ML17219A450

Proposed Tech Specs,Incorporating Revised Pressure/Temp Limits & Results of Recent Low Temp Overpressure Protection Analysis

ML17219A450
Person / Time
Site:	Saint Lucie
Issue date:	03/17/1987
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17219A448	List: ML17219A447 ML17219A449 ML17219A450
References
NUDOCS 8703240086
Download: ML17219A450 (60)

Text

ATTACHMENT4 Marked U Technical S ecification Pa es l-4 3/4 I-8 3/4 I -9a (new) 3/4 l-l2 3/4 4-Id 3/4 4-2 I 3/4 4-22 3/4 4-23a (delete) 3/4 4-23a (new) 3/4 4-23b (delete) 3/4 4-23b (new) 3/4 4-23c (delete) 3/4 4-23c (new) 3/4 4-59 (plus insert) 3/4 4-60 3/4 5-7 B 3/4 l-3 (plus insert)

B 3/4 4-I B 3/4 4-6 (plus insert)

B 3/4 4-7 B 3/4 4-8 B 3/4 4-9 B 3/4 4-I0 B 3/4 4- I I (delete)

B 3/4 4-I5 (plus insert)

B 3/4 5-I 8703240086 870317 PDR ADaCX,0SO00aSS P .

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DEF INITIONS IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank, or Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE. or

c. Reactor Coolant System leakage through a steam generator to the-secondary system.

LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE 7iCSEZP5%'f'.16 The LOM TEMPERATURE RCS OVERPRESSURE PROTE ION RANGE is that operating condition when (1) the cold leg temperature is c W%-f and (2) the Reactor Coolant System has pressure boundary integrity. The Reactor Coolant System does not have pressure boundary integrity when the Reactor Coolant System is open to containment and the minimum area of the Reactor Coolant System opening is greater than 1.75 square inches.

MEMBER S OF THE PUBLIC 1.17 MEMBER(S) OF THE PUBLiC shall include all persons nho are not occupation-'lly associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

Thfs category does include persons who use portions of the site for recrea-tional, occupational or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL ODCM 1.18 The OFFS1TE OUSE CALCULATION MANUAL shall contain the current methodology and parameters used fn the calculations of offsfte doses-due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monftoring alarm/trip setpoints, and shall include the Radiological Envfion-mental Sample 'point locations.

ST. LUCIE - UNIT 1 1-4 Amendment No.$ 9.50,69

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REACTIVITY CONTROL SYSTEMS 3 4.1.2 BORATION SYSTEHS FLOW PATHS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths and one associated heat tracing circuit shall be OPERABLE:

a. A flow path from the boric acid makeup tank via either a boric acid pump or a gravity feed connection and charging pump to the Reactor Coolant System if only the boric acid makeup tank in Specification 3.1.2.7a is OPERABLE, or

b. The flow path from the refueling water tank via either a charging pump or a high pressure safety injection pump*

to the Reactor Coolant System if only the refueling water tank in Specification 3.1.2:7b is OPERABLE.

APPLICABILITY: NODES 5 and 6.

ACTION:

With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one injection path is restored to OPERABLE status.

SURVEILLANCE RE UIREHENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days .by:

1. Cycling each testable power operated or automatic valve in the flow path required for boron injection through at least one complete cycle of full travel, and

2. Verifying that the temperature of the heat traced portion of the flow path is above the temperature limit line shown on Figure 3.1-1 when a flow path from the boric acid make-up tanks is used.

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REACTIVITY'ONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING COttDITIOtt FOR OPERATION 3.1.2.3 At least one charging pump or one high pressure safety injection pump* in the boron injection flow path required OPERABLE pursuant to Speci fi-cation 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no charging pump or high pressure safety injection pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one of the required pumps is restored to OPERABLE status.

SURVEILLANCE RE UIREMENTS 4.1.2.3 At least the above required charging pump or high pressure safety injection pump shall be demonstrated OPERABLE at least once per 31 days by:

a. Starting (unless already operating) the pump from the control room,

b. Verifying pump operation for at least 15 minutes, and

c. Verifying that the pump is aligned to receive electrical power from an OPERABLE emergency bus.

the RCS via ~ HPSI pump shall be

,~he flow path from the established only i RMT to d e¹ -~%

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REACTOR COOLANT SYSTI<.",

COLD SXUTOOWt' LOOPS FILLED L 1IITING CONDITION FOR OPERATION 3.4,1.4.1 Qt least one shutdown cooling loop shall be OPEPABLE and in.

operation:. and ei ther:

a. One additional shutdown cooling loop shall be OPERABLE, or

b. The secondary side wa er level of at least two steam generators shall be greater than 10" of narrow range indication.

APPLICABILITY: MODE 5 with reactor coolant loops filled ACTION:

a. With less than the above required loops OPEPABLE or with less than the required steam generator level, within one (1) hour initfate corrective action to return the required loops to OPEfNBLE status or to restore the required level.

b. With no shutdown cooling loop fn operation, suspend all operations involving a reduction fn boron concentration of the Reactor Coolant System and within one (1) hour initiate corrective action to return the required shutdown cooling loop to .operation.

SURVEILLANCE REOUIRPiENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined.to be wfthfn limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4,4.1.4.1.2 At least one shutdown cooling loop shall be determined to be fn operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

~ The Normal or Emergency Power Source may be inoperable for each shutdown cooling loop.

~ The shutdown cooling pump may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided 1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 10'F belo~ saturation temperature.

> One shutdown cooling loop may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other shutdown cooling loop is OPERABLE and fn operation.

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tha 4S-$ above each of the Reactor Coolant System cold leg temperatures.

~H9e7:W F ST LUCIE - UNIT 1 3/4 4-ld Amendment No. 56

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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION

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3.49.1 The React Coolant System (except the pressurizer) temperature and pressure sha be limited in accordance with the limit lines shown cn Figur~+M44 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing.~i'=

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an analysis to determine the effe'cts of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that th.

Reactor Coolant System remains acc ptable for continued operations or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 200'F- and 500 psia, respectively, within tne following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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4 REACTOR COOLANT SYSTEN POWER OPERATED RELIEF VALVES Su'56cr:

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LINITING CONDITION FOR OPERATION 3.4.15 Two power operated relief valves (PORYs) shall be OPERABLE, with their set oints selected to the low temperature of operationg APPLICABILITY: 'ODES c~ ~+ 5, f ~b5.

mode ACTION:

a. With less than two PORVs OPERABLE and while at Hot Standby during a planned cooldown, both PORVs will be returned to OPERABLE status prior to entering the applicable NODE unless:

1; The repairs cannot be accomplished within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the repairs cannot be performed under hot conditions, or

2. Another action statement requires cooldown, or

3. Plant and personnel safety requires cooldown to Cold Shutdown with extreme caution.

b. With less than two PORVs OPERABLE while in COLO SHUTDOWN, .both c.

PORVs will be returned to OPERABLE status prior to startup.

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SURV'EILLANCE RE UIREMENTS 4.4;13 .The PORVs shall be verified OPERABLE by:

" a. Verifying the isolation valves are open when the PORVs are reset to the low temperature mode of operation.

Performance of a CHANNEL FUNCTIONAL TEST of the Reactor Coolant System overpressurization protection system circuitry up to and including the relief valve solenoids once per refueling outage.

c~ Performance of a CHANNEL CALIBRATION of the pressurizer pressure sensing channels once per 18 months.

iReactor Coolant System cold leg temperature below

+ PORVs are cdee not required below 140'F when RCS does not have pressure boundary integrity.

ST. LUCIE - UNIT 1 3/4 4- 59 Amendment No. 6 6

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- STARTING LIMITING CONOITION FOR OPE 3.4.14 If the earn generator temperature exceeds the primary temperature by A

APPLICABILITY: MOOEQ~~.

sar: 9 J ACTION: ~sr: 90 F 0 If a reactor coolant pump is started whe the steam generator temperature exceeds primary temperature by more than evaluate the subsequent transient to determine compliance with Specification 3.4.9.1.

SURVEILLANCE RE VIREMENTS 4.4.14 Prior to Istarting a reactor coolant pump, verify ~@co that the steam generator temperature does not exceed primary temperature by more tha

~gg: 50 F, i5gsser: 954'f, fReactor Coo'lant System Cold Leg Temperature is less than ~~,

ST. LUCIE - VNIT 1 3/4 4-60 Amendment Ho. 6 P

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EMERGENCY CORE COOL IN G S YST EM" ECCS SUBSYSTEMS - T < 325'F LIMITING CONOI T ION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

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a. In MODES 3* and Q one ECCS subsystem composed of one OPERABLE high pressure safety injection pump and one OPEBBLE flow path capable of taking suction from the refueling water storage tank on a safety injection actuation signal and automatically transferring suction to the containment sump on a sump recirculation actuation signal.

b. Prior to decreasing the reactor coolant system temperature below a maximum of only one high pressure safety injection pump fs

f59'F 444'F to be OPERABLE with its associated header stop valves open.

c. Prior to decrdasfng the reacto~ coolant system temperature below

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• y I t t p 111 t dt PI d and their assocfateP headei stop valves cTosed.

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a. With no ECCS subsystems OPERABLE fn MOOES 3* and Q fmnedfately restore one ECCS subsystem to OPERABLE status or be fn COLO SHlJTOOMN within 20 hours.

seer pep d<<'.

With RCS temperature below 4Ã~ and with more than the allowed high pressure safety injection pump/OPERABLE or injection valves and header isolation valves open, fmmedfately disable the hfgh pressure safety injection pump(s) or close the header isolation valves.

c. In the event the ECCS fs actuated and injects water into the Reactor

'Coolant System. a Special Report shall be pre>ared and submftted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

SURVEILLANCE RE UIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstra'ted OPERABLE per the appl fcable Surveillance Requirements of 4.5.2.

4.5.3.2 =The high pressure safety injection pumps shall be verified inoperable and the assocfated header stop valves closed prior to decreasing below the above specified Reactor Coolant System temperature and once per month when the Reactor Coolant System fs at refueling temperatures.

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'Mfta pressurizer pressure <<1750 psfa.

AEADDDPDIATETE Idtgp T. LUCIE - UNIT 1 3/4 5-7 Amendment No. gSp'g

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RKACTIVITY CONTROL SYSTEMS BASKS 3/4.1.2 BORATION SYSTEMS Continued The boron addition capab111ty after the plant has been placed in HOOKS 5 and 6 requires either 1660 gallons of SX boric acid solutfon from thc boric acid tanks or 1630 gallons of 1720 ppm borated water from the refuel ing water tank to makeup for contraction of the primary coolant that could ASo: occur if the temperature fs lowered from 200'F to 140'F.

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~ 'al 3/4.1. 3 MOVABLE CONTROL ASSEMBLIES The specifications of this sect1on ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOW MARGIN fs maintained, and (3) the potential effects of a CEA tjcctfon accident are limited to acceptable levels.

The ACTION statements which permit limited varfatfons fma the basic requirements are accompanied by additional restrictions which ensure that.

the original criteria arc met.

The ACTION statements applicable to an famovable or untrfppable CEA and to a large misalignment (> 15 inches) of two or more CEAs, require a prompt shutdown of the reactor since either of these conditions may be indicative of a possible loss of mechanical functional capability of the CEAs and fn thc event of a stuck or untrfppablc CEA, the loss of SHUTDOMN NRCIN.

For small mfsalfganents (c 15 inches) of the CEAs. there 1s 1) a small degradatfon fn thc peaking factors relative to those assumed fn generating LCOs and LSSS setpofnts for DNBR and linear heat rate, 2) a small effect on thc t1me dependent. long term peer distributions relative to those used fn generating LCOs and LSSS setpofnts for DNBR and linear heat rate, 3) a small effect on the available SHUTDON( MARGIN, and 4) a small effect on the e)ected CEA worth used fn the safety analysis. Therefore, the ACTION statement associated with the small misalignment of a CEA permits a one hour time interval during which attempts may be made to restore the CEA to within fts alfgreent requirements prior to initiating a reduct1on fn THERMAL POWER. The onc hour time limit 1s sufficient to (1) identify causes of a mfsalfgned CEA, (2) take appropriate corrective act1on to realign the CEAs, and (3) mfnfmize the effects of xenon redistribution.

Overpower margin fs prov1ded to protect the core fn the event of a large misalignment (i 15 inches) of a CEA. However, this mfsalfgenent would cause distortion of the core power distribution. This distributions, fn turn, have a s1gniffcant effect on (1) the available SHUTDON NR&IN, (2) the tfme-dcpcndent long-term power dfstrfbutfons relative to those used fn gcnerat1ng LCOs and LSSS setpoints, and (3) the +ected CEA worth used fn the safety analysis. Therefore, the ACTION statement assoc1ated with the large mis-alignment of the CEA requires a prompt realignment of the mfsalfgned CEA.

ST. LUCIE - UNIT 1 B 3/4 1-3 Anendment No. g7,7I

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3 4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps fn operation, and maintain DNBR above the DKHR limit during all normal operations and anticipated transients . In NODES 1 and 2 with one reactor coolant loop not fn operation, this speci ffca-tion requires that the plant be fn at least HOT STANDBY within hour . 1

.In NODE 3, a single reactor coolant loop provides sufficfent heat ranoval capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.

In NODE 4, and in MODE 5 with reactor coolant loops filled, a single reac-tor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either shutdown cooling or RCS) be OPERABLE.

In NODE 5 with reactor coolant loops not filled, a single shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations and the unavailability of the steam generators as a heat removing component, require that at least two shutdown cooling loops be OPERABLE.

The operation of one Reactor Coolant Pump or one shutdown cooling pump provides adequate flow to ensure afxing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with bor on reductions will, therefore, be within the capability of operator recognftfon and control.

The restrictions on starting a Reactor Coolant Pump are provided to pr event RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR 50. The RCS will be protected against over pressure transients and will not exceed the limits of Appendix 6 by restricting- starting of the Reactor Coolant Pumps to when the secondary water temperature of each steam generator is less tha above each of the Reactor Coolant System cold leg temperatures.

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3/4.4.2 and 3 4.4.3 SAFETY VALVES The pressurizer code 'safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. Each safety valve is designed to relieve 2 x lOs lbs per hour of. saturated steam at the valve setpoint.

relief capacity of a single safety valve is adequate to relieve any over- The pressure condition which could occur during shutdown. In the event that no safety valves're OPERABLE. an operating shutdown cooling loop, connected to the RCS, provides overpressure relief capability and will pr event RCS overpressurfza-tion.

ST. LUCIE - UNI T 1 B 3/4 4-1 Amendment No. gB. 5'g

Pg "e7 V'I

REACTOR COOLANT SYSTEM BASES Reducing Tavg to c 500'F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters..

associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4. 9 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to with-stand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load. cycles used for design purposes are provided in Section 5.2.1 of the FSAR. Ouring startup and shutdown,. the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

Ouring heatup, the thermal gradients in the reactor vessel produ thermal stresses which vary from compressive at the i wall to tensi t the outer wall. These thermal induced compr ve stresses

'e tend to alle the tensile stresses induced by the i nal pressure.

Therefore, a pr ure-temperature curve based on ste state conditions (i.e., no thermal osses) represents a lower bo of all similar curves for finite hea rates when the inner of the vessel is treated as the. governing ation.

The hea0up analysis also c s determination of pressure-temperature limitations for the ca n which the outer wall of the vessel becomes the controlling atio The thermal gradients estab-lished during heatup produce nsile stre at the outer'all of the gcienuf; vessel. These stresses a additive to the p sure induced tensile Apy stresses which are air y present. The therma 'uced stresses at the outer wall of the v el are tensile and are depend on both the rate rg~g of heatup and th >me along the heatup ramp; therefore, lower bound curve similar that described for the heatup of the inne ll cannot be defined onsequently, for the cases in which the outer wa of the vessel omes the stress controlling location, each heatup rate inte t must be analyzed on an individual basis.

ST. LUCIE - UNIT 1 B 3/4 4-6

I Tp, INSERT 4.

Ouring heatup, the thermal gradients through the reactor vessel wall produce thermal stresses which are compressive at the reactor vessel inside surface and are tensile at the reactor vessel outside surface. Since reactor vessel internal pressure always produces tensile stresses at both the inside surface and outside surface locations, the total applied stress is greatest at the outside surface location. However, since neutron irradiation damage is larger at the inside surface location than at the outside surface location, the inside surface flaw may be more limiting. Consequently, for the heatup analysis, both the inside surface and outside surface flaw locations must be analyzed for the specific pressure and thermal loadings to determine which is more limiting.

Ouring cooldown, the thermal gradients through the reactor vessel wall produce thermal stresses which are tensile at the reactor vessel inside surface and are compressive at the reactor vessel outside surface. Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the inside surface.

Since neutron irradiation damage is also greater at the inside surface, the inside surface flaw location is the limiting location during cooldown.

Consequently, only the inside surface flaw must be evaluated for the cooldown ana lys i s.

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REACTOR COOLANT SYSTEM BASES 54 ~ R~A255x42tf The heatup and cooldown limit curves ( ) are composite curves which were prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatu cooldown rate/ of up to 100'F per hour. The heatup and cooldown curves were prepared based upon the most 1;miting value of the predicted adjusted reference temperature at the end of the service period.~

The reactor vessel materials have been tested to determine their initial RTNpT; the results of these test are shown in Table B 3/4.4-1.

Reactor operation and resultant fast neutron (E>l Mev) irradiation will cause an increase in the RTNpT. Therefore, an adjusted reference temperatur based upon the fluence~

he heatup and cooldown limit curves shown on Figurq~3.4-2w~p-'/~

include predicted adjustments for this shift in RTNOT at the end of the applicable service period, as well as adjustments for possible errors in the pressure and temperature sensing instruments.

The actual shift in RTNOT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73, r eactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalcu-lated when the ARTNOT determined from the surveillance capsule is dif-ferent from the calculated IRTNpT for the equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figurg~~ for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature require-ments of Appendix G to 10 CFR 50.

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REACTOR COOLANT SYSTEM BASES 4.4.13 pOWER OPERATED RELIEF VALVES and 3 4.4.14 REACTOR COOLANT PUNP-TART IN G w temperature reactor coolant system overpressure mitigating sys prov prevent RCS overpressurfzatfon above the 10 CFR 50, ix G, operating rves (Figure 3.4-2b or 3.4-2c, as applic at RCS temperatures be

' The RCS overpressurizatf em is based on the g gRAcf use of the pressurfzer for the design basis mass in pressurizer bubble by volume for tra, crated relfef I-V-1402 and I-V-1404) and the formation of a 60%

n basfs energy addftfon transient.

For the case when vo pressu steam bu formed, protection against .

A)Iaf the design basis ener tion transient fs a by limiting the Lt4RA' secondary-to-pr emperature differential below he operability of the R pressurization protection system wfll only red during per heatup and cooldown below RCS temperatures below cold shutdown, when the RCS has pressure boundary integrity.

275'fods 3/4.4.15 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensfble gases and/or steam from the primary system that could inhibit natural circulation core cooling. The OPERABILITY A at least one Reactor Coolant System vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this function.

The redundancy design of the Reactor Coolant System vent systems serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the Reactor Coolant System vent system are consistent with the requirements of Item II.b.l of HUREG-0737, "Clarification of THI Action Plan Requirements," November 1980.

ST. LUCIE - UNIT 1 B 3/4 4-15 Amendment No. g'4/, 6B

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l 3/4.5 EMERGENCY CORE COOLING SYSTEMS ECCS BASES 3/4.5.1 SAFETY INJECTION TANKS The OPERABILITY of each of the RCS safety injection tanks ensures hat a sufficient volume of borated water wi11 be inmediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the safety injection tanks. This initial surge of water into the core provides the initial cooling echanism during large RCS pipe ruptures.

The limits on safety injection tank volume, boron concentration and ressure ensure that the assumptions used for safety injection tank injection in the accident analysis are met.

The limit of one hour for operation with an inoperable safety injection tank minimizes the time exposure of the plant to a LOCA event ccurring concurrent with failure of an additional safety injection tank hich may result in unacceptable peak cladding temperatures.

4.5.2 and 3 4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two separate and independent ECCS subsystems nsures that sufficient emergency core cooling capability will be avail-ble in the event of a LOCA assuming the loss of one subsystem through ny single failure consideration. Either subsystem operating in conjunc-ion with the safety injection tanks is capable of supplying sufficient ore cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended reak of the largest RCS cold leg pipe downward. In addition, each ECCS ubsystem provides long term core cooling capability in the recirculation ode during the accidqnt recovery period.

The Surveillance Requirements provided to ensure OPERABILITY of ach component ensure that at a minimum, the assumptions used in the ccident analyses are met and that subsystem OPERABILITY is maintained.

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T. LVCIE - UNIT 1 B 3/4 5-1 Amendment Ho. 26