L-88-164, Annual Rept Re Plant Changes/Mods Made Between 861006 & 871006

Annual Rept Re Plant Changes/Mods Made Between 861006 & 871006

ML17221A717
Person / Time
Site:	Saint Lucie
Issue date:	10/06/1987
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17221A716	List: L-88-164, Annual Rept Re Plant Changes/Mods Made Between 861006 & 871006 ML17221A715
References
L-88-164, NUDOCS 8804130318
Download: ML17221A717 (60)

Text

PCM 182-283 SYSTEM THERMOCOUPLES INTRODUCTION The folloving thermocouples are installed in such a way that will not allow 'the removal of the element for calibration or replacement:

TE-1 through TE-6 Continuous Containment Filter Train Heaters TE25-59 6 -60 Continuous Containment Filter Train Charcoal Absorbers This PC/M modifies the mounting details of the e thermocouples to facilitate their removal.

The original scope of this package included the remounting of a total of eight (8)thermocouples.

However, the modi-fications to TE 25-59 and TE 25-60 necessitate modifications to TE 25-58 and TE 25-61 (see BCS-182-283.3001, Sh 1 6 2).

Therefore, this PC/M modifies the mounting details for a total of ten (10) thermocouples.

SAFETY ANALYSIS With respect to Title 1Q of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety ques-tion:

(i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The Continuous Containment Purge/Hydrogen Purge System is not required for the safe shutdown of the reactor.

Therefore, the system is not safety related other than the containment isolation valves and penetra-tions, which are Quality Group B and seismic Category I as required by GDC 56.

The system is seismically qualified to preclude damage to neighboring safety-related equipment.

The modifications performed by this PC/M have no effect on the system operation and do not change any equipment locations.

Only mounting details for eight thermocouples are being modified, to facilitate maintenance.

The foregoing constitutes, per 10CFR50.59(b),

the written safety evaluation which provides the basis that this change does not involve an unreviewed safety question; therefore, prior Commission approval is not required for implementation of this PC/M.

eSoelS0SiS SSoeoe PDR ADOCK 05000389 R

DCD

PCM 211-283 SECONDARY SYSTEM PIPING VIBRATION RESTRAINTS M'an during power ascension tests.

MhQe the cause of this vibration is not jaxrwn(systan transients, air in the lines, abmrmal op-erating mades, etc.)

we have been requested by operatians to pro-vide vibration restraints ta protect the syst~ integrity.

Fur-ther ixnrestigatians willbe conducted to determine the causes of the vibration.

SAEVA AHALYSIS Hith respect to Title 10 of the Code of Fed'egulations, Part 50.59, a proposed change shall be dered to involve an unxeviewed safety question; (i) if the probability of occurence or the con-sequences of an accident or malfunctian of equ:iymnt important to safety previuxxsly evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previausly in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specificatian is md+~.

%he Secondary Systems are mn-saf

/mn-seisnu.c.

This PCH~

cribes changes xa restx-dnt design for these systans

hence, m safety function is affected.

~refore, this PC/M does not increase the probability or con-sequences of a previausly analyzed accident mr does it, create any new types of accidents.

In addition it does not reduce the martpin of safety as defined in the bases for any technical spec-ifications.

Acmrdingly, there is m unreviewed safety questions and prior caanissian approval is not required.

PCM 406-283 CONDENSER WATERBOX WORK PLATFORMS INTRODUCTION Various valves, pumps and level indicators located on the east side of the condenser waterboxes require frequent maintenance.

Platforms and ladders are needed to access this area.

PCM 406-283 (Supplements 0 and

1) was issued to provide the required platforms.

Subsequent to the release of PCM 406-283 Supplement 1,

the plant requested a reduction in the sizes of some of the platforms.

SAFETY ANALYSIS With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a

proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

This PCM Supplement provides the details to reduce the size of the condenser waterbox work platforms in accordance with the plant request.

The modifications do not affect the load carrying capability of the platforms and all appropriate load combinations and allowable stresses are still satisfied.

The resultant platforms are non-safety related structures and as such do not perform any functions vital to plant safety.

The platforms are not attached to safety related structures and are located in non-safety areas and therefore, cannot adversely affect safety components or structures.

The proposed change does not involve a

change to the Technical Specifications.

The foregoing constitutes, per 10CFR50.59(b),

the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior Commission approval for the implementation of this PCM is not required.

PCM 008-984 NON-MANUAL PARKING LOT ABSTRACT This engineering package covers the restoration and repaving of the Non-Manual Parking Lot which is on the east side of the St. Lucie Plant. Also included in the package h the removal of the construction fire water tank in the parking lot, as well as the addition of an improved area lighting design.

The parking lot 'is located outside of the plant security fence perimeter.

The modifications included in this design package will not affect any plant safety-related system and are therefore classified as non-nuclear-safety-related.

In addition, the removal of the construction fire water tank wQl not affect the plant fire protection system, since this work cannot be started until PCM 178-985 is implemente*

PCM 178-985 ties the fire water piping downstream from the tank into the plant system.

The restoration and repaving of the Non-Manual Parking Lot and the tank removal do not pose any unreviewed safety questions.

Safet Evaluation The Non-Manual Parking Lot is located outside of the security perimeter fence and willnot be in the vicinityof any plant safety-related'structure or system.

It does not in any way perform or affect a plant safety-related function.

The Non-Manual Parking Lot area lighting does not perform or affect any plant safetywelated systems or function. It is being supplied from LP 260 which is a non~fety related lighting panel and is not loaded on the emergency diesel generator.

The removal of the construction fire water tank and piping does not affect any plant safety-related system or functions.

The city water to the plant is not a safety-related system.

The fire water supply from the tank is not part of the plant fire water system and does not affect that system.

The modifications to the Non-Manual Parking Lot do not change any assumptions made or conclusions drawn in the St. Lucie FSAR.

The repaving of the lot does not adversely affect any site topographic features.

For the above reasons the modifications of the Non-Manual Parking Lot willnot increase the probability of occurrence nor the consequences of a design basis accident or malfunction of equipment important to the safety of the plant.

Additionally, there will continue to be no possibQity of an accident or malfunction different than those already evaluated in the FSAR.

Finally, the margin of safety as defined in the Plant Technical Specifications has not been reduced.

It is therefore concluded that this modification does not pose an unreviewed safety questions pursuant to 10 CFR 50.59 and does not affect any technical specifications.

NOTE:

THIS PACKAGE CONTAINS SAFEGUARD DRAWINGS.

PCM 018-284 SEQUENCE OF EVENTS (SER)

PRINTER INTRODUCTION The summary report for the Detail Control Room Design Review (DCRDR) of the. plant St Lucie No 2

(Docket No 50-389) identified in section 6.7, File No 10, the Human Engineering Discrepancies (HED's) as listed below:

Findings:

(Section 6.7, File No 10)

The sequence of Events Printer is a teletype and has a

speed of approxi-mately 12 lines per minute.

In addition, the printer is very loud and would be quite distracting during a transient.

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

The SER system is non-safety related and performs no safety functions.

By replacing the non-safety related sequence of event recorder (SER)

printer, the probability of occurrence of an accident or malfunction of equipment important to

safety, previously evaluated in the FSAR is not created.

Since the new printer provides a faster printout of events, information vill be available to the operator on a more timely basis.

Technical Specifications are not affected by this PC/M.

The foregoing constitutes, per 10CFR50.59(b),

the vritten safety evaluation vhich provides the basis that this change does not involve any unrevieved safety

question, therefore prior Commission approval is not required for implementation of this PC/M.

KM 028-284 REACTOR COOLANT PUMP SEAL INJECTION SYSTEM INTRODUCTION The Reactor Coolant Pumps

{RCPs) are provided vith mechanical seals on the pump shafts When operating at reactor coolant temperature these seals require cooling in order to prevent damage to the seals and to prolong seal life.

Currently the only means of cooling the RCP Seals is by the Component Cooling Mater System

{CCW) to the seal heat exchanger and the RCP cover heat barrier.

The pump manufacturer has established maximum limit of 10 minutes operating time vithout seal cooling vater.

Operation of the RCP beyond this limitation requires inspection of the seals for any damage.

h problem developed during Post Core Load Hot Functional Testing vhere the RCP seals vere overheating vhen a pump vas idle during the alternate pump combination runs.

This incident required inspection of RCP seals for any degeneration/damage.

I This PCM implements a backup cooling vater system for the RCP seals vhich is intended to eliminate the need for inspection and/or prevent damage to the seals in the event of a total loss of CCW to the RCP seals heat cxchangcr or vhen one or more RCPs are idle during alternate pump combination runs.

This design utiliaes

'some of the existing sections, installed as pert of the temporary system, particularly the hard piped sections at each pump ~

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

~

This modificaion adds a backup cooling ~ster system'rom the CVCS for the RCP seals to preclude degeneration and/or damage to the seals in the event of loss of CCV and during alternate pump combination runs.

The RCPs are discussed in FSAR Section 5.4-1 and the seals and cooling water connections are shown on Figures 5.4-1 and 5.4-3.

The CCW system is discussed in FSAR Section 9.2.2 and the Chemical Volume and Control System in Section 9.3.4.

The proposed change does not involve a unreviewed safety question for the following reasons:

PCM 028-284 1.

The probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased because.')

We system piping and components are designed and stress analyzed in accordance vith the applicable design codes and Regulatory requirements for Safety Class 1 and 2 systems

~

b)

The addition of remote control svitches on the RTGB 203 and local flov indication and local control svitches for the system MOVs have been designed to meet the applicable design codes and seismic Category I requirements.

Tubing and accessories are designed and procurred to Safety Class 2

requirements.

c)

The valves, instrument tubing, piping and supports are classified as Seismic Category I and are designed to vithstand the applicable loading listed in Chapter 3 of the PSAR.

The flov indicating devices used for flov balancing are non-safety, non seismic since they vill be isolated after start-up testing easing safety class root valves..

d)

The foot print loads for those structures used to support piping and components have been evaluated for their effect on the existing structure and found acceptable.

e)

The tvo new motor operated valves are Class IE and are povered from the essential (DG'powered) portion of the 480V AC MCC 2B5.

The impact on the DG loading ha" been evalua ed and found to be negligible due to:

(ii)

Small size of the valve motor operators (2 hp) and short duration of operation (60 seconds).

Valves are being operated remote manually as such they do not affect the automatic DG loading sequence.

The new cabling for the valves vas chosen on the basis of acceptable ampacity, voltage drop and short circuit vithstand.

Cables vere routed in appropriate seismically supported raceways (conduits and trays),

chosen on the basis of separation criteria (voltage level, function, and separation codes)

~

All the equipment is qualified for its intended fuctions.

g)

The isclation of the RCP seal injection system from the interrelated systems (CVCS and RCS) is maintained by appropriate Safety Class 1 (tvo check valves in series) and Class 2 (seal injection actuation valve) valves.

PCM 028-284 2.

The'ossibility for an accident or malfunction of a different type other than any evaluated previously in the safety analysis report is not created because the system will be isolated fry the interrelated systems.

The failure of isolation from interrelated sytems is discussed as bein CVCS: hn inadvertant opening of RCP seal injection actuation valve V2185 will not result in excess RCS inventory since only one charging pump normally operates.

'Ibis failure would cause loss of sufficient cooling of the letdown flow in the letdown cooling heat exchanger.

This would in turn result in e high letdown flow temperature alarm in the control room and subsequent automatic isolation of the letdown flow.

Based on this alarm the operator would be able to take the required corrective action.

RCS:

Two check valves used in aeries assure a

posiiive iaolation from the RCS.

Ho failure of isolation is considered in this case.

Since the RCP Seal Injection System i's part of the CVCS and operates less than 2R of the time the CVCS operates, the RCP Seal Injection System's ccnsidered a moderate energy system.

For moderate energy lines only through wall cracks and water spray need to be postulated'oderate energy lines only crack and spray water.

Systems/equipment inside containment, essential for safe

shutdown, are qualified to operate under such conditions.

3.

Since this system is designed to the same design codes used for original plant construction the margin of safety as defined in the basis for any technical specification is not reduced because:

a)

The isolation of the RCP Seal Injection System from the interrelated systems (CVCS and RCS) is mainta'ned by appropriate Safety Class 1 (two check valves in series) and Class 2 (seal injection actuation valve) valves.

b)

The operation of this system is neither required for plant safe shutdown nor for mitigating consequences of an accident.

The system is installed to protect plant investment.

PCH 028-284 The folloving is the CE statement addressing the impact of this sys em on other plant systems:

The seal injection system is designed to provide seal cooling in the event of a loss of Component Cooling Mater (CO') to the Reactor Coolant Pumps (RCPs).

In addition, the system is used concurrent vith CCM for additional cooling and seal flushing operations.

This is accomplished by providing back-up cooling vater from the Chemical Volume Control Sys em (CVCS) by a remotely controlled isolation valve.

The seal injection system vill therefore minimize the potential damage to the RCP seals.

Strict administrataive controls ~ill ensure that the seal injection system is isolated from the CVCS and RCS during normal plant operation.

The seal injectior, system opera.es by diverting charging pump flov upstream of the regenerative heat ex"hanger.

The diverted flov passes through the seal injection system isolation valve and then bran"hes into a header that supplies eight gallons per minute to each RCF seal.

Once injected into the RCP seal, seven gallons per minute provide cooling to the seal and subsequently enters the RCS and the balance flovs into the RCP-controlled bleedoff.

In order that there is adequate charging flov through the regenerative heat exchanger to cool the letdown fluid, and adequate floe to the RCP seals, tvo charging pumps must be in operation vhen seal injection is used.

This safety evaluation addresses the potential impact of the seal injection system on the integrity of the RCP pressure boundary and the interaction" of the seal injection system vith the nuclear steam supply system (NSSS) auxiliary systems.

This evaluation does not address seal injection system piping or components.

Ihis evaluation applies to Lhe installation of a seal injection system at both St Lucie Units 1 and 2 since the RCPs, seal injection system design, and the NSSS interfaces are the same for both units.

The proposed change does not involve an unrevieved safety question for the folloving reasons:

1.

The probability of occurrence or the consequences of an accident or malfunction of equ-pment important to safety previously evaluated in the safety analysis report is not increased because:

a)

The RCP pressure retaining boundaries have been analyzed in accordance vith the applicable sections of the ASM'ode Section III.

The code stress analysis under vorst case seal injection conditions for the St Lucie Unit 2

RCP shoved that the governing stress occurs at the central portion of the RCP cover.

The remainder of the RCP is virtually unaffected by seal injection.

Under all conditions of seal injection initiation, the requirements of Section III of the ASNE Code are satisfied, and the integrity of the primary boundary is maintained'here is, therefore, no time limit on seal injection

PCM 028-284 tinitiation vith regard to pres ure boundary integri'ty.

h maximum of 1190 allovable cycles (a cycle is defined as the initiation of seal injection at RCS temperatures above 300oF) of seal inje"tion has been calculated.

h conservative value of 597 cycles is used's the operational limit for the initiation of seal injection et RCS temperatures greeter than 300oF vhich is considerably more than expected during the life of the plant.

Ho operational limit is imposed for the initiation of seal injection concurrent vith CCM to the RCPs at RCS temperatures belov 300 F.

b)

The seal injection system. is isolated from the RCS and CVCS during normal plant operation.

Tnis isola.ion is achieved by using tvc Safety Class 1 check valve's in series (RCS isolation) end one Class 2 seal injection isolation valve (CVCS isolation).

c)

Mhen in operation, the seal injection system vill divert 32 gallons per minute of charging flov.

Tvo charging pumps must be in operation to maintain adequate charging flov to the regenerative heat exchanger to cool letdovn.

flov. If inadequate charging flov (i.e., only one charging pump in operation) exists, the letdovn flov vould automatically be isolated on high temperature to protect the CVCS equipment.

d)

Operation of the Emergency Core Cooling System (ECCS) during e Loss of Coolant Accident (LOCA) vould not be effected even if the seal injection vere in operation prior to the start of the event

~

Since charging pump flov is not credited during a large break LOCA event, diversion of charging pump flov to the seal injection system vould not adversely affect this event.

For e small break LOCA, injection flov is credited from e single charging pump.

Zf seal injection vere in operation prior to e small break LOCA, charging flov vould still be delivered to the RCS at the same rate assumed in the safety analysis even though the injection point into the RCS has changed.

Additionally, total Refueling Meter Tank (RMT) inventory usage vould not be affected by seal injection since the RCP controlled bleedoff rate is maintained at 1 gallon per minute for each pump vitb or vithout seal injection.

..PCM 028-284 2 ~

The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created be"-ause:

a)

The seal injection system is isolated during normal 'plant operation.

b) c)

'Ihe seal injectior system vill not adversely affee t RCS chemi s try control.

Since the seal injectior syster uses diverted floe from the charging

pumps, the vater added to the RCS by the seal injection system vill be chemically identical to that added by the CVCS.

Inadvertent boron dilution analyzed in Chapter 15 of the FSAR vill bound the nev configuration.

Inadvertent opening of the RCP seal injection isolation valve V2185 vill not result in a change in RCS inventory.

Total RCS make up vill remain the same even though part is diverted through the seal injection system.

Diversion of charging flov to the seal injection system will result in insufficient cooling of the letdown flov at the regenerative heat exchanger.

Letdo~ will be isolated automatically on high 1etdovn flow temperature to protect the CVCS equipment.

d)

~ Failure to isolate the system from the RCS is not considered credible because tvo check valves are used in series.

3.

The margin for safety as defined in the basis for any technical specification is not reduced because:

a)

The seal injection system is isolated from the RCS and CVCS during normal operation and, therefore, it does not affect parameters controlled by the technical specification.

The above considerations demonstrate that operation qf the seal injection system does not involve an unrevieved safety question or require a technical specification change.

The implementation of this PC.". does not require a change to the plant technical specifications.

The foregoing constitutes, per 10CRF50.59(b),

the vritten safety evaluation vhich provides the bases that this change does not involve an unrev'iewed safety question and prior Com=ission approval for the implementation of this PCH is not required'

PCM 029-284 CABLE-LOFT HVAC SYST IhTROD'JCTIOh This PC/M Supplement provides for the installation of a forced ventilation system for the Cable-Loft Area. (RAB elev. 28.67'),

located above the false ceiling of the'adio Chemistry Labora-

tory, RAB elev. 19.50',

column lines RAH-RAI.

The Cable-Loft Area was part of Main Corridor Area and is now separated from it by' Thermolag partition, in order to comply with 10CFR50, Appendix R, Section III G.

This PCM Supplement incorporates the installation of four new fire damper assemblies furnished with the new design concept of built-in thermal expansion clearance provided at the damper/

expansion frame interface.

The ventilation for the Cable-Loft Area is designed to limit the temperature of the area to 104oF based on an outside air temper-ature of 93 F.

o SAFETY ANALYSIS This modification has been reviewed with respect to Title 10 of the Code of Federal Regulations Part 50.59 which states that a proposed change shall be deemed to involve an unreviewed safety question:

if the possibility of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increasedi or if a possiblity for an accident or malfunction of a differ-ent type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any tech-nical specification is reduced.

(i)

By supplying air from the nuclear-safety related system 2HVS-4h/4B, the area is provided with ven-tilation air in both normal and accident plant operation.

Therefore, the installation of a vent-ilation system for the new Cable-Loft Area does not increase the probability of occurrence of an acc'i-dent previously evaluated in thr Final Safety Analy-sis Report.

Lik wise, the existence of fire dampers in the HVAC system does not increase the possibility of occurrence of a previously evaluated

accident, since damper function is not required to alleviate design basis accidents.

PCM 029-2S4 (ii)

The possibility for an accident or malfuction of a differnt type than any previously evaluated in the Safety Analysis Report is not created since the ad-dition of the ventilation system to the Cable-Loft Area has been performed in accordance with criteria set forth in FSAR Chapter 9 for the entire Reactor Auxiliary Building.

Likewise, the fire dampers in the ventilation system do not create the possibility for an accident of a different type than any evalu-ated in the Safety Analysis Report since the damper is a passive device which performs its functions only on the occurrence of a sufficiently severe fire, which melts the fusible link.

Zn addition to this a fire damper is not required to function during any design ba is accident.

(iii)

The provision of a ventilation system with fire dam-pere in penetrations through fire barriers does not reduce the marcin of safety def'ned in the bases fo Tec".=.ical Specification 3/4.7.12 (Fire Rated Asser><<

lies) or any other Technical Specifica ion.

Thi PC/K doe not involve any change to the Technical Specifica.ions incorporatec in the license.

The forgo'ng cons itutes, pe" 10CFP50.59(b),

the written sa ety eval-uation whic.. provide the ba is t'pat thi change does not include an urrev'ewe" safety ques ion and prio" Commission approval for the implementa.ion of this PC/N is not required.

PCM 038-284 HEAT TRACING SETPOINT MODIFICATIONS INTRODUCTION The standard setpoints for the majority of the Boric Acid Heat Tracing System are listed as follows:

Primary beater high temperature alarm Primary heater maintenance temperature Primary heater low temperature alarm 185'F 170'F 162'F Redundant heater high temperature alarm Redundant heater maintenance temperature Redundant heater low temperature alarm 185'F 155'F 148'F However, the setpoints for the circuits associated with the following systems/equipment deviate from the standard setpoints:

A.

The Boric Acid and Radwaste Concentrators B.

The Charging By-Pass Lines C.

The Radiation Monitoring System (RMS)

D.

The Post Accident Sampling System (PASS)

Therefore, in order to eliminate nuisance alarms in the Control Room and at the Heat Trace Control Panels, this PC/M provides the correct set-points for these systems/equipment.

SAFETY ANALYSIS Qith respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety ques-tion:

(i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

A.

The Boric Acid and Radwaste Concentrators Boric Acid Concentrators 2A 6 2B and the Radwaste Concentrator are non-safety related.

Only the high temperature alarms are being reset from 185'F to 260'F.

This will only eliminate the high temperature nuisance alarm and will not adversely effect any functions of the heat tracing system.

PCM 038-284 B.

The Char in Bv-Pass Lines The Charging By-Pass Lines are safety-related.

Only the low temperature alarms are being reset from 162'F and 148'F to 50'F as described in Section XZ.B.

As discussed with the plant operators (Q Pierce, FPL-NPS), solidification of the boric acid in these lines will not inhibit any safety functions of any equipment.

Also, the operators could detect blockage in these lines by the altered flow path which would result.

Furthermore, the operators make daily temperature checks on these lines to reconfirm process temperature and proper operation of the heat tracing.

C.

The Radiation Nonitorinc System (R.fS)

The lines for the RMS which are heat traced, are sa ety related.

The temperature setpoints are being reset per attached BCS drawings, in order to control these lines at the proper temperature and not at tne same temperature as the boric acid piping system.

D.

The Post Accident Sam lin System (PASS)

The lines for the PASS which are heat traced are non-safety related.

The temperature setpoints are being reset per attached BCS drawings in order to control these lines at the proper temperature and not at the same temperature as the boric acid piping system.

The foregoing constitutes,'er 10CFR50.59(b),

the written safety evaluation which provides the basis that this change does not involve an unreviewed safety question; therefore, prior Commission approval is not required for implementation of this PC/M.

PCM 071-284 THERMOCOUPLE CABLE CHANGEOUT INTRODUCTION The presently installed thermocouple extension cable for eight (8) thermocouples associated with the Shield Building Ventilation System (HVE-6A 6 6B), are "qualified for interim operation" until the first refueling.

Therefore, this PC/M provides the details to change out the existing cable with class IE qualified cable.

SAFETY ANALYSIS With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety ques-tion:

(i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The modifications presented by this PC/M will upgrade the thermocouple extension cable for the Shield Building Ventilation System from cable which is "qualified for interim operation" to cable which is "class IE qualified."

The original design and function of the system remain unchanged.

Tne replacement cable has been provided by CVI Corporation per the o igina'urchase Order

(ÃY-422743).

The cable has been qualified in Actor. Lab kepor: No. 17414 to IEEE 323-1974.

The typical vertical flame tes:

as described in IEEK 383 was not addressed in the report.

However., since the, cables are in dedicated conduit from end to end, this is not considered to be a problem.

The Technical Specifications do not require a revision as a result of this PC/M.

The foregoing constitutes, per 10CFR50.59(b),

the written safety evaluation which provides the basis that this change does not involve an unreviewed safety question; therefore, prior Commission approval is not required for implementation of this PC/M.

<<PCM 072-984 CONDENSER PIT. PLOODING INTRODUCTION During periods of heavy rain the condenser pits in both units have flooded.

The standing water leve1 in both the east and west pits has been as high as elevation +13 ft (C)

One of the contributing factors causing the flooding is that the overf1cnr provisions from the two existing storm water basins shown on drawing 8770-G-687> Rev 1 were not constructed as designed.

With the ditches and overflow basin properly installed, the water level in the storm water~asins would probably not exceed e1evation +7 ft.

however, without these provisions, the water level can rise to about elevation + 14 ft causing the storm drainage system to back-up into low points in the power block area.

SAFETY ANALYSIS With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a

proposed change shall be deemed to involve an unreviewed safety question; (i) if the probabi1ity of occurence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

With the implementation of this PCM the capacity of the storm water drainage system wt11 be substantially increased.

This wi11 prevent backflooding in any power block area and consequently eliminate any damaging effect from a heavy rainstorm.

This change will be such that it will not compromise the integrity of the plant security system.

The proposed changes do not involve a change to the technical specification.

The foregoing constitutes, per 10CFR50.59(b),

the crritten safety evaluation which provides the basis that this change does not involve any unreviewed safety questions, therefore prior Commission approval is not required for implementation of this PCM.

PCM 077-284 PRESSURIZER HEATER BUS DOOR LOUVERS The Pressurizer Heater Load Centers 2A3 and 2B3 presently exceed their temperature specifications vhen all heater banks are utilized.

Spurious trips result due to inadequate heat transfer away from the MCC components.

This PC/M replaces the present solid compartment doors vith louvered doors to provide increased ventilation.

Safety Evaluation This PC/M is classified non-nuclear safety related and does not involve an unrevie~ed safety question.

The pressurizer heaters are non-safety related and this modification does not affect the operation of any equipment.

Greater reliability is achieved due to the decrease in operating temperature.

Safe Anal sis The pressurizer heaters are considered non safety related (FSAR Fig.

8.3-2A).

Modifications are being made only to increase ventilation through the Pressurizer Heater Load Centers.

This modification does not change the operation of any equipment and has no affect on any safety related equipment or operations.

The probability of occurrence or the'onsequences of a design base accident or malfunction of safety related equipment previously evaluated in the FSAR has not been increased.

There is no possibility of an accident or malfunction different than those previously evaluated.

Based on the above this PC/M is non-safety related and does not involve an unreviewed safety question.

PCM 098-284 SA.PL G SYSTEM CONTAINMENT ISOLATION VE REPLACEMENT INTRODUCTION In the cours course of tne St Lucie Unit No.

2 Environmental Qualification

Program, three (3)

Hoke pneumatic operated Containment Isolation Valves the Sampling System supplied by Combustion Engineering (V-5200, ante II II V-5201 and V-5202) were found to have inadequate Harsh environment Equipment ~alification documentation meeting NUREG"0588 Rev.

1 and IEEE-323-1974 requirements for the Micro-Switch limit switches.

These valves, which are normally closed, are located inside the Containment Building and are required to provide Containment Isolation on CIS signal during LOCA/MSLB "Harsh" environmental conditions.

At the time of qualification review, a JIO (10CFR50.49 Equipment List Documentation Package 3.3) was written to justify the use of these valves for interim operation until the first refueling outage.

This PC/M is replacing those interimly qualified valves with fully "Qualified" solenoid operated Target Rock Corp (TRC) valves.

SAFETY ANALYSIS Mich respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety quest@on; (i) if the probability of occurrences or the consequences of an accident or malfunction or equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a difierent type tnan any evaluated previously xn the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis ior any technical specification is reduced.

Tnxs proposed change included within this PC/M does not involve an unreviewed safety question because:

i) The probability of occurrence of an accident has not been increased since the sampling, valves are not utilized in determining the probability of an accident.

The consequences of an accident have not been increased since the sampling valve replacement will not affect operability of the sampling system-The probability of malfunction of the sampling valves has been reduced by this modification since the valve replacement will provide more reliable operation during normal and DBA conditions.

The conse'quences of malfunction of the sampling valves have not been increased since the failure mode of the replacement valves is the same as the original valves.

ii) The replacement of the sampling valves will not create the possibility of an accident of a different type than previously analyzed because the operational characteristics of the sampling system has not been altered.

The possibility of a malfunction of a different type than previously analyzed has not been increased for the reason given (i) above.

PCM 098-284 iii) The margin of safety as defined. in the basis for any Technical Specification has not been reduced.

The mode.fications within this PC/M have been designed with respect to

10CFR50, Appendices A, B, J and R, utilizing the recommendations of the applicable Regulatory Guides 1.11, 1.26, 1.29, 1.68, 1.73, 1.75, 1.89, 1.97, 1.120, Branch Technical Positions PSB-1, ICSB-17, NUREG-0588 Rev.

1 and Industry Codes and Standards.

Specifically, the replacement valves have been designed to ASME Section III, Class 2 (Quality Group B) and seismic Category I requirements and meet the recommendations of Reg.

Guide 1.26 "Quality Group Classifications and Standards for Water,

Steam, and Radioactive Waste containing components of Nuclear Power Plants",

and Reg.

Guide 1.29; "Seismic'esign Classification".

Additionally, the replacement valves meet the requirements set forth in FSAR Sections 3.2, 3.10, 3.11 6.2.4 ana 7.3.

The mechanical ana electrical portions of this PC/M have been designed to meet their anticipated St Lucie Unit 2 environmental and operational conditions and meet the requirements-of IEEE<<323-1974, IEEE-344"1975 and IEEE-382-1972.

The implementation of this modification does not require a change to the Plant Technical Specifications.

All the relevant parameters of Technical Specification Tables 3.6-1 and 3.6-2 remain unchanged.

The foregoing constitutes, per 10CFR50.59(b),

the written safety evaluation which provides the basis that this change does not involve an unreviewed safety question and prior commission approval for the implementation of this PC/M is not required..

PCM 150-984 QSPDS INVERTER/POWERLINE CONDITIONER MODIFICATION INTRODUCTION The Powerline Conditioners are part of the Inadequate Core Cooling (ICC) system power supply.

There are 2 power supply systems lA and 1B for Unit 1 and 2A and 2B for Unit 2.

The power supply systems are composed of a 7.5 KVA inverter and a 25 KVA Powerline Conditioner.

The Powerline Conditioner is the By-pass source used for the maintenance of the 7.5 KVA inverters.

Aft'er the installation of the ICC power supply systems, the fuses protecting the current transformer in the Powerline Conditioners were blowing out during power transfers (ie. transfer to off-site power or diesel generator).

SAFETY ANALYSIS This modification has been reviewed with respect to Title 10 of the Code of Federal Regulations, Part 50.59, which states that a proposed change shall be deemed to involve an unreviewed safety question:

(i) if the probability of occurrence or the consequences of an accident or m+function of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an ac-cident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

Modification covered in this PC/M requires the replacement of resistors only and does not degrade the system nor affect the internal integrity or quali-fication of the Powerline Conditioner.

The installation of the class 1E, 68 ohm resistor in the Units 1 and 2

Powerline Conditioners enhances the operation of these equipment by increasing the reliability during power transients.

This PC/M neither poses the possibility for a new or different accident or malfunction as defined in the safety analysis report nor it affects the Plant Technical Specification, as written.

The foregoing constitutes, per 10CFR50.59 (b), the written safety evaluation which provides the basis that this change does not involve, an unreviewed safety question; therefore, prior Commission approval is not required'or implementation of this PC/M.

PCM 176-284 MODIFICATION OF THE SPENT FUEL HANDLING TOOL

==

Introduction:==

'Ihe bottom tool asserbly and the grapple assenbly [it~ 16

& 19 re-spectively of Reference (B) and (C) ) am to be modified to facilitate fuel handling ape ations and to insure positive loc)cog of a fuel asserrbly to the spent fuel hanQ3iag tool.

Lead-in surfaces are to be provided an the bottom tool asserrbly to permit the spent fuel handling tool to be guided through the portable Fuel Iaading Funnels that. are installed on tcp of the Fuel Storage Racks.

'Ihe grapple assembly is to be rrndified to increase the length of the fuel seating surfaces.

This rradificatian will insure that the Spent Fuel Handling Tool rerrains locked to a fuel asserrbly during handling operations.

With respect to Title 10 of the Code o Federal Regulatians, Part 50.59, a proposed change shall be deme to involve an unreviewed safety question:

'i) if the polity of occru~nce or the consequences of an accident or nalfunction of equiprrant important to the safety previously evaluated in the safety analysis re~rt may be in~ed; or (ii) if a possibility for an accident or malfrrrcmon of a different type than any evaluated previously in the safety analysis report may be created-or (iii) if the margin of safety as defined in the basis for any tec!nical specification is reduced.

~ficatians to the Spent Fuel Handling Tool are consistent with the original design specification reauirere~ts for this tool.

The tool na8ifications are not part of a safety relate'ystem, therefore, a safety analysis is not required.

T.".e iml~tation of ~s PC/2 does not reo~

a c!age in the pl--.t tee?ical refications.

PCM 225-284 t

t CEDM TEMPERATURE SMITCH SUPPORT MODIFICATIONS INTRODUCTION This PCM provides a bolted splice detail for the CEDM temperature switch (TS-25-18 and

19) supports which were cut to allow removal of the reactor seal ring.

SAFETY ANALYSIS Mith respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previous-ly in the Safety Analysis Report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is re-duced.

The CEDM temperature switches TS-25-18 and 19 serve no safety related func-tion.

However, in order to preclude possible interaction with safety related equipment below during a

seismic

event, the switch supports and the new connections have been seismically designed.

Therefore implementation of this PCM will not increase the probability of any accident previously evaluated.

Implementation of this PCM does not involve a change to the St Lucie Unit 2 Technical Specifications.

The foregoing constitutes, per 10CFR50.59(b),

the written safety evaluation which provides the basis that this change does not involve any unreviewed safety

question, therefore prior Commission approval is not required for the implementation of this PCM.

PCM 226-284 APPENDIX R PIRE DAMPERS INSTALLATIONMODIFICATIONS ZYTRODUCTION This plant change/modification, PCM 226-284, is for the replacement and/or modification of some of the existing Appendix "R" Fire Damper installation to provide for the thermal expansion clearances required by 10CFR50 hppendix "R"~

The replacement of some of the existing Fire Dampers vith Fire Damper/Expansion Frame assemblies shall provide the required thermal expansion clearances for these penetrations

~

Modifications to come penetretions to adjust the clearances betveen the sleeve/duct and the fire barrier shall provide the required thermal clearances for these penetrations.

The plant change/modification is also for the replacement of some Fire Damper Sleeves to provide minimum sleeve gage thickness for these penetrations, as required by 10CFR50 Appendix "R".

Nev Fire Damper hssemblies are UL listed and are equipped vith limit cvitches.

Modifications to existing installations to meet minimum gage requirements and thermal expansion clearance requirements are in accordance vith UL555 and vendor 's instructions, as required by 10CFR50 Appendix "R", for installation in the ductvork penetrating the designated Appendix "R" Fire Barriers separating areas containing equipment required for cafe shutdovn of the Plant.

The methodology of thermal clearances adjustments shall be determined during construction and is not included in this PCM.

Generic solutions are provided vithin this package as veil as thermal clearance requirements.

SAFETY ANALYSIS Qith respect to Ti tie 10 of the Code of Federal Regulations, Part 50.59 a proposed change chall be deemed to involve an unrevieved cafety question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the cafety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii)if the margin of safety as defined in the basis for any technical cpecification is reduced.

Installation of the safety-related and non safety related fire dampers does not increase the probability of occurrence of an accident previously evaluated in the Final Safety Analysis Report, but rather, enhances the plant safety by providing asurance that a fire does not cause the accidental loss of equipment required for safe plant chutdovn

~

The closing of fire dampers installed in the air duct and air transfer opening penetrations through fire barriers separating various fire areas, enhances the plant safety, in the event of severe fire in a fire area.

Closed fire dampers prevent the progression of fire across the fire barriers, from the fire affected area to other adjacent areas vhere essential equipment for the cafe chutdovn of the plant may be located'

PCM 226-284 The closing of the fire dampers caused by a severe fire in an area may cause the loss of ducted supply air to other area(s) located downstream of the affected area and/or the loss of ducted exhaust air from other area(s) located upstream of the affected area.

Potential adverse effects, of heat build-up due to loss of supply or exhaust air to adjacent areas housing essential safe shutaown equipment, are averted by operator action to establish alternate ventilation flow paths to these

areas, for duration of fire in the affected area and until the restoration of the ventilation systems

~

Previously evaluated accident consequences are not increased by the fire dampers since, the fire dampers are not required during the spectre of desi@; basis accidents.

Previously evaluated malfunctions of equipment important to safety (i.e., FM:-A') are no: increased since damper function is not required to alleviate design basis accidents, and moreover the fire damper is considered a passive device.

ii - The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety hnalysis Report is not created since the damper is a passive device which performs its function only on the occurence of a sufficiently severe fire which melts the fusible link.

h fire damper is not required to function during any design basis accident.

iii-Provision of fire dampers does not reduce the margin of safety as defined in the basis for any Technical Specification.

Instead, the provision of fire dampers substantially increases the margin of safety defined in the bases for Technical Specification 3/4.7.12, "Penetration Fire Barriers" since the dampers will help to ensure the functional integrity of the fire barrier penetrations so that fires are confined or adequately retarded from spreading to adjacent portions of the facility~

The implementation of this PCH does not require a change to plant technical specifications.

The foregoing constitutes, per 10 CFR50.59 (b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior Commission approval for the implementation of this PCM is not required

~

PCM 253-984 NaOCL GAS DISENGAGING TANK BLOWER REMOVAL SYSTEM DESCRIPTION 1.0

~equation:

Operation of the sodium hypochlorite system will be affected only in that the hydrogen blowers willnot be installed and therefore operation of their associated switches will no longer be part of the startup sequence.

Additionally, all automatic interlocks between the blowers and the remainder of the system have been removed.

Therefore, the remaining portions of the system willoperate as normal.

2.0 Function This modification functions to physically remove both gas disengaging tank blowers and functionally removes all switches, alarm lights and interlocks associated with these blowers.

3.0 Desi n Descri tion This modification provides for the complete physical removal of the gas disengaging tank blowers and provides for installation of a protective screen on top of the tank cover.

Additionally, this modification provides for removing all interlocks associated with 'hese blowers and the remainder of the system.

SAFETY ANALYSIS 1.0 This modification has been reviewed with respect to 10 CFR 50.59 and has been deemed not to involve any unreviewed safety question because of the following:

1.1 The Sodium Hypochlorite System is non-safety'elated, non-seismic and does not perform any function related to plant safety.

1.2 This modification involves removal of the gas disengaging tank blowers which are not required when the system is located outdoors.

Therefore, the original design intent of the system does not change.

1.3 These modifications do not interact with any safety related systems or components.

1.4 No safety related systems or components are compromised by any assumed failure of any existing or new equipment or components.

1.5 No parameters relating to Technical Specifications are adversely affected and no Technical Specifications are altered by this modification.

2.0 This. constitutes the safety evaluation according to 10 CFR 50.59.

Therefore, prior Commission approval is not required prior to implementation of this modification.

PCM 028-285 CONCENTRATOR HI-HI LEVEL FEED PUMP TRIP MODIFICATIONS INTRODUCTION At present~

the Concentrator Feed Pumps will not trip on hi-hx level when both concentrators A & B are on line.

This PC/M will modify the present design to provide a Hi-Hi level t,rip signal from either Concentrator A or B.

This modification is required to prevent boron carry-over into the vapor separator, vapor condenser and distillate system.

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

This modification is for the disconnection and reconnection of wiring between non-safety/non seismic relay contacts.

This change will permit the tripping of the feed pumps from either Concentrator A or B Hl-Hl level.

Therefore this modification will not increase the probability of the occurrence of any accidents whether previously evaluated or of a different type than previously evaluated and will not reduce the safety of the plant.

The foregoing constitutes, per 10CFR50.59(b),

the written safety evaluation which provides the basis that this change does not involve an unreviewed safety question, therefore prior Commission approval is not required for implementation of this PCM.

This PCM does not reduce'the margin of safety as defined in the basis of any technical specification, nor does it require a revision of a technical specification.

PCM 053-285 CEDMCS POWER SUPPLY REPLACEMENT 5anma~

This moci ication provides fo" the re. lacement o.

CEDMCS, Logic, Lamp and relay D.C. voltage pover supplies vhich are no longer available from the-original eouipment m nufacturer.

Safety Analysis a.

The orobabili-v o~ o"currency of an ac=ident nrevio ~sly evaluated in the F

AF, has not be

r. a:=e ted si ce the CcDvCS is not u ilize in determin'ng the probabil'ties of accidents.

b.

The consequences of an accident previously evaluated in the FSAR have not been chan ed since th's modification does no affect the availability, redundancy,

capacity, or function o any equipment required to mitigate the effects of an accident.

c.

This modification does

. not affect any other safety related equipment.

d.

The conseouences of the malfunction of eouioment important to safety reviouslv evaluated in'he FSAR have not been affected.

Redundancy, changed.

function, or failure mode capacity have not been e.

The possibility of an accidnet of a different tvoe tha-,.

analyzed in the FSAR has not been created since this modification does not affect any systems vital to safety or vhose failure could directly'esult in an uncontrolled release of radioactive material.

f.

The possibility of equipment malfunction of a different type tha..

analyzed in the FSAR has not been increased.

The margin of safety as defined in the basis for any Techn cal Specification has no been changed since this modification does not change the performance, capabilities, cr operating characteristics o! the CEDMCS.

In conclusion this change does not involve an unrevieved safety question.

PCM 085-285 INTAKE COOLING WATER PUMP EXPANSION JOINT SYSTEM DESCRIPTION 1.0 Function The intake cooling water pumps provide c'ooling water from the intake structure to the CCN heat exchangers and the TCÃ heat exchangers.

Expansion joints between the pump nozzle and the piping system reduce stresses on the nozzle imposed by the movements of the piping system.

Although they are not necessary to meet code design stresses, the expansion joints reduce vibration induced stresses and ease reassembly of the system.

2.0 Desi Descri tion This PC/M changes the bellows material from Monel (ASME-SB-127) to Inconel 625 (ASME-SB-443).

The liner material is changed fron 316 SS to Inconel 625.

The attachment of the liner and bellows is changed from a welded design to a Van Stone fiange design.

These design changes are made to reduce the current rate of corrosion exhibited by the existing expansion joints.

Inconel 625 is superior with respect to corrosion fatigue

strength, pitting and crevice corrosion resistance when compared with Monel.

The new expansion joints should have a significantly longer service life than the existing Monel expansion joints.

3.0

~Oeratien This modification willnot affect the normal operating procedures or specific operation of the ICN pumps.

4.0 Precautions and Limits r

a r 3 ~

After implementation of this modification, all weld repair procedures" for Inconel 625 shall have engineering concurrence.

~

a

~

SAFETY ANALYSIS This change does not represent an unreviewed safety question" since it does not affect any accident addressed in the FSAR, present any new accident not previously analyzed in the FSAR, nor does it affect the margin of safety for any technical specification.

The operation of the intake cooling water pumps or the piping system has not been affected by the use of an alternate material as specified in this PC/M package, as this alternate material is equal to or better than the original material in all aspects.

Therefore, this material change does not increase the probabilities or consequences of accidents or equipment malfunction important to the safety of the plant previously evaluated in the FSAR.

,. 'L'af

~ g f

PCM 101-285 MA FEEDWATER ISOLATION VALVE SR A CHECK VALVES S stem Descri tion 1.0 Functi<rn The air check valve in the pneumatic supply line of the main feedwater isolation valve actuator closes upon loss of instrument air, preserving a supply of air in the air reservoir for actuation of the hydraulic control valves.

This ensures the operability of the main feedwater isolation valve during loss of instrument air conditions.

2.0 Desi n Descri tion This PC/M provides for the replacement of the actuator air check valve with one of modified design which has been demonstrated to be more reliable.

Anchor/Darling, the manufacturer of the MFWIV actuator and supplier of the instrument air check valve, has informed customers of a potential problem with the check valves.

This particular check valve has failed to seat in slow loss of instrument air conditions at several plants with A/D hydraulic actuators.

Safe The replacement valve for this PC/M is the one recommended by Anchor/Darling.

This modified air check valve is improved over the previous design.

It has a guided poppet and a heavier spring to ensure seating of the poppet under very low reverse flowconditions.

1 a.

The probability of occurrence of accidents previously addressed in the FSAR is not changed by the replacement of the check valve.

The new check valve is similar in design and function to the existing valve.

The modifications to the internal components of the valve serve to increase reliability only.

lb.

lc.

ld.

le.

The consequences of accidents previously evaluated in the FSAR are not increased.

The new check valve helps to ensure the actuator's ability to close the MFWIV and, thus, perform the intended safety function.

The probability of malfunction of equipment important to safety previously addressed in the FSAR is not increased.

The new check valves enhance the reliability of the MFWIV actuators.

For the same reasons as those described in 1 a.,

the consequences of malfunction of equipment important to safety previously evaluated in the FSAR are unchanged.

For the same reasons as those described in la., the possibility of new accidents not considered in the FSAR is not created.

lg.

For the same reasons as those described in la., the possibility of malfunctions of a different type than those analyzed in the FSAR is not created.

The margin of safety defined in the basis for the Technical Specifications is not affected.

The ability of the MFWIVs to close within the time required by T.S.0.7.1A willbe verified by testing upon completion of this PC/M.

Based on the above discussions, it can therefore be concluded that this modification to the MFWIVactuator does not pose an unreviewed safety question.

PCM 112-285 INTRODUCTION ESFAS TEST GROUP ASSIGNMENT REVISIONS The Engineered Safety Features System (ESFAS) circuitry includes the redun-dant initiating variable measurement

devices, trip bistables, the coinci-dence logic matrices, actuation modules, output relays, manual and automatic test circuitry.

The testing criteria is as follows:

IEEE 338-1971, "Criteria for the Periodic Testing of Nuclear Generating Station Protection Systems",

and Regulatory Guide 1.22,

Periodic Testing of Protection Systems Actuation Functions",

(RO) provides guidance 'or development of procedures, equipment and documentation of periodic testing.

Test intervals and their bases are included in the Technical Specifications.

Since operation of the ESFAS is not expected, the systems are periodically tested to verify operability.

The system is tested fromthe sensor signal through the actuation devices.

Complete channels can be individually tested without initiating protective action, without violating the single failure criterion and without inhibiting the operation of the systems.

Those actu-ated

devices, which are not tested during reactor operation (e.g.

main feedwater isolation valves),

are tested during scheduled reactor shutdown to assure that they are capable of performing the necessary functions.

FSAR Table 7.3-9 lists all actuated devices not tested from ESFAS during normal operation.

FSAR Table 7.3-9a lists the ESFAS actuation relay devices not tested during normal operating conditions.

FSAR Table 7.3-2 identifies the Test Group assignments for components actuated on a

Safety Injection Actuation Signal (SIAS).

Operational testing of th% group output relays is accomplished by individu-ally selecting one group (refer to FSAR Table 7.3-2 for the test group assignment).

All test groups are tested during power operation except for Group 0.

Group 0 is tested during shutdown.

For

example, when SIAS Test Group lA is tested the LPSI pump 2A startec but LPSI discharge valves remain'closed since they are assigned to Test Group 2A.

This overlapping test method causes the ESF components'to actuate; therefore the propagation

~ of a valid trip during testing is not impeded

'and the ESF system proceeds to full actuation.

According to the vendor Instruction Manual (Emdrac number 2998-15662),

a Permissive Test Panel in each safety cabinet allows the tsting of the output relays and the associated plant safety equipment by selecting one actuation module for activation.

The Permissive Test switch selects which of the sev'en functions is to be tested by applying a positive voltage to the Group Test Switch wipers.

The Group Test Switch selects which one of the actuation modules will be activated when the module Test switch is depressed, thus activating the selected group of output relays.

For example, if the Group 1 output relays for the SIAS containment pressure channel are to be tested, the Permissive Test Switch is set to the SIAS Cont Press position (switch contact 3) and the Group Test Switch is set to the Group 1 position (switch contact 2).

Under these conditions, only the containment pressure circuit of activation module AM501 can be activated when the module Test switch is depressed, thus activating the 2/4 logic.

The activated 2/4 logic energizes the SIAS output relays associated with AM501 only.

PCM 112-285 SAFETY ANALYSIS This modification has been revised with respect to Title 10 of the Code of Federal Regulations, Part 50.59, which states that a

proposed change shall be deemed to involve an unreviewed safety questions; (i) if the proba-bility of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increasedl or (ii) if a possibility for an accident or mal-function of a different type than any evaluated previously in the safety analysis report may be created',

or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The ES'FAS and ESFAS Testability is addressed in FSAR Sect'on 7.3.

VCT Discharge V-2501 testability i presently addressed in FSAR Table 7.3-9a.

The relay devices associated with pressurizer heater 2A3 41G'reaker and non essential lead which are not tested during normal plant operation are addressed as an attachment to th's PC/M.

This PCM does not affect the criteria established in the FSAR.

The tables have been revised to reflect the change in load groups'his change does not affect the ability of the ESFAS to perform its intended safety function.

No additional equ'pment/cable is added by this PC/M only jumper is required.

Therefore, the probabil$ty of a

previous'y reviewed accident is not in-

creased, the poss'ility of an accident of a differen.

type has no.

been created and the margin of safety has not been reduced.

The imp'enenta:ion of this PC/M does not require a change to the p'ant technica'pec'icat'on.

The foregoing constitutes, per 10CFR50.59(b),

the wr'tten safety evalua o"..

wh'h provides the basis that th' change does not involve any unreviewed safety

question, therefore prior Commission approval is not requ'red for implementation of this PC/M.

PCM 114-985 SMALL CENTRIFUGAL PUMPS OIL SEAL REPLACEMENT ABSTRACT This engineering package covers replacement of lip type oil seals on small centr.'fugal pumps throughout Units /Il and I/2 with labyrinth type seals.

The major feature of this package provides a list of acceptable labyrinth seals and a list cf pumps approved for seal replacement.

This package is classified as non-nuclear safety related for all pumps listed with the exception of the Boric Acid Makeup (BAM) and Fuel Pool pumps on both Unit Nl and //2 and the Diesel Oil Transfer pumps on Unit 2 which are classifed as nuclear safety related.

Modif!cations to these pumps, however, do not constitute an unreviewed safety question.

SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a

proposed change shall be deemed to involve an unreviewed safety question: (1) if the probability of occurence or the conse'quences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased; or {ii)if a possibility for an accident or malfunction of a different type than any evaluated pteviously in the Safety Analysis Report may be created; or (iii)if the margin of safety as defined in the bases for any technical specifictions is reduced.

The subject modifiction provides for replacement of lip type oil seals with labyrinth oil seals in small centrifugal pumps.

These seals are located in the bearing housings of those pumps Identified in this document and do not affect any pressure boundary portions of the pumps.

The use of the labyrinth seals provides a superior sealing system in that the stationary and rotating members do not touch as with the existing lip seals.

Therefore, the increased likelihood for leakage caused by rubbing of the shaft with subsequent localized shaft wear does not occur.

Thus the improved design significantly reduces the likelihood of oil leakage along the pump shaft and pump reliability is increased.

Therefore with respect to 10CFR 50.59 the use of the labyrinth type seals; does not increase the probability of an accident or malfunction of equipment important to safety, does not create possible accident scena! ios not previously addressed by the Safety Analysis Report, and does not affect or require changes to the Technical Specifications.

The foregoing constitutes, per 10 CFR 50.59 {b), the written safety evaluation which provides the basis that this change does not involve an unreviewed safety question.

Therefore, prior commission approval is not req'uired for implementation of this PC/N.

PCM 116-985 t

DEMINERALIZER WASTE NEUTRALIZ~N SYSTEM INTRODUCTION Currently the Water Treatment Plant (WTP) demineralizer regeneration wastes are transferred to the neutralization basin vhere they are neutralized and discharged.

The Resource Conservation and Recovery Act (RCRA) requires that vastes with a pH below 2 or above 12.5 must be handled, stored and disposed of as hazardous wastes.

Certain steps in the regeneration process create a vastestream outside pH limitations of 2 or 12.5

~

Rather than seek exemption from the RCRA permit requirements it has been decided to install a neutralization tank.

Portions of demineralizer regeneration

~astes vhich create vastestreams vith a pH value outside the limitations vill be routed to this tank.

All other portions vill be routed to the existing neutralization basin as in the current practice

~

To expedite installation, this package is being issued in multi-supplements.

Supplement 0 deals with the tank foundation and buried piping.

Supplement 1 vill deal vith the piping, Supports/

Restraints, Electrical, I&C and Civil portions.

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

The water treatment system is a common system for both units and is discussed in the PSL-1 FSAR Section 9.2.5 as part of the makeup vater system.

This task modifies the demineralizer wastes drain lines and also adds a neutralization tank.

The modification is done to comply vith the requirements of RCRA.

The vater treatment system does not interface with any safety related

system, therefore, it is non-safety, non-seismic.

This sub-system does not impact any safety related component.

The WTP is not considered in any accident analysis in the FSAR nor

'ill this modification create an accident or malfunction not previously evaluated in the PSAR.

As a non-safety, non-seismic system it will not reduce the margin of safety as defined in the bases for any technical specifications.

The implementation of this PCM does not require a change to the plant technical specifications.

The foregoing constitutes, per 10CFR 50.59 (b), the written safety evaluation which provides the bases that this change does not involve an unreviewed aatety question and prior Commission approval for the implementation of this PCM is not required.

PCM 132-285 REPLACEMENT OP RIS DEVICES INTRODUCTION This PC/M is for the installation of one (1) new transmitter by Rochester Instrument Systems Model SC-1302-323 to replace an existing unit.

The existing transmitter is reaching its qualified life expectancy.

Therefore, a

new replacement unit is required to satisfy the life expectancy require-ment.

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

This modification does not involve an unreviewed safety question and the following provides the bases for this conclusion.

This new transmitter by RIS is qualified environmentally to IEEE-323-1974 and seismically to IEEE-344-1975.

This PC/M replaces an existing RIS trans-mitter with a

new unit, thus satisfying the life expectancy requirements.

The seismic qualification of the new transmitter has been reviewed and found acceptable for the mounting location in the RTGB.

The seismic integrity of the RTGB is not affected since the device is a replacement for an existing device at the same location/mounting and weight.

Therefore, this modification will not increase the probability of the occur-rence of any

accident, whether previously evaluated of a different type than previousl evaluated and will not reduce the safety of the plant.

This PC/M does not reduce the margin of safety as defined in the basis of any technical specification.

The implementation of this PC/m does not require a

change to the plant technical specification.

The foregoing constitutes, per 10CFR50.59(b),

the written safety evalua-tion which provides the bases that this change does not involve an unreviewed safety

question, and prior Commission approval for the implementation of this PC/M is not required.

PCM 178-985 TIE BETWEEN CONSTRUCTION FIRE MAIN AND'PLANT FIRE LOOP Abstract This Plant Change/Modification is for the connection of the Beckfit Construction Fire Main to the St Lucie Units I and 2 Fire Water Loop.

This connection consists of two separate tie-ins between the fire main anrf the fire loop.

This PCM ie not classified as Safety Related since the fire main and the fire loop do not perform any safety function.

Since the fire loop provides protection for safety related equipment, this PCM is classified as Quality Related.

This PCM provides additional fire protection to the plant since these tie-Vns create an additional fire water supply to other portions of the Plant.

SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a

proposed change shall be deemed to involve an unrev1ewed safety question; (1) if the probability of occurence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (11) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (111) if the margin of safety as defined in the basis for any technical specification is reduced.

The proposed modification affects fire protection 11ne 12-FP-3, which functions to distr1bute a supply of water throughout the St Lucie site to protect nuclear safety related and nonnuclear safety related equipment from the consequences of a fire.

Because the affected equipment is not considered by the FSAR in determining the probability of accidents or possible types of acc1dents or in the evaluation of the consequences of acc1dents, it can be concluded that the probabilty of occurence of accidents previously addressed in the FSAR is unchanged, the possibility of new accidents not considered in the FSAR is not

created, and the consequences of accidents previously evaluated in the FSAR are unchanged.

Line 12-FP-3 provides protection for nuclear safety related equipment and is therefore quality related.

The new installations are quality related from their tie-ins with 12-FP-3 up to and including their isolation valves.

The proposed design of these sections is in compliance with the applicable codes and FSAR requirements for all fire protection equipment.

The potential failure mode of this system and degree of protection provided to nuclear safety related equipment is therefore unchanged.

The probability of malfunction of equ1pment important to safety, previously evaluated 1n the FSAR is thus unchanged, the consequences of malfunction of equipment important to safety previously evaluated in the FSAR are unchanged, and the possibility of malfunctions of a different type than those analyzed in the FSAR is not created.

The new area which will be serviced by the tie-ins to 12-FP-3 is the Backfit Construction area.

This area does not contain any nuclear safety related components or any equipment that is considered in the basis of the technical specifications.

The proposed design includes the capability to isolate the nonnuclear safety related section from the quality related section.

No technical specification changes will be required and no margin of safety defined therein is affected.

Based on the above discussion, it can therefore be concluded that the implementation of quality related PCH 178-985 will not create an unreviewed safety question.

The foregoing constitutes, per 10 CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior Commission approval for the implementation of this PCM is not required.

PCM 199-985 WATER TREATMENT PLANT REGENERATION WASTE NEUTRALIZATION TANK MODIFICATION BOOSTER PUMP ABSTRACT The subject REA requested a neutralization tank be added to the Water Treatment Plant (WTP) to meet current Department of Environmental Regulation (DER) regulations governing discharge of hazardous wastes.

The neutralization tank modification (PC/M 116-985) provides the necessary details for installation of this tank and the associated

piping, equipment and components necessary to allow for regeneration wastes to be automaticaily directed to the tank during the appropriate times in the regeneration process.

During the caustic injection steps of regeneration, caustic solutions must be directed to the tank.

The existing system,

however, is unable to provide the necessary flows and pressures required to accommodate these regeneration steps due to the additional headloss in the new piping runs.

Thus, to accommodate the new arrangement, a booster pump must be added to the caustic dilution water demineralized water supply.

In addition, the caustic dilution water fiow control valve and flow indicator/transmitter must be replaced to accommodate the flow requirements.

This system is not required for plant safe shutdown, therefore this modification is non-nuclear safety related and its implementation does not create an unreviewed safety question.

SAFETY EVALUATION The subject modification provides for'dd1tion of a booster pump and flow control valve in the caustic dilution water supply to the WTP.

In addition, the modification provides for replacement of certain caustic dilution water flow transmitter components to accomodate the required flow rates.

As defined in Section 9 of the Unit 1 FSAR, the WTP and its associated systems are classified as nonnuclear safety related and are not required to perform a safety function.

Based on the failure mode analysis, as addressed in the Design Analysis, the modification has no affect on nuclear safety.

Therefore, the modification is adequately classified as Non-Nuclear Safety Related Quality Group D.

Based on the above evaluation and informat1on. supplied in the design analys1s, it can be demonstrated that an unreviewed safety question as defined by 10CFR 50.59 is not created.

Since the mod1fication affects only the WTP which is classified as Non-Nuclear Safety Related and cannot affect any other safety related equipment or components as addressed the failure mode analysis, the consequences of all analyzed accidents remains unchanged.

Also, with respect to nuclear safety, no new accidents or malfunctions are introduced as a result of this design change.

Additionally, the margin of safety as defined in the Techn1cal Specifications has not been reduced.

Therefore, an unreviewed safety question does not exist.

Since this modification does not involve an unreviewed safety question, nor require a change to the Technical Specifications, this modification is acceptable with respect to nuclear safety thus prior NRC approval is not required for implementation of the modification.

PCM 210-285 INSTRUMENT AIR UPGRADE TIE-INS ABSTRACT This engineering package covers all outage related tie-ins for the Instrument Air (IA) Upgrade.

The IA Upgrade modification willconsist of installation of two new 1009t'ompressors and provide for replacement of the existing dryer with new refrigerant and desiccant dryers.

Accordingly, this modification provides all necessary tie-ins for the new components and consists of those tie-ins required for Turbine Cooling Water (TCW) and (IA) air piping. Allelectrical tie-ins are capable of being implemented during plant operation and are not part of this modification.

This modification is considered non-nuclear safety related and does not constitute an unreviewed safety question.

SAFETY EVALUATION r

The subject modification provides for installation of piping tie-ins to the TCN and IA Systems.

As defined in Section 9 of the FSAR, these systems are considered nonnuclear safety related Quality Group D and are not required to perform a safety function.

Additionally, based on the failure mode analysis as addressed in the Design Analysis this modification has no affect on nuclear safety.

Therefore, this modification is adequately classified as nonnuclear safety related Quality Group D.

Based on the above evaluation and information suppliec in the design analysis, it car, be demonstrated that an urreviewec safety questior, as defined by 10CFH 50.59 does not exist.

Since the modification; affects only the TC4 and 1.-; Systems which are classified as nonnuclear safety related, affects only those portions of the system located in the turbine building, away from any safety related equipment or components such that these compnents cannot fall on or hit such safety related eouipment or components, and does not affect the operability ot'ny safety related equipment or component, the consequences of all analyzed accidents remain unchanged.

Therefore, with respect to nuclear safety no new accidents or malfunctions are introduced as a result of this design change.

Additionally, the margin of safety as defined in the Technical Specifications has not been reduced.

Therefore, an unreviewed safety question docs not exist.

In summary, since this modifcation does not involve an unreviewed safety question nor requires a

change to the Technical Specifications and is acceptable from the standpoint of nuclear safety, prior NRC approval is not required to implement this change.

The foregoing constitutes, per 10 CFR 50.59 (b), the written safety evaluation which provices the basis that this change does not involve an unreviewed safety question.

Therefor e, prior commission approval is not required for implementation of this PC/.'1.

PCM 015-986 TELEPHONE SYSTEM UPGRADE hBSTRACT This Engineering Package covers the modifications and details required to support the installation (by ATILT) of a nev ATILT System 85 PBX Telephone System.

The central equipment for System 85 vill be loca ed in the Telecommunication Equipment Rooms in the Unit 1 Service Building and Unit 2 D-13 Building.

The modifications and details consist of enlargement of the telecommunication rooms to accommodate the new equipment; installation of redundant air conditioner units for each room to satisfy equipment environmental requirements; pover supplies with emergency back"up; racevay between the tvo telecommunication rooms to install the ATdT supplied fiberoptics cable, and raceway betveen the D-13 Building, G-3 Building and Start-up Trailers to accommodate the ATILT supplied multipair telephone cables.

Based on the importance of the telephone system as one of the plant communication means, this Engineering Package has been classified Quality Related to enhance the system design and installation confidence.

The nev "System 85" vill replace the existing Dimension 600 Electroni:

Stored Program PBX located in the Unit 1 Service Building Telecommunications Room and the Private Automatic Telephone Exchange (PAX) located in Unit 1 Reactor Auxiliary Building (Elev 43'-0).

Tnis replacement vill require modification of Section 9.5.2 "Communication Systems" of the Unit 1 and Unit 2 FUSAR, Figures 9.5-1 and 9.5"4, Table 9.5"6 of the Unit 2 FUSAR and Section 3.8 of the Unit 1 and Unit 2 Nuclear Fire Protection Program.

To energize the System 85 telephone equipment and air conditioners located in the Unit g Service Building upon loss of normal off-site power will require manual svitching at Pover Panel PP-135 located in the Security and Records Building.

Resetting vill also be required upon returning of normal off-site power.

Tne Sys em 85 telephone equipment modules and air conditioners located in the Unit 2 D-13 Building vill be automatically supplied by the Non-Class lE diesel supplying the D-13 Building upon loss"of normal pover.

PCM 015-986 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed'to involve an unreviewed safety question; (i) if the probability of occurrence of the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a

different type tnan any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is redu'ced.

The telephone system is not a safety related system.

The replacement of the ATILT Dimension 600 System and the PAX System with a new System 85 has no impa"t on any plant systems and/or operations except for the replacement of the black intraplant telephone sets (PAX) with standard telephone sets to perform the same function.

Tne new System 85 increases the capacity and reliability of the telephone system to allow for future telephone growth within the plant.

The modifications to the telephone equipment rooms in the Unit 1

Service Building and Unit 2 D-13 Building are not safety related and do not interfere with any plant system.

Routing of the fiberoptics cable was designed in dedicated

conduit, along the west side of the Turbine Building, does not require any seismic support and does not interfere with any safety related equipment.

Power supplies are from non Class lE power panels.

Tne non-emergency section of PP-135 requires manual actuation and resetting upon loss of normal AC power.

The new System 85 is independent from the Emergency Notification System.

The installation of the new System 85 has an impa"t on Se"tion 9.5.2 of the Units 1 and 2 FUSAR, Figures 9.5-1 and 9.5-4, Table 9.5-6 of the Unit 2 FUSAR and Section 3.8 of the Units 1 and 2 Nuclear Fire Protection Program.

Revision to the above sections will be required during the update of the Unit 1 and Unit 2 FUSAR.

The implementation of this PC/M does not require a change to the plant Technical Specifications.

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

the written safety evaluation which provides the basis that this change does not involve any'nreviewed safety question, therefore prior Commission approval is not required for implementaton of this PC/M.

PCM 016-286 RCP MOTOR ARRD BACKSTOP PIN DESIGN ENHANCEMENTS ABSTRACT This engineering package covers design enhancements and installation guidelines for the Unit 2 reactor coolant pump (RCP) assemblies Anti-Reverse Rotation Device (ARRD) back-stop pins.

The intent of the design enhancements and installation guidelines is to mitigate the potential for the backstop pins to mushroom and become stuck in their respective retaining ring(s).

The major features of this package are:

1.

Document the manufacturers design changes to the ARRD back-stop pins.

2.

Emphasize the importance of proper installation of the back-stop pin access plugs.

These modifications are classified as quality related and do not constitute an unreviewed safety question.

SAFETY EVALUATION The Unit 2 FSAR Section 5.0 discusses the design bases of the RCPs and RCP motors.

The RCP motors function to provide motive force for the pumps during normal plant operation and provide inertia to improve coastdown characteristics during a loss of pump power condition.

The motors also contain an anti-reverse rotation device to preclude reverse rotation caused by backflow through the pump impeller.

Potential causes of backflow considered are:

(1) Reversed power leads (2) Loss of power to one RCP, with others operating (3) and RCP suction line break; LOCA.

The greatest challenge to the anti-reverse rotation device (ARRD) is the reverse torque due to a large break LOCA (depicted in C.E. Spec.

13172-PE-180, Rev 05, Figure 12). Combustion Engineering in enclosure 1 to F-CE-8706 has analyzed the ARRD in regards to the reverse torque from LOCA loads.

Their analysis shows that a minimum of four equally spaced ARRD pins per motor of SA 193 B7 material are adequate to withstand the maximum specified reverse torque.

The motor manufacturer's material change to Reyerson steel EDT 150 increases the tensile strength of the pins by approximately 25,000 psi, and the chamfering at edge has made the pins less susceptible to "mushrooming," (see FPL letter EPO-86-0268),

and therefore the ARRD's ability to function as required has been enhanced by a greater margin of safety in the design.

e PCM 016-286 The RCP motor manufacturer, Siemens-Allis, has evaluated the proposed modifications and has concluded that the capability of the motor to perform its function will not be affected.

Since the normal operation of the RCP motor is not affected by these modifications, the probability of occurrence of a

design basis accident or malfunction of equipment important to safety as previously evaluated in the FSAR will not be increased.

The modifications do not change the postulated failure modes of the RCP motors which would result in loss of reactor coolant flow. The consequences of a design basis accident or malfunction of equipment important to safety as previously evaluated in the FSAR will therefore not be increased.

The possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR willnot be created.

In conclusion, the changes proposed in this design package are acceptable from the standpoint of nuclear safety; do not involve an unreviewed safety question; do not require NRC approval prior to implementation.

PCM 017-286 RCP MOTOR OIL SEALS MODIPICATIONS ABSTRACT I

This engineering package covers design enhancements and installation guidelines for the Unit 2'eactor Coolant Pump (RCP) motor lower bearing oil sealing assemblies.

The intent of the design enhancements and installation guidelines is to mitigate the potential for the oil sealing assemblies to fail as occurred in August 1985 to 2A2 RCP. - (Reference PCM 127-285)

The major feature of this package is the modification of the oil slinger assemblies on RCPs 2A1, 2B1, 2B2.

These modifications are classified as quality related and do not constitute an unreviewed safety question.

SAFETY EVALUATION The Unit 2 FSAR Section 5.0 discusses the design bases of the RCP's and RCP motors.

The RCP motors function to provide motive force for the pumps during normal plant operation and provide inertia to improve coastdown characteristics during a loss of pump power condition.

The motors also contain an anti-reverse rotation device to preclude reverse rotation caused by backflow through the pump impeller.

The proposed modifications do not (1) change the design intent as described in the FSAR or (2) adversely affect the operation or motor qualification provided by Siemens-Allis.

No safety analysis discussed in Chapter 15 of the Unit 2 FSAR is affected.

There is no motor/drive to pump/driven interface affected by these proposed modifications and, therefore, the safety related aspects of the pumps pressure boundary are unaffected.

The probability of occurrence or the consequences of a design basis accident or malfunciton of equipment important to safety as previously evaluated in the FSAR is therefore not increased, and the possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR will not be created.

The margin of safety as defined in the basis for a Technical Specification has not been reduced.

This PC/M is considered not safety related QA/QC required.

The foregoing constitutes, per 10 CFR 50.59 (b), the written safety evaluation which provides the basis that this change does not involute an unreviewed safety question and prior Commission approval for the implementation of this PC/M is not required.

e pcM 036-286 TURBINE THRUST BEARING ADDITION,OF OIL SEALS ABSTRACT Due to excessive oil leakage on the LP turbine at St.

Lucie Unit 2, an investigation was performed by the site Westinghouse representative.

Based upon his review, Westinghouse developed and fabricated a supplemental oil seal to be used in conjunCtion with the primary oil seal (i.e., outer sea9.

This new seal was then installed between the thrust bearing and existing outer seal on the LP turbine rotor.

To support this modification, a 10CFR50.59 review was completed and the respective safety analysis which is now part of this package was prepared to verify that a safety concern was not created.

This design package functions to endorse the use of the additional seal and provide for adequate document updating.

This modification is considered Non-Nuclear Safety Related and does not create an unreviewed safety question.

SAFETY EVALUATION The Unit 2 No.

2 LP turbine is located in a Non-Nuclear Safety Related system and as such is not required to function during any existing analyzed accident scenario.

Therefore, modifications to the LP turbine affects only Non-Nuclear Safety. Related Qxality Group D equipment.

Based on a failure mode analysis, failure of the labyrinth seal ring added by this modification will not inhibit the operation of any existing safety related equipment or components.

The new seal ring is internal to the turbine bearing housing and as such cannot fall onto or hit any safety related equipment or components.

Based on the failure mode analysis, failure of the oil seal can occur from improper installation or development of improper clearances resulting from equipment vibration or rubbing.

In all cases, failure will result in an inability of the seal to perform its sealing function.

Failure of the internal seal would require appropriate sealing from the outer seal which is part of the original design.

Therefore failure of the new seal with subsequent oil leakage will not affect any safety related equipment.

Also, this extra seal will not increase the probability or consequences of the turbine missile analysis in the FSAR.

Based on this information, it can be demonstrated that an unreviewed safety question as defined by 10CFR50.59 does not exist since the consequences of all analyzed accidents remains unchanged.

Additionally, with respect to nuclear

safety, no new accidents or malfunctions are introduced as a result of using the additional seal ring.

Finally, the margin of safety as defined in the Technical Specifications has not been reduced nor have changes to the Technical Specifications been required.

In conclusion, this modification is acceptable from the standpoint of nuclear safety since it does not involve an unreviewed safety question and does not require changes to the Technical Specifications.

Thus implementation of this modification does not require prior NRC approval.

PCM 054-286 TORQUE SEATING-ADV ISOLATION VALVES ABSTRACT This Engineering Package movers a modification to the isolation valves (MV-08-14, 15, 16, 8

17) for the PSL' Atmospheric Dump Valves (ADV's). This modification changes the control circuit for these valves to allow the closing direction to be limited by the torque switch instead of by the limit switch as presently designed.

This PCM is classified Is Nuclear Safety Related, and does not constitute an unreviewed safety question.

The atmospheric dump valves (ADV's) provide a

means of decay heat removal and cooldown capability when the MSlV's are closed.

The ADV's can also be modulated to control primary plant temperature during startup and shutdown.

Each ADV has an upstream DC motor operated isolation (block) valve which is normally maintained in the "locked open" position via a key operated switch in the control room.

The ADV isolation valves serve to isolate flow through a stuck open or leaking ADV.

The valve manufacturer, Anchor/Darling, was consulted and concurs that these valves n ay be torque seated.

Although the Anchor/Darling letter (attachment

3) references one valve serial number, all four ADV isolation valves were supplied under the same purchase order and are identical in construction and application.

The proposed change ls therefore acceptable for all four ADV isolation valves.

The Llmltorque Instruction and Maintenance Manual (Bulletin SMB1-180-A) also states that the closing direction for wedge type gate valves is typically limited by the torque switch.

The proposed design, although different than the original, does not change the operation of the atmospheric dump system as discussed in the PSL Unit 2 FSAR Sections 5.4, 6.3, and 10.3.

Since the ability of the ADV isolation valves to close and isolate a stuck open ADV has not been adversely affected by this change, the probability of occurrence or the consequences of a design basis accident or malfunction of equipment important to safety as discussed in the FSAR Chapter 15 has not been increased.

Previously analyzed failure modes for the ADV isolation valve remain valid, and thus the possibility for an accident or malfunction of a differenct type than any evaluated previously in the FSAR Chapter 15 is not created.

PSL 2 Technical Specificiations Section 3.7.1.7 provides the "Limiting Condition for Operation" for the ADV's and associated isolation (block) valves.

The Technical Specification requirements are not changed by this modification and the margin of safety as defined in the bases for this Technical Speclficiation will not be reduced.

The safety evaluation thus demonstrates that an unreviewed safety question does not exist.

In conclusion, the change proposed in this design package is acceptable from the standpoint of nuclear safety; does not involve an unreviewed safety question; and does not require NRC approval prior to implementation.

PCM 067-286 REPLACEMENT OF VALVE V3740 INTRODUCTION The solenoid vent valve, V37~0, on Safety Injection Tank 2B2 was found to be inoperable.

The failed valve was manufactured by Garrett P eumatic System Division and a direct replacement is not available

~ It has been decided to replace the failed Garrett valve Model No 373>003 n

with a Target Rock valve Model No.

80B-001 S/N 5.

SAFETY ANALYSIS Pith respect to Title 10 of the Code of Federal Regulations, Part 50.590 a proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurrence or the

onsequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created; or (iii) if the margin of safety as defined in the basis for any techni=al specification is reduced'his modification replaces a failed valve manufactured by Garrett

'neumatic Systems Division for which a direct replacement is not available.

The failed valve is being replaced by a Target Rock Valve of similar charac teri s t ic s.

The Safety Injection Tanks and vent valves are discussed in FSAR Sections 6.3.2.2.1 and 5.4.7.5.

The replaced valve is lighter in weight than the existing valve.

This change will not impact the existing, stress analysis and design of supports.

The modifications includel in the PCM do not involve an unreviewed safety question because:

i-The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since the failed valve has been replaced with a ASME-III Class 2 valve.

The solenoid valve has been evaluated for the environment that it vill be subjected to and found to be acceptable.

Seismic qualification of similar valves has been evaluated and found to be acceptable.

Documentation to this effect is included in Do Pac 35.6.

PCM 067-286 ii-There is no possibility for an accident or malfunction of a different type than any evaluated since the failed valve has been replaced vitn an acceptable ASNE-III Glass 2 valve and the system function or capability has not been modified iii-This modification does not change the margin of safety as defined in the basis for any technical specification.

The implementation of this PCM does not require a change to the plant technical specifications.

The foregoing constitutes, per 10CFR50.59(b),

the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior Commission approval for the implementation of this PCM is not required.

PCM 072-286 FHB HVAC PENETRATION BARRIER ABSTRACT It has been determined that steel barriers are required for the two (2)

HVAC penetrations located at elevation 48 ft of the Fuel. Handling Building (west exterior wall).

Tne barriers are required in order to prevent unauthorized access into the FHB.

Both HVAC penetrations are protected by a

continuous L-shaped concrete tornado missile barrier, cantilevered two (2) feet from the FHB exterior wall and extending down to approximately five (5) feet below the bottom of the penetration.

For ease of, construction, the access barriers will be located at the bottom of the two (2) foot opening which exists between the missile barrier and the FHB exterior wali'he existing HVAC system has not been affected by this modification.

Based on the review of the existing HVAC system, a 40K reduction of the missile barrier opening is acceptable.

As a result of the addition of the access

barriers, the missile barrier openings have been reduced by

'nly 17K.

Failure of the access barriers will not adversely affect the function of any safetymelated systems or components.

However, since the barriers are being installed in a

tornado missile barrier and the FHB exteiior

wall, this PCM has been classified as Quality Related.

This modification does not affect the structural capability of the missile barrier or the FHB wall, nor does it pose any safety hazards.

This PCM does not constitute an unreviewed safety question and has no effect on plant safety.

The addition of the access barriers has no impact on plant operation and does not affect any safety related equipments SAFETY EVALUATION With respect to title 10 of the Code of Federal Regulations, Part 50.59, a

proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to.

safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The modifications included in this PCM do not involve an unreviewed safety question because:

The probability of occurence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since:

a - Tne failure of the access barriers for the two (2)

H~'AC penetrations located at elevation 4c ft of the Fuel Handling Building will not adversely affect the structural capacity of the missile barrier nor the FHB wall, for which certain quality control inspections (e.g. hole size and verification that no rebar is cut) will be performed.

b - No effect on equipment or components performing a safety function are located beneatn this access barrier.

c - Tne HVAC ventilation system operation has not been affected by the reduction in the missile barrier opening.

Tnere is no possibility for an accident or malfunction of a different type than any previously evaluated since this modification will nave no impact on the plant safe snutdown.

This modification does not change the margin of safety as defined in the basis for any Technical Specification by the addition of these access barriers.

There is no change on the existing technical specification due to the implementation of this PCA.

The foregoing constitutes, per 10CFR50.59(b),

the written safety evaluation which provides the basis that this change does not involve an unreviewed safety

question, therefore prior Coaanission approval is not required for implementation of this PCH.

PCM 078-286 ENVIRONMENTAL QUALIFICATION UPDATE ABSTRACT This engineering design package provides the vehicle for updating several areas of equipment qualification.

This package includes corrections to the 10CFR50.49 list, changes in maintenance requi'rements, and various documentation corrections.

This design package is considered nuclear safety related because it affects equipment falling under the scope of 10CFR50.49.

This package does not represent an unreviewed safety question since it deals strictly with enhanceing the'resent documentation used to qualify equipment at St. Lucie.

Safety Evaluation This engineering design package provides for several documentation changes to the present St. Lucie Unit No. 2's equipment qualification documentation.

This documentation wQl affect the future procurement of various safety related components and assist in validating the components'bQity to function during a design basis accident.

Therefore, this design package is considered safety related.

This PC/M is the vehicle for the following documentation changes:

1)

Removal of equipment from the scope of 10CFRSO-49-removal is based on (A) the equipment is not required to function in a harsh environment to mitigate or monitor the consequences of a design basis accident; and (B) faQure of the equipment wQl not interfere with the operation of any safety related equipment.

2)

Changes to existing documentation - changes are based on the analyses contained herein.and affect maintenance requirements, tag

, number inconsistencies, correction of model numbers on the 10CFR50.49 Est and correction of Doc. Pac. deficiencies.

None of the changes require physical modifications to any plant system.

The probabQity of malfunction of equipment important to safety has not been increased nor has an equipment malfunction of a different type than previously analyzed been created.

The surveQlance requirements of the Technical Specifications were reviewed against the equipment qualification maintenance requirements addressed in this package in the design analysis.

No Technical Specification changes are required by this design package.

In conclusion, the changes proposed by this design package are acceptable from the standpoint of nuclear safety because they do not involve an unreviewed safety question and no Technical Specification changes are requirede

PCM 102-286 REGULATORY GUIDE 1.75 ABSTRACT In accordance with FPL commitments to the NRC to satisfy the requirements of Reg Guide 1.75 Rev 1, a second (redundant) fault current interrupting device shall be installed in series with the existing breaker on all non safety loads fed from an essential section of a safety MCC.

This second fault current interrupting device will protect the safety bus in case of failure of the existing qualified breaker that feeds a non class 1E load.

The original design bases for implementation of Reg Guide 1.75 (PCM 015-283) excluded the requirements for double isolation for valve MOV V2504 and Diesel Generator Compressor Motors 2A and 2B.

Further investigation has determined that this equipment now requires a redundant (second) isolation device.'dditionally, Reg Guide 1.75 requires that non-Class 1E wiring should be separated from Class 1E wiring by the minimum separation

distance, or by a barrier.

During the design walkdown phase of this EDP, two (2) safety cables were fouad to be routed in the non Class 1E wireways without adequate separation as required by Reg Guide 1.75 (Rev 2).

This Engineering Design Package (EDP) covers 'the modifications and details necessary'o bring these loads into conformance with Reg Guide 1.75 requirements by the addition of redundant fault'current interrupting devices and flexible metal conduit over the Class lE cables where they pass through.

aon-Class 1E wireways.

A review of the changes to be implemented by this PC/M was performed against the requirements of 10CFR50.59.

As indicated in Section 3.0 of this EDP, this PC/M does aot involve an unreviewed safety question, nor does it require a

revision to the Technical Specification; therefore, prior Com 'ssion approval is not required for implementation of this EDP.

This PCM is safety related in that the equipment to be modified, MCC 2A-7, MCC 2B-7, and MCC 2B-5, is safety related

PCM 102-286 With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (i) I.f the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased, or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii)if the margin of safety as defined in the bases for any technical specification is reduced.

The plant modifications described in this PCM are to provide:

(1) double isolation for NOU V2504 and Diesel Generator Air Compressor Motors 2A and 2B in accordance with Reg Guide 1.75 (Rev 2); (2) installation of flexible metal conduit over Class 1E cables 21123C-SA and 21133C-SB where they pass through the non-Class 1E section of their respective NCC's 2A-7 and 2B-7 ~

These modifications bring the plant design for these loads into conformance with Regulatory Guide 1.75 (Rev 2) as described in PSAR Section 8.3.1.2.2.

Por the installation of the second redundant isolation device, Regulatory Guide 1.75 requires that all circuit components (1E circuit breakers) be treated as safety related up to the downstream side of the second fault current interrupting device, or adequate 5ustification be provided.

Pursuant to this requirement, the fault current Interrupting devices added by this PCM are Class lE (safety related).

They are seismically mounted and qualified.

I MOU V-2504 Evaluation Due to physical constraints in MCC 2B-5, the cables associated with the second fault current interrupting devi.ces are the same safety classification as the original feeder cables (non-safety).

Thus in the case with MOU V2504 the cable between the primary and backup fault current interrupting devices'or this non-safety load fed from a safety source is routed as non-safety.

Justification for this is provided below:

l.

'All cables (safety or non"safety) are Class lE (IEEE-383-1974) qualified for both vendor as well as Ebasco supplied cable.

2 ~

All cables connected to the (safety related)

MCCs are routed through seismically supported raceway regardless of the fact that the load is non-class 1E.

3 ~

Regulatory Guide 1.75 concerns itself with the fact that the non-Class lE loads may fail during a seismic or other DBA and threaten an upstream breaker.

The non-Class 1E load is always downstream of the second fault current interrupting devices.

4 ~

The cable between the primary fault current Interrupting device and the backup device is kept as short in length as possible.

PCM 102-286 MOV V250 nd MCC 2B-5 appear on the Essen al Equipment list and are noted as such on the Pire Protection Checklist (Attachment 7.2).

This modification has no impact on the operation of the valve.

The additional cables routed between the primary fault current interrupting device and the redundant fault current interrupting device will be added to the Essential Cable Lis't in that they are associated with the operation of MOV V2504.

Diesel Generator Com ressor Motors 2A and 2B Evaluation The Diesel Generator Air Compressor Motors 2A and 2B were determined to be non-Class 1E components as a result of the Environmental Qualification Study for the Emergency Diesel Generator Sets (Reference Ebasco letter P-M-SL-85-0489).

This results in the need to provide a

redundant isolation device and the reclassification of the respective circuits to "associated" in accordance with the Reg Guide 1.75.

By definition, non Class 1E circuits that share power supplies, enclosures or raceways with Class lE circuits or are not physically separated from Class lK circuits or equipment shall be designated as associated circuits.

Since the power and control circuits as well as the circuits associated with the new fault current interrupting devices are routed in the MCC wireways, with Class lE circuits, they will be redesignated as associated circuits and will comply with the requirements of such circuits.

One of the requirements is that they be separated from non"Class 1E circuits.

In the non safety MCC wireway where these circuits are partially routed, a flexible metal conduit will be installed to provide for separation from the non safety circuits.

Cables 21123C-SA and 21133C-SB Evaluation These cables feed the Class 1E Turbo Lube Oil AC Pumps for the EDG.

This modifification only provides for the installation of flexible metal conduit over the cables where they are routed with non-Class lE cables in their respective MCC wireways.

This will satisfy the separation requirements for Class lE cables routed with non-Class 1E cables as required by Reg Guide 1.75 (Rev 2) such that the failure of the non-Class lE cables will not adversely affect the Class 1E cables.

The modification has no impact on the operation of the components.

Conclusions The addition of these isolation devices and the flexible conduit will prevent non-safety loads from interrupting the safe operation of Class lE components.

This dual isolation and addition of flexible metal conduit to satisfy separation criteria conforms to Reg Guide 1.75 Rev 2 and will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety.

This PCN does not create a possibility for an accident or malfunction of a different type than previously evaluated.

The margin of safety as defined in the bases for any technical specification is not reduced.

The implementation of this PQi does not require a change to the plant technical specifications.

The foregoing constitutes, per 10CFR 50.59, the written safety evaluation which provides the bases that this change does not involve any unreviewed safety questions, and prior commission approval for the implementation of this PCM is not required.

PCM 110-286 TURBINE LUBE OIL/HYDROGEN SEAL OIL ABSTRACT This Engineering Package covers a control circuit modification to the (1) turbine lube oil'eservoir and conditioner, (2) turbine lube oil piping and (3) hydrogen seal oil unit deluge systems on Unit 2 to allow for periodic testing of the release solenoid with the system isolation valve closed.

Upon completion of this modification, the operation of these deluge systems will be consistent with the other plant deluge systems.

Since the affected fire protection systems do not affect safety related equipment, this engineering package is classified as no~afety related.

The proposed modification to the turbine lube oil/hydrogen seal oil deluge systems will allow for actuation of the release solenoid during testing with the system isolation valve closed.

Presently, with the isolation valve closed, a tamper switch opens the control circuit to the release solenoid thus preventing the solenoid's operation.

As demonstrated by the failure modes analysis in the design

analysis, the subject deluge systems are located in the turbine building and have no affect on safety related equipment.

Since equipment important to safety or equipment used to mitigate accidents is not affected by the proposed

design, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR has not been increased, and the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR has not been created.

The subject deluge systems are not addressed in the Technical Specifications.

The proposed change will therefore not reduce the margin of safety as defined in the basis for any Technical Specification.

10CFR50.59 allows changes to a facility as described in the FSAR if an unreviewed safety question does not exist and if a change to the Technical Specifications is not required.

As shown in the preceding sections, the change proposed by this design package does not involve an unreviewed'safety question because each concern posed by 10CFR50.59 that pertains to an unreviewed safety question can be positively answered.

Also, no Technical Specification changes are required.

n In conclusion, the change proposed in this design package is acceptable from the standpoint of nuclear safety, does not involve an unreviewed safety question, and does not require NRC approval prior to implementation.

PCM 112-986 TURBINE BUILDING CRANE GIRDER CONNECTION ANGLE MODIPICATIONS

Recently, cracking and ezcessive prying deformation vere noted at some of the crane girder connections in the laydown area between the Vait 1

and 2 turbine buildings.

ha evaluation of the condition concluded that the failures vere attributable to the inability of the connections at column liae 20 to slide as originally designed.

This PC/M will provide modifications to the crane girder connections at column line 20 to restore independent thermal movement between the unite.

Modifications will also be implemented at the other crane girder connections in the laydown

, area to provide reinforcement for those connections which may have been sub)ected to overstress as a result of the thermal restraint of the crane girders.

This PC/H does aot constitute aa unreviewed safety question and hae no effect on plant 'safety.

The turbine buildings are classified as nonnuclear safety related structures and therefore the modification does not affect any safety related equipment.

The connection modifications have no impact on plant operation ezcept for restrictions on the movement of the turbine gantry cranes while the modifications are in progress.

The turbine buildings have been designed for seismic loading to preclude interaction with ad)scent Seismic Category I structures during a seismic event.

Consequently, this PC/M is classified as Quali.ty Related.

Safet Anal sis With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a

proposed change shall be deemed to involve an unreviewed safety question; (1) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if-the margin of safety as defined in the basis for any technical specification is reduced.

This PC/M provides modifications to the crane girder connections in the laydown area between the Un1t l and Unit 2 turbine buildings to restore 1ndependent thermal movement between the units. It does not involve an unreviewed safety question.

The following are the bases for this conclusion:

(1)

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased s1nce this modification will be performed in accordance with Quality Related requirements, hence the seismic capability of the turbine buildings is not compromised and there can be no impact on any ad)acent Seismic Category I

structures,

systems, or equipment.

(11)

There is no possibility for an accident or malfunction of a

different type than any evaluated previously since the turbine buildings are non-safety related structures containing no safetymelated equipment, hence this modification can have no impact on any safety-related system.

(iii)

This modification does not change the margin of safety as defined in the basis foi any technical specification.

The implementation of this PC/M does not requ1re a

change to plant technical specifications.

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

the written safety evaluation which provides the bases that 'this change does not involve an unreviewed safety question and prior Comm1ssion approval for the implementation of this PC/M is not required.

PCM 120-286 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION LIST REVISION ABSTRACT This Engineering Package provides the vehicle for updating several 'areas of equipment qualification.

This package includes corrections to the 10CFR50.49 list, changes in maintenance requirements, and various documentation package corrections.

This Engineering Package (EP) is considered Nuclear Safety Related because it affects equipment falling under the scope of 10CFR50.49.

This package does not represent an unreviewed safety question since it deals strictly with enhancing the present documentation used to qualify equipment at St Lucie Unit No 2

and no physical plant modifications are required by the Engineering Package.

The safety evaluation of this package indicates that a change to the Plant Technical Specifications i.s not required.

The equipment removed from the 10CFR50.49 list are listed in Section 2.1.2 of this PCM.

Removal of equipment from the 10CFR50.49 list does not affect plant safety and operation.

SAFETY EVALUATION With respect to Title 10 'of the Code of Federal Regulations, Part 50.59, a

proposed change shall be deemed to involve an unreviewed safety question:

(i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created, or (iii) if the margin of safety as defined in the bases for any technical specification is reduced.

This Engineering Package provides for several changes to the present St Lucie Unit No. 2's 10CFR50.49 list.

This documentation will affect the future procurement of various safety related components--and-~@4,st

-in validating the components'bility to function before, during and after a design basis event.

Therefore, this EP is considered Nuclear Safety Related.

The documentation changes addressed in this package range from corrections of typographical errors on the 10CFR50.49 list to additions and deletions of equipment as a result of EQ documentation packages reviews.

None of the changes require physical modification to any plant system.

They do,

however, affect the future maintenance of various equipment.

The possibility of new Design Basis Events (DBEs) not considered in the FSAR is not created since this change does not alter any equipment used to mitigate accidents.

This modification is an enhancement of the environmental qualification documentation of various equipment and in no way affects the plant design.

Due to the fact that this EP.does not affect or modify any cables essential to safe reactor shutdown or systems associated with achieving and maintaining shutdowns, this package has no impact on 10CFR50 Appendix "R" fire protection requirements.

Therefore the proposed design of this package is in compliance with the applicable codes and PSAR requirements for fire protection equipment.

Since this modification involves no physical modifications to safety related equipment and changes in the maintenance schedules will not result in failure of equipment, the degree of protection provided to Nuclear Safety Related equipment is unchanged.

Removal of equipment from the 10CFR50.49 list does not affect the plant's safety since the equipment being removed has been shown to be installed in a

mild environment or not required to mitigate and monitor the consequences of an accident.

The probability of malfunction of equipment is unchanged.

The probability of malfunction of equipment important to safety previously evaluated in the FSAR remains unchanged.

The consequences of malfunction of equipment important to safety previously evaluated in the FSAR are unchanged.

The possibility of malfunctions of a different type than those analyzed in the FSAR is not created.

Based oa the

above, the modifications included in this Engineering Package do not involve an unreviewed safety question because of the following reasons:

(i)

The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report will not be increased by this modification because it does not affect the availability, redundancy,

capacity, or function of any equipment required to mitigate the effects of an accident.

(ii)

The possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report will not be created by this modification.

Function, mounting and the ability to withstand harsh environmental conditions have not been altered and this modification does not affect any other safety related equipment.

(iii) The margin of safety as defined in the bases for any technical specification is not reduced since this modification does not change the requirements of the Technical Specifications.

The implementation of this PCM does not require a change to the Plant Technical Specifications.

The foregoing constitutes, per 10CFR50.59(b),

the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior Commission approval for the implementation of this PCM is not required.

PCM 130"986 NEUTRALIZATION BASIN CLOSURE MONITORING WELLS ABSTRACT This engineering package covers the installation of two temporary ground water monitoring wells in the vicinity of the St. Lucie Water Treatment Plant.

These wells will be used to demonstrate to the State Department of Environmental Regulation (DER) that the operation of our acid/caustic neutralization basin has not resulted in any ground water contamination.

The temporary monitor wells perform no safety related func'tion and are located away from, and have no effect on, any safety related system.

This PC/M is non-safety related, but has been classified as 'Quality Related'o ensure the wells are located as specified by the enclosed design drawings.

The addition of these wells does not pose an unreviewed safety question.

SAFETY EVALUATION The Neutralization Basin ground water monitoring wells do not perform any plant safety - related function.

They will not be located in the vicinity of any safety - related equipment and therefore well drilling operations cannot adversely impact safety - related functions.

A complete well failure or collapse will not impair the structural integrity of plant fillmaterial; accordingly, safety related structures or equipment supported by the plant fillwill not abc affected.

Based on the above evaluation and the information supplied in the design analyis it can be demonstrated that an unreviewed safety question as defined by 10 CFR 50.59 does not exist.

The probability of occurence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased.

I The possibility of an accident or malfuntion of a different type than any evaluated previously in the safety analysis report has not been created.

The margin of safety as defined in the basis for any Technical Specification has not been reduced.