ML20058A469

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Proposed Tech Specs Re Low Temp Overpressure Protection
ML20058A469
Person / Time
Site: Calvert Cliffs Constellation icon.png
Issue date: 10/22/1990
From:
BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20058A468 List:
References
NUDOCS 9010260217
Download: ML20058A469 (36)
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Text

{{#Wiki_filter:.- i 0 ATTACilMENT (3) l BG&E Letter Ctted October 22, 1990 License Amet dment Request l Low Temperature Ovttpressure Protection i PROPOSED TECilNIC AL SPECIFICAT!ON Cil ANG ES f 1 l l.^ i l l l l= l 9010260217 901022 PDR ADOCK 05000310 p PNV

REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORAT10N SYSTEMS FLOW PATHS SHUTDOWN LIMLUNG CONDlHON FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths and one associated heat tracing circuit shall be OPERABLE: a. A flow path from the boric acid storage tank via either a boric acid pump or a gravity feed connection and charging pump to the Reactor Coolant System if only the boric acid storage tank in specification 3.1.2.7a is OPERABLE, or b. The flow path from the refueling water tank via either,-a charging pump or a high pressure safety injection pum(V to the Reactor Coolant system if only the refueling water tank in Specification 3.1.2.7b is OPERABLE. APPLICABillTY: MODES 5 and 6. Afil0H: With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one injection path is restored to OPERABLE status. SURVEILLANCE REQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE: a. At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path is above the temperature limit line shown on figure 3.11 when a flow path from the concentrated boric acid tanks is used, b. At least once per 31 days be verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise securea in position, is in its correct position. v v y At 305UF and less, the required OPERABLE HPSI pump shall be in pull-to lock and will not start automatically. At 305 F and less, HPSI 0 pump use will be conducted in accordance with Technical Specifica-tion 3.4.9.3 s _A --CALVERT CLIFFS - UNIT 2_ 3/41-8 Amendment No.

REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.1.3 At least one charging pump or one high pressure safety injection pumD* in the boron injection flow path required OPERABLE pursuant to Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus. APPLICABillTY: MODES 5 and 6. ACTION: With no charging pump or high pressure safety injection pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one of the required pumps is restored to OPERABLE status. EURVEILLANCE REQUIREMENTS 4.1.2.3 No additional Surveillance Requirements other than those required by Specification 4.0.5. 1 v N 1 At 305 F and less, the required OPERABLE HPSI pump shall be in pull-to-lock and will net start automatically. At 305 F and less, HPSI 0 pump use will be conducted in accordance with Technical.Specifica-tion 3.4.9.3 A ^ -CALVERT. CLIFFS -. UNIT 2 3/4 1-10 Amendment No.

's. TABLE 3 3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 9 G9 MINIMUM TOTAL MO. CHANNELS CHANNELS APPLICABLE P FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION ~ 6 G 1. SAFETYINJECTION(SIASj8 - a. Manual (Trip Buttons) 2 1 2 1, 2, 3, 4 6 C .5 b. Containment Pressure- [ High 4 2 3 1, 2, 3 7* c. Pressurizer Pressure - Low. 4 2 3 1,2,3(a) 7* 2. CONTAllWENT SPRAY (CSAS) a. Manual (Trip Buttons) 2 1 2 1, 2, 3, 4 6 g m b. Containment Pressure -- w ] High 4 2 3 1, 2, 3 11 3. CONTAllOIENT ISOLATION (CIS)# a. Manual CIS (Trip Buttons) 2 1 2 1, 2, 3, 4 6 b. Containment Pressure - High 4 2 3 1, 2, 3 7* N Containment isolation of non-essential penetrations is also initiated by SIAS (functional units 1.a and the RCS Tempera $ is: 8 (a) Greater than 350 F, the required OPERABLE HPSI pumps must be able to start automatically upon I receipt of g SIAS signal, 0 (b) Between 350 F and 305 F, a transition region exists where the OPERABLE HPSI pump will be placed in M pull-to-lock on a cooldown and restored to automatic status on a heatup, l (c) At 305 F and less, the required OPERABLE HPSI pump shall be in pull-to-lock and will not start 0 ~ I i automatically. A A .A ^

REACTOR COOLANT SYSTEM COOLANT LOOPS 8ND COOLANT CIRCULATf 0N HOT STANOBY I LIMITING CONDITION FOR OPERATION ~

3. 4.1. 2 a.

The reactor coolant loops Iisted below shall WOFE.hin,t.: s 1. Reactor Coolant Loop M4M#21)(and at least one associated reactor coolant pump. Reactor Coolant Loop 4M-(#22/ and at leaTt-Unt-associated reactor coolan)t pump. 2. b. At least one of the above Reactor Coolant Loops shall be in operation'. APPLICABILITY: MODE 3 #[ ACTION: a. With less than the above required reactor, coolant loops OPERABLE. restore the required loops to OPERABLE status within 72 hours or de in HOT SHUTOOWN within the next .12 hours. b. With no reactor coolan'. ioop in operation, suspend'all operations involving a reduction in boren concentration 3g i of the Reactor Coolant System and initiate corrective 4ction to return the required loop to operatido 410ii n one hour. SURVEILLANCE REOUIREMENTS 44.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 dafs by verifying correct breaker alignments and indicated power availability. 4.4*.1.2.2 ' At least one cooling loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours. t 'All reactor coolant pumps may be de-energized for u'p to l' hour (up to 2 \\ hours for low flow test) provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron cogcentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature gg gp Ay

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REACTOR COOLANT SYSTEM COOLANT LOOPS AND COOLANT CIRCULATION SHUTDOWN SUkVEILLANCE REQUIREMENTS 4.4.1.3.1 The required shutdown cooling loop (s), if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and intiicated power availability for pumps and shutdown cooling loop valves. 4.4.1.3.2 The required steam generator (s), if it is being used to meet 3.4.1.3.a. shall be determined OPERABLE by verifying the secondary side water level to be above -50 inches at least once per 12 hours. 4.4.1.3.3 At least one coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours. 1 CALVERT CLIFFS - UNIT 2 3/4 4-2b Amendment No. 38

REACTOR COOLANT SYST[M COOLANT LOOPS AND COOLANT CIRCULATION SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the coolant loops listed below shall be OPERABLE: 1. Reactor Coolant loop #21 and its associated steam generator and at least one associated reactor coolant

pump, 2.

Reactor Coolant loop #22 and its associated steam generator and at least one associated reactor coolant

pump, 3.

Shutdown Cooling loop #21*, 4. Shutdown Cooling loop #22*, b. At least one of the above coolant loops shall be in i operation **. APPLICABILITY: MODES 4***# and 5***#. ACTION: a. With less than the above required coolant loops OPERABLE, initiate corrective action to return the required coolant i loops to OPERABLE status within one hour or be in COLD SHUTDOWN within 24 hours, l b. With no coolant loop in operation, suspend all operations I involving a reduction in boron concentration of the Reactor i Coolant System and initiate corrective action to return the required coolant loop to operation within one hour. l. l The normal or emergency power source may be inoperable in MODE 5. All reactor coolant pumps and shutdown cooling pumps may be de-energized for up to I hour provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is 0 m intained at least 10 F below saturati m erature. A reactor coolant Tdl!Fp shall not be s arted w the RC3 N 0 temperature'less t1an or equal to 305 F unless (1) the i pressurizer water level is less than or equal to 170 inches, i and (2) the secondary water temperature of each steam generator i 0 l-( is less than or equal to 30 F above the RCS temperature, and 1 \\ (3L the pressurizer orassure is__less than or equal to 320 psia. l ) 5ee Special Test Exception 3.10.5^ N ^ 1 P CALVERT CLIFFS - UNIT 2 3/4 4 2a Amendment No. M. J

i REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS 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 line shown on Figure 3.4-2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with: 7f A maximum heatup of % 'F in any one hour period, a. b. A ma ooldown of in any on period wit "Y e 250' and a mum e oldown 20'F 'n any o g our, period th below O' ,yg c. A maximum temperature change of 5'F in any one hour period, during hydrostatic testing operations above system design 3

,\\.

pressure. APPLICABILITY: At all times. I ACTION: With any of the above limits exceeded, restore the temperature and/or F"- pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determ.ine. that the Reactor Coolant System remains acceptable for continued operations. or be in at least HOT STANDBY within the next 6 hours and reduce the RCS T and pressure to less than 200*F and MQ psia, respectively, within tAE9fo11owing 30 hours. 300 W h' % },- p + , g, g 4, 4 A << A 4, s s. n cs w& l 00'f 5 G M >/Po'F 40Y Y Q W,bW fY I Sc)*f 4 / qo *p /5fN M**JPM Y b Y 4 l + p * /= l CALVERT CLIFFS-UNIT 2 3/4 4-23 AUG 13 B/6

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titMAXIMUM PRESSSUREF M

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,-ny 4. \\( REACTOR COOLANT SYSTEM PRESSURI'ZER 4 LIMITING CONDITION FOR OPERATION c. -. 3.4.9.2 The pressurizer temperature shall be limited to:. A maxic.om heatup of 100'F. in any one hour'pe,riod,' a. b. A maximum cooldown of 200'F in any one hour period, and A maximum spray water temperature differential of 400',F. c. APPLICABILITY: At all 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; t(' of-limit condition on the fracture toughness properties of the pressurizer; perform an engineering evaluation to determine the effects of the out-detennine -that the pressurizer remains acceptable for continued operation or be in'at least HOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than,309-pg3g within the following 30 hours.. 3 p o p,' A SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer. temperatures shall be determined to be within the limits at least once per 30 minutes durfng system heatup or cooldown. The spray water -temperature differential shall be determined to be within the limit at least. once.per 12 hours during auxiliary spray operation. (. U CALVERT CLIFFS-UNIT 2-3/4 4-27

k l I REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDIT!0fi FOR OPERA ION 3. 9.3 At least one of the following overpressure protection s stems sha be OPERABLE: a. Two power operated relief valves (PORVs) with a ft setting of 450 psig, or b. Ar ctor coolant system vent of 1 1.3 squ e inches. APPLICABILITY: en the temperature of one or m e of the RCS cold legs is 1 275'F. v ACTION: a. With one PORV i erable, eith restore the inoperable PORV to OPERABLE status w hin 7 day or depressurize and vent the RCS through a 1 1.3 squ inch ent(s) within the next a hours; i maintain the RCS in a en d condition until both PORVs;have been restored to OPERAB status. b. With both PORVs inop able, epressurize and vent the RCS l through a-1 1.3 squ e inch v t(s) within 8 hours; maintain the RCS in a vent condition u il both PORVs-have been n restored to OPE LE status, c. In the event ther the PORVs or the CS vent (s) are used to mitigate a S pressure transient, a S cial Report shall be prepared a d submitted to the Commission ursuant.to Specifica-- tion 6.0 within 30 days. The report sha 1 describe the cir-cunista es initiating the transient, the of et of the PORVs or ve.(s) on the transient and any correctiv action necessary 'to event recurrence. -d. e' provisions of Specification 3.0.4 are not appl' cable. 4 4 __._x____

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION ~ ~ [3.4.9.3 The following overpressure protection requirements shall be met:. a. One of the following three overpressure protection systems shall be in place:

1.. ' Two power-operated relief valves (PORVs) with a lift setting of s 430 psia, or 2.

A single PORV with a lift setting of s 430 psia and a Reactor Coolant System vent of 21.3 square inches, or 3. A Reactor Coolant System (RCS) vent 2 2.6 square inches. b. Two high pressure safety injection (HPSI) pumps # shall be disabled by either removing (racking out) their motor circuit breakers from the electrical power supply circuit, or by . locking shut their discharge valves, c. The HPSI lcop motor operated valves (MOVs)# shall be prevented from automatically aligning HPSI pump flow to the RCS by placing their handswitches in pull-to-override, d. No more than one OPERABLE high pressure safety injection pump with suction aligned to the Refueling Water Tank may be used to inject flow into the RCS and when used, it must be under manual control and one of the following restrictions shall apply: 1. The total high pressure safety injection flow shall be -limited to 1 210 gpm OR 2. A reactor coolant system vent of 2 2.6 square inches shall exist. 0 APPLICABILITY: When the RCS temperature is 1 305 F and the RCS is vented to < 8 square inches. ALIIM: a. With one PORV inoperable, either restore the inoperable PORV to ' OPERABLE status within 5 days or depressurize and vent the RCS through a 2 1.3 square inch vent (s) within the next 48 hours; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status, b. With both PORVs inoperable, depressurize and vent the RCS through 'a 2 2.6 square inch vent (s) within.48. hours;' maintain the RCS-in a vented condition until either one OPERABLE PORV and a vent of 21.3 square inches has been established'or both PORVs have been restored to OPERABLE status. EXCEPT when required for testing. A-I ~ w_ x-CALVERT CLIFFS - UNIT 2 3/4 4-27a Amendment No. 5,

T g s REACTOR COOLANT SYSTEM LIMITING CDNDITION FOR OPERATION (Continued) v v v v v c. In the event either the PORVs or the RCS vent (s) are used to mitigate a RCS )ressure transient, a Special Report shall be prepared and su)mitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe L the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurrence, i d. With less than two HPSI pumps # disabled, place at least two HPSI pum) handswitches in pull-to-lock within fifteen minutes i j and disaale two HPSI pumps within the next four hours. 8 e. With one or more HPSI loop MOVs not prevented from automatically aligning a HPSI pump to the RCS, immediately place the MOV handswitch in pull-to-override, or shut and disable the affected MOV or isolate the affected HPSI header flowpath within four hours, and implement the action requirements of Specifications 3.1.2.1, 3.1.2.3, and 3.5.3, as l applicable. f.- With HPSI flow exceeding 210 gpm while suction is aligned to 'the RWT and an RCS vent of < 2.6 square inches exists, 1. Immediately take action.to reduce flow to less than or equal to 210 gpm. 2. Verify the excessive flow condition did not raise pressure above the maximum allowable pressure for the given RCS temperature on Figure 3.4-2b or Figure 3.4-2c. 3. If a pressure limit was' exceeded, take action in. L accordance with Specification 3.4.9.1. L g. The provisions of Specification 3.0.4-are not applicable. } A A A L k i a L EXCEPT when required for testing. e 3 t. k .4 .CALVERT CLIFFS UNIT 2 3/4.4-27b Amendmen', No. #,. l 4 ... o

s = REACTOR COOLANT SYSTEM EURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by: a. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, 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 PORV is required OPERABLE. b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months. c. Verifying the PORV isolation valve is open at least once per 72 g hours when the PORV is being used for overpressure protection, d. Testing in accordance with the inservice test requirements for ASME Category C valves pursuant to Specification 4.6.5 4,4.9.3.2 The RCS vent (s) shall be verified to be open at least once per 12 rs* g (s y)e tection. .4.9.3.3 All high pressure safety injection pumrs except the above c i OPERABLE pump, shall be demonstrated inoperable at least once per 12 hours by verifying that the motor circuit breakers have been removed from their electrical power supply circuits or by verifying their discharge valves are locked shut. The automatic opening feature of the high pressure safety injection loop MOVs shall be verified disabled at least once per 12 hours. \\hsvvwvv% i i 9 .b - Except when the vent pathway is locked,Qeiit'patN_a'fg open at sealedy or otherwise secured 1 in the open position, then verify these oncelper 31 days. ^^~~ m CALVERT CLIFFS-- UNIT'2 3/4 4-27c Amendment No. #,

) L EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE0VIREMENTS Each ECCS subsystem shall be demonstrated OPERABLh -4.5.2 a. At least once per 12 hours by verifying that the following valves are in the indicated positions with power-to the valve operators removed: 1: N -Valve Number Valve Function Valve Position L( l. MOV-659 Mini-flow Isolation Open 2. MOV 660 Mini-flow Isolation Open L 3. CV-306 Low Pressure SI Open Flow Control i o

b..

At least once per 31 days by: 1. Verifying that upon a Recirculation Actuation Test Signal, the containment sump isolation valves open. L~ 2. Verifying-that each valve (manual, power operated or automat c) in the flow path that is not locked, sealed, or m otherwise secured in position, is in itr, correct position. L c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment. sump and cause l restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed: 1. For all accessible areas of the containment prior to l establishing CONTAINMENT INTEGRITY, and 2. Of the areas affected within containment at the completion of containment entry when CONTAINMENT INTEGRITY is f. established. i d.- Within 4 hours prior to increasing the RCS pressure above 1750 j psia by verifying, via local indication at the valve, that CV-306 is open. i v y Whenever flow testing into the RCS is required at RCS temperatures 0 of 305 F and less, the high pressure safety injection pump shall l recirculate RCS water (suction from RWT isolated) or the controls of Technical Specification 3.4.9.3 shall apply. l A l :!,. l-CALVERT CLIFFS - UNIT _2-3/4 5-4' Amendment No. i

.. = EMERGENCY CORE COOLING SYSTEMS 0 ECCS SUBSYSTEMS - Tavg < 300 F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE: One# OPERABLE high-pressure safety injection pump, and l a. b. An OPERABLE flow path capable of taking suction from the refueling water tank on a Safety Injection Actuation Signal and i automatically transferring suction to the containment sump on a Recirculation Actuation Signal, 1 APPLICABILITY: MODES 3* and 4. ACTION: a. With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within I hour or be in COLD SHUTDOWN within the next 20 hours, b. In the event the ECCS 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 describing the circumstances of the actuation

and the. total accumulated actuation cycles to date.-

SURVEILLANCE RE0VIREMENTS l 4.5.3.1 The ECCS subsystem.shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2. s a p et e n a whe OPERABLE HPSI pump will be placed in pull-to-lock on a cooldown and restored'to automatic status on.a heatup. At 305 F and less, the required OPERABLE HPSI pump ghall be-in pull-to-lock and will not start' automatically. At 305 F and less, HPSI pump use' will be 'i i L conducted in accordance with Technical Specification 3.4.9.3. ~ L f L

CALVERT' CLIFFS ONIT 2 3/4 5-6 Amendment No. J%,

e lf l]Ii_ ' i

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 COOLANT CODES 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-.195 during all normal operations and anticipated transients. y A single reactor coolant loop with its steam generator filled above the low level trip setpoint provides sufficient heat removal capability for core cooling while in MODES 2 and 3; however, single failure consid-erations require plant shutdown if component repairs and/or corrective actions cannot be made within the allowable out-of-service time. In MODES 4 and 5, 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 be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two shutdown cooling loops to 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 restrictions on startina a tor Coolant Pump during MODEh 4 and 5 with the(TtCSJempefaturF < 3 5 F are provided to prevent RCS pressure transients, caused by energy additions-from the secondary system, which could exceod the limits of Ap e _ndix G to 10 CFR Part 50 J ' f see Bases 3/4 %.9) 7 r operation of the TIFactor Toolant T mps, The following criteria apply: (1) restrict the water volume in the pressurizer (170 inches) and thereby providing a volume for the primary coolant'to expand into.and (2) restrict starting of the RCPs to when the [L indicated secondary water temperature of each steam generator is less 0 than or equal to 30 F' above Reactor Coolant-System temperature,- and (3)' 1 limitsthe initial indicated pressure of the pressurizer ~ to less than or (equal to 3/4.4;2 SAFETY VALVES -The pressurizer. code safety valves operate to prevent the RCS from 1 being pressurized above its Safety Limit of 2750 psia.. Each safety valve 5 is-designed to. relieve approximately 3 x 10 lbs per hour of saturated a steam at the. valve setpoint. The 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 1 operating shutdown cooling loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.- 3-i During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized nove its safety limit of.2750 psia. The combined relief capacity of these valves is sufficient 3 to CALVERT CLIFFS ' UNIT 2 B 3/4 4-1 Amendment No. JE#E#E#E,

REACTOR COOLANT SYSTEM BASES o steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM and a concurrent loss of offsite electrical power. The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Calvert Cliffs site, such as site boundary location and meteorologi,:a1 conditions, were not considered in this evaluation. The NRC is fina*,izing site specific criteria which will be used as the basis for the r'.reva16ation of the specific activity limits of this site. ~This reev.tluation may result in higher limits. The ACTION statement pennitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 uCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 uCi/ gram DOSE EQUIVALENT I-131 but within

s the ifmits shown on Figure 3.4-1 must be restricted to no more than 10 percent of the unit's yearly. operating time since the activity levels allowed by Figure 3.a-1 increase the 2 hour thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture.

Reducing Tavg to < 500*F prevents the release of activity should a steam generator tube rupture since the. saturation pressure of the primary coolant is below the lift pressure of the_ atmospheric steam relief. valves. The surveillance requirements provide adequate-assurance that excessive specific activity levels in.the primary coolant will be detected in sufficient time to take corrective action.: Information obtained on iodine spiking will be used to _ assess. the parameters associated-with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. 3h9 PRESSURE / TEMPERATURE LIMITS All componen c+ o ant System are designed to with-stand the effects of cyc oads due temperature and pressure changes. These c' cads are introduced by no. transienes,- reactor tr nd-startup and shutdown operations. The variou.

  • egories v

AE I ' " CALVERT CLIFFS-UNIT 2 3 3/4 4-5 l l 1

3,,, i REACTOR COOLANT SYSTEM BASES r s -g load cycles used for design purposes are provided in Section 4.. of t FSAR. During startup and shutdown, the rates of temperature nd pres ure changes are limited so that the maximum specified hea

anc, coold rates are consistent with the design assumptions and atisfy the.str s limits for cyclic operation.

Durin heatup, the thermal gradients in the reactor vessel wall produce the 1 stresses which vary from compressive a the inner wall to tensile at e outer wall. These thennal induced ompressive stresses tend to allevia the tensile stresses induced by e internal pressure. Therefore, a pres re-temperature curve based on eady state conditions (i.e., no thennal s esses) represents a lower und of all similar curves for finite hea up rates when the inner all of the vessel is l 1 treated as the governi location. The heatup analysis a so covers the etermination of pressure-L temperature limitations for he case in hich the outer wall of the vessel-becomes the controllin locati The thennal gradients.estab- .1 lished during heatup produce te sile tresses at the outer wall of the L vessel. These stresses are addi v to the pressure induced tensile stresses which ye already presen. The thermal induced stresses at the outer wall of the vessel are te ile nd are dependent on both the rate of heatup and the time along t heat ramp; therefore, a. lower bound 4 curve similar to that descri d for the eatup of the inner wall cannot y L be defined.- Consequently, r the cases 'n.which the outer wall of the L vessel.becomes the stress ontrolling loca on, each heatup rate of - f interest must be analyze on an individual b is. 1. .The heatup and c 1down limit curves (Figu e 3.4-2) are composite curve _s which were pr ared'by determining the mo conservative case, with either the ins de or outside' wall controlling for any heatup or P cooldown rates of p:to 100*F.per hour. The heatup nd cooldown curves were prepared ba d upon the most limiting value of e predicted adjusted reference temp ature at the end-of the service. period indicated on Figure 3.4-2. The r ctor vessel materials have-been-tested to date ine their-initial R DT; the results'of these test are shown in Table: 3/4.4-1. Reactor-eration and resultant fast neutron. (E>l Mev) irradia ion will cause increase-.in the RINDT. Therefore.'an adjusted referen tempe ture, based upon the fluence can be predicted using Figure B 3/4. -1. The heatup and.cooldown limit curves shown on Figure 3.4 o in ude predicted adjustments for this shift in RTNDT at the end of t e, applicable service period, as well as adjustments for possible ~~ rrors in the pressure and: temperature sensing instruments. CALVERTJCLIFFS-UNIT 2 B 3/4 4-6 AUG 1 a sM i'

c 3_

(

^ + -f_; r-TABLE 8 3/4.4-1 l REACTOR VESSEL TOUGHNESS Mini pper li ' Shelf v energy Drop CHARPY V-NOTCH f omgitudina E LPC.-No. Code No. Heat No. Vessel Location Weight 930 ft-lb 950 ft-lb rection - ft-- a Reference Dw9. E 233-761-2 w 203-02 D-8903 58306 4P2710-4 -VI Vessel Flange +30* -48' -25* 158 204-02 D-8912 C5505-5 Bottom Head Dome -30" -43* 125 204-03'A-D-8911-3 C5176-3 Bo om Head Peel -20* -42* -08* 153 [ 8 D-8911-2 C5505-4 -20' -40* -11* 130 S C D-8911-2 C5505 -20' 0* -11* 130 [ D D-8911-1 C5505-3 -20* -34' -10* 131 -42* -08* 123 E D-8911-3 05176-3 F D-8911-1 C5505-3 -2 -34* -1 G* 131 205-02 A D-7203 AV3280-9A-9133 Inlet Nozzles -30* -62* -32* 130 i 8 D-7203-6 AV3288-9A-9254 -10* -65* - 132 l C 7203-5' AV3283-9A-9134 +20' -62* -52* 135 1 D D-7203-8 AV3285-9A-9253 -10' -50' CS* 126 205-03 A D-8920-3 AV3176-8L213 Inlet Nozzle 0 -10' 41 150 3 Extensions U 8 D-8920-2 AV317 2137 0* -10* 412* 150 j C D 8920-1 176-8L2136 0* -10* +12* 150 D D-8920-4 AV3176-8L2139 0* -10* 412* 159 .5 i 4 In l I _..,_..s_

m ~ ' j G TABLE B 3/4.4-1 (Cont'd) ' ~ - 2* ' REACTOR VESSEL TOUGHNESS ~P M nimm upper 3 Shelf Cv energy Drop CHARPY V-NOTCH for Longitudina Heat No. Vessel Location Weight 930 ft-lb 950 ft, Direction --ft-PC. No. . Code No. c-205-06 A D-7204-3 . 9- -001' Outlet Nozzles -20* -50' 132 8 D-7204-4 9-6512-0* -34* +06* 108 t 205-07 A D-8921-1 AV3282-8L21 ' Outlet Nozzle +10* -16* +15* 128 Extensions 8 D-8921-2 AV3282-8L2128 +10' -16* +15* 128 215-01 A 'D-8905-2 C5312-2 Upper 5 1 -46' +05* 133 8 D-8905-3 C5286-2 -20* -06* +16* 125 l ", C D-8905-1 C5312-1 -20* -40* ,20* 125 + fD 215-02 A D-8906-1 - A4463-1 Inte ate Shell -39' 432* 118 j

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8 D-8906-3 A4463-2 -10' +04* +35* 116 i C D-8906-2 B9427-2. +10' 3* +04* 126 215-03 A D-8907-1 C5804-1 . Lower Shell -10* 0 +28* 140 I 8 -D-8907-2 C52 -1 +10* -20' +14* 145 C' D-8907-3 5803-3 -20* .-18* + 130 B c-Ge ^ .g i .n ~.- .~

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~ M TABLE B 3/4.4-1 (Cont'd) E REACTOR VESSEL TOUGHNESS n r- -[ Shelf Cv energ) nimum upper -}~ Drop CHARPY V-NOTCH [ for Longitudint E -PC. No. Code No. I No. Vessel Location Weight 930 ft-lb 950 fFTb - Direction - ft- -i ' Reference Dwg. E 233-762-1 to - 209-02 -D-8904-1 3P2339-AZU-Closure Head +10* -68* -55* 185 lange 209-03 A D-8909-3 05389-3. Clo e Head Peels -10* -20* -02* 147 B D-8909 C5524-4 +14* +48* 135 C D-8909-2 A4700-2 -10' -5* +30* 130 D ~D-8909-1 C5524-4 0* +14* +48* 135 E D-8909-2 A4700-2 -5* +30* 130 g [ F D-8909 C5389-3 -10* -20* -02* 147 lo 209-04 08910-1 A4700-1 Closure Head Dome 0 +24* 123 b 9 m. o-a L3 4$ ~ 5

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REACTOR COOLANT SYSTEM BASES The actual shift in RT of the vessel material will be es' blished dically during operati$'by removing and eveluating, in cordance per witn TM E185-73, reactor vessel material irradiation surv lance i specime installed near the inside wall of the reactor v sel in the core area. Since the neutron spectra at the irradiatio samples and vessel insi radius are essentially identical, ne m ssured transition L shift for a s ole can be applied witn conficence tr :ne adjacent section of the reactor ssel. The heatup and c:oidown c.ves must be recalcu-lated wnen tne aR determined from the surve~ilance capsule is dif-ferent from the cat.Tlated 4RT for the equ alent capsule radiation ND7 exposure. The pressure-tempera. re limit iin snewn on Figure 3.4-2 for reactor criticality anc for #nservice 'aak and hydros ctic testing have ) been provided to assure conti, nce w'.h tne minimum temperature reovire-ments of Appendix G to 10 CFR 5 ihe maximum RT for all ceac r coolan':'s7 stem precsure-rs:aining materials,wi-h:ne'ybeeptic of :ne,anc:or :ressure vessel, has been determined to be 50*F. Th towest Serv *e Tempere ure limi-line ofSectionl'Iof'N.SinceArticleNE-2332 l asec upon this T. ( shown on Figure 3.4-2 is ASME Boiler and pressure q (Summer Addenda of 1972 Sebw nis ter:e ature, the :),.n + 100*F Service Tempe ature to se RT Vessel Cece requi-es t.e Lowes 5iem cressere l fer piping, pumps an valves. mt. s,be limited to mar,imum of E0', of the sys te... s hydrostatic tes i pressure of 3125 sia. The numbe of reactor vessel irradiation surveilla ce spe:imens and 1 the frequenti s for removing and testing these specirens re provided in Table 4.c-5 o assure ccmpliance witn the recuirements'of n endix H to-10.CFR art 50. and e ray water temperature differential are provided to assure tho' the ~ T iimitations imposed on the pressuri:er heatup and coo'id n rates pre surizer is operated within the design criteria assumed for the f *i-g analysis performed in accordance with the ASME Code re::uirements. l e OPERABILITY of two PORVs or an RCS vent opening of greater jh_ar f - l ~ 1.3 souar es ensures that the RCS will be protected from sure transients which exceed the limits of Appendix G,.t en Part 50 2 when one or more of the a d legs are < 27j5'J g i er PORV has adequate .f ra_ erpressurization when the-relieving capability to protect u transient is limited to either 0 -th' at " an idle RCP with the / measured by a surface:(e,yr-esof,.e s team generato secondary water temperat __ ' 'F (34*F when ) ntact ins: ument) aDeve the coo 4.. ' m erature ( in the rea 4Msel or (2) the start of a HPSI oump and its in;..eMagt in+ ater solio RCS. CALVERT CLIFFS-UNIT 2 B 3/4 a.11 Ar mdman: No. 15 m

a. REACTOR COOLANT SYSTEM BASES steam generator tube rupture accident in conjunction with an assumed steady state primary-to secondary steam generator leakage rate of 1.0 g>m and a concurrent loss of offsite electrical power. The values for tie limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Calvert Cliffs site, such as site boundary location and meteorological conditions, were not considered in this evaluation. The NRC is finalizing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site. This reevaluation may result in higher limits. The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's s)ecific activity >l.0 uti/ gram DOSE EQUIVALENT l-131, but within tie allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 uti/ gram DOSE EQUIVALENT I-131 but within the limits shown_on Figure 3.4-1 must be restricted to no more than 10 percent of the unit's yearly operating. time since the activity levels - allowed by Figure 3.4-1 increase the 2 hour thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture. 0 Reducing T to < 500 F prevents the release of activity should a steamgeneratorNberupturesincethesaturationpressureoftheprimary a coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity' levels in the primary coolant will be . detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtaiiled. 3/4.4.9 PRESSURE / TEMPERATURE LIMITS-W v v-Operation within the appropriate heatup and cooldown curves assures the integrity-of. the reactor vessel against fracture' induced by, combinative thermal and pressure stresses. As the vessel is subjected to l h increasing. fluence, the toughness of the limiting material continues to ( decline,.and even more restrictive Pressure / Temperature limits must be . observed E.The current limits,' Figures 3.4-2b and 3.4-2c, are for up_.to and including 12 Effective Full Power Years'(EFPY) of operation. The shift.in the material fracture toughness, as represented by is calculated using Regulatory Guide 1.99, Revision.2. For 12 .EFPY, aRTunT, t1the 1/4 T' position, the adjusted reference temperature (ART) l A A. sw 'CALVERT CLIFFS - UNIT 2 B 3/4 4-5 Amendment No.

1 REACTOR COOLANT SYSTEM -BASES 0 0 value is 171 F. At the 3/4 T position the ART value is 125 F. These values are used with procedures developed in the ASME Boiler and Pressure Vessel Code, Section III, A)pendix G to calculate heatup and cooldown limits in accordance with tie requirements of 10 CFR Part 50, Appendix G. To develop composite pressure-temperature limits for the heatup transient, the isothermal, 1/4 T heatup, and 3/4 T heatup pressure-temperature limits are compared for a given thermal rate. Then the most restrictive pressure temperature limits are combined over the complete temperature interval resulting in a composite limit curve for the reactor vessel beltline for the heatup event. To develop a composite pressure-temperature limit for the cooldown event, the isothermal pressure-temperature limit must be calculated. The L isothermal pressure-temperature limit is then compared to the pressure-temperature limit associated with a cooling rate and the more restrictive allowable pressure-temperature limit is chosen resulting in a composite limit curve for the reactor vessel beltline. l l Both 10 CFR Part 50 Appendix G and ASME, Code Appendix G require the development of pressure-temperature limits which are applicable to E inservice hydrostatic tests. The minimum tem)erature for the inservice hydrostatic tett pressure can be determined )y entering the curve at the test pressure (1.1 times normal operating pressure) and locating the corresponding temperature. This curve is shown for 12 EFPY on Figures 3.4-2b and 3.4-2c. Similarly,10 CFR Part 50 specifies that core critical limits be established based on material considerations. This limit is shown on the l heatup curve, Figure 3.4-2b. Note that this limit does not consider-the core reactivity safety analyses that actually control the temperature at which the core can:be brought critical. The Lowest Service Temperature is the minimum allowable temperature at-pressures above 20%' of the pre-operational system hydrostatic test l pressure (625 psia). This temperature is defined as equal to the most limiting RT for-the balance of the Reactor Coolant System components l plus 100 F,NDIr Article NB 2332 of Section III of the ASME Boiler and 0 p Pressure Vessel Code. The horizontal line between the minimum boltup. temperature and the-Lowest Service Temperature is defined by the ASME Boiler and Pressure Vessel Code as.20% of the pre-operational. hydrostatic test pressure. The change'in the line at 150 F on the cooldown curve is due'to a cessation of'RCP flow induced pressure deviation, since no RCPs are permitted to 0 operate during a cooldown below 150 F. k A. -w A A / t .CALVERT CLIFFS - UNIT 2 B 3/4 4-6 Amendment No. L

REACTOR COOLANT SYSTEM BASES v v w (,The minimum boltup temperature is the minimum allowable temperature at pressures below 20% of the pre-operational system hydrostatic test pressure. The minimum is defined as the initial RTNDT for the material of the higher stressed region of the reactor vessel plus any eti9 cts for irradiation per Article G-2222 of Section 111 of the ASME Boi16r &nd Pressure Vessel Code. The initial reference temperature of the rer.ctor i vessel and closure head flanges was determined using the certified I material test reports and Branch Technical Position MTEB 5-2. The head flange is 30 F,T associated with the stressed region of the closure maximum initial RTND 0 The minimum boltup temperature including 0 0 temperature instrument uncertainty in 30 F + 10 F = 40 F. However, for 0 conservatism, a minimum boltup temperature of 70 F is utilized in the analysis to establish the low temperature PORV lift setpoint. The design basis events in the low temperature region assuming a water solid system are: A RCP start with hot steam generators; and, An inadvertent HPSI actuation with concurrent charging. Any measures which will prevent or mitigate the design basis events are sufficient for any less severe incidents. Therefore, this section will discuss'the results of the RCP start and mass addition transient analyses. Also discussed is the effectiveness of a pressurizer steam bubble and a single PORV relative to mitigating the design basis events. The RCP start transient is a severe LTOP challenge for a water solid RCS. fherefore, during water solid operations all four RCPs are tagged out of service. Analysis indicates the transient is adequately controlled by placing restrictions on three parameters: initial pressurizer pressure and level, and the secondary-to-primary temperature difference. With these restrictions.in pl' ace, the transient is adequately controlled without the assistance of the PORVs. The inadvertent actuation of one HPSI pump in conjunction with one charging pump is the most. severe mass addition overpressurization event. . Analyses were performed for a single HPSI pump and one charging pump 2 assuming one PORV available with the existing orifice area of 1.29 in, For the limiting case, only a single PORV is considered available due to single _ failure criteria. A figure was developed which shows the calculated RCS pressures versus time that will occur assuming HPSI and charging pump mass inputs, and the expansion of the RCS following. loss of decay heat removal. Sufficient overpressure protection results when the i CALVERT CLIFFS --Ut.IT 2 B 3/4 4-7 Amendment No.

REACTOR COOLANT SYSTEM BASES b equilibrium pressure does not exceed thelimiting Appendix G curve ' ] pressure. Because the equilibrium pressure exceeds the minimum Appendix G limit for full HPSI flow, HPSI flow is throttled to no nore i e than 210 gpm indicated when the HPSI pump is used for mass addition. The HPSI flow limit includes allowances for instrumentation uncertainty, charging pump flow addition and RCS expansion following loss of decay heat removal. The HPSI flow is injected through only one HPSI loop MOV to limit instrumentation uncertainty. No more than one charging pump (44 gpm) is allowed to operate during the HPSI mass addition. i{ Comparison of the PORV discharge curve with the critical pressurizer pressure of 471.2 psia indicates that adequate protection is provided by 0 a single PORV for RCS temperatures of 70 F or above when all mass input is limited to 380 gpm. HPSI discharge is limited to 210 gpm to allow for one charging pump and system expansion due.to loss of decay heat removal. The low temperature PORV pressure lift setpoint is set to protect the I most restrictive Appendix G pressure limit (471.2 psia). A FORV setpoint I of 430 psia, which includes instrumentation uncertainties and sufficient margins for PORY response time requirements necessary for the protection of-471.2 psia, was selected. To provide single failure protection against a HPSI pump mass addition transient, the HPSI loop MOV hand 3 witches must be placed in pull-to-override so the valves do not automatically actuate upon receipt of a SIAS signal. Alternative actinns, described in the ACTION STATEMENT, are to disable the affected M0V (by racking out its motor (..rcuit breaker or equivalent), or to isolate the affected HPSI header. Examples of HPSI header isolation actions include; (1) de-energizing and tagging shut the HPSI header isolation valves; (2) locking shut and tagging all three HPSI pump discharge MOVs; and (3) disabling all three HPSI pumps. Three 100% capacity HPSI pumps are installed at Calvert Cliffs. Procedures will require that two of the three HPSI pumps be disabled I (breakers racked out) at RCS temperatures less than or equal to-305 F and L that the remaining HPSI pump handswitch be placed in pull-to-lock. Additionally, the HPSI pump normally in pull-to-lock shall be throttled .to less than or equal to 210 gpm when used to add mass to the RCS. I Exceptions are provided for ECCS testing and for response to LOCAs. l A pressurizer steam volume and a single PORV will provide satisfactory control of all mass addition. transients with the exception of a spurious b actuation of full flow from a HPSI pump. Overpressurization due to this. 0 I transient will be precluded for temperatures 305 F and less by disabling. two HPSI pumps, placing the third in pull-to-lock, and by throttling the ~ l third pump to less than or equal' to 210 gpm flow when it is used to add ~ mass to the RCS. Note that only the design bases events are discussed in detail since the less severe transients are bounded by the RCP start and inadvertent HPSI actuation analysis. A ~ A A

CALVERT CLIFFS - UNIT 2 B 3/4 4-8 Amendment No.

~ REACTOR COOLANT SYSTEM BASES I RCS temperature, as used in the applicability statementi is determined as ) follows: (1) with the RCPs running, the RCS cold leg temperature is the l appropriate indication, (2) with the shutdown cooling system in operation, the shutdown cooling temperature indication is appropriate, (3) if neither the RCPs or shutdown cooling is in operation, the core It exit thermocouples are the appropriate indicators of RCS temperature. g Y 1 s CALVERT CLIFFS - UNIT 2 B 3/4 4-9 Amendment No.

DELETED Y \\; CALVERT CLIFFS - UNIT 2 B3/44-10 Amendment No,

f4 p S DELETED I i CALVERT CLIFFS - UNIT 2 B 3/4-4-11 Amendment No. _ = _ _-_____ _ __- -_________ _-__ - -___-_____-____ ___

.t EMERGENCY CORE COOLING SYSTEMS BASES The trisodium phosphate dodecahydrate (TSP) store in dissolving baskets located in the containment basement is provided to minimize the possibility of corrosion cracking of certain metal components during operation of the ECCS following a LOCA. The TSP provides this protection by dissolving in the sump water and causing its final pH to be raised to 2 7.0. The requirement to dissolve a re)resentative sample of TSP in a sample of RWT water provides assurance t1at the stored TSP will dissolve in borated water at the postu' lated post LOCA temperatures. The Surveillance Requirements provided to ensure OPERABILITY of each component ensure that a minimum, the assumptions used in the safety analyses are met and the subsystem OPERABILITY is maintained. The surveillance requirement for flow balance testing provides assurance that L proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points p in accordance with the assumptions used in the ECCS-LOCA analyses, and l. (3) provide an acceptable level of total ECCS flow to all injection L points equal. to or aboyf _Lhat assumed in the ECCS-LOCA analyses. Minimum l-HPSI flow-requirements (for tempiratures a_boveWare based upon small l break LOCA calculations which credit charging punip flow following an i SIAS. Surveillance testing includes allowances for instrumentation and i system leakage' uncertainties. The 470 gpm requirement for minimum HPSI flow from the three lowest flow legs includes instrument uncertainties i but not system check valve leakage. The OPERABILITY of the charging pumps and the associated flow paths is assured by the Boration System Specification 3/4.1.2. Specification of safety injection pump total developed head ensures pump performance is consistent with safety l analysis assumptions. e-~- l g 0 P ( - At-temperatures of 305 F and less, HPSI injection flow is limited to l less than or equal to 210 gpm except in response to excessive reactor coolant.~ leakage.- With excessive RCS leakage (LOCA), make-up requirements could exceed a HPSI flow of 210 gpm. Overpressurization is prevented by controlling other parameters,- such as-RCS pressure and subcooling. This l provides overpressure protection in the low temperature region. An L analysis has:been performed which shows this flow rate is more than . adequate to meet ~ core coo _ ling safety analysis assumptions. HPSI pumps-are not required to auto-start when the RCS is in the MPT ~ enable- [ condition. The Safety Injection Tanks provide immediate injection of i borated water into the core in the event of an accident, allowing i . adequate-time for an operator to take action to start a HPSI pump. I Surveillance testing of HPSI pumps is. required to ensure pump operability.. Someisurveillance testing requires that the HPSI pumps )

1 deliver flow to the RCS.' To allow this testing to be done without 1

l increasing the potential for overpressurization of the RCS, either the RWT must be isolated or the HPSI; pump flow must be limited to less than or equal to 210 gpm orl an RCS vent greater-than or equal to 2.6 square inches must be provided. 4 n x ~~ CALVERT CLIFFS - UNIT 2 B 3/4 5-2 Amendment No. J%/EJ/J), j l i

i EMERGENCY CORE COOLING SYSTEMS BASES /4.5.4 REFUELING WATER TANK (RWT) The OPERABILITY of the RWT as part of the ECCS ensures that a M sufficient supply of borated water is available for injection by the ECCS '(f in the vent of a LOCA. The limits on RWT minimum volume and boron ' concentration ensure that 1) sufficient water is available within bp pS containment to permit recirculation cooling flow to the core, and 2) the c Ms reactor will remain subcritical in the cold condition following mixing of '9 the RWT and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical -Qharacteristics. 7 i: CALVERT CLIFFS - UNIT 2 B 3/4 5-2a Amendment No. JE/EJ/99, i i-um-- --um- -i-----}}

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