Text

Amends 137 & 120 to Licenses NPF-4 & NPF-7,respectively, Revising Tech Specs to Enhance RHR Reliability by Including Reduction in Min RHR Flow Rates from 3,000 Gpm to 2,000 Gpm
	ML20059F204
Person / Time
Site:	North Anna
Issue date:	08/27/1990
From:	Berkow H; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20059F209	List: ML20059F204; ML20059F217;
References
	NUDOCS 9009110139
	Download: ML20059F204 (28)
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UNITED STATES

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W NUCLEAR REGULATORY COMMISSION 3

WASHING TON, D. C. 20666 VIRGINIA ELECTRIC AND POWER COMPANY j

OLD DOMINION ELECTRIC COOPERATIVE i

DOCKET N0. 50-338 NJRTHANNAPOWERSTATION.UNITN0.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 137 License No. NPF-4 1.

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

A.

The application for amendment by Virginia Electric and Power Company 1

{

etal.,(thelicensee)datedJune8,1990,assupersededJune 13, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and l

regulations set forth in 10 CFR Chapter I; 8.

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

Thereisreasonableassurance(1)thattheactivitiesauthorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities wi,11 be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied, t

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Accett'.ngly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.0.(2) of Facility Operating License No. NPF-4 is hereby amended to read as follows:

(2) Technical Specificatiens The Technical Specifications contained in Appendices A and B, as revised through Arendment No.137

, 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 He ert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications

. Date of Issuance: August 27, 1990 n

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ATTACHMENT TO LICENSE AMENDMENT NO. 137

_TO FACILITY OPERATING LICENSE NO. NP,?-4 DOCKET N0. 50-338 l

l Replace th'e following pages of the Appendix "A" Technical Sptcifications-with the enclosed pages as indicated. The revised pages are identified by i

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

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

.P., ale 3/4 4-3a 3/4 9-8 3/4 9-8a B 3/4 4-1 B 3/4 9-2 f

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i REACTOR COOLANT SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a.

At. least two of the coolant loops listed below shall be OPERABLE:

1.

Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,*

2.

Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,*

3.

Reactor Coolant Loop C and its associated steam generator and reactor coolant pump,"

4.

Residual Heat Removal Subsystem A,**

5.

Residual Heat Removal Subsystem B.**

1 b.

At least one of the above coolant loops shall be in operation.***

APPLICABILITY: MODES 4 and 5.

ACTION:

With less than the above required loops OPERABLE, immediately a.

initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTOOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant j

loop to operation.

"A reactor coolant pump shall not be started with one or more of the RCS cold' leg temperatures less than or equal to 324*F unless the secondary water temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures.

- The offsite or emergency power source may be inoperable in MODE 5.

- - All reactor coolant pumps and residual heat removal pumps may be de-energized for up to I hour provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10*F below saturation temperature.

NORTH ANNA - UNIT 1 3/4 4-3 Amendment No. 16,,77, 117

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FIEACT.QR_CDO.LANT RYRTEM

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SURVEILLANCE REQUIREMENTS l

4.4.1.3.1 The required RHR subsystems shall be demonstrated OPERABLE per Specification 4.7.9.2.

4.4.1.3.2 The required reactor coolant pump (s), if not in operation, shall be determined l'

to be OPERABLE once per 7 days by venhing correct breaker alignment and indicated power availability.

4.4.1.3.3 The required steam generator (s) shall be determined OPEFtABLE by ventying secondary side water level to be greater than or equal to 17% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4 4.4'.1.3.4 At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , venfy at least one coolant loop to be in operation and circulating reactor coolant by:

a. Verifying at least one Reactor Coolant Pump is in operation.

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b. Venfying at least one RHR Loop is in operation and, j

1.

if the RCS temperature >140' F or the time since entry into MODE 3 is

<100 hours, orculating reactor coolant at a flow rate 23000 gpm.

l or i

2. if the RCS temperature 5140' F and the time since entry into MODE 3 is 2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months , circulating reactor coolant at a flow rate 22000 gpm to -

remove decay heat.

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NC4TH ANNA UNIT 1 3/4 4 3a Amendment No, 32, j

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i REFUELING OPERATIONS l

s/a.s.a RHIDUAL MEAT REMOVAL (RNR1 AND COOLANT CIRCULATIO,g l

NORMAL. WATER LEVEL l

l L!MITING CONDITIONS FOR OPERATION 3.9.8.1 At least one RHR loop shall be OPERABLE

and at least one RHR loop shall be in operation.

l APPLICABILITY' MODE 6 With the reactor vessel water level greater than or equal to i

l 23 feet above the top of the reactor pressure vessel flange, ACTION: a. With less than one RHR loop OPERABLE,immediately truttate corrective actions to return the required RHR :. oops to OPERABLE status as soon as possible,

b. With less than one RHR loop in operation, except as provided in c. below, l

l suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

c. The RHR loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months l

period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

d. The provisions of Specification 3.0.3 are not applicable.

l SURVEILL ANCE REQUIREMENTS 4.9.8.1.1 Venfy the required RHR loop to be OPERABLE per Specification 4.0.5.

l 4.9.8.1.2 At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , ve'nfy at least one RHR Loop is in operation and,

a. If the RCS temperature >140' F or the time since entry into MODE 3 is

<100 hours, circulating reactor coolant at a flow rate 23000 gpm.

I b if the RCS temoerature s140' F and the time since entry into MODE 3 is 2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months , circulating reactor coolant at a flow rate 22000 gpm to remove decay.

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The normal or emergency power source may be inoperable for each RHR loop.

NORTH ANNA UNIT 1 3/4 9 8 Amendment No. 32,137,

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FIEFUELING OPERATIONS 1

RESIDUAL H2AT REMOVAL (RHR) AND COOLANT CIRCULATION l

LOW WATER LEVM l

l LIMITlMG CONDITION FOR OPERATION 4

1 3.9.8.2 Two independent RHR loops shall be OPERABLE' with 31 least one loop in i

operation.

(

l APPLIC AtlLITY: MODE 6 with the reactor vessel water level less than 23 feet above l

the top of the reactor pressure vessel fiange, l

i ACTION: a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.

b. With less than one RHR loop in operation, suspend all operations involving an l

t increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations provxling direct access from the containment atmosphere to the outside atmosphere l

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

c. The provisions of Specification 3.0.3 are not applicable.

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i SURVElLLANCE REQUIREMENTS 4.9.8.2.1 Venty the required RHR loops to be OPERABLE per Specification 4.0.5.

l 4.9.8.2.2 At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , venty at least one RHR Loop is in operation and, i

a

!! the RCS temperature >140' F or the time since entry into MODE 3 is i

<100 hours, circulating reactor coolant at a flow rate 23000 gpm.

t b if the RCS temperature s140' F and the time since entry into MODE 3 is j

2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months , circulating reactor coolant at a flow rate 22000 gpm to remove decay.

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The normal or emergency power source may be inoperable for each RHR loop.

l NORTH ANNA UNIT 1 3/4 9 8a Amendment No. H,137

5/4.4 MACTOR COOLANT SYSTEM BASES 3/a.a.1 REACTOR COOLANT LOOPS i

The plant is designed to operate with all reactor coolant loops in operation and maintain I

DNBR above 1.30 during all normat operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this spoofication requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

in MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single f ailure considerations require that t vo loops be j

OPERABLE.

In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat, 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 0.0278 hours 1.653439e-4 weeks 3.805e-5 months a

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minimum RHR system flow rate of 2000 gpm in MODE 5 is permitted, provided there is sufficient i

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 GHR i

system decay heat remova! also decrease. Adequate decay heat removalls provided as long as

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the reactor has been shutdown for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after entry into MODE 3 and RHR flow is sufficient to maintain the RCS temperature less than or equalto 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 requirement to maintain a 3000 gpm flow rate p*0vides sufficient coolant circulation to minimize j

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

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

50. The RCS wil! be protected against overpressure transients and will not exceed the limits of Appendix G by restncting 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 flow to ensure mixing, prevent stratification, and produce gradual reactivity changes during boron.

concentration reductions in the Reactor Coolant System. 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. Venfication of the boron concentration in an idle loop prior to opening the cold leg stop valve provices 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 prior to opening its cold leg stop valve ensures adequate mixing of the coolant in this loop and prevents any reactivity effects due to boren 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 its temperature is NORTH ANNA UNIT 1 B 3/4/ 41 Amendment No. 16,32.777,137,

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i 3/4.4 REACTOR COOLANT SYSTEM BASES i

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I within 20'F of the operating loops.

Making the reactor subcritical prior i

to loop startup prevents any power spike which could result from this cool water induced reactivity transient.

j 3/4.4.2 AND 3/4.4.3 SAFETY VALVES l

The pressurizer code safety valves operate to prevent the RCS from l

l being pressurized above its Safety Limit of 2735 psig.

Each safety valve i

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

The relief capacity of a single safety valve is adequate i

to relieve any overpressure condition which could occur during shutdown.

l In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be

)

OPERABLE to prevent the RCS from being pressurized above its safety j

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 l

load assuming no reactor trip until the first Reactor Protection System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only I

during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

l The power operated relief valves and steam bubble function to relieve '

RCS pressure during all design transients up to and including the design step i

load decrease with steam dump.

Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.

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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 l

of operation assumed in the SAR.

The limit is consistent with the 1

initial SAR assumptions.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance is suf-ficient to ensure that the parameter is restored to within its limit following expected transient operation.

The maximum wat** volume also i

ensures that a steam bubble is formed and thus the RCS :s hat a hydraulically solid system.

NORTH ANNA - UNIT 1 B 3/4 4-2 Amendment No. 32

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3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

1) the reactor'will remain suberitical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the accident analyses.

The value of 0.95 or less for K,Ne boron includes a 1%

Similarly.

ak/k conservative allowance for uncertainties.

concentration of 2300 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

t 3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products.

This decay time is consistent with the assumptions used in the accident analyses.

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration cle:ure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment.

The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon a lack of containment pressurization potential while in the REFUELING MODE.

3/4.9.5 COMMUNICATIONS The requirement for communication capability ensures that refueling station personnel can be promptly informed of significant changes in the l

facility status or core reactivity conditions during CORE ALTERATIONS.

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NORTH ANNA - UNIT 1 B 3/4 9-1 Amendment No. 93

REFUELING OPERAft0NS BASES l

3/4.9.6 MANIPULATOR CRANE OPERABILITY i

The OPERABILITY requirements for the manipuletor cranes ensure that:

1)mani-pulator cranes will be used for movement of control rods and fuel assemblies; 2))e crane has sufficient load capacity to lift a control rod or fuel assembly, and 3 the core internals 'nd pressure vessel ara protected from excessive lifting force in the a

l event they are inadvertently engaged during lifting operations.

l 3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped, 1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuel in the storage racks will not result in a critical array.- This i

assumption is consistent with the activity release assumed in the accident analyses.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal-(RHR) loop be in opera-tion ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140'F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron l

stratification.

After the reactor has shutdown and entered into MODE 3 for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , a minimum RHR system flow rate of 2000 gpm in MODE 6 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 recuirements for RHR system decay heat removal also decrease. Adequate decay heat removal is provided as long as the reactor has been shutdown for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after entry into MODE 3 and RHR flow is sufficient to maintain the RCS temperature less than or equal to 140'F. The reduced ficw 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 requirenent to maintain a 3000 gpm flow rate provides sufficient coolant circulation to minimize the effect of a boron dilution incident and to prevent boron stratifi-cation.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capa-bility. With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

3/9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

NORTH ANNA - UNIT 1 B 3/4 9-2 Amendment No. 32,137, i

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UNITED STATES NUCLEAR REGULATORY COMMISSION a

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VIRGINIA ELECTRIC AND POWER COMPANY DLD 70 MINION ELECTRIC COOPERATIVE DOCKET N0. 50-339 NORTH ANNA POWER STATION.' UNIT N0. 2 AMENDMENT TO FACILITY. OPERATING LICENSE Amendment No. 120 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 etal.,(thelicensee)datedJuneA,1990,assupersededJune13, 1990, complies with the standards and re EnergyActof1954,e.samended(theAct)quirementsoftheAtomic

, and the Comission's rules and regulations set forth in 10 CFR Chapter-I; r

B.

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

Thereisreasonableassurance(1)thattheactivitiesauthorized 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 comon 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 Comission's regulations and all applicable requirements have been satisfied.

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-Accordingly. the license is amended by changes to the Technical Spect -

fications as indicated in the: attachment to this license amendment, and paragr.aph 2.C.(2) of Facility Operating License No. NPF-7 is hereby l

amended to read as follows:

i (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 120, 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 icsuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION H bert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office f Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: August 27, 1990 e

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ATTACHMENT TO LICENSE AMENDMENT N0.' 120 l

TO.iACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-339 i

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

l' with the enclosed pages as indicated. The revised pagesare identified by I

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

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

]

o ll III 3/4 1-4 3/4.)-4a 3/4 4-3a 3/4 9-9 3/4 9-9a B 3/4 4-1 B 3/4 9-2 s

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INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS u

SECTION Pa2' 2.1 SAFETY LI'4ITS Kw d c to r C o re...................................................

2-1 Reactor' Coolant System Pressure...............................

2-1

'..ao 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip Setpoints........................................

2-5 BASES SECTION Page 2.1 SAFETY LIMITS Reactor Core..................................................-

B 2-1 Reactor Coolant System Pressure...............................

B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS i

Reactor Trip Setpoints........................................

B 2-3 NORTH ANNA - UNIT 2 II m

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LIMITING CONDmONS FOR OPER ATION AND SURVEILLANCE REQUIREMENTS SECTION EAQt 3/40 A P P L I C A B i L l TY....................................................................................... 3/4 0 1

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3/41 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown M argin. Ta vg > 200' F........................................................... 3/4 1 1 Shutdown M argin Tavg s 200' F............................................................ 3/4 1 3 Boron Dilution Reactor Coolant Flow........................................................ 3/4 1 4 Bornn Dilution Valve Po sition...................................................................

3/4 1 4a Moderator Temperature Coefficient.......................................................... 3/4 1 5 Minimum Temperature for Criticality............................................................

3/4 1 7 l

3/4.1.2 BORATION SYSTEMS -

Flow P at h s - S hutdown............................................................................. 3/4 1 8 l

Row P at h s O p e rating.............................................................................. 3/4 1 9 Charging Pump S h u tdow n.............................................................. 3/4 1 1 1 Ch arging Pu mps Ope rating..................................................................... 3/4 1 12 :

Borated Water Sources Shutdown....................................................... 3/4 1 13 E2 orated Water Sources Operating..

..................................................3/4114 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group H eig h t............................................................................................ 3/4 1 1 6 Position Indicator Channels Operating................................................... 3/41 19 Position Indicator Channels Shutdown.................................................... 3/41 20 Rod D rop Ti me........................................................................................ 3/4 1 21 Shutdown Rod Inse rtion Limit................................................................. 3/4122 Co nt rol Rod Insertion Limits.................................................................. 3/4 1 23 NORTH ANNA UNIT 2 III Amendment No.120 t

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REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T,yg LESS THAN OR EQUAL TO 200'F SURVEILLANCE REQUIREMENTS I

3.1.1.2 The SHUTOOWN MARGIN shall be greater than or equal to 1.77% delta k/k.

APPLICABILITY: MODE 5.

ACTION:

With the SHUTDOWN MARGIN less than 1.77% delta k/k, insnediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 12.950 ppm. boron or equivalent until. the required l

SHUTDOWN MARGIN is restored.

S_ULRVEILLANCEREQUIREMENTS 4.1.1.2' The SHUT 00WN MARGIN shall be determined to-be greater than or equal to 1.77% delta k/k:

Within one hour after detection of an inoperable control rod (s) and a.

at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod (s).

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by considaration of the following factors:

1.

Reactor coolant system boron concentration,.

2.

Control rod position, 3.

Reactor coolant syster. average temperature, 4.

Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and l

6.

Samarium concentration.

NORTH ANNA - UNIT 2 3/4 1-3 Amendment No.54 l

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i REACTIVITY, CONTROL SYSTEM 3/a.1.1.3 EORON DILUTION REACTOR COOLANT FLOW LIMITINd ONDITIONS FOR OPERATION 3.1.1.3.1 The flow rate of reactor coolant through the reactor coolant system shall be 23000 gpm whenever a reduction in Reactor Coolant System boron concentration is being made.

APPLICABILITY: All MODES ACTION: With the flow rate of reactor coolant through the reactor coolant system <3000 gpm.

' t' immediately suspend all operations involving a reduction in boron concentration of the Reactor Coolant System.

SURVEILLANCE REQUIREMENTS-4.1.1.3.1 The flow of reactor coolant through the reactor coolant system shall be determined to be 2 3000 gpm within one hour prior to the start of and at least once per hour during a reduction in the Reactor Coolant System boron concentration by either:

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a. Verifying at least one reactor coolant pur@ is in operation.

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b. Verifying that at least one RHR pump is in operation ard supplying 23000 gpm through the reactor coolar't system.

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i Amendment No.120,')

NORTH ANNA UNIT 2 3/414 4

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f REACTIVITY CONTROL SYSTEM BORON DILUTION VALVE POSITION LIMITING CONDITION FOR OPERATION l'

3.1.1. 3. 2 The following valves shal'1 be locked, sealed or otherwise secured iin I

the closed position except during planned boron dilution or makeup activities:

a. CH-140 or r

b.

2-CH-160, 2-CH-156, FCV-2114B and FCV-21138.

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APPLICABILITY: MODES 3, 4, 5, and 6.

ACTION:

With the above valves not locked, sealed or otherwise secured in the closed l

position:

1) suspend all operations involving positive reactivity changes or CORE ALTERATIONS, 2) lock, seal or'otherwise secure-the valves in the closed position within 15 minutes, and 3) verifyJthat the SHUTDOWN MARGIN is greater than or equal to 1.77% delta k/k within 60 minutes, i

SURVEILLANCE REQUIREMENTS r

4.1.1. 3. 2 The above listed valves shall be verified to be-locked, sealed or l

otherwise secured in the closed position within 15 minutes after a planned boron dilution or makeup activity.

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i NORTH ANNA - UNIT 2 3/4 1-4a Amendment No. 120, b

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REACTOR COOLANT SYSTEM SHUT 00'.l?!

SURVEILLANCE REQUIREMENTS

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4.4.1.3.1 The required RHR subsystems shall be demonstrated OPERABLE PER SPECIFICATION 4.7'9.2.

4.4.1.3.2 The required reactor coolant pump (s) if not in operation, shall be determined to be OPERA 8LE once per 7 days by verifying correct breaker alignments and indicated power availability..

4.4.1.3.3 The required steam generator (s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 17% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.3.4 At least once per 12. hoursa verify at least one coolant loop to be in operation and circulating reactor coolant by:

Verifying at least one Reactor Coolant Pump is in operation..

a.

or s

b.

Verifying at least one RHR toop is in operation and, s.

1.

if the RCS temperature > 140'F or the time since entry into MODE 3 is < 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , circulating reactor coolant

'at' a flow rate 1 3000 gpm, or 2.

if the RCS temperature i 140*F and the time since entry into MODE 3 is 1 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , circulating reactor coolant at a flow rate 1 000 gpm to remove decay heat.

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NORTH ANNA - UNIT 2 3/4 4-3a Amendment No'. 120, 1

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. REACTOR COOLANT SYSTEM l

I ISOLATED LOOP j

v LIMITIN3 CON 0fT!ON FOR OPERATION j

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3.4.1.2 The boron concentration of an isolated loop shall be maintained greater than or equal to the boron concentration of the operating loops, i

unless the loop has been drained for maintenance, i

APPLICABILITY:

MODES 1, 2, 3, 4 and 5.

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ACTION:

3 With the requirements of the above specification not satisfied, do not open the isolated' loop's stop valves; either increase the boron concentri, ion of the isolated loop to within the limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least-HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months with the unisolated portion of the RCS borated to a SHUT 00WN MARGIN equivalent to at least 1.77% ok/k at 200*F.

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i SURVEILLANCE REQUIREMENTS 1

J 4.4.1.2 The boron concentration of an isolated loop shall be determined to be greater than or equal to the boron concentration of the operating loops at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and within 30 minutes prior to opening either the hot leg or cold leg stop valves of an isolated loop.

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NORTH ANNA - UNIT 2' 3/4 4-4 1

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' REFUELING OPERATIONS sla.s.a RmalDUAL HEAT REMOVAL (RNR) AND' COOLANT CIRCULA~rlON' l2 NORMAL WATER LEVEL l-L!MITING CONDITION FOR OPERATION 3.9.8.1 At least one RHR loop shall be OPERABLE' and at least one RHR loop shall be in operation, APPLICABILITY: MODE 6 With the reactor vessel water level greater than or equal to 23 feet above the top of the reactor pressure vessel flange.

ACTION' a. With less than one RHR loop OPERABLE. Immediately initiate corrective actions to retum the required RHR loops to OPERABLE status as soon as posS:ble.

b. With less than one RHR loop in operation, except as provided in c. below, l

suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

c. The RHR loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months l

l penod during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

d. The provisions of Specification 3.0.3 are not aoplicable.-

l SURVE!LLANCE R E Q'UIR E ME NTS 4.9.8.1.1 Venty the required RHR loop to be OPERABLE per Specification 4.0.5.

l 4.9.8.1.2 At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . verify at least one RHR Loop is in operation and, l

a. if the RCS temperature >140* F or the time since entryinto MODE 3 is

<100 hours, circulating reactor coolant at a flow rate 23000 gpm.

b. - if the RCS temperature 5140* F and the time since entry into MODE 3 is l

2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months , circulating reactor coolant at a flow rate 22000 gpm to l

remove decay heat.

The normal or emergency power source may be inoperable for each RHR loop.

NORTH ANNA UNIT 2 3/4 9 9 Amendment No.120, e

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REFUELING OPERATIONS l

RESID11ALMEAT. REMOVAL.. fRHR) AND COOLANT CIRCULATION' l

LOW WATER LEVELB l

LIMITING CONDITION FOR OPERATION l

3.9.8.2 Two independent RHR loops shall be OPERABLE' with at least one loop in -

i operation, j

'q APPLICABILITY! MODE 6 with the reactor vessel water levelless than 23 feet above the top of the reactor pressure vessel flange.

ACTION: a. With less than the required RHR loops OPERABLE. Immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.

b, With less than one RHR loop in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment peretrations providing direct access from the contalament atmosphere to the outside atmosphere -

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

c. The provisions of Specification 3.0.3 are not applicable.

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i SURVEILL ANCE REQUIREMENTS 4.9.8.2.1 Venfy the required RHR loops to be OPERABLE per Specification 4.0.5.

.l 4.9.8.2.2 At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , verify at least one RHR Loop is in operation and,

a. if the RCS temperature >140' F or the time since entry into MODE 3 is

<100 hours, circulating reactor coolant at a flow rate 23000 gpm.

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b. If the RCS temperature 5140' F and the time since entry into MODE 3 is l

2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months , circulating reactor coolant at a flow rate 22000 gpm to j

remove decay heat.

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The normal or emergency power source may be inoperable for each RHR loop.

NORTH ANNA.tJNIT 2 3/4 9 9a Arnendment No.120, 1

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t REFUELING OPERATIONS l

CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM-V l

LIMITING CONDITION FOR OPERATION l

3.9.9 The Containment Purge and Exhaust isolation system shall be OPERABLE.

APPLICABILITY:

MODE 6.

ACTION:

With the Containment Purge and Exhaust isolation system inoperable, close each of the Purge and Exhaust penetrations providing direct access from the contain-ment atmosphere to the outside atmosphere.

The provisions of Specification 3.0.3 are not applicable.

1 SURVEILLANCE REQUIREMENTS I

4.9.9 The Containment Purge and Exhaust isolation system shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of and at least once.per 7 days during CORE ALTERATIONS by verifying that containment Purge and Exhaust isola-tion occurs on manual initiation and on a high radiation test signal from the l

containment gaseous and particulate radiation monitoring instrumentation f

channels.

NORTH ANNA - UNIT 2 3/4 9-10 6

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3/4.4 RE ACTOR OOOLANT SYSTEM BASES 3/a.d.1 -

REACTOR COOLANT LOOPS-l l

The plant is designed to operate w!'h all reactor coolant loops in operation and maintain 5

i DNBR above'1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation this specnictbn requires that the plant be in at least 1

l HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

in MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single f ailuro considerations require that two loops be J

OPERABLE.

In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for rernoving decay Natl but single f ailste considerations require that at least two loops be OPERABLE. Thus,if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops te be OPER ABLE.

After the reactor has shutcown and entered into MODE 3 for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , 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 removalis provided as long as the reactor has been shutdown for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months 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 i

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

The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less 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 agairist 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 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. Verification of the boron concentration in an idle loop prior to oper.ing the cold leg stop valve provides a reassurance of the adequacy of the boron concentration in the isoiated loop. Cperating the isolated loop on recirculating flow for at least 90 minutes prior to opening its cold leg stop valve ensures adequate mixing of the coolant in -

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

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Startup of an idle loop willinject cocl water from the loop into the core. The reactivity transient resulting from this cool water injection is minimized by delaying isolated loop startup until its temperature is within 20'F of the operating loops. Making the reactor subcritical prior to loop startup prevents any power spike which could result from th:2 cool water induced reactivity transient.

i NORTH ANNA UNIT 2 8 3/4 41 Amendment No.123, 1

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REACTOR COOLANT SYSTEM BASES 3/4.4.2 and 3/4.4.3 SAFETY AND RELIEF VALVES i

The pressurizer code safety valves operate to prevent the RCS from being i

pressurized above its Safety Limit of 2735 psig.

Each safety valve is designed to relieve 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 shutdown.

In the event that no safety valves are OPERA 8LE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpres-surization.

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 all of these valves is greater than the maxi-~

mum surge rate resulting from a complete loss of load assuming no reactor trip -

until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also i

assuming no operation of the power operated relief valves or steam dump i

valves.

Demonstration of the safety valves' lift settings will occur only during I

shutdown and will be performed in accordance with the provisions of Section XI of the ASM Boiler and Pressure Code.

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Th' ' power operated relief valves (PORVs) and steam bubble function to i

relieve RCS pressure during all design transients up to and including the i

design step load decrease with steam dump.

Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer code safety valves, i

Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.

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3/4.4.4 PRESSURIZER l

The limit on the maximum water volume in the pressurizer assures that the-i 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-hour p'riodic surveillance is sufficient to ensure that the parameter e

l is restored to within its 10.ait following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus tne l

RCS is not a hydraulically solid system.

The requirement that a minimum number of pressurizer heaters be OPERA 8LE ensures that the plant will be able to establish natural circulation.

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3 NORTH ANNA - UNIT 2 8 3/4 4-2 i

1 3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION i

The limitat.fons on reactivity conditions during REFUELING ensure that:

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1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boroti dilution incident in the accident analyses. The value of 0.95 or less for K,Ne boron includes a 1%

ak/k conservative allowance for uncertainties.

Similarly.

concentration of 2300 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS i

The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containmer.t will be restricted from leakage to the environment.

The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.

3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling I

station personnel can be promptly informed of significant changes inithe facility status or core reactivity conditions during CORE ALTERATIONS.

NORTH ANNA - UNIT 2 B 3/4 9-1 Amendment No. 78 l

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REFUELING OPERATIONS l-1 l

BASES 3/4.9.8 MANIPULATOR CRANE OPERABILITY The OPERABILITY requirements for the manipulator cranes ensure that: 1) manipulator cranes will be used for movement of control rods and fuel assemblies,2) each crane has sufficient load capacity to ift a control rod or fuel assembly, and 3) the core Intemals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TR AVEL SPENT FUEL PIT The restriction on movement of loads in excess of the nominalweight that of a fuel and -

control rod assemblies and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped,1) the activity release will be limited to that -

contained in a single fuel assembly, and 2) any possible distortion of fuelin the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the l

accident'.

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3/4.9.8 RESIDU AL' HE AT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay hect and maintain the water in the reactor pressure vessel below 140'F as required during the REFUELtNG MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.

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

l minimum RHR system flow rate of 2000 gpm in MODE 6 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 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after entry into MODE 3 and RHR flow is sufficient to maintain the RCS temperature less than opequal to 140'F. The reduced flow rate,

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

During a reduction in reactor coolant system boron conc 6ntration the Specification 3.1,1.3.1 requirement to maintain a 3000 gpm !!ow rate provides sufficient coolant circulation to minimize the effect of a boron dilution incident and to prevent boron stratification.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capabilty. With the reactor vessel.

head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus,in the event of a f ailure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOL ATION' SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated uper, detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive matenal from the containment atmosphere to the environment.

NORTH ANNA UNIT 2 8 3/4 9 2 Amendment No. 120,

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ML20059F204

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