ML20246F696

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Amend 117 to License NPF-4,revising Heatup & Cooldown Curves & Associated Low Temp Overpressurization Setpoints to Be Valid for Up to 10 EFPYs
ML20246F696
Person / Time
Site: North Anna 
Issue date: 06/30/1989
From: Berkow H
Office of Nuclear Reactor Regulation
To:
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ML20246F698 List:
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NUDOCS 8907130369
Download: ML20246F696 (34)
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{{#Wiki_filter:_ -_ _ _ __-- _-__ - _- l p nrooy\\ / UNITED STATES f g NUCLEAR REGULATORY COMMISSION r,, j WASHINGTON, D. C. 20'55 \\....*/ VIRGINIl. 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.117 License No. NPF-4 1. The Nuclear Regulatory Consission (the Comis 'on) has found that: A. The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated November 30, 1988, as supplemented June 19, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as anended (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 Comission; 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. pmum iL_____________

L , 2. Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendnent, and paragraph 2.D.(2) of Facility Operating License No. NPF-4 is hereby amended to read as follows: l-(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.117. ere hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. '3. This license amendnent is effective as of the date of issuance and shall be implemented within 30 days. FOR THE NUCLEAR REGULATORY COMMISSION H rbert N. Berkow, Directof Project Directorate II-2 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: June 30, 1989 l }

1 ATTAC%ENT TO LICENSE AMENDMENT NO.117 ) TO FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 1 Replace the following pages of the Appendix "A" Technical Specifications w',th 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. Page 3/4 1-9 3/4 1-12 3/ ' 4-3 3/4 4-27 3/4 4-28 3/4 4-31 3/4 5-3 3/4 5-6 3/4 5-6a B3/4 1-2 B3/4 1-3 B3/4 4-1 B3/4 4-6 83/4 4-7 B3/4 4-8 B3/4 4-9 83/4 4-10 B3/4 4-11 B3/4 5-2

"~ ~ I REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 Each of the following boron injection flow paths shall be OPERABLE: a. The flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System, and b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System. APPLICABILITY: MODES 1, 2, 3 and 4. 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 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 least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE: At least once per 7 days by verifying that the temperature of a. the heat traced portion of the flow path from the boric acid tanks is > 115 F.

  1. nly one boron injection flow path is required to be OPERABLE whenever 0the temperature of one or more of the RCS cold legs is less than or equal to 324 F.

NORTH ANNA-UNIT 1 3/4 1-9 Amendment t!o. 48,68, 117

r. p c l' REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. c. At least once per 18 months during shutdo, by verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal. l NORTH ANNA-UNIT 1 3/4 1-10

7> REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1. 2. 3 At least one charging pump 'a the boron injection flow path required by Specification 3.1.2.1 shan be OPERABLE. APPLICABILITY: MODES 5 and 6 ACTION: With no charging pump OPERABLE, suspend all operations involving CORE I a. ALTERATIONS or positive reactivity changes until one charging pump is restored to OPERABLE status, 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 0PERABLE status as soon as possible. SURVEILLANCE REQUIREMENTS 4.1. 2. 3.1 At least the above required charging pump shall be demonstrated OPERABLE by verifying that, on recirculation flow, the pump develops a discharge pressure of > 2410 psig when tested pursuant to Specification 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 by verifying that the switches in the Control Room have been placed in the pull to lock position. l l NORTH ANNA - UNIT 1 3/4 1-11 Amendment No. 75, 24

.A '_.C REACTIVITY CONTROL SYSTEMS l CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2. 4 At least two charging pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4*. 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% ak/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 pro-visions of Specification 3.0.4 are not applicable for one hour following heatup above 324 F or prior to cooldown below 324'F. l SURVEILLANCE REQUIREMENTS 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 > 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 324"F by verifying that the switches in the Control Room have been placed l 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 324'F. l 10RTH ANNA-UNIT 1 3/4 1-12 Amendment No. 3, yg, 117 i

~. REACTOR COOLANT SYSTEM l SHUTOOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the coolant loops listed below shall be OPERABLE: 1. Reactor Coolant Loop A and its nisociated steam generator and reactor coolant pump,* 2. Reactor Coolant Loop 8 and its associated steam generator and reactor enolant pump,* 3. Reactor Coolant Loop C and its associated stea? generator and reactor coolant pump,* 4. Residual Heat Removal Subsystem A,** 1 5. Residual Heat Removal Subsystem B.** b. At least one of the above coolant loops shall be in operation.*** APPLICABILITY: MODES 4 and 5. ACTION: With less than the above required loops OPERABLE, immediately a. initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD 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 i loop to operation. "A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 324*F unless the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures.

    • The offsite or emergency power source may be inoperable in MODE 5.
      • All reactor coolant pumps and residual heat removal pumps may be de-energized for up to 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 No. 15, 77, 117

. AC E L REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required residual heat removal loop (s) shall be determined OPERABLE per Specification 4.0.5. 4.4.1.3.2 The required reactor coolant pump (s), if not in operation, shall be determined to be OPERAELE once per 7 days by verifying correct breaker alignment and indicated power availability. 4.4.1.3.3 The required steam generator (s) shall be determined OPERABLE by' verifying secondary side water level to be greater than or equal to 17% at least once per 12 hours. 4.4.1.3.4 At least one coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours. 1 NORTH ANNA - UNIT 1 3/4 4-3a Amendment No. 32

~ 9 Material Procery Basis Controlling Material: Circumferential Wald Copper Content: 0.086' WTI Nickel Content: 0.glWT5 Initial RT 19 F NOT. RTNOT 1/4T. 136.3,F 3/47, 116.1 F Curves Applicable For Service Periods Up To 10 EFPY And Contain' Margins Of 20 0F And 80 nsi For Possible Instrument Errors I'* I'* 5 l l l~ l l j l l 1 i / ( / t.m timitOstis ildrettsticf g i f t.m Testing pareties I'* i l- /- I / I'*)) I -l / / l / / ? I'* i l \\ l { f I l use: gi 'i. - ;- l UNACCEPTABLE OPERAll0N [i g 0 '" j l /

1. M :

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1. m i 7

I I C use = 1.tu : f h 1 / m ,,l_ limit for liestep Isles / l i

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v r.serar / i ._ ACCEPTABLE i i i / l_ _ OPERAil0ll iG ,l l / ^ m: Mi ni: e-m: e.. 100 150 200 250 300 350 1 l Cold 1.egTemperature(degreesf? Figure 3.4.2 Reactor Coolant System Pressure-Temperature Heatup Limitations NORTH ANNA - UNIT 1 3/4 4-27 Amendment.No. 15.74.117 {

..d t Material Procertv Basis Controlling Matertal: Cire m ferential Wald k-Cooper Content: 0.088 WT! Nickal Content: 0.11 WTI Initial RT 19 0F MDT. RT.iDT 1/47, 136.3 *F 3/4T, 115.1 F Curves Aoplicable For Service Periods Up To 10 EFPY And Contain Marg 1ns Of 20 0 F And 80 psi For Possible Instrument Errors 2"a i i i i i.. - I I i i / ,'yI l I I I I 3, i / 1 ,,, a i i i i i/ ['ll2 i i i i r 1 i i i i / g, .,,,4 i i i / ,,,j I-l 1 i / 3 i.m UNACCEPTABLE i / i-I = '~i l OPERAil0X lf i E i. _._. i 1 I I I / i i i i/ I I / m 3,, e cas 1.15 i 8 r= : ACCEPTABLE -- i: 3 I 8 / L - OPERATIOR = 5 bh m+- unit rir c,..is.e. a.i s IIN -f/Ir g o i I f W~/A d \\ i

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1i l' l 4 50 108 150 til 154 3H 3M i ColdLet) Temperature (decreesF) ~ Figure 3.4.3 Reactor Coolant Systes Pressure-Temperature Cooldown Limitations NORTH ANIIA - UNIT 1 3/4 4-23 Amendment No. J$,7#,1 17 I \\

} c, l1 REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS ~ LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE: Two power operated relief valves (PORVs) with a lift setting of: a.

1) less tiu or equal to 450 psig whenever any RCS cold leg temperaf re is less than or equal to 261'F, and 2) less than or equal ts 190 psig whenever any RCS cold leg temperature is less than 1509, or b.

A reactor coolant system vent of greater than or equal to 2.07 square inches. I APPLICABILITY: When the temperature of one or r're of the RCS cold legs is less than or equal to 261*F, except when the reactor vessel head is removed. ACTION: With one %1V inoperable, either restore the inoperable PORY a. to OPERAE.t status within 7 days or depressurize and vent the RCS thrr,ugS 2.07 square inch vent (s) within the next 8 hours; maintcin 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. NORTH ANNA - UNIT 1 3/4 4-31 Amendment No. 4,7A,117

_ REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by: Perfomance of a CHANNEL FUNCTIONAL TEST on the PORV actuation a. channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE. b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel, at least once per 18 months. Verifying the PORV keyswitch is in the Auto position and the c. 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. 4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once per 12 hours

  • when the vent (s) is being used for overpressure protection.

Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days. 8 l 4 i i j 10RTH ANNA - UNIT 1 3/4 4-32 Amendment No. yp,3 4

I { EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - Tava " 350 F LIMITINGCONDITIONFOR0_PEPATION i 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of: One OPERABLE centrifugal charging pump, a. 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: With one ECCS subsystem inoperable, restore the inoperable sub-a. system to OPERABl.E 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 accumlated actuation cycles to date. c. The provisions of Specification 3.0.4 are not applicable to 3.5.2.a and 3.5.2.b for one hour following heatup above 324 F or prior to cooldown bescw 324 F. I NORTH ANNA-UNIT 1 3/4 5-3 AmendmentNo.3,M, l 117

r _ -_ ,x j EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 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: Valve Number Valve Function Valve Position

a. MOV-1890A
a. LHSI to hot leg
a. closed
b. MOV-1890B
b. LHSI to hot leg
b. closed
c. MOV-1836
c. Ch pump to cold leg
c. closed
d. MOV-1869A
d. Ch pump to hot leg
d. closed
e. MOV-1869B
e. Ch pump to hot leg
e. closed b.

At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. c. By a visual inspection which verifies that no loose debris (rags, 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 establishing 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 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. 1 NORTH ANNA-UNIT 1 3/4 5-4 j w

i EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2. Verifying that each of the following pumps start auto-matically upon receipt of a safety injection test signal: a) Centrifugal charging pump, and b) Low head sefety injection pump. f. By verifying that each of the following pumps develop the indicated discharge pressure (after subtracting suction pres-sure) on recirculation flow when tested pursuant to Specifica-tion 4.0.5. 1. Centrifugal charging pump >_2410 psig. 2. Low head safety injection pump > 156 psig g. By verifying that the following manual valves requiring adjustment to prevent pump " runout" and subsequent compenent 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 when the ECCS subsystems are required to be OPERABLE. 2. At least once per 18 months. 1. 1-51-188 Loop A Cold Leg 2. 1-SI-191 Loop B Cold Leg 3. 1-SI-193 Loop C Cnid Leg 4. 1-SI-203 Loop A Hot Leg 5. 1-SI-204 Loop B Hot Leg 6. 1-51-205 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 characteristics and verifying tnat: 1. For high head safety injection lines, with a single pump running: a) The sum of the injection line flow rates, excluding the highest flow rate, is > 384 gpm, and b) The tott,1 pump flow rate is 1 650 gpm. NORTH ANNA-UNIT 1 3/4 5-5 Amendment No. 6, 19 o

g .s 4 l EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T,yg < 350 F LIMITING CONDITION FOR OPERATION l 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE: One OPERABLE centrifugal charging pump #, a. 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 spraf pump. APPLICABILITY: MODE 4. ACTION: a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 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 heatremovalmetMEs.essthan350'Fbyuseofalternate Coolant System T l c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Comission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

  1. A maxirnum of one centrifugal charging pump and one low head safety inject on pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 324'F.

l [ NORTH ANNA-UNIT 1 3/4 5-6 Amendment No. 3. M, 84, 117

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2. 4.5.3.2 All charging pumps and safety injection pumps, except the above required OPERABLE pumps, shall be demonstrated inoperable at least once per 12 hours whenever the temperature of one or more of the RCS cold legs is less than or equal to 324*F by verifying that the switches in the Control Room are in the pull to lock position. l l l NORTH ANNA-UNIT 1 3/4 5-6a Amendment No. Jg,117 I L

} 2 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity tran-sients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained suf-ficiently subcritical to preclude inadvertent criticality in the shut-down condition. SHUTDOWN MARGIN requirements vary throughout core life as a fun-ction of fuel depletion, RCS boron concentration, and RCS Tat no 18XE. The most restrictive condition occurs at EOL, with T operating a temperature, and is associated with a postulated lEeam line break ac-cident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.77% ak/k is initially required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T <200*F, the reactivity transientsresultingfromapostulatedst$lElinebreakcooldownare minimal. A 1.77% ak/k shutdown margin provides adequate protection for the boron dilution accident. 3/4.1.1.3 BORON DILUTIO_N_ A minimum flow rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 3000 GFM will circulate an equivalent l Reactor Coolant System volume of 9957 cubic feet in approximately 30 ) minutes. The reactivity change rate associated with baron reductions will therefore be within the capability for operator recognition and control. 3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC) l l l The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting conditions assumed for this 1 parameter in the FSAR accident and transient analyses. The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at l NORTH ANNA - UNIT 1 B 3/4 1-1

'3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.4 MODF_RATOR TEMPERATURE COEFFICIENT (MTC) (Continued) conditions other than those explicitly stated will require extrapolation to i those conditions in order to permit an accurate comparison. j .l The most negative MTC value was obtained by incrementally correcting the MTC used in the FSAR analyses to. nominal operating conditions'. These corrections involved adding the incremental. change in the MTC associated with a core condition of Bank D inserted to an all rods withdrawn condition and an incre. mental change in MTC to account for (measurement uncertainty.at RATED THERMAL POWER cond -5?0x10jtions. These corrections result in thg limiting MTC value of Ak/k/*F. The MTC,value of -4.0 y 10- ak/k/'F' represents a conservative value (with corrections'for burnup and soluble boron) at a core condition-of 300 ppm equilibruim boron concentration and js obtained by making' i these corrections to the limiting MTC value of -5.0 x 10- Ak/k/?F. Once the equilibrium boron concentration falls below about 60 ppm, dilution operations take an extended amount of time and reliable MTC measurements become more difficult to obtain due to the potential for fluctuating core conditions cter the test interval. For this reason, MTC measurements may be suspended provided the measured MTC value at an equ < 60 ppm is less negative than -4.7 x 10-{ librium full power boron concentration delta k The difference between This value and the Ifmiting MTC value of -5.0 x 10 fk/'F. delta k/k/*F conservatively bounds the maximum credible change in MTC between the 60 ppm equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER conditions) and the licensed end-of-cycle, including the effect of boron concentration, burnup, and end-of-cycle coastdown. The surveillance requirements for measurement of the MTC at the beginning and near the end of each fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541'F. This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, and 3) the P-12 interlock is above its setpoint, and

4) compliance with Appendix G to 10 CFR Part 50 (see Bases 3/4.4.9).

r 3/4.1.2 BORAT!ON SYSTEMS The boron injection sy: tem 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.

NORTH ANNA - UNIT 1 B 3/4 1-2 Amendment No. W. III,117

4 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.2 BORATION SYSTEMS (Continued) With the RCS average temperature above 200*F, a mini.':am 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 r_epair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period. The boration capabioity of either system is sufficient to provide a StiUTDOWN MARGIN from expected operating conditions of 1.77% ak/k after xenon decay and cooldown to 200*F. This expected.boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 6,000 gallons of 12.950 ppm borated water from the boric acid storage tanks or 54,200 gallons of 2300 ppm borated water from the refueling water storage tank. i The limitation for a marimum 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 324'F provides assurance l that a mass addition pressure transient 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. The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1.77% ak/k after xenon 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 l The contained water volume limits include allowance for water not I available because of discharge line location and other physical characteristics. The OPERABILITY of one boron injection system during REFUELING insures that this system is available for reactivity control while in MODE 6. l NORTH ANNA - UNIT 1 B 3/4 1-3 Amendment No. E, 76, 68,75,117 )

L REACTIVITY CONTROL SYSTEMS BASES 3/4.1.2 B0 RATION SYSTEMS (Continued) The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within the containment after a LOCA. This pH minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. 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 required to maintain the charging pump cross-connect system operational. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained. (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalignment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors or a restriction in THERMAL POWER; either of these restrictions provides assurance of fuel rod integrity during continued operation. In addition those accident analyces affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation. Control rod positions and OPERABILITY of the rod position in'dicators are required to be verified on a nominal basis of once per 12 hours with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied. The maximum rod drop time restriction is consistent with the assumed roddgoptimeusedintheaccidentanalyses. Measurement with T

2. 500 F and with all reactor coolant pumps operating ensures thaP9 the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

NORTH ANNA UNIT 1 B 3/4 1-4 Amendment No.16, 24 l

y 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in 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 considera-tions require that two loops be OPERABLE. In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat, but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE. The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 324 F are provided to prevent RCS pressure l transients, caused by energy additions from the secondary system which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold 1eg 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, l be within the capability of operator recognition and control. i The requirement to maintain the boron concentration of an isolated loop greater than or equal to the boron concentration of the operating loops ensures that no reactivity addition to the core could occur during startup of an isolated loop. Verification of the boron concentration in ( an idle loop prior to opening the cold leg stop valve provides a reassurance { of the adequacy of the boron concentration in the isolated loop. Operating l the isolated loop on recirculating flow for at least 90 minutes prior to j opening its cold leg stop valve ensures adequate mixing of the coolant in l this loop and prevents any reactivity effects due to boron concentration stratifications. Startup of an idle loop will inject cool water from the loop into the core. The reactivity transient resulting from this cool water injection is minimized by delaying isolated loop startup until its temperature is j

1 l 1 3/4.4 REACTOR COOLANT SYSTEM BASES within 20*F of the operating loops. Making the reactor subcritical prior to loop startup prevents any power spike which could result from this cool water induced reactivity transient. 3/4.4.2 AND 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from beine pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 380,000 lbs per hour of saturated steam at the valve set point. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety 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 during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. The powar operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code 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 l following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. J NORTH ANNA - UNIT 1 B 3/4 4-2 Amendment No. 32

U REACTOR COOLANT SYSTEM L BASES ~ 1 l 3/4.4.7 CHEMISTRY l The limitations on Reactor Coolant System chemistry ensure that I corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural in-tegrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that opera-tion may be continued with contaminant 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 the struc-tural integrity of the Reactor Coolant System. The time interval per-mitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restora the con-taminant concentrations to within the Steady State Limits. The surveillance requirements provide adequate assurance that con-centrations 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 specific activity of the primary coolant ensure that 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 assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent interi.m 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 which will be used as the basis for the reevaluation of the specific activity limits of this site.. This reevaluation may result in higher limits. NORTH ANNA - UNIT 1 B 3/4 4-5

e 1 REACTOR COOLANT SYSTEM BASES The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 pCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Reducing T to < 500'F prevents the release of activity should a steam generator tube $ture sir.ce 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 / TEMPERATURE 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 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. l NORTH ANNA - UNIT 1 B 3/4 4-6 Amendment No.Pf,117 l'

q 0 o ' REACTOR COOLANT SYSTEM BASES l The heatup analysis also ccvers the determination of pressure-l temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thennal 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 alrea(y present. The thennal 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. Conse-quently, for the cases in which the outer wall of the veMel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis. The heatup limit curve Figure 3.4.2, is a composite curve which was l prenared 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 composite curves which were l 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 cooldewn curves were prepared based upon the most ifmiting value of the predicted adjusted reference temperature at the end of 10 EFPY. The adjusted reference temperature was cal.culated using results from a capsule removed after the sixth fuel cycle. The results are documented in Westinghouse Reports WCAP-11777. February 1988 and WCAP-11791, May 1988. The reactor vessel materials have been tested to determine their initial RT The results of these tests are shown in the UFSAR and WCAP-11777 Reactor l opNtion and resultant fast neutron (E>1 Mev) irradiation will cause an increase in the RT Therefore, an adjusted reference temperature, based upon the fluence anST. copper content of the material in question, can be predicted using N US NRC Regulatory Guide 1.99, Revision 2. The haatup and cooldown.1imit curves (Figure 3.4.2 and Figure 3.4.3) include predicted adjustments for this shift in RT at the end of 10 EFPY, as well as adjustments for possible errors in thepbsureandtemperaturesensinginstruments. The actual snift in RT of the vessel material will be establish 2d periodically during operatiggTby 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 NORTH ANNA - UNIT 1 B 3/4 4-7 Amendment No.117 l

1 l I l ' REACTOR COOLANT SYSTEM BASES vessel inside radius are essentially identical, the measure:I transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The'heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule is different from the cal $$IatedART for the equivalent capsule radiation exposure. NDT 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 specified in T.S. 3.1.1.5 aseures 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 Part 50. The limitation: 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. The OPERABILITY of two PORVs or an RCS vent opening of greater than 2.07 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 261'F. l Either PORY has adequate relieving capability to protect the RCS from overpressurization when the transient is linited 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 261'F. This temperature is the water temperature corresponding to a metal temperature of at least the limiting RT 90'F + instrument uncertainty. Above 261*F administrative control is ahku+te protection to ensure the a limits of the heatup curve (Figure 3.4.2) and the cooldown curve (Figure 3.4.3) are not violated. The concept of requiring automatic LTOP at the lower end and administrative control at the upper end, of the Appendix G curves is further discussed in NRC Generic Letter 88-11. NORTH ANNA - UNIT 1 B 3/4 4-8 Amendment No. 7#,117 L_____

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?. REACTOR COOLANT SYSTEM BASES 3/4.4.10 STRUCTURAL INTEGRITY 3/4.4.10.1 ASME CODE CLASS 1, 2 AND 3 COMPONENTS The inspection programs for ASME Code Class 1, 2 and 3 Reactor Coolant System components ensure that the 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 Vessel Code. l l wi l \\ NORTH ANNA - UNIT 1 B 3/4 4-12 Amendment No. 75, 58

L i h qy 3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES l 3/4.5.1 ACCUMULATORS The OPERABILITY of each RCS accumulator ensures that a sufficient volume I of borated water will be immediately forced into the r?. actor 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 sing 7 failure criteria, removal of power to the valver.is required. The limits for operation with an accumulator inoperable for any reason 1 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 immediately 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 3/4.5.3 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. NORTH ANNA - UNIT 1 B 3/4 5-1

.' 6 EMERGENCY CORE COOLING SYSTEMS BASES 1 ECCS SUBSYSTEMS (Continued) 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 324 F l provides assurance that a mass addition pressure transient can be relieved by. the operation of a single PORV. The Surveillance Requirement.s provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. 3/4.5.4 BORON INJECTION SYSTEM The OPERABILITY of the boron injection system as part of the ECCS ensures that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS system cooldown. RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident or a stcam line rupture. The limits on injection tank minimum contained volume and boron concentra-tion ensure that the assumptions used in the sten line break analysis are met. The OPERABILITY of the redundant heat tracing channels associated with the boron injection system ensure that the solubility of the boron solution will be maintained above the solubility limit of 111 F at 15,750 ppm boron. i NORTH ANNA - UNIT 1 B 3/4 5-2 Amendment No. M. (g, 117 _______.__..-_-__.__________-m. _ _ _ _ - _ _,}}

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