Text

Proposed Tech Specs Re Heatup/Cooldown Curves for RCS Pressure/Temp Limits
	ML20042H065
Person / Time
Site:	Calvert Cliffs
Issue date:	05/14/1990
From:	; BALTIMORE GAS & ELECTRIC CO.
To:	;
Shared Package
ML20042H062	List: ML20042H061; ML20042H065; ML20042H068;
References
	NUDOCS 9005170149
	Download: ML20042H065 (51)
	v • d • e

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. _. -. _ _ _.. - - -.. ~... FIGURE 3.4 20 CALVERT CLIFFS UNIT 1 HEATUP CURVE,12 EFPY REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS 2500 + HEATUPr _ j ~.. 5 INSERVICE N DROSTATIC TEST 5 q 2000 3 = 5 i k i 5 5 5 5 5 5 E 1500 __ LOWEST j j -' CORE! I E g _ SERVICE CRITICAL: g E TEMPERATURE Eif ~5 i E 160'F= '/ [ /' l1000 l l h 'e j f RCS TEMP. H/U RATE. 4 / / -{ 70T TO 3059 ' $609/1 HR f ./ 305T TO 3274 s109/1 HR f 2:327T 560T/1 HR 500 T MIN. BOLTUP TEMP. 70 *F MAXIMUM PRESSURE FOR SDC OPERATION 0 100 200 300 400. 500 600 INDICATED REACTOR COOLANT TEMPERATURE T ' T C The minimum boltup. temperature is the temperature i of the-reactor vessel flange, not the. coolant. temperature. .i i 1 4 s d 3/4 4-24

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~---- FOR SDC OPERATION ~ ~.. .== 1 O~ 100 200 300 400, 500 600 j INDICATED REACTOR COOLANT TEMPERATURE T ,'F i C The minimum boltup temperature is the temperature of the reactor vessel flange, not the coolant temperature. l t 3 h l J . ~ 3/.4 4 24a. ~..

i REACTOR COOLANT SYSTEM '3/4.4.9 PRESSURE / TEMPERATURE-LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION' l h 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit-lines shown on-Figure 3.4-2 during heatup, cooldown, _ criticality, and inservice leak and hydrostatic testing with: '1 A m,axi 'a. m Maximum Allowable Heatuo Rate RCS Temoerature g i 0 0 0 60 F in any one hour period 70 F to 305 F 0 0 0 10 F in any one hour period _- 305 F to 327 F 0 60 F in any one hour period 2 327 F j vve%%W 0 b.- A maximum cooldown of 100 F in any one hour period with T above 250 F and a maximum cooldown of 20 F in Lany one houf9 0 0 0 period with T below 250 F. avg 0 c. A maximum. temperature change of 5 F in any one hour period, during hydrostatic testing operations above system design-pressure. APPLICABILITY: At all times. ACTION: 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 fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant' System remains acceptable for continued operations or be in at least HOT STANDBY.within the next and reduce the RCS T 0 respectively,wi8inand.pressuretolessthan1200Fand300 psia, the following 30 hours, v SURVEILLANCE RE0VIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing-operations. 4.4.9.1.2 The reactor vessel material-. irradiation surveillance specimens shall' be removed and examined, to determine changes in material properties, at the intervals shown in Table 4.4-5. The results of these examinations shall be used to update Figure 3.4-2. i i 't CALVERT CLIFFS - UNIT 1 3/4 4-23 Amendment No.

1 REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited'to: 0 a. A maximum heatup of 100 f in any one. hour period, 0 b. 'A maximum cooldown of 200 F in any one hour' period, and 0 c.- A maximum spray water temperature differential of 400 F. . APPLICABILITY: At.all times. J ACTION:- With the pressurizer temperature limits.in excess of any of the above limits,. restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of-the-out-of-limit condition on the-fracture toughness properties of the pressurizer; determine that the pressurizer remains. acceptable for continued operation or be in at least HOT STANDB hi the:next 6 hours and reduce the pressurizer pressure to less tha 300 psia ithin the following 30 hours. SURVEILLANCE RE0VIREMENTS j 4.4.9.2 The pressurizer temperatures shall be determined.to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential'shall be determined to be within the limit at least once per 12 hours during auxiliary-spray operation. 4. i l 4 .l -CALVERT CLIFFS - UNIT 1 3/4 4-26 Amendment No.

[ \\ REACTOR COOLANT SYSTE!4 i ~ D P.PPESSURE PROTECTION SYSTEMS 'I S t. c. O k C M LIMITI CONDITION FOR OPERATION ',NI / i 3.4.9.3 At least one of the following overpressure protection syst s shall be OPE LE: I a. Two p er operated relief valves (PORVs) pith a' 'lif t etting ~ of 1 45 psig, or j .b. A reactor. lant system vent of > 1.3 square i hes.- hPPLICABILITY: When th temperature of one or more of heRCS[.goldlegs j is 1 275'F. ACTIO5:~ L a. With one PORY inopera le, either re.cre the inoperable.PORY to j OPERABLE status within days or pressurize and vent the RCS 1 through a > 1.3 square ch ven a) within the next B hours; ? maintain tWe RCS in a van dc dition until both PORVs have been restored to OPERAELE s us. b. With both PORVs inoperab , de ressuri:e and vent the RCS-through a > 1.3 square ch ven s)~within 8 hours; maintain the RCS in a vented condi'.on until b th PORVs have been restored l to OPERABLE status. c. In the event sit r the PORVs or the S vent (s) are used to i mitigate a RCS ressure transient, a Sp ial Report shall bs prepared and ubmitted to the Co::uission 'ersuant to 'S;:ccifice-tion.6.9.2 ithin 30 days. The report sha T describe the cir-cumstance initiating the transient, the of et of the PORVs er l vent (s) n the transient and any corrective a

ion necessary to preve recurrence, f.

d. Th provisions of Specification 3.0.4 are,not appl cable. p j c 1 i / j CALVERT CLIFFS - UNIT 1 3/4 4-26a Amendmant 2. -E

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OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9 e ements shall be met: a. One of the following three overpressure protection systems shall be in place: 1 1. Two power-operated relief valves (PORVs) with a lift [ setting s 424.5 psia or 2. A single PORV with a lift: setting of s 424.5 psia and a: reactor coolant system vent of 21.3 square inches, or l 3. A reactor coolant system (RCS) vent 2 2.6-square inches, b. Two high pressure safety injection (HPSI) pumps # shall'be M disabled by either removing'(racking out) their motor circuit breakers from the electrical power supply circuit, or_ by j I locking shut their discharge valves. { The HPSI loop motor operated valves-(MOVs)# shallibe prevented c. from automatically aligning.HPSI pump flow to the RCS:by - ( placing their hand switches in pull-to-override. 4 d. No more than one OPERABLE high pressure safety injection pump l 2 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 i control and one of the following restrictions shall apply: 1. The total high pressure safety injection flow shall be-limited to s 350 gpm OR 2. A reactor coolant system vent of 2 2.6' square inches'shall exist. APPLICABILITY: When the RCS temperature is 1 327 F and the RCS is vented 0 to < 8 square inches. ACTION: With one PORV inoperable, either restore. the' inoperable PORV to: a. \\ 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 m'S in a vented condition until either one OPERABLE PORV and a vunt of > 1 PORVs t' ave beeii re.3 square inches has-been established or both stored to OPERABLE status. % >~ w w w % CALVERT CLIFFS - UNIT 1 3/4 4-26a Amendment No. M, o

REACTOR COOLANT SYSTEM -LIMITINGCONDITIONFOROPERATIONiContinuedf c~~ ~ ~ - s%%4 c. In the event either:the PORVs or the RCS. vent (s) are used to- - mitigate a RCS pressure transient, a -Special Report shall_ be prepared and submitted to-the-Commission pursuant to S)ecification 6.9.2 within 30 days. The report shall describe t1e circumstances _ initiating the transient, the effect of.the PORVs or vent (s) on the transient and:any corrective action necessary to prevent recurrence.- d. With less than two HPSI pumps # disabled, place at least two HPSI put) handswitches in pull-to-lock within. fifteen minutes W disaale-two HPSI pumps within the next fourLhours, 'WithoneormoreHPSIloopMOVsfnot-preventedfrom e. 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,- d implement the action requirements of' Specifications-3.1.2.1, 3.1,2.3,'and 3.5.3", as ( applicable. I f. With HPSI flow exceeding 350 gpm while suction is aligned to j the RWT and an RCS vent of.< 2.6 square inches. exists, 1. Immediately take action to reduce flow to less than 350 gpm. 2. Verify the excessive flow condition did not raise pressure above the maximum allowable pressure for the 'given RCS l 1 temperature on Figure 3.4-2b or Figuret3.4-2a. 3. If a pressure limit:was exceeded,- take action-in - accordance with Specification 3.^4.9.1. i g. The provisions of specification 3.0.4 are not applicable. yl l l 1 EXCEPT when required for testing, i CALVERT CLIFFS - UNIT 1 3/4 4-26b Amendment No, M, k

J REACTOR COOLANT SYSTEM-SVRVEILLANCE REOUIREMENTS 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 b. Performance of a CHANNEL CALIBRATION on the'PORV-actuation. L channel'at least' once per 18 months. c. Verifying the PORV isolation valve 'is open at least once'per.72' r -hourswhenthePORVisbeingused.foroverpressure; protection. d. ' Testing in accordance with the inservice test requirements for. I ASME Category C valves pursuant to Specification 4.0.5 4.4.9.3.2 The RCS vent (s shall be verified to'be open at least once per-12 hours

wha N the ven -is being used for overpressure.orotection.

J enema 4.4.9.3.3 All high pressure safety injection pumps,iexcept the'above. OPERABLE pump, shall be demonstrated inoperable at least once per 12 y hours by verifying that the motor circuit breakers have been removed froe their electrical power supply r.ircuits or by verifying their discharge valves are locked shut. T" automatic opening. feature of the high. pressure safety injection < op MOVs shall be verified disabled at 'least i once per 12 hours. 1 WyvVVWW WU ) 1 t 1 I i i 1 N Except when the vent pathway is locked, eAe# otherwise secured in the open position, then verify thes vent pa w open at least once.per 31 days. a l 4 i CALVERT CLIFFS,- UNIT.1 3/4 4-26c AmendmentNo.j, 4

=r 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 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 i are conservative in that specific site parameters of:the Calvert Cliffs-l~' 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' ? e in higher limits. The ACTION statement permitting 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 i ) ( activity levels exceeding.1.0 pCi/ 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.. to < 500 F prevents the release of activity should a Reducing Ta steam generator Nbe 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 l that excessive-specific activity levels in the primary coolant will be d I. detected in sufficient time to take corrective action. Information L obtained on iodine spiking will be used to assess the parameters associated l with. spiking phenomena. A reduction'in frequency of isotopic analyses following power changes may be permissible if justified by the' data i obtained. 44JL9 PRESSURE / TEMPERATURE LIMITS ( l i All components ctor Coolant S . esigned'to with-stand the effects of cyclic 1 etem temperature and pressure changes. Thes a s are introduced by-no transients, 'i; yfac.tse o ps, and startup and shutdown operations. The varl ries CALVERT CLIFFS. UNIT 1-B 3/4 4-5 = = ~..

q 1 (see.cLbo.cd gu REACTOR COOLANT SYSTEM S of ica ycles used for design. purposes are provided in Section 4.2.1 f the FSAR. During startup and shutdown, the rates of, temperature and pressure c nges are limited so that the maximum specified heatup d cooldown ra s are consistent with'the design assumptions _and sat sfy j the stress li its for cyclic operation. During heat .-the. thermal-gradients in the reactor ves 1 wall produce thermal s esses which vary from compressive _at th inner wall to tensile at the o ter wall. These thermal. induced comp ssive. stresses: tend to alleviate th tensile stresses induced by the i ernal pressure. Therefore, a pressure-perature curve based on' stead state conditions (i.e.. no thermal stres s) represents a lower bound f all similar curves for finite heatup tes when the inner wall the vessel is treated as the governing 1 ation. The heatup analysis also overs the dete nation of pressure-temperature limitations for the ase in which he outer wall of the vessel becomes the controlling;1 ation.- T thermal gradients estab-lished during heatup produce tensi stres es at the outer wall of the vessel. These* stresses are additiv to e pressure induced tensile stresses which are already present.- thermal: induced stresses-at the outer wall of the vessel are tensile-are dependent.on both:the rate of heatup and the time along the he up mp; therefore, a lower bound-curve similar to that described fo the he tup of the inner wall a nnot' be defined. Consequently, for t casestin hich the outer wall of the vessel becomes the stress contr ling locatio each heatup rate of interest must be analyzed on individual bas -The heatup and cooldo limitcurves(Figure .4-2);are composite curves which were prepar by determining the most onservative case, with either the inside outside wall contro111 rig, r any heatup or-cooldown rates of up 100*F per hour..The heatup an 'cooldown curves-were prepared based on the most^ 1imiting value of the. redicted adjusted reference temperat e at the end of the service period in icated on Figure 3~.4-2. The. react vessel materials have been tested to determi their initial RT , the results of these test are shown in Table B -3 4.4-1. g Reactor ope tion and resultant fast neutron (E>l Mev) irradiati will i cause an crease in the RTNDT. Therefore, an adjusted reference tempera re, based upon the fluence can be predicted using Figure.B 3/4.4 The heatup and.cooldown limit curves-shown on Figure 3.4-2 inc1 e predicted adjustments for this shift in RTNDT at the end of-the policable service period, as well as adjustments-for possible er ors in the pressure and temperature sensing instruments. CALVERT CLIFFS-UNIT 1 B 3/4 4-6 j

[ ~ (n j v ? L, g f;_ 3 TABLE B 3/4.4-1 V nI-REACTOR VESSEL TOUGMESS n Minimum upper - Shelf. Cy energy E i) rop V-NOTCH for Longitudinal _ l Q PC. No. Code No. Hea Vessel Location Weight 930 f b 950 ft-lb Direction - ft-Ib-1 Reference Dwg. E 233-426-1 203-02 -D-7201 123W171VA1 Vessel Flange -84* .-40* 153 l L 204-02 D-7212 .B8504-4 tton Head Dome -20' +5* +21* 135 204-03 A D-7211-3 A3641-3 Bott Head Peel -30* -26* +09' 136 B D-7211-2 A3641-4 -30' -20' +20* 147 l ];2 C D-7211-3 A3641 - -30*- -26* +09' 136 D D-7211-2 A3641-4 -30* -20' +20' .147 1 { E D-7211 -1 B8504 30* ' -35' -14* 140 F D-7211-1. B8504-1 -30 -35* -14' 140; 205-02 A D-7203-4 -AV2358-N8 43 Inlet Nozzles -10' -43* -15* 124 -10* 118 B D-7203-1 AV23 - 8C-6384 0* C D-7203-2 330-N8C-6385 0* -38* -15

123 l

D-b-7203-AV2354-N8C-6442 -20' -60* 140 205-03 A D ~' 0-4 AV3176-8L-2135 Inlet Nozzle Extensions .O' ' 12* +16* 150 ,B -7220-1 AV3176-8L-2132 0* -12* +16* 150 D-7220-2 AV3176-8L-2133 .O -12* +16* ~ D D-7220-3 AV3176-8L-2134 10' . -12 * - +16' 150 ~

m w l m S o3 i P ~ TABLEB3/4.4-1(Cont'd)_ -e REACTOR VESSEL TOUGHNESS j l 7 Minimum upper j

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^ . Shelf Cv energy - j 2 Drop . CHARPY -NOTCH for Longitudinal U PC. No. . Code No. a t No. Vessel Location Weight 930 ft-950 ft-lb Direction - ft-lb d 205-06 A D-7204 9-591 1 Outlet Nozzles - +10* -33* -Ol*' 115 B D-7204-1 9-5949-002 0* -46* -10* ' 101 205-07 A D-7221-2 AV3282-8L-2131 Outlet Nozzles -12* +14* 128 xtensions B D-7221-1 AV3282-8L-2126 215-01 A D-7205-3 C4420-2 Upper Shel +10* -51* +04* 135 B D-7205-1 C4423-2 +10* ~ +10* 442*- 120 3 C D-7205-2 C4679-2 +10* -38* +03* 119 f 215-02 A D-7206-2 C4441-2 ntermediate Shell -35* -04* . 125 o B D-7206-3 C4441-1 0* -09' +37' . 128-C D-7206-1 C4351-2 -10* + +48* 139 215-03 A -D-7207 C4420~. Lower Shell 0* --28' +06* 118 B. D-7207-3 89-1 -20*' -33' 125 C D-7207-2' . B8489 -10* -38* -06* 139 1 .u r ..mu. 7,-

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A3641-2

+10* -16* +16* 133 D D-7209-1 A3641-1 0* -14* +52*, 115 209-03 E D-7209-3 B85043- -10 -10" +26* 135-w F D-7209-1 A3641-1 0* 14* +52*- 115 'o 209-04 D-7210 C3376-1' Closure Head Dome -20* -16 +42* 110 4 - i s I t 4 -...,..m.

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lgy j I PT5 t s REACTOR COOLANT SYSTEM BA S s The ctual shift in RT of the vessel material will be estab shed periodica yduringoperatiggTby removing and evaluating,.in accor ance with ASTM 5-73, reactor vessel material irradiation surveilla e speci-mens install near'the inside wall of the reactor vessel in th core area. Since t e_ neutron spectra at the-irradiation samples a vessel inside radius a essentially icentical, the measured'transi ion shift for a sample can

e. applied with confidence to the adjacen section of the reactor vessel.- T heatup and cooldown ' curves must be r calculated when the ART r dete..ined from the surveillance capsul is-different fromthecal$0 lated 4*DT_orjheequivalentcapsule adiation exposure.

The pressure-temper 'ure limit lines shown on igure 3.4-2 for reactor criticality and fo inservice leak and ny estatic testing have been provided to assure com liance with the min..um temperature require-i ments of Appendix G to 10 CF 50. l The maximum RT for all actor cool nt system pressure-retainiilg materials,withtheNceptionof. e reac N r pressure vessel, has been g determined to be 50*F.- The Lowest 'ervi e Temperature limit line shown on Figure 3.4-~2 is based-upon-RT since Article NS-2532 (Summer Addenda of 1972) of Section t. of N ASME Soiler and Pr-essure vessel Code requires the Lowest Ser' ice Temperature to be RT,wthistempera for piping, pumps and valves. Be em pressure i must be limited to a maximum of- % of ths ystem's hydrostatic test pressure of 3125 psia. The number of reactor ,.ssel irradiation s rveillance specimens and tne frequencies for removi g and testing these s cimens are provided in Table 4.4-5 to assure co liance with the recuire., nts of Appendix H to 10 CFR part 50. I The limitations imposed on the pressurizer heatup nd cooldown rates and spray water te erature differential are provided-t assure that ne pressurizer is oo rated within the design criteria assume ' for the fa:i-gue analysis pe.ormed in accordance with the ASME~ Code re irements. The OPE BILITY of two PORVs or an RCS vent ooening of g

ter tnan 1.3 square.nches ensures that the RCS will be protected from p. essure transient which could exceed the limits of Appendix G to 10 CFR art 50 t

when on r more of the RCS cold legs are 1 275*F. Either PORV has adequate reliev# g cacability to protect the RCS from overpressuri:ation w e. the tran ent is limited to either (1) -he start of an idle RCP with :ne sec ndary water temperature of the steam generator 1 46*F (3c'F wnen m sured by a surface contact instrunent) above the coolant temperature j 'n the reactor vessel or (2) ne star of a MPSI ou o anc its 'njecion 4 into a water solid RCS. i CALVE?.T CLIFFS VMT 1 E 2/4 4-11 M endmen:.!c. 24

L 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 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 ano 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 specific activity'>1.0 i uCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on i Figure 3.4'-1, accommodates possible iodine spiking phenomenon which may I occur following changes in THERMAL POWER. Operation with specific 1 activity levels exceeding 1~.0 uCi/ 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-j 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 l tube rupture. 0 Reducing T to < 500 F prevents the release o'f act'ivity should a. steam generator U be rupture since the-saturation pressure of the. primary a coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance-requirements provide adequate assurance that-excessive rpecific activity levels in the-primary coolant will be detected in sufficient time to take corrective action. 'Information 1 obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic _l analyses following power changes may-be permissible if ~ justified by the =j data obtained. 3/4,4.9 PRESSURE / TEMPERATURE LIMITS I mm< w nmn Operation within the appropriate heatup and cooldown. curves assures-the integrity of the reactor vessel against fracture induced:by j combinative thermal and pressure stresses. As the vessel is subjected to ( increasing fluence, the toughness of ~the limiting material continues to decline, and ever more restrictive Pressure / Temperature limits must be observed. The current limits, Figures 3.4-2a and 3.4-2b, are:for up to and including 12 Effective Full Power Years (EFPY) of operation. The shift in the material fracture toughness, as represented by RT nT, t the 1/4 T position, the-adjusted reference-temperature-(ART N is calculated using Regulatory Guide 1.99, Revision 2. For 12 EFPY, a mvWwv i CALVERT CLIFFS - UNIT 1 B 3/4 4-5 -Amendment No. l

REACTOR COOLANT SYSTEM BASES ,_ q 0 ~ 0 value is 222 F. At the 3/4 T position the ART value is 162.5 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 f limits in accordance with tie requirements-of 10 CFR Part 50,- Appendix G. l ) ) To develop composite pressure-temperature limits for the heatup j transient, the isothermal,1/4 T heatup, and 3/4 T heatup pressure-1 j. temperature limits are compared for a given thermal rate. Then the most 4 restrictive pressure-temperature limits are combined over the complete temperature interval resulting in a composite limit curve for the reactor j vessel beltline for the heatup event. To develop a composite pressure-temperature limit for=the cooldown event,- 4 the isothermal pressure-temperature limit must be calculated. The isothermal. pressure-temperature limit is then compared to the pressure-temperature limit associated with a-cooling rate and the more restrictive i allowable pressure-temperature limit is chosen resulting in a composite limit curve for the reactor vessel beltline. j Both 10 CFR Part 50 Appendix G and 7SME, Code Appendix G require the-i development of pressure-temperature limits which are applicable to ) inservice hydrostatic tests. The minieum temperature for the inservice-l T hydrostatic test pressure can be detereined by entering the curve at the I test pressure (1.1 times normal operating pressure) and locating the { i corresponding temperature. _This curve it shown for,12 EFPY on' Figures j 3.4-2a and 3.4 2b. Similarly,10 CFR Part 50 specifies that core critical limits be-established based on material considerations. This limit is shown on the i heatup curve, Figure 3.4-2a. Note that this limit does not consider the 3' l core reactivity safety analyses that actually' control the temperature-at } which the core can be brought critical, i l The Lowest Service Temperature is the minimum allowable temperature at i j pressures above 20% of the pre-operational system hydrostatic test i pressure (625 psia). This temperature is defined as equal to. the most limiting RT for the balance of the Reactor Coolant System components plus 100 F,NDT 0 per Article NB 2332 of Section III ~of the ASME-Boiler and 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-ocerational hydrostatic test pressure. 3 The change in the line at 150 F on the cooldown curve is due to a cessation 0 of RCP flow induced pressure deviation, since no RCPs are permitted to operate during a cooldown below 150 F. 0 w wv I d CALVERT CLIFFS - UNIT 1 B 3/4 4-6 Amendment No.

t + REACTOR COOLANT SYSTEM BASES gem -- v y v ~ v w n The minimum boltup temperature is the minimum allowable temperature at pressures below the 20% of the pre-operational system hydrostatic-test-pressure. The minimum is defined.as the initial RT nT for the material N I of the higher stressed region of the reactor vessel plus' any effects for \\ irradiation per Article G-2222 of Section III of the ASME Boiler and-Pressure Vessel Code.. The initial. reference temperature of the reactor vessel and closure head flanges was determined using the certified material test reports and Branch Technical Position MTEB 5-2. The. maximum initial RT nT associated with the stressed region of the closure 0 head f1ange_is -10 F, -The minimum boltup temperature including 0 0 0 temperature instrument uncertainty is -10 F + 10 F - 0 F. However, for 0 conservatism, a minimum boltup temperature of 70 F is utilized. k. The design basis events in the low temperature region assuming a water solid system are: j T A RCP' start with hot steam generators; and, - - 1 .t An inadvertent HPSI actuation with concurrent charging. ( .Any measures which will prevent or mitigate the design basis events are. f sufficient for any less severe incidents. Therefore,.this section will discuss the results of the RCP-start and mass addition-transient j \\ 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. [.. Therefore, during water solid operations all 4 RCPs are tagged out of: service. However, analysis indicates that the transient:is adequately i mitigated by. a pressurizep steam bubble..The approximate size of the-required volume is 820 ft (170 in, water level).. The steam bubble allows the operators at least 10 minutes to initiate corrective action. Other overpressurization incidents which are mitigated by maintaining a-1 pressurizer bubble are: l - Charging pump input with insufficient letdown; Letdown isolation; Shutdown cooling isolation; and, Inadvertent actuation of all pressurizer heaters. The inadvertent actuation of one or more HPSI pumps in conjunction with j up to three charging pumps is the most severe mass addition i i overpressurization event. Analyses were performed for a single HPSI aump. and either one or 3 charging pumps assuming one PORV available'with t1e 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 that occur when the discharge of one PORV reaches equilibrium with the HPSI and charging-i- / mass. inputs. Sufficient overpressure protection.results when the ( equilibrium pressure does not exceed the limiting Appendix G curve Ewuw%%-vv CALVERT CLIFFS - UNIT 1 B 3/4 4-7 Amendment No. D

REACTOR COOLANT SYSTEM BASES j ~%e pressure. Because the_ equilibrium pressure exceeds the LTOP PORV setpoint, HPSI flow is throttled to no more than 350 gpm when the HPSI pump'is used for mass addition. ~No more than one charging pump (45 gpm)' .is allowed to operate during the HPSI mass addition. ~ Comparison of the PORY discharge curve with the critical pressurizer pressure of 424.5 psia indicates that adequate protection is-provided by: ,l 0 a single PORV for RCS temperatures above 70 F when all 'massiinput is ~1 limited to 470 gpm. - HPSI discharge is limited to 350 gpm to allow for one-charging _ pump and system expansion due to decay heat, j \\ To provide single' failure' protection against a HPSI pump. mass addition q l transient, the HPSI loop M0V handswitches must be placed in-j pull-to-override so the. valves do not automatically actuate upon receipt j of a SIAS signal. Alternative actions, described in the. ACTION STATEMENT, are to disable the affected M0V (by racking out its motor circuit breaker or ecuivalent), or to isolate the affected HPSI header. Examples of HPSI. heacer isolation actions include; (1) de-energizing and' tagging shut the HPSI header isolation valves; (2) locking shut and j tagging all three HPSI pump discharge MOVs; and'(3) disabling all' three i HPSI pumps. l Three 100% capacity HPSI pumps are installed at Calvert Cliffs. k Procedures will require that two of.the three HPSI pumps be disabled i j (breakers racked out) at RCS temperatures-less than or-equal to 327 F and f-0 t' that the remaining HPSI pump handswitch be placed~in pull-to-lock. Additionally,-the HPSI pump normally in pull-to-lock shall be throttled / -l to less than 350 gpm when used to add mass to_the RCS. Exceptions are ) (_ provided for ECCS testing and for response-to LOCAs.- \\ A pressurizer steam volume or a single PORY will provide satisfactory / control of all transients with the exception of a spurious. actuation of l full flow from a HPSI pump. Overpressurization due to this transient will ) be precluded for temperatures below 327 F by. disabling. two HPSI-pumps, 0 placing the. third in pull-to-lock, and'by throttling the third pump to- \\ less than 350 gpm flow when it is used to add mass to the RCS. RCS temperature, as used in the applicability' statement, is determined-as-l follows: (1).with the RCPs running, the RCS cold leg temperature is the i appropriate indication, (2) with the shutdown cooling. system in-operation, the shutdown cooling temperature indication is appropriate, a l i l (3) if neither the RCPs or shutdown cooling is in operation, the core exit thermocouples are the appropriate indicators of RCS temperature. l ic CALVERT CLIFFS - UNIT 1 B 3/4 4-8 Amendment.No. t I;

l' L l REACTOR COOLANT SYSTEM l BASES l e. .Nhn%nnm 1 pressure. Because the equilibrium pressure exceeds the LTOP PORV setpoint, HPSI flow is throttled to no more than 350 gpm when.the HPSI (pumpisusedformassadditi9n.'Nomorethanonechargingpump(45.gpm) is allowed to operate during the HPSI mass addition. 3 Comparison of the PORV disch'arge-curve with the critical pressurizer pressure of 424.5 psia indicates that'adequgte protection is provided by a single PORY for RCS temperatures above 70 F when all' mass input is limited to 470 gpm. HPSI discharge is limited to 350 gpm to allow for.- one charging pump and system expansion due to decay' heat.- 1 \\ To provide single failure protection against a HPSI pump mass addition transient, the HPSI loop M0V handswitches must be placed in pull-to-override so the valves do not automatically actuate upon receipt of a SIAS signal. Alternative actions,L described in the ACTION ~- STATEMENT, are to' disable the affected MOV- (by racking out its motor J Examples of HPSI header isolation actions include;- (1) de-energizing an circuit breaker or equivalent), or to isolate the affected HPSI header. l 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 ) umps are installed at Calvert Cliffs. .[ Procedures will rcquire tlat two of the three HPSI pumps be disabled (breakers racked-out).at RCS temperatures less than or equal to 327 F and 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 350 gpm when used to add mass to the RCS. Exceptions are ) provided for ECCS testing and for: response to LOCAs. A pressurizer steam volume or a single PORV will provide satisfactory j .i / control of all transients with the exception of a spurious actuation of full flow from a HPSI pump. Overpressurization due to this transient will ) be precluded for temperatures below-327 F by disabling two HPSI-pumps, 0 placing the third in pull-to-lock, and by throttling the third pump to \\ less than 350 gpm flow when it is used to add mass to the RCS. / l RCS temperature, as used in the applicability statement, is determined.as follows: (1) with the RCPs running, the RCS cold leg temperature is the appropriate indication, (2) with the. shutdown cocling' system in. s operation, the shutdown. cooling-temperature indication is appropriate, I (3) if neither the RCPs or shutdown cooling is in operation, the core exit thermocouples are the appropriate indicators of RCS temperature. L l www CALVERT CLIFFS - UNIT 1 B 3/4 4-8 Amendment'No. N

.: t t f i -f DELETED -) a i i t i i J t t [ E 1 l CALVERT CLIFFS - UNIT 1 B 3/4 4-9 Amendment No.

r:. 1 C DELETED i l 4 1 i l i !'l 1 q 4 i i CALVERT CLIFFS - UNIT 1~ B3/4.4-10 Amendment No.

REACTOR COOLANT SYSTEM BASES .I DELETEDJ i i .i l f I .i t -i i 1 I i i 0-CALVERT CLIFFS - UNIT 1 B 3/4 4-11 Amendment No. Jf, m

- REACTOR COOLANT SYSTEM BASES 1 DELETED ) J 1 i i I i l CALVERT CLIFFS . UNIT'l B 3/4 4-11 Amendment.No. Jf, s

l t REACTOR COOLANT SYSTEM ~ COOLANT LOOPS AND COOLANT CIRCULATION l SHUTDOWN a LIMITING CONDITION FOR 0PERATION ~ 3.4.1.3 a.. At least two of the coolant loops listed below shall be i OPERABLE: I 1. Reactor Coolant Loop #11-(#21) and its associated steam generator and at least one associated reactor coolant i pump. 2. Reactor Coolant Loop #12 (#22) and its associated steam generator and at-least one associated reactor coolant

pump, 3.

Shutdown Cooling Loop #11 (#21)*, 4. Shutdown Cooling Loop #12 (#22)*. b. At least one of-the above coolant loops shall be in operation **. L APPLICABILITY: MODES 4***# and 5***#, ACTION: a. With less than the-above required coolant loops OPERABLE, initiate corrective action to return.the required coolant loops ( to OPERABLE status within one hour to be in COLD: SHUTDOWN within 24 hours. i b. With no coolant loop in operation, suspend:all operations-involving a reduction in boron concentration of the Reactor-l_ Coolant System and initiate corrective -action to return the required coolant loop-to operation within~one hour. 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 )ermitted that would cause dilution of the reactor coolant system -)oron concentration, and (2) core outlet temperature is' maintained at-least 10 F below saturation temperature.1 . gto 327 F unless (! Rr kd g rAagoAcooa jlot be,_. ture-s less than or equa

1) the pressurizer indicated water level is less than 170 inches, and (2) the secondary water t mper ture of each steam generator is less than 150 F above the:RCS.

e peratur. See-Special Test Exception 3.10.5. i CALVERT CLIFFS - UNIT 1 3/4 4-2a Amendment No. M, i

REACTOR COOLANT SYSTEM v i COOLANT LOOPS AND COOLANT CIRCULATION SHUTDOWN -i SURVEILLANCE REQUIREMENTS 4 l' 4.4.1.3.1 1The required shutdown cooling loop (s), if not in operation, ydh shall be determined OPERABLE once per 7 days by verifying correct breaker rog Oojn 90i. alignments;and indicated power availability for pumps and ,t> 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. 3 l . -j i ( il'

i r

t. l-CALVERT CLIFFS - UNIT 1 3/4 4-2b Amendment No. )),-

t 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 COOLANT CODES AND COOLANT CIRCULATION [ The Ant is designed to operate with both reactor coolant loops and i associated reactor coolant pumps in operation, and maintain DNBR above i 1.195 during all normal operations and anticipated transients. 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 considerations require plant shutdown if component repairs and/or corrective actions cannot be made within the allowable out of-service time. i In MODES 4 and 5, a single reactor coolant loop or shutdown cooling loop provides sufficient heat romoval capability for removin decay heat; butsinglefailureconsiderationsrequirethatatleasttwokoopsbe 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 1 I 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 &tiqK ctor Coolant Pump during MODES 4 and 5 wit e RCS tem a ure re provided to prevent RCS pressure trens dditions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50 (see Bases 3/4.4.9. The RCS will be protected against overpressure transients and wil)l not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting se gn a es o s temperature. %s 3/4.4.2 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS fran being pressurized above its Safety Limit of 2750 psia. Each safety vi ve i t is designed to relieve approximately 3 x 105 lbs per hour of saturated steam at the valve setpoint. The relief capacity of a single safety l valve is adequate to relieve any overpressure condition which could occur l during shutdown. In=the event that no safety valves are OPERABLE, an 3 operating shutdown cooling loop, connected to the RCS, provides t overpressure relief capability and will prevent RCS overpressurization. During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia. The. combined relief capacity of these valves is sufficient to CALVERT CLIFFS - UNIT 1 B 3/4.4-1 Amadment No. ##f/A5#f,

TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATIOi, SYSTEM INSTRUMENTATION 9G MINIMUM 9 TOTAL NO. CHANNELS CHANNELS APPLICABLE i FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION M 1. SAFETY INJECTION (SIAS)9 2 a. Manual (Trip Buttons) 2 1 2 1, 2, 3, 4 6 h b. Containment Pressure - High 4 2 3 1, 2, 3 7* c. Pressurizer Pressure - Low 4 2 3 1,2,3(a) 7* 2. CONTAIISIENT SPRAY (CSAS) a. Manual (Trip Buttons) 2 1 2 1, 2, 3, 4 6 b. Containment Pressure - High 4 2 3 1, 2, 3 .11 w1 3. CONTAIfGIENT ISOLATION (CIS)# 4 w a. Manual CIS (Trip Buttons) 2 1 2 1, 2, 3, 4 6 b. Containment Pressure - High 4 2 3 1, 2, 3 7* Containment isolation of non-essential penetrations is also initiated by SIAS (functional units 1.a and 9 ^ g- .r (a) Greater than 350'F, the required OPERABLE HPSI pumps must be able to start automatically upon l receipt of a SIAS signal = [ h (b) 8 Between 350'F and 327 F (inclusive), a transititwi region exists where the OPERABLE HPSI pump will be placed in pull-to-lock on a cooldown and restored to automatic status on a heatup, o 8 ( (c) Below 327 F, the required OPERABLE HPSI pump shall be in pull-to-lock and will not start automatically. w v

EMERGENCY CORE COOLING SYSTEMS l 0 ECCS SUBSYSTEMS - Tayg < 300 F LIMITING CONDITION FOR OPERtTION ( 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE: a. One# OPERABLE high-pressure safety injection pump, and b. An OPERABLE flow path capable of taking suction from the refueling water tank on a Safety Injection Actuation Signal and t automatically transferring suction to the containment sump on a Recirculation Actuation Signal. APPLICABILITY: MODES 3* and 4. ACTION: n 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. i SVRVEILLANCE RE0VIREMENTS-4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2. I high-pressure safety injection pumps, except the V required OP br;, 11 be demonstrated inopdiv II'least once per 12 hours whenever the te e or more of the RCS cold legs 0 is 1 275 F b ar4f at the motor circu e been removed l _ _ fam r electrical power supply circuits. J [W$th_ppssupFand,327f(A1154pinclusive, a transition region exists l3ecpcqsgur.e %e % u Between 356 ( where the OPERABLE HPSI pump will be placed in )ull-to-lock on a 0 cooldown and restored to automatic status on a leatup. Below 327 F, the required OPERABLE HPSI pump shall be in pull-t -lock and will 0 1 not start automatically. Below 327 F, HPSI pump u.a will be l conducted in accordance with Technical Specification 3.4.9.3. wwwvwww -i y i CALVERT CLIFFS - UNIT 1 3/4 5-6 Amendment No M/##,

REACTIV1H CONTROL SYSTEMS, i 3/4.1.2 BORATION SYSTEMS i FLOW PATHS - SHUTDOWN l LIMITING COND.lTJON FOR Of"(I!ON 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 ficw 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 eith i charging pump or a high pressure safety injection pu to the '\\ Reactor Coolant System if only the refueling water ta in 1 Specification 3.1.2.7b is OPERABLE. aEPLICABILITY: MODES 5 AND 6. ACTION: With none of the s' ave flow paths OPERABLE, suspend all operations involving CORE ALM RATIONS or positive reactivity changes until at least one injection path is restored to OPERABLE status. SURVEltt.ANCE REOUIREMENTS I 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 r.bove the temperature liniit line shown on Figure 3.1-1 when a flow path l-from the concentrated boric acid tanks is.used. 1 j b. At least once per 31 days be verifying that each. valve (manual, i power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in'its correct position. ~ i e50w eMr b pull-to-lock and will not start automatically. Below 327 F, HPSI 0 ( pump use will be conducted in accordance with Technical (Specification 3.4.9.3. l W- ^ Yw wvvy l t CALVERT CLIFFS - UNIT 1 3/4-1-8 Amendment No. l ~ ~..

REACTIVITY CONTROL SYSTEMS t CHARGING PUMP - SHUTDOWN LllilTING CONDITION FOR OPERATION 3.1.A3 At least one charging pump or one high pressure safety injection pumfi.JAn 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. APfLLCABILIII: MODES 5 and 6. AL11DN: 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. EURVEllLANCE REOUIREMENTS 4.1.2.3 No additional Surveillance Requirements other than those required by Specification 4.0.5. W mmmmm Below 327uf, the required OPERABLE HPSI pump shall be in pull-to-lock and will not start automatically. Below 327 F, HPSI-0 pump use will be conducted in accordance with Technical Specification 3.4.9.3. s v sawvwwW CALVERT CLIFFS - UNIT 1 3/4 1-10 Amendment No. I ___i 2

l EMERGLFCY CORE COOLING SYSTEMS ) SURVEILLANCE REOUIREMENTS 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. i Valve Number Valve Function Valve Position i 1. MOV 659 1. Mini-flow Isolation 1. Open 2. MOV-660 2. Mini-flow Isolation 2. Open I 3. CV-306 3. Low Pressure SI 3. Open Flow Control b. At least once per 31 days by: 1. Verifying that upon a Recirculation Actuation Test Signal, the containment sump isolation valves open. 2. Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or i otherwise secured in position, is in its correct position, 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 restriction of the pump suctions during LOCA conditions. This l visual inspection shall be performed: 1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and 2. Of the areas affected within containment at the completion of containment entry when CONTAINMENT INTEGRITY is established. L d. Within 4 hours prior to increasing the RCS pressure above 1750 psia by verifying, via local indication at the valve, that CV-306 is open.

I sw m mmm m.

Whenever flow testing into the RCS is required at RCS temperatures 0 below 327 F, the high pressure safety injection pump shall recirculate RCS water (suction from RWT isolated) or the controls of (% Technical Specification 3.4.9.3 shall apply. d# CALVERT CLIFFS - UNIT 1 3/4 5-4 Amendment No'.-

i i EMERGENCY CORE COOLING SYSTEMS { BASES The trisodium phosphate dodecahydrate (TSP) store in dissolving baskets located in the containment basement is provided to minimize the i possibility of corrosion cracking of certain metal components during i 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 1 7.0. The requirement to dissolve a resresentative sample of TSP in a sample of RWT water provides assurance t1at the stored TSP will dissolve in borated water at the postulated post LOCA temperatures, i 1 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 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 i configuration, (2) provide the proper flow salit between injection points in accordance with the assumptions used in t1e ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the A analyses. Minimum HPSI flow requirements for temperatures.abov 27 re based upon small break LOCA calculations which credit charging ow following an SIAS. Surveillance testing includes allowances for instrumentation and system leakage uncertainties. The 470 gpm requirement for minimum HPSI flow from the three lowest flow legs includes instrument uncertainties but not system check valve leakage. The OPEkABILITY of the charging i 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 i ""k e 0 At temperatures below 327 F, HPSI injection flow is -limited to less than or equal to 350 gpm except in res)onse to excessive reactor coolant i leakage. With excessive RCS leakage (.0CA), make-up requirements could l exceed 350 gpm. Overpressurization is prevented by. controlling other-parameters, such as RCS pressure and subcooling. This provides overpressure protection in the low temperature region.. An analysis has been performed which shows this flow rate is more than adequate to meet core cooling safety analysis assumptions. HPSIs are not required.to auto-start when the RCS is in the MPT enable condition. The Safety Injection Tanks provide immediate injection of borated water into the core in the event of an accident, allowing adequate time for an operator to take action to start a HPSI. ( Surveillance testing of HPSI pumps is required to ensure pump operability. Some surveillance testing requires that the HPSI pumps deliver flow to the RCS. To allow this testing to be done without increasing the potential for overpressurization of the RCS, either the RWT must be isolated or the HPSI pump flow must be limited to les.s than or equal to 350 gpm or an RCS vent greater than 2.6 square inches must be provided. w w vJJ CALVERT CLIFFS - UNIT 1 B 3/4 5-2 Amendment No. JA/Jpf/JJ7, k

EMERGENCY CORE COOLING SYSTEMS BASES [b4.5.4 REFUELING WATER TANK (RWT) l The OPERABILITY of the RWT as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the vent of a LOCA. The limits on RWT minimum volume and boron gs.b concentration ensure that 1) sufficient water is available within r0g containment to permit recirculation cooling flow to the core, and 2) the

3, reactor will remain subcritical in the cold condition following mixing of W (7, the RWT and the RCS water volumes with all control rods inserted except 65,

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 (characteristics. s b s e CALVERT CLIFFS - UNIT 1 B 3/4 5-2a-

A*ITACRMENT ii) l LTOP SYSTEM DESCRIPTION CALVERT CLIFFS NUCLEAR POWER PLANT UNIT I l

1.0 INTRODUCTION

l On July 21, 1977, a plant specific report on Reactor Coolant System (RCS) f overpressure protection (LTOP) at low temperatures (Reference 8.1) was submitted to the NRC. That report detailed the administrative controls and hardware i modifications which were necessary to protect Calvert Cliffs from an LTOP event I for 10 effective full power years (EFPY). This document describes additional measures required to continue adequate 10 CFR Appendix G prote: tion for 12 EFPY, i It addresses the same issues a> the 1977 report and therefore ;ould be considered an extension of the original submittal. The focus of this description is not on the technical-detail of supporting

analyses, although some information is provided. Rather, a general overview of LTOP protective measures at CCNPP is presented. These measures (for 12 EFPY) supersede those established in 1977 Since no. major modifications to systems affecting LTOP have been made, the important overpressurization events are the same as those postulated in the 1977 report; i.e.,

letdown isolation, safety injection (SI) pump start, charging pump start, reactor coolant pump - (RCP) start, and full pressurizer heater actuation. The re-analysis of these events is similar to the analysis discussed in the 1977 report. 2.0 GENERAL APPROACII TO OVERPRFEURIZATION PROTECI' ION BG&E's approach to LTOP is based primarily on the fact that the potential for c t pressurization of the RCS can be minimized by a combination of administrative p,o.:edures and operator action, liowever, because operator action cannot always be assumed, and because possible equipment malfunctions must be considered BO&E has put in place additional controls to ensure adequate protection exists for all postulated events. Analyses have been performed which demonstrate that a combination of administrative controls and hardware modifications provide this protection. In general, this protection includes the - following: Procedural precautions and controls; i Disabling of non-essential components whenever LTOP is required (below MPT enable temperature and RCS not vented); 1 Maintenance of non-solid system whenever practical; and, Use of the low relief setpoint in the PORY control logic, e t i 0 _

A*ITACHMENT (1) [ l 4 2.1 Desien Criteria I The basic criteria to be satisfied in determining the adequacy of i Overpressure protection is that no single equipment failure or operator error shall result' in violation of the operating curve limitations. This is in accordance with the criteria as originally stated.in Reference 8.1. The applications of these criteria are addressed in Section 6.0, after the i specific means of overpressure protection have been presented. 2.2 Basis for Pressure-Temocrature Limits Pressure-Terroarature ' (P-T) limits (Technical Specificatkn Figures 3.4-2a and 3.4-2b) we e calculated per the requirements of 10 CFR 50, Appendix G as supplernented by the Appendix G to Section 111 of the. ASME Boiler and Pressure Vessel Code, 1986 Edition. Pressure-Temperature limits for 12 EFPY were calculated using' Adjusted Reference temperatures developed from the procedures of Regulatory Guide 1.99, Revision 2. In addition, these ' P-T - limits were corrected for pressure drops and for pressure and temperature instrument uncertainties (Attachment 2). 2.3 pasis for Low Pressure PORY Setooint j The low temperature PORY. pressure lift setpoint is based on protecting the most restrictive pressure of both the heatup and cooldown curves. The I most restrictive pressure limitation is for the 20 F/hr cooldown at 0 0 70 F in s the RCS. The allowable pressurizer pressure (not including pressure instrument uncertainty) is 424.5 psia. A setpoint of 384A +/- 5.2 pria is implemented in the field. This includes the uncertainties necessary for I the protection of 424.5 psia in the pressurizer (Reference 8.2). 2.4 Basis for MPT Enable Temocrature and Prdsure Setnoints r l The LTOP enable temperature (W T enable) has been developed using the i l guidance found in NRC Standard Review Plan 5.2.2,. Revision 2. This SRP defines MPT enable as the water temperature correspo~ ing to a metal 0 temperature of at least RT + 90 F at the beltline, location (1/4 T or is controlling he Appendix G limit calculations. MPT enable 3/4 T) that ~ i temperature was calculated accordingly by-using specific heatup transients with changing thermal rates to accurately - determine stress distributions. This method credits the soaking out of thermal stresses to meet the SRP criteria and is described in Attachment (2). The MPT enable temperature l 0 for Unit 1 is 327 F.

3.0 DESCRIPTION

OF ANALYTICAL MODELE s Overpressurization analyses were performed as follows: The worst case overpressurization scenarios were identified for both mass and energy addition events; 6

A*ITACHMENT f1) e The effectiveness of the PORY to terminate an overpressurization event was evaluated; and, The effectiveness of maintaining 'a pressurizer steam volume as a means of preventing or delaying overpressurization was investigated. The worst case events were identified and reported. In Reference 8.1. To determine the worst case events, solid water RCS conditions were considered. This was/is a conservative assumption since the time delay in the transient due to a non :olid ' system is eliminated. Also, all letdown flow paths which could - initigate or. terminate a particular overpressurization event were assumed isolated. The following subsections discuss the solid system mass and energy i' input analysis, the reactor coolant pump transient model, and the effectiveness l of either the PORVs or pressurizer steam bubble to mitigate an LTOP event. 1 t 3.1 Solid RCS Mass Addition Analysig The following mass addition events were postulated: Inadvertent High Pressure Safety injection (HPSI) pump start; l Inadvertent HPSI and Charging pump start; and. Inadsertent m'ismatch of charging and letdown flow. 3.2 Solid RCS Enernv Addition Andvsis The following energy addition events were postulated: l Decay heat addition due to shutdown cooling system isolation; Inadvertent pressurizer heater actuation; and, Energy addition from the steam generator secondary side to the RCS due to a start of an RCP pump when the steam generators are at a higher temperature than the reactor vessel inventory. Energy additions which are constant with time include inadvertent pressurizer heater actuation and decay heat addition. Hand calculations were sufficient to model the resulting transients. 3.3 RCP Start Transient ModcJ Energy addition analysis of a single RCP _ start with a positive secondary-to p %ary delta T was performed to determine the RCS pressure ~ response ' as a function of time using the RETRAN computer code-(Reference 8.3). Assumptions include isolated letdown and no sensible heat absorption by the RCS component metal mass. These assumptions yield results which are considered the upper bounds of anticipated RCS pressures. -3 ~

ATTACHMENT f1) i 3.4 Effect of PORY Protection The effectiveness of a single PORY was examined for (1) the RCP start 5 transient, and (2) an landvertent mass addition from a itPSI and charging i pump. Thne incidents are considered the Design Basis Events as will be discussed in Section 4.0. i 3.4.1 For the case of mass addition, the' equilibrium pressure at which the IIPSI and charging pump deliveries match a single PORY l discharge is determined. 3.4.2 For the RCP pump s; art transient, the analysis examined the effect of _ a single PORY upon the pressure transient. Saturated i liquid conditions. are assumed in the' pressurizer. l In both transients, valve-discharge rates are determined as a function of upstream RCS pressure, using a maximum backpressure (P ) of 115 psia. { g i 3.5 Pressurizer Steam Bubble Analysis in 1977, analyses were performed to determine the effectiveness of a pressurizer steam volume relative to terminating or mitigating the overpressurization events noted in Sections 3.1 and 3.2. The analyses were based on the assumption that the pressurizer steam will not instantaneously condense with a sudden mass insurgo, and consequently pressure will rise. L The steam bubble, therefore, was considered to be a non-condensible volume, l resulting in considerable conservatism. An update of this analysis for the RCP start transient was performed by _ BG&E in 1987 using the RETRAN computer code. A non-equilibrium pressurizer model was used to confirm the effectiveness of the steam bubble in mitigating LTOP events. 4.0 RTSUl!IS OF ANAIMIS t ( As indicated, the design basis events assuming a water W:1 system are: I A RCP start with hot steam generators; and, An inadvertent ilPSI actuation with concurrent charging. Any measures which will prevent or mitigate the design basis events are sufficient for any of the less severe incidents. Therefore, this section will l discuss the results of the RCP start and mats addition transient analyses. Also discussed is the effectiveness of a procedurally controlled pressurizer steam i bubble and a single PORV relative to mitigating the design basis events. l 4.1 RCP Start Transient The RCP start transient it a severe LTOP challenge for a water solid RCS. Therefore, during water solid operations all four RCPs are tagged out of

service, flowever, analysis indicates that the transient is adequately 8 !

.~. -

ATTACHMENT (1) mitigated by a pressurizy ' steam b @ hle. The approximate size of the required volume is 820 ft. The steam bubble allows the operators at least 10 minutes to initiate corrective action. Other overpressurization incidents which are mitigated by maintaining a pressurizer bubble are: I Charging pump input with insufficient letdown; Letdown isolation; Shutdown cooling isolation; and, Inadvertent attuation of all pressurizer heaters. 4.2 Inadvertent Safety Iniection Transient The inadvertent actuation of one or more HPSI pumps in conjunction with up to three chargirig pumps is the most severo mass addition overpres-surization event. Analyses were performed for a single HPSI - pump and either one or 3 charging pumps assum ng one PORY available with the l 2 existing orifice area of 1.29 in. For the 1imiting case, only a single l PORY is considered available due to single i failure criteria. Figure 1 i shows the calculged RCS pressures that' occur when the discharge of one i PORY-reaches equilibrium with the HPSi and : charging mass inputs. Sufficient overpressure protection results when the equilibrium pressure i does not exceed the. limiting Appendix G curve pressure. Because the equilibrium pressure exceeds the LTOP PORY setpoint, HPSI flow is throttled to no more than 350 gpm when the HPSI pump is used for mass addition. No raore than one charging pump (45 gpm) is alleved to operate during the IIPSI mass addition. 4.2.1 Three 100% capacity HPSI pumps are installed at' Calvert Cliffs. Procedures will require that two of the three HPSI pumps be disabled (breakers racked out) - at RCS temperatures less than or 0 coual to 327 F and that the remaining HPSI pump handswitch bc placed in pull tm bck. A dditionally, the IIPSI pump in pull to -i lock shall be throttled to less than 350 gpm when used to add mass to the RCS. Exceptions are provided for ECCS testing and for response to LOCAs. These cases are discussed in Sections 5.4.1 and 5.4.3. 4 a 4.2.2 Comparison of the PORY discharge curve of Figure I with the critical pressurizer pressure of 424.5 pAa indicates that adequate l protection it provided by a single PORY for RCS temperatures 0 above 70 F ' when all mass input is e limited to 470 spm. MPSI discharge is limited to 350 gpm to allow for one charging pump and system expansion due to decay heat. 9 I l . 3

ATTACHMENT (1) i i 4.3 Summary of Results A pressurizer steam volume or a single PORY - will provide satisfactory l control of all transients with the exception of a spurious actuation of i full flow from a HPSI pump. Overpressurization due to this transient will 0 be precluded for temperatures belew 327 F by disabling two HPSI pumps, placing the third in pull-to-lock, and by throttling the third pump when used. to add mass to less than 350. gpm flow. 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. \\ 5.0 PROVISIONS FOR OV.ERPRESSURE PROTECTION l 7 Low temperature overpressurization is provided-at Calvert Cliffs by a combination of administrative controls and hardware provisions. The hardware provisions - include the incorporation of a dual setpoint capability in the PORV. control circuitry end enabling the' PORVs during low temperature operations. 'Although the PORVs are the primary means of protection, it is desirable to avoid challenging i them. Therefore, maintenance of a pressurizer steam bubble whenever possible durnng heatup and cooldown operations is integral to the overpressure protection t measures, especially when the RCPs are available for= service during heatup. Finally, disabling components when unnecessary for plant operation will prevent their inadvertent actuation and therefore minimize their potential for causing overpressurization. This section discusses specific administrative and hardware modifications including procedural limitations 'for-plant operation during startup, shutdown, surveillance testing, and RCS filling. + 5.1 Administrative Mensures This subsection discusses the administrative measures being taken to i preclude RCS overpressurization. 3 5.1.1 Maintenance of a Pressurizer Steam Volume i Where RCS pressure, temperature, and other operating considerations permit, a maximum 'evel of 170 inches will be maintained. This is considered the most positive means of overpressure protection, encompassing nearly all contributing cases. Limitations which govern pressurizer operatior.s are heatup and cooldown rates, spray valve temperature differentials, and pressurizer-to-hot les temperature differential;. A steam bubble may be formed and maintained as long the pressurizer operatior.s do - not exceed these limits. There is a general precautiori in applicable procedures to instruct operating personnel to minimize the time in which the RCS is in a water solid condition. r ~

A*ITACHMENTE}

l 1

5.1.2 De-activation of Non-Essential Cutoonents in general, any component capable of an. energy or mass input which would result in :RCS overpressurization will be disabled when its t operation is not essential to plant operations. The following are specific limitations: l 5.1.2.1 - Reactor Coolant Pumps shall be disabled during water solid operations. A pressurizer steam volume l will be drawn and secondary to primary delta T less 0 than. or equal to 150 F will be verified prior to operation of an RCP. Primary temperature is read ' using Shutdown Cooling System temperature indication in the ' Control Room; steam generator secor.dary i temperature is - determined by either reading the steam generator pressure and using the corresponding l saturation temperature, or vsing the steam line [ temperature indication, or using a hand-held surface instrument at the steam generator head. 5.1.2.2 IIPSI Pumps - Two of the HPSI pumps are disabled and one is placed in pull to lock at RCS temperatures 0 equal to and below 327 F Also,' the eight HPSI loop motor operated valves are prevented from operating automatically, typically by placing their handswitches in pull-to-override..This ensures that i no SIAS can cause flow to the RCS given a single failure. In addition, when the RCS is solid and

cold, either - the HPSI header-isolation valves (SI-654-MOV and SI-656-MOV) or equivalent valves in the HPSI discharge flowpath are locked shut or equivalent protection is provided L by rackirg out and tagging the-third HPSI pump breaker. Caution tags are used where operation of a pump or valve ceuld result in RCS overpressurization.

5.1.2.3 Pressurizer heaters are disabled and tagged during solid system. operations, except' as provided by procedure to allow drawing a bubble.- 5.1.2.4 Charging pumps that are not required during water solid operations are disabled and tagged. Typically only one charging pump is required to be operating under cold. shutdown conditions. Whenever the RCS is below MPT enable temperature and a HPSI pump is being used to inject into the RCS for testini, at least two charging pumps shall be maintained in pull-to-lock. ~ + l l i

a ATTACHMENT fi) i 5.1.3 Secondary to Primary TemDerature i To ensure that an RCP start does not challengt, MPT ilmits, a l secondary to primary ternperature differential of less than or equal to 150 F is verified prior to RCP start. Preventive measures are taken as follows: l 5.1.3.1 Procedures require that steam generator temperatures 4 0 be reduced to 220 F concurrently with allowing the l RCS to be cooled by the shutdown cooling system. 0 Once 220 F has been achieved, a delta T of less than 0 150 F is assured since-the lowes? RCS service i 0 temperature is 70 F, t 5.1.3.2 Verification of differential temperature 'less than 0 150 F is ' required prior to starting a pump. This is accomplished by comparing RCS temperature to either the saturation-temperature corresponding to steam generator'

pressure, or steamline temperature indication, or a direct temperature measurement made t

with a hand-held instrument, j 5.1.3.3 Reactor coolant pumps will be secured and tagged 0 when RCS temperature is less than 150 F during a cooldown. i 1 5.1.4 Testine of ECCS Comnonents No testing of components is permitted during. water solid operations. 5.2 liardware Features This subsection discusses the hardware provided to mitigate overpressuriza-tion events. I!igh setpoint (2400 psla) PORVs and Code Safety Valves g prevent overpressurization at temperatures above 327 F, Below this 0 temperature, the low setpoint relief capabilities of the system must be enabled. A discussion of 'W operation and-related hardware considerations follows. 5.2.1 Indication and Alugg An automatic computer-activated high pressure alarm is set to alarm at an increasing RCS pressure. The alarm is automatically enabled, by the plant computer, whenever the RCS temperature is less than the MPT enable temperature. This alarm provides audible and visual alarm on C06 and a typewritten message on the alarm typewriter. The pressure sensors used for this alarm function are PT103 and PT103-1. Each sensor-loop provides a separata input to the computer. ; i u

.~. ATTACHMENT (1) i r i Additionally, a computer-activated high-pressure alarm is manually set to alarm at an increasing pressure based on existing RCS temperature. By procedure, the plant operator resets this i alarm to correspond with plant conditions as RCS temperature i changes. This alarm provides the plant operator with a flashing display on the plant computer and is designed to provide three alarm levels upon sensing increasing. RCS pressure; i.e., ' warning,' ' alert,' and ' critical.' The pressure sensors for this alarm function are also PT103 and PT103-1. ( 5.2.2 The mitigation system. against RCS overpressurization at low l RCS temperatures is based on the use of the existing PORVs - (ERV-402 and ERY-404) enabled to provide relief capability at low l pressures. In conjunction with specific procedural

controls, each PORV will provide -sufficient and therefore redundant relief capr. city to ensure that RCS pressure remains within the opere'ing limit curves. The PORY low pressure setpoint will be 384.4 osia, which will be manually aligned when RCS temperature 0

decreses below 327 F. Assurance of preventing inadvertent PORY r 0 actuation at RCS temperatures above 330 F is provided by the inclusion of a temperature interlock in. the circuitry which prevents the low pressure setpoint from actuating the PORVs at RCS temperatures above MPT enable. The mitigating system is provided with separate and independent P-T rignals, bistables and power supplies to cach PORV. This approach is consistent with separation and single failure criteria used in the original l design of the plant. t 5.3 Summarv of Oneration i The following discussion summarizes the sequence of events that ensures overpressure protection is available: l I 5.3.1 By normal plant cooldown procedures the RCS temperature and r pressure are decreased to 330 F and 360 psia, respectively. An annunciator light will come on to indicate that-MPT enable is required. Prior to cooling the RCS below 327 F, normal operating 0 procedures will-require the activation of the manual computer-l generated high pressure alarm, the resetting of the hand switch s to the "MPT Enable" position, checking that the PORY block valves . indiente "open", disabling of two HPSI pumps by racking out I their supply

breakers, placing the third HPSI pump in pull-to-lock, and placing the HPSI loop MOV handswitches in

= pull-to-override. When the PORVs are reset to the LTOP setpoint the annunciator - window light will. clear, indicating that the low temperature PORY mode of operation is in service. The setpoint of the plant computer high pressure alarm -is manually adjusted as called for in procedures so that the operator will be alerted whenever - RCS pressure approaches the operating limits. Upon entering MODE 4, shutdown cooling may be used to cool the

A*ITACMMENT fI) } i i 0 RCS. Steam generators must continue to be coeled to 220 F. RCPs are disabled by locking and tagging out their supply 0 breakers at RCS temperatures less than 150 F. This ensures assumptions in the LTOP analysis are valid. i 5.3.2 During plant heatup,. normal operating procedures will maintain the RCS pressure below 360 psia until the RCS temperature is 0 greater than 327 F. When the RCS temperature exceeds MPT enable, normal operating procedures will require that the PORVs be reset to the normal (high) rellef setpoint of 2,383 psig. At the same time, alarms will be deactivated by procedure, and the temperature interlock will activate; thereby preventing the i lifting of the PORVs at the low. setpoint. Prior to exceeding 0 350 F, two HP3I pumps must be returned to automatic service. { Prior to starting an RCP for the heatup, secondary to primary-0 150 F is verified. l delta T less than i [ 5.4 Doeratinn Guidelines f i 5.4.1 Earveillance and Component Testine When ECCS system HPSI testing is required at RCS 0 temperatures less than 327 F, testing will be performed such that no new mass is introduced to the RCS unless HPSI flow is throttled to no more than 350 gpm and a pressurizer bubble exists or an adequate vent exists. When HPSI suction is taken from the shutdown cooling system, no limit is placed on discharge to the RCS since no new' mass is being added, if addition of non-recirculated mass to the RCS in excess of' 350 ' spm is required for testing. : then the reactor coolant system mugt ~ be i vented through at least 2.6 in. for one pump or 8 in for l multi-pump testing. Testing of Safety injection and CVCS system components - (i.e., pumps, valves, automatic signals, etc.) that are affected by LTOP ccatrols will only be accomplished with a non-Lolid RCS, Such testing is only performed in accordance with approved procedures which establish adequate overpressure protection prior to component testing, No ECCS testing is allowed when water solid. l L i 1 + l. 5.4.2 Reactor Fillinn Reactor coolant system fillind operations during a heatup are normally accomplished by using the containment spray pumps. which have a shutoff head - that is well below the limiting pressure of the MPT curve. To collapse the steam-bubble during a cooldown, a j single charging pump is used. =f ( f

ATTACilMENT (1) 5 5.4.3 LOCA Response In response to unidentified RCS leakage HPSI flow will be i controtted to maintain pressuriter level and avoid overpressurization events. Depending on the size cf the RCS leakage, flow greater than 350 spm may be required. 5.4.4 Ooerator Trainina [ Operator training through required reading and/or m shift briefings will ensure adequate operator awareness of the latest approved LTOP controls. i 6.0 DESIGN CRITERIA The design criteria for LTOP protection system were addressed in Reference 8.1, A brief discussion of the criteria follows: 6.1 Onerator Action i in each of the transient e.calyses, operator action was not credited for the: first 10 minutes. The pressure alarms detailed in Section 5.2.1, in addition to other plant condition indications, will make the operator aware I of the transient, t 6.2 Sinnle Failure A single failure must be considered in the overpressure mitigation system response to an initiating event. { 6.2.1 The sensing / actuating / relieving system consists of two redundant and independent trains. I 6.2.2 For the energy addition design basis event (RCP start with a hot steam generator), the PORY setpoint is not challenged within 10 minutes provided primary-to-secondary delta T is less than 150 F 0 and a pressurizer bubble is present; failure of a PORY does not t result in overpressurization. At least two procedural controls have to be violated for starting an RCP with secondary to primary i 0 delta T greater than 150 F i.e., failure to cool the. steam i 0 generators to 220 F and failure to measure the secondary to t primary differential temperature. Thus the single failure criteria is met - for the energy addition design basis event. t,

{ ATTACHMENT (1) 6.2.3 For the mass addition design basis event (llPSI actuation), a single PORY provides _ protection provided that 2 of the 3 HPSI pumps are disabled and the remaining pump's flow is throttled, if we asseme that the LTOP system single failure is failure to throttle the HPSI, then two PORVs are available to limit the pressurization below MPT limits. 6.3 Seismic and IEEE-279 Demian Criteria i Presently installed PORVs meet seismic criteria consistent with the basic l objective of preventing a potential LOCA pathway. Design of equipment i added for overpressure mitigation is consistent with existing plant design 1

criteria, and with the single failure criteria previously _ discussed.

Design is such that (1) no additional risk of LOCA or other accident is i imposed, and (2) design criteria of existing safety related systems are maintained, and these systems are not degraded. 6.3.1 in addition, the intent of seismic and IEEE-279 criteria is met for the operabiMty and effectiveness of the mitigating system in that a single failure which initiates an overpressurization event does not disable - the mitigating system. 6.3.2 Power is supplied to the PORVs from vital supplies designed to f operate during a seismic event and following loss of off-site I power. Cable raceways for this equipment are supported to withstand a seismic event. 6.4 Testability The system is designed to be tested 4 th a frequency that will ensure the i system is operable when needed, 7.0

SUMMARY

Overpressure protectioti is provided by a combination of hardware and procedural controls. Two PORVs are set to lift at 384.4 psia (protecting 424.5 psia in the 0 pressurizer) for temperatures at and below 327 F, Alarms are provided to the operators to alert them to implement LTOP protective measures and to warn them i when pressure limits are - being approached. Components that can challenge MPT - limits are disabled when not needed and in particular are. disabled for water solid operations. Testing of components is controlled so as to minimize any potential challenge to MPT limits and testing is prohibited during water solid operations. i s 'v, 4 e,--w-

ATTACHMENT (1) 8.0 REFFRENCES 8.1 Letter from V. R. Evans (BGAE) to D. K. Davis (NRC), dated July 21,1977 8.2 Design Engineering Calculation I-89-118, ' Loop Uncertainty Estimate - i Loop wP103-l* 8.3 NEU Calculation 100-TH-8701, ' Low Temperature Overpressure Protection - f - LTOP Thermal Hydraulle Analysis" i i l 6 w f D l I i l .y

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l ATTACWlENT (2) ABB ASEA BROWN BOVERI '{ k 1 May 9, 1990 + B-MPS-90-115 1 1 [ b Mr. Trevor Cook l Nuclear Engineering Department Baltimore Gas & Electric Company Calvert Cliffs Nuclear Generating station i Lusby, MD 20657 i l Subjects Calvert Cliffs 1 Reactor Vessel Pressure-Temperature Limits Final Report References (1) P. J. Hijeck to T. Cook, " Reactor Vessel Pressure-Temperaturc l Limits for Calvert Cliffs Unit 1 and Unit 2 Proposal No. 90-241-59A," Letter B-MPS-90-052, dated February 23, 1990. l (2) Baltimore Gas & Electric Purchase Order 26400FX-01 SOR No. [ 115, dated March 14, 1990. (3) P. J. Hijeck to T. Cook, " Reactor Vessel Pressure-Temperature ( Limits for Calvert Cliffs Unit 1," Letter B-MP8-90-099, dated I April 13, 1990. l I l (4) Baltimore cas & Electric Purchase Order 26400FX-01 SOR No. 119. E j (5) P. J. Hijeck to T. Cook, "Calvert Cliffs 1 Reactor Vessel Pressure-Temperature Limits," Letter B-MP3-90-107, dated April 23, 1990. Attachments: (1) Final Report on Reactor Vessel Pressure-Temperature Limits for Calvert Cliffs Unit i for 12 Effective Full Power Years, Revision 01, May 1990. 1 (2) Technical specification Figures 3.4-2a and 3.4-2b, Reactor Coolant system Pressure-Temperature Limits. [ l

Dear Mr. Cook:

The purpose of this letter is to transmit Revision 01 to the Final Report on . Pressure-Temperature Limits for Calvert Cliffs Unit I for 12 Effective Full Power Years. This report supersedes the report sent to you in Reference (5) and is enclosed as Attachment (1). ABB Combustion Engineering Nuclear Power Combuston Engineemo inc. 1000 Prospect Ha Road Telephone (203) 088-1911 Post Offce Box 500 Fax (203) 2859512 Wodsor. Connectcut 060950500 Telex 99297 COMBEN WSOR

... - - In addition, please find enclosed as Attachment (2), revised figures for the Plant Technical Specifications. These figures have been modified to incorporate editorial modifications requested by the SG&E staff. Figures 3.4-2a and 3.4-2b have been QA verified in accordance with combustion Engineering procedures and are provided for immediate use by BG&E. It has been a pleasure working with BG&E on their current reactor vessel integrity issues and if you should require future support, please do not hesitate to contact me at (203) 285-3115. Sincerely, COMBUSTION ENGINEERIPaG, INC. P. J. Hijeck, P.E. Supervisor, Reactor Vessel Integrity PJHetm Attachments ces J. Connolly M. S. Mcdonald E. A. Siegel C. D. Stewart

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}}

ML20042H065

Contents

Text

:

==

1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

Dear Mr. Cook:

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ML20042H065

Text

:

==

1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

Dear Mr. Cook:

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