible due to an automatic makeup feature and city water storage tank level alarm set points.

43

SAFETY EVALUATION PSL-ENG-SENS-97-101 REVISIONS 0 6 1 ALTERNATE METHODS FOR MONITORING REACTOR VESSEL INVENTORY The Technical Specifications require that if the number channels of the reactor vessel level monitoring system (RVLMS) is of less than the minimum number required, then an alternate method of monitoring the reactor vessel inventory must be initiated or the plant must be shut down. This evaluation provides an alternate method of monitoring reactor vessel inventory in the event that all channels of the RVLMS become inoperable. The alternate method involves using other plant parameters to provide a method of determining key points in the reactor vessel inventory to support operations decisions within the emergency operating procedures should an event occur.

The alternate method consists of: 1) monitoring and comparing charging and letdown flows, pressurizer level and level behavior;

2) core exit thermocouple temperatures, 3) reactor coolant system hot leg temperatures, and; 4) reactor coolant system sub-cooling.

SAFETY EVALUATION PSL<<ENG-SEIS-97-108 REVIS ION 0 UNITS 1 & 2 REACTOR PROTECTION SYSTEM LOGIC MATRIX TEST STATUS LIGHTS The design of the reactor protection system (RPS) includes built-in test features and status indication lights, which facilitate periodic surveillance testing of the actuation logic circuits.

The lights are also used to provide the status of each stage of the actuation logic when surveillance testing is not being performed on the system. The UFSAR (both units) notes that "Proper operation of all coils and contacts is verified by lights on a trip status panel."

At various times .in the past, status lights have become inoperable for short periods of time. Typically, the status lights are repaired on an expedited basis and returned to service before the grace period expires on the required RPS surveillance; however, in some cases the repair effort may extend beyond the allowed grace period for the RPS surveillance. As such, this evaluation provides the option of performing the surveillance with the use of a digital voltmeter in lieu of an inoperative status light.

SAFETY EVALUATION PSL-ENG-SEMS-98-005 REVISION 0 FSAR DISCREPANCIES ON ECCS PUMP NPSH

~summa .

This evaluation addresses and resolves discrepancies identified between the UFSAR, Design Basis Documents and design calculations related to emergency core cooling system (ECCS) pump available net positive suction head (NPSHa) values. The corrected values have no significant affect on analysis assumptions.

46

SAFETY EVALUATION PSL-ENG-SEFJ-98-005 REVISIONS 0 & 1 REEVALUATION OF ST. LUCIE UNIT 2 CONTROL ELEMENT ASSEMBLIES LIFETIME BASED ON TWO CEA INSPECTION RESULTS UFSAR Section 4.2 implies that the operational lifetime of a control element assembly (CEA) i's ten years. The vendor (ABB/CE) clarified that this limit ien calendar years, which is approximately eight effective full power years. This limit is based on a different CEA design than that currently employed at St. Lucie Unit 2. The current CEA operational lifetimes were reported in a previous evaluation report and were determined by evaluating the CEA in-service performance data taken from the 1992 inspection.

This evaluation reevaluates the CEA operational lifetime using the 1992 and 1997 CEA inspection results. The new evaluation results have concluded that:

a) all full strength CEAs have an operational lifetime of 12 cycles of operation; and b) all reduced strength (lead bank) CEAs have an operational lifetime of 10 cycles of operation.

47

SAFETY EVALUATION PSL-ENG-SENS-98-007 REVISION 0 IMPLEMENTATION OF ICW/CCW OPERATIONAL TEMPERATURE CURVES

~SUIMIIR This safety evaluation was issued to evaluate the implementation of intake cooling water (ICW)/component cooling water (CCW) operational temperature curves which provide ICW temperature limits based on CCW heat exchanger fouling.

The UFSAR and the Technical Specification Bases provide a maximum value of 95'F for the ultimate heat sink temperature. The ICW/CCW system was designed for the specific ability to remove post accident decay heat with only a single train in service.

The 95 F value for the ultimate heat sink provided an upper temperature limit in order to size the system (i.e., procurement of heat exchangers, such that sufficient cooling capacity would exist given typical in-service conditions without compromising any safety margin). Operational temperature curves incorporated into plant procedures provide assurance that sufficient cooling capacity exists and that system performance remains within accident analyses.

48

SAFETY EVALUATION PSL-ENG-SENS-98-011 REVISIONS 0 & 1 UNITS 1 & 2 ONSITE STORAGE OF RADIOACTIVE MATERIALS IN A REMOTE RCA

~Summa The purpose of this evaluation is to identify the conditions under which a separate radiation controlled area (RCA) for the storage of contaminated materials may be developed within the FPL owner controlled area of the St. Lucie site, but, situated non-contiguous with the existing RCA that encompasses the Unit 1 and 2 reactor containment, auxiliary and fuel handling buildings.

Certain byproduct materials (as defined by 10 CFR 20.1003) typically used during refueling outages are currently stored in the existing radiation controlled areas. It is desired to retain this material for future use; however, while not in use, storage of this material requires sufficient space such that its presence is judged to interfere with the normal operational and maintenance functions occurring inside the existing RCA. The remote RCA will be used to store equipment necessary for outage work onsite. Its location and design ensure that dose to the members of the public remains within acceptable limits.

49

SAFETY EVALUATION PSL-ENG-SEMS-98-012 REVISION 0 UNITS 1 G 2 ADDITION OF DIMETHYLAMINE TO THE SECONDARY SYSTEM This evaluation addresses the addition of dimethyilamine (DMA) to the secondary system during all modes of operation for both units. DMA will aid in reducing corrosion product build-up in the steam generators and it has also been shown to remove corrosion product buildup from the feedwater flow venturies and assist in the removal of sludge from the steam generators at other operating nuclear plants. The addition of DMA will be in conjunction with the hydrazine and ammonia chemistry program, not as a replacement. DMA has been evaluated and endorsed for on-

• line use in all ferrous plants by the Combustion Engineering Owners Group.

50

SAFETY EVALUATION PSL-ENG-SEFJ-98-013 REVISION 0 AMENDMENT TO CORE OPERATING LIMITS REPORT This safety evaluation provides justification to change Core Operating Limits Report (COLR) Figure 3.2-3 of the UFSAR. This figure provides the limits for the combinations of Thermal Power with F 'nd F,',/ respectively. ~ In this figure, the the Fvalue F~ tradeoff at each curve is proposed to be changed to make T power level equal to the corresponding F, value multiplied by 1.0294. This adjustment addresses an identified discrepancy in the setpoint analysis calculations related to the F~ tradeoff curve. The proposed change will result in an increase in the available margin to the DNB, while maintaining consistency between the limits of the revised Figure 3.2-3 and the existing setpoint analysis.

The license amendment for implementing COLR submitted to the NRC was already approved by the NRC. The COLR defines cycle-specific parameter limits for certain Technical Specifications. These limits are developed using NRC approved methodologies, and are consistent with all applicable limits of the safety analysis.

NRC Generic Letter 88-16 is the guidance document for relocating cycle-specific limits for certain TS parameters into COLR. Per GL 88-16, changes to COLR can be implemented under the provisions of 10 CFR 50.59 when the modified limits are generated using previously NRC approved methods listed in the TS. The proposed change is, therefore, in accordance with GL 88-16.

51

SAFETY EVALUATION PSL-ENG-SEMS-98-017 REVISION 0 UNITS 1 & 2 BIOCIDE TREATMENT OF CLOSED COOLING WATER SYSTEMS This safety evaluation addresses the addition of biocides on an "as needed basis" as determined by sampling and analysis of the various closed cooling water systems. Microbiological contamination of closed cooling water systems has been identified as a problem by many of U.S. nuclear utilities. Testing at St.

Lucie and Turkey Point plants has identified the presence of microbiological organisms in several of the closed cooling water systems. At St. Lucie, the following systems are included in this evaluation: component cooling water, turbine cooling water and steam generator blowdown cooling wat:er. Some utilities have experienced equipment degradation and/or failures as a result of material corrosion that was caused by microbiological activity.

Biocide treatment is a standard practice in other industries.

Two non-oxidizing biocides were evaluated: Glutaraldehyde and Isothiazolin. They have both been evaluated and successfully used at other nuclear sites. Both of these biocides produce a rapid kill of both aerobic and anaerobic microbes and each has a short effective active response time, typically less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, after which they are no longer present in the system. Two biocides are required to develop an effective biological contaminant control program. A single biocide can prevent significant reduction in biological contaminant; however, some bacteria develop a resistance to a single biocide and a second biocide becomes necessary to eliminate the resistant bacteria.

The effectiveness of the biocide selected is also dependent on the ability of the biocide to contact the microbiological organism. As such, this evaluation also addresses the use of a biodispersant to ensure the effectiveness of the biocide is maximized.

52

SAFETY EVALUATION PSL-ENG-SEIS-98-020 REV.ISION 0 PROTECTION SYSTEM TEST CONFIGURATION WITH A POSTULATED SINGLE FAILURE CONCERN OF THE CONTAINMENT PRESSURE TRANSMITTER INPUT SIGNALS Condition Report (CR) 98-0512 documented a single failure concern with the four safety-related containment pressure transmitters.

Each transmitter has a dedicated containment penetration with a normally energized, fail closed solenoid valve in its process sensing line. Per the CR, a single power failure can be postulated that will cause a loss of two redundant containment pressure measurement channels. Loss of two channels could prevent actuation of the reactor protective system (RPS) and engineered safety features system (ESFAS) when in a two-out-of-three configuration (i.e., with a channel placed in bypass). As a result, the CR disposition prohibited the placing of a channel in bypass.

As described in the UFSAR, the bypass feature is normally used for required surveillance testing of the affected channels.

Since the CR prohibited use of the bypass feature, this evaluation was necessary to document the acceptability of performing required monthly surveillance testing without placing the affected channel in bypass.

Note plant modification/change (PC/M) 98033 was subsequently implemented to restore the original plant design basis.

53

SAFETY EVALUATION PSL-ENG-SEIS-98-021 REVISION 0 UNIT 1 & 2: RSPT SIGNAL RESTORER EVALUATION

~Summa Two non-Class 1E systems of control element assembly (CEA) position indication are provided for display of CEA position on the main control panel. The systems are the pulse counting CEA position indication system and the reed switch CEA position indication system. The reed switch CEA position indication system obtains its signals from a reed switch position transmitter (RSPT) which consists of a network of resistors connected in series. As the CEA moves up and down, a magnet moves with it causing the reed switches to change state and the resistance value to change proportionately. With voltage applied to the RSPT, it acts as a stepwise potentiometer that provides an output voltage proportional to CEA position.

An open circuit or high resistance of the RSPT resistor bank or a stuck reed switch will cause the RSP indication to be inoperable for the applicable CEA, which requires a Limiting Condition for Operation in accordance with the Technical Specifications. A RSPT signal restorer device is available, which electronically compensates for these two failure mechanisms. This safety evaluation documents the acceptability of utilizing the RSPT signal restorer to reinstate partial or full operability of a failed RSPT and to establish plant restrictions with device installed.

54

SAFETY EVALUATION PSL-ENG-SEMS-98-024 REVISION 0 UNITS 1 G 2 CONTROL ROOM PAINTING The plant has requested various upgrades to the control rooms and the simulator. One of these upgrades involves repainting of the control room panels. This evaluation addresses the impact of the painting activities and the final installed paint on plant operation and nuclear safety. Relevant restrictions are identified to ensure that the criteria in the UFSAR and Technical Specifications are met.

'J SAFETY EVALUATION PSL-ENG-SES J-98-026 REVISION 0 ANALYSIS OF CAPSULE 263 FROM THE REACTOR VESSEL IRRADIATION SURVEILLANCE PROGRAM On May 4, 1997, FPL removed surveillance capsule 263'rom the reactor vessel. The capsule contained material specimens representative of the reactor vessel limiting plate and non-limiting intermediate to lower shell girth weld. These specimens were tested in accordance with 10 CFR 50 Appendix H and ASTM E-185-82. The results indicate that the limiting plate material measured 30 ft-lb transition temperature shift was greater than predicted by NRC Regulatory Guide 1.99, Rev. 2. The weld material and standard reference material plate transition temperature shift was less than predicted by the Reg. Guide.

Results of the capsule test and analysis were documented in a Westinghouse report.

This safety evaluation was prepared using RG 1.99 guidance to assess the conclusions of the Westinghouse report and the potential impact on the existing pressure temperature limit curves and LTOP setpoints. The P/T limit curves and LTOP setpoints remain valid for the 15 EFPY period of applicability based on use of surveillance data.

The evaluation concluded that surveillance capsule data from the Westinghouse report is credible by comparing it to the results of previous capsule, which are also summarized in the report. The capsule 263'eport also meets the reporting criteria in 10 CFR 50 Appendix H and ASTM E185-82.

56

SAFETY EVALUATION PSL-ENG-SENS-98-027 REVISION 0 UNITS 1 6 2 SHARED SYSTEMS AND INTERCONNECTIONS BETWEEN UNIT 1 AND UNIT 2

~Summa Several systems, structures and components are interconnected and shared between both units. According to NRC General Design Criterion 5, structures, systems and components important to safety shall not be shared among the nuclear units unless such sharing will not significantly impair their ability to perform their . safety functions. An on going UFSAR review effort determined that discussion about shared and interconnected systems for both units is inaccurate and inconsistent.

The current Unit 1 UFSAR description states that all interconnections have locked closed valves to isolate each unit.

Also, the Unit 1 UFSAR lists several systems and components that are interconnected but not normally shared. However, some of the interconnected systems do not have locked closed valves and are frequently or continuously used by both units during plant operations (e.g., hydrogen, nitrogen and make-up demineralizer rege'neration systems).

For Unit 2, the UFSAR states that locked closed isolation valves are provided for the liquid waste management (hold-up tanks and aerated waste storage tank AWST), and station service air interconnections. However, no locks are provided in the plant for the AWST interconnection and only one valve of the service air system is locked closed. This evaluation resolves these discrepancies and provides the necessary UFSAR changes. In addition, the write-ups for both units'FSARs are being revised to provide consistent formats and content.

57

SAFETY EVALUATION PSL-ENG-SENS-98-028 REVISION 0 UNITS 1 & 2 DUMMY FUEL ASSEMBLY USE UFSAR CHANGES An on going UFSAR review identified a discrepancy related to the use of dummy fuel assemblies to check the alignment of spent fuel storage racks and fuel handling systems. Current plant procedures do not check the alignment with dummy fuel assemblies as specified in the UFSAR. Instead, funnels are provided to guide the fuel assemblies into the spent fuel pool rack locations.

Safety evaluation JPN-PSL-SENS-94-025 was issued to provide UFSAR corrections for Unit 1 since dummy fuel assemblies were only used to test the fuel handling systems during the manufacturing and post-installation periods. Subsequently, the Unit 2 UFSAR was revised with safety evaluation JPN-PSL-SENS-95-021, Rev. 1, for similar inconsistencies related to the use of dummy assemblies during periodic testing of the spent fuel storage racks and fuel handling equipment.

This safety evaluation justifies and provides the necessary Unit 1 and Unit 2 UFSAR change packages to address the alignment checks of spent fuel pool storage racks. In addition, the Unit 1 text is being written in a similar manner to Unit 2 to add a discussion about the use of a test weight for fuel handling crane testing prior to lifting the fuel assemblies.

An editorial change is also addressed in both UFSAR change packages related to liquid penetrant testing, which was a one-time-test performed during the pre-operational testing period for both units.

58

SAFETY EVALUATION PSL-ENG-SENS-98-030 REVISION 0 IMPLEMETATION OF AN EXPERIMENT TO DETERMINE NEW OPERATIONAL SETTINGS FOR FHB VENTILATION

~Summa Conditi'on Report 98-0262 documented excessive cycling of the shield building ventilation system (SBVS) outside air intake dampers when aligned to the fuel handling building (FHB). The Condition Report determined that settings controlling operation of the outside air intake dampers were not optimal for the SBVS when aligned to the FHB, resulting in excessive cycling of these dampers. The SBVS was demonstrated to be operable in the interim, except for the function of evacuating filtered air from the FHB. In accordance with Technical Specification 3.6.6.1.c(2), the movement of irradiated fuel and crane operation with loads over irradiated fuel was suspended, pending final resolution of the damper cycling problem. This safety evaluation addressed conducting a test/experiment to determine the correct settings (for operation in the FHB mode) to be used as a basis for a permanent modification.

Note Plant Change/Modification 98035 was implemented to install new controlling set points for system dampers.

59

SAFETY EVALUATION PSL-ENG-SENS-98-038 REVISIONS 0 & 1 TEST OF P-10 & P-11 CONTAINMENT PENETRATIONS WITH OPEN SHIELD BUILDING DOOR ACCESS PORT

~SIIINIIR I

Testing of containment penetration nos. 10 & 11, containment purge exhaust and supply, is performed periodically during power operations. This testing requires access to the 40 ft and 60 ft elevations in the annulus between the reactor containment building and the shield building. In the past, testing personnel have taken the required e'quipment into the annulus prior to beginning surveillance testing and have remained in the annulus during testing with the shield building access door closed, thus preserving shield building integrity as defined by Technical Specifications. Typically, testing of this penetration requires approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, including setup and dismantling.

Review of site testing practices with respect to MB~ concerns has concluded that the presence of personnel in the annulus for this testing during power operation is undesirable. The primary containment consists of a relatively thin welded steel vessel.

The substantially greater thickness of the shield building and its material'f construction provide significant attenuation of the ex-core gamma and fast neutron flux during power operations.

As a result, from an AId~ standpoint it is desirable to minimize personnel stay time in the annulus during power operations.

The desire to minimize stay time, combined with the requirements of the Technical Specifications concerning shield building integrity and operability of the shield building ventilation system, have led to a proposed modification of the testing protocol. This evaluation provides the justification for: 1) keeping test personnel outside the shield building during test;

2) keeping the shield building access doors closed during the test; and 3) routing test wires and cabling through a temporary, small-diameter hole in the shield building access door.

60

SAFETY EVALUATION PSL-ENG-SENS-98-051 REVISION 0 TEMPORARY INSTALLATION OF A DIAGNOSTIC SUBROUTINE FOR DIGITAL DATA PROCESSING SYSTEM (DDPS)

This evaluation addresses the temporary installation of a software diagnostic subroutine to the digital data processing system (DDPS). The subroutine will be used to facilitate identification of the source of a "high" incore alarm that has been randomly occurring. The high alarms were short duration, singular, and from random detectors. The subroutine will be in place for a limited period in order to identify and record more detailed information about the source of the alarm. The source information will be used to determine if corrective actions are needed for the hardware or software related to the incore detectors, communication wiring, or DDPS system. Once enough data is collected to identify the source of the random alarm, the diagnostic subroutine will be removed and the system returned to its original design configuration.

61

SAFETY EVALUATION PSL-ENG-SENS-98-056 REVISION 0 UNITS 1 & 2 UFSAR UPDATE FOR GRID STABILITY ANALYSIS

~StttmR The UFSAR Review Project identified the present grid stability analyses contained in the Unit 1 & 2 USFARs are out of date. It was also determined that NRC Information Notices (INs) suggested that licensees should periodically review their grid analysis.

The latest IN on the subject states that NRC Standard Review Plan (SRP) Section 8.2 provides current guidance for assessing the adequacy of the offsite power system; therefore, FPL's Transmission System Planning group was requested to update this analysis.

The SRP, Section 8.2 III.l.f, requires that grid stability analysis show the preferred source of power is not completely lost as a result of the outage of a single element which disconnects: a) the grids largest source of power, b) the grids largest load, or c) most critical transmission circuit. An updated dynamic stability analysis has been performed and the results of that analysis are being incorporated into both UFSARs via this safety evaluation.

Additionally, all references to 240 kV system are being revised to 230 kV. FPL's 230 kV system was referred to as 240 kV prior to 1988. The nominal design voltage of these facilities has always been 230 kV, which is a standard transmission voltage level in North America. FPL now refers to these transmission facilities as 230 kV to avoid confusion on the line ratings, which are based on amperes. Also, Chapter 8 of Unit 2 UFSAR refers to outdated statistics for FPL transmission line mileage.

This is also being updated per input from Transmission System Planning.

62

SAFETY EVALUATION PSL-ENG-SENS-98-059 REVIS ION 0 SPENT RESIN TANK DEWATERING OPERATIONS UFSAR CHANGES A Condition Report identified discrepancies between spent resin dewatering operations described in plant procedures and those described in the UFSAR. Specifically, three discrepancies were identified:

1. The UFSAR states that water which results from sluicing of spent resin into the spent resin tank (SRT) is directed to the holdup tanks for processing. However, according to plant procedures, draining and dewatering of the SRT is directed to floor drains or a safeguards system sump.

2. The UFSAR states that spent resins are expected to be sluiced to an outside shipping container and dewatered via the resin dewatering pump. However, according to plant procedures, a portable dewatering system is used instead of the resin dewatering pump.

3. The UFSAR states that liquid in the holdup tanks is used as flushing water during resin sluicing of the SRT and pre-concentrator ion exchangers. However, according to plant procedure, sluicing operations are performed using primary makeup water instead.

This evaluation justifies and provides the necessary UFSAR change package to correct the above discrepancies and make the UFSAR consistent with actual plant operation.

63

SAFETY EVALUATION PSL-ENG-SENS-98-062 REVISION 0 1998 FSAR REVIEW FINDINGS REQUIRING CHANGES OR CLARIFICATIONS TO THE FSARs IN ACCORDANCE WITH 10 CFR 50. 59 The purpose of this safety evaluation is to update, correct, or add clarifications to the UFSAR following a review of selected systems. The subject matter of these changes involves miscellaneous changes to UFSAR details associated with the following systems and components: containment isolation, safety injection tank valves and piping, Quality Group A components, NSSS active valves, emergency diesel generator loading, auxiliary feedwater system components, safety related annunciator panels and circuits, and engineered safety features actuation system (ESFAS) sensors. Each change has been individually evaluated and justified.

64

SAFETY EVALUATION PSL-ENG-SECS-98-064 REVISION 0 UNITS 1 6 2 OPERATION OF SPENT FUEL CASK CRANE "RESTRICTED ZONE" ADJACENT TO FUEL HANDLING BUILDING Administrative Procedure 0010438 restricts the spent fuel cask crane from entering an area adjacent to the fuel handling building (both units). This "restricted zone" is located on the east and, north sides of the buildings. The cask cranes are restricted from traveling into the "restricted zone" unless both hooks are in the full up position or it travels south through the two-foot access located at the center of the L-shaped door.

Currently, crane limit switches are set to prevent movement of the cranes in accordance with the above restriction.

Maintenance has requested permission to allow movement and use of the spent fuel cask cranes within the "restricted zone" in order to remove material, casks, and equipment that may be located within the "restricted zone." This evaluation justifies the temporary alteration of the Unit 1 and 2 spent fuel cask crane limit switches to allow movement of either crane within the "restricted zone" while the "L-shaped" door is closed. The changes allowed by this evaluation are consistent with the requirements of NUREG-0612.

65

SAFETY EVALUATION PSL-ENG-SENS-98-065 REVISION 0 RADIATION PROTECTION UFSAR CHANGES The 1996 UFSAR/procedure consistency review process identified several minor discrepancies between UFSAR Chapter 12, Radiation Protection, and current plant procedures, practices'nd configuration. These discrepancies include:

1. out of date description of the responsibilities of the Health Physics (HP) department;

2. incorrect frequency of radiation surveys performed by HP;

3. incorrect description of the HP office as one location for radiation monitoring system operators'onsole;

4. computer storage in hard drives rather than nine track magnetic tape;

5. instrument calibration performed per Tech Specs rather than just annually; and

6. discontinued use of hand and foot radiation monitors at the containment and fuel handling building exit locations.

Additionally, Condition Report (CR) 98-1222 documents .chronic operational and maintenance problems associated with the mobile airborne radiation monitors and associated equipment, resulting in reduced availability of the PC-11 radiation monitoring system.

Per the CR, the mobile monitors have not been used since Unit 2 startup. This evaluation justifies the removal of the mobile radiation monitors from the UFSAR. Airborne radioactivity monitoring is performed in accordance with HP procedures.

This safety evaluation provides justification for the UFSAR changes necessary to resolve the above discrepancies.

Acceptability of these changes is predominantly based on the continued ability of the radiation protection program to meet 10 CFR part 20 requirements.

66

SAFETY EVALUATION PSL-ENG-SENS-98-066 REVISION 0 INSTRUMENT RANGE AND ACCURACY IN UFSAR Twenty-six different tables in the UFSAR contain instrument ranges and accuracies. These values were originally included in the FSAR during the licensing process to demonstrate that appropriate instrumentation was being used. There appears to be no specific requirement to maintain these values in the UFSAR.

Furthermore, having these values listed in the UFSAR causes an additional burden in the plant change process by forcing 10 CFR 50.59 evaluations for insignificant instrument accuracy and range changes. This evaluation has been written to replace all range and accuracy values from the UFSAR with a note identifying the general design requirement.

A review of various licensing documents determined that requests in those documents for range and accuracy details is specific for the Preliminary Safety Analysis Report stage. In fact, Appendix A of R.G. 1.70 specifically states, "this appendix describes safety-related interfaces ... that should be presented at the reliminar desi n sta e," (emphasis added). Appendix 7.1-C of the "Guidance for Evaluation of Conformance to IEEE Std 603" states "tables in Sections 7.2 and 7.3 should provide accuracies and ranges..." (emphasis added). The identification of specific instrument ranges and accuracies is considered to be a detail necessary only for the plant licensing stage and is not considered a necessary detail to be maintained in the current UFSAR.

New instruments or modified instrument parameters are provided in accordance with Nuclear Engineering Quality Instructions (QIs).

As such, whenever an instrument change is made, instrument ranges and accuracies are properly considered and selected. Instrument ranges are selected in accordance with standard engineering practices. Likewise, instrument accuracies are selected such that existing instrument loop performance and safety analysis assumptions remain valid. Where applicable, instrument accuracies are also evaluated for their impact on setpoints in accordance with the FPL Setpoint Methodology. For post-accident monitoring instrumentation, the instrument range and accuracies required by Regulatory Guide 1.97 are considered.

67

SAFETY EVALUATION PSL-ENG-SENS-98-068 REVISION 0 LIQUID WASTE MANAGEMENT SYSTEM UFSAR CHANGES

~Summa The 1996 UFSAR/procedure consistency review effort identified a number of discrepancies related to liquid waste management system operations described in the UFSAR. Specifically, the operation of the boric acid concentrators, waste concentrator .and associated components is not performed as described in the UFSAR.

This equipment and associated components have not been used for many years, and alternative waste processing methods have been implemented. It is the intent of this safety evaluation to analyze and evaluate these methods for regulatory compliance and to revise the UFSAR accordingly.

68

SAFETY EVALUATION PSL-ENG-SENS-98-072 REVISION 0 WARMING STARTING AND OPERATING THE SDC SYSTEM Shutdown Cooling (SDC) operation is initiated at a point in the plant cooldown when the reactor coolant system (RCS) conditions drop below SDC design conditions. The SDC system removes RCS decay and sensible heat until the RCS reaches a nominal temperature of 140'F. The method of aligning the SDC system for cooldown operation is described in the UFSAR.

Condition Report (CR) 96-2840 identified that the method of SDC initiation described in the Unit 1 UFSAR disagreed with the method employed by the Unit 1 operating procedure. While addressing this issue it was discovered that, based on historical plant data, vendor requirements for the SDC heat exchangers had been exceeded. This resulted in the initiation of CR 98-0643.

The disposition of this CR resulted in actions to revise the operating procedures for Units 1 and 2, specifically, to monitor heatup rates for the SDC heat exchangers. Further review by Engineering determined that operating procedures for both units required significant changes to preclude unacceptable thermal transients to the SDC heat exchangers.

This evaluation justifies and provides a revised procedural sequence for Unit 2 SDC system initiation. In addition, the descxiption of actions required for SDC initiation in the Unit 2 UFSAR was revised to agree with procedure changes incorporated as part of this evaluation.

69

SAFETY EVALUATION PSL-ENG-SENS-98-073 REVISION 0 UNITS 1 6 2 USE OF SUPPLEMENTAL GAGES TO MONITOR THE CCW HEAT EXCHANGER INLET STRAINER DIFFERENTIAL PRESSURE

~Summa This safety evaluation was issued to examine the use of additional gages to supplement the monitoring of the intake cooling water (ICW) inlet strainer differential pressure to the Unit 1 and 2 component cooling water (CCW) heat exchangers.

Additionally, pressure gages will be installed to supplement the monitoring of ICW differential pressure across the inlet strainer for the turbine cooling water heat exchanger. These additional gages will be installed on the upstream side of the existing differential pressure gages. Use of supplementary pressure gages with a greater range will allow for additional monitoring capability of the pressure drop between the pump discharge and the inlet to the heat exchanger. The instrumentation application for strainer differential pressure is described in UFSAR Table 9.2-3 for each unit. This configuration does not alter the instrumentation described in the UFSAR.

70

SAFETY EVALUATION PSL-ENG-SEMS-98-074 REVISION 0 OPERATION AND TESTING OF THE AFW PUMP SUCTION CROSS-TIE BETWEEN UNIT 1 AND UNIT 2 AT THE UNIT 2 CST

~SUIMIIR As part of the original licensing commitment for St. Lucie Unit 1, a cross-tie was established between the Unit 2 condensate storage tank (CST) and the Unit 1 CST. This cross-tie and dedicated water supply in the Unit 2 CST were provided for use in the unlikely event that a vertical tornado missile disabled the Unit 1 CST.

During an NRC audit it was identified that, while this was installed, it had never been tested to establish cross-tie its flow capability. Subsequently, testing of the cross-tie piping was performed; however, the piping from the 2A/2B and 2C pump suction headers to the auxiliary feedwater (AFW) cross-tie was not tested. This, evaluation addresses the testing of these two sections of pipe and their isolation valves.

The evaluation examines the test configuration in'hich the cross-tie isolation valves for the 2A/2B and 2C pump suction headers are opened and portions of the cross-tie piping are used to verify AFW flow capability. Considerations are given to both Unit 1 and Unit 2 operating conditions and requirements during the test.

71

SAFETY EVALUATION PSL-ENG-SENS-98-076 REVISION 0 UFSAR CHANGES TO FUEL POOL SYSTEM PERFORMANCE MONITORING DESCRIPTIONS The 1996 UFSAR/procedure consistency review identified a discrepancy between the UFSAR and existing plant procedures regarding the periodic taking of data to confirm heat transfer capabilities, purification efficiency and component differential pressures for the fuel pool system. Formal data collection and analysis, as alluded to in the UFSAR, is not required by plant procedures; however, existing practices, including the use of plant instrumentation and alarms, are sufficient to ensure proper system operation and performance. This evaluation justifies changing the UFSAR to be consistent with plant practices.

72

SAFETY EVALUATION PSL-ENG-SENS-98-083 REVISION 0 SOLID WASTE MANAGEMENT SYSTEM UFSAR CHANGES The 1996 UFSAR/procedure consistency review identified several discrepancies related to equipment in the solid waste management system no longer available or installed at 'he plant. Also, solid waste handling procedures described in the UFSAR are not performed as described, and other procedures and methods have taken their place.

This safety evaluation justifies and provides the necessary UFSAR changes to make the UFSAR consistent with the actual plant operation. These changes include:

1. deletion of discussions about the use of a portable spent resin solidification unit, which is not available at the plant;

2. deletion of discussion about the dry waste baler compactor, which is no longer used at the plant;

3. addition of discussion about the use of portable spent resin dewatering systems (also addressed in PSL-ENG-SENS-98-059);

and

4. labeling solid radwaste generation tables as "estimated" values developed prior to initial plant startup.

73

SAFETY EVALUATION PSL-ENG-SENS-98-085 REVISION 0 UNITS 1 6 2 FLASH TANK OPERATION UFSAR CHANGES The 1996 UFSAR/procedure consistency review identified a discrepancy related to the current operation of the flash tank.

According to the findings, the flash tank is bypassed to the holdup tanks during normal plant operations. This is contrary to the UFSAR description, which describes the normal use of the flash tank. This issue is applicable to both units.

This evaluation justifies the current operating practice of bypassing the flash tank and provides the necessary Unit 1 6 2 UFSAR changes. Also, the option for operating with the flash tank will be kept in the UFSAR for plant operation when reactor coolant activity levels exceed a predetermined value, or whenever the hydrogen or fission gas stripping function of the tank is required.

74

SAFETY EVALUATION PSL-ENG-SEMS-98-087 REVISION 0 INCREASED SDC PURIFICATION FLOW

~Summa Purification of reactor coolant system (RCS) inventory during outages has a direct impact on RCS activity and, consequently, on personnel dose and outage duration. Purificat:ion of RCS fluid is normally performed by t: he chemical volume and control system (CVCS) letdown path when the RCS is pressurized. During periods of shutdown cooling (SDC) system operation with the RCS pressurized, primary system purification is performed by aligning the SDC system to the CVCS purification loop. The flow is processed by the CVCS ion exchangers and then returned to the suction of the LPSI pumps. It has been identified that increasing the maximum letdown flow during SDC operation would significantly benefit the time required to achieve the desire outage activity levels and reduce overall personnel dose. This evaluation justifies increasing SDC purification flow by up to 17% to a nominal value of 150 gpm.

75

SAFETY EVALUATION PSL-ENG-SENS-98-089 REVISION 0 INTAKE COOLING WATER STRAINER UFSAR CHANGES

~SUltltllR The 1996 UFSAR/procedure consistency review process identified a discrepancy with respect to backwashing of the intake cooling water (ICW) strainers. According to the discrepancy, the UFSAR states the ICW strainers are operated under a continuous backwash, whereas plant procedures do not normally align the backwash feature. Furthermore, the strainers are designed for manual backwashing and are not capable of continuous backwash.

The ICW system is a safety related system that provides the heat sink for the component cooling water (CCW) system. The subject strainers we located upstream of the ICW/CCW heat exchangers and are designed as manual back flush type single element basket strainers. The differential pressure across the strainers is monitored and an alarm is provided in the control room upon a high differential pressure (DP). Per plant procedures, the strainer(s) are manually backwashed upon high DP. This evaluation justifies elimination of the continuous backwash description in the UFSAR.

SAFETY EVALUATION PSL-ENG-SENS-98-090 REVISION 0 UNITS 1 6 2 USE OF PRIMARY SYSTEM IN-LINE SAMPLE ANALYSIS FOR DISSOLVED GASES

~SUIMIIR This safety evaluation justifies the use of a primary system in-line sample analyzer for monitoring dissolved gases in the reactor coolant system (RCS) during normal operations, plant shutdown, and plant startup. Oxygen is monitored to minimize system corrosion and eliminate the possibility of creating an explosive gas mixture. Hydrogen is monitored to prevent explosive gas mixtures, ensure the radiolytic reaction of any oxygen that may enter the system, and maintain a reducing environment in the RCS. Currently, these gases are analyzed by grab sample analysis, as described in both the Unit 1 and 2 UFSARs.

Use of the in-line analyzer for monitoring dissolved gases in the RCS does not alter the configuration of the sampling system.

The in-line analyzer utilizes connections normally used for grab sample analysis. As such, the Unit 1 and 2 UFSARs will be updated to describe the new in-line analyzer as an acceptable means for monitoring dissolved gases in the RCS in addition to gases analyzed by grab sample analysis.

77

SAFETY EVALUATION PSL-ENG-SENS-98-091 REVISION 0 UNITS 1 6 2 USE OF CVCS PURIFICATION/DEBORATING ION EXCHANGERS UFSAR CHANGES The 1996 UFSAR/procedure consistency review identified four discrepancies related to the current operation of purification and deboration ion exchangers for both units. According to the noted discrepancies, the UFSAR states that each of the chemical and volume control system (CVCS) ion exchangers are used for a specific purpose; however, per current plant operating procedures, all three CVCS ion exchangers are interchangeable.

Also, the UFSAR states that one ion exchanger (the deborating) is used to reduce boron concentration at the end of cycle (EOC),

when in fact current plant operations allow the use of any one or two ion exchangers for deborating activities. This evaluation provides the necessary Unit 1 & 2 UFSAR Change Packages to address these discrepancies and to justify current operating practice.

78

SAFETY EVALUATION PSL-ENG-SEIS-98-093 REVISION 0 INSTALLATION OF JUMPERS ACROSS VARIOUS ISOLATION CABINET CIRCUITS DURING MODES 5 AND 6 The plant design includes isolation cabinets which provide electrical separation between safety related and non-safety circuits per Reg. Guide 1.75. Preventive maintenance of system power supplies necessitated de-energizing the isolation cabinets.

This work was to be done during modes 5, 6 or refueling. Since de-energization would cause a loss of various signals that are normally passed through these cabinets and since several of these signals are important to plant operation during the modes in which the work was to take place, this safety evaluation was written to justify the installation of jumpers in the isolation cabinets to support continued operation, during the maintenance activity, of the following circuits:

1. interlocks between the containment purge fans and the purge exhaust valves, also, the containment isolation signal (CIS) input signal to trip the containment purge fans;

2. the high temperature input signal to close the control room North outside air intake valve;

3. the fuel handling building (FHB) high radiation trip signal to the normal FHB ventilation fans;

4. high containment radiation input signals to the containment evacuation alarm; and

5. input signals to various non-safety annunciators.

79

SAFE TY EVALUATION PSL-ENG-SENS-98-094 REVISION,O LETDOWN RADIATION MONITOR UFSAR CHANGES

~SUIBHIR The 1996 UFSAR/procedure consistency review identified discrepancies between the UFSAR and plant procedures with respect to the letdown radiation monitor. Specifically, the UFSAR describes the function and use of the letdown radiation monitor, whereas plant operating procedures isolate this monitor, with no instructions to align it for use.

The letdown radiation monitor, also referred to as the chemical volume and control system (CVCS) process monitor, was originally installed to alert plant operators of an increase in reactor coolant system (RCS) radioactivity, primarily to indicate potential fuel failure(s). However, the monitor is not being used due to ALARA concerns associated with its installed location. In contrast to the same monitor in Unit 1, which is installed in a shielded area on the 19.5 foot elevation, the Unit 2 monitor is located on the 43 foot elevation and is unshielded.

The pipe routing to the Unit 2 monitor poses an ALARA problem since it contains radioactive fluid in an unshielded area. This monitor does not perform any functions related to safe plant shutdown or functions that are required to mitigate the consequences of an accident. The option of relocating the monitor to a more suitable location was not pursued since other means exist to monitor RCS activity. Therefore, this evaluation provides the justification to revise the UFSAR to acknowledge the isolation and discontinued use of the letdown radiation monitor.

80

SAFETY EVALUATION PSL-ENG-SEES-98-097 REVISION 0 TEMPORARY POWER CONNECTIONS TO P.A. SYSTEM PWR. DIST. CAB.

DURING MODES 5 AND 6 Plant design includes 120 volt vital ac buses 2A, 2B, 2A-1 and 2B-1. These buses provide low voltage electrical power for instrumentation and control for various services. Preventative maintenance activities associated with the 120 volt ac system required de-energizing the vital ac buses. This work was to be done during modes 5, 6 or with the reactor defueled.

Since de-energization would cause a loss of various power supplies that are normally fed from these buses, Engineering performed a review of the devices and annunciation to be lost during the evolution and concluded that temporary power connections were required to maintain operability of the plant P.A. system power distribution panel. This panel provides power to the containment evacuation alarm and input signals to various non-safety annunciators that have been determined to be important to plant operation during the maintenance activity. As such, this safety evaluation was written to justify the installation of temporary power connections to the P.A. system power distribution cabinet to support normal operation of the containment evacuation alarm and input signals to various non-safety annunciators.

81

SAFETY EVALUATION PSL-ENG-SENS-98-098 REVISIONS Og 1 & 2 UPPER GUIDE STRUCTURE LIFT RIG USE WITHOUT POSITIONING BUSHING The 0 guide bushing was damaged as a result of being set down on an upper guide structure (UGS) storage pad. The guide bushing at 180 is still intact and is providing guidance.

have a very close tolerance The bushings with the alignment pin installed in the reactor vessel flange. The such that it tilt of the damaged bushing was could jam on the alignment pin. It was decided to remove the 0 bushing to preclude potential jamming and additional lift rig damage. The bushing was attached to the, mounting plate by fillet welds. The welds were removed and the bushing discarded. The surface to which the bushing was attached has been ground smooth. Any shims that may have been between the bushing and mounting plate were also removed.

The lift rig is a tool used to remove and install the UGS into the reactor vessel, it does not perform a safety function. The guide bushing is not in the load path of the UGS attached to secondary structural members of the lift lift rig, it rig. There is was no gross damage to this secondary structure.

This evaluation and justifies reinstallation of the using the UGS, lift rig, including removal with the 0 guide bushing removed and with an alignment sleeve installed over the 0 reactor vessel alignment pin. An assessment of fuel assembly and CEA potential loads is also provided.

82

SAFETY EVALUATION PSL-ENG-SEIS-98-099 REVISIONS 0 & 1 INSTALLATION OF JUMPERS ACROSS SB/MB/MD ISOLATION CABINET CIRCUITS DURING MODES 5 AND 6 The plant design includes isolation cabinets which provide electrical separation between safety related and non-safety circuits per Reg. Guide 1.75. Preventive maintenance of system power supplies necessitated de-energizing the isolation cabinets.

This work was to be done during modes 5, 6 or refueling. Since de-energization would cause a loss of various signals that are normally passed through these cabinets and since several of these signals are important to plant operation during the modes in which the work was to take place, this safety evaluation was written to justify the installation of jumpers in the isolation cabinets to support continued operation, during the maintenance activity, of: (1) the high temperature input signal to the control room North outside air intake valve, and (2) the sump level input signal to the emergency core cooling system sump pump automatic start circuit.

SAFETY EVALUATION PSL-ENG-SEMS-98-101 REVISION 0 UNIT 2 STEAM GENERATOR SECONDARY SIDE FOREIGN OBJECTS This evaluation addresses the safety significance of operating the steam generators with foreign objects inside. The foreign object that has caused the largest through-wall defect has not been characterized, but a detailed history of the tubes in the region has been investigated. The foreign object is currently located at the top edge of eggcrate 03C adjacent to tube R16/L106 in steam generator 2B. This tube showed approximately a 20% through-wall defect in October 1990 with no significant change until the most recent outage. During that outage showed approximately a 32% through-wall defect with the bobbin coil and it 30$ through-wall with the plus point probe. Tube R18/L106 also showed a defect in 1990 (approximately 20% through-wall) with no significant change until 1997, when it was plugged following an inspection that showed the defect had grown to approximately 30% through-wall. No other adjacent tubes have shown defects at this elevation.

Since the foreign object is located at the top of eggcrate 03C, no visual examination can be performed and no access is available for removal. Further, bobbin coil and plus point eddy current inspections have not been successful in characterizing the foreign object. Thus, the foreign object is addressed by investigating its effect on those tubes that have been degraded and assessing its impact for long term operation. Since the subject tubes are now plugged but not staked, this evaluation also provides a basis for continued operation without staking.

This evaluation focuses on tubes R16/L106 and R18/L106 because of the more severe damage associated with them; however, the results are also applicable to foreign objects previously observed in the steam generators. None of these foreign objects caused wear damage to adjacent tubes except for the weld rod located between lines L129 and L130, which caused damage to tubes R114/L130 and R112/L130 in steam generator 2A. Loose part monitoring of these tubes, began in 1989 and, although there was no significant growth rate in the defects, a conservative decision was made to plug these tubes. At the time of plugging these tubes had through-wall defects that were less than 20%.

It is likely that the relatively soft material hardness of the weld rod

~

(as compared to the foreign object adjacent to tubes R16/L106 and R18/L106) has resulted in the low growth rate. Thus, since these tubes showed lower growth in the depth of the through~wall defect, the evaluation of tubes R16/L106 and R18/L106 bounds all of the other foreign objects examined in this evaluation.

SAFETY EVALUATION PSL-ENG-SENS-98-104 REVISION 0 "B" TRAIN ESF TESTING WITH THE "B" ICW HEADER OUT OF SERVICE This safety evaluation evaluates conducting the engineered safeguards features (ESF) testing for the "B" train in Mode 5 with reactor coolant system loops not filled and with the "B" intake cooling water (ICW) discharge piping out of service for code repair due to corrosion and through-wall leaks on the pump discharge piping immediately downstream of the pump discharge isolation valve.

In the evaluated configuration, the "C" ICW pump is aligned to the "A" discharge header and the "AB" electrical bus (which powers the "C" pump) is powered from the "B" bus. The "A" ICW loop is in operation, powered from its normal source. This electrical configuration is normal for ESF testing of the "B" train in accordance with the existing test procedure; however, the mechanical configuration is unusual since the "C" pump is aligned to the "A" header, along with the operable "A" pump.

85

0 SAFETY EVALUATION PSL-ENG-SEMS-98-119 REVISION 1 OPERATION WITH MOTOR OPERATED VALVE V3523 OPEN Condition Report 98-2096 identified a degraded condition of valve V3523, a containment isolation valve on the hot leg injection line of the high pressure safety injection (HPSI) system. The hot leg injection piping is used for post-accident long term cooling (per UFSAR Section 6.3.2.2.3). The HPSI pumps are manually re-aligned for simultaneous hot and cold leg injection.

This ensures flushing and ultimate sub-cooling of the core coolant independent of break location.

The change proposed by this safety evaluation is to lock open V3523 to its throttled position and use the second, redundant valve in the line, V3551, as both the containment isolation valve and the hot leg isolation valve. The change places V3523 in its "safe position" for alternate hot leg and cold leg injection. In addition, the safety function of V3523 for containment isolation and hot leg isolation is transferred to V3551, which will be de-energized to provide single failure protection under specific plant scenarios. Since the quality and level of protection of the original design basis have been maintained, this change has no effect on plant operation or safety.

The proposed change is an adequate measure until the next outage of sufficient duration, at which time V3523 will be repaired.

86

SAFETY EVALUATION PSL-ENG-SENS-99-004 REVISION 0 PROCESS SAMPLING SYSTEM UFSAR CHANGES The 1996 UFSAR review project identified a discrepancy related to UFSAR Section 9.3.2 for current operation of the process sampling system. According to the noted discrepancy, the UFSAR "has numerous conflicts with plant procedures." Since the UFSAR finding was not specific about the conflicts found, the Chemistry department was asked to review the UFSAR section and identify any discrepancies. The following discrepancies were identified:

the secondary sampling system's chemical analyzer panel sample drain line is not routed to the steam generator feedwater pump leak-off storage tank, but rather, to the storm drain system;

2. the volume of the primary sample system's sample vessels is 300 cc instead of 1000 cc as noted in the UFSAR; and

3. the primary sampling system local valve controls are not located in the sample room as indicated in the UFSAR.

This evaluation provides necessary justification to revise the UFSAR to reflect actual plant configuration. A valve tag number correction is also provided.

In summary, routing of the sample drain line to the storm drain system is acceptable since it is consistent with the routing of other secondary system drains. The use of a 300 cc sample vessel is acceptable since it is of sufficient size and capacity for system sampling. The current location of valve controls is acceptable since the associated UFSAR discussion is intended to discern between local and control room control of the valves.

87

SAFETY EVALUATION PSL-ENG-SENS-99-005 REVISION 0 UNITS 1 & 2 CHANGES TO REACTOR COOLANT AND REACTOR MAKEUP WATER CHEMISTRY SPECIFICATIONS

~SUIMIIR This safety evaluation addresses changes to Unit 1 UFSAR Table 9.3-8 and Unit 2 UFSAR Table 9.3-5, Reactor Coolant and Reactor Makeup Water Chemistry. The changes involve updating the current reactor coolant and reactor makeup water chemistry specifications to be consistent with current industry standards. The revision also reformats the tables so that the data are consistent between both units. The only difference in chemistry parameters listed for the two units is the fluoride limit. This reflects the difference in the fluoride limits listed in the Unit 1 and Unit 2 Technical Specifications.

The chemistry limits for the RCS parameters are being tightened within the cu'rrent limits or revised to reflect current Technical Specifications or current industry recommendations. These chemistry specifications provide guidance for maintaining the reactor coolant chemistry program. The identification of chemistry conditions that lead to long term NSSS integrity is a dynamic process. Some of the specifications and chemistry control practices given in the UFSAR are now obsolete and need to be revised to reflect current industry research and experience.

The Electric Power Research Institute and ABB Combustion Engineering have both issued revisions for primary chemistry control parameters. Typically, the revised limits are more restrictive than the original values and rely on consistency between the parameters, such as the relationship between boron/lithium, pH, and conductivity.

88

SECTION 3 RELOAD SAFETY EVALUATIONS 89

PLANT CHANGE/MODIFICATION 98016 ST. LUCIE UNIT 2 CYCLE 11 RELOAD

~SUIMIIR This engineering package provided the reload core design of the St. Lucie Unit 2 Cycle 11. The Cycle 11 core is designed for cycle lengths between 11,463 and 12,530 EFPH, depending upon variation in the Cycle 10 length of between 11, 900 and 12, 700 EFPH, respectively. The design of Cycle 10 included an end of cycle inlet temperature coastdown to 535'F followed by a coastdown in power to approximately 85% power. Cycle 10 was expected to reach an EOC exposure of approximately 12, 550 EFPH.

The primary design change to the core for Cycle 11 is the replacement of 64 irradiated fuel assemblies with fresh Region N fuel assemblies. The fuel is arranged in a low leakage pattern similar to the design of the Cycle 10 core. The mechanical design of Region N fuel is functionally the same as that of Region M (Cycle 10). The primary changes associated with Region N are the incorporation of the ABB-CE "Value-Added" fuel pellet and the Guardian grid designs. The new pellet design benefits fuel cycle economics and the Guardian grid provides enhanced debris resistance.

The safety analysis of this design was performed by Asea Brown Boveri Combustion Engineering Nuclear Operations (ABB CENO) using NRC approved methodology and was independently reviewed by Florida Power and Light Co.

90