ML20096H556

Safety Evaluation Supporting Amends 159 & 140 to Licenses NPF-4 & NPF-7,respectively

ML20096H556
Person / Time
Site:	North Anna
Issue date:	05/12/1992
From:	Office of Nuclear Reactor Regulation
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ML20096H545	List: ML20096H541 ML20096H556
References
NUDOCS 9205270284
Download: ML20096H556 (5)

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i UNITED STATES 5

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• s e s * $,8 WAsHINoTON, o.C. 2%66 VIRGINIA ELECTRIC AND POWER COMPANY QLD DOMINION ELECTRIC COOPERATIVE QQEET W. 50-339 NORTH ANNA POWER STATION. UNIT NO. 2 AMENDMENT TO FACILITY QPERATING LICEt[51 Amendment No. 140 License No. NPF-7 1

1.

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

A.

The application for amendmer.t by Virginia Electric and Power Company i

et al., (the licensee) dated November 7, 1991, complies with the st.ndards and requirements of the Atomic Energy Act of 1954, as amender fthe Act), and the Commission's rules and regulations set forth it. 10 CFR Chapter I:

B.

The facility will o)erate in conformity with the application, the provisions of tie Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) Gat 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':; regulations; D.

The issuance of this amendment will not b inimical to the common defense and stsurity or to the health and safety of the public; and E,

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

9205270284 920512 PDR ADOCK 05000338 P

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Au ordingly, the license is amended by chauges to the Technica.1 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 amended to read as follows:

(2) Technical Specificaligni The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 140, 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 issuaner and shall be implemented within 30 days.

FOR T

NUCLEAR REGULATORY COMMISSION

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11 rbert N. Berkow, Director Project Directorate 11-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Attachment.

Changes to the Technical Specifications Cate of Issuance: May 12, 1992 l

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AllAGMENT TO LICENSE AMENDMENT N0, 140 TO FACILITY OPERATING. LICENSE N0. NPF-7 DOCKET NO. 50-339 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 PJLqe1 Insert Pa4es 3/4 7-14 3/4 7-14 3/4 7-14a B 3/4 7-4 8 3/4 7-4 l

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PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRES $URE/ TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2.1 The temperatures'of both'the primary and secondary coolants in the steam generators shall be greater-than 70'F when the pressure of either coolant in the steam generator is greater than 200;psig.

APPLICABILITY:

At all times.

ACTION:

-With the requirements of the above specification not satisfied:

a.

Reduce the steam-generator pressure of the applicable side to less-than or equal to 200 psig within 30 minutes, and b.

Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator.-

Determine that the steam generator remains acceptable-for continued

. operation prior to increasing its temperatures above 200'F.

SURVEILLANCE REQUIREMENTS.

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4.7.2.1 The-pressure in each side of the steam. generator shall be

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determined to be'less than 200 psig at least once per hour _ when the tempera-ture of either-the primary or secondary coolant is less than 70*F.

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PLANT SYSTEMS 2'4.7.3 CCMPONENT CCOUNG WATER SYSTEM 3'4.7.3.1 CCffPONEN1_COOUNG WATER SUBSYSTEM - OPERATING LIMITING COND(TtON COR OPERATION 3.7.3.1 Three component cooling vwir subsystems (shared with Unit 1) shall be OPERABLE

• With each subsystem consisting of:

a Ore OPERABLE component cooling water pump ard,

b. One OPERABLE component cooling wr.ter heat exchanger.

APPUCAB!LITY:

Either Unit in MODES 1, 2,3, or 4.

ACTON:

a With one required component cooling water subsystem inoperable, return the component cooling subsystem to OPERABLE status within the next 7 days, or place both units in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. With two required component cooling water subsystems inoperable, place both

, units in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and within the next hour, initiate actions to place both units in COLD SHUTDOWN and continue until COLD SHUTDOWN is achieved, i

c, With no component evoling water available to supply the residual heat removal heat exchangers to cool the units, place both units in HOT SHUTDOWN within the nert 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and remain in HOT SHUTDOWN until afternate means of decay heat removal can be implemented. Continue actions until both units are in COLD SHUTDOWN.

SURVEILLANCE REOUlREMENTS 4.7.3.1 Three component cooling water subsystems shall be demonstrated OPERABLE:

l a At least once per 31 days by veri 4ing that each vatve (manual, power ope sted or automatic) servicing :n the flow path of the residual heat removal system that l

is not locked, sealed, or otherwise secured in position, is in its correct position.

b. Each component cooling water pump shall be tested in accordance with Specification 4.0.5.

For the purpose of this Technical Specification, each subsystem is considered OPERABLE if it is operating or if it can be placed in sernce from a standby condition by manually unisolating a standby heat exchanger and/or manually starting a standby pump.

NORTH ANNA UNIT 2 3/4 7 14 Amendment No. 140,

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PLANT SYSTEMS T4.7.3 COMPONENT COOUNG WATER SYSTEM T4.712 COMPOf;ENT COOUNG WATER SUBSYSTEM SHUTDOWN P

UMITING CONDITION FOR OPERATION 3.7.3.2 Two component cooling water subsystems (shared with Unit 1) shall be OPERABLE' with each subsystem consisting of:

a. One OPERABLE component coollng water pump and,

b. One OPERABLE component cooling water heat exchanger.

APPUCABILITY:

Both Units in MODES S or 6.

i ACnON:

With one required component cooling water subsystem inoperable, immediately suspend all_ operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reador Ceolant System.

SURVEILLANCE REQUIREMENTS 1

4.7.3.2 At least two component cooling water subsystems shall be demonstrated OPERABLE:

- a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing in the flow path of the residual heat removal system that is not locked, sealed. or otherwise secured in position, is in its correct position.

b.- Each component cooling water pump shall be tested in accordanco with Specification 4.0.5.

For the purposes of this Technical Specification, each subsystem is considered OPERABLE if it is operating or if it can be placed in service from a standby condition by manually

. unisolating a standby heat exchanger and/or manually starting a standby pump.

P NORTH ANNA UNIT 2 3/4 714a Amendment No. 140, i

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PLANT SYSTEMS l

BASES Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 340 gpm at a pressure of 1064 psig to the entrance of the steam generators.

The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a pressure of 1064 psig to the entrance of the steam generators.

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 S1ANDBY 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 total loss of off-site power. The contained water volume limit in:ludes an allowance for water not usuable 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 rup'ture.

This. dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in I

the steam generator of the affected steam line.

These values are consistent l

with the assumptions used in the accident analyses.

l 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 I

l restriction is required to 1) minimite the positive reactivity effects of the 2eactor Coolant System cooldown associated with the blowdown, and 2) limit the l

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 w

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PLANT SYSTEMS BASES 3/4.71.6 and 3'4.71. 7 STEAM TUR9INE and OVER9 DEED DROTECTION 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 00 curs as a result of destructive overspeed, potentially damaging missies 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 furbine material which are the major contributors to turbine failures.

3'4.7.2 STEAM GENERATOR PRESSUAE'TEVDERATURE LIMITATION 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 COMDONENT COOLING WATER SUBSYSTEM ODERATING The component cooling water system normally operates continuously to remove heat from variuJs piant components and to transfer the heat to the service water system. The system consists of four subsystems sh.1 red oetween units, with each subsystem containing one pump and one heat exchanger.

The curreni design basis for the componant cooling water system is a fast cooldown of one Unit while main.aining 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 ccmponent cooling water subsystems a slow cooldown on one unit while esintaining normalloads on the opposite Unit can be accomphshed.

The compone-* coling water system is designed to reduce the temperature of the reactor coolant sy.; tem from 20'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 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 OPER ABLE.

Because subsystems are placed in stancby by shutting down pumps and isolating neat exchangers and tMs system serves no a:cicent mitigation functions, the subsystem is considered OPERABLE in the standby conditions since it can be easily p! aced in service quickly by manual operator actions.

3'4 ' 3 2 COYPONENT COOUNG W ATER SUE 9v97EM. SWUTDOWN The OPERABILITY of the component cochng water system when both units are in COLD SHUTDOWN or REFUELING ensures that aa aceouate heat sink is maintained for the residual heat remcval system.

NORTH ANNA - UNIT 2 B 3'4 74 Amendment No. 7#,140,

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j' SAFETY EVALUATIOR BY THE OFFICE Oi i4VCLEAR REACTOR REGULATION RELATED TO AMENDMENT N05.159 AND 140 TO FACILITY OPERATING llCIBSE NOS. NPF-4 AND NPF-7 ylRGJNIA ELECTRIC AND POWLR COMPANY OLD DQMINIOP ELECTRIC COOPERATIVE

[(QRTH ANNA POWER STATJ0N. UNITS NO.1 AND NO. 2 DOCKET NOS. 50-338 AND 50-339

1.0 INTRODUCTION

By letter date( November 7,1991, the Virginia Electric and Power Com)any (the licensee) proposed changes to the Technical Specifications (TS) for tie North Anna Power Station, Units No. I and No. 2 (NA-l&2). The proposed changes would revise the current NA-l&2 TS to ensure the design basis is met for the component cooling water system (CCWS).

The proposed changes are being made as a result of an NRC violation reg:rding the NA-l&2 service water system (SWS).

In the Notice of Violation dated February 1, 1991, the Nkt also observed that operation of the CCWS was different from that described in the NA-l&2 Updated Final Safety Analysis Report (UFSAR).

In the licensee's response to the NRC Notice of Violation o sed March 1, 1991, the licensee committed to changes in the NA-l&2 TS to clarify both the SWS and CCWS operability require'ments and, in the interim, to provide adequate administrative controls to ensure the SWS and CCWS design bases are met.

By letter dated October 3, 1991, the licensee proposed the p-appropriate changes to ensure the design basis is met for the SWS.

On December 13, 1991, the NRC issued Amendment Hos. 152 and 136 for NA-l&2, respectively, which addressed the operability requirements for the SWS.

A description of the NA-1&2 CCWS, a discussion of the CCWS TS changes as proposed in the licensee's letter of Mamber 7,1991, and the staff's evaluation are provided below.

2.0 CCWS BACKGROUND The CCWS is a closed cycle system.

Cool water is circulated through various components in the plant for cooling and returned to the heat exchangers for heat rejection to the SWS.

The CCWS consists of four subsystems shared between NA-l&2, with each subsystem containing one pump and one heat exchanger.

The major components in the CCWS are a surge tank, four pumps,

. four heat exchangers, and a radiation monitor.

CCW must be supplied to various components to accompl.sh the following functions:

a.

Removal cf resiaual hc

~m the reactor coolant system (RCS) through the residual i moval (RHR) system during unit cooldown.

b.

Cooling of letdm flow to the chemical and volume control system (CVCS) during power generation, and c.

R(emoval of heat from various nuclear steam supply system (NSSS) components during power generation and normal unit cooldown.

The CCWS ensures that sufficient coolin;, capacity is available for continued operation of various equipment during normal unit cooldown.

The CCWS performs no design basis accident mitigation function.

The CCWS is not a system which functions to mitigate the failure of or presents a challenge to the integrity of a fission product barrier.

Complete redundancy to meet single failure cri+.eria is not a design basis feature of this syt6em. The CCWS supports operation of the RHR system.

The RHR system does not perform a design basis accident mitigation function.

The current design basis for the CCWS is a fast cooldown of one unit while maintaining normal loads on the other unit.

Three CCW 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 CCW subsystems operable, a slow cooldown on one unit while maintaining normal loads on the opposite unit can be accomplished without reliance on the main steam system below 350*F.

The CCWS is designed so that when one t. nit is being cooled down, two component cooling pumps and two component cooling heat exchangers for that unit supply the RHR, reactor coolant pumps, and nonregenerative and seal water heat exchanger flow paths while the common loeds header is supplied from the other unit.

If only one component cooling pump and component cooling heat exchanger are operating, cooldown can be accomplished but requires more time.

3.0 DISCUSSION The numbering of the TS would be changed from 3/4.7.3 to 3/4.7.3.1 to allow for the addition of a new CCW TS when both units are in modes 5 or 6 (which is numbered 3.7.3.2).

In addition, the title of this TS would be changed to add the word " operating" to identify that this TS applies to whenever either unit is in modes 1 through 4.

The limiting condition for operation (l.C0) would be changed to require three subsystems (shared between both units) to be operable and to define what an operable CCW subsystem consists of.

Two CCW subsystems provide the minimum heat removal capability to accomplish a slow cooldown on one unit while

t maintaining normal loads on the opposite unit.

To ensure the design basis require.nent of a fast cooldown on one unit and normal operational loads on the opposite unit is met, three subsystems of CCW must be operable, in addition, a footnote would be added to further clarify when a subsystem is considered o pe ra b*. e.

The applicability statement would be changed to clarify that this TS applies if either or both units are in modes ' through 4.

This change will ensure that sufficient cooling capacity is available for both units.

Action statement 3.7.3.1.a would be added to require that if one of the three required subsystems becomes inoperable, then the subsystem must be returned to an operable status within 7 days.

If the required CCW subsystem cannot be restored in the required completion time, both units must be placed in a mode in which the risk to the unit is minimized.

This would be done by placing both units in hot standby in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> :iid in cold shutdown in the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

i Action statement 3.7.3.1.b would be added to require _that if two of the three required subsystems become inoperable, then both units must be placed in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and that actions be initiated within the next hour to place both units in cold shutdown and continue to cold shutdown if CCW is available to supply the RHR exchangers to further cool the units.

The units are first placed in a condition where decay heat can be removed by the steam generators. 'ihis can be achieved in hot shutdown.

Action statement 3.7.3.1.c would be added to require that with no CCW available to supply the RHR heat exchangers to further cool the units, both units must be placed in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The. units may remain in hot shutdown until alternate means of decay heat removal can be implemented.

If com)onent cooling is available and a heat sink to further cool the units is availa)1e, unit cooldown would continue until cold shutdown is achieved under action statement 3.7.3.1.b.

Surveillance requirement 4.7.3.1.a would be modified to replace " safety

- related equipment" with "in the flow path of the residual heat removal-systems." Also, Surveillance Requirement 4.7.3.1.b would be added to specify surveillance testing for operability determination of the CCW pumps in i

accordance with TS 4.0.5, the ASME Section XI program.

TS 3.7.3.2 would be added to support the current NA-l&2 UFSAR design bases l

when both units are in modes 5 or 6.

When both units are in cold shutdown refueling, the design basis requiras that the CCW system be operable.

This is to ensure an adequate heat sink is maintained for the RHR system.

L A new LCO would be established for modes 5 and 6.

This new LCO differs from i

TS 3.7.3.1 by requiring that-two of the four CCW subsystems be operable.

CCW is required to provide a heat sink for the RHD system to remove decay heat from the reactor core.

However, there is a significant reduction in potential

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4 heat loading on CCW with the reduced operational requirements of the other systems that are cooled by CCW.

The major reduction in heat loads is due to the fact that by mode 5 reactor decay heat has already dropped off significantly and that reactor coolant pumps and control rod drive mechanisms are not required to be operating in modes 5 and 6.

Therefore, only two CCW subsystems are required to be operable, if only one of the two required CCW subsystems is operable, action statement 3.7.3.2 would require all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the RCS must be immediately suspended.

This is consistent with the action requirements for a total loss of RHR capability during shutdown conditions.

The surveillance requirements would be the same as those of TS 3.7.3.1 as noted above.

Ff-y, tha existing Bases section (3/4.7.3) would be expanded to provide a more, cetailed description of the CCWS.

This Bases section would also be split to provide a description for operating and shutdown conditions (3/4.7.3.1 and 3/4.7.3.2, respect ively).

Bases page B 3/4 7-4a for both units were changed by Amendment Nos. 152 and 136, and therefore are not included with these amendments.

4.0 EVAEJJMi The proposed changes document the licensee's commitment to clarify the CCWS

.merability requirements.

The propcsed changes enhance the availability of che CCWS and ensure that sufficient cooling capacity is available for ontinued operation of various equipment during normal unit cooldown for bnth

'V.

The proposed changes further ensure the availability of a heat sink toi

.te RHR system to remove decay heat from the reactor core by requiring that two of the four CCW subsystems be operable when NA-l&2 are in modes 5 and 6.

Finally, the proposed changes to the NA-1.&2 TS ensura consistency with the UFSAR design basis and result in additional limitations not currently specified in the NA-l&2 TS.

Based on all of the above, the staff finds the propoced changes to be acceptable.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Virginia State official was notified of the proposed issuance of the amendment.

The State official had no comment.

6.0 E(11LRONMENTAt CONSIDERATION These amendments change a requirement witn respect to installation or use of a facility component located within the restricted area as defined in 10 CFR

-Part 20 and changes surveillance requirements.

The NRC staff has determined

. that the amendments involve no significant increasc in the amounts, and no significant change in the types. of any ef fluents that may be released offsite, and that there is no significant increase in individual or cumulative The Commission 'as previously 'esued a occupational radiation exposure.

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proposed finding that the amenu.nents involve no sigr.ficant ha;: arils

ideration and there has been no pubic comment ur + uch finding (56 FR et.,

>3). Accordingly, the amendments meet the eligib.lity criteria for

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c.gorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuarit to 10 CFR

%,y 22(b) no environantal impact statement or environmental assessment need be

'N epared in connection with the issuance of the amendments.

9 L.QN.C.LMilQM

.; s-Cournission has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endar.gered by operation in the proposed manner, (2) such activities wii' be conducted in compliance with the Commission's regulations,

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and (3) the 'cscance of the amendments will not be i imical to the common

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n defense and security or to the health and safety of the public.

Prirripal Contributor; Leon B. Engle Date: May 12, 1992 5

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DATED:

May 12, 1992 AMENDMENT NO. 69 TO FACILITY OPERATitlG LICENSE NO. NPF-4-NORTH ANNA UNIT 1 AMENDMENT NO.140 TO FACILITY OPERATING LICENSE NO. NPF-7-NORTH Af1NA U11T 2 Docket File NRC & Local PDRs PDil-2 Reading S. Varga, 14/E/4 G. Lainas, 14/H/3

11. Perkow D. Miller L. Engle W. Lefave C. McCracker, OGC D. Hagan, 3302 MNB8 G. Hill (8), P-137 Wanda Jones, MNBC-7103 C. Grimes, ll/f/23 ACRS (10)

OPA OC/LFMB M, Sinkule, R-Il