Text

Proposed Tech Spec Changes Incorporating TMI-2 Lessons Learned Category A,Revised Maint Requirements for Min Water Levels Above Fuel Assemblies.Provides for Redundancy in Decay Heat Removal Capabilities
	ML19338G077
Person / Time
Site:	North Anna
Issue date:	10/15/1980
From:	; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML19338G072	List: ML19338G071; ML19338G077;
References
	NUDOCS 8010280415
	Download: ML19338G077 (36)
	v • d • e

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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WY ATTACHMENT 1 PRCPOSED TECHNICAL SPECIFICATION CHANGE NORTH ANNA UNIT NO. 1 B 010280q\\

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TABLE 3.3-4 (Continued) 3;g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRIMENTATION TRIP SETPOINTS e

FUNCTIONAL UNIT TRIP SETPolNT Att0WABLE VALUES 2.

CONTAINMENT SPRAY a.

Manual Initiation Not Appilcable Not Applicable b.

Automatic Actuation Logic Not Applicable Not Applicable c.

Containment Pressure--High-High i g psia i g psta 27.75 29.25 3.

U NTAINMENT ISOLATION Phase "A" Isolation b

Manual Not Applicable Not Appilcable f.

From Safety injection Not AppItcable Not Appitcable Automatic Actuation logic b.

Phase "B" Isolation 1.

Manual Not AppItcable Not Appilcable 2.

Automatic Actuation logic Not AppItcable Not Applicable 1 / psia 1)(ps1a lg I

3.

Containment Pressure--High-High 27.75 2't. 25

,_ ;3 8

6 i

5 g

TABL53.3-4(Continued) g ENGINEEREO SAFETY FEATURE ACTUATION SYSTEM INSTRIMENTATION TRIP SETPo*J!TS l

5 FUNCTIONAL UNE TRIP SETFolig ALLOWABLE VALUES a

6.

AUXILIARY FEEDWATER PUNP START

c. g.

Steam Generator Water

> 51 of narrow range

> 41 of narrow range Level Low-Low Tnstrument span each Tnstrument span each steam generator steam generator d F.

S.I.

See 1 above (All S.I. Setpoints)

e. r.

Station Blackout

> 57.51 Transfer Dus Voltage > 52.51 Transfer Bus Voltage F. #

Trip of Main Feed Pump I?. A.

H.A.

R 7.

LOSS OF POW R n

a.

4.16 kw Emergency Bus Undervoltage 2999 1 60 volts with a 2912 1 60 volts with a w

g (Loss of Voltage) 2.2 1 0.03 second time delay 3 1 0.03 second time delay h.

4.16 kv Emergency Dus tindervoltage 3744 + 1.4 volts with a 3619 + 1.4 volts with a (Degraded Voltage) 60

I second time delay 75 1 I second time delay i I O

- a.

Monua\\

14.A.

N. A.

j (

n I

~f' p

- b.

haheMchcb & n Lyc N. A.

N.A.

g l4 i

4

n 3e,,

l l

TABLE 3.3-5 (Continued) 1 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

(

6.

Steam Flow in Two Steam Lines-High Coincident with Steam Line Pressure-Low a.

Safety Injection (ECCS)

S 13.04/23.0#

b.

Reactor Trip (from SI) 5 3.0 c.

Feedwater Isolation

< 8.0 d.

Cont'ainment Isolation-Phase "A" 1 18.0#/28.04 e.

Auxiliary Feedwater Pumps 6

1 0.0 f.

Essential Service Water System Not Applicable g.

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< 8.0 7.

Containment Pressure--High-High a.

Containment Quench Spray 1 60.0 b.

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Containment Pressure-Intermediate High-High a.

Steam Line Isolation

_ 7.0 9.

Steam Generator Water Level Low-Low a.

Auxiliary Feedwater Pumps

_ 60.0

10. Station Blackout

( Co.o a.

Auxiliary Feedwater Pumps M t ^;;?f::57:

11. Main Feedwater Pumo Trip g go,o a.

Auxiliary Feedwater Pumps "0t '-?? iC tI l

12. Steam Generator Water Level--High High a.

Turbine Trip - Reactor Trip

_ 2.5 b.

Feedwater Isolation

_ 11.0 NORTH ANNA - UNIT 1 3/4 3-29

l

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TAllLE 4.3-2 (Continuedl E!

ENGlHEERED SAFETY FEATllRE ACTllATION SYSlfM INSTRUMENTATION wg SUTtWTITAlitf~HTT)TlII~fWNT$

CllANNEL MODES IN WillCH CllANNEL CllANNEL FUNCTIONAL SURVEILLANCE c}

fuNCI10NAL UNIT CilECK CALIllitAT 10N TEST REQtilRED S.

TURRINE 1 RIP AND fEEI) WATER

~

ISOLATION a.

Steam Generator Water S

R H

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AUXILI ARY FEEDWATEll PUNPS w

c. p.

Steam Generator Water S

R H

1, 2, 3

'f Level--Low-Low U

d..E.

S.I.

See I above tall S.I. Surveillance Requirements) e, 4.

Station Blacbut N.A.

R N.A.

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H.A.

R 1, 2 7.

t055 Of POWER 4.16 KV Emergency Dus a.

l.oss of Voltage N.A.

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h.

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[

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TABLE 3.3-10 ACCIDENT HONITORING INSTRUMENTATION g-TOTAL NO.

MINIMUH ge OF CHANNELS CHANNELS OPERABLE e

E 1.

Containment Pressure 2

1 4

, 2.

Reactor Coolant Outlet Temperature-That (wide range) 2 1

3.

Reactor Coolant Inlet Temperature-Tcold (wide range) 2 1

4.

Reactor Coolant Pressure-Wide Range i

1 5.

Pressurizer Water Level 1

1 6.

Steam Line Pressure 2/ steam generator 1/ steam Generator 7.

Steam Generator Water Level-Harrow Range 2/ steam generator 1/ steam geneiator o

hk 8.

Refueling Water Storage Tank Water Level 1

1 w

9.

Boric Acid Tank Solution Level 1

1

)

EI 10.

Auxiliary feedwater Flow Rate 1/ steam Generator 1/ steam generator l

11.

Reactor Coolant System Subcooling Mergin Monitor 2

1 12.

PORY Position Indicator 2/ valve 1/ valve 13.

PORV Block Valve Position Indicator 1/ valve 1/ valve 14.

Safety Valve Position Indicator 1/ valve 1/ valve

l A

TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILI ANCE REQUIREMENTS Y

CilANNEL CilANNEL INSTRllHENT CilECK CALIBRATION E

1.

Containment Pressure M

R 4

2.

Reactor Coolant Outlet Temperature-That (wide range)

M R

3.

Reactor Coolant Inlet Temperature-Tcold (W e ranDe)

M R

4.

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R 5.

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R 6.

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R 7.

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R sa f

8.

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R 9.

Boric Acid Tank Solution Level M

R 10.

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R 11.

Reactor Coolant System Subcooling Margin Monitor M

R 12.

PORV Position Indicator M

R 13.

PORV Block Valve Position Indicator H

R 14.

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R i

l I

l 3/4.4 REACTOR {V MT S.YSTE4 m

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1 j

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SURVEILLANCE REQUIREMENTS The.scLbog rep. _ red re,s.ckor ccc\\ad. bcps shan he. verMed b 4.4.1.1

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At least once per 31 days, with the reactor coolant locos in operation verify that power is removed from the loop stop valve operators.

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NORTH ANNA. UNIT 1 3/4 4 3

'.. -.. '. ~ - ^. "... '. ~.,

I r

REACTOR COOLANT SYSTEM HOT STAND 8Y ilMITING CONDITION FOR OPERATION 3.4.1.2 a.

At least two of the reactor coolant loops listed beice : hall be OPERABLE:

1.

Reactor Coolant Loop A and its associated steam generator and reactor coolant pump, 2.

Reactor coolant Loop 8 and its associated steam generator and reactor coolant pump, 3.

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

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

APPLICABILITY: MODE 3 ACTION:

With less than the above required reactor coolant loops OPERABLE, a.

restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDCWN within the next 12 nours.

b.

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

SURVEILLANCE RECUIREMENTS 4.4.1.2.1 At least the above required rea.ctor coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 At least one cooling loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"All reactor coolant pumps may be de-energized fcr up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause dilution of the reactor coolant ~ system boren concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

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

i REACTOR COOLANT SYSTEM SHUT 00WN LIMITING CONDITION FOR OPERATION 3.4.1.3 a.

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

1.

Reactor Coctant Loop A and its associated steam generator and reactor coolant pu=p,"

2.

Reactor Coolant Loop 8 and its associated steam generator and reactor c:olant pump,*

3.

Reactor Coolant Locp C and its associatad steam generator and reactor coolant pump,"

4.

Residual Heat Removal Subsystem A,""

5.

Residual Heat Removal Subsystam 3.""

b.

At least one of the above c:clant loops shall ce in operation."""

APOLICABILITY: McCES 4 and 5.

~

ACTION:

I With less than the above required locps GPERABLE, immediately a.

initiate c:rrective action to return the required loops to 07ERAELE status as soon as possible; be in COLD SHUTUO*nN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

3.

With no coolant loop in operation, suspend all operations involving a reduction in baron concentration of tne Reactor Coolant System and immediacaly initiata corrective action ts return the required c:clant loop to operation.

31D*F "A reac:cr coolant pump shall not te startad with one or more of the RCS cold leg temperatures less than er equal to 4G2' unless 1) the pressurizer watar volume is less than 457 cubic feet or 2) tne secondary watar taccera-ture of each steam generator is less than 50*F aceve each of the RCS cold leg tamperatures.

""The offsite or emergency power scurca may be inoperable in MODE 5.

"*"Ali, eact:r c:elant pumps and rasidual heat removal pumps may be de-energized for-up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided 1) no operations are permitted that would cause i

dilution of the reactor c:olant system boren concentration, and 2) core outlet taccerar.ure is maintained at least 10*F talow saturation temperature.

NORTH ANNA - UNIT i 3/4 4-3

l REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required residual heat removal loop (s) shall be determined OPERABLE per Specification 4.0.5.

4.4.1.3.2 The required reactor coolant pump (s), if not in operation, sha,'1 be determined to be OPERABLE 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 OVd?.ABLE 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 one coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

NORTH ANNA - UNIT 1 3/4 4-3a 4

33 7e REACTOR COOLANT SYSTEM ISOLATED LOOP LIMITING CONDITION FOR OPERATION 3.4.I. 4 3.'

'.2 The boron concentration of an isolated loop shall be maintained greater than or equal to the boron concentration of the operating loops, unless the loop has been drained for maintenance.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

With the requirements of the above specification not satisfied, do not open the isolated loop's stop valves; either increase the boron concen-tration of the isolated loop to within the limits within a hours or be in at least HOT STANOBY 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 barated to a SHUTDOWN MARGIN equivalent to at least 1.77 sk/k at 200*F.

SURVEILLANCE REQUIREMENTS 4.4. l. 4

?

The boron concentration of an isolated icop shall be determined to be greater than or equal to the boron concentration of the cperating 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, i

NORTH ANNA - UNIT 1 3/4 4-4

E :: 7

(

REACTOR COOLANT SYSTEM ISOLATED t.00p STARTUp

, LIMITING CONDITION FOR CpERATION 3.4.1.5

~

0. t.1. 0 A reactor coolant loop cold leg stop valve shall remain closed until:

a.

The isolated 1 cop has been operating on a reci culation flow of > 125 gym for at least 90 minutes and the temcerature at the cold leg of the isolated loco is within 20*F of the highest cold leg tamperature of the operating loops, b.

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

ApplICA8II.ITY: ALL MODES.

ACTION:

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

SURVEILLANCE REOUIREMENTS

4. 4. t. 5.1

" ' 1.2.1 The isolated loop cold leg temperature shall be detarmined to be within 20*F of the highest cold leg tamperature of the operating loops within 30 minutas prior to opening the cold leg stop valve.

A.s. t. G.2.

t.'.!.2.2 The reactor shall be 'detannined to be suberitical by at least l

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

l l

l NORT14 ANNA - UNIT 1 3/4 4-5 9

a_

REACTOR COOLANT SYSTEM RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.3.2 Two power relief valves (PORVs) and their associated block valves shall be OPERA 8LE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With one or more PORV(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore a.

I the PORV(s) to OPERA 8LE status or close the associated b1:ck l

valve (s) and remove power from the block valve (s); ot';<rwise, be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.3.2.1 Each PORV shall be demonstrated OPERABLE:

At least once per 31 days by performance of a CHANNEL FUNCTIONAL a.

TEST, excluding valve operation, and b.

At least once per 18 months by performance of a CHANNEL CALIBRATION.

4.4.3.2.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve trough one complete cycle of full travel.

4 i

l NORTH ANNA - UNIT 1 3/4 4-7a 1

=

3 5 30 PLANT SYSTEMS AUXILIARY FEE 0 WATER SYSTEM LIMITING CONDITION'FOR OPERATION 3.7.1.2 At least three independant steam generator auxili ary feedwater pumps and associated flow paths shall se OPERABLE with:

a.

Two motor driven feedwater pumps.

b.

One feedwater pump capable of being powered fror an OPERABLE steam sup11y system.

APPLICABILITY: MOD 051, 2 and 3.

ACTION:

With one auxiljary Peedwater pump inop erable, restore at l east three auxiliary feedwater pumps (two capable of being powered fi ca separate emergency busses an f one capable of be ing powered by an OF ERABLE steam supply system) to 0 )ERABLE status witr in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be ir HOT SHUTCOWN within the next 12 1ours. Specificati on 3.0.4 is not appl icable during heatup into Mode 3 For specification :.7.1.E.b.

13 SURVEILLANCE REQUIREMENTS 4.7.1.2. In additicn to the requirements of Specification 4.0.5, each auxiliary feedwater pump shall be demonstrated OPERASLE:

a.

At least once per 31 days by:

1.

Veri fying that the ste.tm turbino driven pump develops a discharge pressure of p_1380 psig at a fica of > 35 gpm on recirculation flow.

13 2.

Veri fying that each va lve (m nual, power o perated or autcmatic) in the flow path that is not lecked, sealed, 1

or c therwise secured i l position, is in its correct posi tion, and

't V

gr SEE ATT ACED PMEs NORTH AN,';A - UNIT 1 3/4 7-5 ndment "c.13

S E 20

^

PLAllT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) b.

At least once per 18 months during shutdown by:

1.

- Veri "ying that each automatic valve in the flow path actuittes to its correct position on a safeuy injection test signal.

2.

Veri"ying that each auxiliary feedwater pur:p starts auto-matii: ally upon receipt of each of the following :...

a.

Safety injection b.'

Loss of offsite power c.

Low-law level in any one steam genera :or d.

Trip of all main ieed pumps.

c.

Prior to er try into Mode 3 fo llowing Mode 5 performance of a flow test of each auxiliary feedwater pump to "erify the normal fic',, path from the condensate storage tank through 33 the pump tc its associated stham generator is required.

Y Y

V SEE NTTACHED PAGES NORTH ANNA - UNIT 1 3/4 7-6

'70 d 00t "0.

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEN LIMITING CONDITION FUR OPERATION 3.7.1.2 At least three independent st31s generator auxiliary feedwater pumps and associated flow paths shall be OPEk SLE with:

a.

Two motor driven auxiliary feedwater pumps, each capable of being powered from separate emerger::y busses, and b.

One steam turbine driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.

pPLICA8ILITY:

MODES 1, 2 and 3.

ACTION:

a.

With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to a OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN 'within the fo} lowing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With two auxiliary feedwater pumps inoperable be in at least HOT STANDBY i

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERA 8LE status as soon as possible.

SURVEILLANCE REQUIREMENTS

4. 7.1. 2 In addition to the requirements of Specification 4.0.5, each auxiliary feedwater pump shall be demonstrated OPERABLE:

a.

At least once per 31 days by:

1.

Verifying that each motor driven pump develops a discharge pressure of greater than or equal to 1250 psig at a flow of I

greater than or equal to 53 gpm.

2.

Verifying that the steam turbine driven pump develops a dis-charge pressure of greater than or equal to 1380 psig at a flow of greater than or equal to 35 gpm on recirculation flow. The provisions of Specification 4.0.4 are not appli.able.

l l

NORTH ANNA - UNIT 1 3/4 7-5

PLANT SYSTEF2 SURVEILLANCE REQUIREMENTS (Continued) 3.

Verifying that each valve (manual, power or.erated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

b.

At least once per 18 months during shutdown by:

1.

Verifying that each automatic virve in the flow path actuates to its correct position on an auxiliary feedwater actuation test signal.

2.

Verifying that each auxiliary feedwater puer. starts automatically upon receipt of an auxiliary feedwater actuation test signal.

The auxiliary feedwater systsu shall be demonstrated OPERABLE prior c.

to entry into MODE 3 following each COLD SHUTDOWN by performing a flow test to verify the normal flow path from the emergency condensate storage tank through each auxiliary feedwater pump to its associated steam generator.

9 NORTH ANNA - UNIT 1 3/4 7-6

11 25 "

REFUELING OPERATIONS C^^L?" C "C"'.r!O" RESIDUAL WNT REMOVAL AMD COOLANT ttRCutATtoN ALL WNTER LEVELG LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal loop shall be in operation.

1 APPLICABILITY: MODE 6.

ACTION:

a.

With less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations in-volvir.3 an increase in the reactor decay heat load.or a re-duction in boron concentration of the Reactor Coolant Sy?'.em.

Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosp,here within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

The residual heat removal 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 period during the p'erformance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

c.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.8.1 A residual heat removal loop shall be determined to be in opera-tion and circulating reactor coolant at a flow rate of f_ 3000 gpm at least once per /4 hours.

ca., % oc NORTH ANNA - UNIT 1 3/4 9-8

j REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent Residual Heat Removal (RHR) loops shall be OPERABLE.*

l APPLICABILITY: MODE 6 when the water level above the top of the reactor pressure vessel flange is less than 23 feet.

ACTION:

a.

With less than the required RHR loops OPERABLE, i= mediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible, b.

The provisions of Specification 3.0.3 are not applicable.

o SURVEILLANCE REQUIREMENTS 4.9.8.2 The required Residual Heat Removal loops shall be determined OPERABLE per Specification 4.0.5.

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

NORTH ANNA - UNIT i 3/4 9-8a

.l

REFUELING OPERATIONS WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10 At least, 23 feet of water shall be maintained over the top of the reactor pressure vessel flange.

APPLICABILITY: During CORE ALTERATIONS wnile in MODE 6.

ACTION:

With the recuirements of the above specification not satisfied, suspend all CORE ALTERATIONS. The previsions of Specification 3.0.3 are not appitcable.

SURVEILLANCE REQUIREMENTS

.4.9.10 The water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the startup of and at least once per 24 hpurs thereafter during CORE ALTERATIONS.

.. ~..

NORTH ANNA - UNIT i 3/4 9-10 i

e o., n.

l 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS Th 1 anne ia

.r...,e p!:nt it,de igned t epe-=+= with e.is *=e,c'a-co,nlaa*

2-

..........o._,_,m_--.....-...... n.3...

...m..

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.2

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........,.u...

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tt; 1;;;; will ::a;; ; r:::t:r trip if 0;cr:ti ; :b:ve ' 7 uhil: : le:: Of

'l:= i ::: 1;;; will : u;; ; rc;:::r trip if p;r: ting ab:v; ' 3.

. SEE ATTAC.HED PAGE The r::tricti::: :n : tarting o ::ter C ! nt o p 5:!:e " ' 'ith u

Or ::r: RCS ::1d leg: 1:

th:n :r : ::1 t: 220*c :r pr:vided t

.____,_':.,....;u.,..

_ _ _ _ _.. ; ; 4. 4,. _ _. _ <-._._.. u...

___..__. nee r......__

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h:

..,,,__.'..___;m._RCS will be pr::::ted :g:in:t :::rpre:: r: tr:nci: t: and r.

...u...__

n. 3,...._,_.,__.u..

u..

.... i... unwww.

...i.....

.p7....-

.j g

....ng w:t:r v:1;;; in th; pr ;;;ri::r :nd th:r:by previding : v!u:: f:r the primary c;; lent te ex nd int;, cr (2) by r;;tricting :: rting ;f th:

RC?: t: wh n th ;;;;nd:ry w t:r t: p:r:: r: :f :::5 ::::r gen:r:ter !!

!:': thin 50*"

b::: :::5 Of the eCS :!d leg tr peraturer.

l The requirement to maintain the baron 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 baron 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 prior to opening its cold leg stop valve ensures adequate mixing of the coolant in this loop and prevents any reactivity effects due to baron concentration stratifications.

Startup of an idle loop will inject 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 within 20*F of the operating loops. Making the reactor subcritical prior j

to loop startup prevents any power spike which could result from this cool water indur 9d 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 NORTH ANNA - UNIT 1 B 3/4 4-1

endment "

l '

3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in speration' this specification 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 re-moval capability for removing decay heat; however, single failure considera-tions require that two loops be OPERABLE.

In MODES 4 and 5, a single reactor coolant loop or RER loop provides sufficient heat removal cagt.bility 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 RER loops to be OPERABLE.

The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 3200F 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 pressurizer and 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 generator is less than 500F above each of the RCS cold leg temperatures.

d 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 reduc-tions in the Reactor Crolant System. The reaccivity change rate associated with boror reduction will, therefore, be within the capa-bility of operator r;c gnition and control.

i 1

I i

4

)

s,

,e,,

MACTOR COOLANT SYSTEM BASES 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 OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will 8

prevent RCS overpressurization.

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 :11 of these valves is greater than the maximum surge rate resulting > m a complete loss of load assuming no reactor trip until the first Rt... tor Protective System trip set point is reached (i.e., no credit is ti Tn for a direct reactor 1

trip on the loss of load) and also assuming no :.;,eration of the power cperated 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.

SEE ATRCHEb PAGE 3/4.4.4 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the nonr.a1 steady state envelope of operation assumed in the SAR. The limit is consistent uith the 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 foli:. 'Jg expected transient operation. The maximum water volume also ensures that a steam buble is fonned and thus the RCS is not a hydraulically solid system.

The m er Ope n ted :!!:f v:F::: and :tes.- bubb!: 'un c tf en -to-

)

=11:ve ".CS p=::ur; dur' ; :!' d::f;n tr:n:ier:: u; t e d ' :!ud' ;

th: d::f;n :t:p 10:d d::rr.:.:: with :t::- 6teme. C;;r:ti:n Of th: p:w:-

perated

!f f V F;:: ef '-f::: 2: und::trd!: Op:n' ; Of 2: :pr* g

? d d ; :::u-f::r ::d ::f:ty V:!v::.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam

enerator tubes is based on a modification of Regulatory Guide 1.83, NORTH ANNA - UNIT 1 8 3/4 4-2 l

i t-I 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 undersirable 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.

1 l

f J

l' s-l

P t

l REFUELING OPERATIONS 8ASES 3/4.9.6 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 assemelies, 2) each crane has sufficient load capacity to lift a control rod or fuel assembly, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL _ PIT The restriction on movement of loads in excess of the nominal weight 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 fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident.

3/4.9.8 RESIOUAL HEAT 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 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 baron 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 capability. 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.

3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated uoan 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.

NORili ANNA - UNIT i B 3/4 9-2

o TABLE 6.2 _1 MINIMUM SHIFT CREV CCMPOSITION I

WITN UNIT 1 IN MODE 5 OR 6 OR DE-FUELEO POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODES 1, 2. 3, & 4 MODES 5 & 6 8

SS 1

l' SRO 1

none RO 2

1, D

A0 2

2 i

STA 1

none 1

{

WITH UNIT 1 IN MODES 1, 2, 3, OR 4 POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODES 1, 2, 3, & 4 l MODES 5&6 SS l'

18 8

SRO 1

none b

RO 2

y D

A0 2

1 8

STA 1

none af Individual may fill the same position on Unit 1 bf One of the two required individuals may fill the same position on Unit 1.

NORTH ANNA - UNIT 1 6-4 i

l l

TABLE 6.2-1 (Continued)

SS - Shift Supertisor with a Senior Reactor Operators License on Unit 2.

SRO - Individual with a Senior Recctor Operators License on Unit 2.

l RO - Individual with a Reactor Operators License on Unit 2.

AO - Auxiliary Operator STA - Shift Technical Advisor Except for the Shift Supervisor, the Shift Crew Composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the Shift Crew Composition to within the minimum requirements of Table 6.2-1.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

During any absence of the Shift Supervisor from the Control Room while the unit is in MODE 1, 2, 3 or 4, an individual (other tnan the Shift Technical Advisor) with a valid SRO license shall be designated to assume the Control Room command function. During any absence of the Shift Supervisor from the Control Room while the unit is in HODE 5 or 6, an individual with a valid R0 license (other than the Shift Technical Advisor) shall be designated to assume the Control Room command function.

I Licensed operators shall:*

1.

Not work more than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months straight, 2.

Not work more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, 3.

Not work more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7-day period, 4.

Not work more than 14 consecutive days without having 2 consecutive days off.

" Deviation from these requirements may be authorized by the Station Manager in accordance with established procedures and with documentation of the cause.

Overtime limits do not include shift turnover time.

1 NORTH ANNA - UNIT 1 6-5

~

5 17 30

\\

ADMINISTRATIVE CONTROLS 5.2 "iC!L!'" ST""" OL'*L!"!C??!OME 6.3.1 Each mester of the facility staff shall meet or exceed the minimum qualifications <lf ANSI N18.1-1971 for comparable positi ons, except for the Supervisor-health Physics who shall meet or exceed the qualifica-tions of Regula1 ory Guide 1.8. September 1975.

6.4 TRAINING 4

6.4.1 The Stat-on Manager is respansible for ensuring that retraining and replacement training programs

'or the facility staf f are maintained and that such pr ograms meet or exc red the requirements and recommendations

))

of Section 5.5 c f ANSI N18.1-1971,ind Appendix "A" of Il CFR part 55.

6.5 REVIEW AND AUDIT 6.5.1 STATION PUCLEAR SAFETY AND OPERATING COMMITTEE ( INSOC)

FUNCTION 6.5.1.1 The SN! OC shall function :o advise the Station Manager on all matters related to nuclear safety.

COMPOSITION 6.5.1.2 The SN! 0C shall be conposed of the:

Chainnan:

S ta t-on Manager Vice-Chainuan:

Superintendent - Operations Member:

Superintendent - Maintenance Member:

Superintendent - Technical Services ALTERNATES 6.5.1.3 All al lernate members shall be appointed in writing by the SNSOC Chairman no serve on a temporary basis; however, no more than one alternate shall participate as a vc ting member in SNSOC activities at any one time.

'f V

sf

)

5,EE ATTACHED PAGE l

l NORTH ANNA - UNIT 1 6-5

.;;nd:: t "c i

l

i l

ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions and the supplemental requirements specified in the March 28, 1980 NRC letter to all !*rr rer, except for the Supervisor-Health Physics who shall meet or exceed the[ qualifications of Regulatory Guide 1.8, September 1975.

hceng,,g 6.4 TRAINING 6.4.1 The Station Manager is responsible for ensuring that retraining and replacement training programs for the facility staff are maintained and that such programs meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevent industry operacional experience identified by the M SES.

6.5 REVIEW AND AUDIT

5. 5.1 STATION NUCLEAR SAFETY AND OPERATING CCMMITTEE (SNSOC) g FUNCTION 6.5.1.1 The SNSOC shall function to advise the Station Manager on all matters related to nuclear safety.

COMPOSITION 6.5.1.2 The SNSOC shall be composed of the:

- Anistant Staban Man

,f Chairman:

Station Manager Vice-Chairman:

Sr;: '-te-de-t - Ope :t r :

Member:

Superintendent - Maintenance Member:

Superintendent - Technical Services Member :

Sop Miendenk - oper.%n, ALTERNATES

6. 5.1. 3 All alternate members shall be appointed in writing by the SNSOC Chairman to serve on a temporary ba.is; however, no more than one alternates shall participate as a voting member in SNSOC activities at any one time.

i

' NORTH ANNA - UNIT 1 6-5 1

\\

9? ?? '?

t 8

AMINISTRATIVE CONTROLS 6.8 PROCEDURES ( PROGRWS 6.8.1 Written procedures shall be established, implemented and main-tained covering the activities referenced below:

a.

The applicable procedures recommended in Appendix "A" of Regulatory' Guide 1.33. Revision 2. February 1978.

b.

Refueling operations, c.

Surveillance and test activities of safety related equipment.

d.

Security Plan implementation.

e.

Emergency Plan impimentation, f.

Fire Protection Program Implementation.

6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed by the SNSCC and approved by the Station Manager prior to implementation and reviewed periodically as set forth in administrative procedures.

6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

a.

The intent of the original procedure is not altered.

b.

The change is approved by two mebers of the plant suce-visory staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

c.

The change is documented, reviewed by the SNSOC and accreved by the Station Manager within 14 days of implementation.

(,.6 4 SEE ATTACHED PAGE 6.9 REPORTING REQUIRE *ENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6.9.1 In addition to the applicable reporting requirments of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.

NORTH ANNA - UNIT 1 6-13

?: = c at ";. M

6.8.4 The following programs shall be established, implemented and maintained:

a.

Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside contain-ment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the re-circulation spray, safety injection, chemical end volume control, gas stripper, and hydrogen recombiners. The program shall include the following:

(1) Preventive maintenance and periodic visual inspection reauirements and (ii)

Integrated leak test requirements for each system at refueling cycle intervals or less.

b.

In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

This program shall include the following:

(1) Training of personnel, (ii) Procedures for monitoring, and (iii) Provisions for maintenance of sampling and analysis equipment.

c.

Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation.

This program shall include:

(i)

Identification of a sampling schedule for the critical variables and control points for these variables, (ii)

Identification of the procedures used to measure the values of the critical variables, (iii)

Identification of process sampling points, (iv) Procedures for the recording and management of data, (v) Procedures defining corrective actions for all control point chemistry conditions, (vi) A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action, and (vii) Monitoring of the condensate at the discharge of the condensate pumps for evidenta=of condenser inleakage. When condenser inleakage is confirmed, the leak shall be repaired, plugged or isolated within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months .

i

,

ML19338G077

Text

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