Text

Amend 31 to License NPF-16,changing RCS Pressure/Temp Limit Figures to Be Effective Up to Six Full Power Yrs of Operation
	ML17222A294
Person / Time
Site:	Saint Lucie
Issue date:	06/14/1988
From:	Berkow H; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML17222A295	List: ML17222A294; ML17222A297;
References
	NUDOCS 8806220336
	Download: ML17222A294 (33)
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 FLORIDA POWER 8( LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO FLORIDA AND FLORIDA HUNICIPAL POWER AGENCY DOCKET NO. 50-389 ST.

LUCIE PLANT UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

License No.

NPF-16 1.

The Nuclea~ Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Florida Power 8 Light Company, et al. (the licensee),

dated November 27, 1987, as supplemented Hay 4 and 20, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this aIwendment is in accordance with 10 CFR Part 51 of the CoInnission's regulations and all applicable requirements have been satisfied.

8806220336 880614 PDR ADOCK 05000389 P

PDR 2.

Accordingly, Facility Operating License No.

NPF-16 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 2.C.2 to read as follows:

2.

Technical S ecifications The Technical Specifications contained in Appendices A and B,

as revised through Amendment No. 3>, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION I

Herbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

June 14, 1988

ATTACHMENT TO LICENSE AMENDMENT NO.

31 TO FACILITY Of'ERATING LICENSE NO.

NPF-16 DOCKET NO.

50-389 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

The corresponding overleaf pages are also provided to maintain document completeness.

Remove Pa es Insert Pa es 1-4 3/4 4-3 3/4 4-5 3/4 4"10 3/4 4-29 3/4 4-30 3/4 4-31 3/4 4-32 3/4 4-33 3/4 4-35 3/4 4-36 3/4 4-38" B3/4 4-1 B3/4 4-3 B3/4 4-8 B3/4 4-9 B3/4 4-10 B3/4 4"11 1-4 3/4 4-3 3/4 4-5 3/4 4"10 3/4 4-29 3/4 4-30 3/4 4-31a 3/4 4-31b 3/4 4-32 3/4 4-33 3/4 4-35 3/4 4-36 3/4 4-37a 3/4 4-38" B3/4 4"1 B3/4 4-3 B3/4 4-8 B3/4 4"9 B3/4 4-10 B3/4 4-11 "There is no change to this page.

It is included to maintain document completeness.

DEFINITIONS DOSE E UIVALENT I-131 1.10 DOSE E(UIVALENT I-131 shall be that concentration of I-131 (microcuries/

gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually. present.

The thyroid dose conversion factors used for'this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

E -

AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for

isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95K of the total non-iodine activity in the coolant.,

ENGINEERED SAFETY FEATURES

RESPONSE

TIME 1.12 The ENGINEERED SAFETY FEATURES

RESPONSE

TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel senso~ until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. ).

Times shall include diesel generato:

starting and sequence loading delays where applicable.

FRE UENCY NOTATION

1. 13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table l. l.

GASEOUS RADWASTE TREATMENT. SYSTEM

l. 14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

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 b.

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.

ST.

LUCIE - UNIT 2 1-3

DEFINITIONS 1.16 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE The LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE is that operat'ing condition when (1) the cold leg temperature is less than or equal.tc that specified in Table 3.4-3 for the applicable operatingperiod,'nd (2) the Reactor Coolant System is not vented to containment'y an opening of at least 3.58 square inches.

MEMBER S

OF THE PUBLIC

l. 17 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally 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.

This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the pl ant.

OFFSITE DOSE CALCULATION MANUAL ODCM

1. 18 The OFFSITE DOSE CALCULATION MANUAL shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and shall include the Radiological Environmental Monitoring Sample point locations.

OPERABLE - OPERABILITY

l. 19 A system, subsystem,

train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s),

and when all necessary attendant instrumentation, controls, electrical

power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem,

train, component or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE -

MODE 1.20 An OPERATIONAL MODE (i.e.

MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50. 59, or (3) otherwise approved by the Commission.

ST.

LUCIE - UNIT 2 1-4 Amendment No. i<~ 3I

REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4. 1.3 At least two of the loop(s)/train(s) listed below shall be OPERABLE and at least one Reactor Coolant and/or shutdown cooling loops shall be in operation."

a.

Reactor Coolant Loop 2A and its associated steam generator and at least one associated Reactor Coolant pump,""

b.

Reactor Coolant Loop 2B and its associated steam generator and at least one associated Reactor Coolant pump,""

c.

Shutdown Cooling Train 2A, d.

Shutdown Cooling Train 2B.

APPLICABILITY; MODE 4, ACTION:

a.

With less than the above required Reactor Coolant and/or shutdown cooling loops OPERABLE, immediately initiate corective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is a shutdown cooling loop, be in COLD SHUTDOWN within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With no Reactor Coolant or shuta

~n cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

Al eactor oo ant pumps and shutdown cooling pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet teaperature is maintained at least 104F below saturation temperature.

<<<<A Reactor Coolant pump shall not be started with two idle loops and one or more of the Reactor Coolant System cold leg temperatures less than or equal to that specified in Table 3.4-3 for the applicable operating period unless the secondary water temperature of each steam generator is less than 40'F above each of the Reactor Coolant System cold leg temperatures.

ST.

LUCIE - UNIT 2 3/4 4-3 Amendnent No 44~

31

REACTOR COOLANT SYSTEM HOT SHUTDOWN SURVEILLANCE RE UIREHENTS

4. 4. l. 3. 1 The required Reactor Coolant pump(s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4. 1. 3. 2 The required steam generator(s) shall be determined OPERABLE by verifying the secondary side water level to be

> lOX indicated narrow range level at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4. 1.3.3 At least one Reactor Coolant or shutdown cooling loop shall be verified to be in operation and circulating Reactor Coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ST.

LUCIE - UNIT 2 3/4 4-4

REACTOR COOLANT SYSTEM COLD SHUTDOWN -

LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4. 1.4. 1 At least one shutdown cooling loop shall be OPERABLE and in operation",

and either a.

One additional shutdown cooling loop shall be OPERABLE

, or b.

The secondary side water level."f at least two steam generators shall be greater than lOX indicated narrow range level.

APPLICABILITY:

MODE 5 with Reactor Coolant loops filled ACTION:

a ~

b.

With one of the shutdown cooling loops inoperable and with less than the required steam generator level, immediately initiate corrective action to return the inoperable shutdown cooling loop to OPERABLE status or to restore the required steam generator level as soon as possible.

With no shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required shutdown cooling loop to operation.

SURVEiLLANCE RE UIREMENTS 4.4. 1.4. 1. 1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4. 1.4. 1.2 At least one shutdown cooling loop shall be determined :o be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The shutdown cooling pump may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 104F below saturation temperature.

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

A Reactor Coolant pump shall not be started with two idle loops and one or more of the Reactor Coolant System cold leg temperatures less than-or equal to that specified in Table 3.4-3 for the applicable operating period unless the secondary water temperature of each: steam generator is less than 40'F above each of the Reactor Coolant System cold leg temperatures.

ST.

LUCIE - UNIT 2 3/4 4-5 Amendment No.

REACTOR COOLANT SYSTEM COLD SHUTDOWN -

LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4. 1.4.2 Two shutdown cooling loops shall be OPERABLE and at least one shutdown cooling loop shall be in operation."

APPLICABILITY:

MODE 5 with reactor coolant loops not filled.

ACTION:

a.

b.

With less than the above required loops OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate corrective action to return the required loops to OPERABLE status as soon as possible.

With no shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate corrective action to return the required shutdown cooling loop to operation.

SURVEILLANCE RE UIREMENTS 4.4. 1.4.2 At least one shutdown cooling loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

The shutdown cooling pump may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are perIitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 104F below saturation temperature.

ST.

LUCIE - UNIT 2 3/4 4"6

REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER 0

LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a minimum water level of greater than or equal to 27K indicated level and a maximum water level of less than or equal to 68K indicated level and at least two groups of pressurizer heaters capable of being powered from 1E buses each having a nominal capacity of at least 150 kW.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

b.

With one group of the above required pressurizer heaters inoperable, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS 4.4.3. 1 The pressurizer water volume shall be determined to be within its limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified to be at least 150 kW at least once per 92 days.

4.4.3.3 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by verifying that on an Engineered Safety Features Actuation test signal concurrent with a loss of offsite power:

a.

the pressurizer heaters are automatically shed from the emergency power sources, and b.

the pressurizer heaters can be reconnected to their respective buses manually from the control room after resetting of the ESFAS test signal.

ST.

LUCIE - UNIT 2 3/4 4-9 Amendment No.

4'~

11

REACTOR COOLANT SYSTEM 3/4.4;4 PORV BLOCK VALVES LIMITING CONDITION FOR OPERATION 3.4.4 Each Power Operated Relief Valve (PORV) Block valve shall be OPERABLE.

No more than one block valve shall be open at any one time.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

With one or more block valve(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valve(s) to OPERABLE status or close the block valve(s) and remove power from the block valve(s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With both block valves

open, close one block valve within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The provisions of specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.4.4 Each block valve'shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of Action a. or b.

above.

Vhen the RCS cold leg temperature is above the LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE of Table 3.4-3.

ST.

LUCIE - UNIT 2 3/4 4-10 Amendment No.

31

REACTOR COOLANT SYSTEM 3/4.4. 9 PRESSURE/TEMPERATURE L IMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.

1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2, 3.4-3 and '3.4-4 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing.

APPLICABILITY: At al 1 times.

ACTION:

Mith any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T

and pressure to less than 200'F and 500 psia, respectively, within tlag following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS

4. 4.9. 1. 1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup,

cooldown, and inservice leak and hydrostatic testing operations.

ST.

LUCIE - UNIT 2 3/4 4-29 Amendment No. gg, 31

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (Continued) 4.4.9. 1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50 Appendix H in accordance with the schedule in Table 4.4-5.

The results of these examinations shall be used to update Figures 3.4-2, 3.4-3 and 3.4-4.

ST.

LUCIE - UNIT 2 3/4 4-30 Amendment No. i8, 3l

FIGURE 3.4.2 ST. LUCIE-2 P/T LIMITS, 6 EFPY HEATUP AND CORE CRITICAL 2500 I

50oF/HR 2000 K

o K

1500 N

D Q

1000 V

Oz OK 500 LOWEST SERVICE TEMP. 1680F 50oF/HR 535 PSIA COR E C R ITI GAL MIN, BOLTUP TEMP.

ALLOWABLEHEATUP RATES RATE oF/HR TEMP. LIMIT oF 50 AT ALLTEMPERATURES 100 300 400 500 TC IN ICATED REACTOR COOLANT TEMPERATURE' ST. LJCIE - UiIIT 2 3/4 4-31a Amendnent

.'Io.

31

FIGURE 3.4-3 ST. LUCIE-2 P/T LIMITS, 6 EFPY COOLPOWN AND INSERVICE TEST 2500 2000 KD 1500 N

D Q

1000 I-OZ INSERVICE TEST I

I l

I I

I LOWEST SERVICE I

TEMP. 1680F I

I ISOTHERMAL 100oF/HR & ISOTHERMAL 500 550 PSIA 30oF/HR 50 100 MIN. BOLTUP TEMP.

100 200 300 400 500 TC 'NDICATED REACTOR COOLANT TEMPERATURE oF ST.

LOCI- - 'JI'IIT 2 3/4 4-3lb Anendnent

.)o. 31

FIGURE 3.4-4 ST. LUCIE-2 P/T LIMITS, 6 EFPY MAXIMUMALLOWABLECOOLDOWN RATES 100 80 Kx 60 I-Kz 0

40 O

00U 20 RATE oF(HR 30 50 75 100 TEMP LIMIT oF 4104 104-130 130-146

)146 80 100 120 160 180 200 Tc INDICATED REACTOR COOLANT TEMPERATURE oF NOTE: A MAXIMUMCOOLDOWN RATE OF 100oF/HR IS ALLOWEDAT ANY TEMPERATURE ABOVE 146oF ST.

LUGIE - UNIT 2 3/4 4-32 Anendnent tIo. 31

I C

M m

TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE C

CAR I

CAPSULE NUMER 1

2 3

4 5

6 VESSEL LOCATION 83 97 104 263~

2770 284 LEAD FACTOR

<1.5

<1.5 WITHDRAWAL TIHE EFPY 1.0

24. 0 STANDBY 12.0 STANDBY STANDBY

REACTOR COOLANT SYSTEM PRESSURIZER HEATUP/COOLDOWN LIMITS LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:

a.

A maximum heatup of 100'F in any 1-hour period, and b.

A maximum cooldown of 2004F in any 1-hour period.

APPLICABILITY: At al 1 times.

ACTION:

With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown.

ST.

LUCIE - UNIT 2 3/4 4-34 Amendment No.

16

0 REACTOR COOLANT SYSTEM OVERPRESSUQE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 Unless the RCS is depressurized and vented by at 1'hast 3.58 square

inches, at least one of the following overpressure protection systems shall be OPERABLE:

a.

Two power-operated relief valves (PORVs) wi.th a lift setting of less than or equal to 470 psia and with their associated block valves open.

These valves may only be used to satisfy low temperature overpr essur e protection (LTOP) when the RCS cold leg temperature is greater than the temperature listed in Table 3.4-4.

b.

Two shutdown cooling relief valves (SDCRVs) with a lift setting of less than or equal to 350 psia.

c.

One PORV with a liftsetting of less than or equal to 470 psia and with its associated block valve open in conjunction with the use of one SDCRV with a liftsetting of less than or equal to 350 psia.

This combination may only be used to satisfy LTOP when the RCS cold leg temperature is greater than the temperature listed in Table 3.4-4.

APPLICABILITY:

MODES 3, 4, 5 and 6.

ACTION:

With either a

PORV or an SDCRV being used for LTOP inoperable, restore at least two overpressure protection devices to OPERABLE status within 7 days or:

1.

Depressurize and vent the RCS with a minimum vent area of 3.58 square inches within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; OR 2.

Be at a temperature above the LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE of Table 3.4-3 within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b.

With none of the over pressure protection devices bei'ng used for ESTOP OPERABLE, within the next eight hours either:

1.

Restore at least one'overpressure protection device to OPERABLE status or vent the RCS; OR 2.

Be at a temperature above the'OW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE of Table 3.4-3.

With cold leg temperature within the LOll TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE of Table 3.4-3.

ST.

LUCIE - UNIT 2 3/4 4-35 Amendment No. JS.

31

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued):

c.

In the event either the

PORVs, SDCRVs or the RCS vent(s) are used to mitigate a

RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the

PORVs, SDCRVs or

'ent(s) on the transient and any corrective action necessary to prevent recurrence.

d.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a.

In addition to the requirements of Specification 4.0.5, operating the PORV through one complete cycle of full travel at least once per 18 months.

ST.

LUCIE - UNIT 2 3/4 4-36 Amendment No. JS, 31

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued)

Performance of a CHANNEL FUNCTIONAL TEST on the PORY actuation

channel, but excluding valve oper ation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORY is required OPERABLE.

C.

d.

Performance of a CHANNEL CALIBRATION on the PORV actuation

channel, at least once per 18 months.

Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORV is being used for overpressure protection.

4.4. 9.3.2 The RCS vent(s) shall be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months " when the vent(s) is being used for overpressure protection.

Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

ST.

LUCIE - UNIT 2 3/4 4"37

TABLE 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE Operating

Period, EFPY During

~He at u During Cooldown Cold Le Tem erature F'>6

< 313

< 304 TABLE 3.4-4 MINIMUM COLD LEG TEMPERATURE FOR PORV USE FOR LTOP Operating Period EFPY 4>6

cold, F'uring

~Heatu 156

cold, F'uring Cool down 179 ST.

LUCIE - UNIT 2 3/4 4-37a Amendment No.

31

REACTOR COOLANT SYSTEM 3/4.4. 10 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4. 10 At least one Reactor Coolant System vent path consisting of two vent valves and one block valve powered from emergency buses shall be OPERABLE and closed at each of the following locations:

a.

Pressurizer steam

space, and b.

Reactor vessel head.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

a.

b.

With one of the above Reactor Coolant System vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the vent valves and block valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With both Reactor Coolant System vent paths inoperable, maintain the inoperable vent paths closed with power removed from the valve actuators of all the vent valves and block valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.4. 10. 1 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months by:

1.

Verifying all manual isolation valves in each vent path are locked in the open position.

2.

Cycling each vent valve through at least one complete cycle of full travel from the control room.

3.

Verifying flow through the Reactor Coolant System vent paths during venting.

ST.

LUCIE - UNIT 2 3/4 4"38 Amendment No. 25

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 in operation, and maintain DNBR above 1.20 during all normal operations and anticipated transients.

In MODES 1

and 2

with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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

In MODE 4, 'and in MODE 5 with reactor coolant loops filled, a single reactor 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 MODE 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, requi re 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 mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions. in the Reactor Coolant System.

The reactivity change rate associated with boron reductions will; therefore, be within the capability of operator recognition and control.

The restriction on starting a reactor coolant pump in MODES 4 and 5, with two idle loops and one or. more RCS cold leg temperatures less than or equal to that specified in Table, 3.4-3 for the applicable operating period is provided to prevent RCS pressure transients, caused by energy additions from the secondary system from exceeding the limits of Appendix G to 10 CFR Part 50.

The RCS will be protected against overpressure transients by (1) sizing each PORV to mitigate the, pressure transient of an inadvertent safety injection actuation in a water=solid RCS'ith pressurizer heaters energized, (2) restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 40'F above each of the RCS cold leg temperatures, (3) using SDCRVs to mitigate RCP.start transients.and the transients caused by inadvertent SIAS actuation and charging water, and (4) rendering one HPSI pump inoperable when the RCS is at low temperatures.

3 4.4.2 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 212,182 lbs per hour of saturated steam at the valve setpoint.

The relief capacity of a single safety valve is adequate to relieve any overpres-sure condition which could occur during shutdown.

In the event that no safety valves are

OPERABLE, an operating shutdown cooling loop, connected to the

RCS, provides overpressure relief capability and will prevent RCS overpressurization.

In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

ST.

LUGIE - UNIT 2 B,3/4 4-1 Amendment No. N.

31

REACTOR COOLANT SYSTEM BASES.

SAFETY VALVES Continued During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia.

The combined relief capacity of these valves is sufficient to limit the system pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pres-surizer Pressure-High) is reached (i.e.,

no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of tie pressurizer power-operated relief valve or steam dump valves.

Demonstration of the safety valves'ift settings will occur only during shutdown and wi 11 be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.

3/4.4.3 PRESSURIZER An OPERABLE pressurizer provides pressure control for the Reactor Coolant System during operations with both forced reactor coolant flow and with natural circulation flow.

The minimum water level in the pressurizer assures the pressurizer

heaters, which are required to achieve and maintain pressure

control, remain covered with water to prevent failure, which could occur if the heaters were energized uncovered.

The maximum water level in the pres-surizer ensures that this parameter is maintained within the envelope of operation assumed in the safety analysis.

The maximum water level also ensures that the RCS is not a hydraulically solid system and that a steam bubble will be provided,to'accommodate pressure surges during operation.

The steam bubble also protects the pressurizer code safety valves against water relief.

The requirement to verify that on an Engineered Safety Features Actuation test signal concurrent with a loss of offsite power the pressurizer heaters are automatically shed from the emergency power sources is to ensure that the non-Class 1E heaters do not reduce the reliability of or overload the emergency, power source.

The requirement that a minimum number of pressuri" r heaters be OPERABLE enhances the capability, to control Reactor Coolant System pressure and establish and maintain natural circulation.

ST.

LUCIE - UNIT 2 B 3/4 4-2

BASES 3/4.4.4 PORV BLOCK VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.

Operation of the PQRVs in conjunc-tion with a reactor trip on a Pressurizer Pressure-High signal minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

The opening of the PORVs fulfills no safety-related function and no credit is taken for their operation in the safety analysis for MODE 1, 2, or 3.

Each PORV has a remotely operated block valve to provide a'positive shutoff capability should a relief valve become inoperable.

Since it is impractical and undesirable to actually open the PORVs to demonstate their reclosing, it becomes necessary to verify OPERABILITY of the PORV block valves to ensure the capability to isolate a malfunctioning PORV.

As the PORVs are pilot operated and require some system pressure to operate, it is impractical to test them with the block valve closed.

The PORVs are sized to provide low temperature overpressure protection (LTOP).

Since both PORVs must be OPEPABLE'when used for LTOP, both block valves will be open during'peration within the LTOP range, As the PORV capacity required to perform the LTOP function is excessive for operation in NODE 1, 2, or 3, it is necessary that the operation of more than one PORV be precluded during these HODES'.

Thus, one block valve must be shut during MODES 1, 2, and 3.

The applicability of this technical specification to only a part of MODE 3 is due to the LTOP range slightly overlappi'ng HODF. 3 in..the operating period beyond 15 EFPY.

Both block valves will be 'opeh during operat'ion in these lower temperature portions of MODE 3, 3I4.4.5 STEAM GENERATORS The Surveillance Requirements fot inspection of the steam generator tubes ensure that the structural integt ity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice Inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

ST.

LUCIE - UNIT 2 B 3/4 4-3 Amendment No.

31

REACTOR COOLANT SYST H

BASES STEAM GENERATORS Continued Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system primary-to-secondary leakage

= 1.0 gpm from both steam generators.

Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.

Operating plants have demon-strated that primary-to-secondary leakage of 0.5 gpm per steam generator ccn readily be detected by radiation monitors of steam generator blowdown.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.

However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging will be requir ed for all tubes with imperfections exceeding the plugging limit of 40K of the tube nominal wall thickness.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20K of the original tube wall thickness.

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.

1 LEAKAGE OETECTION SYSTEHS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressu're Boundary.

These detection systems are consistent with the recommendations of ST.

LUCIE - UNIT 2 B 3/4 4-4 Amendment No. 13

REACTOR COOLANT SYSTEH BASES

'PECIFIC ACTIVITY Continued The sample analysis for determining the gross specific activity and E can exclude the radioiodines because of the.low primary coolant limit of 1 micro-curie/gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the radioiodine level is to be determined every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If the gross specific activity level and radioiodine level in the primary coolant were at their limits, the radioiodine contribution would be approximately 1X.

In a release of primary coolant with a typical mixture of radioactivity, the actual radioiodine contribution would probably be about 20K.

The exclusion of radionuclides with half-lives less than 10 minutes from these determinations has been made for several reasons.

The first,consideration is the difficulty in identifying short-lived radionuclides in a sample that requires a significant time to collect, transport, and analyze.

The second consideration is the predictable delay time between the postulated release of radioactivity from the primary coolant to its release to the environment and transport to the SITE BOUNDARY, which is relatable to at least 30 minutes decay time.

The choice of 10 minutes for the half-life cutoff was made because of the nuclear characteristics of the typical primary coolant radioactivity.

The radionuclides in the typical primary coolant have half-lives of less than 4 minutes or half-lives of greater than 14 minutes, which allows a distinction between the radionuclides above and below a half-life of 10 minutes.

For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before thay could be transported from'he primary coolant to the SITE BOUNDARY under any accident cordition.

Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2'ours between sample taking and completing the initial analysis is based upon a typical time necessary to perform the sampling, transport the sample, and perform the analysis of about 90 minutes.

The gross count should be made in a reproducible geometry of sample and counter having reproducible Y or p self-shielding properties.

The counter should be reset to a reproducible efficiency versus energy.

It is not necessary to identify specific nuclides.

The deteraination of the contributors to the E result should be based upon those energy peaks identifiable with a 95K confidence level.

The radio-chemical determination of nuclides should be based on multiple counting of the sample with typical counting basis following sampling of less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , about 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , about 1 day, about 1 week, and about 1 month.

Reducing T

to less 'than 500 F prevents the release of activity should avg a steam generator tube rupture since the saturation pressure of the primary coolant is below the liftpressure 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.

ST.

LUCIE - UNIT 2 B 3/4 4-7

BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand 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 of the FSAR.. During 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.

During heatup, the thermal gradients through the reactor vessel wall produce thermal stresses which are compressive at the reactor vessel inside surface and which are tensile 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 outside surface location.

However, since neutron irradiation damage is larger at the inside surface location when compared to the outside surface, the inside surface flaw may be more limiting.

Consequently, for the heatup analysis both the inside and outside surface flaw locations must be analyzed for the specific pressure and thermal loadings to determine which is more limiting.

During cooldown, the thermal gradients through the reactor vessel wall produce thermal stresses which are tensile at the reactor vessel inside surface and which 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 location.

Since the neutron indication damage is also greatest at the inside surface location the inside surface flaw is the limiting location.

Consequently, only the inside surface flaw must be evaluated for the cooldown analysis.

The heatup and cooldown limit curves Figures 3.4-2, 3.4-3 and 3.4-4 are composite curves which were prepared by determining the most conservative

case, with either the inside or outside wall controlling, for any heatup rate of up to 50 degress F,per hour or cooldown rate of up to 100 degrees F per hour.

The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the service period, and they include adjustments for possible errors in the pressure and temperature sensing instruments.

The reactor vessel materials have been tested to determine their initial RT D, the results of these tests are shown in Table B 3/4.4-1.

Reactor operation anÃesultant fast neutron (E greater than 1 NeV) irradiation will cause an increase in.the RTNDT An adjusted reference temperature can be predicted using a) the initial RT

, b) the fluence (E greater than 1 NeV), including appropriate adjustments for nNFron attenuation and neutron energy spectrum variations through the wall thickness, c) the copper and nickel contents of the material, and d) the transition temperature shift from the curve shown in Figure B 3/4.4-1 as recommended by Regulatory Guide 1.99, Revision 2 (Draft), "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials."

The heatup and cooldown limit curves Figures 3.4=2, 3.4-3 and 3.4-4 include predicted adjustments for this shift in RTNDT at the end of the applicable service period.

ST.

LUCIE - UNIT 2 B 3/4 4-8 Amendment No. 7$.

3I

TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS I

CnM m

C:

M Piece No.

Code No.

Haterial Vessel Location Drop Weight Results Temperature of Charpy V-Notch RT ( F)

NDT 950 ft-lb Hinimum Upper Shelf Cv energy for Longitudinal(1)

Direction Charpy Ft - lb O

122-102A 122-1028 122-102C 124-1028 124-102C 124-102A 142-102C 142-1028 142-102A 102-101 106-101 128-101A 128-101D 128"101B

. 128"101C 128-3018 128-301A 126-101 131-102A 131-102D 131-1028 131-102C 131-101B 131-101A 152-101 154-102 (A to F) 104-102 (A to D)

(1) Repor H-604-1 H-604-2 H-604-3 H-605-1 H-605-2 H-605"3 H-4116-1 H-4116-2 H-4116-3 H-4110-1 H-4101-1 H-4102-1 H-4102-2 H-4102-3 H-4102" 4 H"4103-1 H-4103-2 H-602-1 H-4104-1 H-4104-2 H-4104-3 H-4104-4 H"4105"1 H-4105-2 H-4112-1 H-4111"1 H-4109-1 ted only for SA 5338 Cl 1 SA 533B Cl 1 SA 533B Cl 1 SA 533B Cl 1 SA 533B Cl 1 SA 533B Cl 1 SA 533B Cl 1 SA 5338 Cl 1 SA 533B Cl 1 SA 533B Cl 1 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 1 SA 508 Cl 1 SA 508 Cl 1 SA 508 Cl 1 SA 508 Cl 1 SA 508 Cl 1 SA 5338 Cl 1 SA 5338 Cl 1 SA 5338 Cl 1 bel tline region plates.

Upper Shell Plate Upper Shell Plate Upper Shell Plate Intermediate Shell Plate Intermediate Shell Plate Intermediate Shell Plate Lower Shell Plate Lower Shell Plate Lower Shell Plate Closur e Head Closure Head Flange Inlet Nozzle Inlet Nozzle.--

Inlet Nozzle Inlet Nozzle Outlet Nozzle Outlet Nozzle Vessel Flange Inlet Nozzle Safe End Inlet Nozzle Safe End Inlet Nozzle Safe End Inlet Nozzle Safe End Outlet Nozzle Safe End Outlet Nozzle Safe End Bottom Head Dome Bottom Head Torus Closure Head Torus 0

+10

-10 0

-10

-20

-30

-50

-40

-10 0

-20

-20 0

-10 "20

-30

-20

-10

-50

-40

-60

+50

+10

+30

+10

+20

-+20

+30 0

-20

-20 0

-10

-20

-30

-10

+20

+20 0

0

-40

+40

-10 105

.-113 113 91 105 100

150 lio leap 120 110 r100 9 0 fa 80

.0 tat gp P

(0 gp

~ j

~

o l. rt t I

I';

~

~ of(i,.

.iil

~ "(",

~. If

~ ~ ~

~ o (I

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!!II i! i fi'::

II iii I(ii

(

! I !i'I "'i i;iIri'I i

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!I '.:-I! I

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(fle(!".l !ji( eio(

f)i",')II!!I j}ji ~ ~ ~ ~ ~r" e ~ ~ ..r. ~ re ~ ~ r q;rr~/ I ))tk ff ' ll!Ie((e) o~ ( I~::: ar l(il o fl. le .i:":Ii(ii

I. !I:! i I!i!Ie!(

o'll II), ((" I!.! lj( itic )f(;jill'. )I!i}!!ii I(I(lo( I!!I!'i!I;!i!!(! (Lo I" I!Ili!!!i!i'j:. .f.. i!L I I! '-.f I Ij!(!II )(II l!f (le (!r !( LljI' i i ! (I'. '.ijif Ii}iii.'(1'."! )i~ ~ ,!l'" IiftTji 'ii;(+ ~ ~ !ii:::; ~ ~ ~ I ~ ( I;il ~ ' 'I tll ~r oel.j!l!'. r>> ~ r ~ I lr' '. !F. E(l:I j ..Il E)'i ~ I't! ~ e I !='I H(t(t ii!: o.l ~ ~ ~ e ljjiii!("

'(!l I I I(l( ((((

e }f)if!()!I:i I .'Io f..: }ii

I I-I -:~)oi=

!" I3s t(t II I I j().I +e .Q'f ~ITfI I .=-Li (f l!i!lji!."=.i jijjjfj r t'I, l(L: ~ ~ ~ ~ o('.I ~ etc I I.'" -:.-a-.0 i: !.::::.Il ~ !e ~ )}~ ( ~ ! ilI'l Io (()('o i('Q ~ ltl' ~ ~ I I el( ~l(('l: ~

h. I(l(ifi!I!I i)"l':" (l

~'~F t

I

=.E.::: EE 'I 6'T "E o ~ e ie I'(I !fl 'I l.! -ii N ~.r.:I= ji(i q f)!~ F. I >(I rr) EHI. I'P o ro ol(i e ~ oijfji: ~ ~ ~ ~ ~ r ~ ~ ~ ~ el Ill i }} fffffol.i jill lfl! I!! '..':i! !!!:ir ~ ~ e I'l II)I'I! (I:. o! 0, ~ (,a(a !. jt'" aoce IFL I o \\ o I'ol'(( .'-I! I I I I. <~e (,a'( i;I Su ye< I(>) I", I.I I!:::i '.ll!)':i!i I)l!I!!ii':::lil"!:!itt:.:.::I:-::..:-i'0 0 10 >> ~ I ~ II '"'I'I I':i'i 101 9 Q - Actual 5urveillance Capsule Results Calculated 4 RTNgT Shift - oF I o ~ e I I 1020 NEUTRC(H fI.UEHCK, n/ca (E i 1 HeV) 5T. I.UCIE "UNIT 2 ~re ICli.i 0 Reference Teltperature (RTNpT) tncreases as a function of fast (E 1 QeV) neutron fluence (550 F irradiation) for Reactor Yessel tteltline )aterials ST. LUCIE - UNIT 2 83/4 4-10 Amendment No. 3)

REACTOR COOLANT SYSTEM BASES The actual shift in RT T of the vessel material will be established periodically during operati)lPby removing and evaluating, in accordance with ASTH E185-73 and 10 CFR Appendix H, reactor vessel material irradiation surveil-lance specimens installed near the inside wall of the reactor vessel in the core area. The surveillance specimen withdrawal schedule is shown in Table 4.4-5. 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 recalculated when the delta RT determined from the surveillance capsule is different from the calculated Sita RTN for the equivalent capsule radiation exposure. The lead factors shown 6 Table 4.4-5 are the ratio of neutron flux at the surveillance capsule to that at the reactor inside surface. The pressure-temperature limit lines shown on Figures 3.4-2, 3.4-3 and 3.4-4 for reacto~ criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. The maximum RTN T for all Reactor Coolant System pressure-retaining materials, with the,3xception of the reactor pressure

vessel, has been determined to be 60'F.

The Lowest Service Temperature limit line shown on Figures 3.4-2, 3.4-3 and 3.4-4 is based upon this RT T since Article NB-2332 (Summer Addenda of 1972) of Section III of the ASHE )Piler and Pressure Vessel Code requires the Lowest Service Temperature to be RT DT + 100'F for piping,

pumps, and valves.

Below this temperature, the system pressure must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3125 psia. The limitations imposed on the pressurizer heatup and cooldown rates and spray. water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASHE Code requirements. The OPERABILITY of two PORVs, two SDCRVs or an RCS vent opening of greater than 3.58 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of 'the RCS cold leg temperatures are less than or equal to the applicable maximum LTOP temperatures. The Low Temperature Overpressure Protection . System has adequate relieving capability to protect the RCS from overpressuriza-tion when the transient is limited to either (1) a safety injection actuation in a water-solid RCS with the pressurizer heaters energized or (2) the start of an idle RCP with the secondary water temperatur'e'of the steam generator less than or equal to 40'F above the RCS cold leg temperatures with the pressurizer water-solid. ST. LUCIE - UNIT 2 B 3/4 4-11 Amendment No. 7$, 31

REACTOR COOLANT SYSTen BASES 3/4.4. 10 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling. The OPERABILITY of at least one reactor Coolant System vent path from the reactor vessel head and the pressurizer steam space ensures the capa-bilityy 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. 1 of NUREG-0737, "Clarification of THI Action Plan Requirements," November 1980. 3/4.4. 11 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASHE Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of 'these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a (g) (6) (i). Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through Summer 1973. ST. LUCIE - UNIT 2 B 3/4 4-12}}

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