ML20065K142

Proposed Tech Specs,Modifying P/T Operating Limitations During Heatup & Cooldown & LTOP Sys Pressure Setpoints & Enabling Temps for Units 1 & 2 & Revising Lcos,Action Statements & SRs for Provs & Block Valves,Per GL 90-06

ML20065K142
Person / Time
Site:	North Anna
Issue date:	04/15/1994
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20065K100	List: ML20065K098 ML20065K142 ML20065K148
References
GL-90-06, GL-90-6, NUDOCS 9404190195
Download: ML20065K142 (53)

Text

- - - , - - - . , _ - - - - - - _ _ - - - . - - _ _ - _ - - , _ _ _ _ - - - _ _ _ - _ _ . . - - - , . . _ _ , - - _ _ _ _ - - . - - - - - - - - - , . . , _ _ , - . _ _ . - - . _ , - - .

4 ATTACHMENT 2 1

.1 l

PROPOSED TECHNICAL CPECIFICATIONS CHANGES NORTH ANNA UNITS 1 AND 2

.1 VIRGINIA ELECTRIC- AND POWER COMPANY l

9404190195 940415 PDR ADOCK 05000338 P- -PDR

. _ - _ _ _ _ - _ _ _ _ - _ - _ _ _ _ _ _ - - _ _ _ _ _ _ _ . _ _ _ = _ - .-_-___--_____________________--_2

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.2 S AFETY VA LVES - S HUTDOWN ............................ . ........................... 3/4 4-6 3/4,4.3 SAFETY AND RELIEF VALVES-OPERATING l S afety Valves ... . ... .. .. .. ... . .. .. .. .. .. .. .... . . ..... .. . .. ....... ... .. .. .. .. .. .... .. .. .. .. .... .... .. .. ... 3/4 4-7 Relie f Valve s .. . .. . .. . ... .. .. .. .. .. .. .... .. .... . .. .. . . . . .. .. .. .. .. .. .. .. .. .. . . .. . . .. .... . . .... .. .. . ... 3/4 4-7 a 3/4.4.4 PRES S U RIZE R . .. .. .. . . .. . .. ... . .. . .. .... .. . . . ... .. .. .. .. .. . .. .. . . . . .. .. .. .... .. .. .. .. .... 3/4 4- 8 3/4.4.5 STEAM G ENERATOR S . .... .. ... . .... . . .. .. ... .. . . .... .... .. .. .. .. . .. .. .. .. .. ... .... .. .. .... . 3/4 4-9 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leaka ge De tection System s .................. ........ ........................................... 3/4 4- 16 Ope ration al Leakage .. . . .. .. . .. . . .. .. .. .. . ... . .. .. .. .. .. .. .. .. .. .... .. .. .... .. .. . . .... .... .... .. 3/4 4- 17 Primary to Secondary Leakage ........ . ............... ......... ................... ........ 3/4 4-18b Primary to Secondary Leakage Detection Systems .................................. 3/4 4-18d 3/4.4.7 C H E M I STRY. . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . .. . . .. . . . . . . . .. . . . . . . . . . . . . . . .. . . . . . . .. . . . . . . .. . . . . . . .. . 3/4 4 - 19 3/4.4.8 S PECI FIC A CTI VITY .. .. . . .. .. .. .. . .. .. .... . . .. .. . .. . . . . . .. .. . .. . . . . ... ... .. .. ....... ... . . .... .. 3/4 4-22 3/4.4.9 PRESSURF/ TEMPERATURE LIMITS Reactor Coolant Syste m ..... .................... ...... ................................ ..... ...... 3/4 4-26 Pre s s urizer... . . . . .. . . . .. .. .. .. .. . . .... . . .. .. . ... . .. .. . . ...... ...... .. .. .. .. .. . . .. .. .. .. .. .. .. .... .... .. .. 3/4 4- 30 Ixw-Temperature Overpressure Protection .... ........................................ 3/4 4-31 l 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2 & 3 Components ................................................. 3/4 4-33 l 3/4.4.11 REACTOR VES S EL HEAD VENT ........................................................ 3/4 4-36 3/_44 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 A CCU M ULA'IORS ..... .. . . .. . ... .. . . .. .. .. .. .. .. . . . .. .. . . . . .... .... .. .... .. .... .. .. .. .. .. .. .. . . .. .. 3/4 5- 1 .

3/4.5.2 ECCS SUBSYSTEMS - Tavg 2 3 5 0 F ... .. .. .. .... .. .. .. ..... . ... . .. .. .. .. .. . ... . .. .... .. . 3/4 5 -3 3/4.5.2 ECCS SUBSYSTEMS - Tavg < 3 50 F ... .. .. .. .. .. . . .. . . .... .. .... .. .. .. .. . . .. .. .... ..... .. 3/4 5- 6 o

i 1

NORTH ANNA - UNIT 1 y Amendment No. 64, i

F' E

INDEX BASES SECTION PAGE 3/_43 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION.................................. ................. B 3/4 3-1 3/4.3.2 ENGINEERED SAFTEY FEATURE INSTRUMENTATION................ B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION ....... ....... ...... ........... ............... B 3/4 3-1 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLA NT LOOPS ............... ..... .. .................................... B 3/4 4- 1 3/4.4.2 and 3/4.4.3 S AFETY AND RELIEF VALVES ......................... ......... ............ B 3/4.4-2 3/4.4.4 PR ES S U R IZE R .. .. .. .... .. . . .. .. .. . . .. .. . . .. .. .... . . .. .. .. . . . . . .. . .. . . .. .. .. .. .. . .. .. .. . . .. .. . . . . . B 3/4 4-2a 3/4.4.5 STEA M G ENERATORS ...... ... ............ ...... . . ...... .. .... ....... .. .. .... ........... B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ....................................... B 3/4 4-4 3/4.4.7 CHEMI STRY..... .... .... ...... ... .. . . . . . .. . . .. . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . .. .. .. . B 3/4 4 -5 3/4.4.8 S PECIFI C A CTI VITY . . . . .. . ... .. . . .. .. .. .. .. . . . . .. . .. .. . . . . .. .... .. .. .. .. .. . . .. .. . .. . ... . .. B 3/4 4-5 3/4.4.9 PRES SURE / TEMPER ATURE LIMITS .................... ............................ B 3/4 4-6 3/4.4.10 STR UCTU RAL INTEG RITY ............. ........ ............ ......................... ... B 3/4 4- 12 l 3/4.4.11 REACTOR VES SEL HEAD VENT ....... ......... .. ............................ ...... B 3/4 4- 13 1

NORTH ANNA - UNIT 1 XIII Amendment No. 64,

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPER ATING LIMITING CONDITION FOR OPERATION 3.1.2.2 Each of the following boron injection flow paths shall be OPERABLE:

a. The flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System, and

b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System.

APPLICABILITY- MODES 1,2,3 AND 4#,

AC"G21;

a. With the flow path from the boric acid tanks inoperable, restore the inoperable flow path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1.77% Ak/k at 200 F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the flow path to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With the flow path from the afueling water storage tank inoperable, restom the flow path to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the temperature of the heat trac;d portion of the flow path from the boric acid tanks is 2115 F.

1 l

1. Only one boron injection flow path is mquired to be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 235 F. l 1 l

NORTH ANNA - UNIT 1 3/419 Amendment No. Mr68,117,170, 1

_ _ . _ . . . . . _. _ _ _ _ _ ~ . . _ _ _ . . _ _ _

1 l

i REACTIVITY CONTROL SYSTEMS CIIARGING PUMPS - OPER ATING LIMITING CONDITION FOR OPERATION j

3.1.2.4 - At least two charging pumps shall be OPERABLE.

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

i ACTION:

With only one charging pump OPERABLE, restore a second charging pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least IlOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1.77% Ak/k at 200 F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore a second charging pump -

to OPERA ~BLE status within the next 7 days or be in COLD S11UTDOWN within the next 30 i hours. The provisions of Specification 3.0.4 are not applicable for one hour following heatup above 235 F or prior to cooldown below 235 F. .]

SURVEILLANCE REQUIREMENTS 4l.2.4.1 The above required charging pumps shall be demonstrated OPER.ABLE by verifying, that on recirculation flow, each pump develops a discharge pmssure of 2 2410 psig when tested pursuant to Specification 4.0.5.

4.1.2.4.2 All charging pumps, except the above requimd OPERABLE pump, shall be i

demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperatum of one or more of the RCS cold legs is less than or equal to 235 F by verifying that the switches in the Control Room l have been placed in the pull to lock position.

1 a

• A maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 235 F. l NORTli ANNA - UNIT 1

~

3/4 1-12 Amendment No, h-16d4% 470,

,m

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. Reacto,r Coolant Loop A and its associated steam generator and mactorcoolant 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.**

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

APPLICABILITY: MODES 4 and 5 ACTION: ,

I

a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

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 loop to operation.

I l

l I

• A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 235 F unless the secondary water temperatum of each j steam generator is less than 50 F above each of the RCS cold leg temperatums.

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

• *

• All nactor 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 j maintained at least 10 F below saturation temperature.

NORTH ANNA - UNIT 1 3/44-3 Amendment No. 16,32,117,170, l

REACTOR COOLANT SYSTEM i SAFETY AND RELIEF VALVES-OPERATING SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.4.3.1 All pressurizer code safety valves shall be OPERABLE with a lift setting of l 2485 PSIG i 1%.*

APPLICABILITY: MODES 1,2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.3.1 No additional Surveillance Requirements other than those required by Specification l 4.0.5.

• The lift setting pressure shall correspond to ambient conditions of the valve at nominal -

temperature and pressure.

NORTil ANNA - UNIT 1 3/44-7

REACTOR COOLANT SYSTEM SAFETY AND RELIEF VALVES - OPERATING l RELIEF VALVFJ LIMITING CONDITION FOR OPERATION 3.4.3.2 Both power-operated relief valves (PORVs) and their associated block valves shall be l OPERABLE.

APPLICABILITY: MODES 1,2, and 3.

ACTION:

a. With one or both PORVs inoperable but capable of being manually cycled, within I hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

b. With one PORV inoperable and not capable of being manually cycled, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or capable of being manually cycled, or close its associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With lx)th PORVs inoperable and not capable of being manually cycled, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore at least one PORV to OPERABLE status or capable of being manually cycled, or close its associated block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT S'HUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d. With one or both block valves inoperable, within I hour restore the blxk valve (s) to OPERABLE status or place its associated PORV(s) in manual control. Restore at least one block valve to OPERABLE status within the next hour if both block valves are inoperable; restore the remaining inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

e. The provisions of Specification 3.0.4 are not applicable.

NORTH ANNA - UNIT I 3/4 4-7a Amendment No. 32,

REACTOR COOLANT SYSTEM SAFE 1Y AND RELIEF VALVES-OPERATING i I

RELIEF VALVES I SURVEILLANCE REQUIREMENTS 4.4.3.2.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE:

a. At least once per 31 days by performing a CHANNEL FUNCTIONAL TEST, .

excluding valve operation, and

b. At least once per 18 months by

1. Operating the PORV through one complete cycle of full travel during MODES 3 or 4, and

2. Operating the solenoid air control valves and check valves on the associated accumulators in the PORV control systems through one complete cycle of full travel, and

3. Performing a CHANNEL CALIBRATION of the actuation instrumentation.

4.4.3.2.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION b or c in Specification 3.4.3.2.

NORTH ANNA - UNIT 1 3/4 4-7b 4

m n -

+ .

l. 5 f

i -

Figme 3.4 North Anna Unit 1 Reactor Coolant System Heatup Limitations Material Property Basis Limiting Material: Circumferential Weld Seam L*miting ART at 30.7 EFPY: 1/4-T,162.9 F I 3/4-T,139.9 Heatup Rates (F/hr) - '

20 40 60 2500.00 -

r

.g ,

Leak Test Limit

/ /// .

j j/j 2000.00 l --

lll a

Il 3 B k

Unacceptable Operation O .f

// ,

2 1500.00 g/

a >

$ l l

$ ll g //

.U // >

g 1000.00 ' p.t g Acceptable e d

/, Operation ,

5 Af /

Heatup Rates (F/hr) - g

/ k/

p' 500.00 20 __- r,3 -

40 60 0.00 .., 1 0 50 100 150 200 250 300 350. H Cold Leg Temperature (Deg. F) ,

1

. North Anna Unit 1 Reactor Coolant System Hestup Limitations (Heatup Rates up to i 60 F/hr) Apphcable for the First 30.7 EFPY (Without Margins for Instrumentation Errors) : -

-i

.i NORTH ANNA - UNIT 1 3/4 4-27 Amendment No.16,7d, !!?,170,

,, -, , , , -. . . . . , ~ . . , . --- ,., - , ,

'i h

Figure 3.4 North" Anna Unit 1 Reactor Coolant System Cooldown Limitations.

Material Property Basis . __

Limiting Material: Circumferential Weld Seam Limiting ART at 30.7 EFPY: 1/4-T,162.9 F 3/4-T,139.9 F i

2500.00 r

/

i

/

/

2000.00

/

/

3 / -

a i n.

~

2 1500.00 ,

/

Unacceptable Operation! /

] / -

m a,

2 /

-- /

~

'C 1000.00 -- - -

f S I '

$ df ,

$ g 7 l Acceptable Operation -

Cooldown Rates y Z/ .f

{F/hr) > M /

SCh$ Y / '

0 -

500,00 20 . , ,__, , ,

_, _. 40  % " J 60 -

100 0.00 l 50 '150 350-O 100 200 '250- 300~

(

Cold L'eg Temperature (Deg. F)

• North Anna Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to

,- 100 F/hr) Applicable for the First 30.7 EFPY (Without Mergins for instrumentation Errors) 'I

-NORTH ANNA - UNIT 1 3/4 4 Amendment No.16,7A,117,170, _

REACTOR COOLANT SYSTEM LOW-TEMPERATURE OVERPRESSURE PROTECTION l LIMITING CONDITION FOR OPERATION 3.4.9.3 Two power-operated relief valves (PORVs) shall be OPERABLE with lift settings of (1) less than or equal to 500 psig whenever any RCS cold leg temperature is less than or equal to 235 F, and (2) less than or equal to 395 psig whenever any RCS cold leg temperatum is less than 150 F.

APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 235 F, MODE 5, and MODE 6 when the head is on the reactor vessel and the RCS is not vented through a 2.07 square inch or larger vent.

ACTION:

a. With one PORV inoperable in MODE 4, restore the inoperable PORV to OPERABLE status within 7 days or depmssurize and vent the RCS through at least a 2.07 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b. With one PORV inoperable in MODES 5 or 6, either (1) mstom the inoperable PORV to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or (2) complete depressurization and venting of the RCS through at least a 2.07 square inch vent within a total of 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />.

c. With both PORVs inoperable, complete depressurization and venting of the RCS through at least a 2.07 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

d. With the RCS vented per ACTIONS a, b, or c, verify the vent pathway at least once per 31 days when the pathway is provided by a valve (s) that is locked, sealed, or otherwise secured in the open position; otherwise, verify the vent pathway ever 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

e. In the event either the PORVs or the RCS vent (s) are used to mitigate an RCS q pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall  ;

describe the circumstances initiating the transient, the effect of the PORVs or RCS vent (s) on the transient, and any corrective action necessary to prevent recurmnce. 1 1

f. The provisions of Specification 3.0.4 are not applicable. j u

1 NORTH ANNA - UNIT 1 3/4 4-31 Amendment No. 46,74, ' !?,170,

l 1

l REACTOR COOLANT SYSTEM LOW-TEMPERATURE OVERPRESSURE PROTECTION l SURVEILLANCE REQUIREMENTS l

4.4.9.3 Each PORV shall be demonstrated OPERABLE by: l

a. Perfonnance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE.

l Performance of a CHANNEL CALIBRATION on the PORV actuation channel, at b.

least once per 18 months.

c. Verifying the PORV keyswitch is in the Auto position and the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpressure protection,

d. Testing pursuant to Specification 4.0.5.

i l

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l NORTH ANNA - UNIT 1 3/4 4-32 Amendment No. M,-34,

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS -Tave > 350 F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE centrifugal charging pump,

b. One OPERABLE low head safety injection pump,

c. An OPERABLE flow path capable of transferring fluid to the Reactor Coolant System when taking suction from the refueling water storage tank on a safety injection signal or from the containment sump when suction is transferred during the recirculation phase of operation or from the discharge of the outside recirculation spray pump.

APPLICABILITY: MODES 1,2 and 3.

ACTION:

a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l.

b. In the event the ECCS is actuated and injects waterinto the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the -

actuation and the total accumulated actuation cycles e .iate.

c. The pmvisions of Specification 3.0.4 are not applicable to 3.5.2.a ;.nd 3.5.2.b for one hour following heatup above 235 F or prior to cooldown below 235 F. l NORTH ANNA - UNIT 1 3/45-3 Amendment No. 3,16,117,153r 470,

I l

EMERGENCY CORE COOLING SYSTENG ECCS SUBSYSTEMS-Tavg < 350 F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE centrifugal charging pump # ,

b. One OPERABLE low head safety injection pump", and

c. An OPERABLE flow path capable of automatically transferring fluid to the reactor coolant system when taking suction from the refueling water storage tank or from the containment sump when the suction is transferred during the recirculation phase of operation or from the discharge of the outside recirculation spray pump.

APPLICABILITY: MODE 4.

ACTION:

a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHlTrDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

b. With no ECCS subsystem OPERABLE because of the inoperability of the low head safety injection pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T avg less than 350 F by use of alternate heat removal methods.

c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

1. A maximum of one centrifugal charging pump and one low head safety injection pump shall be OPERABLE whenever the temperatun: of one or more of the RCS cold legs is less than or equal to 235 F. l NORTH ANNA - UNIT 1 3/4.5-6 Amendment No. Gr14, P A,117, 4-70,

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.

4.5.3.2 All charging pumps and safety injection pumps, except the above required OPERABLE '

pumps, shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperature of one or more of the RCS cold legs is less than or equal to 235 F by verifying that the switches in l the Control Room are in the pull to lock position.

NORTH ANNA - UNIT 1 3/4 5-6a Amendment No. 46,117,170,

EEACTIVITY CONTROL SYSTEMS BASES 3/4.L2 BORATION S'YSTEMS (Continued)

With the RCS average temperature above 200 F, a minimum of two separate and redundant boron injection systems are provided to ensum single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The boration capability of either system is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.77% Ak/k after xenon decay and cooldown to 200 F. This expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 6,000 gallons of 12,950 ppm borated water from the boric acid storage tanks or 54,200 gallons of 2300 ppm borated water from the mfueling water storage tank.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 235 F provides assurance that a mass addition pressure transient can be l relieved by the operation of a single PORV.

With the RCS temperature below 200 F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the mactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

The boron capability required below 200 F is sufficient to provide a SHUTDOWN MARGIN of 1.77% Ak/k after zenon decay and cooldown from 200 F to 140 F. This condition requires either 1378 gallons of 12,950 ppm borated water from the boric acid storage tanks or 3400 gallons of 2300 ppm borated water from the refueling water storage tank.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. The OPERABILITY of one boron injection system during REFUELING insures that this system is available for reactivity control while in MODE 6.

NORTH ANNA - UNIT 1 B3/41-3 Amendment No. 5,16,68,93, 44-7r470,

3/14 REACTOR COOL 1NT SYSTEM l BASES 3/4.4.I REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation and maintain the DNBR above the design limit during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be 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 <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, a minimum RHR system flow rate of 2000 gpm in MODE 5 is permitted, provided there is sufficient decay heat removal to maintain the RCS temperature less than or equal to 140 F. Since the decay heat power production rate decreases with time after reactor shutdown, the requirements for RHR -

system decay heat removal also decrease. Adequate decay heat removal 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 /> 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.

During a reduction in reactor coolant system boron concentration the Specification 3.1.1.3.1 requirement to maintain 3000 gpm flow rate provides s'ufficient 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 235 F are provida.i to prevent RCS pressure transients, caused by energy additions l from the secondary system wSch 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 restricting starting of the RCPs to when the secondary water temperature of each steam l generator is less than 50 F above each of the RCS cold leg temperatums.

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. Verification of the boron concentration in an  ;

idle loop prior to opening the cold leg stop valve provides a massurance 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 boron concentration stratification.

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 j untilits temperature is I NORTH ANNA - UNIT 1 B3/44-1 Amendment No. 16,32,1],13{,

_ , . ~ ,

i 3/.L4 REACTOR COOLANT SYSTEM BASES within 20 F of the operating loops. Making the reactor subcritical prior to loop startup prevents any power spike which could result from this cool water induced reactivity transient.

3/4.4.2 AND 3/4.4.3 SAFETY AND RELIEF VALVES l 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 point. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during hot shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, or the power operated relief valves (PORVs) will provide 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 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 ofload assuming no reactor trip until the first Reactor Protective System trip setpoint is n: ached (i.e., no l '

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 during shutdown and will ]

be perfonned in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel 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 load decrease with steam dump. Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded I pressurizer code safety valves. Each PORV has a remotely operated bk)ck valve to provide a positive shutoff capability should a relief valve become inoperable.

The OPERABILITY of the PORVs and block valves is detennined on the basis of their being capable of performing the following functions:

a) Manual control of PORVs to control reactor coolant system pressure. This is a function that may be used to mitigate cenain accidents and for plant shutdown.

b) Maintaining the integrity of the reactor coolant pressure boundary. This function is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.

NORTil ANNA - UNIT 1 B 3/4 4-2 Amendment No. 32r441,

l.

I l

l

!. 3/14 REACTOR COOLANT SYSTEM L

BASES i

l- c) Manual control of the block valve to (1) unblock an isolated PORV to allow it to be j used for manual control of reactor coolant system pressure (Item a, above), and (2) iso-late a PORV with excessive seat leakage (Item b, above). l 1

d) Automatic control of PORVs to control reactor coolant system pressure. This function l L reduces challenges to the code safety valves for overpressurization events.  ;

e) Manual control of a block valve to isolate a stuck-open PORV.

Surveillance Requirements provide the assurance that the PORVs and block valves can perform their functions. Specification 4.4.3.2.1 addresses the PORVs and Specification 4.4.3.2.2 addmsses the block valves. The block valves are exempt from the surveillance requirements to cycle the valves when they have been closed to comply with the ACTION requirements. This ,

precludes the need to cycle the valves with full system differential pressure or when maintenance -)

is being performed to restore an inoperable PORV to operable status.

( Surveillance Req &ement 4.4.3.2.1.b.2 provides for the testing of the mechanical and electrical aspects of control systems for the PORVs.

Testing of PORVs in HOT STANDBY or HOT SHUTDOWN is required in ortler to I simulate the temperature and pressure environmental effects on PORVs. Testing at COLD SH'UTDOWN is not considered to be a representative test for assessing PORV performance under normal plant operating conditions.

3/4.4A PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance is sufficient to l ensure that the parameter is restored to within its limit following expected transient operation. The l maximum water volume also ensures that a steam bubble is fonned and thus the RCS is not a hydraulically solid system.

NORTH ANNA - UNIT 1 B 3/4 4-2a 1-

REACTOR COOLANT SYSTEM l BASES The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 ~pCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figum 3.4-1, accommodates possible iodine spiking l phenomenon which may occur following changes in THERMAL POWER.

Reducing Tavg to < 500 F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressum of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency ofisotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System l

All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 5.2 of the UFSAR. During startup and shutdown, the rates of temperature and pressum changes are limited so that the maximum specified heatup and cooldown rates am consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) n: presents a lower bound of all similar curves for finite heatup rates when the inner wall of the vesselis treated as the governing location.

The heatup analysis also covers the determination of pressure- temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stmsses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stesses which are already present. The thermal induced stmsses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be def'med. Consequently, for the cases in which the outer wall of the vessel becomes the stmss controlling location, each heatup rate ofinterest must be analyzed on an individual basis.

NORTH ANNA - UNIT 1 B 3/4 4-6 Amendment No. 96,117,170, ,

REACTOR COOLANT SYSTEM l l

BASES The heatup limit curve, Figure 3.4-2, is a composite curve which was pmpared by i determining the most conservative case, with either the inside or outside wall controlling, for any l heatup rate up to 60 F per hour. The cooldown limit curves of Figure 3.4-3 are composite curves l which were prepared based upon the same type analysis with the exception that the controlling I

location is always the inside wall where the cooldown thermal gradients tend to produce tensile-stresses while producing compressive stresses at the outside wall. The heatup and cooldown curves wem prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 30.7 EFPY. The most recent capsule analysis results are documented in Westinghouse Report WCAP-11777, February 1988. The heatup and cooldown curves are documented in Westinghouse Report WCAP-13831, Rev.1, August 1993.

The reactor vessel materials have been tested to determine their initial RTNDT. Reactor operation and resultant fast neutron (E > 1 Mev) irradiation will cause an increase in the RTNDT-An adjusted refemnce temperature, based upon the fluence and copper content of the materialin l question, can be predicted using US NRC Regulatory Guide 1.98, Revision 2. The heatup and cooldown limit curves (Figure 3.4-2 and Figure 3.4-3) include predicted adjustments for this shift in RTNDT at the end of 30.7 EFPY. The reactor vessel beltline region material properties are listed on Figures 3.4-2 and 3.4-3.

The actual shift in the RTNDT of the vessel material will be established periodically by -

removal and evaluation of the reactor vessel material specimens installed on the inside wall of the thermal shield. The surveillance capsule withdrawal schedule was prepared in accordance with the requirements of ASTM E-185 and is presented in the UFSAR. Regulatory Guide 1.99, Revision 2, provides guidance for calculation of the shift in RTNDT using measured data. Dosimetry from the surveillance capsule is used to determine the neutron fluence to which the material specimens were exposed, and to support calculational estimates of the neutron fluence to the reactor vessel.

The pmssure-temperature limit lines shown on Figure 3.4-2 for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. The minimum temperature for criticality specified in T.S. 3.1.1.5 assures compliance with the criticality limits of 10 CFR 50 Appendix G.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in the UFSAR to assure compliance with the l mquirements of Appendix H to 10 CFR Part 50.

Pressurigt l The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements..

NORTH ANNA - UNIT 1 B 3/4 4-7 Amendment No. 117,170,

1 REACTOR COOLANT SYSTEM l I BASES ,

Low-Temocrature Overoressure Protection l The OPERABILITY of two PORVs or an RCS vent opening of greater than 2.07 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 235 F. Either PORV has adequate relieving capability to protect the RCS from l overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generatorless than or equal to 50 F above the RCS cold leg temperatures or (2) the start of a charging pump and its injection into a water-solid RCS. l Automatic or passive low temperature overpressure protection (LTOP) is required whenever any RCS cold leg temperature is less than 235 F. This temperature is the water temperature corresponding to a metal temperatum of at least the limiting RT N ur + 50 F +

instrument uncertainty. Above 235 F administrative controlis adequate protection to ensure the limits of the heatup curve (Figure 3.4-2) and the cooldown curve (Figure 3.4-3) are not violated.

The concept of requiring automatic LTOP at the lower end, and administrative control at the upper end, of the Appendix G curves is further discussed in NRC Generic Letter 88-11.

NORTH ANNA - UNIT 1 B3/44-8 Amendment No. 74,117,170,

EMERGEN'CY CORE COOLING SYSTEMS (ECCS) l BASES ECCS SUBSYSTEMS (Continued)

With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The limitation for a maximum of one centrifugal charging pump and one low head safety -

injection pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and low head safety injection pumps except the required OPERABLE pump to be inoperable below 235 F provides assurance that a mass addition pressure transient can be relieved by the l operation of a single PORV.  ;

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.

3/4.5.4 BORON INJ1 TION SYSTEM The OPERABILITY of the boma injection system as part of the ECCS ensures that sufDeient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS system cooldown. RCS cooldown can be caused by inadvertent j depressurization, a loss-of-coolant accident or a steam line mpture.

The limits on injection tank minimum contained volume and boron concentration ensure that the assumptions used in the steam line break analysis are met.

The OPERABILITY of the redundant heat tracing channels associated with the boron injection system ensure that the solubility of the boron solution will be maintained above the solubility limit of 111 F at 15,750 ppm boron.

NORTH ANNA - UNIT 1 ' B 3/4 5-2 Amendment No.46,68, !!7,170,

ADMINISTRATIVE CONTROLS (Continued) _

SL'CIAL REPORTS l

l 6.9.2 Special reports shall be submitted to the Regional Administrat'or, Region II, within the time l period specified for each report. These reports shall be submitted pursuant to the requirement of the applicable specification:

l

a. Inservice Inspection Reviews, Specification 4.0.5, shall be reported within 90 days of completion.

j b. MODERATOR TEMPERATURE COEFFICIENT. Specification 3.1.1.4.

c. RADIATION MONITORING INSTRUMENTATION. Specification 3.3.3.1, Table 3.3-6, Action 35.

i d. SEISMIC INSTRUMENTATION. Specifications 3.3.3.3 and 4.3.3.3.2.

e. METEOROLOGICAL INSTRUMENTATION. Specification 3.3.3.4.

f. Deleted,

g. LOOSE PARTS MONITORING SYSTEMS. Spec;iication 3.3.3.9.

h. Deleted. 4

i. LOW-TEMPERATURE OVERPRESSURE PROTECTION. Specification 3.4.9.3. l

j. EMERGENCY CORE COOLING SYSTEMS. Specification 3.5.2 and 3.5.3.

k. SETTLEMENT OF CLASS 1 STRUCTURES. Specification 3.7.12.

1. GROUND WATER LEVEL - SERVICE WATER RESERVOIR. Specification 3.7.13.

m. Deleted.

n. RADIOACflVE EFFLUENTS. As required by the ODCM.

o. RADIOLOGICAL ENVIRONMENTAL MONITORING. As required by the ODCM.

p. SEALED SOURCE CONTAMINATION. Specification 4.7.11.1.3.

q REACTOR COOLANT SYSTEM STRUCTURAL INTEGRITY Specification 4.4.10. For any abnonnal degradation of the structural integrity of the reactor vessel or the Reactor Coolant System pressure boundary detected during the perfonnance of Specification 4.4.10, an initial report shall be submitted within 10 days after detection and a detailed report submitted within 90 days after the completion of Specification 4.4.10.

NORTH ANNA - UNIT 1 6-21 Amendment .No. -374&r63r%r 403r4-30,440,

INDEX l l

l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.2 S AFETY VALVES - S HUTDOWN ......... ................................................. 3/4 4-6 3/4.4.3 SAFETY AND RELIEF VALVES - OPERATING l S afe ty Va1ve s . .. .. .. .. . . . . . . .. .. .. . . . . . .. .. .. .. . . . . . . .. .. . . .. .. .. .. .. .... .. .. . . .. . . .. .... .. .... .. .. .. .. .. . 3/4 4 -7 Rel ie f Val ves ... . . . . .. . . .. . . . . . . . . . .. . . .. .. . . .. .. . .. . .. . . .. . . .. .. .. . . . . . . . . .. . . .. .. . . . . . . .. .. . . . . .. . . .. .. 3/4 4-7 a 3/4.4.4 PRES S U RIZE R ... . . . . . . . . . .. . . . ... . . . . . . .. . . . . . . .. . . . .. .. . . .. .. . . .. . .. . . . . . . . . .. .. . . . . . . .... . . .. . . .. 3/4 4- 8 3/4.4.5 STEA M G ENE RATORS .. .. .. . . . . . . .. . . . . . . .. .. . .. .. .. . . . . .. .. . . . . . . .. .. .... .. . . . . .. . . . . .. .. .. . . . 3/4 4-9 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE leakags Je tec tion S ys te m s .. . .. .. . . . . .. .. . . .. . . .. .. .. .. .. . . . ... . . .. .. . .. . . . .. . . .. .. . . . . .. . . .. .. . 3/4 4- 16 Opera tion al Le aka ge .. .. .. . .. .. .. .. . . . . . . ... . . . . . . . . .. . . .. . . .. .. .. .. .. . . . . . .. .. . . .. . .. . . . . . .. . .. 3/4 4- 17 Prim ary to S econdary Le akage ........................... .................................... ... 3/4 4- 18 b Primary to Secondary Leakage Detection Systems ................................... 3/4 4-18d 3/4.4.7 CHEM I STRY . . . . . . . . . .. . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 3 3/4.4.8 S PECIFIC A Crl VITY . .. .. .. .. . .... .. . . .... . .. ...... .. .. .. ........ .... ........ .... ............ 3/4 4-22 3/4.4.9 PRESSUREffEMPERATURE LIMITS Re ac tor Coolan t S y s te m .. .. .. .. . .. ... ...... .. .. . ... .. .. . .. .. .... .. .... ...... ... .. ...... .. .... .. .. 3/4 4-26 Pre s s u ri ze r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . .

low-Temperature Overpressure Protection ............................................. 3/4 4-30 3/4.4.10 STRUCTURALINTEGRITY A SM E Code C1as s 1, 2 & 3 Components ................................................. 3/4 4-32 l 3/4.4.11 REACTOR VES S EL H EAD VENT......................................................... 3/4 4-34 3L431 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 A CCU M U LA'IO R S ... . . . . .. . . . . .. . . .. .. .. . . .... . ... . . .. .. .. . . . . . . .. .. .. . . .. .... .. .. .... . . . ... .. . . .. . . 3/4 5 - 1 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN 350 F.......................... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350 F ................................... 3/4 5 6 3/4.5.4 BORON INJECTION SYSTEM B oron I nj ect i on Tank .. . . .. . .. .. . . .. . . . . .. . . . .. . . . . . . .. ... . . ..... .. .. .. . . .. . ... . . . .. .. .. .. . .. . . . 3/4 5 - 8 Hea t Tracin g ... . . . . .. .. .. .. . ... . . . .. .. .. .. . . . . . . . .. . .. . . .. . . .. .. .. .. . . .. .. .. .. .. .. .. . .... .. .. .. .... 3/4 5 -9 3/4.5.5 REFUELING WATER STORAGE TANK . ... ........... ............................. 3/4 5-10 NORTH ANNA - UNIT 2 V Amendment No. 49,

INDEX I

BASES SECTION PAGE 3/13, INSTRUMENTATION 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESP) INSTRUMENTATION ................................... B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION ......................... ........................ B 3/4 3-1 3/14 REACTOR COOLANT SYSTEM 3/4.4.1 REA CTOR COOLANT LOOPS ............................................................. B 3/4 4- 1 l 3/4.4.2 and 3/4.4.3 S A FETY A ND R ELIEF VALVES .... ........................ .................. B 3/4 4-2 3/4.4.4 PR ES S U R IZE R . . . .. .. . . .. . . .. . . . . . . . . . . . . . . . . . . .. .. . . . . . . .. . . .. .. .. .. . . . . . ..... .. . . . .l .. .. .. .. .. . . .. . B 3/4 3/4.4.5 STE A M G E NE RATO RS . .. . . . . .. . . ... . ... . .. . . . . . . . . . . .. .. . . ... . .. . . . . .. . . .. .. . . . . . . . . . . .. .. . . ... B 3/4 4-3 3/4.4.6 REACTOR COOLANT S YSTEM LE A KAGE ............... ....................... B 3/4 4-4 3/4.4.7 C H EM I S TRY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .

3/4.4.8 S PE CI FIC A Cfl V ITY . . .. .. . . . . . .. .. . . .. .. .. . . .. .. . . .. .. . . . . .. .. .. . . .. .. .. .. .. . .. . .. .. .. .. .. .. .. .. B 3/4 4-5 3/4.4.9 PR ES S URE / TEMPER ATURE LIM 1TS ................................................. B 3/4 4-6 3/4.4.10 STRUCTUR AL INTEG RITY ..... .. ............ .. ... ... .............. .... .... ...... ........ ... B 3/4 4-8 3/4.4.11 REACTOR VES SEL HE AD VENT ......................... .............................. B 3/4 4- 17 u.

1 1

1 r

f i

! NORTII ANNA - UNIT 2 XI Amendment No. 49, i <=

L

(

' REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING .

LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following boron injection flow paths shall be OPERABLE:

a. The flow path from the boric acid tanks via a boric acid transfer pump and a j charging pump to the Reactor Coolant System,

b. Two flow paths from the refueling water storage tank via charging pumps to the Reactor Coolant System.

APPLICABILITY: MODES 1,2,3 and 48. l ACTION: ,

With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1.77% delta k/k at 200 F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD '

SIIUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the temperature of the heat traced portion -

of the flow path from the boric acid tanks is greater than or equal to 115 F when it  ;

is a required water source.

1 i

'# . Only one boron injection flow path is required to be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 270 F. l

' NORTil' ANNA - UNIT 2 3/41-9 Amendment No. 5A,129,1 A9, l

. REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1,2,3 and 4#. l ACTION:

With only one charging pump OPERABLE, restom a second charging pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1.77% delta k/k at 200 F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore a second charging pump to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The provisions of Specification 3.0.4 are not applicable for one hour following heatup above 270 F or prior to cooldown below 270 F. l 1

SURVEILLANCE REQUIREMENTS

+

4.1.2.4.1 The above requimd charging pumps shall be demonstrated OPERABLE by verifying, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to 2410 psig when tested pursuant to Specification 4.0.5.

4.1.2.4.2 All charging pumps, except the above requimd OPERABLE pump, shall be E

demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperatum of one or more of the )

RCS cold legs is less than or equal to 270 F by verifying that the control switch is in the pull to l l kek position.

l l

q i

l l l l

1. A maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 270 F. l l

NORTH ANNA - UNIT 2 3/4 1-12 Amendment No. 449,

REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION ,

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

1 Rcactor Coolant Loop A and its associated steam generator and reactor coolant -

pump, l

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

3. Reactor Coolant Imop C and its associated steam generator and reactor coolant pump,

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

APPLICAHILITY, MODE 3 ACTION:

a. With less than the above required loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 l hours.

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 actions to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor 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 coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• All reactor coolant pumps may be de-energized for up to I hour provided (1) no operations _l am 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.

• • The requirement to have one coolant loop in operation is exempted during the performance l of the boron mixing tests as stipulated in License Condition 2.C(15)(f) and 2.C(20)(b).

NORTH ANNA - UNIT 2 3/44-2 Amendment No.49rl49, J

p REACTOR COOLANT SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the coolant loops listed below shall be OPERABII:

1. Reacto,r Coolant Loop A and its associated steam generator and mactor 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.**

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

APPLICABILITY: MODES 4 and 5 ACTION:

a. With less than the above required loops OPERABLE,immediately initiate corrective action to return the required loops to OPERABLE status as'soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

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 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 270 F unless the secondary water temperatum of each l steam generator is less than 50 F above each of the RCS cold leg temperatums.

• • 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 mactor coolant system boron concentration, and 2) core outlet temperatum is maintained at least 10 F below saturation temperature.

NORTH ANNA - UNIT 2 3/44-3 Amendment No. 449,

REACTOR COOLANT SYSTEM SAFETY AND RELIEF VALVES - OPER ATING l RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.3.2 Both power-operated relief valves (PORVs) and their associated block valves shall be l OPERABLE.

APPLICABILITY: MODES 1,2, and 3.

ACTION:

a. With one or both PORVs inoperable but capable of being manually cycled, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restom the PORV(s) to OPERABLE status or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With one PORV inoperable and not capable of being manually cycled, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or capable of being manually cycled, or close its associated block valve and remove power from the block valve; mstore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT  ;

STA.NDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following J 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With both PORVs inoperable and not capable of being manually cycled, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore at least one PORV to OPERABLE status or capable of being i manually cycled, or close its associated block valve and remove power from the i block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT 'j SIlUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. j

d. With one or both block valves inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the block valve (s) to OPERABLE status or place its associated PORV(s) in manual control. Restore at least one block valve to OPERABLE status within the next hour if both block valves are inoperable; restore the remaining inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

e. The provisions of Specification 3.0.4 are not applicable.

NORTH ANNA - UNIT 2 3/4 4-7a

REACTOR COOLANT SYSTEM SAFETY AND RELIEF VALVES -OPERATING RELIEF VALVES i

__ SURVEILLANCE REQUIREMENTS l

4.4.3.2.1 In addition to the requirements of Specification 4.0.5, each PORV shall be I demonstrated OPERABLE:

a. At least once per 31 days by perfonning a CHANNEL FUNCTIONAL TEST, excluding valve operation, and

b. At least once per 18 months by:

1. Operating the PORV through one complete cycle of full travel during l MODES 3 or 4, and 1

2. Operating the solenoid air control valves and check valves on the associated I accumulators in the PORV control systems through one complete cycle of j full travel, and 1

3. Performing a CHANNEL CALillRATION of the actuation instmmentation.

4.4.3.2.2 Each block valve shall be demonstrated OPERABLE at least once per 9'2 days by I operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION b or e in Specification 3.4.3.2. I NORTH ANNA - UNIT 2 3/4 4-7b

~

p e

Figure 3.4 North Anna Unit 2 Reactor Coolant System Heatup Limitations Material Property Basis Limiting Material: ' Lower Shell Plate Limiting ARTat 17 EFPY: '1/4-T,'196 F 3/4-T,172 F Heatup Rates (F/hr) 20 40 '60 2500.00

! l

c/

- I h Leak Test Limit l f

/

//

2000.00 / bl i

lll

. Al b

r,,

//

r/

S 1500.00 2

j U

//

Unacceptable a>

Operation-a / /

k / /

's / /

g 1000.00 , Acceptable --

E  !/

' Operation --

,; f: / .

A f/

Heatup Rates (F/hr)

,- j

- f e '

/

f 500.00 20 40

__--;f <

60 0.00 0 50 100- -150 200 250 300- 350- 400 Cold Leg Temperature (Deg. F)

North Anna Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates'up to 60 F/hr) Applicable for the First 17 EFPY (Without Margins for Instrumentation Errors)

NORTH ANNA - UNIT 2 3/4 4-27 Amendment No. "J, l'9, "

F

P Figun: 3.4-3.- North Anna Unit 2 Reactor Coolant System Cooldown Limitations Material Property Basis Limiting Material: Lower Shell Piste Limiting ART at 17 EFPY: 1/4 T,196 F 3/4-T,172 F i

1800.00 ,

.p i

1600.00

/

/

l 1400.00 /

/ ,

i

/

3 1200.00 ,

• iis /

S Unacceptable

/ .

. /

5 1000,00 a

Operation J

f e I g lM A V 800.00 M/> ,

B /W/

y /////

o / 4'//

E 600.00

' " 9 # ,

f ///s /

--- Cooldown Rates i (Fh) --' ' -/ // / ---

/ / Acceptable 0

,- /,

Operation 400.00 20

~

1 40 ,

60 _,

--. goo y

' I 200.00 R

0.00 0 50 100 150 200 250 ~300 350 -

1 Cold Leg Temperature (Deg.' F)

~

North Anna Unst 2 Reactor Coolant System Cooldown Lirnitations (Cooldown Rates up to ~  !!

100 F/hr) Appheable for the First 17 EFPY (Wrthout Margins for instrumentation Erroral l

-1

.d NORTH ANNA - UNIT 2 3/4 4-28 Amendment No. CA, M9,- l l

.l L

REACFOR COOLANT SYSTEM LOW-TEMPERATURE OVERPRESSURE PROTECTION l LIMITING CONDITION FOR OPERATION 3.4.9.3 Two power-operated milef valves (PORVs) shall be OPERABLE with lift settings of (1)less than or equal to 415 psig whenever any RCS cold leg temperature is less than or equal to 270 F, and (2)less than or equal to 375 psig whenever any RCS cold leg temperatum is less than 130 F.

APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 270 F, MODE 5, and MODE 6 when the head is on the reactor vessel and l the RCS is not vented through a 2.07 square inch or larger vent.

ACTION:

a. With one PORV inoperable in MODE 4, restore the inoperable PORV to OPERAl3LE status within 7 days or depmssurize and vent the RCS through at least a 2.07 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b. With one PORV inoperable in MODES 5 or 6, either (1) mstore the inoperable PORV to OPER ABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or (2) complete depressurization and venting of the RCS through at least a 2.07 square inch vent within a total of 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />.

c. With both PORVs inoperable, complete depressurization and venting of the RCS through at least a 2.07 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

d. With the RCS vented per ACTIONS a, b, or c, verify the vent pathway at least once l per 31 days when the pathway is provided by a valve (s) that is locked, sealed, or otherwise secured in the open position; otherwise, verify the vent pathway ever 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

e. In the event either the PORVs or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient, and any corrective action necessary to prevent recurrence.-

f. The provisions of Specification 3.0.4 are not applicable.

NORTII ANNA - UNIT 2 3/4 4-30 Amendment No. 64r449,

REACTOR COOLANT SYSTEM LOW-TEMPERATURE OVERPRESSURE PROTEC110N l SURVEILLANCE REQUIREMENTS 4.4.9.3 Each PORV shall be demonstrated OPERABLE by: l

a. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE.

b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel, at least once per 18 months.

c. Verifying the PORV keyswitch is in the AUTO position and the PORV isolation l valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpressure protection.

d. Testing pursuant to Specification 4.0.5.

t NORTH ANNA - UNIT 2 3/4 4-31

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS -Tavg GREATER THAN 350 F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE centrifugal charging pump,

b. One OPERABLE low head safety injection pump,

c. An OPERABLE flow path capable of transferring fluid to the Reactor Coolant System when taking suction from the refueling water storage tank on a safety injection signal or from the containment sump when suction is transferred during the recirculation phase of operation.

APPLICABILITY: MODES 1,2 and 3.

ACTION:

a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursur.nt to Specification 6.9.2 within 90 days describing the cimumstances of the -

actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in thh; Special Report whenever its value exceeds 0.70.

c. The provisions of Specification 3.0.4 am not applicable to 3.5.2.a and 3.5.2.b for one hour following heatup above 270 F or prior to cooldown below 270 F. l SURVEILLANCE REQUIREMENTS , 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the following valves are in the indicated positions with power to the valve operators mmoved:

NORTH ANNA - UNIT 2 3/45-3 Amendment No. 449,

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS -Tavg LESS THAN 350 F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum,one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE centrifugal charging pump #,

b. One OPERABLE low head safety injection pump # and ,

c. An OPERABLE flow path capable of automatically transferring fluid to the reactor coolant system when taking suction from the refueling water storage tank or from the containment sump when the suction is transferred during the recirculation phase .

of operation.

APPLICABILITY: MODE 4.

ACTION:

a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

I

b. With no ECCS subsystem OPERABLE because of the inoperability of the low head safety injection pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System Tavg less than 350 F by use of alternate heat removal methods.

c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the cimumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this i Special Report whenever its value exceeds 0.70.

l l

1. A maximum of one centrifugal charging pump and one low head safety injection pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 270 F.

l P

NORTH ANNA - UNIT 2 3/45-6 Amendment No.-747 449, 1:

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS i

4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable .;

Surveillance Requirements of 4.5.2.

4.5.3.2 All charging pumps and safety injection pumps, except the above required OPERABLE pumps, shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperature of one or more of the RCS cold legs is less than or equal to 270 F by verifying that the control switch l is in the pull to lock position.

NORTH ANNA - UNIT 2 3/45-7 Amendment No. 449,-

FEACTIVITY CONTROL SYSTEMS

, BASES 3/4.1.2 BORATION SYST6hiS.

The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources,2) charging pumps,3) separate flow paths,4) boric acid transfer pumps,5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 200 F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN M ARGIN from expected operation conditions of 1.77% delta k/k after xenon decay and cooldown to 200 F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 6000 gallons of 12,950 ppm borated water from the boric acid storage tanks or 54,200 gallons of 2300 ppm borated water from the mfueling water storage tank.

With the RCS temperatum below 200 F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requimment to verify all charging pumps except the mquired OPERABLE pump to be inoperable below 270 F provides assurance that a mass addition pressure transient can be l mlieved by the operation of a single PORV.

The boron capability required below 200 F is sufficient to provide a SHUTDOWN MARGIN of 1.77% delta k/k after zenon decay and cooldown from 200 F to 140 F. This condition requires either 1378 gallons of 12,950 ppm borated water from the boric acid storage tanks or 3400 gallons of 2300 ppm borated water from the afueling water storage tank.

l i

i

! NORTH ANNA - UNIT 2 B3/41-3 Amendment No. 51,78,149,

3L4J RHACTOR COOLANT SYSTEM l BASES 3L4dl REACTQR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation and maintain the DNBR above design limit during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within I hour, in MODE 3, a single reactor coolant loop provides sufficient heat removal capabihy for removing decay heat; however, single failure considerations mquire that two loops be 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 <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, a minimum RHR system flow rate of 2000 gpm in MODE 5 is permitted, provided there is sufficient decay heat removal to maintain the RCS temperature less than or equal to 140 F. Since the decay heat power production rate decmases with time after reactor shutdown, the requirements for RHR system decay heat amoval also decrease. Adequate decay heat removal is provided as long as the mactor has been shutdown for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after entry into MODE 3 and RHR flow is sufficient to maintain the RCS temperature less than or equal to 140 F. The educed 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 3000 gpm flow rate 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 10270 F are provided to prevent RCS pressure transients, caused by energy additions l 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 restricting starting of the RCPs to when the secondary water temperature of each steam l 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 opening the cold leg stop valve provides a reassurance of the adequacy of the baron 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 boron 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 stanup 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 this cool water induced reactivity transient.

NORTH ANNA - UNIT 2 B 3/4 4-1 Amendment No. 120,122,149,

1 REACTOR COOLANT SYSTEM BASES 3/4.4.2 AMD 3/4.4.3 SAFETY AND RELIEF VALVES The pressurizer code safety valves operate to prevent the RCS from being pmssurized 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 mlief capacity of a single safety valve is adequate to relieve any overpmssure condition which could occur during hot shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, or the power operated relief valves (PORVs) will provide overpressure relief capability and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pmssurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss ofload assuming no reactor trip until the first Resctor Protective System trip serpoint is reached (i.e., no l credit is taken for a direct reactor trip on ec h n of load) and also assuming no operation of the power operated relief valves or steam dup ees.

Demonstration of the safety valvea aih .ettings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel C(xle. 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 load decrease with steam dump. Operation of the PORVs 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.

The OPERABILITY of the PORVs and block valves is determined on the basis of their being capable of performing the following functions:

a) Manual control of PORVs to control reactorcoolant system pressure. This is a f t on that may be used to mitigate certain accidents and for plant shutdown.

b) Maintaining the integrity of the reactor coolant pressure boundary. This function is n: lated to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.

c) Manual control of the block valve to (1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item a, above), and (2) iso-late a PORV with excessive seat leakage (Item b, above).

NOR fli ANNA - UNIT 2 B 3/4 4-2 Amendment No. 424,

l 3L44 REACTOR COOLANT SYSTEM i BASES d) Automatic control of PORVs to control reactor coolant system pressure. This function reduces challenges to the code safety valves for overpressurization events.

e) Manual control of a block valve to isolate a stuck-open PORV.

Surveillance Requirements provide the assurance that the PORVs and block valves can perform their functions. Specification 4.4.3.2.1 addresses the PORVs and Specification 4.4.3.2.2 -

addresses the block valves. The block valves are exempt from the surveillance requirements to cycle the valves when they have been closed to comply with the ACrlON requin:ments. This precludes the need to cycle the valves with full system differential pressure or when maintenance is being performed to restore an inoperable PORV to operable status.

Surveillance Requirement 4.4.3.21.b.2 provides for the testing of the mechanical and electrical aspects of control systems for tae PORVs.

Testing of PORVs in IIOT STANDBY or HOT SIlUTDOWN is required in order to simulate the temperature and pressure environmental effects on PORVs. Testing at COLD SilUTDOWN is not considered to be a representative test for assessing PORY performance under nonnal plant operating conditions.

3L4dA PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady state envelope of operation assumed in the S AR. The limit is consistent with the initial S AR assumptions. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE ensures that the plant will be able to establish natural circulation.

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- REACTOR COOLANT SYSTEM l BASES The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 ~ Ci/ gram DOSE '

EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. l Reducing Tavg to less than 500 F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency ofisotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System l All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by .

normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 5.2 of the UFSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel.' These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent.on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Consequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate ofinterest must be analyzed on an individual basis.

NORTH ANNA - UNIT 2 B3/44-6 Amendment No. 83r449,

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REACTOR COOLANT SYSTEM l BASES  !

l ll The heatup limit curve, Figure 3.4-2, is a compesite curve which was prepared by I determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60 F per hour. The cooldown limit curves of Figure 3.4-3 am composite curves which were prepared based upon the same type a..alysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cooldown curves wem prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 17 EFPY. The most recent capsule analysis results are documented in Westinghouse i Reports WCAP-12497, January 1990. The heatup and cooldown curves are documented in Westinghouse Report WCAP-12503, March,1990.

The mactor vessel materials have been tested to determine their initial RTNDT. Reactor operation and resultant fast neutron (E > 1 Mev) Irradiation will cause an increase in the RTNDT-An adjusted reference temperature, based upon the fluence and copper content of the materialin l question, can be predicted using US NRC Regulatory Guide 1.98, Revision 2. The heatup and cooldown limit curves (Figum 3.4-2 and Figure 3.4-3) include predicted adjustments for this shift in RTNDT at the end of 17 EFPY. The reactor vessel beltline region material properties are listed on Figures 3.4-2 and 3.4-3.

The actual shift in the RTNDT of the vessel material will be established periodically by removal and evaluation of the reactor vessel material specimens installed on the inside wall of the thermal shield. The surveillance capsule withdrawal schedule was prepared in accordance with the mquirements of ASTM E-185 and is presented in the UFSAR. Regulatory Guide 1.99, Revision 2, provides guidance for calculation of the shift in RTNDT using measured data. Dosimetry from the surveillance capsule is used to determine the neutron fluence to which the material specimens were exposed, and to support calculational estimates of the neutron fluence to the reactor vessel.

The presstre-temperature limit lines shown on Figure 3.4-2 for inservice leak and hydrostatic testing 1. ave been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. The minimum temperature for criticality specified in T.S. 3.1.1.5 assures compliance with the criticality limits of 10 CFR 50 Appendix G.

The number of reactor vesselirradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in the UFSAR to assure compliance with the l requirements of Appendix 11 to 10 CFR Part 50. ,

14essurizer l The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential am provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

NORTil ANNA - UNIT 2 B3/44-7 Amendment No. 449,

REACTOR COOLANT SYSTEM l BASES Low-Temocrature Overoressure Protection l The OPERABILITY of two PORVs or an RCS vent opening of greater than 2.07 square -

inches ensures that the RCS will be proteued from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 270 F. Either PORV has adequate relieving capability to protect the RCS from -l overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generc,orless than orequal to 50 F above the RCS cold leg temperatures or (2) the start of a charging pump and its injection into a water-solid RCS, _l Automatic or passive low temperature overpressure protection (LTOP) is required whenever any RCS cold leg temperature is less than 270 F. This temperature is the water temperature corresponding to a metal temperature of at least the limiting RTNDT + 50 F +

instrument uncertainty. Above 270 F administrative controlis adequate protection to ensure the limits of the heatup curve (Figure 3.4-2) and the cooldown curve (Figure 3.4-3) are not violated.

The concept of requiring automatic LTOP at the lower end, and administrative control at the upper end, of the Appendix G curves is further discussed in NRC Generic Letter 88-11.

3/4.4.10 STRUCTUR AL INTEGRITY 3/4.4.10.1 ASME CODE CLASS 1. 2 and 3 COMPONENTS The inspection programs for ASME Code Class 1,2 and 3 Reactor Coolant System components ensure that the structuralintegrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.

NORTH ANNA - UNIT 2 B3/44-8 Amendment No. 449,

d EMERGENCY CORE COOLING SYSTEMS (ECCS) l BASES ECCS SUBSYSTEMS (Continued)

With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements, The limitation for a maximum of one centrifugal charging pump and one low head safety injection pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and low head safety injection pumps except the required OPERABLE pump to be inoperable below 270 F provides assurance that a mass addition pressure transient can be relieved by the l operation of a single PORV.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for the throttle valve position stops and .l flow balance testing provide assurance that proper flow msistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS - LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above tha: assumed in the ECCS - LOCA analyses.

3/4.5.4 BORON INJECTION SYSTEM The OPERABILITY of the boron injection system as part of the ECCS ensures that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS system cooldown. RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident or a steam line rupture.

The limits on injection tank minimum contained volume and boron concentration ensure that the assumptions used in the steam line break analysis are met. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

The OPERABILITY of the mdundant heat tracing channels associated with the boron injection system ensure that the solubility of the boron solution will be maintained above the solubility limit of 111 F at 15,750 ppm boron.

NORTH ANNA - UNIT 2 B 3/4 5-2 Amendment No. 54,149,

ADMINISTRATIVE CONTROLS SPECIAL REPORTS ,

1 6.9.2 Special reports shall be submitted to the Regional Administrator, Region II, within the time l period specified for eac.h report. These reports shall be submit'ed pursuant to the requirement of l the applicable specification:

a. Inservice Inspection Reviews, Specification 4.0.5, shall be reponed within 90 days of completion,

b. MODERATOR TEMPERATU.RE COEFFICIENT. Specification 3.1.1.4.

c. Deleted.

d. RADIATION MONITORING INSTRUMENTATION. Specification 3.3.3.1, Table 3.3-6, Action 35.

e. Deleted.

f. LOW-TEMPERATURE OVERPRESSURE PROTECTION. Specification 3.4.9.3.

g. EMERGENCY CORE COOLING SYSTEMS, Specification 3.5.2 and 3.5.3.

h. SE'ITLEMENT OF CLASS 1 STRUCTURES. Specification 3.7.12,

i. GROUND WATER LEVEL - SERVICE WATER RESERVOIR. Specification 3.7.13.

j. Deleted,

k. Deleted.

1. RADIOACTIVE EFFLUENTS. As required by the ODCM.

m. RADIOLOGICAL ENVIRONMENTAL MONITORING. As required by the ODCM.

n. SEALED SOURCE CONTAMINATION. Specification 4.7.11.1.3.

o. REACTOR COOLANT SYSTEM STRUCTURAL INTEGRITY. Specification 4.4.10. For any abnormal degradation of the structuralintegrity of the reactor vessel -

or the Reactor Coolant System pressun: boundary detected during the performance of Specification 4.4.10, an initial mport shall be submitted within 10 days after detection and a detailed report submitted within 90 days after the completion of Specification 4.4.10.

p. CONTAINMENT STRUCTURAL INTEGRITY. Specification 4.6.1.6. For any .

abnonnal degradation of the containment structure detected during the performance of Specification 4.6.1.6, an initial report shall be submitted within 10 days after completion of Specification 4.6.1.6. A final report, which includes (1) a description of the condition of the liner plate and concrete, (2) inspection procedure, (3) the tolerance on cracking, and (4) the corrective actions taken, shall be submitted within 90 days after the completion of Specification 4.6.1.6.

NORTII ANNA - UNIT 2 6 21- Amendment No. 31,47,83,5m, 1 M,123,

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SIGNIFICANT- HAZARDS CONSIDERATION y

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~ VIRGINIA ELECTRIC AND POWER COMPANY ,

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Significant Hazards Consideration The North Anna Units 1 and 2 Reactor Coolant Systems are protected' from material failure by the imposition of restrictions on allowable pressure and temperature, and on heatup and cooldown rate. The Low Temperature Overpressure Protection System (LTOPS) ensures that material integrity limits are not exceeded during the design basis overpressurization accidents. Equipment operability requirements are imposed to ensure that the assumptions of the accident analyses remain valid. The operating restrictions, setpoints, and equipment operability requirements must be revised to extend their applicability to a higher cumulative burnup, and to improve operational flexibility.

The current pressure / temperature operating limits and LTOPS setpoints are valid to 12 EFPY and 17 EFPY for North Anna Units 1 and 2, respectively. According to the most recent estimates, the burnup applicability limits will be exceeded by North Anna Unit 1 in Spring of 1996. The North Anna Unit 2 pressure / temperature operating limits and LTOPS setpoints remain valid well into the year 2002. The proposed North Anna Unit 1 Technical Specifications include revised pressure / temperature operating limits valid to end-of-license. Although the North Anna Unit 2 Technical Specification pressure / temperature operating limits are not being changed,. the Unit 2 LTOPS setpoints and associated reactor vessel integrity protection philosophy are being changed. The reactor vessel integrity protection philosophy which supports the proposed Technical Specifications changes provides improved operational flexibility while maintaining an adequate margin of safety as demonstrated by the safety analysis. The probability and consequences of the design basis low temperature overpressurization accidents are not increased.

Virginia Electric and Power Company has reviewed the Technical Specification changes against the criteria of 10 CFR 50.92 and has concluded that the changes do not pose a significant hazards consideration. Specifically, operation of Surry Power Station.in accordance with the Technical Specification changes will not:

a. involve a significant increase in the probability or consequences of an accident previously evaluated. The safety analysis demonstrates that the proposed reactor vessel protection philosophy, and the associated pressure / temperature limits, LTOPS setpoints, and component operability requirements, ensure that reactor vessel integrity will be maintained during normal operation and design basis accident- 1 conditions. Specifically, adherence to the heatup/cooldown -l rate-dependent pressure / temperature operating limits ensures ,

that the assumed design basis flaw will not propagate during l normal operation. Below the LTOPS enabling temperature, .j automatic. actuation of the PORVs ensures that the assumedi cj design basis flaw will not propagate under design basis' low- j temperature overpressurization accident conditions. Two 1 pressurizer safety valve are sufficient to relieve the overpressurization due to the inadvertent startup of two charging pumps at water solid conditions without propagation of the assumed design basis flaw. The proposed changes to 1

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address the concerns of Generic Letter 90-06 (Generic . Issues 70 and 94) improve' . LTOPS availability ~and reliability by instituting requirements for PORV, block valve, and control system' testing and allowed outage times for.these components.

Although these changes do not reduce the probability, of occurrence or the consequences of the LTOPS design basis (mass.

and heat addition) transients, the changes provide ~ increased assurance that pressure relieving devices will perform their design function when required,

b. create the possibility of a new or different kind of accident-

• from any accident previously evaluated. The proposed Technical Specifications modify pressure / temperature' operating limits, LTOPS setpoints and enabling temperatures, and component operability requirements. The revised pressure / temperature operating limits, and LTOPS setpoints;and, enabling temperatures are only slightly different than those ~ >

currently in the Technical Specifications. No operating limits or setpoints are added or deleted by the proposed changes. Therefore, it may be concluded that the operating limits and setpoint changes do not create the possibility offa:

new or different kind of accident'. With regard to component operability requirements, restrictions on the number of charging pumps which may be operable, the number of PORVs which must be operable, and the allowable temperature difference between the steam generator primary'and secondary ,

remain unchanged. Only the setpoint temperature at 'which-these restrictions apply have been modified. The proposed' ,

changes -are entirely consistent with the reactor ' vessel-integrity protection philosophy which ensures that the design ~

basis reactor vessel flaw will not propagate under normal operation or postulated accident conditions. Further,.the proposed changes do not' invalidate _the any component design criteria or..the assumptions of any'UFSAR Chapter 15 accident n analyses. In addition, modifications have been made to the Technical Specifications to improve availability 'and reliability of PORVs and associated block valves. These changes have been made in accordance with NRC guidance in Generi~c Letter 90-06. It may be concluded that none of the proposed changes creates the possibility of a new or different kind of accident from any previously evaluated.

c. involve ..a significant reduction in - a margin of~ safety. _ As described- above, the reactor vessel ' integrity ' protection .

philosophy' ensures that the design basis assumed flaw will not.

propagate under. normal operation or design basis' accident conditions. Adherence .to the Technical Specification pressure / temperature operating limits 1 ensures.that the margin to vessel fracture provided by the ASME'Section-XI methodology.

is maintained. With regard to LTOPS _ protection, the ~ safety- ;i analysis demonstrates that the proposed LTOPS design ensures. l margins consistent with those prov.ided by ASME Section XI .l Appendix G methods.

experience .

w ith LTOPS This conclusion is based on industry events and engineering evaluation.

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SpecifIcally,. both industry experience and- engineering-  :)

evaluation demonstrate that-LTOPS design basis events. may'be 21 1 i

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expected ' to occur at essentially isothermal conditions. '

Engineering evaluation demonstrates. that 'a n y ' r e d u c t i o n ' i n allowable pressure due to thermal stresses which may be o expected to occur during low temperature- operation is insignificant when compared to ' margins provided by the 'ASME Section .XI Appendix G methods for calculating pressure / temperature operating limits. Use of the. isothermal  !

pressure / temperature limit curve. as the design . . limit for I establishing low temperature PORV lift setpoints has been approved for other utilities by the NRC. This design  !

maximizes- the operating margin above the minimum RCS pressure l for reactor' coolant pump (RCP) operation, thereby minimizing I' the Lprobability of undesired PORV lifts during RCP startup.

The proposed changes to address.the concerns of Generic Letter.

90-06 (Generic Issues 70 and 94) improve' LTOPS . availability' and reliability by instituting. requirements for'PORV, block valve, and control system testing and allowed outage t-imes for these components. Although these changes do not increase the margin of safety demonstrated by the analysis of the LTOPS' design basis (mass and heat addition) transients, the changes provide increased assurance that pressure relieving devices -

. will perform their design function when required.

Virginia Electric and Power Company concludes that the activities associated with these proposed Technical Specification changes satisfy the no significant hazards consideration criteria of 10 CFR 50.92 and, accordingly, a no significant hazards consideration finding is justified.

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