Text

Proposed Tech Specs Re Design Bases for Providing Containment Pressure Signal Input to Reactor Protective Sys
	ML20205G557
Person / Time
Site:	Calvert Cliffs
Issue date:	10/25/1985
From:	; BALTIMORE GAS & ELECTRIC CO.
To:	;
Shared Package
ML20205G540	List: ML20205G535; ML20205G557;
References
	NUDOCS 8511130284
	Download: ML20205G557 (35)
	v • d • e

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'REtCTIVITY CONTRCL SYSTEMS i

CHARGING PUMPS - OPERATING

-'!v! TING CONDITION FOR OPERATION

'! j

. At least two charging pumps shall de OPERABLE.

3.1.2.4 e

APPLICABILITY:

MODES 1 2, 3 and 4.

~ TION:

~ 'ith only one charging pump OPERABLE, restore at least two tharging pumps :c k

OPERASLEstatuswithin72hoursorbein3.tleastHOTSjAND5Yandboratedto

!a SHUTDOWN MARGIN equivalent to at least 3'; ik/k at 200 F within the next 6

.,;nours; restore a: least two charging pumps to OPERAELE status within the

next 7 days or be in COLD SHUTDOWN within tne next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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' SURVEILLANCE REOUIREMENTS 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE:

a.

At least once per 18 months by verifying that each charging pump starts automatically upon receipt of a Safety Injection Activation i

Test Signal.

i b.

No additional Surveillance Requirements other than those required by Specification 4.0.5.

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e li i CALVERT CLIFFS - UNIT 1 3/4 1-11 Amendmer.: No. JE, g 8511130284 851025 PDR ADOCK 05000317 P

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INSTRUMENTATION POST-ACCIDENT INSTRUMEhrATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The post-accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and.3.

ACTION:

s. As showa in Ta.ble 3.3 *io "ith th nuch;r ;f CPEP,faLE p;;t ;;;id:nt ;;;it;ri;; ch;rn:1; le;; th;r..; quired by T;ble 2.0-10, cither.; stere the in;;;r:bl; cheanci t; OPERACLC etet a..i;.:. 30 Jaya, v.

Le in ll0T :"JT00W" ithi.. the..mL 12 :.v. ;.

b.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS -

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4.3.3.6 Each post-accident monitoring instrumentation chane.el shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL..

CALIBRATION operations at the frequencies shown in Table 4.3-10.

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CALVERT CLIFFS - UNIT 1 3/4 3-40 l

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TABLE 3.3-10 POST-ACCIDENT MONITORING INSTRUMENTATION n?

' MINIMUM

u CHANNELS

[

-lNSTRUMENT_

OPERABLE Actsod r-q 1.

Deleted v,

2.

Containment Pressure 2

31' i

3.

Wide Range Logarittunic Neutron Flux Monitor 2

3l

~

4.

Reactor Coolant Outict Temperature 2

3I 5.

Deleted 6.

Pressurizer Pressure 2

3)

~

7.

Pressurizer Level 2

3i S

8.

Steam Generator Pressure 2/ steam generator 3i 9.

_ Steam Generator Level (Wide Range) 2/ steam generator

'3 /

10.

feedwater Flow

[

2 3/

^

11.

Auxiliary feedwater Flow Rate 2/ steam generator 3/

Il 12.

RCS Subcooled Margin Monitor 1

3i

3

'[.

13.

PORV/ Safety Valve Acoustic Flow Monitoring 1/ valve 31 O'

14.

PORV Solenoid Power Indication 1/ valve 31 h

15.

Containmenth'aterLevel(WideRange)

I a, 3433 1

D a

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-o TABLE 3.3-10 (continuedl a' ACTION STATEMENTS ACTION 31 '-

With the number of OPERABLE post accident monitoring channels less than required by Table 3.3-10, either restore the inoperable channel to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 32 With the number of, OPERABLE post accident monitoring channels one less than the minimum channel operable requirement in Table 3.3-10, operation may proceed provided.the inoperable channel is restored to OPERABLE status at the next outage of sufficient duration.

ACTION 33 With the number of OPERABLE post accident monitoring channels two less than required by Table 3.3-10, either restore one inoperable chann9 to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

e CAlvenr cuffs-umrx 3k 3-%

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REACTOR COOLANT SYSTEti PRESSURIZER LIMITING CONDITION FOR OPERATION

3. 4.~ 4.

The _ pressurizer shall be OPERABLE witn a steam bubble and with at least 150 kw of pressurizer heater capacity capable of being supplied by emergency power.

The cressurizer level shall be maintained witnin an operating. band between 133 and 225 inches except when three charging pumps-are operating and letdown flow is.less than 25 GP!1.

If three chargin'g pumps are operating and letdown flow is less than 25 GPM-ma**ensk pressurizer

l level shall-be limited to between 133 and 21'O inches.

APPLICABILITY: MODES 1 and 2.

ACTION:

a.

With the pressurizer-inoperable due to an inoperable energency power supply to the pressurizer heaters eitner restore the inocerable emergency-power supoly within 12 nours or ce in at leas: HOT STAND 3Y within tne next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDDWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

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

SURVEILLANCE REQUIREMENTS 4.4.4 The pressurizer water level shall be determined to be within'the above band at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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CALVERT CLIFFS - UNIT 1 3/4 4-5 Amendment No. 33, /$I

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REACTOR-COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4 5.

Defect means an imperfection of such severity that it exceeds the plugging limit.

A tube containing a defect is defective. Any tube which does not permit the passage of 4

the eddy-current inspection probe shall be deemedja defective tube.

6.

Plugging Limit means the ' imperfection depth at or beyond which the tube shall be removed from service-because it may become unserviceable prior to the next inspection and is equal to 40". of the nominal tube wall thickness.

I 7.

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect!its

~

structural integrity in the event of an Operating; Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.

.I 8.

Tube Inspection means an inspection of the steam ge'nerator tube from the point of entry (hot leg side) completely

- around the U-bend to the top support of the cold jeg.

O b.

The steam generator shall be deteimined OPERABLE after templeting

the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) re' quired by

, Table 4.4-2.

4.4.5.5 Reports

tubes, Following each inservice inspection of steam generator, ll be a.

the number of tubes plugged in each steam generator sha fa.2.a, 9m eri g g

reportedtotheCommissionwithin15 days / Poes w er T6 r

b.

The complete results of the steam generator tube inser,vice

. inspection shall be included in the Annual Operating Report for the period in which this inspection was completed.

Th'is report ll

( _,e sw <..Ac.u.sgi 4

shall include:

1.

Number and extent of tubes inspected.

2.

Location and percent of wall-thickness penetration for each indication of an imperfection.

3.

Identification of tubes plugged.

O CALVERT CLIFFS - UNIT 1 3/4 4-9 Awwrw No.

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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) cf.

Results of steam generator tube inspections which fall.into Category C-3-and-require verbal notification of the Cormission 10' '

4

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Aon,uiswanby telephone within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months g= t t; 3eci.':c.it;m. 0.0.2 prior to resumption of plant operation. The written followup of this report shall provide a. description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence and sh'all be submitted within the next 30 daysf roma^~r n se m m v..a M.2.

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l CALVERT CLIFFS - UNIT 1 3/4 4-10 Amendment No.

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' EMERGE llCY CORE COOLING SYSTEMS v

SURVEILLANCE REQUIREMEilTS (Continued) e.

At least once per 18 montns by:

i 1.

Verifying automatic ~ isolation and interlock action of the shutdown cooling system from the Reactor Coolant System when the Reactor Coolant System pressure is above 300 psia.

2.

A visual inspection of the containment sump and verifying' that the subsystem suction inlets are not restricted!by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or cor{osion 3.

Verifying that a minimum total of 100 cubic feet of !

solid granular trisodium phosphate dodecahydrate (TSP) is contained within the TSP storage baskets.

4.

Verifying that when a representative sample'of 4.0 + 0.1 grams of TSP from a TSP storage basket is submerged without agitation, in 3.5 0.1 liters'of 77'u 10 F borated water from the RWT, the pH of the mixed solution is raisedito

> 6 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

s t.

At least once per 18 months, during shutdown, by:

1.

Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection Actuation test signal,.

2.

Verifying that each_ of the following-pumps start auto-matically upon receipt of a Safety Injection Actuatidn Test Signal:

a.

High-Pressure Safety Injection pump.

b.

Low-Pressure Safety Injection pump.

g.

verifying the correct position of each electrical pojii'viori' stop

' e following Emergency Core Cooling S stem' throttle valves:

1.

During each perfo ne lve cycling required:by Speci on 5 by.observat valve position on '.

trol boards.

CALVERT CLIFFS - UNIT 1 3/4 5-5 AmendmentNoj#,M

l EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) l

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2.

hin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of maintenance on tiie vaTv'e or 1 erator by measurement of stem travel.when tCCS subsystems required to be OPERABLE:-

HPSI SYSTEM

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Valve Number Valve Number l

{

MOV-616 MOV-61 MOV-626 MOV-627 MOV-MOV-637

-646 MOV-647 gE.

By performing a flow balance test curing shutdown following comple-tion of HPSI system modifications that alter system flow character-istics and verifying the following flow rates for a single HPSI Lj.

pump system *:

1.

The sum of the three lowest flow legs shall be greater ;

than _470** gpm.

j hf.

By verifying that the HPSI pumps develop a total head of 29b0 ft.

on recirculation flow to the refueling water tank when tested pursuant to Specification 4.0.5.

l ll

A HPSI pump system is a HPSI pump and one of two safety injection headers.

- These limits contain allowances for instrument error, drift orf fluctuation.

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CALVERT CLIFFS - UNIT 1 3/4 5-Sa Amendment No. E, 73,fyg 1

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3/4.6 CONTAINMENT SYSTEMS L

v 3/4'. 6.1 - PRIMARY - CONTAINMENT CONTAINMENT INTEGRITY

@ERATio@

LIMITING CONDITION FOR GEPRHf6 W l

3. 6.1.1 Pr'imary CONTAINiiENT INTEGRITY shall be maintained.

APPLICABILITY: : MODES.1, 2. 3 and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore' CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .-

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

~ At least once p'er 31 days by verifying' that all penetrations

a.

not capable of being closed by OPERABLE containment automatic isolation valves and. required-to be closed during accident conditions are closed by valves, blind flanges, or deactivated.

automatic valves secured in their positions, except as provided in Table 3.6-1 of Specification 3.6.4.1.

b.

By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3.

By verifying that the equipment hatch is closed and saaled, c.

prior -to entering itode 4 following a shutdown where the equipment hatch was opened, by conducting a Type B test per Appendix J to 10 CFR Part 50.

y f

d

Except. valves, blind flanges, and deactivated automatic -valves which are located inside the containment and are locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

CALVERT CLIFFS - UNIT 1 3/4 6-1 AmendmentNo.gg

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CONTAINMENT SYSTEMS 3/4.6.5 COMBUSTIBLE GAS CONTROL HYOROGEN ANALYZERS LIMITING CONDITION FOR OPERATION 3.6.5.1 Two independent containment hydrogen analyzers shall be OPERABLE.

s APPLICABILITY: ' MODES 1 and 2.

ACTION:

~

With one hydrogen analyzer inoperable *, restore the inoperable analyzer to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

6 I

49 9 4

SURVEILLANCE REQUIREMENTS i

4.6.5.1 Each hydrogen analyzer shall be demonstrated OPERABLE at le'ast once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gases in accordance with manufacturers'

~

recommendations.

c

During the period from fiay 15 to July 15, 1983, one hydrogen analyzer may

/

be made inoperable, at any given time, for the purpose of replacing system

/

solenoid valves with environmentally qualified valves.

During this time, Specification 3.0.4 is not applicable to this requirement.

[/d CALVERT CLIFFS - UNIT 1 3/4 6-26 AmendmentNo.50,7A,s'y n,$

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m 3.6.5.1

_Two independeni Containment Hydrogen Andlyzers shall be OPERABLE.

APPLICABILITY:'

A40 DES I and 2.

ACTION:

With one hydrogen analyzer inoperable, restore the inoperable analyzer to a.

! OPERABLE status within 30 ' days, or 1) verify containment. at osphere grab sampling capability and prepare m

and submit a - specist report to the Commission pursuant to 4,

specification 6.9.2 within the.following 30 days, outlining the ACTION s

taken, the cause for the inoperability, and the plans and schedule for restoring the system to OPERABLE status, or 2) be in at least HOT STANDBY within the next six hours.

b.t With both hydrogen analyzers inoperable, restore at least one inoperable analyzer to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next six hours.

Specification 3.0.4 is not app!'icable to this requirement.

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TABLE 3.7-4 9

G-SAFETY RELATED HYDRAULIC SNUB 8ERS*

m E

p SNUBBER' SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION-ESPECIALLY DIFFICULT "q

NO.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE 2

(A or 1)

(Yes or No)

(Yes or No) c'-5

'l-83-7S AUXILIARY STEAM ISOLATION H

' VALVE B.YPASS 32' A

No No 1-83-76 AUXILIARY FEED PUMP STEAM SUPPLY FROM S.G. #12 40' A

No No 1-83-76A AUXILIARY FEED PUMP STEAM I

SUPPLY FROM S.G. #12 40' A

No No 1-83-77 AUXILIARY FEED PUMP STEAM w

2 SUPPLY FROM S.G. #12 40' A

No No

[

I-0170 ill AUXILIARY FEE 0 PU"P Tean:xc STEA". SUPPLY 15'

^

"^

Snubbers may be' added to safety related systems without prior License Amendment to Table 3.7-4 provided that a revision to Table 3.7-4 is included with the next License Amendment request.

g Snubbers may be removed from safety related systems.for the purpose of replacement by sway struts in accordance with the NRC's Safety Evaluation dated April 19, 1984 provided that a 3

g revision to* Table 3 1-4 is included with the next. License Amendment request.

o

- Modification to this table due to changes in high radiation areas shall be submitted to the g

NRC as part' of' the next License Amendment request.

e 0o n m

t LIMITING SAFETY SYSTEM SETTINGS BASES operation of the reactor at reduced power if one or two reactor coolant pumps are taken out of service. The low-flow trip setpoints and Allowable Values for the various reactor coolant pump combinations have been derived in consideration of instrument errors and response times of equipment involved to maintain the DNBR above 1.23 under normal operation and expected transients.

For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip setpoints, the Power Level-High trip set-points, and the Thermal Margin / Low Pressure trip setpoints are automatically changed when the pump condition selector switch is manually set to the desired two-or three-pump position.

Changing these trip setpoints during two and three pump operation prevents the minimum value of DNBR from going below 1.23 during normal operational transie,nts and anticipated transients when only two or three reactor coolant pumps are operating.

t Pressurizer Pressure-High The Pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-operated relief valves avoids the undesirable opera-tion of the pressurizer code safety valves..

Containment Pressure-High The Containment Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection. The setpciat fer j

.64,

+.4.

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c.7 g gge 3fety 4.j :tje, tpej.t,

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7me A -ro, oA AT LEAS-r Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant.

The setting of 685 psia is sufficiently

'.y below the full-load operating point of 850 psia so as not to interfere with normal operation, but still high enough to provide the required protec-tion in the event of excessively high steam flow. This setting was used with an uncertainty factor of + 85 psi which was based on the main steam line <,

break event inside containment.

Amendment No. 33, JS, $/f, g'p' CALVERT CLIFFS - UNIT 1 B 2-5

W INSTRUMENTATION BASES by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

3/4.3.3.2 INCORE DETECTORS The OPERABILITY of the 'incore detectors sith the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.

3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that suffi-cient capability is available to promptly determine the magnitude of a seismic. event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis forsthe facility and is cons.istent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquakes", April 1974.

3/4.3.3.4.

METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficie'nt meteorological data is available for estimating potential radiation dosos to the public as a result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendationsofRegulatoryGuide1.23,P.:'t. ' ("r:p:::d) " Mete:r:le;f :1 "r:;r::: in Supp;rt :. L';:10 r P:w;r ?l::t;," :pt;;t:r 19e0, 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient' capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room.

This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.

T SvPPeM!ntec$ N l

"Om i te. Me.horo\\o qicst ProgrQ%," Farucvy (9T2,

2 Supplement I

to WREG- 073 7, CALVERT CLIFFS - UNIT 1 B 3/4 3-2 Amendment No.-HEF-

M E'MERGENCY CORE COOLING SYSTEMS

- lI lBASES The trisodium phosphate dodecahydrate (TSP) stored in dissolving baskets located in the containment basement is provided to minimize the possibility of corrosion cracking of certain metal *;omponents during operation of the ECCS following a LOCA. The TSP provides this protection by dissolving in the sump water and ceusing its final pH to be raised to > 7.0.

The requirement to dissolve a representative sample of' TSP in a sample of RWT water provides assurance that the stored TSP will dissolve in borated water at the postulated

,, post LOCA temperatures.

o i

The Surveillance Requirements provided to ensure OPERABILITY of each I

component ensures that at a minimum, the assumptions ed in the safety analyses t

are met and that subsystem OPERABILITY is maintained.

urveillanrg requirement *'

t for thr;tti; ;;1v; p;;iti;n :t;;; :nd flow balance testing provid& assurance f

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 injec-tion 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 spli.t between injection points in accordance with

!! the assumptions used in the ECCS-LOCA analyses, and (3) prctide an acce;: table

" level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. Minimum HPSI flow requirements are based upon small break LOCA calculations which credit charging pump flow 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 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 analysis issumptions.

3/4.5.4 REFUELING WATER TANK (RWT)

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 event of a LOCA. The limits on RWT minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recircula-tion cooling flow to the core, and 2) the reactor will remain subcritical in tne cold condition following mixing of the RWT and the RCS water volumes witn i

L.all control rods inserted except for the most reactive control assembly.

!!These assumptions are consistent with the LOCA analyses.

l1 M

fna contained water volume limit includes an allowance for water not

'Iusable because of tank discharge line location or other physical character-istics.

2 L

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.CALVERT CLIFFS - UNIT 1 B 3/4 5-2 Amendment No. M,19't

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3/4.7 PLUii SYSTEMS

, ESSES O3/4.7.1 TUR5INE CYCLE 3/4.7.1.1 S UETY VALVES u

The OPERASILITY of the main steam line code safety valves ensures that the

secondary system pressure will be limited to within 110'4 of its design cressure

of 1000 psig during the most severe anticipated system operational transient.

j The total relieving capacity for all va'ves on all of the stea; lines.is i

' 12.15 x 106 lbs/nr at 100:: RATED THERMAL POWER.

The maximum relieving capa:ity j p is associated with a turbine trip frca 100% RATED THERML POWER coincident with an assumed loss of condenser heat sink (i.e.,,,no steam bypass to the condenser). ;~

The main steam line code safety valves are tested and maintained in accordance with the requirements of> Section XI of the ASME Boiler and Pressure Vessel Code.

,W EditiM The as-lef t lift settings will be no less tnan gE5 osig to l.

! ensure that the lift setpoints will remain within specification during the l

cycle.

j!

i

.i In MODE 3, two main steam safety valves are required OPERASLE cer stea.T

,, generator.

Tnese valves will provide adequate relieving capacity for removal

of both decay heat and reactor coolant pump heat from the reactor coolant system via either of the two steam generatirs.

This requirement is provided to facilitate the post-overhaul setting and OPERABILITY testing of the safety valves which can only be conducted when the RCS is at or above 5000F.

It allows entry into MODE 3 with a minimum number of main steam safety valves OPERABLE so that the set pressure for.the remaining valves can be adjusted in the plant.

This is the most accurate means for adjusting safety valve. set prcssures since the valves will be in thermal, equilibrium with the operating environment.

STARTUP and/or POWER CPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced I reactor trip settings of the Power Level-High channels.

The reactor trip

!setpoint reductions are derived on the following bases:

For two loop operation h

sp = (X) - (Y)(V) x 105.5 llh For single loop operation (two reactor coolant cumos coerating in the same loop)

SP = IX) - p u) x 46.8 x

where:

M SP reduced reactor trip setpcint in percent of RATE: THERMAL

=

p0WER maximum number of inoperable safe:y valves per stear line V

=

CALVERT CLIFFS - UNIT 1 3 3/4 7-1 Amencment No. p r

, _. = -

' ADMINISTRATIVE CONTROLS F) 4 n

a.

ECCS Actuation, Specifications 3.3.2 and 3.3.3.

t b.

Inoperable 5eismic Monitoring Instrumentation, Specification 3.3JJ.

c.

Inoperable Meteorological Monitoring Instrumentation, Specificatlon 3.3J.4.

i d.

Seismic event analysis, Specification 4.3.3J.2.

e.

Core Barre! Movement, Speelfication 3.4.11.

f.

- Fire Detection Instrumentation, Specification 3.3.3.7.

)

g.

Fire Suppression Systems, Speelfications 3.7.11.1,3.7.11.2,3.7.11.3,3.7.11.4 and 3.7.11.3.

h.

Penetration Fire Barriers, Specification 3.7.12.

L Steam Generator Tube Inspection Results, Specification 4.4.3.3. a. a c..

l J.

Specific Activity of Prin.ary Coolant, Specification 3.4.8.

k.

Containment SW Integrity, Specification 4.6.1.6.

~

1.

P=dlanctive Effluents - Calculated Dose and Total Dose, Specifications N

3.11.1.2,3.11.2.2,3.11.2.3, and 3.11.4.

g

'.?4 m.

P=dla=etive Effluents - Liquid Radwaste, Gaseous Radwaste and Ventilation Exhaust Treatment Systems Discharges, Specificati,ns 3.11.1 3.11.2.4.

n.

Radiological Environmental Monitoring Program, Specificatl

.12.1.

g%,4 n a Ibron.em festnoerraries, sewr. car..ex 3.33.1 o,

(rm s.3 - 0 p,

ememsou %-..a S-5,

5 F'e ' *'^

3 d' 3 i'

fymasca A Muy t m.,

$ruw' d K&I

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a

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4 CALVERT CLIFFS UNIT l' 6-18a Amendment No. 29, 94, I

b

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.%s__..,.,,,,,,,,,._-.,7,,_

,-._,y,,,,-

,m..,,_m

(

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REACTIVITY' CONTROL SYSTEMS

~

SURVEILLANCE REQUIREMENTS (Continued)~

B 4.1~.1.4.2 The MTC shall be determined at the following frequencies and

~

THERMAL POWER conditions during each fuel cycle:

a.

Prior to initial operation above 5% of RATED THERMAL POWER,

'after.each fuel loading.

b.

At' any THERMAL POWER above 90% of RATED THERMAL POWER, within 7 t

EFPD after initially reaching an equilibrium condition at or above 90% of RATED THERMAL POWER after ach fuel loading.

c.

At any THERMAL Power,'within 7 EFPD

_ reaching a RATED l

THERMAL, POWER equilibrium baron concentration of 300 ppm.

I

?

l l

CALVERT CLIFFS-UNIT 2-3/4 1-6 Amendment No.g g l

{

~;

D INSTRUMENTATION POST-ACCIDENT INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The post-accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

5 APPLICABILITY: MODES 1, 2 and.3.

ACTION:

As noW2 IA Ta.ble %3-10 a.

With th...

.L., of 0?; MOLE pest-eccident iisnitering chennels lui, then requicud by Teble 0.0-10, cither rc;tcr: th ineperable chennel ts 0? 00L: status aithir. 20 d;y:, ;r bc

=

in ::0T :lLiOOWN aithin the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;.

b.

The provisions of Specification 3.0.4 are not applicable.

D v

SURVEILLANCE REQUIREMENTS T

2 4.3.3.6 Each post-accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-10.

I i

7 l

l CALVERT CLIFFS - UNIT 2 3/4 3-40 AUG 1, mg i

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h N

i TABLE 3.3-10'

^

9 POST-ACCIDENT MONITORING INSTRilMENTATION~

G m-MINIMUM 4

CHANNELS g

INST f1ENT OPERABLE

.h' n

.l.

Containment Pressure 2

31 3#

2.

Wide Range Logarithmic. Neutron Flux Monitor 2

3.

Reactor Coolant Outlet Temperature 2

4.

Pressurizer Pressure 2

5.

Pressurizer Level 2

31 6.

Steam Generator Pressure 2/ steam generat6r 31 7.

Steam Generator Level (Wide Range) 2/ steam generator 31 s

]

' fl.

Auxiliary Feedwater Flow Rate 2/ steam generator 3l t

9.

RCS Subcooled Margin Monitor 1

31 10.

PORV/ Safety Valve Acoustic Flow Monitoring 1/ valve 3) 6" 11.

PORV Solenoid Power Indication 1/ valve 31 j

R 12.

Fdedwater Flow 2

3I a

5 13.

C mtainment Water Level (Wide Range) 1 3233.

l

/

5 W

M.

V TABLE 3.3-10 (continued) i ACTION STATEMENTS ACTION 31 With the number of OPERABLE post accident monitoring channels less than required by Table 3.3-10, either restore the inoperable channel to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

- ACTION 32 With the number of. OPERABLE post accident monitoring channels one less than the minimum channel operable requirement in Table 3.3-10, operation may proceed provided the inoperable channel is restored to OPERABLE status at the next outage of sufficient duration.

- ACTION 33 With the number of OPERABLE post accide.nt monitoring channels two less than required by Table 3.3-10, either restore one inoperable channel to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Sk 3 ~ YlW B M.VEAT c uffS - UNIT 9, w

o 1

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REAC_ TOR COOLANT SYSTEf1 PRESSURIZER LIMITING CONDITION FCR OPERATION-l 3.4.4.

The pressurizar shall be OPERABLE with a steam bubble and witn at least 150 kw of-pressu.-izer heater capacity capable of being supplied by emergency power.

The pressurizer level shall be maintained within an operating band between 133 and 225 inches except wnen three charging pumps are operating and.. letdown flow is less than 25 GPM.

If three charging pumps are operating and letdown flow is less than 25 GPM-ee*4ense pressurizer l

level shall be limited to between 133 and 21,0 inghes.-

APPLICABILITY: MODES 1 and 2.

ACTIO}t:

a.

With the pressurizer inoperable due to an inoperable energency power supply to the pressurizer heaters either restore the inoperable emergency power supply within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANCSY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN witnin the fol'osing 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

~

b.

With the pressurizer otherwise inoperable, be in at least HOT STAND 3Y Nith the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REGUIREMENTS 4.4.4 The pressurizer water level shall be determined to be within the above band at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

CALVERT CLIFFS - UNIT 2 3/4 4-5 AmendmentNo.26,Jdb' I

4

(

C

.l

]

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

F 5..

. Defect means an imperfection of such severity that it exceeds the plugging limit. _ A tube containing a defect is defective. Any tube which does not permit the passage of

.the eddy-current inspection probe shall be deemed a defective tube.

6.

Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service because it may become unserviceable prior to the next inspect [on and is equal to 40% of the nominal tube wall thickness.

{

7.

Unserviceable-Neribes the condition of a tube if!it leaks or cor a defect large enough to affect its 2

st: ucturC '

.grity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam lline or feedwater line break as specified in 4.4.5.3.c. above.

Tube Inspection means an inspection of the steam gelnerato 8.

tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

0

\\

b.

The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the limitandalltubescontainingthrough-wallcracks)requl plugging ired by Table 4.4-2.

4.4.5.5 Reports a.

Following,each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator sha)1 be reported to the Commission within 15 daysf -um t. srew.um.a(..M.

l 7

b.

The complete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was completed.

This report

{

shall include:

^

( m s r s %lr.or.44.u l6 1.

Number and extent of tubes inspected.

2.

Location and percent of wall-thickness penetration for each indication of an imperfection.

3.-

Identification of tubes plugged.

D CALVERT CLIFFS-UNIT 2 3/4 4-9 wuwnr Mo.

l

/

i:

_=

.h REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) cf.

'Results of steam generator tube inspections which fall into Category C-3.-and-require verbal notification of the Commission M

'^'

b smv.a.by telephone within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months pr:n.r. t: 0; c ificeri = !.0.S

, prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence a'nd shall be

r. hwWos 6.'t.2.

submitted within the next 30 daysf evasva<

m.

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O' c^'vear c'tres - unit 2 3/4 4-io

^=eadme't "o-75

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c.

EMERGENCY CORE COOLING SYSTEMS

-SURVEILLANCE REQUIREMENTS (Continued) e.

. At least once'per 18 months by:

1.

. Verifying automatic isolation and int'erlock action of the shutdown cooling system from the Reactor Coolant System when the Reactor Coolant System pressure is above 300 psia.

2.

A visual inspection of the containment sump and verifying' that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks,' screens, etc.)shownoevidenceofstructuraldistressorcorrosion.

3.

Verifying that a minimum total of 100 cubic feet of solid granular trisodium phosphate dodecahydrate (TSP)-

is contained within the TSP storage baskets.

4.

Verifying that when a representative sample of 4.0 + 0.1 grams of. TSP from a TSP storage basket is submerged ~without agitation, in 3.5 t 0.1 liters of 77' 10'F borated water from the RWT, the pH of.the mixed solution is raised to

> 6 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

f.

~At least once per 18 months, during shutdown, by:

1.

Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection

,3 Actuation test signal.

2.

Verifying that each of the following pumps start auto-matically upon receipt of a Safety Injection Actuation Test Signal:

a.

High-Pressure Safety Injection pump.

'b.

Low-Pressure Safety Injection pump.

P rifying the correct position of each electricaJosritibf 1

stop ro ollowing Emergency Core Coolin ystEm throttle valves:

1.

During each er ance of va ling required by Spe on 4.0.5 by observation o ai m ition

~

n the control boards.

CALVERT CLIFFS - UNIT 2 3/4 5-5 AmendmentNo.16,//

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.

thin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of maintenance on th vaTve I

or rator by measurement of stem travel when CCS subsyst e required to be OPERABLE:

HPSI SYSTEM Valve Number Valve Number MOV-616 MOV-61 MOV-626 MOV-627 MOV-MOV-637

-646 MOV-647

p. By perfonning a flow balance test during shutdown following comple-tion of HPSI system modifications that alter system flow character-istics and verifying the following flow rates for a single HPSI pump system *:

1.

The sum of the three lowest flow legs shall be greater than 470.** gpm.

hg. By ve.rifying that the HPSI pumps develop a total head of 2900 ft.

on recirculation flow to the refueling water tank when tested pursuant to Specification 4.0.5.

A HPSI pump system is a HPSI pump and one of two safety injection headers.

- These limits contain allowances for instrument error, drift or fluctuation.

CALVERT CLIFFS - UNIT 2 3/4 5-Sa Amendment No. J6,/E6,$

3/4-6 CONTAINMENT SYSTEMS 3/4. 6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY

[opEcATioro LIMITING CONDITION FOR GEPRAftoft l

3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1,-2, 3 and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAIN'1ENT INTEGRITY within or.e hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a.

At least on.ce per 31 days by verifying that all penetrations

not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except ar provided in Table 3.6-1 of Specification 3.6.4.1.

b.

By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3.

~

c.

By verifying that the equipment hatch is closed and sealed, prior to entering Mode 4 following a shutdown where the equipment hatch was opened, by conducting a Type B test per Appendix J to 10 CFR Part 50.

Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed, or otherwise secured in the closed position.

These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more of ten than once per 92 days.

CAINERT CLIFFS - UNIT 2 3/4 6-1 AmendmentNo.6,)[g 4

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7 CONTAINMENT SYSTEMS 3/4.6.5 COMBUSTIBLE GAS CONTROL HYOROGEN ANALYZERS LIMITING CONDITION FOR OPERATION 3.6.5.1 Two independent containment hydrogen analyzers shall be OPERABLE.

1 APPLICABILITY: MODES 1 and 2.

1 ACTIE:

With one hydrogen analyzer inoperable *, restore the inoperable analyzer to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

6W~~

go'gh gnGL SURVEILLANCE REQUIREMENTS 4.6.5.1 Each hydrogen analyzer shall be demonstrated OPERABLE at least once per 92 days on a' STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gases in ace'ordance with manufacturers' g;

recommenda tions.

7 l

f i

0uring the period from itay 15 to July 15, 1983, one hydrogen analyzer may I

be made inoperable, at any given time, for the purpose of replacing system 6

solenoid valves with environmentally qualified valves.

Curing tnis time,

)

l Specification 3.0.4 is not applicable to this requirement.

l t(

)

Amendment No. 42, 55,k CALVERT CLIFFS - UNIT 2 3/4 6-26 91 I

I.

.s w e..

3.6.5.1 Two independent Containment Hydrogen Andlyzers shall be OPERABLE.

APPLICABILITY.

MODES I and 2.

ACTION:

a.

With one hydrogen analyzer inoperable, restore the inoperable analyzer to OPERABLE status within 30 days, or 1) verify containment atmosphere grab sampling capability and prepare and submit a special report to the Commission pursuant to

- specification 6.9.2 within the following 30 days, outlining the ACTION taken, the cause for the inoperability, and the plans and schedule for restoring the system to OPERABLE status, or 2) be in' at least HOT STANDBY within the next six hours.

b.

With both hydrogen analyzers inoperable, restore at least one inoperable analyzer-to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY. within the next six hours.

c.

Specification 3.0.4 is not applicable to this requirement.

w gy b,

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-INSTRUMENTATION BASES-l by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

3/4.3.3.2 INCORE DETECTORS The OPERABILITY of the incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.

3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERASILITY of the seismic instrumentation ensures that suffi-cient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permi,Leomparison of the meg _sured response to that used in the design basis for the facility and _is consistent with the recomendations of Regulatory Guide 1.12. " Instrumentation for Earthquakes", April 1974.

3/4.3.3.4.

METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient metecrological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recomendations of Regulatory Guide 1.23, %;.1 (Pr:p:::d), "M::::=1 g e:1 j

"=;r: : ' r.

rt Of i

!

: r P:w;r Pier,t:," k ;;;;t:r 100C.

l 3/4.3.3.5 REMOTE SHUTOOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of-the facility from locations outside of the control room.

/

]

This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.

'*PP W ediA %

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id UwE6 -O'7 3'7, CALVERT CLIFFS - UNIT 2 B 3/4 3-2 Amendment No.+r,-

r1 0

3

.w l LIMITING'SAFETYSYSTEM~ SETTINGS BASES operation of the reactor at reduced power if one or two reacto'r ccolant pumps are taken out of service.

The low-flow trip setpoints and Allowable

- s Values for the various reactor coolant pump combinations have been-derived in consideration of instrument errors and response times of equipment involved to maintain the DNBR above 1.23 under normal operation d'.

and expected transients ~.

For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip set-points, the Power. Level-High trip setpoints, and the Thennal Margin / Low Pressure trip setpoints are automatically changed when the pump condition selector switch is manually set to the desired.two-' or three-pump position.

Changing these trip setpoints during two and three pump operation prevents the minimum value of DNBR from going below 1.23 during d

normal operational transients.and anticipated transients when only two or three reactor coolant pumps are operating.

Pressurizer Pressure-Hich The Pressurizer Prossure-High trip, backed up by the pressurizer code safety' valves and main steam line safety valves, provides reactor coolan.t system protection against overpressurization in the event of loss of load without reactor trip.

This trip's~setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-operated relief valves avoids the undesirable operation of the pressurizer code safety valves.

Containment. Pressure-High The Containment Pressure-High trip provides assurance that a reactor tripisinitiateg4 concurrently-with a safety injection.

Me seta^iat j

Faw +k4e-.4n

4. e- - antieml'en the e2fatu injaetinn emen,ninL tw r

Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant.

The setting of 685 psia T

is sufficiently below the full-load operating point of 850 psia so S

as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively hiah steam flow.

This setting was used with an uncertainty factor of + 83 psi n

in the accident analyses.which was based on the Main Steam l~ine Break

)

avent.

c CALVERT CLIFFS - UNIT 2 B 2-5 Amendment No. JE,37, gg u

EMERGENCY CORE COOLING SYSTEMS BASES The trisodium phosphate dodecahydrate (TSP) stored 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 > 7.0.

The requirement to dissolve a representative sample of TSP in a sample of RWT water provides assurance that the stored TSP will dissolve in borated water at the postulate!

post LOCA temperatures.

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

gurveillance requirementX for thr;ttic vel;c ps;iti;r, :t;;; :-d flow balance testing providh 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 configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the 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 ECCS-LOCA analyses. Minimum HPSI flow requirements are based upon Small Break LOCA calculations which credit charging pump flow following a 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.g The OPERABILITY of the charging pumps and the associated flowpaths is assured by the Boration System Specifications 3/4.1.2.

Specification of safety injection pump total developed head ensures pump performance consistent with safety analysis assumptions.

3/4.5.4 REFUELING WATER TANK (RWT)

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 event of a LOCA. The limits on RWT minimum volume and boron concentration ensure that1)sufficientwaterisavailablewithincontainmenttopermitrecircula-tion cooling flow to the core, and 2) the reactor will remain subcritical in the cold condition following mixing of 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 characteristics.

CALVERT CLIFFS - UNIT 2 B 3/4 5-2 AmendmentNo.H/(g'

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3/4.7 PLANT SYSTEMS wi$wasa.Nl.arcel,,b.

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BASES g-3/4.7.1 TURBINE CYCLE G/4.7.1.1 SAFETY VALVES The OPERA 81LITY of the main team line code safety valvos ensures that the secondary system pressur will be limited to withini'its design pressure of 1000 psig during the st severe anticipated system opera-The maximum r lieving capacity is associated with a tional transient,100% RATED T turbine trip frca RMAL POWER coincident with an assumed

~

loss o co enser heat sink (i e., no teambypasstothecondenser).ad uk s.g als n.

-tucen e

De :;;;i'W =2.:.11 T' i.ii;..,; e. 4 re' i:ti-- Teit9._;s__in

- )[1.

accordance with the requir ts af taction @DM%iler and he total reliev' ng capacit for all valves Pressure code. M?' O.;,,.

ines is 12.18 x 105 lbs/hr Ji;T.15 {.

Con all of the steam Er ^;'2? ::::t. ;^_ r r :" II.^^

at 1005 RAT THEMAL PWER.fn c' '

" ': r_- m : t y ret:: ;:r !*r? :-- - 9t-s

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. in tr: e.TT;c;.i reite.t..ii c.;;;tt, ;;b is on Tn NW ME I

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f gg givi FMAMA?H h HEAL; STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of.the ACTIM requirements on the basis

/

of the reduction in secondary system steam flow and THERMAL PWER required

~

by the reduced reactor trip settings of the Power Level-High channels.

The reactor trip setpoint reductions are derived on the following bases:

For two loop operation

$p, (X) - (Y)(V) x 106.5 For single loop operation (two reactor coolant pumps operating in the same loop)

$p, (X) - (Y)(U) x 46.8 where:

reduced reactor trip setpoint in percent of RATED SP

=

THERMAL POWER maximum number of inoperable safety valves per steam V

=

line y,-.

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k' h.4 serm WG

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,CALVERT CLIFFS-UNIT 2 8 3/4 7-1 26^ociE6 m e"

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~

t ADMINISTRATIVE COPCROLS i

D a.

ECCS Actuation, Specifications 3.3.2 and 3.3.3.

k b.

Inoperable Selsmic Monitoring Instrumentation, Specification 3.3.3.3.

c.-

Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.

d. - Seismic event analysis, Specification 4.3.3.3.2.

e.

Core Barrel Movement, Specification 3.4.11.

f.

Fire Detection Instrumentation, Specification 3.3.3.7.

g.

Fire Suppression Systems, Speciflettions 3.7.11.1, 3.1.11.2, 3.7.11.3, 3.7.11.4 and 3.7.11J.

i h.

Penetration Fire Barriers, Specification 3.7.12.

L Steam Generator Tube Inspection Results, Specification 4.4.3.3. o.. A c.

l J.

Specific Activity of Primary Coolant, Specification 3.4.8.

j k.

' Containment Structural Integrit/, Specification 4.6.1.6.

1.

P=diametive Effluents - Calculated Dose and Total Dose, Specifications N

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