ML18005A979

Proposed Tech Specs Re RCS Cooldown & Heatup Limitations

ML18005A979
Person / Time
Site:	Harris
Issue date:	06/30/1989
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML18005A977	List: ML18005A976 ML18005A979
References
NUDOCS 8907070181
Download: ML18005A979 (58)

Text

ENCLOSURE 4 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATION PAGES 8907070i ai s9o630 pgp +DOCK 05000400 pgg p

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LIMITING CONOITIONS FOR OPERATION ANO SURVHLLANCE RE UIREMENTS SECTION PAGE 3/4.4. 9 PRESSURE/TEMPERATURE LIMITS Re C4l C00 1'y5%o ~ ~ ~ o~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-33 FIGURE 3. 4-2 REACTOR CQQLAN SYSTEM CQQLQQMN LINITATIQNS-APPLICABLK UP TQ ~ ~ ~ ~' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-35 FIGURE 3.4'3 REACTOR CQO SYSTBI HEATUP LIMITATIONS AP PLICASLE UP TO ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-36 3

TABLE 4. 4-5 REACTOR VESSEL TERIAL SURVEILLANCE PROGRN-MITHQRAMAL SCHEOULEe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ o ~ ~ 3/4 4-31 TAHLE 4.4-6 MAX9%N HEATUP ANO CQOLDQMN RATES FQR MOGES 4, 5 ANO 6 (MITH REACTOR VESSEL HBD QN)............... .---" ~ 3/4 4 38 P 55S!lf f 28'pe ~ e ~ ~ ~ ~ ~ ~ ~ e ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o o e ~ e o ~ ~ ~ e ~ ~ ~ ~ ~ o ~ ~ ~ 3/4 4-39 Qverpressure Protactfon Syste.......................... 3/4 4 40 FIGURE 3.4-4 MAXIMUM ALLQMEO PORV SETPOINT FOR THK UN TEMPERATURE QVERPRESSURK SYSTEM...................................... 3/4 4 41 3/4e 4e 10 STRUCTURAL XEGRITYe ~ e e ~ o e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ e e o ~ ~ ~ ~ ~ ~ '

~ ' ' ~ 3/4 4 43 3/4e 4e 11 REACTOR COOLANT SYSTEM VQTSe ~ ~ ~ o ~ ~ ~ ~ ~ oo ~ ~ ~ ~ ~ e ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4 44 3/4.5 EMERGBCY CORE COOLING SYSTEMS 3/4.5.1 ACCULTQRSe o oovo4 e ~ e e ~ ~ ~ ~ ~ e e ~ ~ e e

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 5-1 3/4.S.2 ECCS SUBSYSTEMS - T4VQ GREATER THAN QR @NL TO 350 F. 3/4 5'3 3/4.5.3 ECCS QFSTEMS - T4vg LESS THAN 350 t-e ~ ~ ~ ~ o ~ o eoo ~ ~ ~

~ ~ 3/4 5 7 3/4.5.4 REFUELING MATER STORAGE TANKS o o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ 3/4 5-9 SHEARQN HARRIS UNIT 1 vfff

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BIO EX BASES SEETEGN PAGE TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS.......................... B 3/4 4-8 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>iHeV) AS A FUNCTION OF FULL POWER SERVICE LIFE.................................. B 3/4 4-9 FIGURE B 3/4.4-2

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~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 4-10 3/4.4. 10 STRUCTURAL INTEGRITYe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ B 3/4 4-'15 3/4.4. 11 REACTOR COOLANT SYSTEH VENTS........................... ~ ~ B 3/4 4-15 3/4. 5 BIERGENCY CORE COOLING SYSTBIS 3 I4 5 1 e e ACCUHU LATORS o ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEHS..........................e... B 3/4 5-1 3/4e Se 4 REFUELING WATER STORAGE TANKe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 5-2 3/4. 6 CONTA INHENT SYSTEHS 3/4. 6. 1 PRIMARY CONTAINHENTo ~ ~ o ~ ~ ~ o ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 6-1 B 3/4 6-3 3/4. 6. 2 OEPRESSURIZATION ANO COOLING SYSTEHSe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

ISOLATION VALVES...................... .. -. B 3/4 5-4 3/4. 6..3 ~ ~ ~ ~ 'QNTAINHENT B 3/4 6-4 3/4. 6. 4 COHBUSTIBLE GAS CQNTROLo ~ ~ ~ ~ ~ ~ ~ ~ ~ o o ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

B 3/4 6-4 3/4.6. 5 VACUUH RELIEF SYSTEHo o o ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ e ~ o o o ~ o o ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~

SHEARON HARRIS - UNIT 1 X]V

. REACTIVITY CONTROL SYSTEMS 3/4. 1. 2 BORATION SYSTEMS FlOW PATH - SHUTDOWN LIMITING CONQITION FOR OPERATION 3.1.2.1 As a minimum, ane of the following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

a. A flaw path fram the boric acid tank via either a boric acid transfer pump or a gravity feed connection and a charging/safety injection pump to the Reactor Coolant System if the boric'acid tank in Speci-fication 3.1.2.5a. or 3.1.2.6a. is OPERABLE, or

b. The flow path fram the refueling water storage tank via a charging/

safety injection pump to the Reactor Coolant System if the refue1ing water storage tank in Specification.3. 1.2.5b. or 3. 1. 2.6b. is OPERABLK.

APPLICABILITY: MOOES 4", 5", and 6".

ACTION:

With none of the above flow paths OP'ERABLE or capable af being powered fram an OPERABLE emergency power saurce, suspend all operations involving CORK ALTERATIONS or positive reactivity changes.

SURVEILLANCE RE UIREMENTS 4.1.2.1 At least one of the above required flaw paths shal1 be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the temperature of the flow path between the boric acid tank and the charging/safety injec tion pump suction header is greater than or equal to 654F when a flow path fram the boric acid tank is used, and

b. At least once per 31 days hy verifying that each valve (manual, power operated, or automatic} in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

~A maximum of one charging/safety injection pump shall be OPERABLE whenever the temperature of one or mom af the RCS cold legs is less than or equal to F and the reactor vessel head is in place.

9R5 SHEARON HARRIS - UNIT 1 3/4 1-7

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REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTOOMN LIMITING CONOITIQN FOR OPERATION 3.1.2.3 One charging/safety injection pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

APPLICABILITY: MOOES 4", 5"0, and 6"8.

ACTION:

With no charging/safety injection pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORK ALTERATIONS or positive reactivity changes.

SURVEILLANCE RE UIREHENTS 4.1.2.3.1 The above required charging/safety injection pump shall be demon-strated OPERABLE by verifying, on recirculation flow or in service supplying flow to the reacto~ coolant system and reactor coolant pump seals, that a differential pressure across the pump of greater than or equal to 2446 psid is developed when tested pursuant to Specification 4.0.5.

4. 1. 2.3.2 All charging/safety injection pumps, excluding the abave required OPERABLE pump, shall be demonstrated inoperable~

by veri ing that 4'- motor ~~~t's circuit breaker 'ecu in the open position+ p>oQ +< fe~peyg~y~

ozone or more. oF We RQS cold legs decreasing gael,~ gag'p~nd av /easF once p<y 3l days +ereaAer, u~~ aber the, reactor vessel head is ~oveJ.

"A maximum of one charging/safety injection pump shall be OPERABLE whenever the temperature of one or more of the RCS cold 'legs is'ess than or equal to and the reactor vessel head is in place.

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~An in erable pump may be energized for testing provided the discharge of

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the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator or by a manual isolation valve secured in'the closed position.

SFor periods of no more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, when swapping pumps, it is permitted that there be no OPFRABLE charging/safety injection pump. No CORE ALTERATIONS or positive reactivity changes are permitted during this time.

SHEARON HARRIS

- UNIT 1 3/4 1-9

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HOT SHUTDOWN LIMITING CONOITION FOR OP'ERATION 3.4.1.3 At least two of the loops listed below shall be OPERABLE and at least one of these loops shall be in operation:"

a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,""

b: Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,""

c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump,~

d. RHR Loop t:A], or

e. RHR Loop [Bj.

APPLICABILITY: MODE 4.

ACTION:

a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is an RHR, loop, be in COLO SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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

• All reactor coolant pumps and RHR pumps may be deenergized 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 104F below saturation temperature.

~A reactor coolant pump shall not be started with one or mo Reactor unless 32S Coolant System cold leg temperatures less than or equal to F 504F the secondary water temperature of each steam generator is less than above each of the Reactor Coolant System cold leg temperatures.

SHEARON HARRIS - UNIT 1 3/4 4-4

REACTOR COOLANT SYSTEM COLO SHUTOOMN - LOOPS FILLED LIMITING CONDITION FOR OPERATION

3. 4.1.4.L At least one residual heat removal (RHR) loop shall be OPERABLE and in operation", and either:

a. One additional RHR loop shall be OPERABLE"", or

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

APPLICABILITY: HOOK 5 with reactor coolant loops filled""".

ACTION:

a. Qith one of the RHR loops inoperable and'with less than the required steam generator water level, imnediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator water level as soon as possible.

b. 'lith no RHR 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 RHR loop to operation.

4 SURVEILLANCE RE UIREMENTS 4.4.1.4.L.l The narrow range secondary side water level af.at least two steam generators when required shall be determined to be within 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 RHR loop shall be determined to be in operation and circulating reactor coolant at least once per I2 hours.

"The RHR pump may be daenergized 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 opera-tions are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.

""One RHR 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 RHR loop is OPERABLE and in operation.

"" A reactor coolant pump shall not be started with one or mor the Reactor Coolant System cold leg temperatures less than or equal to F un ess 82,5 the secondary water temperature of each steam generator is less than 50 F above each of the Reacto~ Coolant System cold leg temperatures.

SHEARON HARRIS - UNIT 1 3/4 4-6

REACTOR COOLANT SYSTEH 3/4.4.4 RELIEF VALVES LIHITING CONOITION FOR OPERATION 3.4.4 All power operated relief valves (PORVs) and thefr associated block valves shall be OPERABLE.

APPLICABILITY: HOOfS 1, 2, 3, and 4~

ACTION:

a. With one or more PQRV(s) inoperable, because of excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLE status or close the associated block valve(s); otherwise, be fn at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN withfn the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With one PORV inoperable as a result of causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block val ve.

C. With two PORVs inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLE status or close the associated block valve(s) and remove power from the black valve(s); restore the PQRV to OPERABLE status wfthfn the fol-lowing 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be fn HQT STANOBY within the next,6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLO SHUTOOWN wfthfn the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

With all three PORVs fnoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PQRV(s) to OPERABLE status or close their associated block valve(s) and remove power, frotm the block valve(s) and be in HOT STANOBY wfthfn the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLO SHUTOOWN'wfthfn the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

With one or more block valve(s) inoperable, wfthin 1 hour:

(1) restore the block valve(s) to OPERABLE status, or close the block valve(s) and remove power froa the block valve(s), or close the PQRV and remove power from its assocfated solenoid valve; and (2) apply the ACTION b., c. or d. above, as appropriate, for the isolated PORV(s).

The provfsfons of Specff'fcatfon 3.0.4 are not applicable.

" HOOE 4 when the temperature of all RCS cold legs is greater than 3%4F.

~S25 SHEARON HARRIS - UNIT 1 3/4 4-11

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3/4.4. 9 PRESSURE/TB!PERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONOITION FOR- OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and 3.a2 d1.4-51 ill hydrostatic testing with:

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pressure shall be limited in accordance with the limit lines shown on Figures di i* 1 d

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

b. A maximum cooldown .of 1004F in any 1-hour period, and

c. A maximum temperature change of less than or equal to 10'F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICAHILITY: MOOES 1, 2, and 3.

ACTION: g pg9~g

+g aressure ~d fcrnpemhcre. Iimi+ lines sho~n 3.g-g an J 8.f-g aver exceeded>

e on With 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 operation or be in at least HOT STANOBY'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 Tavg and pressure to less than 2004F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS 4.4.9. 1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during cooldown, and inservice leak and hydrostatic testing operations.

system'eatup, SHEARON HARRIS - UNIT 1 3/4 4-33

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LIMITING CONDITION FOR OPERATION

3. 4. 9. 2 The Reac.or Coolant System (except the pressurizer) temperature and pressure, shall be limited in accordance with the limit lines shown on .Figures 3.4-2 and 3.4-3 during heatup, cooldown, and inservice leak and hydrostatic testing with: 1

a. A maximum heatup rate as shown on Table 4.4-6.

b. A maximum cooldown rate as shown on Table 4,4-6.

c. A maximum temperature change of less than or equal to 10'F in any 1-hour period during inservice hydrostatic and leak testing opera-tions above the heatup and cooldown limit curves.

APPLICABILITY: MOOES 4, 5, and 6 with reactor vessel head on.

ACTION: lilnibli'nbsshngrlo<

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R'gures 8.g-g and g.4-$ gyer8 With any of the above limits exceeded, res ore the temperature an'd/or pressure to wsthin 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 operation or maintain the RCS T and pressure at less than 200'F and 500 psig, respectively. avg SURVEILLANCE REOUIRBfENTS

4. 4. 9. 2. 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.

4.4.9.2.2 Oeleted.

SHEARON HARRIS - UNIT 1 3/4 4-34

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FIGURE 3.4-2 REACTOR COOLANT SYSTEM COOKS'HTTATIONS - APPLLCABLE UP TO 4 EFPY SHEARON HARRIS - UNIT 1 314 4-35

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FIGURE 3 '-3 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 3 EFPY SHEARON HARRIS - UNIT 1 3/4 4-36

NXIiMUH COOLGQW FOR'DOGES <.. ANO a 'a ANO HEATUP RATES

>H rc-A i V>>e,i HEAQ ON)

COOLQQiN RATES TEMPERATURE" COOLQO'AN IN ANY 1 HQUR PERIOO 350-20 4 504F ZOO- Z04F F II 0

/40-I/O F /toit=

HEAl P RATES TBSP ERATURE" HEATUP IN ANY 1 POUR PeRIOO 4F /35 104F is~-i<<&~&YF ~eA;F-

-3504F 504F 04 Is5-zoo 'F

+ 7~ mature used 4hiiuld ke based on /mes RQ< cold'lg

'ribbon '/pe0 u~e an opeia lian <+<

value heat m~

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in Or--"I o~hf v'em/el'~~.

Temperature r ange used should he based on the liest RCS cold leg value.

SHEARON HARRIS - UNIT 1 3/4 4-38

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REACTOR COOLANT SYSTEM OVERPRESSURE PRQTECTIQN SYSTBtS LIMITING CONOITION FOR OPERATION 3.4.9.4 At least one of the following Overpressure Protection Systems shall be OPERABLE:

-a. Two power operated relief valves (PORVs) with setpofnts which do not exceed the limits established fn Figure 3.4-4, or

b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2.9 square inches.

APPLICABILITY: bCOE 4 when the temperature af any RCS caid leg fs less than or equal to 4F, HOOK 5 and HOOK 6 with the reactor vessel head on.

325 ACTION;

a. With one PORV inoperable, restore the inoperable PORV to OPERABLE status within 7 days or depressurize and vent the RCS through at least a 2. 9 square inch vent within the next S hours.

b. With both PORVs inoperable, depressurfze and vent the RCS through at least a 2.9 square inch vent within 8 haurs.

c. In the event either the PORVs'r the RCS vent(s) are used to mitigate an 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 PQRVs or RCS vent(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 4.4.9.4. 1 Each PORV shall be demonstrated OPERABLE by:

a. Performance of an ANALOG CHANNEL OPERATIONAL'EST on the PORV actua-tion channel, but excluding valve operation, within 31 days prior ta entering a condition fn which the PORV is required OPERABLE and at least once per 31 days thereafter when the PQRV fs required OPERABLE;

b. Performance'f'a CHANNEL CALIBRATION on the PORV actuation channeI at least once per M months; and

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

SHEARON HARRIS UNIT 1 3/4 4-40

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500 CA l

400 300 200 300 400 MEASURED RCS TEMPERATURE < F)

RCS TEMP LOW PORV W HIGH PORV ~

OF PSIG 0 PSIG 6 F 100 383~F870 ~38 8 125 400 410 250 4~00 410

~300 440

< VALUES BASED ON KEFPY REACTOR VESSEL DATA AND CONTAINS MARGINS F AND +60 PSIG FOR POSSIBLE INSTRUMENT ERROR

-Ib FIGURE 3.4-4 NXIMUM ALLO%ED PORV SETPOINT FOR THE LQM TEMPERATURE OVERPRESSURE . SYSTE M SHEARON HARRIS - UNIT 1 3/4 4-41

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KHERGKNCY CORE COOLING SYSTB1S 3 4. .3 KCCS SUBSYSTEMS - T LESS THAN 0 F LIMITING CONOITION FOR OPERATION 3.5.3 As a minimum, one KCCS subsystem comprised of the following shall be OPKRABI ~:

a. One OPERABLE charging/safety injection pump,"

b. One OPERABLE RHR heat exchanger,

c. One OP'ERABLE RHR pump, and

d. An OPERABLE flow path capable of taking suction from the refueling

~ater storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: HOOK 4.

ACTION:

Nth no ECCS subsystem OPERABLE because of the inoperability'f either the charging/safety injection pump or the flow path from the refueling water storage tank, restore at least one KCCS subsystem to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLQ SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. Mith no ECCS subsystem OPERABLE because of. the inoperability of either the residual heat removal heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reac-tor Coolant System T v less than 350'F by use of alternate heat removal methods.

C. In the event the KCCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6. 9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated fot actuation cycles to date. The current value of the usage factor each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

shall be whenever 'the, "A maximum of one charging/safety injection pump OPERABLE temper ature of one or more of the RCS cold legs is less than or equal to SHEARON HARRIS

- UNIT 1 3/4 5-7

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l EMERGENCY CORE COOLING SYSTEHS SURVEILLANCE REQUIREMENTS 4.5.3. 1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2.

Deleted) 4.5.3.2 <4 SHEARON HARRIS - UNIT 1 3/4 5-8

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REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS (Continued) condition of the reactor and the additionaL restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single boron injection flow path becomes inopera'ble.

The Limitation for a maximum of one charging/safety injection pump (CSIP) to be OPERABLE and the SurveiLLance Requirement to verify all CSIPs except the required OPERABLE pump to be inoperable below ~'F provides assurance that a mass addition pressure transient can be reLieved by the operation of a single PORV. 9&

The boron'apability required .below 200'F is sufficient to provide the required SHUTDOWN MARGIN as defined by Specification 3/4.1 1.2 after xenon

~

decay and cooldown from 200'F to 140'F. This condition requires either 7100 gaLlons of 7000 ppm borated water be maintained in the boric acid storage tanks or 106,000 gallons of 2000"2200 ppm borated ~ater be maintained in the RWST.

The gallons given abo've are the amounts that need to be maintained in the tank in the various circumstances'o get the specified value, each value had added to it an allowance for the unusable volume of water in the tank, aLlowances for othez identified needs, and an allowance for possible instrument error. In addition, for human factors purposes, the percent indicated Levels were then raised to either the next whoLe percent or the next even percent and the gallon figures rounded off, This makes the LCO values conservative to the analyxed values. The specified percent level and gallons differ by less than 0.3Z ~

The Limits'n contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the soLution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimixes the effect of chloride and caustic stress corrosion on mechanicaL systems and components.

The BAT minimum temperature of 65'F ensures that boron solubility is maintained for concentrations of at Least the 7750 ppm Limit The RWST minimum temperature

~

is consistent with the STS value and is based upon other considerations since solubility is not an issue at the specified concentration leveLs. The RWST high temperature was seLected to be consistent with analytical assumptions for containment heat Load.

The OPERABILITY of one Boron Injection System during REFUELING ensures that

.this system is available for reactivity contr'oL while in MODE 6 ~

3/4 '.3 MOVABLE CONTROL ASSEMBLIES The spec'ifications of, this section ensure'hat: (1) acceptable power distri-bution Limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misaLignment on associated accident analyses are Limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

SHEARON HARRIS - UNIT 1 B 3/4 1-3 Amendment No.

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3/4. 4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS ANO COOLANT CIRCULATION The plant is designed to operate with all reactor coolant Iaops in operation and maintain ONBR abave 1.30 during all normal operations and anticipated tran-sients. In HOOES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANOBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

In MOOE.3, two reactor coolant loops provide sufficient heat removal capability for removing care decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e.; by opening the Reactor Trip System breakers. Single failure considerations require that two loops be OPERABLE at all times.

In HOOE 4, and in MOOE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for remov-ing decay heat; but single failure considerations require that at least two laaps (either RHR or RCS) be OPERABLE.

In NIOBE 5 with reactor coolant loaps not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat remov-ing component, require that at least twa RHR loops be OPERABLE.

The operation of one reactor coolant pump (RCP) or one RHR pump provides ade-quate flow ta ensure mixing, prevent stratification,and produce gradual re-activity changes during boron concentration reductions in the Reactar Coolant System. The reactivity change rate associated with boron reductian will, there-fore, be within the capability of operator recognition and control.

The restrictions-an starting an RCP with one or more RCS cold legs less than or B2s equal to 'F are provided to prevent RCS pressure transients, caused by energy additions fram the Secondary Caolant System, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against over pressure transients ynd will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 504F above each of the RCS cold leg temperatures-3/4.4.2 SAFETY VALYES The'pressurizer Code safety valves operate ta prevent the RCS fram being pres-surized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 380,000 Ibs per hour of saturated steam at the valve Setpoint. 'he relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition,against the Overpressure Protection System provides- a diverse means of protection RCS overpressurization at law temperatures.

SHEARON HARRIS " UNIT 1 B 3/4 4-1

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REACTOR COOLANT SYSTBI BASES SPECIFIC ACTIVITY Continued distinction between the radionuclides above and below a half-life of 15 minutes.

For these reasons the radionuclides that are excluded fram cansideration are expected to decay ta very Iow levels before they could be transported from the reactor coolant to the SITE 80UNOARY under any, accident condition-Based upon the abave considerations for excluding certain radfanuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample taking and completing the initial analysis is based upan a typical time necessary ta per form the sampling, transport the sample, and perform the analysis of about 90 minutes. After 90 minutes, the gross caunt should be made fn a reproducible gecmetry of sample and counter having reproducible beta or gamma self-shielding properties. The counter should be reset ta a reproducible efficiency versus energy. It is not necessary tc identify specific nuclides. The radiachemical determination cf nuclides should be based on multiple counting cf the sample within typical counting basis fallowing sampling of less than L hour, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about L day, about 1 week, and abaut 1 mcntti.

Reducing 7 avg to less than 5004F prevents the release of activity should a steam generator tube rupture occur, since the saturation pressure of the reactor cool-ant is below the lift pressure of the atmospheric steam relief valves. The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time ta take corrective action. A reduction in frequency of isotopic analyses following power changes may ba permissible if )ustified by the data abtained.

3/4.4. 9 PRESSURE/TEMPERATURE LIMITS and H The temperature and pressure changes during heatup and cooldcwn are limited ta be consistent with the requirements given in the AQIE Bafler and Pressure Vessel Cade, Section III, Appendix G, and 10 CFR 50 Appendix . 10 CFR 50, Appendix G also addresses the metal temperature of the closure head flange and vessel flange regions. The minimum metal temperature of the closure flange region should be at least 120 F higher than the limiting RT NOT far these regions when the pressure exceeds 20 (621 psfg for Mestfnghousi plants) of the preservice hydrastatic test pressure. For Shearcn Harris Unit 1, the minimum temperature of the closure flange and vessel flange regions is L20~F because the limiting RT NOT is 0 F (see Table B 3/4 4-1). The Shearcn Harris Unit 1 cooldawn and heatup Iimftatians shawn fn Ffgures 3.4-2 and 3.4-3 and Table 4.4-6 are not impacted by the 1204F lfmft.

L. The reactor coolant temperature and pressure and sys em cooldawn and heatup rates (with the exceptian of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 and Table 4.4-6 for the service period specified thereon:

a. Allowable,combinations of pressure and temperature for specific temperature change rates are below and ta the right of the limit I-ines shown. Limit lines for cooldown rates between those pre-sented may be obtained by interpolatfon; and SHEARON HARRIS - UNIT 1 . B 3/4 4-6

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'EACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS Continued

b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only.. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2. These limit lines shall be calculated periodically using methods pro-vided belo~,

3. The secondary side of the steam generator must not-be pressurized above 200 psig 704F, if the temperature of the >team generator is below 4, The pressurizer heatup and cooldown rates sha11 not exceed 100 F/h and 2004F/h, respectively. The spray shall'not be used if the tem-perature difference between the pressurizer and the spray fluid is greater than 6254F, and

5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness testing of the ferritic materials in the reacto~ vessel was performed in accordance with the 1971 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code; These properties are then evalua ed in accordance with the NRC Standard Review Plan.

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNOT', at the end of ffective full power years (EFPY) of service life. The M'+Sf service Iife~eriod is chosen such that the limiting RTNOT at the 1/4T location in the core region is greater than the RT NOT OT of the limiting unirradiated material. The selec ion of such a limiting RTNOT OT assures that all components in the Reac or Coolant System will be operated conservatively in accordance with applicable Code requireaents.

The reactor vessel materials have been tested to determine their initial RTNOT,

., the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resuItant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT . Therefore, an ad d reference temperature, based upon the copper content, and NOT'luence, content of the material in question, I can be predicted using Figure B 3/4.4-1 an the largest value of LBTNOT computed by ~~ 2 Regulatory Guide 1.99, Revision, "Effect of Residual E1ements on Predicted Radiation Ramage to Reacwr Vessei Ratariaisg SHEARQN HARRIS - UNIT 1 B 3/4 4-7

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t" REACTOR VESSEL TOtIGIIIIESS Cl AVCi. SIIEI.F LHERGY

'WD IIEAT CU, TNOT "TNDT I/HWD CQHPONENT GRADE ND ~F ~F fT-Ln FT" tlat Closure Ild. noae A533,0,CLl A9213-1 -10 8 ll4 Ilead Flange A508, CL2 5302-V2 135 I

Vessel Flange 5302"Vl -10 110 Inlet II Nozzle ll 4380-4 "20 "20 169 4380-5 0 0 120 ll ll 4388-6 "20 -20 149 Outlet Nozzle 4398-4 -10 -10 151 Il ll 4390-5 "10 <<]0 152 ll ll 4390-6 "10 -10 150 4J D

Nozzle Shell A5330,CL] C0224-1 .12 .OQS "20 90 N 11 CO]23-1 .12 .006 0 42 84 Inter. Shell A9153-1 .09 . 007 "]0 106 ll ll 84197-2 .006 -10

.10 112 74 Lower Shell ll 11 C9924-1 C9924-2

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.08

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005

-]0

-20 54 57 147 148 M ~

Bottom Ild. Torus A9249-2 -40 14 94 ll Il BOge A9213-2 -40 -8 125 Meld (Inter L Lower She'l Vert, ical Meld Seams)'- .06 . 013 "20 -20 >94 Meld (Inter. to Lower Shell Girth Seam) .04 . 013 "20 -20

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FIGURE 8 3/4.4-1 FAST NEUTRON FLUENCE (&IÃtV) AS 1 FUNCTION OF FULL POSER SERVICE LIFE SHEARON HARRIS UNIT 1 1 3/4 4 9

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FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (P'1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE SHEARON HARRIS - UNIT 1 B 3/4 4-9

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~5 REACTOR COOLANT SYSTEM 8ASES PRESSURE/TEHPERATURE tFig 3.4-R this shift in, RTN>T d .4-3 ~i LIMITS Continued as 1d The cooldown and heatup Ch d dj limits well as adjustments for possible errors in the pressure and temperature sensing instruments.

Values of 4RTNOT determined in this manner may be used until the results from the material surveillance program, evaluated according to ASTH E185, are available. Capsules will be removed and evaluated in accordance with the requirements of ASTH E185-82 and 10 CFR Part 50, Appendix H.. The results obtained from the surveillance specimens can be used to predict future radiation damage to the reacto~ vessel material by using the lead factor and the with-drawal time of the capsule. The cooldown and heatup curves must be..recalculated when the ~NOT determined from the surveillance capsule exceeds the calculated rGTNOT for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various cooldown and heatup rates are calculated using methods derived from Appendix G in Section IEI of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 5IQ '

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFH) technology.

En the calculation procedures a semielliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as. at the outside of the vessel wall.

The dimensions of this postulated crack, referred to in Appendix G of ASHE Section IjI as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure. To assure that the radiation embrittlemen+ effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference tempera-ture, RTNOT, is used and this includes the radiation-induced shift, &TNOT, corresponding to the end of the period for which cooldown and heatup curves are generated.

Tne ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K>, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K>R, for the SHEARON HARRIS " UNIT 1 B 3/4 4-11

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aASES PRESSURE/TEMPERATURE LIMITS Continued heatup and the time (or coolant temperature) along the heatup ramp. Further more, since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined. Rather, each heatup rate of interest must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve 'is constructed based on a point;by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limita-tions because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASIDE Code requirements.

LOW TBIPERATURE OVERPRESSURE PROTECTION The OPERABILITY of two PORVs or an RCS vent, opening of at least 2.9 square inches ensures that the RCS wiIT be protected from pressure transients which could exceed the limits of'ppendix G to 1 F art 50 when one or more of the RCS cold legs are less than or equal to 'F ~ ither PORV has adequate relieving capability to protect the RCS from overpressurization when the tran-sient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than 50~F above the RCS cold Ieg temperatures, or (2) the start of a charging/safety injection pump and its injection into a water solid RCS.

The maximum allowed PORV setpoint for the Low Temperature Overpressure Protec-tion System (LTOPS) is derived by analysis which models the performance with of the LTOPS assuming various mass input and heat input transients. Operation a PORV setpoint less than or equal to the maximum setpoint ensures that Appendix G criteria will not be violated with consideration for a maximum pressure over shodt beyon'd the PORV setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. To ensure that mass and heat input ransients more severe than those SHEARON HARRIS - UNIT 1 8 3/4 4-14

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the reactor vessel head fnstalled and disallow start af an RCP temperature is more than 504F above primary temperature.

The maximum a11owed

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Technical Speciffcatians require ackout af all but ane if secondary PORV setpoint for the LTOPS wf11 be updated based on the results of examinations of reactor vessel material irradiation surviillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule fn Table 4.4>>5.

3/4. 4. 10 STRUCTURAL- INTEGRITY The inservice inspection and testing programs far ASME Code Class 1, 2, and 3 components ensure that the structural integrity and aperatfanal readiness af these campanents 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 Cade and applicable Addenda as required by 10 CFR 50.55a(g) except ~here specific written relief has been granted by the Commis-'ion pursuant to 10 CFR 50.55a(g)(5)(f).

Components of the Reactor Coalant System were desfgned to provide access to permit fnservice inspectfons in accordance with Section XI of the ASME Boiler and Pressure Vessel Cade, 1977 Kditfan and Addenda through Summer 1978.

3/4.4. 11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncandensfble gases and/or steam from the Reactor Coolant System that could fnhfbft natural cft culatfon core,caoling. The OPERABILITY of least one Reactor Coolant System vent path fram the reactor vessel head and the pressurfzer steam space ensures that, the capability exists to perfarm this functfon.

The valve redundancy of the Reactor Coolant System vent paths 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 nat prevent isolation of the, vent path.

The function, capabflftfes, and testfng requirements of the Reactor Coolant System vents are consistent wfth the requirements of Itea II.8.1 of NUREG-0737, "Clarfffcatfon of TMI Actfon Plant Requfrements," November 1980.

SHEARON HARRIS UNIT I B 3/4 4 15

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3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5. 1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the. pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met:

The value of 66>> indicated level ensures that a minimum of 7440 gallons is maintained in the accumulators. The maximum indicated level of 96~ ensures that an adequate volume exists for nitrogen pressurization.

The accumulator power operated isolation valves are considered to be "operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of. power to the valves is required.

The limits for operation with an accumulator inoperable for any reason except an isolation 'valve closed minimizes the time exposure of the plant to a LOCA event occur ring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the..full capability of one accumulator is not available and prompt action is r'equired to place the reactor in a mode where this capability is not required.

3/4.5.2 ANO 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.

subsystem operating in conjunction with the accumulators is capable of 'ither supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period.

With the RCS temperature below 3504F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The limitation for a maximum of. one charging/safety injection pump to be OpERABLE and the Surveillance Requirement to verify one charging/safety injec-tion pump OPERABLE below F provides assurance that a mass addition pressure transient can be relieve by the operation of a single PORV.

325 SHEARON HARRIS - UNIT. 1 B 3/4 5-1

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