ML17227A316

St Lucie Annual 10CFR50.59 Rept for 1991

ML17227A316
Person / Time
Site:	Saint Lucie
Issue date:	12/31/1991
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17227A317	List: ML17227A316 ML17227A319 ML17309A673
References
NUDOCS 9203030215
Download: ML17227A316 (48)

Text

SECTION ST~LUCIE ANNUAL 10 CFR 50'9 REPORT A summary of changes to the facility as described in the Final Safety Analysis Report (FSAR)(10 CFR 50.59 (A)(1)(i))is submitted by separate letters at the same time as the annual FSAR update for each unit (July 22 for St.Lucie Unit 1 and April 6 for St.Lucie Unit 2).Safety Evaluations (for 1991)that were approved by the Facility Review Group (FRG)and those associated with Jumper/Lifted Leads are attached.P2030302i5 920225 l PDR*DOCK 05000335 R PDR 10 CFR 50.59 Evaluations Summaries of Evaluations Approved by the St.Lucie Facility Reviev Group ST~LUCIE UNIT 2 SAFETY EVALUATION FOR AUXZL1ARY FEEDWATER PUMP 2C TURBINE COUPLING BEARING LUBE OIL PIPING TEMPORARY MODIFICATION ZNTRODUCTIONI This Safety Evaluation is prepared to document the acceptability of the temporary use of piping and fittings in place of the originally installed tubing on the 2C Auxiliary Feedwater Pump Turbine Coupling Bearing Lube Oil system.The original tubing was damaged during maintenance activities and exact replacements are not readily available.

The Auxiliary Feedwater Pump performs a safety related function and is designed as a Quality Group C component.

This evaluation concludes that the temporary modification describes herein does not represent an unreviewed safety question and has no impact on plant safety or operations.

A review of the Plant Technical Specifications and the FSAR has shown that there are no Technical Specification changes involved.This evaluation is valid through the end of the 1992 refueling outage.SAFETY EVALUATION:

The temporary modified configuration does not create a new path for uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function.The temporary modified piping configuration will not adversely impact the ability of the 2C Auxiliary Feedwater Pump (or any other equipment) to perform its accident mitigating design function and will not create any new failure modes for the 2C Auxiliary Feedwater Pump.The modified configuration will not inhibit or otherwise adversely affect the operation of any equipment important to safety.Therefore:

1)The probability of occurrence of an accident previously evaluated in the FSAR has not been increased.

2)The consequences of an accident previously evaluated in the FSAR have not been increased.

3)The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR has not been increased.

The temporarily modified piping configuration is equivalent to the original tubing configuration, and meets or exceeds the system design pressure, temperature, material, and flow characteristics and does not modify any active components.

4)The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR have not been increased.

ST~LUCIE UNIT 2 SAFETY EVALUATZON FOR AUXILIARY FEEDWATER PUMP 2C TURBINE COUPLING BEARING LUBE OIL PIPING TEMPORARY MODIFICATION PAGE 2SAFETY EVALUATION (Continued):

5)The possibility of an accident of a different type than any previously evaluated in the FSAR has not been created since the temporarily modified configuration does not add or affect any equipment capable of initiating an accident.6)The possibility of a malfunction of a different type than any evaluated previously in the FSAR has not been created.7)The existing configuration does not reduce the margin of safety as defined in the basis for any Technical Specification since the temporarily modified coupling bearing piping configuration will not impact the operation of the 2C Auxiliary Feedwater Pump as required per the Technical Specifications or the FSAR.The temporarily modified configuration is functionally equivalent to the original configuration, flow characteristics of the oil system are not changed.

ST LUCIE UNIT 1 SAFETY EVALUATION FOR CONTAINMENT FAN COOLER UNQUALIFIED COATINGS-REVISION 1 INTRODUCTION:

This safety evaluation addresses the presence of unqualified coatings on the 1A, 1B, 1C and 1D Containment Fan Cooler coil flanges.New cooling coils were installed under PC/M 081-189 during the Unit 1 1990 refueling outage.Upon inspection of the coils prior to installation, the coating on the coil flanges was determined to be improperly applied (i.e.-unqualified).

Some of these unqualified coatings could not be removed and replaced due to their proximity to the copper coils and the resultant potential for damage of the copper coils.The presence of the unqualified coatings inside containment will not affect Plant safety or operation since their is'o potential for failure of the coatings during a loss of coolant accident (LOCA)to adversely affect the operation of any structure, system, or component important to safety.A review of the Technical Specifications and the FSAR has shown that there are no unreviewed safety questions or Technical Specification changes involved.During the 1991 Unit 1 refueling outage, the Containment Fan Cooler coil flanges were inspected.

This revision incorporates the findings of the inspection.

A review of the plant Technical Specifications and the FSAR has shown that there are no unresolved safety questions or Technical Specifications changes involved in this revision.SAFETY EVALUATION:

The unqualified coatings do not perform a safety function and their failure during a LOCA will not adversely affect the function of any structure, system, or component important to safety, or affect any accident initiating events.The failed coatings cannot adversely affect the ECCS since they will not clog the containment sump, or affect the performance of ECCS pumps and containment spray nozzles.Degradation of the component cooling water system is not a concern since appreciable corrosion of the flanges due to the lack of a protective coating would be identified and addressed before a loss of function of the flange occurred.Therefore:

1)The probability of occurrence of an accident previously evaluated in the FSAR has not been increased.

2)The consequences of an accident previously evaluated in the FSAR have not been increased.

ST~LUCIE UNIT 1 SAFETY EVALUATION FOR CONTAINMENT FAN COOLER UNQUALIFIED COATINGS-REVISION 1 PAGE 2 SAFETY EVALUATION (Continued):

3)The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR has not been increased.

4)The possibility of a malfunction of equipment of a different type than previously evaluated in the FSAR has not been created.5)The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR have not been increased.

6)The possibility of an accident of a different type than any evaluated previously in the FSAR has not been created.7)The proposed modification does not reduce the margin of safety as defined in the basis for any Technical Specification since the unqualified coatings cannot affect the basis for any Technical Specification.

ST~LUCIE UNIT 1 SAFETY EVALUATION FOR MV-07-3B STEM REPLACEMENT REVISION INTRODUCTION:

The valve stem of Containment Spray B Header Isolation Valve MV-07-3B is damaged and must be replaced.The existing valve stem material is ASTM A473 type 316 stainless steel.An inspection of the proposed replacement item in stock (M&S 577-67526-7) identified the material to be ASTM A276 type 410 stainless steel.This Safety Evaluation is a revision and was issued to extend the duration of the evaluation until the 1993 refueling outage and revises the administrative portions of the evaluation to comply with current requirements.

The conclusions of the Safety Evaluation remain valid and have not changed as a result of this revision.This evaluation does not involve an unreviewed safety question nor does it require a change to the Technical Specifications.

SAFETY EVALUATION:

The containment spray system is designed to assist in the mitigation of a Loss of Coolant Accident (LOCA), assuming a single active or passive failure.Furthermore, the valve that is the subject of this evaluation is designed to Quality Group B and Seismic Class I requirements.

Based on the above description, this evaluation and associated modifications are considered Nuclear Safety Related.This report does not involve an unreviewed safety question based on the following conclusions:

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

The use of the 410 SS will not change the operational ability or the seismic qualification of the subject valve in that the corrosion experienced during one refueling cycle will not significantly degrade the integrity of the valve stem.2)The possibility for an accident or malfunction of a type than any previously evaluated in the FSAR is not created since the temporary modification does not change the operational design of the system.3)The margin of safety as defined in the bases for any Technical Specification is not affected by this change since the temporary modification does not compromise the valve integrity, reliability, or affect its operational characteristics.

ST~LUCIE UNIT 1 CVCS AND WASTE MANAGEMENT BORIC ACID HEAT TRACING CIRCUIT DE-ENERGIZATION INTRODUCTIONS This safety evaluation addresses the effect of de-energizing specific circuits of the heat tracing system associated with the Boric Acid Makeup System.Engineering packages PC/M 336-189 Revision 2,"Boric Acid Concentration Reduction Modifications", and PC/M 094-188 Revision 0,"Boric Acid Concentration Reduction" reduced the boric acid concentration in the Boric Acid Makeup System.Portions of the Boric Acid Makeup System with boric acid concentrations of 3.5 weight percent or less do not require heat tracing.This evaluation focuses in on the Chemical and Volume Control System (CVCS)and Waste Management System (WMS)(or Boric Acid Makeup and Recovery Systems)for the purpose of de-energizing heat trace circuits that have been identified by St.Lucie Plant Maintenance as not being required.Per FSAR section 9.3.4.1 portions of the Boric Acid Makeup System are designed and built to meet the requirements of seismic class I, hence this safety evaluation is classified as safety related.SAFETY EVALUATION Based on the St.Lucie Unit 1 FSAR, the Chemical and Volume Control System (CVCS)is designed to perform the following:

a)maintain the chemistry and purity of the reactor coolant within the limits specified in FSAR Table 9.3-8.b)maintain the required volume of water in the reactor coolant system by compensating for coolant contraction or expansion due to plant step load changes of (+/-)104 of full power and ramp changes of (+/-)54 of full power per minute between 15 and 1004 power and for reactor coolant losses or additions.

c)accept out-flow from the reactor coolant system when the reactor coolant is heated at the administrative rate of 75 degrees F/hr and to provide the required makeup when the reactor coolant is cooled at the administrative rate of 75 degrees f/hr using two charging pumps.d)accommodate the reactor coolant system water inventory change for a full-to-zero power decrease with no makeup system operation and with the volume control tank initially at the normal operating level band.e)inject concentrated boric acid into the reactor coolant system upon a safety injection actuation signal (SIAS).

ST~LUCZE UNIT 1 CVCS AND%ASTE MANAGEMENT BORIC ACID HEAT TRACING CIRCUIT DE-ENERGIMTZON PAGE 2 INTRODUCTION (Continued):

f)control the boron concentration in the reactor coolant system to obtain optimum control element assembly (CEA)positioning to compensate for reactivity changes associated with large changes in reactor coolant temperature, core burnup, and xenon concentration variations, and to provide shutdown margin for maintenance and refueling operations.

g)inject boron in sufficient quantity to counteract the maximum reactivity increase due to cooldown at 75 degrees/hr and xenon decay using one charging pump.h)automatically divert the letdown flow to the waste management system (HMS)when the volume control tank is at the highest permissible level.assure that the radioactivity due to corrosion and fission products in the reactor coolant system does not exceed Technical Specification limits for an assumed 1%failed fuel condition.

i)provide continuous on-line measurement of reactor coolant boron concentration and radioactivity due to fission and corrosion products.j)k)provide auxiliary pressurizer spray for operator control of the reactor coolant system pressure during the final stages of shutdown and to allow for the cooling of the pressurizer.

Based on a review of the above items, the effect of de-energizing heat trace circuits has been determined to have no impact on safety functions or regulatory requirements.

The probability of an accident previously evaluated in the FSAR has not been increased because de-energizing heat trace circuits do not affect the initiation of an accident evaluated in the FSAR nor increase the probability of occurrence.

Desired boric acid concentration is maintained even with the de-energizing of heat tracing circuits.The consequences of an accident previously evaluated in the FSAR are not increased by the de-energizing heat tracing circuits.As described above, de-energizing heat trace circuits will not change, degrade, or prevent system functions described in, or assumed to occur in the mitigation of any FSAR accident.

ST~LUCIE UNIT CVCS AND WASTE MANAGEMENT BORIC ACID HEAT TRACING CIRCUIT DE-ENERGIZATION PAGE 3 SAFETY EVALUATION (Continuect):

This proposed activity has no impact on the LOCA analysis and the radiological consequences of an accident evaluated in the FSAR will not be increased.

The probability of occurrence of a malfunction of equipment, important to safety previously evaluated in the FSAR has not been increased because the proposed activity will not result in new performance requirements being imposed on any system or components such that any design criteria will be exceeded.The Boric Acid Makeup System functional requirements are unchanged, therefore no new probability of malfunction has been imposed.As described above, de-energizing heat trace circuits do not change, degrade, or prevent actions described in, or assumed to occur in the mitigation of any FSAR accident.Therefore, de-energizing heat tracing circuits will not increase the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR.The de-energizing of heat trace circuits has been evaluated and does not impact the structural integrity or performance capability of CVCS and Waste Management System.The possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the FSAR is not increased by this proposed activity.The proposed activity to de-energize some heat trace circuits in the boric acid system at, St.Lucie Unit 1 does not introduce failure modes of a different type than any previously analyzed in the FSAR.The system configuration and the design basis of the boric acid system has not been changed or affected, therefore the proposed activity does not create the possibility that an accident may be created that is different from any already evaluated in the FSAR.Removal of heat tracing in areas containing boric acid concentration of 3.5 weight percent or less does not reduce any margins of safety for boration sources and flow path requirements since the concentration of boric acid is within the requirements of the Technical Specification.

De-energizing heat tracing does not affect the Technical Specification basis for borated water sources.

8T~LUCIE UNIT 1 8AFETY EVALUATION OF A BLIND FLANGE ON LINE I-3"-CW-160 8PENT FUEL POOL MAKEUP 8ALT WATER BACKUP INTRODUCTION This safety evaluation is prepared to document the acceptability of the installation of a blind flange on line I-3"-CW-160 at valve I-SB-21386.This line is the backup salt water supply for Spent Fuel Pool Makeup.The proposed configuration is acceptable on a temporary basis to support maintenance on line I-2 1/2"-CW-176.

This safety evaluation is required to allow the implementation of the necessary repairs within the constraints of the Limiting Conditions of Operation (LCO)for the ICW system.The ICW and Spent Fuel Pool Makeup systems are safety related and are designated as Seismic Class 1 and Quality Group C systems.This evaluation concludes that the proposed configuration described herein does not represent an unreviewed safety question and has no impact on plant safety or operations.

A review of the Plant Technical Specifications and the FSAR has shown that there are no Technical Specification changes involved.This evaluation is valid through the end of the 1991 refueling outage.SAFETY EVALUATION:

The probability of occurrence of an accident previously evaluated in the FSAR has not been increas'ed since the proposed configuration does not affect any accident initiating components.

The proposed configuration does not create any new failure modes for any equipment or systems capable of initiating an accident.The consequences of an accident previously evaluated in the FSAR have not been increased since the proposed configuration does not create a new path for uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function.The proposed configuration does not impact any equipment which is required to initiate actuation of any safety systems.The proposed configuration will not adversely impact the ability of the Spent Fuel Pool Makeup or ICW systems to perform their safety related design functions.

The design function of the affected line is to provide a minimum of 150 gallons per minute (GPM)of salt water makeup to the spent fuel pool in the event that a loss of fuel pool cooling capability occurs.This function will be retained by use of a hose connection to I-SH-21241 or I-SH-21338.

The probability of occurrence of a malfunction of equipment.

important to safety previously evaluated in the FSAR has not been increased.

No new failure modes for active equipment are introduced by the proposed configuration.

Valve I-SH-21241 or valve I-SH-21338 is now required to open to supply spent fuel pool makeup.

ST~LUCIE UNIT 1 SAFETY EVALUATION OF A BLIND FLANGE ON LINE I-3"-CW-160 SPENT FUEL POOL MAKEUP SALT WATER BACKUP PAGE 2 SAFETY EVALUATION (Continued):

However, the probability of a malfunction of that valve is no greater than that of I-SB-21386.

The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR have not been increased since the proposed configuration does not create a new path for uncontrolled radioactive releases and will not adversely affect any radiation monitoring equipment or equipment which performs a containment isolation function.The proposed configuration will not impact any equipment which is required to initiate actuation of any safety systems.The proposed piping configuration will not adversely impact the ability of the Spent Fuel Pool Makeup or ICW systems to perform their safety related design functions.

A failure of the alternate spent fuel pool makeup flowpath has been evaluated and determined to have no significant impact on the ability of the ICW system to perform its other Safety Related functions.

The possibility of an accident of a different type than any previously evaluated in the FSAR has not been created since the proposed configuration does not add or affect any equipment capable of initiating an accident.The possibility of a malfunction of a different type than any evaluated previously in the FSAR has not been created since the proposed configuration will not inhibit or otherwise adversely affect the operation of any equipment important to safety.A malfunction of the passive blind flange is not likely.The alternate spent fuel pool makeup flowpath is effectively equivalent (ball valve vs.butterfly valve)to the flanged line and does not create the possibility of a different type of malfunction.

The proposed activity does not reduce the margin of safety as defined in the basis for any Technical Specification since the proposed configuration of the piping system will not impact the operation of the ICW or Spent Fuel Pool Makeup systems as required per the Technical Specifications or the FSAR.The proposed configuration, with identified backup salt water supply to the Spent Fuel Pool Makeup system, is functionally equivalent to the original configuration.

ST~LUCIE UNIT 1 SAFETY EVALUATION OF COMPONENT COOLING WATER HEAT EXCHANGER 1A"TUBEGARDS" INTRODUCTION:

The purpose of this safety evaluation is to allow the installation of"Tubegards" into the 1A Component Cooling Water Heat Exchanger (CCW HX).The TubeGards are installed into each unplugged tube at the upstream tubesheet to reduce the effects'f macrofouling (marine growth)on the HX tubes.The Tubegards will be installed on a test basis for a period not to exceed one operating fuel cycle (Cycle 11)., The performance of the Tubegards will then be evaluated to determine if permanent installation into one or both (1A and 1B)CCW HX's is warranted.

The installation of TubeGards into the 1A CCW HX will have no impact on plant operation and safety.Therefore, NRC approval is not required prior to implementation.

This evaluation concludes that the installation of the TubeGards does not represent an unreviewed safety question, nor require a change to the Technical Specifications.

SAFETY EVALUATION:

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

The CCW HX's are utilized for accident mitigation and are not considered to be accident initiating components.

The TubeGards act as a strainer, similar in function to strainers already installed in the ICW system.Installation of the Tubegards within the 1A CCW HX physically prohibits the Tubegards from increasing the probability of previously evaluated accidents.

The consequences of an accident previously evaluated in the FSAR have not been increased by the installation of TubeGards into the 1A CCW HX.No failure modes of Tubegards have been identified which prevent the ICW and CCW systems from performing their design Safety Related functions.

Tubegards are designed for use in heat exchanger applications.

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR has not been increased by the installation of Tubegards.

The Tubegards are passive devices contained within the ICW side of the 1A CCW heat exchanger and are physically separated from any equipment outside the ICW system.No credible failure mechanisms of the Tubegards have been identified which would cause the failure of the 1A CCW HX or the malfunction of any ICW system components.

Additionally the installation of Tubegards does not alter the function of any existing components.

ST~LUCIE UNIT 1 SAFETY EVALUATION OF COMPONENT COOLING WATER HEAT EXCHANGER 1A"TUBEGARDS" PAGE 2 SAFETY EVALUATION (Continued):

The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR have not been increased by the installation of the TubeGards.

The TubeGards and associated failure modes are isolated within the ICW piping system.The failure modes and their effects are enveloped by the existing failure analysis in the FSAR.The Tubegard failure modes and effects have been evaluated for the potential to create an accident of a different type than previously evaluated in the FSAR.The Tugegards are contained within the ICW system at the channel head of the 1A CCW HX, and neither the ICW system or the 1A CCW HX are considered to be accident initiating components.

The installation of Tubegards does not create the possibility of a malfunction of a different type than evaluated previously in the FSAR, since, all credible failure modes are enveloped by existing analyses which considers loss of an ICW train.The installation of Tubegards within the 1A CCW HX does not reduce the margin of safety as defined in the bases for any Technical Specification since the Tubegards do not negatively affect the capability of the ICW and CCW systems to provide the required cooling capacity for the continued operation of vital components and Engineered Safety Feature equipment during normal and accident conditions.

The overall positive benefits demonstrated in the use of Tubegards within the condenser waterboxes, i.e., reduced tube damage due to microfouling induced pitting, and increased system availability due to reduced fouling rates, provide reasonable assurance the Tubegards will perform as intended in the 1A CCW HX and provide similar positive results.

ST~LUCIE UNIT 1 SAFETY EVALUATION FOR THE USE OF P-54 AND E-4 FOR STEAM GENERATOR DURING OUTAGE INTRODUCTION The use of P-54 to provide access to cables and hoses used to support steam generator activities such as ECT and sludge lancing has been previously evaluated and Facility Review Group (FRG)concurrence obtained and documented.

During the Unit 1 outage, a similar configuration is being utilized on P-54 and E-4 with minor modifications to the closure blind flange for P-54.The modification on P-54 entails the use of an extended spool piece with a flange adapter to serve as the blind flange.The intended purpose of the blind flange is still being met by the modified"spool-piece blind".The use of P-54 and E-4 during refueling Mode: The basis for the Technical Specification is to provide air tight closure such that there is no direct path between the containment atmosphere and the outside atmosphere.

The present closure configuration of P-54 and E-4 complies with the Technical Specification by providing a seal (RTV seal)on P-54 at the outside containment side and air tested, and by providing a seal (RTV seal)on E-4 at the containment and outside containment sides of the penetration.

Use of P-54 and E-4 during reduced inventory:

The use of P-54 and E-4 during reduced inventory is addressed in general maintenance procedure 1-M-0060 which provides specific instructions to rapidly close these penetrations during loss of shutdown cooling while at reduced inventory.

SAFETY EVALUATIONS The original intent on the use of P-54 as previously evaluated during the previous outage has not been altered.In conclusion, the present arrangement of P-54 and E-4 does not: 1)Increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, 2)Create the possibility of an accident or malfunction of a different type than previously analyzed, or 3)Reduce the margin of safety as defined in the bases for any Technical Specification.

ST~LUCIE UNIT 1 SAFETY EVALUATION FOR RAB ELECTRICAL EQUIPMENT AND BATTERY ROOM HVAC INTRODUCTION:

This safety evaluation addresses changes to the description of the St.Lucie Unit 1 Reactor Auxiliary Building (RAB)Electrical Equipment and Battery Room Ventilation System as presently stated in the FSAR Section 9.4.2.2.2.

The changes are a result of inconsistencies between the FSAR and other design documents which were discovered during preparation for the Electrical Distribution System Functional Inspection (EDSFI)and are as follows: System flow rates provided on drawing 8770-G-862 (FSAR Figure 9.4-1)were not adjusted following the addition of fire dampers installed by PC/M 269-183 and PC/M 260-183 in the subject ventilation system to meet Appendix R requirements.

The installation of fire dampers increased system resistance which reduced the supply fans'apacities of the reduced system flow rates.Upon loss of offsite power (LOOP), only the battery room exhaust fans are automatically connected to the emergency diesel generators.

The electrical equipment rooms supply and exhaust fans are manually restarted by administrative control.Presently the FSAR states,"Upon loss of off-site power, the system is automatically connected to the on-site emergency diesel generator sets".This implies all the fans in the system are automatically connected to the emergency generator sets, which is not correct, per the EDG Electrical Load Calculation.

An FSAR Change Package (FCP)has been developed that corrects FSAR Section 9.4.2.2.2 to agree with plant CWD's and will accurately describe the system operation upon a LOOP.During an emergency condition, which involves a LOOP, the temperature in the electrical equipment, static inverter, and battery rooms may exceed 104 degrees fahrenheit.

The FSAR currently does not address the acceptability of this condition.

An FCP had been developed to provide additional description to FSAR Section 9.4.2.2.2 and states it's acceptability.

The description in the FSAR does not agree with the as-built condition concerning the number of rooms ventilated, the flow path, and the use of non-safety related air conditioning units.An FCP has been developed with FSAR 9.4.2.2.2 revised to provide the correct description of the system.The FSAR states,"Electrical equipment room temperatures exceeding 110 degrees Fahrenheit are annunciated in the control room".The basis for the 110 degrees fahrenheit setpoint is not provided in the FSAR.A FCP has been developed which adds to the FSAR Section

ST~LUCIE UNIT 1 SAFETY EVALUATION FOR RAB ELECTRICAL EQUIPMENT AND BATTERY ROOM HVAC PAGE 2 INTRODUCTION (Continued):

9.4.2.2.2 the bases for the 110 degrees Fahrenheit setpoint.This information will preclude future confusion concerning this setpoint.SAFETY EVALUATION The probability of occurrence of an accident previously evaluated in the FSAR has not been increased since this change does not affect any accident initiating components.

The RAB Electrical Equipment and Battery Room Ventilation System does not contain or affect any accident initiating component.

The consequences of an accident previously evaluated in the FSAR have not been increased by this change since this change does not have a detrimental affect on any equipment required to mitigate the effects of an accident.The RAB Electrical Equipment and Battery Room Ventilation System has been shown to still perform its safety related function assuming a single active failure of a supply fan and will not alter the radiological consequences of an accident evaluated in the FSAR.All safety related equipment serviced by this ventilation system have been evaluated for the expected room temperatures under normal and emergency conditions and shown to be acceptable.

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased since this change does not affect the function of any existing components, and thus does not increase the possibility of their failure.The safety related electrical equipment in Electrical Equipment Rooms 1A, 1B, 1C, the static inverter room and Battery Rooms 1A and 1B have not been impacted by this change.The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR have not been increased since this change does not have a detrimental effect on any safety related equipment or components.

The possibility of an accident of a different type than any evaluated previously in the FSAR has not been created since this change will not inhibit or otherwise adversely affect the operation of the RAB Electrical Equipment and Battery Ventilation System.The components of the change are in compliance with the FSAR requirements for the system elements.

ST~LUCIE UNIT 1 SAFETY EVALUATION FOR RAB ELECTRICAL EQUIPMENT AND BATTERY ROOM HVAC PAGE 3 SAFETY EVALUATION (Continued):

The proposed change does not reduce the margin of safety as defined in the basis for any Technical Specification since the RAB Electrical Equipment and Battery Room Ventilation System is not addressed in the Technical specifications.

The changes have been shown to not have a detrimental effect on the safety related equipment serviced by this ventilation system.

ST~LUCIE UNIT 1 SAFETY EVALUATION FOR REMOVAL OF VALVE I-TCV-14-4A AND-4B 45 DEGREES STOP DEVICE INTRODUCTION:

Air operated temperature control valves I-TCV-14-4A,-4B are butterfly type valves located in lines I-, 30"-CW-77 at the outlet of the Component Cooling Water Heat Exchangers (CCWHE)1A and 1B.The valves automatically control ICW flow from the exchangers.

They are modulated opened and closed according to the outlet water temperature of the shell side of the CCWHE.Valve closure is limited to 254 from full closed position (by pneumatic relay)to prevent turbulent flow and valve damage.There is no design limitation on the maximum valve opening, however a mechanical stop device is installed on the valves to limit the valve opening to a maximum of 45 degrees (90 degrees represents valve fully open).The plant desires that the mechanical stops be removed during CCW heat exchanger testing and for the duration of the outage.The testing is being performed in response to Generic Letter 89-13.The purpose of this change involves removal of the 45 degree mechanical stop associated with ICW temperature control valves (I-TCV-14-4 A&B).The proposed change is necessary to ensure proper testing of the CCW heat exchanger heat removal capability.

The plant intends to maintain the current calibration on the controllers for these valves.Thus, the valve will still modulate between 45 degrees open and 254 open.However, for testing purposes the plant intends to fail the valve to the full open position.St.Lucie Unit 2 currently successfully operates with these valves modulating between full open and the minimum stop.SAFETY EVALUATION~

The proposed activity does not increase the probability of occurrence of an accident previously evaluated in the FSAR because the subject TCV's are not accident initiating devices.The proposed activity does not increase the consequences of an accident because the ICW flow rate is increased, and the system remains capable of delivering the minimum flow requirements for accident conditions.

All components retain their functions and capabilities with the increased flow.The proposed activity does not increase the probability of occurrence of a malfunction of equipment important to safety because the ICW pumps remain capable of operating within their performance-curve and the valves'etpoints and operation are not affected.

ST~LUCIE UNIT 1 SAFETY EVALUATION FOR REMOVAL OF VALVE I-TCV-14-4A AND-4B 45 DEGREES STOP DEVICE Page 2 SAFETY EVALUATION (continued):

The proposed activity does not increase the consequences of a malfunction of equipment important to safety because the primary equipment, ICW pumps, and the TCV s still perform within their design with no new failure modes introduced.

The proposed activity does not increase the probability of an accident of a different type than any previously evaluated because component replacement does not take place and the operation of the TCV's is unchanged in that valve opening based on temperature is maintained.

The proposed activity does not increase the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the FSAR because component replacement does not take place and functionally there is no change to the response of the system.The proposed activity does not reduce the margin of safety as defined in the basis for any technical specification because the valves'etpoints remain the same and the flow rate increases providing greater heat sink capabilities during accident conditions.

ST~LUCZE UNIT 2 SAFETY EVALUATION FOR CLAMP FOR 4-WAY HYDRAULIC VALVES ON MAIN FEEDWATER ISOLATION VALVES INTRODUCTION Several of the endcap capcrews on the 4-way hydraulic valves have been found broken on the St.Lucie Unit 2 Main Feedwater Isolation Valves (MFIV, HCV-09-1A/2A/1B/2B).

Preliminary inspection of the capscrews by the ZPN-ESI lab indicate overload as the failure mechanism.

As a result of this preliminary investigation, the remaining capscrews are deemed suspect or indeterminate until a thorough investigation of root cause can be completed.

As a prudent measure, a clamp has been designed to replace the function of the capscrews, to assure the MFIV 4-way hydraulic valve will remain operable per the original design.Installation of this clamp will have no impact on plant safety or operation.

A review of the plant Technical Specifications and the FSAR has shown that there are no unreviewed safety questions or Technical Specification changes involved.SAFETY EVALUATION:

The probability of occurrence of an accident previously evaluated in the FSAR has not been increased since installation of the clamp does not affect any accident initiating components.

Installation of the clamp acts to replace the function of the original endcap capscrews and is considered equivalent.

Continued reliable nondegraded operability of the MFIV 4-way hydraulic valve is therefore assured.The consequences of an accident previously evaluated in the FSAR have not been increased since installation of this clamp does not change or alter the ability of the MFIV to respond to a MSIS or AFAS signal (i.e., to close and remain closed).Installation of the clamp does not adversely affect any other equipment required to mitigate the effects of an accident.The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR has not been increased.

The clamp and the effect of the additional weight on the 4-way hydraulic valve mounting capscrews have been evaluated as acceptable to assure the valve can perform its function during a DBE event.Installation of the clamp does not alter the function of any existing components and thus does not increase the possibility of their failure.

ST~LUCIE UNIT 2 SAFETY EVALUATION FOR CLAMP FOR 4-RAY HYDRAULIC VALVES ON MAIN FEEDMATER ISOLATION VALVES PAGE 2 SAFETY EVALUATION (continued):

The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR have not been increased since the clamp does not alter the response or function of the MFIV's during an accident, nor interact with any other equipment important to safety.The possibility of an accident of a different type than any evaluated previously in the FSAR has not been created since installation of the clamp does not add or affect any equipment capable of initiating an accident.The clamp function is equivalent to the function of the endcap capscrews.

The additional weight of the clamp has been evaluated for seismic considerations with respect to the MFIV valve and actuator, and was determined to be acceptable.

The possibility of a malfunction of a different type than previously evaluated in the FSAR has not been created since the addition of the clamp will not inhibit or otherwise adversely affect the operation of the MFIV or the 4-way hydraulic valve.The clamp is external to the moving parts in the 4-way hydraulic valve.The clamp will not adversely affect the function of any components within the valve, i.e., the valve components are metal to metal along the axis, and the valve components are of sufficient thickness.

The addition of the clamp does not reduce the margin of safety as defined in the basis for any Technical Specification since the clamp function is to ensure the 4-way hydraulic valves remain operable as per the original design.This is a prudent measure which provides greater assurance that the function of the MFIV is maintained.

4 ST~LUCIE UNIT 2 SAFETY EVALUATION FOR INSTALLATION OF BLIND FLANGE ON PIPING AT CONTAINMENT PENETRATION P-56 INTRODUCTION Recent Local Leak Rate Test (LLRT)results on PSL Unit 2 Penetration P-56 have shown increasing leakage rates through valve FCV-25-26 and/or FCV-25-36, which have been within the acceptable limits for this penetration.

However, upcoming LLRT surveillance may result in an unsatisfactory leakage rate for the penetration.

As part of a contingency plan for restoration of the penetration, this safety evaluation will evaluate installation of a blind flange (s)as necessary to achieve a satisfactory LLRT.This penetration is the makeup path for the continuous containment purge/hydrogen purge system.The implementation of this temporary modification will have no adverse affect on plant safety or operation.

A review of the plant Technical Specifications and the FSAR has shown that there are no unresolved safety questions or Technical Specifications changes involved.Penetration P-56 and the associated valves are required for containment isolation, therefore this Safety evaluation is classified as Safety Related.SAFETY EVALUATION:

The probability of occurrence of an accident previously evaluated in the FSAR has not been increased since this temporary modification does not affect initiating components.

The consequences of an accident previously evaluated in the FSAR have not been increased by this temporary modification since this modification does not affect any equipment required to mitigate the effects of an accident.The Safety Related function of the system is to maintain containment integrity at penetration P-56.This function is accomplished by the installation of the blind flange and a successful LLRT.The penetration will still be required to meet the LLRT and containment isolation requirements.

Installation of a blind flange in lieu of a valve is acceptable as a passive barrier for containment isolation.

The blind flange configuration is properly specified for the system design conditions and the makeup function of the system may be accomplished by use of the containment vacuum relief system.The exhaust function of the system remains intact and operational.

Therefore, the Technical Specifications relating to containment isolation and containment pressure remain unaffected.

This temporary modification does not alter the function of any existing equipment important to safety, and thus does not increase the probability of their failure.The installation of the blind ST LUCIE UNIT 2 SAFETY EVALUATION FOR INSTALLATION OF BLIND FLANGE ON PIPING AT CONTAINMENT PENETRATION P-56 PAGE 2 SAFETY EVALUATION (Continued):

flange does not increase the loading (weight)on the penetration, therefore the analysis for the penetration loading is unchanged.

The installation of the blind flange serves to enhance the containment isolation function since it is a passive device.The rigging off of the 48" penetration does not adversely affect the analysis of this penetration.

The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR have not been increased since the valves are normally open, fail closed valves which close on CIAS signal.Installation of the blind flange is a passive barrier for containment isolation.

No additional failure modes are introduced, since the potential for leakage through the gasket currently exists for the valve and seat leakage would be eliminated.

The possibility of an accident of a different type than any previously evaluated in the FSAR has not been created since this temporary modification does not add or affect any equipment capable of initiating an accident.The penetration is still subject to the same LLRT acceptance criteria for containment isolation.

The possibility of a malfunction of a different type than any evaluated previously in the FSAR has not been created since this temporary modification will not inhibit or otherwise adversely affect the operation of other components.

The valves are fail closed, the blind flange passively serves this same function.This temporary modification maintains the margin of safety of the Containment Isolation Valve Technical Specification since it replaces an active device with a passive device that is designed to the system design parameters.

With regards to the Containment Pressure (normal)Technical Specification, this Technical Specification is unaffected since the exhaust portion of the system will still function to reduce pressure inside containment.

ST~LUCIE UNITS 1&2 SAFETY EVALUATION FOR CVCS PURIFICATION FILTER PARTICULATE RATING UPGRADE INTRODUCTION:

This Safety Evaluation addresses the technical implications for the use of 6, 2 or 1 (1)micron absolute filter elements in the Unit'1 and Unit 2 Chemical and Volume Control Systems (CVCS)Purification Filters on a test basis.The current design requirements call for 95%and 984, for Unit 1 and 2 respectively, retention by weight of particulate 2 microns and larger per St.Lucie Unit 1 FSAR Amendment 10 and St.Lucie Unit 2 FSAR Amendment 6.The higher efficiency 1 micron absolute filter elements will capture all particulate larger than 1 micron in size, plus capture 99%of the particulate between 0.6 microns and 1.0 micron.The elimination of this particulate will have the following positive effects on both Units systems: 1.)2.)3.)4~)reduce out-of-core radiation, reduce the formation of crud deposits, minimize resin fouling, reduce personnel radiation exposure.Although the CVCS Purification Filter 1A and 2A, for Unit 1 and 2 respectively, does not perform any safety function, it is located in a Quality Group C system.Therefore, this Safety Evaluation is classified as Nuclear Safety Related.The use of smaller particulate rated, higher efficiency, filter elements in the CVCS Purification Filter 1A does not adversely impact plant safety nor operation.

This Safety Evaluation concludes that there are no unreviewed safety questions or Technical Specification changes involved with this modification.

SAFETY EVALUATION:

The probability of occurrence of an accident previously evaluated in the FSAR has not been increased since the test basis modification does not adversely affect any accident initiating components.

The differential pressure drop of the new filters is less than the pressure drop of the original filters.The test basis modification does not alter the function of any existing components, and thus does not increase the possibility of failure.The installation of 6, 2 or 1 (1)micron absolute filters will result in a reduction of particulate in the RCS and the CVCS, leading to increased reliability of system components.

The consequences of an accident previously evaluated in the FSAR have not been increased by this test basis modification since it does not adversely affect any equipment required to mitigate the effects of an accident.

ST.LUCIE UNITS 1 8 2 SAFETY EVALUATION FOR CVCS PURIFICATION FZLTER PARTICULATE RATING UPGRADE PAGE 2 SAFETY EVALUATION (Continued):

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased because the test basis modification does not impact the function of any existing components, does not alter the high differential pressure drop across the filter alarm set point and does not increase the possibility of their failure.The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR have not been increased since this test basis modification does not create a new path for uncontrolled radioactive releases and will not adversely=affect any equipment required to mitigate the consequences of an equipment malfunction.

The possibility of an accident of a different type than any evaluated previously in the safety analysis report has not been created since this test basis modification has no adverse effect on any equipment capable of initiating an accident and no new failure modes are introduced through the installation of the above-mentioned absolute filters.The possibility of a malfunction of a different type than any evaluated previously in the safety analysis report has not been created since this test basis modification will not inhibit or otherwise adversely affect the operation of any equipment important to safety.The test basis installation of the above-mentioned absolute filter elements does not reduce the margin of safety as defined in the basis for any technical specification since the function of the system components remain the same.The new filters provide better filtration than the original design calls for and the filters will be changed at the same differential pressure requirements.

ST~LUCZE UNITS 1 AND 2 GAGGING OF SAFETY RELIEF VALVE V3483 THAT PROVIDES OVERPRESSURE PROTECTION OF SHUTDOWN COOLING PIPING INTRODUCTION On October 20, 1991, Unit 1 Shutdown Cooling (SDC)return relief valve V3483 was reported to be leaking RCS inventory while the unit was in mode 4 for refueling.

A leaking SDC relief could not be, repaired quickly in the field, and prevented the use of the"A" train of SDC, and subsequent plant cooldown.To terminate the reported loss of RCS inventory and to make both SDC trains available, a 50.59 was required to evaluate the plant response to temporary gagging this relief valve.FSAR section 6.3.2.2.6.d describes the SDC return relief valves V3468 and V3483 as redundant overpressure protection devices for the shutdown cooling system during solid RCS operations with all charging pumps running.The setpoint of these valves is 300 psig and each valve has a capacity of 155 gpm.This value is less than total charging pump capacity of 132 gpm.Section 9.3 of the Unit 1 FSAR, table 9.3-27 lists the design pressure of the suction line to the Low Pressure Safety Injection pumps to be 300 psig, the same value as the relief's setpoint.The hydro pressure for the SDC suction line is 440 psia.FSAR section 9.3.5.2.2 states that"(the SDC suction isolation) valves V3651 and V3652 in the 1B loop and V3480 and V3481 in, the 1A loop automatically close whenever the RCS pressure exceeds the design pressure of the shutdown cooling system." SAFETY EVALUATION:

The probability of occurrence of an accident is not increased by the gagging of a single SDC relief.The accident of concern is the failure of the backpressure regulating valves while in solid RCS plant operations.

This may possibly rupture the SDC system suction piping during an overpressurization of the RCS while SDC is in service, which would result in the loss of decay heat removal capability through SDC.The probability of this accident is predicated on the failure of the backpressure regulating valves during solid RCS operation, the start of charging pumps, HPSI pumps or RCS pumps, or the energization of Pressurizer heaters.As a compensatory measure, HPSI, RCS pumps and Pressurizer heaters are de-energized by procedure OP 1-0020127 prior to solid operation.

Therefore, the only credible accident of concern is the failing closed of the backpressure regulators with the pressurizer filled solid with water and one or more charging pumps still in operation.

The initiation of this scenario is independent of SDC relief valve status.

ST~LUCIE UNITS 1 AND 2 GAGGING OF SAFETY RELIEF VALVE V3483 THAT PROVIDES OVERPRESSURE PROTECTION OF SHUTDOWN COOLING PIPING PAGE 2 SAFETY EVALUATION (Continued):

The gagging of a single SDC relief does not increase the probability of occurrence of a malfunction of equipment important to safety because the SDC system still retains overpressure protection from the other operable train's SDC relief valve.Additional overpressure protection is afforded by the Low Temperature Overpressure Protection system, which relieves RCS pressure at less than 350 psia as per Technical Specifications.

The consequences of an accident is not increased by the gagging of a single SDC relief.FSAR section 6.3.2.2.6 d states that each SDC valve has the capacity to relieve the flow from three charging pumps operating.

Therefore, with the compensatory measure of having both SDC trains in service, at least one SDC relief valve can relieve pressure for both SDC trains before the SDC suction isolation valves fully stroke closed in approximately 50 seconds.The rate of pressure increase in this scenario is dependent upon the compressibility of water, and the amount of compressible gases in system high points.Therefore, the consequences are independent of gagging a single SDC relief valve because of the opposite train's SDC relief valve.The consequences of a malfunction of equipment important to safety as previously evaluated in the FSAR is not increased by the gagging of a single SDC relief valve.Assuming the gagging of one relief valve and the failure of the other relief valve, overpressure protection of the SDC system is afforded by the LTOP system.During normal SDC operation, with the pressurizer solid, RCS temperature will require the LTOP setpoint to instantaneously open both PORV's at a pressure not to exceed 350 psia.This is below the SDC system hydro pressure of 440 psia.Therefore, the PORV's provide overpressure protection until the SDC suction isolation valves shut.The possibility for an accident or malfunction of a different type than previously evaluated in the FSAR is not created by gagging a single SDC relief based upon the above information.

The margin of safety for the SDC system is not explicitly stated in the Technical Specifications.

The margin is assured to be the continued availability of having an operable SDC system to use in removing decay heat from the RCS.

10 CFR 50.59 Evaluations Temporary Changes via Jumper/Lifted Leads Recpxests 10 CFR 50'9 Evaluation for Temporary Changes via Jumper/Lifted Leads Unit: 1 Request Number: 1-1-9 Components/Systems Affected: Radiation Monitor Cabinet.Install jumper to obtain control of FCV-6627X.

Description of Change: This jumper removes the signal to FCV-6627X from the Liquid Radwaste Effluent Line's gross radioactivity monitor.This monitor is discussed in Technical Specification 3.3-12 which states that if the minimum channels operable is less than required, effluent releases may continue for up to 14 days provided that at least two different independent samples are analyzed and at least two qualified staff members verify the release rate calculations and discharge line valving.Safety Evaluation Summary: The proposed activity does not increase the probability of occurrence of an accident previously evaluated in the FSAR.The proposed activity does not increase the consequences of an accident previously evaluated in the FSAR The proposed activity does not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR.The proposed activity does not increase the consequences of malfunction of equipment important to safety previously evaluated in the FSAR.The jumper removes the signal from the radwaste monitor which per FSAR continuously monitors discharge and auto terminates if exceeded.However action per Technical Specification for an out of service radwaste monitor was taken prior to release.The proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the FSAR.The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.Calculation of release were performed prior to release ensuring radioactivity would be to low to require verification of release.

10 CFR 50 59 Evaluation for Temporary Changes via Jumper/Lifted Leads Unit: 1 Request Number: 1-1-9 Safety Evaluation Summary (Continued):

The proposed activity does not reduce the margin of safety as defined in the basis for any Technical Specification.

Action taken were the same precautions as those for an out of service monitor.

10 CFR 50.59 Evaluation for Temporary Changes via Jumper/Lifted Leads Unit: 1 Request Number: 1-1-18 Components/Systems Affected: Feedwater Regulating Control System Description of Change: The reason for this jumper was to isolate a leaking transmitter line.This jumper will isolate a leading section of instrument tubing supplying FT-8011.Installation of this jumper will retain all functions of FT-8011, as it will still be in service.Safety Evaluation Summary: The proposed activity does not increase the probability of occurrence of an accident previously evaluated in the FSAR.The proposed activity does not increase the consequences of an accident previously evaluated in the FSAR.The proposed activity does not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR.The proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR.This change will isolate a leaking instrument line while maintaining the operability of that instrument.

The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.The flow transmitter will remain operable with this jumper installed.

The proposed activity does not create the probability of an accident of a different type than any previously evaluated in the FSAR.The proposed activity does not reduce the margin of safety as defined in the basis for any Technical Specification.

No loss of plant function or control will occur as a result of this jumper.

10 CFR 50.59 Eva1uation for Temporary Changes via Jumper/Lifted Leads Unit: 1 Request Number: 1-1-54 Components/Systems Affected: Safety Evaluation to allow the use of spare CEDM Reactor Head Power Cables supplied by ABB-CE.Description of Change: This evaluation allows the use of new spare Control Element Drive Mechanism (CEDM)Reactor Head Power Cables between the drive and the Refueling Disconnect Panels (RDPs).The spare cables are to be used only if one (or more)of the existing cables fail and requires replacement.

These cables are not safety related and are not required to be seismic class I, but are located over and around safety related equipment and must be seismic class II.They are therefore classified as Quality Related.Based on the following evaluation, the use of these spare cables during the next operating cycle will not pose any safety hazard to the plant.Safety Evaluation Summary: The purpose of this evaluation is to allow the use of the new spare CEDM Reactor Head Power Cables if one (or more)of the existing cables fails and requires replacement before or during the next operating cycle (after the 1991 refueling outage).The only accident evaluated in the FSAR that could be affected by the spare cables is a Control Element Assembly (CEA)drop.This accident is evaluated in FSAR section 15.2.3.A failure of one of these spare cables could cause a CEA to drop.The probability of a CEA drop event will not increase as a result of using these spare cables because the spare cables meet or exceed the requirements of FPL Specifications EN-2.14 except as noted and evaluated in ABB/CE Certificate of Conformance.

This specification was written and approved to ensure that the replacement CEDM power cables would comply with all operating requirements for their intended use.This as a result, the spare cables will be better able to perform their intended function then the cables they replace and, will be less likely to fail.Also, these spare cables will provide the same function in the same manner as the original cables and have the same electrical characteristics.

Therefore, using the spare cables will not increase the probability of occurrence of an accident previously evaluated in the FSAR.

0 10 CFR 50'9 Evaluation for Temporary Changes via Jumper/Lifted Leads Unit: 1 Request Number: 1-1-54 Safety Evaluation Summary (Continued):

The failure of one of these spare cables could cause a CEA drop event.However, the plant response to this transient is not altered by the replacement of these cables.The new cables are designed to withstand a seismic event and not degrade to the point that they will affect any safety related equipment and are also designed to withstand the effects of a loss of coolant accident without a loss of integrity.

Thus, the spare cables will not block the containment sump screens and will not impact the available NPSH for the ECCS pumps.Therefore, the use of the spare cables will not increase the consequences of accident previously evaluated in the FSAR.The CEDM cables are used to provide controlled movement of the control element assemblies (CEA's)into and out of the core.However, the CEDM's are fail safe.That is, they are designed to fall into the core upon failure of a CEDM (including interruption of power to the reactor trip switch gear breakers).

There is no credible cable failure that would prevent the CEA's from falling into the core.The fiberglass braid will maintain the cable integrity because it is capable of withstanding the effects of a LOCA while keeping the conductors together and keeping the cable filler and binder tape contained inside.This precludes the possibility of containment sump screen blockage by the cable filler and/or binder tape.This braid will also provide additional abrasion protection to the individual conductor insulation.

Therefore, using the spare cables will not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR.These new spare cables will perform the same function in the same manner as the original cables.The spare cables will not interact with any equipment in any manner that the original cables did not interact with.As such, use of the spare cables will have no effect on the function of equipment important to safety.Therefore, using the spare cables will not increase the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR.Again, these new spare cables will perform the same function in the same manner as the existing cables.As such, all equipment and systems will function in the same manner as is currently described in the FSAR.Therefore, using the spare cables will not create the possibility of an accident of a different type than previously evaluated in the FSAR.

0 10 CFR 50'9 Evaluation for Temporary Changes via Jumper/Lifted Leads Unit: 1 Request Number: 1-1-54 Safety Evaluation Summary (Continued):

These cables will not degrade and affect any safety related equipment.

They will function the same as the existing cables and the failure modes for the existing cables have been analyzed in the FSAR.or have been protected against by using the fiberglass braid.Therefore, using the spare cables will not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the FSAR.The operation of the CEDM system will not change as a result of using the new spare cables.Therefore, the methods used to meet the requirements of the Technical Specifications are not changed.The bases behind the Technical Specifications are still valid and the margin of safety as defined in those bases is not reduced.