Text

Amends 67 & 49 to Licenses DPR-53 & DPR-69,respectively, Requiring Operability & Verification of Auxiliary Feedwater Sys Throttle Valves
	ML20041D128
Person / Time
Site:	Calvert Cliffs
Issue date:	02/08/1982
From:	Clark R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20041D129	List: ML20041D128; ML20041D137; ML20041D139;
References
	NUDOCS 8203040334
	Download: ML20041D128 (15)
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f,j UNITED STATES

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p, NUCLEAR REGULATORY COMMISSION M

E WASHINGTON, D. C. 20555

$?

N g *.ss f BALTIMORE GAS AND ELECTRIC COMPANY DOCKET N0. 50-317 CALVERT CLIFFS NUCLEAR POWER STATION, UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 67 License Nos. DPR-53 1.

The Nuclear Regulatory Comission (the Commission) has found that:

A.

The application for amendment by Baltimore Gas and Electric Company (the licensee) dated December 21, 1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; b.

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

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

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

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

s 8203040334 820208 PDR ADOCK 05000317 P

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2.. Accordingly, the license is amended by. changes to the Technical Specifications as inqicated in the attachment to this license amendment, and paragraph 2.C.(2).of Facility Operating License No. DPR-53 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 67, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specificatior.s.

3.

This license amendment is effective within 30 days after entering MODE 1, as defined by Technical Specification 1.4, following completion of the reload for operation during Cycle 6.

l FOR THE NUCLEAR REGULATORY COMMISSION I

k\\~

Cw Robert A. Clark, Chief Operating Reactors Branch #3 Division'of Licensing l

Attachment:

i Changes to the Technical Specifications Date of Issuance:

February 8,-1982 i

3

ATTACHMENT TO LICENSE AMENDMENT NO. 67 FACILITY OPERATING LICENSE NO. DPR-53 DOCKET NO. 50-317 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

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

Pages 3/4 7-5 3/4 7-Sa 3/4 7-Sb B 3/4 7-2 B 3/4 7-3 4

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least two steam turbine driven steam generator aa.liary feedwater pumps and associated flow paths shall be OPERABLE and capable of automatically initiating flow, within the area of acceptable operation of Figure 3.7-1, to each steam generator.

APPLICABILITY: MODES 1, 2 and 3*.

ACTION:

With one auxiliary feedwater pump inoperable, restore at least two auxiliary a.

feedwater pumps to 0FERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , b.

Whenever a subsystem is inoperable for the performance of periodic testing (i.e., manual discharge valve clo: ed for pump discharge head test) a dedicated

?

operator will be stationed at the local station (i.e., closed valve), with direct communication to the Control Room, to return the subsystem to normal upon instruction from the Control Room.

Upon completion of any testing, the subsystem (valve) will be returned to its proper position and verified in its proper position by an independent operator check.

SURVEILLANCE REQUIREMENTS 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

'I' a.

At least once per 31 days by:

1.

Verifying that each steam turbine driven pump develops a Total Dynamic Head of > 2800 ft. on recirculation flow when the secondary steam supply pressure is greater than 800 psig.

2.

Cycling each testable, remote operated valve that is not in its operating position through at least one complete cycle.

3.

Verifying that each valve (manual, power operatad or automatic) in the direct flow path is in its correct position.

b.

Before entering MODE 3 after a COLD SHUTDOWN of at least 14 days by completing a flow test that verifies the flow path from the condensate storage tank to the steam generators.

+

c.

At least once per 18 months by:

ll 1.

Verifying that each automatic valve 'a the flow path actuates to its correct position upon receipt of each auxiliary feedwater act'uation

I test signal.

ll

2., Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of each auxiliary feedwater actuation test signal.

Automatic flow initiation need not be OPERABLE during MODE '3.

l CALVERT CLIFFS - Unit,1 3/4 7 5 Amendment No. 54, c. 7 1

l s

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 1 3.

Verifying, upon automatic initiation of auxiliary feedwater, a flow within the acceptable operating limits of Figure 3.7-1, Steam Generator Pressure Versus Auxiliary Feedwater Flow.

d.

Upon repositioning of 1/2-CV-4511 and/or 1/2-CV-4512 the valve shall be realigned to provide flow consistent with Figure 3.7-1.

I i

l

Only applicable during tiODES 1 and 2.

l l

CALVERT CLIFFS '- UNIT 1 3/4 7-Sa A:nendment No. 6 7 l

(

4 s

5 1000 X'

AREA OF ACCEPTABLE OPERATION E

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$ 900 R

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g 800 l

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1 i

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80 90 100 110 120 130 140 150 160 AUXILIARY FEEDWATER FLOW (GPM)

Figure 3.7-1 Steam Generator Pressure vs. AumHiery Feedwater Flow CALVERT CLIFFS - UNIT 1 3/4 7-5b Amendment No. 6 7

PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The No. 12 condensate storage tank (CST) shall be OPERABLE with a minimum contained water volume of 150,000 gallons per unit.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With the No.12 condensate storage tank inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a.

Restore the CST to OPERABLE status or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or b.

Demonstrate the OPERABILITY of the No. 11 condensate storage tank as a backup supply to the auxiliary feedwater pumps and restore the No. 12 condensate storage tank to OPERABLE status within 7 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 .

SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The No. 12 condensate storage tank shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the contained water volume is within its limits when the tank is the supply source for the auxiliary feedwater pumps.

4.7.1.3.2 The No. 11 condensate storage tank shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the tank contains a minimum of 150,000 gallons of water and by verifying that the flow path for taking suction from this tank is OPERABLE with the manual l

valves in this flow path open whenever the No.11 condensate storage l

tank is the supply source for the auxiliary feedwater pumps.

l l

CALVERT CLIFFS-UNIT 1 3/4 7-6 i

3/4.7 PLANT SYSTEMS 4

BASES 3/4.7.1 TURBINE CYCLE i

3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within its design pressure of 1000 psig during the most severe anticipated system opera-tional transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

4 The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code,1971 Edition. The total relieving capacity for all valves on all of the steam lines is 12.18 x 106 lbs/hr which is 108 percent of the total secondary steam flow of 11.23 x 106 lbs/hr at 100% RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam generator i

ensures that suffic ent relieving capacity is available for removing decay heat.

i STARTUP and/or POWER OPERATION is allc able 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 reactor trip settings of the Power Level-High channels.

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

I For two loop operation 3p, (X) - (Y)(V)'x 106.5 For single loop operation (ho reactor coolant pumps operating in the same loop) i 3p, (X) - (Y)(U) x 46.8 where:

reduced reactor trip setpoint in percent of. RATED-SP

=

THERMAL POWER

}-

_V maximum number of inoperable safety valves per steam

=

line L

CALVERT CLIFFS - UNIT 1 B 3/4 7-1 i

4

PLANT SYSTEMS BASES maximum number of inoperable safety valves per U

=

operating steam line 106.5 =

Power Level-High Trip Setpoint for two loop operation 46.8 =

Power Level-High Trip Setpoint for single loop operation with two reactor coolant pumps operating in the same loop Total relieving capacity of all safety valves per X

=

steam line in lbs/ hour Maximum relieving capacity of any one safety valve Y

=

in lbs/ hour 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 300 F from normal operating conditions in the event of a total loss of offsite power.

Each steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a Total Dynamic Head of 2490 ft to the entrance of A capacity of 450 gpm, however, is sufficient to ensure the steam generators.

that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System tenperature to less tha-300 F when the shutdown cooling system may be placed into operation.

A minimum flow of 88 gpm (plus 12 gpm for instrument error) and a maximum flow of 142 gpm (minus 12 gpm for instrument error) to each steam generator when automatically initiating AFW flow for a steam generator pressure of 900 psia, is required to ensure sufficient time for Operator action to maintain an adequate heat sink for the reactor.

Figure 3.7-1 shows the allowable minimum and maximum flows.

The minimum flow is adequate enough to allow 20 minutes before the operator is required to increase AFW flow to 450 gpm. At the-same time the maximum flow is low enough to ensure 20 minutes for the Operator to turn off AFW flo'w if main feedwater is available. - Failure to turn off AFW, in this case, would result in safety injection actuation due to the rapid cooldown of _the RCS.

CALVERT CLIFFS - UNIT 1 B 3/4 7-2 Amendment No. 59, 6 7

PLANT SYSTEMS BASES 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months with steam discharge to atmosphere with concurrent and total loss of offsite power.

The contained water volume limit includes an allow-ance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the result-ant off-site radiation dose will be limited to' a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line and a concurrent loss of offsite electrical power.

These values are consistent with the assumptions used in the accident analyscs.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the, event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the cicsure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

3/4.7.1.6 SECONDARY WATER CHEMISTRY The secondary water chemistry program is designed to provide maximum protec-tien to both the steam generator and secondary system internals. The most damag-ing chemical reactants enter the system via condenser cooling water ingress.

Accumulation of these impurities in the steam generators may lead to. loss of metallurgical integrity and/or eventual component failure. The limits presented in Table 3.7-3 are those prescribed by the NSSS supplier 'as " limited-operation" chemistry parameters and are consistent with the most recent industry standards.

By routine monitoring of these parameters, plant personnel are able to rapidly detect and limit the duration of ingress of chemically detrimental species and thereby maintain steam generator tube integrity.

i 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.

The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allow-able fracture toughness' stress limits. The limitations of 70 F and 200 psig are of 50*F and are sufficient to prevent brittle based on a steam generator RTNDT fracture.

CALVERT CLIFFS - UNIT 1 B 3/4 7-3 Amendment No. H, 6 7 a

PLANT SYSTEMS BASES 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available for continued operation of The safety related equipment during normal and accident conditions.

redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses'.

3/4.7.4 SERVICE WATER SYSTEM The OPERABILITi of the service water system ensures that sufficient cooling capacity is available for continued operation of equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

3/4.7.5 SALT WATER SYSTEP1 The OPERABILITY of the salt water system ensures that sufficient cooling capacity is available for continued operation of equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is ccnsistent with the assumptions used in the accident analyses.

3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY 'of the control room emergency ventilation system ensures.that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumenta-tion cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General

~

Design Criteria 10 of Appendix "A",10 CFR 50.

3/4.7.7 ECCS PUMP ROOM EXHAUST AIR FILTRATION SYSTEM The OPERADILITY of the ECCS pump room exhaust air filtration system ensures that radioactive materials leaking from the ECCS equipment within the pump room following a LOCA are filtered prior to reaching the CALVERT CLIFFS-UNIT 1 B 3/4 7-4

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION

$ [ \\,,

"E WASHINGTON, D. C. 20555

... 4 BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER STATION, UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 49 License Nos. DPR-69 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Baltimore Gas and Electric Company (the licensee) dated December 21, 1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; b.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Coamission; There is r' asonable assurance (i) that the activities authorized C.

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

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

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

9 Y

+,

.. =.

, 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license i

i

. amendment, and paragraph 2.C.2 of Facility Operating License No. DPR-69.is hereby amended to read as follows:

2.

Technical Specifications 4

The Technical Specifications contained in Appendices A and B, as revised through Amendment No.49, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective within~ 30 days after entering MODE 1, as defined by Technical Specification 1.4, following completion of the reload for operation during Cycle 5 FOR THE NUCLEAR REGULATORY COMMISSION (s.

+

v Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:

February 8, 1982 I

I i

Q

' ATTACHMENT TO LICENSE AMENDMENT NO. 49 FACILITY OPERATING LICENSE NO. DPR-69 1

1 DOCKET NO. 50-318 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment i

number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document concleteness.

Pages l

3/4 7-5 k

3/4 7-Sa 3/4 7-5b B 3/4 7-2 B 3/4 7-3 1

1

PLANT SYSTEMS AUXILIARY FEE 0 WATER SYSTEll SURVEILLANCE REQUIREMENTS (Continued) 3.

Verifying, upon automatic initiation of auxiliary feedwater, a flow within the acceptable operating limits of Figure 3.7-1 Steam Generator Pressure Versus Auxiliary Feedwater Flow.

Upon repositioning of 1/2-CV-4511 and/or 1/2-CV-4512 the valve

d4 shall be realigned to provide flow consistent with Figure 3.7-1.

I i

1 i

Only applicable 7uring it0 DES 1 and 2.

CALVERT CLIFFS - UNIT 2 3/.4 7-Sa Amendment No. 4 9

... - ~ - -

l,.

lPLANTSYSTEMS i

I AUXILIARY FEEDWATER SYSTEM i

l LIMITING CONDITION FOR OPERATION q

l 3.7.1.2 At least two steam turbine driven steam unerator auxiliary feedwater pumps and associated flow paths shall be OPERABLE and capable of automatically initiating flow, within the area of acceptable operation of Figure 3.7-1, to i

each steam generator.

APPLICABILITY: MODES 1, 2 and 3*.

l l

ACTION:

With one auxiliary feedwater pump inoperable, restore at least t:o auxiliary a.

i feedwater pumps to OPERABLE status within 72 nours or be in HOT SHUlDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

Whenever a subsystem is inoperable for the performance of periodic testing (i.e., manual discharge valve closed for pump discharge head test) a dedicated operator will be stationed at the local station (i.e., closed valve), with direct communication to the Control Room, to return the subsystem to normal upon instruction from the Control Room.

Upon completion of any testing, the j

subsystem (valve) will be returned to its proper position and verified in its proper position by an independent operator check.

SURVEILLANCE REQUIREMENTS 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

l a.

At least once per 31 days by:

1.

Verifying that each steam turbine driven pump develops a Total Dynamic Head of > 2800 ft. on recirculatier, flow when the secondary j

steam supply pressure is greater than 800 psig.

2.

Cycling each testable, remote operated valve that is not in its operating position through at least ene complete cycle.

3.

Verifying that each valve (manual, power operated or automatic) in the direct flow path is in its correct position.

b.

Before entering MODE 3 after a COLD SHUTDOWN of at least 14 days by completing a flow test that verifies the flow path from the condensate storage tank to the steam generators.

f c.

At least once per 18 months by:

1.

Verifying that each automatic valve in the flow path actuates to its

~ '

l correct position upon receipt of each auxiliary feedwater actuation test signal.

2.

Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of each auxiliary feedwater actuation test signal.

Automatic flow initiation need not be OPERABLE during MODE 3.

l CALVERT CLIFFS - UNIT 2 3/4 7-5 Amendment No.-37,1g

i i

i l

j

$ 1MO 1

E f

AREA OF ACCEPTABLE OPERATION g

l

/

1 i

E 3

/

l 900 l

g O

l l

5 1

I zw o

)

E l

g 800 I

N l

u l

I i

i a

I I

I f

I i

j 80 90 100 110 120 130 140 150 160 i

AUXILIARY FEEDWATER FLOW (GPM) 1 I

Figure 3.71

)

Steam Genarator Pressure vs. Auxiliary Feedwater Flow i

CALVERT CLIFFS - UNIT 2 3/4 7-5b Amendment No. 4 9 i

~

m_

PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The No. 12 condensate storage tank (CST) shall be OPERABLE with a minimum contained water volume of 150,000 gallons per unit.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With the No.12 condensate storage tank inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

I a.

Restore the CST to OPERABLE status or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or b.

Demonstrate the OPERABILITY of the No. 21 condensate storage tank as a backup supply to the auxiliary feedwater pumps and restore the No. 12 condensate storage tank to OPERABLE status within 7 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 .

SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The No. 12 condensate storage tank shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the contained water volume is within its limits when the tank is the supply source for the m iliary feedwater pumps.

4.7.1.3.2 The No. 21 condensate storage tank shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the tank contains a minimum of 150,000 gallons of water and by verifying that the flow path for.taking suction from this tank is OPERABLE with the manual.

valves in this flow path open whenever the No. 21 condensate storage tank is the supply source for the auxiliary feedwater pumps.

I CALVERT CLIFFS-UNIT 2 3/4 7-6

3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES l

The OPERABILITY-of the main steam line code safety valves ensures that the secondary system pressure will be limited to within its design pressure of 1000 psig during the most severe anticipated system opera-tional transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss'of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is 12.18 x 106 lbs/hr which is 108 percent of the total secondary steam flow of 11.23 x 106 lbs/hr at 100% RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for removing decay heat.

STARTUP and/or POWER OPERATION 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 P'WER 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 l

3p, (X) - (Y)(V) x 106.5 For single loop operation (two reactor coolant pumps operating in the same loop) 3p, (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 steaa V

=

line lCALVERT CLIFFS-UNIT 2 B 3/4 7-1 m

PLANT SYSTEMS BASES maximum number of inoperable safety valves per U

=

operating steam line

-106.5 =

Power Level-High Trip Setpoint for two loop operation 46.8 =

Power Level-High Trip Setpoint for single loop operation with two reactor coolant pumps operating

.in,the same loop Total relieving capacity of all safety valves per X

=

steam line in lbs/ hour Maximum relieving capacity of any one safety valve Y

=

in lbs/ hour 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 300 F from normal operating conditions in the event of a total loss of offsite power.

Each steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a Total Dynamic Head of 2490 f t to the entrance of the steam generators. A capacity of 450 gpm, however, is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 300 F when the shutdown cooling system may be placed into operation.

A minimum flow of 88 gpm (plus 12 gpm for instrumer.* error) and a maximum flow of 142 gpm (minus 12 gpm for instrument error) to each steam generator when automatically initiating AFW flow for a steam generator pressure of 900 psia, is required to ensure sufficient time for Operator action to main 6ain an adequate heat sink for the reactor.

Figure 3.7-1 shows the allowable minimum and maximum l

flows, The minimum flow is adequate enough to allow 20 minutes before the operator is required to increase AFW flow to 450 gpm. At the same time the maximum flow is low enough to ensure 20 minutes for the Operator to turn off AFW flow.if main feedwater is available.

Failure to turn off AFW, in this case, would result in safety injection actuation due to the rapid cooldown of the RCS.

l e

CALVERT CLIFFS - UNIT 2 B 3/4 7-2 Amendment No. AI,4 9

PLANT SYSTEMS BASES _

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months with steam discharge to atmosphere with concurrent and total loss of offsite power. The contained water volume limit includes an allow-ance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the result-ant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture.

This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line and a concurrent loss of offsite electrical power. These values are consistent with the assumptions used in the accident analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture.

This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

3/4.7.1.6 SECONDARY WATER CHEMISTRY The secondary water chemistry program is designed to provide maximum protec-tion to both the steam generator and secondary system internals. The most damag-ing chemical reactants enter the system via condenser cooling water ingress.

Accumulation of these impurities in the steam generators may lead to loss of metallurgical integrity and/or eventual component failure. The limits presented in Table 3.7-3 are those prescribed by the NSSS supplier as " limited-operation" chemistry parameters and are consistent with the most recent industry standards.

By routine monitoring of these parameters, plant personnel are able to rapidly detect and limit the duration of ingress of chemically detrimental species and thereby maintain steam generator tube integrity.

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allow-able fracture toughness stress limits. The limitations of 70 F and 200 psig are f 50'F and are sufficient to prevent brittle based on a steam generator RTNDT fracture.

CALVERT CLIFFS - UNIT 2 B 3/4 7-;

Amendment No.16, 4 g e

t l

l PLANT SYSTEMS i

BASES 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available for continued operation of

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safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

3/4.7.4 SERVICE WATER SYSTEM The OPERABILITY of the service water system ensures that sufficient cooling capacity is available for continued operation of equipment during normal and accident conditions. The redur. dant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

3/4.7.5 SALT WATER SYSTEM The OPE 1 ABILITY of the salt water system ensures that sufficient cooling capacity is available for continued operation of equipment during normal ar.d accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the control room emergency ventilation system ensures that 1) the ambient air temperature does not exceed the allow-able temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. The OPERABILITY of this system in con-junction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control roor. to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of Gener.1 Design Criteria 10 of Appendix "A",10 CFR 50.

3/4.7.7 ECCS PUMP ROOM EXHAUST AIR FILTRATION SYSTEM The OPERABILITY of the ECCS pump room exhaust air filtration system ensures that radioactive materials leaking from the ECCS equipment within the pump room following a LOCA are filtered prior to reaching the CALVERT CLIFFS-UNIT,2 B 3/4 7-4

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