ML20058E721

Amends 141 & 124 to Licenses NPF-04 & NPF-07,respectively, Deleting Operability Requirement for One Pressurizer Safety Valve in Mode 5 (Cold Shutdown)

ML20058E721
Person / Time
Site:	North Anna
Issue date:	11/01/1990
From:	Berkow H Office of Nuclear Reactor Regulation
To:
Shared Package
ML20058E723	List: ML20058E721 ML20058E726
References
NUDOCS 9011070296
Download: ML20058E721 (14)

Text

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I'6j UNITED STATES

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NUCLEAR REGULATORY COMMISSION l

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WASHINGTON D. C. 20555

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VIRGINIAklECTRICANDPOWERCOMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET N0. 50-338 NORTH ANNA POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 141 License No. NPF-4 1.

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

L A.

The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated May 16,.1990, as supplemented August 31, l

1990, complies.with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Comission; 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 Comission'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; l

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, r

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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)~TechnicalSpecifications

-The Technical Specifications contained in Appendices A and B, as revised through Amendment No.141, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technica1' Specifications.

3.

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

r FOR THE NUCLEAR REGULATORY COMMISSION erbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - 1/II l

Office of Nuclear Reactor Regulation

~ /sttachment:

Changes to the Technical Specifications Date of Issuance: November 01, 1990 L

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ATTACHMENT TO LICENSE AMENDMENT NO.141 TO FACILITY OPEP,ATING LICENSE NO. NPF i DOCKET-NO.-50-338-

-r Replace the following pages of the Appendix "A" Technical S'pecifications' with the enclosed pages as indicated. The revised pages are identified by 4

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

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

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REACTOR C0OLANT SYSTEM 7

ISOLATED LOOP STARTUP t

' LIMITING CONDITION FOR OPERATION 3.4.1.5 A reactor coolant loop cold leg stop. valve shall remain closed until:

a.

The isolated loop has been operating on a recirculation flow of greater than or equal-to 125 gpm for at least 90 minutes and the temperature at the cold leg of the isolated loop is within 20*F of the highest cold leg temperature of the operating loops.

b..

The reactor is suberitical by at least 1.77 percent Ak/k.

APPLICABILITY: ALL MODES.

ACTION:

With the requirements of the above specification not satisfied, suspend startup.

of the isolated loop.

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SURVEILLANCE REQUIREMENTS 4.4.1.5.1. The isolated loop cold leg temperature shall be determined to be within 20*F of:the highest cold leg temperature of. the operating loops within 30 minutes prior to opening the cold leg stop valve.

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4.4.1.5.2 The reactor shall be determined to be subcritical by at least 1.77 percent Ak/k within 30 minutes prior to opening the cold leg stop valve.

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l NORTH' ANNA - UNIT 1 3/4 4-5 Amendment No. 32 l'

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REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOWN l

LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of'one pressurizer code safety valve shall be OPERABLE with a lif t setting of 2485 PSIG i 1%.*

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APPLICABILITY:" MODE 4..

ACTION:

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With no pressurizer code safety valve OPERABLE, immediately suspend all operations _ involving positive reactivity changes and place an OPERABLE l

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RHR loop into operation.

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l SURVEILLANCE REQUIREMENTS L

4.4.2 No' additional Surveillanca Requirements other than those required i

by Specification 4.0.5.

• The lif t setting pressure shall correspond to' ambient condition of the valve at nominal operating temperature and pressure.

NORTH ANNA - UNIT 1 3/4 4-6 Amendment No. 141, l

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'3/4.4 REACTOR' COOLANT SYSTEM-

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i 3/4.4.1 REACTOR COOLANT' LOOPS The plant-is designed to operate with all reactor coolant loops in operation and maintain the DNBR above the design limit during all normal op'erations and l

anticipated transients.

In MODES:1 and 2 with one reactor -coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY l:

within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

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.In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for I

removing decay heat: however, single f ailure considerations require that two loops be

' OPERABLE.

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In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat, but single f ailure considerations require that at least

- two loops be OPERABLE. Thus,if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

After the reactor has shutdown and entered into MODE 3 for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, a minimum RHR system flow rate of 2000 gpm in MODE 5 is permitted, provided there is sufficient.

decay heat removal to maintain the RCS temperature less than or equal to 140'F. Since the decay heat power production rate decreases with time after reactor shutdown, the requirements for RHR system decay heat removal also decrease, Adequate decay heat removalls provided as long as.

the reactor has been shutdown for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after entry into MODE 3 and RHR flow is

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sufficient to maintain the RCS temperature less than or equal to 140'F. The reduced flow rate -

provides additional margin to vortexing at the RHR pump suction while in Mid Loop Operation.

During a reduction in reactor coolant system boron concentration the Specification 3.1.1.3.1 i

requirement to maintain a 3000 gpm flow rate provides sufficient coolant circulation to minimize the effect of a boron dilution incident and to prevent boron stratification.

o The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 324*F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system which could exceed the limits of Appendix G to 10 CFR Part i

50. The RCS wJl be protected against overpressure transients and will not exceed the limits of.

Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures.

The operation of one Reactor Coolant Pump or one RHR pump provides adequate f!ow to ensure mixing, prevent stratification, and produce gradual reactivity changes during boron

. concentration reductions in the Reactor Coolant SystemJ The reactivity change rate associated

with boron reduction will, therefore, be within the capability of operator recognition and control.

The requirement to maintain the boron concentration of an isolated loop greater than or

, equal to the boron concentration of the operating loops ensures that no reactivity addition to the

. core could occur during startup of an isolated loop, Verification of the boron concentration in an idle loop. prior to opening the cold leg stop valve provides a reassurance of tne adequacy of the boron concentration in the isolated loop. Operating the isolated loop on recirculating flow for at

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least 90 minutes prior to opening its cold leg stop valve ensures adequate mixing of the coolant in i

this loop and prevents any reactivity effects due to boron concentration stratification.

- Startup of an idle loop willinject cool water from the loop into the core. The reactivity transient resulting from this cool water injection is minimized by delaying isolated loop startup until l

its temperature is NORTH ANNA - UNIT 1 B 3/4 4-1 Amendment No. 16,32,777,737, 139-l

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3/4.4 REACTOR -COOLANT SYSTEM BASES within 20'F of the operating loops.

Making the reactor subcritical prior to< loop startup prevents any. power spike which could result from this cool water induced reactivity transient.

3/4.4.2 AND 3/4.4.'3 SAFETY VALVES The pressurizer code safety valves operate to prevent-the RCS from being pressurized above its Safety Limit of 2735 psig.

Each safety valve-

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is designed to relieve.380,000 lbs per hour of saturated steam at the valve set point.

Tha relief capacity of a single safety valve i_s adequate to. ' relieve any overpressure condition which could occur during hot shutdown.

In the event that' no safety valves are OPERABLE, an operating RHR loop.

connected to the RCS, or the power operated relifef valvese(PORVs) will provide overpressure relief capability and will prevent RCS overpressurization.

LDuring operation, all pressurizer code safety valves must be OPERABLE to' prevent the RCS from'being pressurized above its safety limit:of 2735 psig.

The' combined relief capacity of all of-these valves is greater than the maximum-surge rate resulting from a complete' loss of load assuming no reactor trip until the first Reactor Protection System trip set poiat is, reached (i.e.,_ ne credit is taken for a direct reactor trip on the loss of load)' and also assuming no operation of the power operatedirelief_ valves or steam dump = valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will:be performed in accordance with the provisions

- e of Section XI of the ASME Boiler and Pressure Code.

The power operated relief valves and steam bubble function to relieve RCS' pressure during all design transients up to and including'the design step load decrease with steam. dump.

Operation of the power operated relief valves minimizes'theLundesirable opening.of the spring-loaded pressurizer code safety valves.. Each PORV has a remotely operated block valve to provide.a positi.ve shutoff capability'should a relief valve become inoperable.

3/4.4.4 -PRESSURIZER' l

The lioit on the maximum water volume in the' pressurizer assures m

that the parameter is maintained within the normal steady state envelope of operation assumed in the SAR.

The limit is consistent with the initial SAR assumptions.' The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance is suf-ficient to ensure that the parameter is restored to within its limit following expected transient operation.

The maximum water volume also ensures that a steam bubble ~is formed and thus the RCS is not a hydra'ulically solid system'.

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NORTH ANNA - UNIT 1 B 3/4 4-2 Amendment No. 37,141,

0 UNITED STATES 4

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L VIRGINIA ELECTRIC AND POWER COMPANY l

OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-339 NORTH ANNA POWER STATION, UNIT NO. 2

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h AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 124 l

License No. NPF-7 1.

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

A.

The application for amendment by Virginia Electric and Power Company, et al., (the licensee) dated May 16, 1990, as supplemented August 31, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B..

The facility.will operate in conformity with the application, J

the provisions of the Act, and the rules and regulations of the.comission; C.

Thereisreasonableassurance(i)thattheactivitiesauthorized-L by this amendment can be conducted without endangering the health L

and safety of the public, and Ni) that such activities will be l-conducted in compliance with the Comission's regulations; 1

D.

The issuance of this amendment will not be inimical to the comon l

l 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.

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Accordingly, the license is amendad by changes to the Technibal Speci-1 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:

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(2) Technical Specifications,

'The Technical Specifications contained in' Appendices A and B, as revised through Amendment No. 124, 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 the date of issuance and shall be. implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION 1

H rbert N. Berkow, Director Project Directorate 11-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: November 01, 1990 i

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~9tB ATTACHMENT TO LICENSE AMENDMENT NO. 124

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1 T0 FACILITY OPERATING LICENSE NO. NPF-7

' DOCKET NO. 50-339 1

' Replace the following.pages of the Appendix "A" Technical Specifications

'with the. enclosed pages,as indicated. The revised pages are-identified by

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

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

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REACTOR COOLANT SYSTEM -

ISOLATED LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.4.1.3 A reactor coolant loop cold. leg stop valve shall remain closed until:

The isolated loop has been operating on a recirculation flow of a.

. greater than or squal to 125 gpm for at least 90 minutes and the temperature at' the cold leg of the isolated loop is within 20*F of the highest cold leg ~ temperature of the operating loops.

.b.

The reactor is subcritical by at least 1.77 perce. t ok/k.

' APPLICABILITY: ALL MODES.

ACTION:

With the requirements of the above specification not satisfied, suspend startup ci-the isolated loop.

SURVEILLANCE REQUIREMENTS 4.4.1.3.1-The isolated loop. cold leg temperature shall be determined to be within 20*F of the highest cold leg temperature of the operating loops within 30 minutes prior to opening the cold leg stop valve.

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4.4.1.3.2 The reactor shall be determined to be subcritical by at least 1.77 L

percent ok/k within 30 minutes prior to opening the cold leg stop valve.

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NORTH ANNA - UNIT 2 3/4 4-5 l

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k SAFETY VALVES-SHUTDOWN UMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a' lift setting of 2485 PSIG A 1%.*

APPLICABILITY: MODE 4.

1 ACTION:

h With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactMty changes and place an OPERABLE RHR loop into operation.

i SURVEILLANCE REQUIREMENTS - 4.4.2 No additional Surveillance Requirements other than those required by Specification

'4.0.5.

'The-lift setting pressure shall correspond to ambient condition of the valve at nominal-operating temperature and pressure, a

NORTH ANNA UNIT 2 3/4 46 Amendment No. 124,

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3/4.4 REACTOR C001. ANT SYSTEM BASES-3/4.4.1 REACTOR COOLANT LOOPS The plant is' designed to operate with all reactor coolant loops in operation and maintain the DNBR above the design limit during all normal operations and

(

anticipated transients.

In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be l

OPERABLE.

In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat l but single failure considerations require that at least two loops be OPERABLE. Thus,if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

After the reactor has shutdown and entered into MODE 3 for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, a i

minimum RHR system flow rate of 2000 gpm in MODE 5 is permitted, provided there is sufficient decay heat removal to maintain the RCS temperature less than or equal to 140'F. Since the decay:

heat power production rate decreases with time after reactor shutdown, the requirements for RHR system decay heat removal also decrease. Adequate decay heat removalis provided as long as the reactor has been shutdown for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after entry into MODE 3 and RHR flow is

. sufficient to maintain the RCS temperature less than or equal to 140'F. The reduced flow rate provides addihonal margin to vortexing at the RHR pump suction while in Mid Loop Operation.

'During a reduction in reactor coolant system boron concentration the Specification 3.1.1.3.1 requirement to maintain a 3000 gpm flow rate provides sufficient coolant circulation to minimize

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the effect of a boron dilution incident and to prevent boron stratification.

1 The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less I

than or equal to 340'F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system which could exceed the limits of Appendix G to 10 CFR Part

50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressunzer and thereby providing a

. volume for the pnmary coolant to expand into or (2) by restricting starting from the RCPs to when the secondary water temperature of each steam generator is less than 50'F above each of the l

RCS cold leg temperatures.

The requirement to maintain the boron concentration of an isolated loop greater than or equal to the boron concentration of the operating loops ensures that no reactivity addition to the core could occur during startup of an isolated loop. Venfication of the boron concentration in an idle loop prior to opening the cold leg stop valve provides a reassurance of the adequacy of the boron concentration in the isolated loop. Operating the isolated loop on recirculating flow for at least 90 minutes pnor to opening its cold leg stop valve ensures adequate mixing of the coolant in this loop and prevents any reactivity ettects due to boron concentration stratification.

1 Startup of an idle loop willinject cool water from the loop into the core. The reactivity

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transient resulting from this cool water injection is minimized by delaying isolated loop startup until E

its temperature is within 20*F of the operating kops. Making the reactor subentical prior to loop startup prevents any power spike which could result from this cool water induced reactivity transient.

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NORTH ANNA - UNIT 2 B 3/4 4 1 Amendment No. 729 122 I

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BEACTORCOOLWTSYSTEM l

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3/4 41 AND 3/4 4.3 SAFETY AND RELIEF VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relleve'380,000 lbs per hour of saturated steam at the valve set point. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during hot shutdown. In the.

1

- event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, or the power operated relief valves (PORVs) will provide overpressure relief capability and will R

prevent RCS overpressurization.

1 During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of i

all of these valves is greater than the maximum-surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protection System trip set point is reached l

(i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no j

operation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

xThe power operated relief valves (PORVs) and steam bubble function to relieve RCSs pressure during all design transients up to and including the design step load decrease with steam dump, Operation of the PORVs minimizes the undesirable opening of the spring loaded pressurizer code safety valves. Each PORV hes a remotely operated block valve to provide a positive shutoff capability should a relief valve become Inoperable.

3/4.4.4 PRESSURIZER

~

The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady state envelope of operation assumed in the SAR. The limit is consistent with the inillal SAR assumptions. The 12-hour periodic surveillance is suffident to ensure that the parameter is restored to within its ' limit following expected transient i

operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid-system.

The requirement that a minimum number of

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pressurizer heaters be OPERABLE ensures that the plant will be able to establish natural circulation.

NORTH ANNA UNIT 2 83/442 Amendment No. 124, I