Text

Amends 170 & 149 to Licenses NPF-4 & NPF-7,respectively, Revising P/T Operating Limitations During Heatup & Cooldown & Ltops Setpoints for 12 & 17 EFPYs for NA-1 & 2, Respectively
	ML20035B874
Person / Time
Site:	North Anna
Issue date:	03/25/1993
From:	Berkow H; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20035B875	List: ML20035B874; ML20035B880;
References
	NUDOCS 9304050227
	Download: ML20035B874 (61)
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{{#Wiki_filter:~ ha 4 E ; 'g _ 4 J 't IJNITED STATES j ki ij NUCLEAR REGULATORY COMMISSION g mi f WAsmGTov D C 20555 0001 VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-338 NORTH ANNA POWER STATION. UNIT NO. 1 [ AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 170 License No. NPF-4. 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated December 27, 1991,. complies with the standards and requirements of the Atomic Energy Act of-1954, as amended (the Act), and the Commission's rules and regulations set l forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; i C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the_ public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. l 1 9304050227 930325 PDR ADOCK 05000338 P PDR i

2. Accordingly, the license is amended by changes to the Technical Spect-fications as indicated in the attachment to this license amendment, and i paragraph 2.D.(2) of Facility Operating License No. NPF-4 is hereby amended to read as follcws: (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 170, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented by August 17,1993 or 10 effective full power years, whichever occurs first. FOR THE NUCLEAR REGULATORY COMMISSION / h{ Her ert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 25, 1993 )

~ ATTACHMENT TO LICENSE AMENDMENT NO. 170 TO FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness. Remove Paaes Insert Paoes 3/4 1-9 3/4 1-9 3/4 1-12 3/4 1-12 3/4 4-3 3/4 4-3 3/4 4-26 3/4 4-26 3/4 4-27 3/4 4-27 3/4 4-28 3/4 4-28 l 3/4 4-31 3/4 4-31 3/4 5-3 3/4 5-3 3/4 5-6 3/4 5-6 3/4 5-6a 3/4 5-6a B 3/4 1-3 -B 3/4 1-3 8 3/4 4-1 B 3/4 4-1 B 3/4 4-6 B 3/4 4-6 t B 3/4 4-7 8 3/4 4-7 B 3/4 4-8 B 3/4 4-8 8 3/4 5-2 B 3/4 5-2 I r I I

~ REACTMTV CONTAOL SYSTCVS FLOW DATHS - OPER ATING i UMITING CONDITION FOR OPERATION 3.1.2.2 Each of the following boron injection flow paths shall be OPERABLE: i

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.

APPUCABILITY: MODES 1,2,3 AND 4#. ACTION: a. With the flow path from the boric acid tanks inoperable, restore the inoperable flow path to OPERABLE status within 72 hours or be in at least 4 i HOT STANDBY and borated to a -SHUTDOWN MARGIN equivalent to at least 1.77% ak/k at 200'F within the next 6 hours; restore the flow path to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours.

b. With the flow path from the refueling water storage tank inoperable, restore the flow path to OPERABLE status within one hour or be in at teast HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

SURVEILLANCE REOUREMENTS 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 2115*F. 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 316*F. l NORTH ANNA - UNIT 1 3/4 1-9 Amendment NoA6,48,4+7,170

9

ESC **VIT< ::NT3CL Sv5~Ev5 S'C:VEILLANCE :EOUIDEwENTS (Continued) l b.

at least cnce Der 31 days by verifying that each valve (manual,

cwer c: erated or automatic) in the. flow path that is not locked, sealed, or otnerwise secured in position, is in its correct position.

At least once per 18 months during shutcown by verifying that c. each automatic valve in tre flow path actuates to its correct ocsition on a safety injection test signal, t i d f 1 A 1 t i i i 1 I i f I t NCRTH ANNA-UNIT I 3/4 1-10 T

il;EACTIV!TV ::NT:0L 5<5~!"5

-A:3IN3 ocu:. Sau :0WN L:w!'*N3 CONDIT:0N FOR OPERAT CN 3.1. 2. 3 At least one cnarging pump in the boron injection flow path repaired by Specification 3.1.2.1 shall te OPERABLE.

LICABILITY: M00E5 3 and 5 ACT:CN:

With no charging ;umo OPERABLE, sus:end all o;erations involving CORE t a. ALTERATIONS or ositive reactivity changes until one charging pump is restored to CPERABLE status. b. With no charging pump OPERABLE and the opposite unit in MODE 1, 2. 3 or a, immeciately initiate corrective action to restore at least one caarging pumo to CPERABLE status as soon as possible. SURVEILLANCE ECUI.E*ENTS 4.1. 2. 3.1 At least the above required charging ; ump shall be demonstrated 3 0:ERABLE by verifying inat, on recirculation flow, the pump cevelocs a discharge pressure of 3,2410 psig anen tested pursuant to 5:ecification 4.0.5. a.l.2.3.2 All charging pumps, except the acove required OPERABLE pump, - sna11 te demonstrated inoperacle at least once per 12 nours by verifying that the swittnes in the Control Room have been placed in the pull to lock oosition. ll NORTH ANNA - UNIT 1 3/4 1-11 Amendment No. 76, 24

.l REACTIVTTY CONTAOL SYSTC_MS CHARGtNG PUMPS - OPERATING l UMITING CONDITION FOR OPERATION l 3.1.2.4 At least two charging pumps shall be OPERABLE. I APPLICABILITY: MODES 1,2,3 and 4. I ACTION: -i i With only one charging pump OPERABLE, restore a second charging pump to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a l SHUTDOWN MARGIN equivalent to at least 1.77% auk at 200'F within the next l 6 hours; restore a second charging pump to OPERABLE status within the next 7 j days or be in COLD SHUTDOWN witt.in the next 30 hours. The provisions of Specification 3.0.4 are not applicable for one hour following heatup above 316'F [ or prior to cooldown below 316*F. { SURVEILLANCE REQUIREMENTS i 4.1.2.4.1 The above required charging pumps shall be demonstrated OPERABLE by verifying, that on recirculation flow, each pump develops a discharge pressure of 2 2410 psig when tested pursuant to Specification 4.0.5. 4.1.2.4.2 All charging pumps, except the above required. OPERABLE pump, shall be i demonstrated inoperable at least once per 12 hours whenever the temperature of one or more of the RCS cold legs is less than or equal to 316*F by verifying that l 1 the switches in the Control Room have been placed in the pull to lock position. ] i i ~ 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 316'F, .l -1 ,) I i

l NORTH ANNA - UNIT 1 3/4 1-12 Amendment No.4,44,4+7.170 i

m m

PEACTOA CCOJNT SWEA SMmXMN UMIT!NG CONDITION FOR OPERATION 3.4.1.3

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

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

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

3. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, 4.

Residual Heat Removal Subsystem A,

5. Residual Heat Removal Subsystem B.

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

A.PPLICABfLiTY: MODES 4 and 5 ACT10N: l

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

corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 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 action to return the required coolant loop i

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 316*F unless the secondary water temperature of l each steam generator is less than 50*F above each of the RCS cold leg temperatures. The offsite or emergency power source may be inoperable in MODE 5. All reactor coolant pumps and residual heat removal pumps may be de-energized for up to 1 hour provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10*F below saturation temperature. NORTH ANNA - UNIT 1 3/4 4-3 Amendment NoA6,G2,4+7,170 i i

REACTOR COOLANT SYSTEM SHUTDOWN j i l SU AVEIL L ANCE GEQUIREMENTS 4 4.1 3.1 The recuired RHR sucsystems shall be cemonstrated CPERABLE per Sceoticaton 4.7 9 2. 4 4 1.3.2 The required reac:or coolant pumots), if not in operaton, shall be cetermined to be CPERABLE once per 7 cays cy ventytog cCrrect Dreaxer alignment and incicated power availacihty. 4 4 1 3.3 The recuired steam generator (s) shall be determined CPERABLE cy ventying secencary sice water :evel to te greaterinan or equal to 17% at least once per 12 hours. 4.4 1 3.4 At feast once per 12 hours. venty at ' east one coolant loco to te in operaton and c:rculatng reac:cr coolant by: a. Ventying at least one Reac:cr Coolant Pumo is an operation. or b. Ventying at feast ene RHR Loco is in operaten and, 1. if the RCS temoerature >140' F orthe time since entry into MCCE 3 is <100 hours, crculatng reacer coolant at a flow rate 23000 gpm. or

2. if the RCS temcerature 5140' F and the tirne since entryinto MOCE 3 is 2100 hours crculatng reacer coolant at a tiow rate 22000 gpm to remove decay heat.

NCRTH ANNA - UNIT 1 3/4 4-3a AmendmentNo.32,g/ 1 AUG 2 71990 )

i N 7,

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250 \\ \\ 1 .iii D \\ . i ! ! ! i i x 1 3 U ' k, l l l ,,ll l,,[ y L 4 >= \\ i .e t, i. .,i .i,, g \\ i,.. 1 g U \\'

' ' UNACCEPTABLE 4

OPERATION .x . i u T -g ( l t I I N -v N E i i N 4, \\ i e i.- A >= jg '\\.l l'" l'll .... 1.. ; i. a, a \\.,i g 1 i. ..x i .., i c: . si i ,.T .. i. e 5 2 i i,.T g og . 8 i. ' i i, i I . 9 i e i} f. \\s g ) . i * , t i i i ii j i j \\ t o M .i 8 e i ) i i 6 +t si e 7 i.. , e i 3 ,i.i I e i i ! I i 1 i 4; e, s . i e i 1 i ii4 4 i e i t il , e i i i. ACCEPTABLE + ' + i ! +i +ia llll ll'l "j OPERATION 50 5 - +>> i,, i .. i . i e, 4 i + 4 ii i + 4 i u 1 i i. i ! 8 # 1 i i i i 6 e {, i e . 4 il i ! i j6 i i , i. . i g . 4 8 j 6 4 i t . 6 1 ( } , e t i i t. i. i e i a ,. i 20 30 40 50 80 70 to 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Umit Versus Peroent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 1.0gCl/ pram Dose Equivalent 1-131 NORTH ANKA - UNIT 1 3/4 4-25 1

W f REACTOR COOLANTSWTmA i 374 4.9 PRESSURE /TEVDEAATURE LPAITS i REACTOR COOLAVTSYSTmA j l UMITING CONDITION FOR OPERATION I 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure I shall be limited in accordance with the limit lines shown in Figures 3.4-2 and 3.4-3 during heatup, cooldown, and inservice leak and hydrostatic testy, with: l 1

a. A maximum heatup of 60'F in any one hour period.

b. A maximum cooldown of 100*F in any one hour period.

c. A maximum temperature change of less than or equal to 10*F in any one hour I

period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves. APPLICABILITY: At all times. eQQN: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains l acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours and reduce the RCS Tavg and pressure to less than 200*F and 500 t psig, respectively, within the following 30 hours. SURVEll1ANCE REOUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be deter *hined to be l within the limits at least once per 30 minutes during system heatup, cooldown j and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50, Appendix H. The results of these examinations shall be { used to update Figures 3.4-2 and 3.4-3. L l l NORTH ANNA - UNIT 1 3/4 4-26 Amendment No.-86,170 F 1

l i l i e i Figure 3.4-2 Unit 1 RCS HEATUP P/T Lirnits t Valid to 12 EFPY Heatup Rates: 0-60*F/Hr. t (Margins for Instrurnent Errors NOT included) t 1 i 2750 - -< d- ... h.... MATERIAL PROPERTY BASIS..'-[.-.......".)'...,... r- ) ..I... ... [.. ... l....... [p. :.. . Centrollin9 Material: l

Lewer Shell Plate Forging 2500 - Copper Content: 0.15 W1.%

...h 9 Y - -:- ' H - k.. - h...:".

Nickel Content: 0.80 W1.%
initial RTndt: 38'F Leak Test -;-

4 ---- '-'fP-- { R t RTnct After 12 EFPY: '.... Li m i t C 2250j 1/4T,145'F ...i... .!3/4T.122*F ..m. g

Heatuo Rate s60*F/Hr.

... {.....:.. .;..p... .r --f-----

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....l... ...g....... ....r... i 2. .;.4 i _.p.......). 3.. 7..., ..,....;...s.4...y... p 750 .g L.. ACCEPTABLE -~ L....:.... OPERATION g... .g...q...).,. 500 ..........d...].:.{... ... i......:. a..... ;...;.,.. '.... :.. .....s. 7.. 4...:..... p.;._.;...:....:.. 7.... :....:.....q...q. 29 i 80 100 120 140 160 180 200 220 240 260 280 300 320 i i Minimum RCS Cold Leg Temperature - degrees F r i I NORTH ANNA-UNIT 1 3/4427 Amendment No. 44,.74,447,17o a r 4 t h k i

t t Figure 3.4-3 Unit 1 RCS COOLDOWN P/T Limits i Valid to 12 EFPY Cooldown Rates: 0-100*F/Hr. (Margins for instrument Errors NOT Included) r 2750 i MATERIAL PROPERTY BASIS '^...".3 7.,.'.' '. ' 7,

I ~ :~ h. *..'

Controlling Material: *-~'"'-I"*J"'?-" 2500 Lower Shell Plate Forging

Copper Content: 0.15 W1.%

. l. :...

4... F...:. J. ;...:. {. 4. i. ;..

l .!Nickel Content: 0.80 W1.% . !..-.".. J. :.-.. :...r. -. -.. - 4-I initial RTndt: 38'F ...J... R 2250 - RTnct After 12 EFPY: ~. +g "...' '. 7...~ ~. ~. ' T. '. '. ' '. }* '. Q" " V.. g ~". ~ i -- - f.. C 1/4T.145'F i S ! 3/4T.122'F -.< ~ y t'-~.

Cooldown Rate: s100'F/Hr.

T~~'''~~'~

'1

2000 p......{...j....J...[.3...t.. ...L.. l...,..[. 4...t... y..........,....... ......t..... P ...L.. . q...~... l z............t.

p..., 4...l......,...........p..;...}.....,....j...q....q......

r e 1750 i.~...4...,,... ..i......t. .. 7...,... L. ;.. ..J.,.,... s

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I ... T.. l. <.........T....... ..i ..J.,. ,.s. ..J.,. s .,....j.....a,..,,,,.........q...;,,J.,',,, u 1500 i i e i i L...s... F UNACCEPTABLE C. e p...{... OPERATION .:.;..:..:...:.q......:.. 1250 p..... ..... 4...p.'......z.l....:.....L.:..'.4 ..2...... ...t...... L..........................-{........ i P 7 ... J.... p..... . :. 3. l. s 1000 t ......t.....a...,...,..... L.....!.....2...[..J... .. 1 .,...[......} 1... . J,... - 3 9 [...... . :..... J. 3. p.:. 4 .:.....].;....p.... 750 t p. a...t.....:....t ..;.. -. ;....:. J... [... [. ...r...,..,... rthr f.... q.............,../.,...[,.q...{... o. 500 +p. 20 [. . Xl. ,o. 4 ACCEPTABLE ,o.

p. Soo-4

.s. OPERATION I 7 ..l.....7..q... 250 ........a........:..........t.. .... a.,.. ....,.t... r ... j. ['.. . s. ..s. .4.. 0 80 100 120 140 160 180 200 220 240 260 280 300 320 Minimum RCS Cold Leg Temperature - degrees F i NORTH ANNA. UNIT 1 3/4428 Amendment No.-M,44,4+7,170 t h

REACTOR COO!. ANT SYSTAA OVERPPFSSURE PROTCCTION SYSTEMS UMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE:

a. Two power operated relief valves (PORVs) with a lift setting of: 1) less than or equal to 450 psig whenever any RCS cold leg temperature is less than or equal to 270'F, and 2) less than or equal to 390 psig whenever any RCS cold l

ieg temperature is ess than 150'F, or

b. A reactor coolant system vent of greater than or equal to 2.07 square inches.

APPLICA9fuTY: When the temperature of one or more of the RCS cold legs is less than or equal to 270'F, except when the reactor vessel head is removed. l ACTION: a. With one PORV inoperable, either restore the inoperable PORV to OPERABLE status within 7 days or depressurize and vent the RCS through 2.07 square inch vent (s) within the next 8 hours; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status.

b. With both PORVs inoperable, depressurize and vent the RCS through a 2.07 square inch vent (s) within 8 hours; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status.

c. In the event either the PORVs or the RCS vent (s) are used to mitigate a 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. d. The provisions of Specification 3.0.4 are not applicable. l l i NORTH ANNA-UNIT 1 3/4 4-31 Amendment No.-4,74,447,170

i 1 .l l l l 1 i

EACTD4 C)0LANT SYS*EM

~ SURVEIttANCE REOUIREWENTS 4 4.4.9.3.1 Each PCRV shall be demonstrated CPERABLE by: Derformance of a CHANNEL FUNCTIONAL TEST on the P a. cnannel, but excluding valve operation, within 31 days crior to entering a condition in which the PORV is required OPERABLE and at least once er 31 days thereafter when the PORV is required i OPERABLE. b. Performance of a CHANNEL CALIBRATION on the PORY ac cnannel, at least once per 18 months. i 3 c. Verifying the PCRV keyswitch is in the Auto position and the PORV isolation valve is open at least once per 72 hours when the PORV is being used for overpressure protection. d. Testing pursuant to Specification 4.0.5. I 4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once :er i 12 hours' when the vent (s) is being used for overpressure protection.

Except wnen the vent pathway is provided with a valve which is locked, sealed, or othentise secured in the open position, then verify these valves open at least once per 31 days.

l t i I iORTH AMMA - UNIT 1 3/4 4-32 Amendment No. 4,3 4 i i

OJERGENCYCORE COOLFG SYSTOAS ECCS SUBSYSTEMS - Tavg 2 350*F UMmNG CONDmON 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 hours or be in HOT SHUTDOWN within the next 12 hours. 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 pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. c. The provisions of Specifications 3.0.4 are not applicable to 3.5.2.a and 3.5.2 b for one hour following heatup above 316*F or prior to cooldown below 316*F. l Adherence to ACTION *a" shall require the following equipment OPERABILITY for the period of operation until steam generator replacement: With one low head safety injection pump inoperable, two centrifugal charging pumps (one in each subsystem) and their associated flow paths shall be OPERABLE or be in HOT STANDBY within the next 6 hours, and be in HOT SHUTDOWN within the next 6 hours. NORTH ANNA - UNIT 1 3/4 5-3 hendment *!o. h M, , M4.170

t L I t .EME:.GENCY CCRE CCCL:NG 5YS~ EMS SURVEILL*NCE :E;UIREuENTS (Continued) i 2. Verifying that eacn of tne following pumps start auto-- matically upon receipt of a safety injection test signal: a) Centrifugal charging pump, and b) Low need safety injection pump. f. By verifying that eacn of the following pumps develop the incicated discharge pressure (after subtracting suction pres-sure) on recirculation flow -nen tested pursuant to Specifica-tion 4.0.5. 1. Centrifugal charging pump 1 2410 psig. 2. Low head safety injection pump 1 156 psig g. By verifying tnat the following manual valves requiring adjustment to prevent puma " runout" and subsecuent-camponent damage are locked and tagged in the proper position for injection: 1. Witnin a hours following completion of any repositionin, f or maintenance on the valve when the ECCS subsystems are recuired to be CFE:.A5LE. 2. At least once per 18 months. l. 1-SI-lE3 Loop A Cold Leg l 2. 1-5 !-1 91 Loop B Cold Leg l 3. 1-51-193 Loop C Cold Leg i 4 1-51-203 Loop A Fot Leg 5. 1-51-2n Loop B Fot Leg 6. 1-SI-205 Loop C Hot Leg h. By performing a flow balance test, during shutdown, following ~ completion of modifications to the ECCS subsystems that alter the subsystem flow cnaracteristics and verifying that: 1. For high head safety injection lines, with a single pumo running: ? a) The sum of the injection line flow rates, excluding the hignest flow rate, is 3 384 gpm, and i b) The total rums ficw rate is 1 650 gpm. NORTH ANNA-UNIT 1 3/4 5-5 Amendment No. 6,19

t OMAGPCYCOAE COCU#3 SYSTPAS l I ECCS SUBSYSTEMS - Tave < 350'F UMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shat! be OPERABLE:

a. One OPERABLE centrifugal charging pump #,

t

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.

APPUCAB!UTY: MODE 4 ? ACTON:

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 hour or be in COLD SHUTDOWN within the next 20 hours.

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 Repor1 shall be prepared and submitted to the Ccmmission pursuant to Specification 6.9.2 within 90 days describing the c:rcumstances of the actuation and the total accumulated actuation cycles to date. 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 316'F. l [ NORTH ANNA - UNIT 1 3/4 5-6 Amendment No.4,4-6,8 4,4-4-7.170 l t

j I EMERGENCYCORFCOOLN3 SMAS l e SURVEILLANCE REQUIREMENTS -l 4.5.3.1 The ECCS ' subsystem shall,be demonstrated OPERABLE per the applicable ~ Surveillance Requirements of 4.5.2. j 4.5.3.2 All charging pumps and safety injection pumps, except the above required i OPERABLE pumps, shall be demonstrated inoperable at least once per 12 hours-t whenever the temperature of one or more of the RCS cold legs is less than or j equal to 316*F by verifying that the switches in the Control Room are in the pull l to lock position. l l l f I t -l ~f i ' t i i i a i t I NORTH ANNA - UNIT 1 3/4 5-6a Amendment No.-44,4+7,170 } r

REACTMTV CONTROL SYST;MS i ) BASES r 3 /4.1. 2 BOR ATION SYSTEMS (Continud With the RCS average temperature above 200'F, a minimum of two separate and redundant boron injection systems are provided to ensure 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 20mF. This expected boration capability requirement occurs at EOL from full power equilibrium i 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 refueling 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 316*F provides assurance that a mass addition pressure transient l can be 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 reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable. i 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 f 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 B 3/413 Amendment No.6.Hi.6-8,MA+7,170

EACTMTY CN SYSTEMS SASEs 14 5 2 9CAAWON SYSTEVS rCemmed)

The limits on contained water volume and boron concentration of the RWST ensuI value of cerween 7.7 and 9.0 for the solution recirculated within tne contain This pH minimizes the evolution of ioctne and minim corrosion on mecnanical systems and components. Zes the effect of chlonde and caustc stress l At least one charging pump must remain operable at all times when the opposite unit is in MODE 1. 2. 3. or 4 This is requirec to maintain the enarging pump cross-connect system operational. 1413 uCVABLE CONTAOL ASSEMBLIES The specifications of this section (1) ensure that acceptable power distribution limits are maintained. (2) ensure that the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalignment on associated accident anatyses. OPERABILITY of the movable control assemblies is established by observing rod motion and determining that rods are positioned within : 12 steps (incicated position) of the respectNe demand step counter position. The OPERABluTY of the individual rod position indication system i.s estabilshed oy "l ar.W-iate perode CHANNEL CHECKS. CHANNEL FUNCTIONAL TESTS, and CHANNEL CAUBRATIONS. OPERABILITY cf the individual rod position indicaers is required to determine control rod position and thereby ensure compliance with the control rod alignment and insanion limits. The OPERABLE condition for the individual rod position indicators is defined as being capable of indicating rod position within : 12 steps of the assoc:ated demand position indicator. For power levels below 50 percent of RATED THERMAL POWER, the specifications of this section permit a maximum one hour in every 24 stabilization period (thermal " soak time *) to j allow stabilization of known thermal drift in the indNidual rod position indicator channels dunng which time the indicated rod position may vary from demand position indic& tion by no more than : 24 steps, This *1 in 24' feature is an upper limit on the frequency of thermal soak allowances and is availab6e both for a continuous one hour period or one consisting of several discrete intervals. During this staDilization penod, greater reliarce is placed upon the demand position indicators to determine rod position. In addition, the : 24 stepMour limit is not applicable when the control rod position is known to be greater than 12 steps from the rod group stop counter demand posnion indicaten. Above 50 percent of RATED THERMAL POWER, rod motion le not eWad to induce thermal transients of sufficient magnitude to exceed the indtvidual rod position indbetor in:trument accuracy of : 12 steps. Companson of the demand position indicssors to the bank insertion limits with venficat6on of rod position by the individual rod position indicators (after thermal soak following rod motion below 50 percent of RATED THERMAL POWER) is sufficient venfication that the control rods are above the insertion limits. The control bank FULLY WITHORAWN position can be varied within the interval of 225 to 229 steps withdrawn, inclusive. This interval pemits periodic repositioning of the parked RCCAs to minimize wear, while having minimal impact on the normal reload core physics and safety evaluations. Changes of the RCCA FULLY WITHDRAWN position within this band are administratively controlled, using the rod insertion limit operator curve. NORTH ANNA-UMT 1 8 3/4 14 Amendment No. 75.M.738. Y4 e

[ T4 4 REACTMTV CONTAOL SYSTEMS t i_ 3 /4. 4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation and maintain the DNBR above the cesign hmit 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 hour. 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 cociant 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 tne 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 hours, 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 Nv heat removalis provided as long as the reactor has been shutdown for at least 100 hours 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 t requirement to maintain a 3000 gpm flow rate provides sufficient coolant circulation to minimize the effect of a boron dilution incident and to prevent boron stratification. The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 316*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 staring of the RCPs to when the secondary water temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures. The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will therefore be within the capability of operator recognition and control. The requirement to maintain the boron concentration of an isolated loop greater than or equal to the boron concentration of the operating loops ensures that no reactivity addition to the core could occur during stattup of an isolated loop. Verification of the boron concentration in an 4 idle loop prior to opening the cold leg stop valve provides a reassurance of the adequacy of the boron concentration in the isolated loop. Operating the isolated loop on recirculating flow for at least 90 minutes prior to opening its cold leg stop valve ensures adequate mixing of the cooiant in this loop and prevents any reactivity effects due to boron concentration stratification. j Startup of an idle loop will inject cool water from the loop into the core. The reactivity transient resulting from this cool water injection is minimized by delaying isolated loop startup until its temperature is NORTH ANNA-UNIT 1 B 3/4 4-1 Amendment NoA6,42,447,W,449, 170 l

t i k i .l3/4.4 REACTOR COOLANT SYSTEM BASES k within 20*f of the operating loops. Making the reactor subtritical prior I to loop startup prevents any power spike which could result from this cool I water induced reactivity transient. 3/4.4.2 AND 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized'above its Safety Li nit 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 j to relieve any overpressure concition which could occur during hot shutdown. In the event that ne safety valves are OPERABLE, an operiting RHR loop, connected :o the RCS, or the power operated relief valves-(PORVs) will provide l overpressure relief capability and will prev:nt 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 of load assuming no reactor trip until the first Reactor Protection System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves. Demonstration of the safety valves' lift settings will occur only f ~ during shutdown and will be performed in accordance with the provisions i of Section XI of the ASME Boiler and Pressure Code. i The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and inclue ng the design step load decrease with steam dump. Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer. : ode safety valves. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable. 3/4.4.4 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12 hour periodic surveillance is suf-ficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also i ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. i NORTH ANNA - UNIT 1 B 3/4 4-2 Amendment No. U,141, i NOV 1 890

i I i i I oEACTOR COOLANT SYSTEM BASES 3/4.4.7 CHEu!STRY t The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the e ootential for Reactor Coolant System leaka" or failure due to. stress ' corrosion. Maintaining the chemistry within the Steady. State Limits i provides adecuate corrosion protection to ensure the structural in-tegrity of the 3eactor Coolant System over the life of the plant. rhe associated effects of exceeding the oxygen, chloride and fluoride limits .i are time and te oerature deoendent. Corrosion studies show that opera-l tion may te continued with contaminant concentration levels in excess of 'l the Steady State Limits up to the Transient Limits, for the specified limited tire intervals without having a significant effect on the struc-tural integrity of the Reactor Coolant System. The time i_nterval per-mitting crntinued coeration within the restrictions of the Transient Limits pr ides time for taking corrective actions to restore the con-taminant concentratiens to within the' Steady State Limits. j The surveillance requirements provide adecuate assurance that con-centrations in excess of the limits will be detected in sufficient time i to take corrective action. 3/a.4.3 SPECIFIC ACTIVITY The limitations on the soecific activity of the primary coolant ensure that the resulting 2 hour doses at the site boundary will not exceed an aopropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secor.dary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC'of typical site-locations. These values are conservative in that specific site parameters of the North Anna site such as site boundary location and meteorological conditions, were not considered in this evaluation. The NRC is finalizing site specific criteria wnich will be used as the basis for the reevaluation of the specific activity limits of this site. .i This reevaluation may result in higher limits. i i 1 NORTH ANNA - UNIT 1 B 3/4 4-5

t i REACTMTV CONTROL SYSTEMS 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 EOUlVALENT l-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine phenomenon which may occur following changes in THERMAL POWER. l 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 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 i 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 of isotopic analyses following power changes may be permissible if justified by the data obtained. 3/4.4 9 PAESSUAE/TEMPEAATURE LIMITS 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 j 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 defired. Consequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis. I t NORTH ANNA - UNIT 1 B 3/4 4-6 Amendment No.g44-7,

PEACTivrTV CON 70L SYSTMS .N The heatup limit curve, Figure 3.4-2. is a composite curve which was prepared by l determining tne 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 are l composite curves which were prepared based upon the same type analysis with the exception j 1 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 are prepared based upon the most limiting value of the i predicted adjusted reference temperature at the end of 12 EFPY. The adjusted reference l temperature was calculated using results from a capsule removed after the sixth fuel cycle. l The results are documented in Westinghouse Report WCAP-11777, February 1988 and Babcock and Wilcox Report BAW-2146, October,1991. The reactor vessel materials have been tested to determine their initial RTNDT. The results of these tests are shown in the UFSAR and WCAP-11777. Reactor operation and resultant fast neutron (E>1 Mev) irradiation will cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence and copper content of the material in 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 12 EFPY. The reactor vessel beltline I region material properties are listed in Table B.3.4-1. The actual shift in RTNDT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-70, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cocidown curves must be recalculated when the ARTNDT determined from the surveillance capsule is different from the calculated ARTNDT for the equivalent capsule radiation exposure. The pressure-temperature limit lines shown on Figure 3.4-2 for inservice leak and l i hydrostatic testing have been provided to assure compliance with the minir7m tamperature requirements of Appendix G to 10 CFR 50. The minimum temperature for criticality specified in T.S. 3.1.15 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 and WCAP-11777 to assure compliance with the requirements of Appendix H to 10 CFR Par 150. The limitations imposed on pressurizer heatup and cooldown and spray water temperature differentia! 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. P NORTH ANNA-UNIT 1 B 3/4 4-7 Amendment NcA+7,170

r ICEE CTIVITY CONTROL SYSTP 4S i EMSES The OPERABILITY of two PORVs or an RCS vent opening of greater than 2.07 square l 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 270*F. Either PORV has adequate relieving capability to protect the RCS l from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less 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. Automatic or passive low temperature overpressure protection (LTOP) is required i whenever any RCS cold leg temperature is less than 270*F. This temperature is the water l temperature corresponding to a metal temperature of at least the limiting RTNDT+90*F+ instrument uncertainty. Above 270*F administrative control is 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. i Table B.3.4-1 I MATEAfAL PAOPERTY BASIS Controlling Material: Lower Shell Plate Forging Copper Content: 0.15 W1.% Nickel Content: 0.80 Wt.% initial RTndt: 38"F RTndt After 12 EFPY: 1/4 T, 145'F 3/4T, 122*F Cooldown Rate- < 10 0' F/ H r. Heatup Rate- < 6 0

F/H r.

4 NORTH ANNA - UNIT 1 B 3/4 4 8 Amendment NoA4.W.170

3/1.5 EuER3ENCY : E C:0L**3 Sv5'EuS ? 5ES 3/a.5.1 AC:UMLLATCRS t The CPE: ABILITY of each RCS accumulator ensures that a sufficient volume of bcratec water will be irrediately forced into tne reactor core through eacn of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provices tre initial cooling mecnanism caring large RC5 pipe ruptures. The limits en accumulatcr volume, coran concentration and pressure ensure tnat tne assumptions usec for accumulatcr injection in the safety analysis are ret. The accumulator ocwer acerated isolation valves are considered to be "cserating bypasses' in :ne centext of IEEE Std. 279-1971, wnicn recuires that cypasses of a protective function ce removed automatically whenever cermissive ccncitions are not met. In addition, as these accumulater isolation valves fail to meet single failure criteria, removal of power to tne valves is recaired. The limits for operation with an accumulater inoperable for any reason excect an isolation valve closed minimi:es the tire expcsure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator wnich may result in unacceptable peak cladding temperatures. i !f a closed isolation valve cannot te im eciately Opened, the full cacacility of one accumulatcr is not available and promet action is recuired to place the reactor in a mode where this capability.is not recuired. 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that suf-ficient emergency core cooling capability will be availacle in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Eitner subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cocling to limit the peak cladding teinperatures within acceptable limits for all post-ulated treak sizes ranging from the double ended break of' the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long term core cooling capability in the recirculatien mode during the accident recovery period. NCRT" cNNA - UNIT 1 3 3/4 5-1

~ ~, =; _ } j

l REACTMTV CONTAOL SNMS N

IMSES ECCS SUBSYSTEMS (Continued) ,fj With the RCS temperature below' 350*F, one OPERABLE'ECCS subs) stem is acceptable - 4 without single Tailure consideration on the basis of the stable reactivity condition of the j reactor and the limited core cooling requirements. j 1 The limitation for a maximum of one centrifugal charging pump and one low head safety l injection pump to be OPERABLE and the Surveillance Requirement to verify all charging Li pumps cnd low head safety injection pumps except the required OPERABLE pump to be inoperable below 316*F provides assurance that a mass addition pressure transient can be l 'l relieved by the operation of a single PORV. s The Surveillance Requirements provided to ensure OPERABILITY cf each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. 314 5 4 BORON INJECTION SYSTEM The OPERABILITY of the boron injection system as part of the ECCS onsures that sufficient f negative reactivity is injected into the core to counteract any positive increase in reactivity i caused by RCS system cooldown. RCS cooldown can be caused by inadvertent .j depressurization, a loss-of-coolant accident or a steam line rupture. .j The limits on injection tank minimum contained volume and boron concentration ensure that .. I the assumptions used in the steam line break analysis are met. I 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 E solubility limit of 111*F at 15,750 ppm boron. 1 1 } I I f 9 -i-! I NORTH ANNA - UNIT 1 B 3/4 5-2 Amendment No.-46,6 8,W, } 170 '[ v

3'**H l a 7". UNITED STATES ((k/J. j } NUCLEAR REGULATORY COMMISSION gfyg j WASWNGTON, O C. 20555M r %,....s VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-339 NORTH ANNA POWER STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 149 License No..NPF-7 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated December 27, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as i amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of _the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. t

i I 2. Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-7 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 149, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented by August 1, 1993, or 10 effective full power years, l whichever occurs first. FOR THE NUCLEAR REGULATORY COMMISSION D ( ( Her er N. Berkow, irector Project Directorate 11-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 25, 1993 s

i i ATTACHMENT TO LICENSE AMENDMENT NO. -149 i TO FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-339 Replace the fcilowing pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by i amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document conipleteness. Remove Paaes Insert Paaes 3/4 1-9 3/4 1-9 3/4 1-12 3/4 1-12 3/4 4-2 3/4 4-2 3/4 4-3 3/4 4-3 3/4 4-26 3/4 4-26 3/4 4-27 3/4 4-27 i 3/4 4-28 3/4 4-28 1 3/4 4-30 3/4 4-30 3/4 5-3 3/4 5-3 i 3/4 5-6 3/4 5-6. 3/4 5-7 3/4 5-7 B 3/4 1-3 B 3/4 1-3 B 3/4 4-1 B 3/4 4-1 i B 3/4.4-6 B 3/4 4-6 B 3/4 4-7 B 3/4 4-7 8 3/4 4-B B'3/4 4-8 i B 3/4 4-9 thru B s/4 4-16 B 3/4 4-9 B 3/4 5-2 8 3/4 5-2 h h

f REACTMTY CONTROL SYSTA/S i FLOW PATHS - OPERATING UMITiNG CONDITION FOR OPERATION 3.1. 2. P. At least two of the following three 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.

b. Two flow paths from the refueling water storage tank via charging pumps l

to the Reactor Coolant System. l APPL!CABlUTY: MODES 1, 2, 3# AND 4# 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 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1.77% detta k/k at 200*F within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEldANCE 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.

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 358'F. l NORTH ANNA - UNIT 2 3/4 1-9 Amendment No.-64,+Gs,149

3Ea:~:,':'" '.D. sis ~EWS _ SURVEILLANCE REQUIREMENTS (Continued) At least once per 31 days by verifying that each valve (manual, b. power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. At least once per 18 months during shutdown by verifying that each c. automatic valve in the flow patn actuates to its correct position on a safety injecticn test signal. AF NORTH ANN: - UNIT 2 3/4 1-10

EEACTIV'TY CONT?OL SYSTEMS CHAPGING PUMP - SHUTDCWN LIMITING CONDITION FOR OPERATION i 3.1. 2. 3 One charging pump in the boron injection flow path required by i Ipecification 3.1.2.1 shall be OPERABLE. APPLIC'BILITY: MODES 5 and 6. i ACTION: With no charging pump OPERABLE, suspend all operations -involving CORE a. 1 nLTERATIONS or positive reactivity changes until one charging pump is restored to OPERABLE status. t b. With no charging pump OPERABLE and the opposite unit in MODE 1, 2, 3 or 4, immediately initiate corrective action to restore at least one charging pump to OPERABLE status as soon as possible. i SURVEILLANCE REQUIREMENTS i 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying that, on recirculation flow, the pump develops a discharge oressure of.-eater than or equal to 2410 psig when tested pursuant to l 5peci fication 4.0.5. 4.1.2.3.2 All charging pumps, except the above required OPERABLE pump, shall be demonstrated inoperable at least once per 12 hours b; verifying that the control switch is in the pull to lock position. ,4 t i NORTH ANNA - UNIT 2 3/4 1-11 Amendment No. 5 i MAE l' %21 k 5 i

REACTMTVCOhlTAOL SYSTEMS CHARG!NG PUMPS. OPERAT!NG UMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps sha!! be OPERABLE. APPL!CABILITY: MODES 1, 2, 3# and 4. ACTION: With only one charging pump OPERABLE, restore a second charging pump to OPERABLE status within 72 hours 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 hours: restore a second charging pump to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. The provisions of Specification 3.0.4 are not applicable for one hour following heatup above 358'F or prior to cooldown below 358'F. l SURVE!LLANCE REOUIREMENTS 4.1.2.4.1 The above required 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 required OPERABLE pump, shall be demonstrated inoperable at least once per 12 hours whenever the temperature of one or more of the RCS cold legs is less than or equal to 358'F by verifying that l the control switch is in the pull to lock position. 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 358",F. l NORTH ANNA - UNIT 2 3/4 1-12 Amendment No.149

i i i i 3/4.4 REACTOR CCCLANT SYSTEM 3/4.4.1 REACTOR CCOLANT LOCPS AND COCLANT CIRCULATICN STARTUP AND POWER OPERATION LIMITINC CONDITION FOR OPERATION

3. 4.1.1 All reactor coolant loops shall be in operation with power removed from the loop stop valve operators.

APPLICABILITY: MODES 1 and 2.* ACTION: With less than the above required reactor coolant loops in operation, be in at i least HOT STANDBY within 1 hour. i f i i i SURVEILLANCE REOUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours.

4. 4.1. 2 At least once per 31 days, with the reactor coolant loops in operation by verifying that the power is removed from the loop stop valve operators.

l

See Special Test Exception 3.10.4.

NCRTH ANNA - UNIT 2 3/44-1 e . = = = e

REACT ^ ^tY1At# SPA HDT STi E UMITING CONDITION FOR OPERATION 3.4.1.2 a. At least two of the reactor coolant loops listed below shall be OPERABLE:

1. Reactor 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 Loop C and its associated steam generator and reactor coolant pump, l

l b. At least one of the above coolant loops shall be in operation. APPUCABILITY: MODE 3 ACTIOrt a. With less than the above required loops OPERABLE, restore the required loops to OPERABLE status with in 72 hours or be in HOT SHUTDOWN within the next 12 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 hours. A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 358'F unless the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures. All reactor coolant pumps may be de-energized for up to 1 hour provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature. The requirement to have one coolant loop in operation is exempted during the performance of the boron mixing tests as st:pulated in License Condition 2.C(15)(f) and 2.C(20)(b). NORTH ANNA - UNIT 2 3/4 4-2 Amendment No.-M, %9

R=ACTCP COCUNT SVSTDA SMf1TCONN I UMmNG CONDITION FOR OPERATION i 5 3.4.1.3 a. At least two of the coolant loops sisted below shall be OPERABLE:

1. Reactor Coolant Loop A and its associated steam generator and reactor I

coolant pump,

2. Reactor Coolant Loop B and its associated steam generator and reactor

{ coolant pump, i

3. Reactor Coolant Loop C and its associated steam generator and reactor f

coolant pump, l 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.

I i APPL!CA9iL!TY: 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 t

possible; be in COLD SHUTDOWN within 20 hours. l i

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 I

operation. ~ A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 358'F unless the secondary water temperature of l each steam generator is less than 50*r above each of the RCS cold leg temperatures. The offsite or emergency power source may be inoperable in MODE 5. All reactor coolant pumps and residual heat removal pumps may be de-energized for up to 1 hour provided 1) no operations are permitted that would cause dilution of the j reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10*F below saturation temperature, t i 4 i NORTH ANNA-UNIT 2 3/4 4-3 Amendment No.149

i N g -n._q - g ~ 0 250 3 \\ 1 H ~ \\ i _5 A' k .g i -J \\ y s p x X e ,i4 ,ii, h \\\\-

UNACCEPTABLE O

.x ,.i! OPERATION x i O s i ua N: ~ .i, m w 1 \\

150 z

e x J 1 o 1-,,i i, . i i o ,3 U s,,, .,ii \\,, ..s. ,,,4 e ,,s, , i, i,is - 100 i+,, i X c i t m . i., s i p iie 6 i, 6, '\\ ' n ..ii , ji a i 1 i i.

=

i i i > i i i 2 i\\ a , is. y ICC'ENd8LE_. l !\\ i OPERATION g ,i .,i. i. i i i. g ,, i, i. g ) a ,4 1 i, j 4 e i I e , i t i, 8 '.8 i [ iiii ii,i ,e ii 4 e i iit ; i.. + i 8. i 20 30 40 50 80 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Act.ivity >1.0 Ci/ gram Dose Equivalent 1131 i NORTH ANNA - UNIT 2 3/4 4-25

.~ REACTOR COO 3M SYSimA 3/4.4 9 PAESSUAETEVpERATURE UMITS REACTOR COOLtNT SYST9A UMITitjG CONDITlON FOR OPERATIOBJ __ _ _ _ _ 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown in Figures 3.4-2 and 3.4-3 during heatup, cooldown, and inservice leak and hydrostatic testing with: l

a. A maximum heatup of 60*F in any one hour period.

b. A maximum cooldown of 100*F in any one hour period.

c. A maximum temperature change of less than or equal to 10*F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

AoPLICABILITY: At all times. ACTIOtj With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours and reduce the RCS Tavg and pressure to less than 200*F and 500 psig, respectively, within the following 30 hours. SURVEILLANCE REOUiREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50, Appendix H. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3. NORTH ANNA UNIT 2 3/4 4-26 Amendment No.4g,149

Figure 3.4-2 Unit 2 RCS HEATUP P/T Lirnits Valid to 17 EFPY Heatup Rates: 0 60*F/Hr. (Margins for instrument Errors NOT included) 2750 .'.' i - - b--4-'- MATERIAL PROPERTY BASIS r.

Controlling Material

. :. r.. - , Lower Shell Plate Forging p.. p....... e 2500 : Copper Content: 0.13 W1.% -a. Lenk Test .1... .-Nicket Content: 0.83 W1.% Limit ',1,- ; '

initial RTndt: 56'F R

RTndt At:er 17 EFPY: ' - - - - } - - -l ' - f - - j " - 2250 7 C 1/4T.196'F 's/4T,172*F . '..L. ....L...t.... S t v,Heatup Rate s60*F/Hr. ..r-- -~ ~- ). P r'.. ). . l.. j.. ...,i..,... i.. ...,. L.. ..i.,... 1... L.. . 1... L.. r T.. 1 - - - 1.- *. ;i- - i -* t-i i .. ;... #.. q ..... p.' -..... r -'. e p. S 1750 p. q..L.. !.. q., ... ;. J...L ..j..J...... 3 p. 3......t. 3.....l..q... u p..a.. .[..a... -}.~............ 4... I-J --'-f-J- i--d-'- -l- - '-i-- r 1500 e L.. . ;. [ -.. :-.. ...L.. L.. h.. f-UNACCEPTABLE L-- (..J.. (.. s e..J.. OPERATION F-. 1250 P p..q.,......J.,.............-....p. ,,...q.,. S e.. .e. .d...... g .q........ .l..... .....[.. 9 1000 ..J..L.....[,....... 2...;.. ... q...l... }.. J.1. ..j...... .J... ....]..... L 7.L. 7.... . :...p. + i .....,..,....... 3.. 7..... ~. 750 ...)... ..J... . i..ACCEPTABLE .~..... ..q.,..s. .....q... .3a.. .. ~u .s. OPERATION 500 ..a.;. . :... ;. L.:... :.. ..J... .s. ... q... .s. ..q.......,.... :...:... :.. 3 250 80 100 120 140 160 180 200 220 240 260 280 300 320 340 360 Minimum RCS Cold Leg Temperature - degrees F NORTH ANNA-UNIT 2 3/4 4 27 Amendment No.-60,149

A a 5 6 i ( t 1 +a Figure 3.4-3 Unit 2 RCS COOLDOWN P/T Limits Valid to 17 EFPY Cooldown Rates: 0100*F/Hr. (Margins for Instrument Errors NOT included) !.t t 2750 i f 4 f .."p..- * ^.. '1 MATERIAL PROPERTY BASIS f 7 Controlling Material: . 71".-.,- 1 ". J..,. .. r. ! - i .r l Lower Shell Plate Forging - -- - F--f.-

i h "

2500} Copper Content: 0.13 WtX ...l... 4. [.. .. s. :. r ..<.-=.q..... i Nickel Content: 0.83 W1A ~-:- r-- R 2250 - RTnct After 17 EFPY: .. 1 .- Initial RTnct: 56*F 4. ~. .- ; _ y'.. j,.. 3-. C 1/4T,196'F .. 1. 7..-..-.- - -'.:.-'".'". 4.. ~~ g . 3/4T,172'F '1--].-~~-----,***--t*)4 ..... 7-;. ..J... 2000. Coolcown Rate: --t-s100'F/Hr. P .,....,.3.... .,.......r..,...,...... ...p.. ..i..i...:. ...g.. .....3..i...,,....,i.,. ....i...,... ..i. L.....J.... .i ...,.. $...i... I...........q......2.L..L..'.....J... r i 4...- .......i... i .,4...... e 1750,' i 4... i S W-1 -. J,...... ...L.. .*J.*. s r - ~ ' ' j y-i -

i' - -

3 . i s - '..r ...a... t.. l.. a..,. u 1500 UNACCEPTABLE .....t. [..,.. v. 1 OPERATION 4 -"f-f#n + -+- i r e i V... t.-..". r. 3..,. .,...". ".. t. t.:. r,.."" 1 ". 7 7 7 3..,. 1250 5,..>..q..>.;. ... t'.. ..J....;. 4 ..J....;. l .....4.,... p ...t,. 2,..J... ..J....:. t 3..,.........;.....3..;...... ........ 3. S 1000

... [>... J. a...

...... [.. ......q..'....L.:...4.J..a... ..a..J. ... L.. t... i p......J... ..j..J... 9 [... ....,. J... ;... 4 .. j. :..... 750 .r-- d------ ----'..'-:- i- - ]..- ......- f....- '....'"J..i - t ..[.. L. ' F/hr .. q... i ,...z. 2. y o. 3. 3.........y....... 500 -- - i. 1. >J... p. . 7g..,... +.. 4.. L. i ,o. ,o. ACCEPTABLE + p.50o. . q....... OPERATION L. 4,....s1......... .m L..,.. ............[........... :. 3. ;... ;..,., ;.. 7.. .c. i 0 80 100 120 140 160 180 200 220 240 260 280 300 320 340 360 .,t Minimum RCS Cold Leg Temperature - degrees F I NORTH ANNA - UNIT 2 3/4 4-28 Amendment No.40.149 i i

t ~ REACTOR COOLANT SYSTEM PRESSURIZER I i LIMITING CONDITION FOR OPERATION f 3.4.9.2 The pressurizer temperature shall be limited to: A maximum heatup of 100'F or cooldown of 200'F, in any one hour a. period, and b. A maximum spray water temperature and pressurizer temperature differential of 320'F. l APPLICABILITY: At all times. ACTION: With the pressurizer temperature limits in excess of any of the above limits. restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer i remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig within the following 30 hours. 4 i SURVEILLANCE REQUIREMENTS L 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The f spray water temperature differential shalf be determined to be within the limit at least once per 12 hours during auxiliary spray operation. l i l i i 'l I r f NORTH ANNA - UNIT 2 3/4 4-29 I i i

REACTOR CODLANT SYSTRA ERPRESSUPE PROT:CTION SYSTEMS UMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE: a. Two power operated relief valves (PORVs) with a lift setting of: 1) less than or equal to 510 psig whenever any RCS cold leg temperature is less than or equal to 321*F, and 2) less than or equal to 360 psig whenever any RCS cold leg temperature is less than 210*F, or

b. A reactor coolant system vent of greater than or equal to 2.07 square inches.

APPUCA91UTY: When the temperature of one or more of the RCS colo legs is less than or equal to 321*F, except when the reactor vessel head is removed. l ACTON: i

a. With one PORV inoperable, either restore the inoperable PORV to OPERABLE i

status within 7 days or depressurize and vent the RCS through 2.07 square inch vent (s) within the next A hours; maintain the RCS in a vented condition l until both PORVs have been restored to OPERABLE status.

b. With both PORVs inoperable, depressurize and vent the RCS through a 2.07 square inch vent (s) within 8 hours; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status.

c. In the event either the PORVs or the RCS vent (s) are used to mitigate a 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. d. The provisions of Specification 3.0.4 are not applicable. t r i NORTH ANNA - UNIT 2 3/4 4-30 Amendment NoAO,149

EV RGENCYCOAECOOUNGSYSTPAS ECCS SUBSYSTEMS Tm GREATER TWAN 950*F UMITING CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE centrifugal charging pump #,

l

b. One OPERABLE low head safety injection pump #,

l

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.

APPUCABIUTY: MODES 1,2 and 3. ACTION: a With one ECCS subsystem inoperab!e. restore the inoperable subsystem to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours. b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shal' 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. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

c. The provisions of Specification 3.0.4 are not applicable to Specifications 3.5.2.a and 3.5.2.b for one hour following heatup above 358'F cr prior to cooldown below 358'F.

SURVEILLANCE REOUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At least once per 12 hours by verifying that the following valves are in the indicated positions with power to the valve operators removed:

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 358'F. NORTH ANNA - UNIT 2 3/4 5-3 Amendment No.149

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) Valve Number Valve Function Valve Position

a. MOV-2890A

a. LHSI to hot leg

a. closed

b. MOV-2890B

b. LHSI to hot leg

b. closed

c. MOV-2836

c. Ch pump to cold leg

c. closed

d. MOV-2869A

d. Ch pump to hot leg

d. closed

e. MOV-2869B

e. Ch pump to hot leg

e. Closed b.

At least once per 31 days by verifying that each valve (manual, t power operated or automatic) in the flow path that is not locked, .ealed, or otherwise secured in position, is in its correct position. By a visual inspection which verifies that no loose debris (rags, c. trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed: 1. For all accessible areas of the containment prior to establish-ing CONTAINMENT INTEGRITY, and 2. Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established. d. At least once per 18 months by: 1. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no 4 evidence of structural distress or corrosion. e. At least once per 18 months, during shutdown, by: 1. Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal. 2. Verifying that each of the following pumps start automatically upon receipt of a safety injection test signal: a) Centrifugal charging pump, and b) Low head safety injection pump. NORTH ANNA - UNIT 2 3/4 5-4

j ~ 1 { I r EMERGENCY CORE COOLING SYSTEMS I 1 SURVEILLANCE' REQUIREMENTS (Continued) I l f. By verifying that each of the following pumps' develop the indicated discharge pressure (after subtracting suction pressure) on recircula-j tion flow when tested pursuant to Specification 4.0.5. i 1. Centrifugal charging pump greater than or equal to 2410 psig.- { 2. Low head safety injection pump greater than or. equal to 156' psig l g. By verifying that the following manual valves requiring adjustment. to prevent pump " runout" and subsequent component damage are' locked and tagged in the proper position for injection: 1. Within 4 hours following completion of any repositioning or maintenance on the valve wMn the ECCS subsystems.are required 1 to be CPERACLE. i 2. At least once per 18 months. i 1. 2-51-89 Loop A Cold Leg ~ 2. 2-51-97 Loop B Cold Leg 3. 2-51-103.. Loop C Cold Leg i 4, 2-51-116 Loop A Hot Leg. 5. 2-51-111 Loop 8_. Hot Leg i 6. 2-51-123 Loop C Hot Leg h. By perfonsing a flow balance test, during shutdown, following comp' etion. l of modifications to the ECCS subsystems that alter the subsystem. l flow characteristics and verifying that:. i 1. For high head safety injection lines, with a single pump running: j a) The sum of the. injection line flow rates, excluding the I highest flow rate, is > 384 gps, and

l b)

The total pump flow rate is < 650 gps. j = NORTH ANNA - UNIT 2 .3/4 5-5

i 5 EVERGCNCY COAE CMJ/3 SYSTAAS ECCS SU9 SYSTEMS - Tavo LESS THAN 350*C UMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABl.E: a. One OPERABLE centrifugal charging pump #,

b. One OPERABLE low head safety injection pump #, and An OPERABLE flow path capaole of automatically transferring fluid to the c.

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. APPUCAB'UTY: MODE 4. ACTION: With no ECCS subsystem OPERABLE because of the inoperability of either the a. centnfugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next 20 hours. i

b. With no ECCS subsystem OPERABLE because of the inoperability of the low head safety injection pump, restore least one ECCS subsystem to OPERABLE status or maintain the Reactor Cc-1t System Tavg less than 350*F by use of alternate heat removal methods.

c. In the event the ECCS is actuate id injects water inte the Reacte Coolant System, a Special Report shall be epared and submitted to the Ccmmission pursuant to Specification 6.9.2 within 90 days desenbing the circumstances 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 Special Report whenever its value exceeds 0.70. i A maximum of one centrifugal charging pump and one low head safety injection pump l shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 358'F. l ) I NORTH ANNA - UNIT 2 3/4 5-6 Amendment No.44,149 i l f

,= -... -. _ 7 I i i EMMEC(COPE ODCUJG SYSTB*S ? i SURVEluRJCE REOU1REMENTS t 4.5.3.1 The UCCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2. i 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 hours whenever the temperature of one or more of the RCS cold legs is less than'or equal to 3FB'F by verifying that the control switch is in the pull to lock position. l l i i I i I i V l q -) e i l l i NORTH ANNA-UNIT 2 - 3/4 5-7 Amendment No.149 j i

EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 BORON INJECTION SYSTEM BORON INJECTION TANK LIMITING CONDITION FOR OPERATION 3.5.4.1 The boron injection tank shall be OPERABLE with: A contained borated water volume of at least 900 a.

gallons, b.

Between 12,950 and 15,750 ppm of boron, and g c. A minimum solution temperature of Il5'F. APPLICABILITY: MODES 1, 2 and 3. ACTION: With the boren injection tank inoperable, restore the tank to OPERABLE status within 1 hour or be in HOT STANDBY and borated to a SHUTDOWN PARGIN equivalent e to 1.77%.tk/k at 200*F within the next 6 hours; restore the tank to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMEfGS l 4.5.4.1 The boron injection tank shall be demonstrated OPERABLE by: Verifying the contained borated water volume at least once per 7 a.

days,

.{ b. Verifying the boron concentration of the water in the tank at least once per 7 days, and Verifying the water temperature at least once per 24 hours. c. { i NORTH ANNA - UNIT 2 3/4 5-8 Amendment No. 54 i

REACTMP/ CONTROL SYST VS EMSES 3/41.2 90 RATION SYSTEMS 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 MARGIN 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 refueling water storage tank. With the RCS temperatu:e below 200'F, one injection system is acceptable without single failure consideration on the basis of the stabic reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive react;vity 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 Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 358'F provides assurance that a mass addition pressure transient l can be relieved 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 refueling water storage tank. i t 2 NORTH ANNA - UNIT 2 B 3/4 1-3 Amendment No.-64.78,149 i

A L a a i OE AC'" "Y O ~~rt Sv5?E"5 l FASE5 l 3/4.1.2 PODATION SYSTE"5 (Continvec) Tre contained water volume limits 1r.clude allcwance for w6'er not available. because of discharge lire iccaticn and other physical charscteristics. The OPERABILITY ; of ore baron injection system during PTFUELING insures that this system is available for i reactivity control while in ft0DE 6. i The lin.its on containec water solume and boron concentration nf the RWST also ensure a pH vdlue of between 7.7 and 9.C for the solution recirculated within the-containment after a LOCA. This pH minimizes the evolution of looir.e and minirizes the ef fect of chloride ard caustic stress corrosion on mechanical systeras and components.j At least one charging pump riust remain cperable at all times when the opposite unit, is in MODE 1, 2, 3, or 4 This is required to maintain the charging pump cross-connect system operatioral. j 3/4.1.3 M0VAktE C0t: TROL ASSE!'BLIES I The specifications of this section (1) ersure that acceptable power distribution i limits are maintained, (2) ensure that tre minimum SHUTDOWN MARGIf1 is maintained, and (3) limit the potential effects of rod misalignment on associated accident analyses. OPERABILITY of the movable control assemblies is established by observing rod motion and determining that rods are positioned within + 12 steps (indicated position) of the respective derand step counter pcsition. The OPERABILITY of the individual rod position indication system is established by appropriate pericalc CHANNEL CHECKS,- CFANNEL FUNCTIONAL TESTS, and CFANNEL CALIBRATIONS. OPERABILITY of the individual rod position indicators is required to determine control rod position and thereby ensure compliante with the control rcd alignment and insertion limits. The OPERABLE tonoltlon for the indisittal rod position indicators is defined as being cepable of incicating rod position within + 12 steps of the associated demand position indicator. For power levels below 50 percent of RATED THERMAL POWER, the specifications of this section permit a maximum cr.e hour stabilization in every 24 period (thermal " soak i tire") to allow sieb1112ation of known therral drift in the individual' rod position inoicator channels during which time the indicated rod position may vary from demand pcsition indication by no more than + 24 steps. This "1 in 24" feature is an upper limit on the frequency of thermal soak allowances and is available for aoth a continuous one bour period or one consisting of several discrete intervals.. During j this stabilization period, greater reliance is placec upon the demand position indicators to determine rod position. In addition, the + 24 step / hour limit is-not applicable when the' control rco position is known to be greater than 12 steps from:, the rod group step counter demand positien indication. t.bove 50 percent of RATED j THERMAL POWER, rod motion is not expected to induce thermal transients of sufficient magnitude to exceed the individual rod pcsition indicator instrument accuracy of + 12 steps. ' Comparison of the demand position indicators to the bank insertion lib.11s '! with verification of rod position by the individual rcc'pcsition indicatcrslafter thermal soak following rod motion below 50 percent of RATED THERMAL POWER)'is sufficient verification that the control rods are above the insertion limits ~ The control bank FULLY WITHDRAWN position can be. varied within the-interval of 225 to 229 steps withorawn, inclusive. This interval permits periodic repositioning. of tFe parked RCCAs to minimize wear, while having minimal impact on the normal l reload core physics and safety evaluations. Changes of the RCCA FULLY FITHDRAWN i position within this band are acministratively controlled, using the rod insertion l limit creratcr curve. j NCRTH AN!!A - Ui.IT 2 B 3/4 1-4 Amendment No. I2I,I29, 133 ;

I 14.4 PCACTIVITY CONTROL SYSTCMS i EMSES i 3/4.4.1 REACTOR COOLANTLOOPS The y W 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 hour. 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 t OPEf%BLE. -{ in MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat i 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. .j After the reactor has shutdown and entered into MODE 3 for at least 100 hours, a minimum i 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 - l power production rate decreases with time after reactor shutdown, the requirements for RHR j system decay heat removal also decrease. Adequate decay heat removalis provided as long as the i reactor has been shutdown for at least 100 hnuts after entry into MODE 3 and RHR flow is I 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. l During a reduction in reactor coolant system boron concentration the Specification 3.1.1.3.1 } requirement to maintain a 3000 gpm flow rate provides sufficient coolant circulation to minimize the 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 E or equal to 358'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 staring of the RCPs to when the secondary water temprature of each steam generator is less than 50*F above each of the RCS cold leg temperatures. [ The requirement to maintain the boron concentration of an isolated loop greater than or l equa' 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

i idle loop prior to opening the cold leg stop valve provides a reassurance of the adequacy of the

} boron concentration in the isolated loop. Operating the isolated loop on recirculating flow for at ) least 90 minutes prior to opening its cold leg stop valve ensures adequate mixing of the coolant l in this loop and prevents any reactivity effects due to boron concentration stratification. j Startup _of an idle loop will inject cool water from the loop into the core. The reactivity [ f transient resulting from this cool water injection is minimized by' delaying isolated loop startup until its temperature is within 20*F of the operating loops. Making the reactor suberitical j prior to loop startup prevents any power spike which could result from this cool water induced i reactivity transient. NORTH ANNA - UNIT 2 B 3/4 4 1 Amendment No.-4GO,442,149 ' f i

F ACTOACOOLA?FSM1 E%SES 14 4 2 A ND l'4 d j SA:m AND :EL':: VALVES The pressurizer code safety valves ocerate to prevent the RCS from being pressunzed above its Safety Limit of 2735 psig. Each safety valve is designec to relieve 380,000 lbs per hour of saturated steam at the valve set point. The relief capacity of a single safety valve is adecuate 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 pressunzed above its safety limit of 2735 psig. The combined relief capacity of I all of inese valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protection System inp set point is reached I (i.e.. ro 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 performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Ccxn. The power operated relief valves (PORVs) 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. ?444 c:ESSUAIZER 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-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expe6ed transient eceration. The maximum water volume also ensures that a steam bubble is formed and thus the ACS is not a hydraulically solid system. The requirement that a minimum number of pressunzer heaters be OPERABLE ensures that the plant will be able to establish natural circulation. F NCRTH ANNA - UNIT 2 B 3/4 42 Amendment No. 124, fini 1 30

i ~ i l DEaCTCR ;0LANT SYSTEM j l 8aSES 3/4 a. 7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion. of the Reactor Coolant System is minimized and reduces the potential for

eactor Coolant System leakage or failure due to stress corrosion.

Maintaining tne enemistry within the Steacy State Limits provides adeguate corrosion protection to ensure the structural integrity of the Reactor Coolant Systes over the life of the plant. The associated effects of exceeding the oxygen, enloride and fluoride limits are time and tescerature dependent. Corrosion studies snow that operation may be continued with centaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on tne structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the: Transient Limits provides time for taking corrective actions to restore the contaminant con-centrations to within the Steacy State Limits. The surveillance requirements provide adequata assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action. 3/4.4.8 SPECIFIC ACTIVITY The limitations on the soecific activity of the primary coolant ensure tnat the resulting 2 hour doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assmed steady state primary-to-secondary steam generator leakage rata of 1.0 GPM. The values for the limits on specific activity represent interie lief ts based upon a parametric evaluation by the NRC of typical sita locations. These values are conservative in that specific-site parameters of the North Anna site suct, as site boundary location and esteorological conditions, were not considered in this evaluation. The NRC is finalizing sita specific criteria which will be used as the basis for r the reevaluation of the specific activity limits.of this sita. This reevaluation may result in higher limits. l e i a b t NORTH A80E - UNIT 2 8 3/4 4-5

REACTtViTV CONTAOL Svcm:MS ... +..... -. - - -.... --...._--------..-.----.n.-----. s---------- . -. - - - - -. - - -... -. ~ 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 l-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine phenomenon which may occur following changes in THERMAL POWER. Reducing Tavg to less thar; 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 of isotopic analyses following power changes may be permissible if justified by the data obtained. 3'4 4 9 PRESSURE /TEMPEAATUAE UMITS r 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 vesselis treated as the goveming 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 wa'i 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 j the stress controlling location, each heatup rate of interest must be analyzed on an individual basis. I t NORTH ANNA-UNIT 2 B 3/4 4-6 Amendment No -88,149 i

RCACTM Y CONTAOL SYST ','S E%SES The heatup limit curve, Figure 3.4-2, is a composite curve which was prepared by 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 are i composite curves which were prepared based upon the same type analysis 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 are prepared based upon the mos? limitir.g value of the predicted adjusted reference temperature at the end of 17 EFPY. The adjusted reference temperature was calculated using results from a capsule removed after the sixth fuel cycle. The results are documented in Westinghouse Report WCAP-12497, January 1990 and WCAP-12503, March,1990. The reactor vessel materials have been tested to determine their initial RTNDT. The results of these tests are shown in the UFSAR and WCAP-12497. Reactor operation and resultant f ast neutron (E>1 Mev) irradiation will cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence and copper content of the material in 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 17 EFPY. The reactor vessel beltline 2 region material properties are listed in Table B.3.4-1. The actual shift in RT DT of the vessel material will be established periodically during N operation by removing and evaluating, in accordance with ASTM E185-70, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiathr. samples and vessel inside radius are essentially identical, the measured transition sha for a sample can be i applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the aRTNDT determined from the surveillance capsule is different from the calculated ARTNDT for the equivalent capsule radiation exposure. 4 The pressure-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 3 specified in T.S. 3.1.1.5 assures compliance with the criticality limits of 10 CFR 50 Appendix G. I The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in the UFSAR and WCAP-12497 to assure compliance with the requirements of Appendix H to 10 CFR Part 50. 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 2 8 3/4 4-7 Amerdment No. 149

1 i i i R:ACTMTV CONTROL SYW:YS BASES 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 i than or equal to 321'F. Either PORV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less 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. Automatic or passive low temperature overpressure protection (LTOP) is required whenever any RCS cold leg temperature is less than 321*F. This temperature is the water temperature corresponding to a metal temperature of at least the limiting RTN DT+ 90* F+ instrument uncertainty. Above 321*F administrative control is 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 tne lower end. I and administrative control at the upper end, of the Appendix G curves is further discussed in NRC Generic Letter 88-11. Table B.3.4-1 MATERIAL PROPERTY B ASIS Controlling Material: Lower Shell Plate Forging Copper Content: 0.13 Wt.% Nickel Content: 0.83 Wt.% Initial RTndt: 56*F RTndt After 17 EFPY: 1/4T, 196*F 3/4 T, 172*F Cooldown Rate: s 100

F/ H r.

Heatup Rate: 56 0

F/ H r.

1 1 3/4.4.10 STRUCTURAL INTEGRfTY 3/4.4.10 1 ASVE CODE CLASS 1. 2 and 3 COMPONENTS The inspection program for ASME Code Class 1, 2 and 3 Reactor Coolant System components ensure that the structural integrity 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 VesselCode. i r NORTH ANNA - UNIT 2 B 3/4 4 8 Amendment No.149

L i r l r i Pages B 3/4 4-9 thru 8 3/4 416 have been deleted. i t i ? ? { t .!i t i t i NORTH ANNA - UNIT 2 B 3/4 4-9 Amendment No. 149 a

3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each RCS accumulator ensures that a sufficient volume of borated water wiil be imediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met. The accumulator power operated isolation valves are considered to be " operating bypasses" in the context of IEEE Std. 279-1971 which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required. The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the-plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be imediately opened, the-full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required. 3/4.5.2 and 3/A.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that suf-ficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single. failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all post-ulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period. NORTli ANNA - UNIT 2 B 3/4 5-1

REACTIVITY CONTAOL SYST:VS l%cFS ECCS SUBSYSTEMS (ConFnuad) l l 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 358'F provides assurance that a mass addition pressure transient can be l relieved by the operation of a single PORV. ihe 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 throttle valve position stops and flow balance testing provide assurance that proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow j 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 that 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 j reactivity caused by RCS system cooldown. RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident or a steam line rupture. 1 The limits on injection tank minimum contained volume and boron concentratio'n ensure that the assumptions used in the steam line break analysis are met. The contained water j -l volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. l 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 2 B 3/4 5 2 Amendment No.64,149 i}}

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