ML20056H448

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Amends 174 & 155 to Licenses NPF-4 & NPF-7,respectively, Increasing Allowable as-found Main Steam Safety Valve Setpoint Tolerance from 1% to 3%
ML20056H448
Person / Time
Site: North Anna  Dominion icon.png
Issue date: 09/07/1993
From: Berkow H
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20056H449 List:
References
NUDOCS 9309090340
Download: ML20056H448 (18)
v • d • e


Text

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WASHINGTON. D.C. 20555-0D01 i

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VIRGINTA ELECTRIC AND POWER COMPANY f

OLD DOMINION ELECTRIC COOPERATIVE t

DOCKET NO 50-338 l

NORTH ANNA POWER STATION. UNIT NO 1 i

AMENDMENT TO FACillTY OPERATING LICENSE Amendment No.174 6

License No. NPF-4 l

i i

1.

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

A.

The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated May 24, 1993, 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 i

provisions of the Act, and the rules and regulations of the l

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 i

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.

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Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.D.(2) of Facility Operating License No. NPF-4 is hereby amended to read as follows:

(2) Technical Specifications

[

i The Technical Specifications contained in Appendices A and B as I

revised through Amendment No. 174, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3 This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

l FOR THE NUCLEAR REGULATORY COMMISSION g

H rbert N. Berkow, Director i

froject Directorate 11-2 l

Division of Reactor Projects - 1/11 l

Office of Nuclear Reactor Regulation j

Attachment:

[

Changes to the Technical Specifications i

Date of Issuance: September 7, 1993 i

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ATTACHMENT TO LICENSE AMENDMENT'NO. 174-TO FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 i

Replace the following pages of the Appendix "A" Technical Specifications with j

the enclosed pages as indicated. The revised pages ar9. identified by i

amendment number and contain vertical lines indicating the area of change.

j The corresponding overleaf pages are also provided to meintain document i

completeness.

l Remove Paaes Insert Paoes l

3/4 7-4 3/4 7-4 l

l B 3/4 7-2 B 3/4 7-2 I

B 3/4 7-3 B 3/4 7-3 l

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TABLE 3.7-1 (Continued)

=-

53 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM 3!

LINE SAFETY VALVES DURING 2 LOOP OPERATION E

z Maximum Allowable Power Range Maximum Number of Inoperable Safety Neutron Flux High Setpoint 55 Valves on Any Operating Steam Generatcr*

(Percent of RATED THERMAL POWER)

-e Loop Stop Loop Stop Valves Closed **

Valves Open**

1 2

3 w1 7

w

    • Values dependent on NRC approval of ECCS evaluation for these operating conditions.

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se TABLE 3.7-2

n STEAM LINE SAFETY VALVES PER LOOP e

b

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VALVENUMBER LIFT SETTING f 3%P ORIFICE S!ZE a.

SV-MS 101 A, B, C 1085 psig 16 in2 b.

SV-MS 102 A, D, C 1095 psig 16 in2 c.

SV-MS 103 A, B, C 1110 psig 16 in2 d.

SV-MS 104 A,0, C 1120 psig 16 in2 c.

SV-MS 105 A, B, C 1135 psig 16 in2 w

i

  • The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. All steam line safety valves shall be retumed to an "as left" lift setting of their nominal lift setting f_1%.

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l 3/4.7 PLANT SYSTEMS BASES S/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensure that the secondary system pressure will be limited to within 110% of the system design pressure, 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 i

loss of condenser heat sink (i.e., no steam bypass to the cor, denser).

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 cap city for all safety valves on all of the steam lines is 12.83 x 10 lbs/ht which is i

greater than the total secondary steam flow of 12.77 x 10 lbs/hr.at 6

100% R4TED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam-generator ensures that sufficient relieving capacity is avail-able for the allowable THERMAL POWER restriction in Table 3.7-1.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable withir. the limitations of the ACTION requirements on the basis of the reduction in cecondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels. The reactor trio setpoint reductions are derived on the following bases:

r For 3 loop operation SP = (X) - (Y)(V) x 109 X

For 2 loop operation with stop valves closed 3p, (X) - (Y)(U)

, 7)

X For 2 loop operations with stop valves open Sp, (X) - (Y) (U) x 66 X

NORTH ANNA - UNIT 1 B 3/4 7-1 Amendment No. 84 s

r

PLANT SYSTEMS EMSES Where:

SP-reduced reactor trip setpoint in percent of RATED THERMAL POWER V-maximum number of inoperable safety valves per steam line U-maximum number of inoperable safety valves per operating steam line 109 -

Power Asnge Neutron Flux-High o

' point for 3 loop operation 71 - Maximum percent of RATED THERMAL POWER perruissible by P-8 Setpoint for 2 loop operation with stop valves closed 66 - Maximum perrent of RATED THERMAL POWER permissible by P-8 setpoint for 2 loop operation with stop valves open X-Total relieving capacity of all safety valves per steam line in Ibs/ hour -

4,275,420 Y-Maximum relieving capacity of any one safety valve in Ibs/ hour -

855,084 3'4.7.1.2 AUXILIARY FEEDWATER SYSTEM i

The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss of off site power.

The original design basis of the AFW system provided for two motor driven AFW pumps (AFWP) each capable of delivering 340 ppm and a single turbine driven AFWP capable of delivering 700 gpm to the steam generators during accident conditions. The design basis accidents for the AFW system are the loss of normal feedwater (LONF), the loss of offsite power (LOOP), which are ANS Condition 11 events, and the main feedline break (MFLB), which is an ANS Condition IV event.

Current analyses of the design basis accidents for the AFW system have shown that the applicable acciderit analysis acceptance criteria are met, including the effects of a single active 4

failure of any AFWP to start, if each AFWP is capable of delivering a 300 ppm to its respective steam generator at the safety valve set pressure (including the effects of setpoint drift).

NORTH ANNA-UNIT 1 B 3/4 7 2 Amendment No.174

PLANTSYSTEMS E%SES i

Survei!!ance testing of the AFWPs is performed in accordance with the requirements of ASME Boiler and Pressure Vessel Code Section XI for inservice testing of ASME Code Class 1,2, and 3 pumps and valves. Periodic tests to verify that each pump develops adequate discharge pressure and flow as defined in the ASME Code are designed to assess the operability of the pumps against their design performance characteristics.

Calculations have shown that successful performance of these tests demonstrates that the current AFW system design, including pumps, valves, piping, etc. meets the performance requirements for AFW flow delivery established by the design basis accident analyses with adequate margins to accommodate measurement and test uncertainties.

This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removal System may be placed into operation.

3'4 7.1.3 EMERGENCY CONDENSATE STORAGE TANK The OPERABILITY of the emergency condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with steam discharge to the atmosphere concurrent with totalloss of off.

site power. The contained water volume limit includes an allowance 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 resultant 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. These values are consistent with the assumptions used in the accident analyses.

St4.7.1.5 MAIN STEAM TRIP VALVES The OPERABILITY of the main steam trip 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 pcsitive reactivity effects of the Reactor Coolant System cooldown associated with r

the blowdown, and 2) limit the pressurs rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam trip valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

NORTH ANNA UNIT 1 B 3/4 7-3 Amendment No.174 i

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L

I Pl.N# SYSTEMS BGES I

3/4_716 and 3/4 7.1_7 STEAM TURBtNE and OVERSoEED PROTECTION The turbine generator at the North Anna facility is arranged in a nonpeninsular orientation. Analysis has shova that this arrangement is such that if a turbine failure occurs as a result of destructive overspeed, potentially damaging misstes could impact the auxiliary building, containment, control room and other structures housing safety related equipment. The requirements of these two specifications provide additional assurance that the facility will not be operated with degraded valve performance and/or flawed turbine material which are the major contributors to turbine failures.

3!4.7.2 STEAM GENERATOR PRESSUREfTEMPERATURE UMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70'F and 200 psig are based on average steam generator impact values at 10'F and are sufficient to prevent brittle fracture.

3/4 711 COVDONENT COOUNG WATER SUBSYSTEM - OPERATING The component cooling water system normally operates continuously to remove heat from various plant components and to transfer the heat to the service water system. The system consists of four subsystems shared between units, with each subsystem containing one pump and one heat exchanger.

The current design basis for the component cooling water system is a fast cooldown of one unit while maintaining normal loads on the other unit. Three component cooling water subsystems need to be OPERABLE to accomplish this function. The fourth subsystem is a spare and may be out of service indefinitely. With only two component cooling water subsystems a slow cooldown on one unit while maintaining normalloads on the oppocite unit can be accomplished.

The component cooling water system is designed to reduce the temperature of the reactor coolant system from 350'F to 140'F within 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> during plant cooldown, based on a service water temperature of 95'F and on having two component cooling water pumps and two heat exchangers in service for the unit being cooled down. Therefore, to ensure cooldown of one unit within 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> and maintain the other unit in normal full power operation three of the four subsystems must be OPERABLE.

Because subsystems are placed in standby by shutting down pumps and isolating heat exchangers and this system serves no accident mitigation functions, the subsystem is considered OPERABLE in the standby conditions since it can be easily placed in service quickly by manual operator actions.

3'4.712 COVPONENT COOUNG WATER SUBSYSTEM - SHUTDOWN The OPERABILITY of the component cooting water system when both units are in COLD SHUTDOWN or REFUELING ensures that an atiequate heat sink is maintained for the residual heat removal system.

NORTH ANNA - UNIT 1 B 3/4 74 Amendment No. 76.757.159,

SR Rf C y

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[,E UNITED STATES 4 'jD i i

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yf NUCLEAR REGULATORY COMMISSION 4/

wAswincton, o c. rosss-coci VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMIN!0N ELECTRIC COOPERATIVE DOCKET NO. 50-339 NORTH ANNA POWER STATION. UNIT NO. ?

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 155 License No. NPF-7 1.

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

A.

The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated May 24, 1993, 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 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.

2.

Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-7 is hereby j

amended to read as follows:

(2) Technical Specifications l

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 155, are hereby incorporated in the j

license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION l

/

K erbert N. Berkow, Director i

Project Directorate 11-2 i

Division of Reactor Projects - 1/11 l

Office of Nuclear Reactor Segulation j

Attachment:

l Changes to the Technical l

Specifications Date of Issuance: September 7,1993 i

I i

ATTACHMENT TO LICENSE AMENDMENT NO. 155 i

TO FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-339 P

Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. 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.

Remove Paaes Insert Paaes 3/4 7-4 3/4 7-4 B 3/4 7-2 B 3/4 7-2 j

1 B 3/4 7-3 B 3/4 7-3 i

9 h

i L

1 l

z TABLE 3.7-1 (Continued)

r jE MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 2 LOOP OPERATION EII Maximum Allowable Power Range Maximum Number of Inoperable Safety Neutron Flux High Setpoint N

Valves on Any Operating Steam Generatcr*

i_P tent of RATED THERMAL POWER)

Stop loop Stop i_ ves Closed **

Valves Open**

1 R

2 a

Y 3

ea "At least two safety valves shall be OPERABLE on the non-operating steam generator.

    • Values dependent on NRC approval of ECCS evaluation for these operating conditions.

Es TABLE 3.7 2 I

!Es STEAM LINE SAFETY VALVES PEF 1 LOOP e

EQ VALVENUMBER LIFT SETTING (+ 3%)*

ORIFICE SIZE a.

SV-MS 201 A, B, C 1085 ps'g 16 in2 i

b.

SV-MS 202 A, B, C 1095 psig 16 in2 c.

SV-MS 203 A, B, C 1110 psig 16 in2 d.

SV-MS 204 A, B, C 1120 psig 16 in2 D

i

[

e.

SV-MS 205 A, B, C 1135 psig 16 in2 1.

"The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. All steam line safety valves shall be retumed to an "as left' lift setting of their nominal lift setting A.,1%.

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3/4.7. PLANT SYSTEMS l

BASES i

3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensure that the I

secondary system pressure will be limited to within 110% of the system design pressure, during the most severe anticipated system operational 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).

t The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and i

Pressure Code,1971 Edition. The total relieving capacity for all safety 6

valves on all of the steam lines is 12.83 x g0 lbs/hr which is greater than the total secondary steam flow of 12.77 x 10 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 the allowable THERMAL POWER restriction in Table 3.7-1.

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 POWER required i

by the reduced reactor trip settings of the Power Range Neutron Flux channels. The reactor trip setpoint reductions are derived on the following i

bases:

For 3 loop operation 2

SP = (X)

(Y)(V) x 109 X

J For 2 loop operation with stop valves closed SP = (X)

(Y)(U) x 71 s

NORTH ANNA - UNIT ?

B 3/4 7-1 Amendment No. 71

4 PLAPRSYSTEMS For 2 loop operations with stop valves open

~

SP =

x 66 X

Where:

SP - reduced reactor trip setpoint in percent of RATED THERMAL POWER V-maximum number of inoperable safoty valves per steam line U-maximum number of inoperable safety valves per operating steam line 109 -

Power Range Neutron Flux-High Trip Setpoint for 3 loop operation 71 - Maximum percent of RATED THERMAL POWER permissible by P-8 Setpoint for 2 loop operation with stop valves closed 66 - Maximum percent of RATED THERMAL POWER permissible by P 8 setpoint for 2 loop operation with stop valves open X-Total relieving capacity of all safety valves per steam line in Ibs/ hour -

4,275,420 Y-Maximum relieving capacity of any one safety valve in Ibs/ hour -

855,084 1

34.7.1.2 AlJX!LIARY FEEDWATER SYSTEM 1

The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss of off-site power.

The original design basis of the AFW system provided for two motor driven AFW pumps (AFWP) each capable of delivering 340 ppm and a single turbine driven AFWP capable of delivering 700 ppm to the steam generators during accident conditions. The design basis accidents for the AFW system are the loss of normal feedwater (LONF), the loss of offsite power (LOOP), which are ANS Condition ll events, and the main feedline break (MFLB), which is an ANS Condition IV event.

Current analyses of the design basis accidents for the AFW system have shown that the applicable accident analysis acceptance criteria are met, including the effects of a single active failure of any AFWP to start, if each AFWP is capable of delivering 2.300 gpm to its respective steam generator at the safety valve set pressure (including the effects of setpoint drift).

NORTH ANNA - UNIT 2 B 3/4 7 2 Amendment No.155

4 PLANTSYSTEMS EMSES Surveillance testing of the AFWPs is performed in accordance with the requirements of ASME Boiler and Pressure Vessel Code Section XI for inservice testing of ASME Code Class 1,2, and 3 pumps and valves. Periodic tests to verify that each pump develops adequate discharge pressure and flow as defined in the ASME Code are designed to assess the operability of the pumps against their design performance characteristics.

Calculations have shown that successful performance of these tests demonstrates that the current AFW system design, including pumps, valves, piping, etc. meets the performance requirements for AFW flow delivery established by the design basis accident analyses with adequate margins to accommodate measurement and test uncertainties.

This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350'F when the Residual Heat Removal System may be placed into operation.

3'4.7.1.3 EMERGENCY CONDENSATE STORAGE TANK The OPERABILITY of the emergency condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with steam discharge to the atmosphere concurrent with totalloss of off.

site power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 ACTIVITY j

The limitations on secondary system specific activity ensure that the resultant 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. These values are consistent with the assumptions used in the accident analyses.

3/4.7.1.5 MAIN STEAM TRIP VALVES The OPERABILITY of the main steam trip 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 trip valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

NORTH ANNA-UNIT 2 B 3/4 7-3 Amendment No. 155 J

PLANT SYSTEMS WSES 3/4.7.1.6 and 3/4.7.1.7 STEAM TUR9fNE and OVERSPEED PROTECTION The turbine generator at the North Anna facility is arranged in a nonpeninsular orientation. Analysis has shown that this arrangement is such that if a turbine failure occurs as a result of destructive overspeed, potentially damaging missles could impact the auxiliary building, containment, control room and other structures housing safety related squipment. The requirements of these two specifications provide additional assurance that the facility will not be operated with degraded valve performance and/or flawed turbine material whlch are the major contributors to turbine failures.

3/4 7.2 STEAM GENERATOR PRESSURE / TEMPERATURE UMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70'F and 200 psig are based on average steam generator impact values at 10*F and are sufficient to prevent brittle fracture.

3'4 7.3[1 COUDONENT COOUNG WATER SUBSYSTEM - ODERATING The component cooling water system normally operates continuously to remove heat from various plant components and to transfer the heat to the service water system. The system consists of four subsystems shared between units, with each subsystem containing one pump and one heat exchanger.

The current design basis for the component cooling water system is a fast cooldown of one Unit while maintaining normal loads on the other unit. Three component cooling water subsystems need to be OPERABLE to accomplish this function. The fourth subsystem is a spare and may be out of service indefinitely. With only two component cooling water subsystems a slow cooldown on one unit while maintaining normalloads on the opposite unit can be accomphshed.

The component cooling water system is designed to reduce the tempcrature of the reactor coolant system from 350'F to 140'F within 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> during plant cooldown, based on a service water temperature et 05'F and on having two component cooling water pumps and two hect exchangers in service for the unit being cooled down. Therefore, to ensure cooldown of one unit within 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> and maintain the other unit in normal full power operation three of the four subsystems must be OPERABLE.

Because subsystems are placed in stai:dby by shutting down pumps and isolating heat exchangers and tWis system serves no accident mitigation functions, the subsystem is considered i

OPERABLE in the standby conditions since it can be easily placed in service quickly by manual l

operator actions 3'4 7.3 2 COYPONENT COOUNG WATER SUBSYSTEM - SHUTDOWN l

The OPERABILITY of the component cboting water system when both units are in COLD SHUTDOWN or REFUEUNG ensures that an adequate heat sink is maintained for the residual heat removal system.

1 NORTH ANNA-UNIT 2 B 3/4 74 Amendment No. 7M,140,

. -

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