ML20059G730

Amends 139 & 122 to Licenses NPF-4 & NPF-7,respectively, Increasing Max Allowable Rod Drop Time Needed to Accommodate New Fuel Assembly Design Designated North Anna Improved Fuel

ML20059G730
Person / Time
Site:	North Anna
Issue date:	09/06/1990
From:	Berkow H Office of Nuclear Reactor Regulation
To:
Shared Package
ML20059G733	List: ML20059G730 ML20059G735
References
NUDOCS 9009130120
Download: ML20059G730 (14)

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i UNITED STATES :

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VIRGINIA ELECTRIC AND POWER COMPANY l

s OLD DOMINION ELECTRIC COOPERATIVE J

DOCKETNO.50-321 NORTH ANNA POWER' STATION'. UNIT NO. 1

'I AMENDMENT TO FACILITY OPERATING LICENSE O

r Amendment No.139

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License No..NPF-4

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

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-A.

The application for_ amendment by Virginia Electric and Power Company et a*., (the licensee) dated Janaury 15, 1990, as supplemented by; letter dated August 15, 1990, complies with the standards and require-1 ments 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; v

y B..

The facility will operate in conformity with the application, the provisions of the Act,uand the rules and regulationstof the Comission -

C.

There is reasonable assurance'(i)'that the activities authorized-by this amendment can-be conducted without endangering-the health f

and safety'of the public, and (ii) that such activities will:be.

conducted-in compliance with the Comission's regulations; D.

LThe issuance of this' amendment will not'betinimical to the common

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-defense and security or to-the health and safety of the public;;

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E.

The issuance of this amendment <is in accordance with 10 CFR Part 51 of.the Comission's regulations and all applicable requirements

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have been satisfied.

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'Accordingly, the license.is' amended by changes to thel echnical Speci-T fications as indicated in the attachment to this~1icense amendment,.

and paragraph 2.D.(2) of Facility Operating License No. NPF-4 is hereby:

. amended to read as follows:

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._(2) Technical; Specifications' i

The Technical Specifications contained in Appendices A-and B,.-

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. as revised through Amendment No.139, are hereby incorporated in the= license. The license 9 shall operate the facility in-x 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 H rbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II 0ffice of Nuclear Reactor Regulation

Attachment:

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Changes to the Technical Specifications-Date of Issuase:

September 6, 1990 J

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k ATTACHMENT TO LICENSE AMENDMENT NO. 139' d

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L TO FACILITY OPERATING LICENSE NO. NPF-4 i

DOCKET NO. 50-338 l

Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified hy.

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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REACTIVITY: CONTROL SYSTEMS R00 OROP TIME.

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LIMITING CONDITION FOR OPERATION J

p 3.1.3.4-The individual full' length (shutdown and control) rod drop; time-.

E from the fu1~1y withdrawn position shall be <'2.7. seconds from beginning l.

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• l of decay of; stationary gripper coil voltage to dashpot ~ entry with:

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p avg-500*F, and a.

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All reactor coolant pumps operating.

3 APPLICABILITY: MODES 1 and 2.

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ACTION:

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4 a.

With the drop time of any fus1 length rod determined toc exceed t

the above limit, restore tb s rod drop time to within the above limit prior to proceeding to MODE 1 or'2~.

"k b.

With the rod drop times within limits but determined with 2

- reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to:

1.

< 66% of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are open, or 2.

< 71%'of RATED-THERMAL POWER when the reactor coolant stop

-c Valves in-the nonoperating loop are closed.

'I SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full length rods'shall be demonstrates through measurement prior to reactor criticality:

a.

For all' rods following each removal of the reactor vessel head, b.

For specifically affected individual rods following any main-tenance on or modification to the control rod ~ drive system which could affect the drop time of those specific rods, and c.

At least once per 18 months.

l NORTH ANNA-UNIT 1 3/4 1-23 Amendment No.139 i

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r prarTIVITY CONTROL SYSTEMS-1 1HUTDOWN~RODINSERTIONLIMIT LI -

4 LIMITING ~ CONDITION ~FOR OPERATION 3.1.3.5' All shutdown rods shall be fully withdrawn.

.c x APPLICABILITY: MODES 1* and 2*#

ACTION::

u With a maximum of one shutdown rod n'ot fully withdrawn, except for s'urve111ance testing pursuant to Specification.4.1.3.1.2, within one hour either:

a..

Fully withdraw'the rod, or 3

.b.

Declare: the rod.to be ' inoperable and apply Specification 3.1.3.1 ~

j SURVEILLANCE' REQUIREMENTS t

4.1.3.5.Each shutdown rod shall be-determined to be fully withdrawn:

'a.

Within 15 minutes prior to initial-control red bank withdrawal during an approach to. reactor criticality, and l'

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-i At least once per 12.l.:"rs thereaf ter.

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?See Special Test Exceptions 3.10.2 and 3.10.3.

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NORTH AN'A-UNIT 1 3/4 1-24 Amendment No.138, N

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3/4,4 REACTOR COOLANT SYSTEM ~

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' BASES'

, 3/4.4.1 REACTOR COOLANT L0 ops

The plant is designed to operate with all' reactor coolant loops in operation,

'l and maintain the DNBR above the design limit during all normal operations _ and anticipated transients.

In M00ES 1 and 2 with one reactor coolant loop not in ~

b.

' 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 capaixtity for removing decay heat; however, single failure 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 Ic remuing decay heat, but single failure considerations require that at least

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two loopc be OPERABLE. Thus,if the reactor coolant loops are not OPERABLE, this w

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

_~i 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 the e11ect of a boron dilution incident and to prevent boron stratification, i

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 will be protected against overpressuie 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 generatoris 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.

i 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 l core could occur during startup of an isolated loop Verification of the boron concentration in an i

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 prior to opening its cold leg stop valve ensures adequate mixing of the coolant in this loop and prevents any reactivity effects due to bomn concentration stratification.

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

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NORTH ANNA - UNIT 1 8 3/4 4-1 Amendment No. 16,32.777.737, 139

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f 3/4.4~ REACTOR COOLANT SYSTEM i

BASES 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.

3/4.4.2 AND 3/4.4.3 SAFETY VALVES f

a 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 relieve'380,000 lbs per hour of saturated steam at the A

valve set point.

The relief-capacity of a single safety valve is adequate i

to relieve any overpressure condition which could occur during shutdown.

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

.. 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 valve's

~

is greater than.the maximum surge rate resulting from a complete loss of i

load' assuming no reactor trip until the first Reactor Protection System trip set point is reached (i.e., no credit is taken for a' directs reactor a'

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 during shutdown and will be performed in accordance with the provisions 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 the undesirable opening of the' spring-loaded pressurizer code i

safety valves.

Each PORV has 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'operati" sssumed in the SAR.

The limit is. consistent with the 4

initia sumptions.

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-x ficier.

! that the parameter is restored to within its limit follow ed transient operation.

The maximum water volume also

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ensur steam-bubble is formed and thus the RCS is not a hydra solid system.

h NORTH ANNA - UNIT 1 B 3/4 4-2 Amendment No. 32 v A s

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UNITED STATES -

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' NUCLEAR REGULATORY. COMMISSION i

WASHINGTON, o. C. 20566

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.VIRGINIAELECTRIC:ANDPOWERCOMPANY[

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.0LD DOMINION ELECTRIC COOPERATIVE i

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' NORTH ANNA POWER STATION." UNIT NO. 2-AMENDMENT TO FACILITY OPERATING LICENSE-Amendment No.122 ;

License No'. NPF-7 l

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

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'A.

The application fo: emendment by Virginia Electric and Power Company, L

Let al., dated August (the licensee) dated Janaury 15, 1990, as supplement

-letter 15, 1990,. complies with the standards and require-;

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i ments of-the Atomic Energy Act of 1954, as ' amended (the Act), and -the-L Comission's. rules'and regulations set. forth in 10 CFR Chapter I; L

B.-

The facility will operate in conformity with'the application, the' provisions of the Act, and the rules and regdations of

.I theLComission; C.

Thereisreasonableassurance-(1)that'the~activitiesauthorized

.by this. amendment can be conducted without endangering the health jf;'

.and ssfety of_ the public, and-(ii).that such activities will be.

conducted in compliance'with the Comission's' regulations;y 4

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 T E.

'he issuance of this amendment is in ac'cordance with 10 CFR Part-o

!51 of the Comission's regulations and all applicable' requirements

~have~been satisfied.

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Accordinglp,dhe'licenseis.amendedbychanges"tothe'TechnicalSpeci-

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'fications.as indicated in the attachment to this; license amendment,.

t and. paragraph 2.C.(2)- of Facility Operating License No. NPF-7 is hereby _

m amended to read as-follows:

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i (2)'TechnicalSpecifications

'The Technical' Specifications contained in Appendices At and B,._

as revised-through Amendment No.

122, are hereby incorpora'ted in the license.- The licensee shall operate the facility;in g

accordance with the Technical-Specifications, e

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3. <:This. license amendment is effective as of the date'of. issuance and shall.

' be impicmented~within.30 days.-

y FOR THE NUCLEAR REGULATORY' COMMISSION.

- Herbert N.-Berkow, Director Project Directorate'II-2

. Division of Reactor Projects - I/II 2

' Office of Nuclear Reactor Regulation Attschment:

. Changes to the Technical

.: Specifications JDate of Issuance:- September 6.1990 1

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ATTACHMENT TO LICENSE AMENDMENT NO. 122

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TO FACILITY OPERATING LICENSE N0. NPF-7 DOCKET N0. 50-339-

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

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

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

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

Page-3/4 1-21 t

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-i REACTIVITYCONTROLJETjg AfDOR0PTIME i

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LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) rod drop time from

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- the fully withdrawn position shall be less then.or equal to 2 7. seconds from

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beginning of decay of stationary gripper coil voltage to deshpot entry with:

T,yg gre4ter than or equal to 500'F, and a.

Allreaciorcoolantpumpsoperating.

b.

APPLICA8ILITY: HODG$ 1.and 2.

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ACTION:

1 With the' drop time of any full length rod determinen to exceed a.-

the above limit, restore the rod drop time to within the above limit prior,to proceeding to MODE 1 or 2.

b.

With the rod drop times within limits but determined with 2 reactor-coolant pumps operating, operation may proceed provided THERMAL-POWER is restricted to:

1.-

Less than or equal to 66% of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are open, nr 2.

Loss' than or equal to 71% of RATED THERMAL POWER when the 'nector

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coslant stop valves in the nonoperating loop are closed.

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l-SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full longth rods shall be demonstrated through measurement prior to reactor criticality:

For all rods following each removal of the reactor vessel head, a.

b.

For specifically affected 'ndividual rods following any main-tenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and c.

At least cnce per 18 months.

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L NORTH ANNA - UNIT 2 3/4 1-21 Amendment No.122

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t REACTMTYCONTROL5YSTEMS S SHUTDOWN RODINSEfmON UMff UMrTING CONDITION FOR OPERAN 3.1.3.5 All shutdown rods shall be fulY withdrawn.

l-APPLICARILITY: MODES 1' and 2's.

ACIX2d:

With a maulmum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hout either:

a.

Fully withdraw the rod. or l

b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVE!LIANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be 4termined to be fully withdrawn:

Within 15 minutes prior to inillal control rod bank withdrawal during an l

a.

approach to reactor criticality, and b.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

s See Special Test Exceptions 3.10.2 and 3.10.3.

1. With Kett at 1.0 i

NORTH ANNA. UNIT 2 3/4 1 22 Amendment No. 121, i

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• 3/4.4 REACTOR COOLANT SYSTEM M

i BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed _ to operate with all reactor coolant loops in operation 1

and maintain the DNBR above the design limit during all normal operations and

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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 />.

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

removing decay heat; however, single f ailure considerations require that two loops be OPERABLE.

In MODES 4 and 5, a single reactor coolant loop or RHR loop proWes sufficient heat two loops be OPERABLE. Thus,if the reactor ecolant loops are not OPEf TABLE, this

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removal capability ter removing decay heat'but single failure considerations. require that # >last 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 tne decay i

heat power production rate decreases with time after reactor shutdown, the requirements for RHR system decay heat removal also decrease. Adequate decay heat tornovat is 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 /> attar entry into MODE 3 and RHR flow is sutticient to maintain the RCS temperature less than or equal to 140'F. The reduced flow rate-i prov! des adstional margin to vortexing at the RHR pump suction while in Mid t.oop Operation.

'Dunng a reduction in reactor coolant system beton concentration the Specification 3.1.1.3.1 requirement to maintain a 3000 gpm flow rate provides sufficient coolant circulation to ninimize 5

the effect of a boron d;tution 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 340r F are provided to prevent RCS pressure transients, caused by energy additions f rom 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) restncting the water volume in the pre.surizer end thereby providing a volume for the primary coolant to expand into or (2) by restricting starting from the RCPs to when the secondary water temperature of each steam generatoris less than 50'F above each of the RCS cold leg temperatures.

L The requirement to maintain the boron concentration of an isolated Ir. :p 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 j

i_

i 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 minute;t prior to opening its cold leg stop valve ensures adequate mixing of the coolant in j

'this loop and prevents any reactivity ettec!$ due to boron concentration stratification.

Start"o of anidle loop willinject coolwater from the loop into the cort The reactivity transient resuting from this cool water injection is minimized by delaying isolated loop startup until its temperature is within 20'F of the operating loops. Ma~ng the reactor suberitical 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 l

'd REACTOR C00LMT SYSTEM EASES 3/4.4.2 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 relieve 380,000 lbs per hour of saturated steam at the valve set L : r 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 OPERABLE, 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 aust be OPERA 8LE 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-aus surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective Systes 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 reitef 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.

The power operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to end including the des'.D step lead dscrease with steam dump. Operation of the P0RVs minimizes the undesirable rpening of the spring-loaded pressurizer code safety valves.

Each PORV has a remotely operated block valve to provide a positive shutoff

/

capability shoi.ld a relief valve become inoperable.

3/4.4.4 PRE 55URIZER 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 initial $AR assumptions.

The 12-hour p'eriodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation.

The maximus water volume alsn ensures that a steam bubble is formed and thus the RCS is not a hydraulicaliy solid systes.

The requirement that a einimum number of pressurizer heaters be OPERABLE ensures that the plant will be able to establish natural circulation.

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

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