ML20073L576
| ML20073L576 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/05/1994 |
| From: | Mccree V Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20073L581 | List: |
| References | |
| NUDOCS 9410130185 | |
| Download: ML20073L576 (74) | |
Text
- _ _ _ - _ _ _ _ _ _ _ _ - _
[gs strop UNITED STATES
[
j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON D.C. m1 k....,/
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 Amendsent No.189 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 April 15, 1994, complies with the standards and requirements of the Atomic Energy Act of 1554, as 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 (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (11) 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.
9410130185 941005 PDR ADOCK 05000338 P
- 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.D.(2) of Facility Operating License No. NPF-4 is hereby amended to read as follows:
(2) Technical Soecificationi j
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.189, 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 within 60 days.
'FOR TH CLEAR EGULATORY COMMISSION s
b Vid: tor M. McC ee, Acping Dirtetor 1
Project Direcporate III-2 Division of Reactor-Projects - I/II Office of Nuclear Relrctor Regulation
Attachment:
}
Changes to the Technical Specifications Date of Issuance: October 5, 1994 i
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ATTACHMENT TO LICENSE AMENDMENT NO.189 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 Paaes V
V XIII XIII 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 4-7 3/4 4-7 3/4 4-7a 3/4 4-7a 3/4 4-7b 3/4 4-27 3/4 4-27 3/4 4-28 3/4 4-28 3/4 4-31 3/4 4-31
,w 3/4 4-32 3/4 4-32 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 8 3/4 1-3 8 3/4 4-1 B 3/4 4-1 B 3/4 4-2 B 3/4 4-2 B 3/4 4-2a B 3/4 4-6 B 3/4 4-6 8 3/4 4-7 B 3/4 4-7 8 3/4 4-8 B 3/4 4-8 B 3/4 5-2 8 3/4 5-2 6-21 6-21
l EDE.X LIMrIING CONDITIONS FOR OPERATION AND SURVmr1 ANCE REQUIREMENTS SECTION PAgg V4.4.2 SAFETY VALVES-SH1TIDOWN.
3/4 4-6 3/4.4.3 SAFETY AND RELIEF VALVES-OPERATING
[
Safety Valves.......
. 3/4 4-7 Relief Valves.
.. 3/4 4-7a 3/4.4.4 PRESSURIZER
.. 3/4 4-8 3/4.4.5 STEAM GENERA *IORS
....~
3/4 4 9 3/4.4.6 REACIOR COOLANTSYS'IEM LEAKAGE Leakage Deection Systems 3/4 4-16 Opermeinnal Leakage..
2 3/4 4-17 Primary to R=-yi=y Leakage
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3/4 4-18b Primary to W y f_*aba= Detection Systems 3/44181 3/4.4.7 CHEMISTRY 3/4 4-19 3/4.4.8 SPECIFIC ACIIVITY 3/4 4-22 3/4.4.9 PRESSURE /IEMPERA'IURE LIMITS Reactor ('nalant Sysem 3/4 4 26 Pressuriser 3/4 4-30 Low-TemperarmeCwHureProtection 3/44-31 l
3/4.4.10 STRUCIURALINTEGRITY ASME Code Class 1,2 & 3 Components
.. 3/4 4-33 l
3/4.4.11 REACTOR VESSEL HEAD VENT 3/4 4-36 j&1 EMER(WNCY CORE COOUNG SYSTEMS E(Y't) 3/4.5.1 ACCUMULA'IORS.
3/4 51 3/4.5.2 ECCS SUBSYSTEMS -T,yg 2 350*F...
. ~. 3/4 5 3 3/4.5.2 ECCS SUBSYSTEMS - T.,, < 350*F _______________._
3/4 5 6 NORTH ANNA -UNIT 1 V
Amendment No.16, 32, 58,66,189,
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l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.5.4 BORON INJECTION SYSTEM Baron Injection Tank.....................................
3/4 5-7 Heat Tracing..............................................
3/4 5-8 3/4.5.5 REFUELING WATER STORAGE TANK.............................
3/4 5-9 s
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NORTH ANNA - UNIT 1 yI
INDEX
=
4 BASES
{
SECTION g
3/.O INSTRUMENTAT10N
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1 3/4.3.1 PROIECTIVE INSTRUMENTATION..
3 3/4 3-1 3/4.3.2 ENGINEERED SAFTEY FEATURE INSTRUMENTATION....... B 3/4 3-1 3
3/4.3.3 MONITORINGINS'IRUMENTATION.......
B 3/4 3-1
{
34M REACTOR COOLANTSYSTEM i
3/4.4.1 REACIOR COOLANTLOOPS
..... B 3/4 4-1 i
1 3/4.4.2 and 3/4.4.3 SAFETY AND RELIEF VALVES.
B 3/44-2 l
i 3/4.4.4 PRESSURI2ER.
B 3/4 4-2a 4
1 3/4.4.5 STEAM GENERA ~IORS
.. B 3/4 4-3 i
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3/4.4.6 REACIDR COOLANT SYS'IEM LEAKAGE B 3/44-4 J
l 3/4.4.7 CHEMISTRY B 3/44-5 i
3/4.4.8 SPECIFIC ACIIVITY B 3/44-5
]:
3/4.4.9 PRESSURE / TEMPERATURE LIM 1TS B3/44-6 s
3/4.4.10 STRUCTURALINTEGRITY...
B 3/4 4-12 l
1 3/4.4.11 REACIDR VESSEL HEAD VENT...'.
.. B 3/4 4-13 4
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5 3
4 1
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4 NORTH ANNA - UNIT 1 XIII AmendmentNo. 32.
H H.189, j
INDEX BASES SECTION P2fiE E
EMERGENCYCORE COOI TNG SYSTEMS GCCS) 3/4.5.1 ACCUMULA10RS
................................. B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBS YSTEMS....................................................... B 3/4 5-1 3/4.5.4 BORON INJECTION SYSTEM
....................... B 3/4 5-3 l
3/4.5.5 REFUELING WATER STORAGE TANK (RWST).................
... B 3/4 5-3 1
i 3L44 CONTAINMENT SYSTEMS 3/4.6.1 CONTAINMENT...
B 3/4 6-1
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3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS..................
..... B 3/4 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES.......
... B 3/4 6-3 i
3/4.6.4 COMBUSTIBLE GAS CONTROL.............
............................. B 3/4 6-3 l
j 3/4.5.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM.................... B 3/4 6-4 l
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NORTH ANNA - UNIT 1 XIV Amendment No. 48,188
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REACITVITY CONTROL SYSTEMS FLOW PATHS - OPER ATING LIMITING CONDITION FOR OPERATION 3.1.2.2 Each of the following boson injection flow paths shall be OPERABLE:.
The flow path fnun the boric acid tanks via a boric acid transfer pump and a a.
charging pump to the Reactor Coolant System, and b.
'Ihe flow path from the refueling water storage tank via a charging pump to the Reactw Coolant System.
8 APPLICABILITY:
MODES 1,2,3 AND 4.
ACITON:
With the flow path from the boric acid tanks inoperable, restore the inoperable flow a.
path to OPERABLE stams within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> w be in at least HOT STANDBY and barated to a SHUTDOWN MAROIN equivalent to at least 1.77% Ak/k at 200'F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the flow path to OPERABM status within the next 7 days a be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, b.
With the flow path frorn the refueling water storage tank inoperable, restore the flow path to OPERABLE status within one hour a be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:
At least once per 7 days by venfying that the temperature of the heat traced portion a.
of the flow path from the boric acid tanks is 2: 115'F.
Only one baron 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 235'F.
l NORTH ANNA - UNIT 1 3/4 1 9 Amendment No. If, f*. !!?. !?O.
- 189,
l REACTIVITY CONTROL SYSTEMS a
SURVEILLANCE REQUIREMENTS (Continued) l b.
At least once per 31 days by verifying that-each. valve (.nanual, power operated or automatic) in the flow path that is not locked, sealed, or otherw:se secured in position, is in its correct position.
1 1
c.
At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.
NORTH ANNA-UNIT 1 3/4 1-10 i
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REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1. 2. 3 At least one charging pump in the boron injection flow path l
required by Specification 3.1.2.1 shall be OPERABLE.
-~
APPLICABILITY: MODES 5 and 6 ACTION:
a.
With no charging pump OPERABLE, suspend all operations involving CORE I
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 OPERABLE status as soon as possible.
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SURVEILLANCE REQUIREMENTS l
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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the switches in the Control Room have been placed in the pull to lock position.
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NORTH ANNA - UNIT 1 3/4 1-11 Amendment No. 16, 24
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REACITVITY CONTROL SYSTEMS 1
CHARGING PUMPS - OPERATING l
}
LIMITING CONDITION FOR OPERATION l
3.1.2.4 At least two charging pumps shall be OPERABLE.
{
APPLICABILITY:
MODES 1,2,3 and 4 *.
i ACHON:
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With only one charging pump OPERABLE, restore a second cha3 ng pump to OPERABE status i
within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borand to a SHUTDOWN MARGIN I
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equivalent to at least 1.77% Ak/k at 200T within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore a second charging pump j
to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 l
hours. The provisions of SM6=h 3.0.4 are not applicable for one hour following beamp above 235T orprior to cooldown below 2359.
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SURVEILLANCE REQUIREMENTS j
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i 4.1.2.4.1 The above required charging pumps shall be h==ased OPERABG by verifying,
)
l that on reciWah flow, each pump develops a dischargo penseurmoflL2410 psig when tested j
pursuant to Sprendon 4.0.5.
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4.1.2.4.2 All charging pumps, except the above required OPERABLE pump, shall be j
demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperanne of one or more of the i
RCS cold legs is less than or equal to 2357 by verifying that the swithes in the Control Room l
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have been placed in the pull to lock position.
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l A maximum of one centrifugal charging pump shall be OPERABM whenever the
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temperature of one or more of the RCS cold legs is less than or equal to 2357.
l NORTH ANNA - UNIT 1 3/4 1-12 Amendment No. 3. It !!?,470 q
- 189,
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REACIOR COOLANT SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a.
At least two of the coolant loops listed below shall be OPERABIE:
i
- 1. Reactw Coolant Imp A and its associated s' team generator add reactor coolant
- Pump, 1
- 2. Reactw Coolant Imp B and its associated steam generator and reactw coolant.
Pump,*
- 3. Reactor Coolant I. cop C and its associated steam generator and reactw coolant l
pump,*
{
- 4. Residual Heat Removal Subsysem A,**
- 5. Residual Heat Removal Subsyrem B.**
b.
At least one of the above canlant loops shall be in operation.***
APPLICABIIrrY-MODES 4 and 5 ACTION-With less than the above requimd 'aops OPERABLE, immediately. initiate a.
corrective action so return the requiredloops no OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
b.
With no coolant loop in operation,==g-: +1 all operations involving a reduction in baron concentration of the Reactor Cociant Sysem and immediately initiate corrective action to return the rgo;.4 coolant loop so operation.
A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 235*F unless the W7watertemperanne ofeach l
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 reacta coolant pumps and residual heat removal pumps may be de-energized for up to I hour provided 1) no operations are permitted that would cause diludon of the reactor coolant system boton concentration, and 2) cost outlet temperstme is maintained at least 10*F below saturation temperature.
NORTH ANNA - UNIT 1 3/44-3 Amendment No. !? 311!?, !".
- 189,
SHUTDOWN l
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SURVEILLANCE REQUIREMENTS j
4.4.1.3.1 The required RHR subsystems shall be demonstrated OPERABLE per j
Specification 4.7.9.2.
i 4.4.1.3.2 The required reactor coolant pump (s),if nct in operation, shall be determined to be OPERABLE once per 7 days by venfying correct breaker alignment and indicated power availabilty.
{
4.4.1.3.3 The required steam generator (s) shall be determined OPERABLE by verifying j
secondary side water level to be greater than or equal to 17% at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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4.4.1.3.4 At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, venfy at least one coolant loop tok in operation j
and circulating reactor coolant by i
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- a. Venfying at least one Reactor Coolam Pump is in operation.
1 or
- b. Venfying at least one RHR Loop is in operation and.
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1.
if the RCS temperature >140' F or the time since entry into MODE 3 is
<100 hours, circulating reactor coolant at a flow rate 23000 gpm.
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- 2. if the RCS temperature 5140' F and the time since entry into MODE 3 is 2100 hours0.0243 days <br />0.583 hours <br />0.00347 weeks <br />7.9905e-4 months <br />, circulating reactor coolant at a flow rate 22000 gpm to remove decay heat.
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NORTH ANNA UNIT 1 3/4 4-3a Amendment No. M,)37
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SAFETY AND RFT TFR VALVES -OPERATING i
SAFETYYALYES
=
l LIMITING CONDITION FOR OPERATION 4
i 3.4.3.1 All pressurizer code safety valves shaE be OPERABLE with a lik seeing of
[
2485 PSIG 1%.*
APPLICABILITY:
MODES 1,2 and 3.
4 4 4
ACTION:
j With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4 SURVEILLANCE REQUIREMENTS I
l 4.4.3.1 No additional Sury,ma-Regsw..ama other than those required by S,9end=
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4.0.5.
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'Ihe lift setting pressure shall correspond to ambient conditions of the valve at nominal J
temperstme and pressure.
i NORTH ANNA - UNIT 1 3/44-7 Amendment No. 189,
e REACTOR COOLANT SYSM SAFETY AND RFI TFE VALVES -OPERATING l
RELIEF VALYEi
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LIMITING CONDITION FOR OPERATION 3.4.3.2 Both power +M relief valves (PORVs) and their associated block valves shs31 be
[
f OPERABLE.
i APPLICABILITY:
MODES 1,2, and 3.'
ACITON:
r--
l With one or both PORVs inoperable but capable of being manually cycled, within a.
I hour either restore the PORV(s) to OPERABM stams or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHU'IDOWN within the l
following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With one PORV inoperable and not capable of being manually cycled, within I hour either restore the PORV to OPERABLE stams or capable of being manually cycled, or close its===Matad block valve and remove power from the block valvet restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT -
STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
With both PORVs inoperable and not capable of being manually cycled, within I c.
hour either restore at least one PORV to OPERABLE status or capable of being manually cycled, or close its associated block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
l d.
With one or both block valves inoperable, within I hour restore the block valve (s) to OPERABLE stams or place its associased PORV(s) in manual conaol. Restore at least one block valve to OPERABLE status within the next hour if both block
/
valves are inoperable; restore the remaining inoperable block valve to OPERABLE stams within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
'Ihe provisions of Specification 3.0.4 are not applicable.
c.
NORTH ANNA - UNIT 1 3/44-7a Amendment No. M,189,
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SAFETY AND RFT TFN VALVES - OPERATING 1
i RELIEF VALVES SURVFn1ANCE REQUIREMENTS i
4.4.3.2.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE:
]
At least once per 31 days by performing a CHANNEL FUNCTIONAL TEST, a.
{
excluding valve operation, and -
j b.
Atleast once per 18 months by i
1.
Operating the PORV through one complete cycle of full travel dunng MODES 3 or 4, and
, 2.,
Operating the solenoid aircontrol valves and check valves on the associated l
accumulators in the PORV control systems through one complete cycle of full travel, and j
3.
Perfonning a CHANNEL CALIBRATION of the. actuation i
instrumentation.
4.4.3.2.2 Each block valve shall be demonstrated OPERABIE at least once per 92 days by i
operating the valve through one complete cycle of full tmvel unless the block valve is closed in order to meet the requirements of ACTION b or c in Specification 3.4.3.2.
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NORTH ANNA - UNIT 1 3/4 4-7b Amendment No. 189, i
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~ - - - -
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Figure 3.4 North Anna Unit 1 Reactor Coolant Sys, tem Heatup Umitations i
m Material Property Basis Umiting Meterlek Circumferential Wold Seem Umiting ART at 30.7 EFPY: 1/4-T,162.9 F 3/4 T,139.9 Hesme Rates (FAw) mem' l
2500.00 1
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Leek Test U.TA
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2000.o0
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Unacceptable
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- 1500.00 g
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dperation - '
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500.00 20
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0.00 0
50 100 150 200 250 300 350 1
Cold Leg Temperature (Dog. F)
Norm Anne Una 1 Reacter Cootem System Mesme Ummenene (Heeme Ames e m 60 F/ho Aephcomes for me First 30.7 EPPY (WIshout Morgms for Inswumentsoen Erreral i
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NORTH ANNA - UNIT 1 3/4 4-27 Amendment No. I',7', !!?, !?0,189, i
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Pigm 3.4 North Anna Unit 1 l
Reactor Coolant System Cooldown Umitations Material Property Basis
% M Circumferential Wald Seene r is. P *e,
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_N"9, ART st 3o.7,EFPY: 1/4-T,162.9 F 3/4-T,139.9 F
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Cold Leg Temperstwo (Deg. F) i
- ^aae un. i nesee, c
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NOR1H ANNA - UNIT 1 3/4 4-28 Amendrnant No. !i ?f, !!?, !?O, j
- 189, 1
=_
REACTOR COOLANT SYSTEM ILW. TEMPERATURE OVERPRFRSURE PROTEC1 TON g
l LIMITING CONDITION FOR OPERATION i
3.4.9.3 Two power operated relief valves (PORVs) shall be OPERABG with liA seeings o i
i (1) less than or equal to 500 psig whenever any RCS cold leg temperanseis.less than or eq l
2359, and (2) less than or equal to 395 psig whc any RCS cold leg temperanno is less than 1507.
.i j
APPLICABIT ITY-MODE 4 when the temperature of any RCS cold leg is less than or equal to 4
{
2359, MODE 5, and MODE 6 when the head is os the reactor' vessel and the RCS is not vented through a 2.07 square inck erlarger vent.
ACITON:
1 1
a.
With one PORV inoperable in MODE 4, restoss the %ble PORY to.
j OPERABLE stams within 7 days or depressurise and vent the RCS through atleast a 2.07 square inch vent within the next 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />su,.
b.
With one PORV inoperable in MODES 5 or 6, either'(!) restore % f6 operable PORV to OPERABG status'within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,' oH2)'cinnpleis depr==Wa= and j
venting of the RCS through at least a 2.07 square inch vent within a total of 32 j
hours.
}
With both PORVs ivble,compless andventing of the RCS c.
{
through at least a 2.07 square inch vont within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
4 l
d.
With the RCS vented per ACDONS a, b, or c, verify the vent pathway at least once j
per 31 days when the pathway is provided by a valve (s) that is locked, scaled, or otherwise secured in the open pasirinn; otherwise, verify the vent pathway ever 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i e.
In the event either the PORVs or the RCS vent (s) ame used to mitigam an RCS pressure transient, a Special Report shall be prepamd and submined to the j
Commission pursuant to S%".c.dse 6.9.2 within 30 days. The report shall j
describe the circumstances initiadng the transient, the ef! bet of the PORVs or RCS
{
vent (s) on the transient, and any corrective action necessary to prevent recurrence.
4 j
f.
'Ihe provisions of Spari& Mon 3.0.4 are not applicable.
l l
l 1i i
f 1
NORTH ANNA - UNIT 1 3/4 4-31 Amendmaar No. !'. 74. ! !?. !?^.
- 189,
i REACTOR COOLANT SYSTEM IDW-TEMPERATURE OVERPRFMSURE PROTECTION _
l SURVEILLANCE REQUIREMENTS 4.4.9.3 Each PORV shall be demonstrated OPERABLE byr
[
Perforenaam of a CHANNEL FUNCI10NAL TEST on'the PORV actuation
^
a.
channel, butexcluding valve operadon, within 31 dais prior to enering a condition in which the PORV is required OPERABLE and at least once per 31 days thereaher when the PORVis required OPERABLE."
b.
Performance of a CHANNEL CALIBRATION on the PORV acmation channel, at least once per 18 months.
Verifying the PORV by..<iich is in the Auto position and the IDRV'innlarinn c.
valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for omresure protection.
d.
Testing pursuant to Specification 4.0.51 i
NORTH ANNA -UNIT I 3/4 4-32 Amewhnaar No. 44r44,
- 189,
I EMERGENCY CORE COOI ING SYSTEMS J
{
ECCS SUBSYSTEMS--Tave 2 350*F i
LIMITING CONDITION FOR OPERATION 1
3.5.2 Two ird;At ECCS subsystems shall be OPERABLE with each subsystem comprised of:
One OPERABM centrifugal charging pump, a.
I b.
One OPERABM low head safety injectico pump, j
c.
An OPERABLE flow path capable of transferring fluid to the Reactor Coolant
{
System when taking suction from the refueling water storage tanic on a safety
)
injection signal or from the containment sump when suction is transferred during
{
l the reciH *iaa phase of operation or from the discharge of the outside 4
- da+ ion spray pump.
f, APPLICABILITY:
MODES 1,2 and 3.
j ACTION:
i With one ECCS subsystem inoperable, restore the inoperable subri._ to a.
)
OPERABE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHU1DOWN within the next 12 i
hars.
l l
b.
In the event the ECCS is actuated and injects waterinto the Reactor Coolant j
System, a Special Report shall be piepared and submined to the C-i=*iaa i
pursuant to 5,Mation 6.9.2 within 90 days describing the circumstances of the j
actuation and the total accumulated acnindan cycles to dase.
j c.
De provisions of Speci6 cation 3.0.4 are not applicable to 3.5.2.a and 3.5.2.b for j
one hour following beamp above 2357 or prior to cooldown below 235T.
l.
i, i
l l
NORTH ANNA-UNIT I 3/45-3 Amendment No. 3. If, !!? 153
- 189, 4M.
EMERGFNCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS -Tavr < 350*F i
LIMITING CONDITION FOR OPERATION 4
i
{
3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
1 One OPERABG centrifugal charging pump',
a.
j b.
One OPERABG low head safety injection pump', and An OPERABLE flow path capable of automatically transferring fluid to the reactor c.
coolant sysum when taking suction from the refueling water storage tank or from the conminment sump when the suction is transferred during the recirculation phase of operation or from the discharge of the outside recirculation spray pump.
APPLICABIT 3TY-MODE 4.
ACTION:
With no ECCS subry.
OPERABM because of the inoperability rf either the a.
centrifugal charging pump or the flow path frtun the refheling water storage tank, restore at least one ECCS subsystem to OPERABG status within I hors or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
b.
With no ECCS subsysum OPERABLE because of the inoperability of the low head safety injection pump, restose at least one ECCS subsysum to OPERABG status or maintain the Reactor Coolant Synaam T l
g ess than 3507by useof alernas heatremovalmethods.
In the event the ECCS is actuated and injects water into the Reactor aatnar c.
r System, a Special Report shall be ptepamd and submined to the a==Wia=
c pursuant to Sp+T==~- 6.9.2 within 90 days describing the circumstances of the acmation and the notal accumulated actuation cycles to dase.
A maximum of one centrifugal charging pump and one low head safety injection pump shall be OPERABG whw the temperamre o(one or more of the RCS cold legs is less than or equal to 2359.
[
NORTH ANNA -UNIT 1 3/45-6 Amendramar No. 3. I', *4, ! !?,
- 189, 4M,
i i
3 i
EMERGENCY CORE COOUNG SYSTEMS
}
SURVEILLANCE REQUIREMENTS 4.5.3.1
'Ibe ECCS subsystem shall be demonstrated OPERABLE per the applicablo l
SurWUa= Requirements of 4.5.2.
4.5.3.2
)
All charging pumps and safety injection pumps, exmpt the above requimd OPERABLE pumps, shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> wLvs the temperature of f
one or more of the RCS cold legs is less than or equal to 2357 by verifying that the switches in j
[
the Control Room are in the pull lo lock posidon.
4 i
i at 1
6 1
l NORTH ANNA -UNIT 1 3/45-6a Amend = war No. !i !!?, !?O.
- 189,
4 l
i t
j REACTIVTTY CONTROL SYSTEMS BASES j
M BORATION SYSTEMS (Cnneinned)
With the RCS averap temperature above 2007, a minimum of two separate and radandene i
baron injection systems are provided to ensure single Anactional capability in the event an assumed j
failure renders one of the sysems inoperable. Allowable out-of-service periods ensure that minor j
component repair or corrective action may be completed without undue ' risk to everall facility j
safety from injection system failures during the tepair period.
j
'Ibe baration capability of either system is suf5cient to g4 a SHUTDOWN MARGIN i
from *W operating conditions of 1.77% AirA aAer xenon decay and cooldown to 2007. *Ihis exy.ci.4 baration capability requirement occurs at EOL fkcun fhl! power equilibrium menon conditions and requires 6,000 gallons of 12,950 ppm barated water froen the baric acid storage j
tanks or 54,200 gallons of 2300 ppm barased water fran the refbeling water storses tank.
1 The limitation for a innvirnum of one centrifhgal charging pump a be OPERABLE and the j
Surveillanem Regia to verify all charging pumps sacept the
- M OPERABLE pump a j
be inoperable below 235T provides assurance that a maa nMirW pressure transient can be
[
mlieved by the operation of a single PORV.
{
With the RCS temperstme below 2007, one injection sysaem is acceptable without single failure consideration on the basis of the stable reactivity enndirian of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and posidve reactivity change in the event the single injection system hart = nae inoperable.
h baron capability required below 200T is municiane to provide a SHUTDOWN i
l MARGIN of I.77% AlrA after menos decay and cocidown frons 2007 to 1407.11 is condition i
requires either 1378 gallons of 12,950 ppen borated waner from the beric acid sacrase tanks or 3400 gallons of 2300 ppm barated waner from the refueling waser storags tank.
t h contained waser volume limits include allowance for water not available because of j
discharge line location and other physical characteristics.1he OPERABILITY of one baron injection system during REFUELING insures that this system is available for reactivity control l
while in MODE 6.
4 i
i i
1 a
i 1
i NORTH ANNA-UNIT 1 B3/41-3 Amendment No. 5, IU". "E
- 189,
!!?.!?O.
i i
1 REACTMTYCONTROLSYSTEMS f
l i
3/412 BORATION SYSTEMS (Continued) i The limits on contained water volume and boron concentration of the RWST-ensure a pHs value of between 7.7 and 9.0 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 j
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
,[_
u The specifications of this section (1)' ensure th'at 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 analyses. CPERABILITY of the movable control assemblies is established by observing rod motion and determining that rods are positioned within i 12 steps (indicated position) of the respective demand step counter position. The CPERABILITY of the individual rod position indication system is established by,
appropriale penodic CHANNEL CHECKS, CHANNEL FUNCTIONAL TESTS, and CHANNEL CAllBRATIONS. OPERABILITY of the individual rod position indicators is required to determine control rod position and thereby ensure compliance with the control rod alignment and insertion limits. The OPERABLE condition for the individual rod position indicators is defined as being capaw of indicating rod position within i 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 one hour in every 24 stabilization period (thermal " soak time") to allow stabilization of known thermal drift in the individual rod position indicator channels during which time the indicated rod position may vary from demand position indict. tion by no more than t 24 steps. This "1 in 24" feature is an. upper limit on the frequency of thermal soak allowances and is available both for a continuous one hour period or one consisting of several discrete intervals. During this stabilization period, greater reliance is placed upon the demand position indicators to determine rod position. In addition, the t 24 step / hour limit is not applicable when the control rod position is known to be greater than 12 steps from the rod group step counter demand position indication. Above 50 percent of RATED THERMAL POWER, rod motion is not expected to induce thermal transients of sufficient magnitude to exceed the individual rod position indicator instrument accuracy of t 12 steps. Comparison of the demand position indicators to the bank insertion limits with verification of rod position by the individual rod position indicators (after thermal soak following rod motion below 50 percent of RATED THERMAL POWER) is sufficient verification that the control rods are above the insertion limits.
i The control bank FULLY WITHDRAWN position can be varied within the interval' of 225 to 229 steps withdrawn, inclusive. This interval permits 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 administrative 1y controlled, using i
the rod insertion limit operator curve.
NORTH ANNA-UNIT 1 B 3/4 14 Amendment No. M,M,7M, )W.
- 149,
,---...------c---
c-
i 1
34M
g BASES l
Ef,fJ, REACTUR COOLANT LDOPS
)
'I1r plant is designed to operate with alt reactor coolant loops in operation and maintain the DNBR above the design limit during all normal operations and anticipated transients. In MODES I and 2 with one reactor coolant loop not in operation, this==~incation requires that the plant m atleast HOT STANDBY within I hour.
In MODE 3, a single reactor coolant loop provides sufScient heat removal capability fo removing decay heat; however, single failure caa*id-ations require that two loops be OPERABLE.
i In MODES 4 and 5, a single reactor coolant loop or RHR loop provides suf5cient heat removal eaa=Mi y for removing decay heat, but single failure considerations require that at least t
two loops be OPERABI.E. 'I1ms, if the reactor coolant loops are not OPERABIE, this 4
specification requires two RHR loops to be OPERABLE.
After the reactor has shutdown and entred into MObE 3 for at leam 100 houra, a minimum RHR system flow rase of 2000 spm in MODE 5 is r,.... :4 provided there is sumcient decay heat removal to maintain the RCS temperature less than or equal to 1407. Since the decay heat
{
i powerr. * + + rate decreases with time after reactor abundown, the requirements for RHR sysum decay heat removal also decrease. Adequate decay heat removal is provided as long as the reactor has been shutdown for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after enny into MODE 3 and RHR flow is -
suf5cient to maintain the RCS temperature less than or equal to 1407. '!he reduced flow rate provides additional margin so vertexing at the RHR punp section while in Mid I. cop Operation.
During a ree'iaa in reactor caalant system baron meration the Sp+Weion 3.1.1.3.1 requirement to maintain 3M g3m Bow rate provides suf3cient canlant Cire"I*'iaa to minimim the effect of a baron dilution incident and to prevent baron strari&arinn.
'Ihe restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less 3
than or equal to 2354 are provided to prevent RCS pressa aansients, caused by energy additions l
from the secondary system which could exceed the 16.4 of Apa-adi G to 10 CPR Part 50. "Ihe j
RCS will be protected against w eme transients and will not exceed the limits of Ag =-W r
i G by restricting starting of the RCPs to when the i-zid y water semperature of each steam l,
generator is less than SOT above each of the RCS cold leg semperannes.
4
'Ihe operation of one Reactor raalant Pump or one RHR pump provides adequate flow to ensure mixing, prevent strariharian and produce gradual reactivity changes during bcron concentration reductions in the Reactor Coolant System. 'Ihe reactivity change ram associated with boron reduction will therefore be within the capability of operator recognition and control
'Ihe requirernent to maintain the baron concentration of an isolated loop greater than or equal to the baron concentration of the operating loops ensures that no reactivity mMirian to the core could occur during startup of an isolated loop. Veri 5 cation of the baron concensmoon m en idle loop prior to opening the cold leg stop valve provides a reassurance of the ackgancy of the boron concentration in the isolated loop. Operating the isolated loop on recirculating flow for at least 90 minutes prior to opening its cold leg stop valve ensures adequate mixing of the coolant in this loop and prevents any reactivity effects due to baron canemneration strati 5 cation.
Startup of an idle loop will inject cool waar from the loop into the core. 'Ibe reactivity t
ransient resulting from this cool water injection is minimimi by delaying isolated loop startup t
j untilits temperatureis NORTH ANNA - UNIT 1 B3/44-1 g,odment No. I', 3, I ',7.,! 3],
n l
3/M REACTOR COOLANT SYS'mM BASES within 20*F of the operating loops. Making the reactor suberitical prior to loop startup p power spike which could result from this cool water induced reactivity transient..
J j
3/4.4.2 AND 3/4.4 3 SAFETY 1ND RFr'rs* VALVES
~
l
[
ne prenurizer code safety valves operam to prevent the RCS from being pressurized j
above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 380,000 lbs pe i
of saturated steam at the valve set point. De relief capacity of a single safety valve is adeq relieve any overpressure condition which could occur during hot shutdown. In the event that no l
safety valves are OPERABLE, an operating RHR loop, conneced to the RCS, or the power j
opermed relief valves (PORVs) will provide overpressure relief capabihty and will prevent RCS j
overpressurization.
i 3
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 capac i
all of these valves is greaser than the maximem surge ram resulting from a complete loss ofloa assuming no reactor trip until the first Reactor Protective Sysum trip setpoint is reached (Le., no l
credit is taken for a direct reactor trip on the loss ofload) and also assuming no operation of the i
power operated relief valves or steam dump valves.
4 l
Demonstration of the safety valves' lift settings will occur only during shutdown and will
{
be performed in accu dence with the provisions of Section XIof the ASME Boiler and Pressure j
Vessel Code.
l De poweroper.;e4 relief valves and seam bubble function to relieve RCS pressure dur 3
all design transients up to and including the design step load decs with steam dump. Opera of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. Each PORY has a remotely operated block valve to provide a i
positive shutoff capability should a relief valve become inoperable.
i The OPERABILITY of the PORVs and block valves is determined on the basis o being capable of performing the following functions:
2; a) Manual control of PORVs to concol reactorcoolant system pressure, nis is a function i
that may be used to mitigas certain accidents and for plant shutdown.
i j
b) Maintaining the integrity of the reactor coolant pressure hoaaA*y. His function is j
related to controlling identified leakage and ensuring the ability to detect unidentified j
reactor coolant pressure boundary leakage.
i.
)
I 4
1 1
NOR'ITI ANNA - UNIT I B3/44-2 Amendment No. 31 !'!.
4
- 189, i
4
)
3LM REACTOR COOLANTSYSTEM BASES c) Manual contml of the block valve to (1) unblock an isolated PORY used for manual contml of reactor coolant system preuure (Item a above), and (
late a PORV with excessive seat leakage (Item b, above).
d) Automatic control of PORVs to control reactor coolant system pressure. His funcdon reduces challenges to the code safety valves fw overpressurization events.
e) Manual control of a block valve to isolate a stuck-open PORV.
Survd11== Requirements provide the assurance that the PORVs and block valves can perform their f-- +-M5. Sp+ -:F='=- 4.4.3.2.1 addresses the PORVs and S;="-==*4.4.3.2.2 addresses the block valves. De block valves are exempt from the surveillance requirements t cycle the valves when they have been closed to comply with the ACTION requirementa. His precludes the need to cycle the valves with full system differential pressure w when maintenan is being performed to restore an inoperable PORV to operable status.
Survai11== Reqs es 4.4.3.2.1.b.2 provides for the testing of the mechank al and electrical aspects of control sysems fw the PORVs.
Testing of FORVs in HOT STANDBY w HOT SHUIDOWN is rq44 in order to simulate the temperature and pressure environmental effects on PORVs. Testing at COLD SHUTDOWN is not considered to be a representative est fw assessing PORVmi mance u normal plant operating cciddons.
3/M4 PRESSURIZER Be limit on the maximum water volume in the pressurizer assures that the parameter maintamed within the normal saady stas envelope of operation assmnad in the SAR. The lim consistent with the initial SAR assumptions. De 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic survaill=~ is sufficient to ensure that the parameter is restored to within its limit following eW transient operation. De maximum water volume also ensures that a steam bubble is formed and thus t hydraulically solid system.
4; 1
. NORTH ANNA - UNIT 1 B3/44-2a Amendment No. 189,
)
REACTOR COOLANT SYSTEM BASES 1
1 3/4.4.7 CHEMISTRY
~
1 i
The limitations on Reactor. Coolant System chem.istry ensure that 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 j
associated effects of exceeding the. oxygens chloride and fluoride limits i
are time and temperature dependent. Corrosion studies show that' opera-1 tion may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient, Limits, for the specified 4
limited time intervals without having a signif'icant 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 actionsto restore the con-taminant concentrations to within the Steady State Limits.
The surveillance requirements provide adeq'uate assurance that con-centrations in excess of the limits will be detected in sufficient time
)
i to take corrective action.
I i
1 i
3/4.4.8 SPECIFIC ACTIVITY i
i The limitations on the specific activity of the primary coolant i
ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 interim a
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..
j This reevaluation may result in higher limits.
1 NORTH ANNA - UNIT 1 B 3/4 4-5 6
r
l i
l j
BASES l
I
{
De ACTION stamment permitting POWER OPERATION to continue forlimited time periods with the primary coolant's specific acdvity > 1.0-pCi/ gram DOSE EQUIVALENI l
but within the allowable limit shown on Figant 3A-1, accommodates possible iodine s
[
i phenomenon which may occur following changes in THERMAL POWER.
]
Reducing T, to < 500T prevents the release of activity should a steam generator tube i
{
rupture since the saturadon pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance tha j
excersive =ari& activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking wiB be used to assess the parameters[
associatedwithspikingpha~=aaa= Areductioninfrequencyofisosopicanalysesfollowingpowe{
changes may be permissible ifjustified by the data obemined.
t
- 3LMj, PRRAKURN / TEMPERATURE UMTTS Reartnr cantene Svu _..
g All compnnante in the Reactor Coolant System are designed to withstand the effecs of i
cyclic loads due so sysem temperanse and pressure changes. These cyclic loads are int
{
nonnal load transients, reactor trips, and startup and shusdown operations. De various catego 1
of load cycles used for design purposes me provided in Section 5.2 of the UFSAR. Dur i
and shutdown, the rates of temperature and pressure changes are liminna so that the maximum l
specified heatup and cooldown rates are consistent with the design assumptions and satisfy t
stress limits forcyclic operation.
4 j
During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses j
which vary from compressive at the inner wall to tensile at the outer wall Dese thermal ind=d j
compressive snesses tend to alleviam the tensile snesses indami by the internal pressure.
{
Therefore, a pressure-temperature curve based on steady stas enndhinne (i.e., no thermal sne f
represents a lower bound of all Wrnilar curves for finite heatup rates when the inner wall of the vessel is treated as the governing lar=*iaa he heatup analysis also covers the d,s...;narina of pressure-temperstme limitations for l
the case in which the outer wall of the vessel hecarnes the consolling location. The thermal j
gradients established during heatup produce tensile stresses at the ouer wall of the vessel These I
j stresses are additive to the pressure induced tensde stresses which are aheady present. ne thermal j
induced stresses at the outer wall of the vessel are tensile and are dap ad at on both the rate of I
heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described i
for the heatup of the inner wall cannot be defined. Consequendy, for the cases in which the outer wall of the vessel t+x+ =5 the stress controlling laearinn, each heatup rate ofinserest must be j
analyzed on an individual basis.
i i
f NORTH ANNA - UNIT 1 B3/44-6 Amendment No. ^' ? !? !?^.
I
- 189, 4
BASES i
The heatup limit curve, Figure 3.4-2, is a composim curve which was prepamd i
deterrnining the most conservative case, with either the inside or outside wall c
)
heatup rate up to 60*Fper hour. The cooldown limit curves of Figure 3.4-3 are compos j
which were prepared based upon the same type analysis with the exception that t
{
location is always the inside wall where the cooldown thermal gradients tend to y.vd c
{
stresses while producing compressive stresses at the outside wall. The heatup and cooldown 4
were pmywM based upon the most limiting value of the predicted adjusted reference te at the end of 30.7 EFPY. The most recent capsule analysis results are t-:-:=*ned in W j
Report WCAP-11777, February 1988. The heatup and cooldown curves are documented in j
Westinghouse Report WCAE-13831, Rev.1. August 1993.
The reactor vessel materials have been tested to determine their initial RTmyr. Reac operation and resultant fast neutron (E > 1 Mev) irradiad= will cause an increase in t
An adjused reference semperature, based upon the fluence and copper content of the materi
[
question, can be predicted using US NRC Regulatory Guide 1.98, Revision 2. The heat j
cooldown limit curves (Figure 3.4-2 and Figure 3.4 3) include predicted adjustments for this j
in RTmyr t the end of 30.7 EPPY. 'Ibe reactor vessel beltline region material properties a
on Figures 3.4-2 and 3.4-3.
j The actual shift in the RTmyr of the vessel material will be established Mdically by 1
i removal and evaluation of the reactor vessel material specimens installed on the inside wall of th j
thermal shield. The survein..c. capsule withdrawal schedule was prepared in accordance wit requirements of ASTM E-185 and is presented in the UFSAR. Regulatory Guide 1.99, Revision j
2, provides giMaac> for cabladaa of the shift in RTmyr using measured data. Dosimet 7
the survei11aara capsule is used to determine the neutron fluence to which the mater were exposed, and to support ral& adanal estimmaan of the neutron fluence to the reactor vesse t
}
'Ihe pressure-temperstme limit lines shown on Figure 3.4 2 for inservice leak and j
hydrostatic testing have been provided to assme compliance with the minimum temperatu
{
j requirements of Ap;-:- "- O to 10 CPR 50. 'the minimum temperature for criticality specified T.S. 3.1.1.5 assures compliance with the criticality limits of 10 CFR 50 Apd G.
i
'Ihe numberof reactor vesselirradiation survailla specimens and the frequencies for i
removing and testing these specimens are provided in the UFSAR to assure compliance with the
[
j equirements of Appendix H to 10 CFR Part 50.
r 1
Prmurim t
I L
i The limitations imposed on pressurizer heatup and cooldown and spray water temperature
{
differential are provided to assure that the pressuriseris operated within the design criteria assumed i
for the fatigue analysis performed in accordance with the ASME Code requirements.
NORTH ANNA - UNIT 1 B 3/4 4-7 Amendment No. !!?. !?O,
- 189, 4
l
REACIUR COOLANT SYSTEM BASES Imw.T;...n-... e O._,amre Pmtectinn l
The OPERABIIIIY 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 CPR Part 50 when one or more of the RCS cold legs are less than or equal to 2357. Either PORY has adequase relieving capability to protect the RCS fma
[
overpressurizadon when the transient is limited to either (1) the start of an idle RCP with the Wa y water temperature of the steam generatorless than or equal 1o 507 above the RCS cold leg temperatures or (2) the start of a charging pump and its injection into a water. solid RCS.
[
Autornatic or passive low temperstme overpressure protection (L10P) is required whenever any RCS cold leg temperstmo is less than 2359. This temperamre is the waser emperatus s. Ag to a metal temperature of at least the Ma-RTmyr + 50T +
instrument uncertainty. Above 235T adminierative commlis adequam protection to ensure the limits of the beamp curve (Figme 3.4-2) and the cooldown curve (Figme 3.4-3) are not via8M The concept of mquiring auenenarie L10P at the lower end, and adminierative control at the upper end, of the Appendix G curves is further discussed in NRC Generic Letter 88-11.
I l
l l
l l
NORTH ANNA - UNIT 1 B3/44-8 Amendmant No. 71. !!? !?0,
- 189,
. =.
.~
1 EMEROFNcy CORN (Yvy rNo Sysi egg (pcgg) g BASES i
ECCS SUBSYSTEMS (CnntinneA) l l
With the RCS temperature below 3507, one OPERABLE ECCS subsystem is i
without single failure consideration on the basis of the stable reactivity condition of the rea f
the limited core cooling requirements.
I t
he limitadon for a maximum of one centrifugal charging pump and one low head injection pump to be OPERABLE and the Surveillance Requirement to verify all ch and low head safety injection pumps except the required OPERABLE pump to be inopemb i
below 235T provides assurance that a mass addition pressure ansient can be relieved b l
l operation of a single PORV.
t l
He Surveillance Requirements provided to ensure OPERABILITY of each component i
ensures that at a minimum, the assumptions used in the safety analyses are met and that su l
OPERABILITY is maintained.
a i
In the event of modifications to an ECCS subsystem that could alter the subsystem flow l
characteristics, a flow balance test shall be performed. De flow balance test criteria are estal>.
{
lished based on the system performance assumed in the safety analysis (minimum flow lim
{
on HHSI pump runout ywwtion (maximum flow limit). In performing the flow balance, the j
effects of flow measurement instrument uncertainties accounting for system configuration a i
variability between installed pumps must be properly considered.
I
~
i Numerical acceptance criteria for the flow balance test are specified in the surveillance test
{
procedure. Hese criteria are established based on the following considerations:
I,
- 1) ne totalinjected flow to the core (assuming spillage of the branch line with the high flow) must meet or exceed that assumed in the safety analysis. He limiting safety i
analysis is the loss of coolant accident (LOCA) analysis. His criterion may vary, j
particularly since the inputs to the safety analysis controlled by LCO 6.9.1.7 may vary j
with reload cycle. %e safety analysis flow requirements are thus established by the currently applicable LOCA analysis which has demonstrated compliance with the i
ECCS acceptance limits of 10 CFR 50.46.
1 i
'l
- 2) De total pumped flow must be less than the HHSI pump runout limit. His flow varies l\\
with the specific HHSI pump assumed to operate during the accident. Since the HHSI pumps also function as normal charging pumps, their characteristics, including runout
{
limits, will vary over servicelife.
1
- 3) The requirements for reactor coolan't pump seal injection must be met daring normal
}
operation, and the effects of seal injection during accidents must be considered in l
meeting constraints 1) and 2) above.
I i
l 1
NORTH ANNA - UNIT 1 B 3/4 5-2 Amendment No. If, f* !!?, !?^, i SS i
- 189, a
t
1 1
ADMINISTRATIVE CONTROLS (Continued)
)
SPECIALREPORTS 6.9.2 Special mports shall be submitted to the Regional Administrator, Region II, within th period Whd for each report. Ihese reports shall be submitted pursuant to the requirem the applicable WMestion:
- a. Inservice Ta==adon Reviews, Specification 4.0.5, shall be hhE.M within 90 da ofcompletion.
- b. MODERATOR TEMPERA ~IURE COEFFICIENT. S,=eiM~daa 3.1.1.4.
- c. RADIATION MONITORING INSTRUMENTATION. Specification 3.33.1 Table 33-6, Action 35.
- d. SEISMIC INSTRUMENTATION. S ---:~-:='=-5 33.33 and 433.3.2.
i
- e. METEOROIAGICAL INSTRUMENTATION. S,=eine=daa 333.4.
f.
Deleted.
- g. 140SE PARTS MONITORING SYSTEMS. Specification 33.3.9.
- h. Deleted.
i.
IDW-TEMPERATURE OVERPRESSURE PROTECTION. Sg-- f- =' 1
+
3.4.93.
- j. EMERGENCY CORE COOLING SYSTEMS. Sh--:'- ='m 3.5.2 and 3.53.
- k. SETILEMENT OF Q ASS 1 STRUCIURES. Sp+47-= tion 3.7.12.
1.
GROUND WATER LEVEL - SERVICE WATER RESERVOIR. Specification 3,7,1 3,
~..
- m. Deleted.
- n. RADIOACTIVE EFFLUENTS. As required by the ODCM.
- o. RADIOIDGICAL ENVIRONMENTAL MONTIORING. As requhed by the ODCM.
i
- p. SEALED SOURCECONTAMINA110N. S=aine-daa 4.7.11.13..
- q. REACIOR COOLANT SYSTEM STRUCTURAL INTEGRITY. S,aaaiMeadon 4.4.10. For any abnormal degradation of the structural integrity of the reactor vessel
{
cr the Reactor Coolant System pressure boundary deteced during the p Tv.mance of Specification 4.4.10, an initial report shall be submitted within 10 days after detection and a detailed report submined within 90 days after the completion of Specification 4.4.10.
NORTH ANNA - UNIT 1 6 21 Amendment No. 3. 4*, fL ^',
- 189,
! ^? "^. ! '^.
l s>3 *t00 9
y UMTED STATES I
j j
j NUCLEAR REGUL.ATORY COMMISSION i
4 WASHINGTON, D.C. 20066-0001 e
l VIRGINIA ELECTRIC AND POWER COMPANY i
OLD DOMINION ELECTRIC COOPERATIVE' '
DOCKET NO. 50-339 I
NORTH ANNA POWER STATION. UNIT N0. 2 J
i j
AMENDMENT TO FACILITY OPERATING LICENSE 4
j Amendment No.170 License No. NPF-7 I
i 1.
The Nuclear Regulatory Commission (the Commission) has found that:
i l
A.
The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated April 15, 1994, complies with the j
standards and requirements of the Atomic Energy Act of 1954, as j
amended (the Act), and the Commission's rules and regulations set 5
forth in 10 CFR Chapter I; j
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the 1
Conssission; C.
There is reasonable assurance (1) that the activities authorized by 4
this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be j
conducted in compliance with the Commission's regulations; I
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 4
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 l
of the Commission's regulations and all applicable requirements have been satisfied.
i l
1 j
3 I
i
)
e, m.-
l
. 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 Soecifications 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 within 60 days.
~~
l
,FOR THE CLEAR REGULATORY COMMISSION k
Victor M. McCree Acti pg. Director Project Directordte IIQ Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation j
Attachment:
Changes to the Technical i
Specifications Date of Issuance: October 5, 1994 j
l
ATTACHMENT TO LICENSE AMENDMENT NO.170 TO FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-339 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 Paaes V
V XI XI 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 4-7a 3/4 4-7a l
3/4 4-7b 3/4 4-27 3/4 4-27 3/4 4-28 3/4 4-28 3/4 4-30 3/4 4-30 3/4 4-31 3/4 4-31 3/4 5-3 3/4 5-3 3/4 5-6 3/4 5-6 i
3/4 5-7 3/4 5-7 l
B 3/4 1-3 8 3/4 1-3 B 3/4 4-1 B 3/4 4-1 B 3/4 4-2 8 3/4 4-2 B 3/4 4-2a B 3/4 4-6 B 3/4 4-6 B 3/4 4-7 B 3/4 4-7 B 3/4 4-8 B 3/4 4-8 B 3/4 5-2 8 3/4 5-2 l
6-21 6-21 l
i I
f mur.A
+
i4 LIMITING CONDITIONS FOR OPERATION AND SURVEnI ANCE REQUIR SEC1'10N I
MCig 4
3/4.4.2 SAFETY VALVES-SHU1DOWN...
3/4 4 6 I
E l
3/4.4.3 SAFETY AND RELIEF VALVES-OPERATING l
)
Safety Valves 3/4 4-7 Relief Valves 3/4 4-7a 4,
i j
3/4.4.4 PRESSURIZER.
3/4 4 8 4,
3/4.4.5 STEAM GENERATORS..
3/4 4 9 V4.4.6 REACTOR COOLANTSYSTEM LEAKAGE 1
leakassDemetion Sysmas 3/4416 j
Operarianal Leakage 3/4417 i
Primary to SGf >=b:=
3/4 418b T
Primary to h - '=rf >=h== Detection Sysmas
_.-- - W4 418d 1
1 I
3/4.4.7 CHEMISTRY 3/4 4-19 1
m j
3/4.4.8 SPECIFICAC11VITY 3/44-22
\\
j 3/4.4.9 PRESSURF/IEMPERATURE LIMITS I
ReactorCoolantSyimm 3/44 26 1
Pressuriser 3/4 4 29 low-Temperarme C-ph Protection 3/4 4 30 t
j 3/4.4.10 STRUCIURALINTEGRITY ASME Code Class 1,2 & 3 Components.
.~
3/4 4 32
[
3/4.4.11 REACIORVESSELHEADVENT 3/4 4-34 3/11 EMERGRNCY mRR COOUNG SYS1TMS (ECCS)
J 3/4.5.1 ACCUMULATORS 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS -T GREA*1ER *IHAN 350'F 3/4 5-3 1
ses 3/4.5.3 ECCS SUBSYS'IEMS - T,, LESS THAN 350'F 3/4 5 6 4
3/4.5.4 BORONINJECTION SYSTEM BaronInjectionTank 3/4 5 8 HeatTracing 3/4 5 9 i
3/4.5.5 REFUELING WATER STORAGE TANK 3/4 510 J
j NORTH ANNA - UNIT 2 y
Amendment No. IB M.
- 170, 1
i INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 CONTAINMENT l
Containment Integrity.....................................
3/4 6-1 Containment Leakage......................................".'
3/4 6-2 Containment Air Locks.....................................
3/4 6-4 1
l Internal Pressure................_.........................
3/4'6-6 Air Temperature...............
3/4 6-8 Containment Structural Integrity..........................
3/4 6-9 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS
~
(
Containment Quench Spray System...........................
3/4 6-10 Containment Recirculation Spray System....................
3/4 6-11 Chemical Addition System..................................
3/4 6-13
)
3/4.6.3 CONTAINMENT ISO LATION VALVES..............................
3/4 6-14 3/4.6.4 COMBUSTIBLE GAS CONTROL l
Hydrogen Analyzers..................................
3/4 6-32 Electric Hydrogen Recombiners.............................
3/4 6-33 l
Waste Gas Charcoal Filter System..........................
3/4 6-34 l
3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM i
Steam Jet Air Ejector...................................
3/4 6-36 l
l VI l
NORTH ANNA - UNIT 2 l
i i
i
=
m mu 4
BASES SECTION PAGE 4
3/4,1 INSTRUMENTATION t
i 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY j
FEATURES (ESF) INSTRUMENTATION B 3/4 3-1 5
3/4.3.3 MONITORINGINSTRUMENTATION B 3/4 3-1 2
l 34(d MCTOR CODI ANTSYSTEM 3/4.4.1 REACIOR COOLANTLOOPS
. B 3/4 4-1 l
3/4.4.2 and 3/4.4.3 SAFETY AND RELIEFVALVES B 3/44 2 j
3/4.4.4 PRESSURI2ER
_ B 3/4 4 2a l
2 1
3/4.4.5 STEAM GENERATORS _
.. B 3/4 4 3 1
1 3/4.4.6 REACIOR COOLANT SYSTEM LEAKAGE B 3/44-4 j
3/4.4.7 CHEMISTRY B 3/44 5 i
3/4.4.8 SPECIFIC ACI1VITY.
B 3/44-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS B 3/44 6
- e j
3/4.4.10 STRUC113RALIN1EGRITY B 3/4 4 8 J
3/4.4.11 REACIOR VESSEL HEAD VENT B 3/4 4-17 e
i s
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4 k
1 s
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1 1
4 e
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)
i h
NORTH ANNA -UNIT 2 XI Amendment No. 40, i
- 170, I
l t
I i
REACTIVITY CONTROL SYSTEMS 4
l 1
FLOW PATHS-OPERA *ITNG i
LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following baron injection flow paths shall be OPERABM:
I The flow path from the boric acid tanks'via a $3ric acid transfer pump and a a.
)
charging pump to the Reactor Coolant Sysem.
=
~
j b.
Two flow paths from the refueling waar storage tank via charging pumps to the Reactor Coolant Sysem.
APPLICABILITY:
MODES 1,2,3 and 4'.
l
}
ACTION:
With only one of~the'above recinimd boson injection flow paths to the Reactor Coolant Syst
)
{
OPERAB2, restore at least two baron injection flow paths to the Reactor aalant Sysum to r
l OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HCYr STANDBY and barased to a i
SHU'IDOWN MARGIN equivalent to at least 1.77% dels k/k at 200T within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; 1
restore at least two flow paths,to OPERABLE stams within the next 7 days or be in COI.D
~1
)
SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
j l
SURVEILLANCE REQUIREMENTS 1
i j
4.1.2.2 Each of the above r@id flow paths shall be demonstrated OPERABLE:
5 i
At least once per 7 days by verifying that the temperstme of the heat traced portion a.
l of the flow path from the boric acid tanks is greater than or equal to 115T when it i
j is a requimd water source.
t l
l 1
l 1
1 1
Only one baron injecdon flow path is required to be OPERABLE wk ; vs the j
temperature of one or more of the RCS cold legs is less than or equal to 2707.
[
d 4
}
i NORTH ANNA-UNIT 2 3/419 Amendment No. "',129, !!?,
o
- 170, i
l l
l REAC"*VITY CONTROL SYSTEMS l
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 shutdown by verifying that each automatic valve in the" flow path actuates to its correct position on l
a safety injection test signal.
I 4
l l
NORTH ANNA - UNIT 2 3/4 1-10
i REACTIVITY CONTROL SYSTEMS f
CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION l
l 3.1. 2. 3 One charging pump in the boron injection flow path required by i
Specification 3.1.2.l'shall be OPERABLE.
l APPLICABILITY: MODES 5 and 6.
i
~
ACTION:
a.
With no charging pump OPERABLE, suspend.all. operations involving, CORE I
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 OPERABLE status as soon as possible.
l SURVEILLANCE REQUIREMENTS 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 pressure of greater than or equal to 2410 psig when tested pursuant to l
Specification 4.0.5.
4.1. 2. 3. 2 All charging pumps, except the above required OPERABLE pump, l
shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the control switch is in the pull to lock position.
l NORTH ANNA - UNIT 2 3/4 1-11 Amendment No. 5
\\
i f
i REACTTVTTY CONTROL SYSTEME 1
}
CHARGING PUMPS - OPERATING k
LIMITING CONDITION FOR OPERATION 4
i
'!.1.2.4 At least two charging pumps shall be OPERABLE.
APPLICABILITY:
MODES 1,2,3 and 4e,,.
m....
g
{
ACTION:
j With only one charging pump OPERABM, restore a second charging pump to OPERABM sta i
within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and boissed to a SHUTDOWN
~
)
equivalent to at least 1.77% delta k/k at 2007 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore a second i
pump to OPERABM status within the next 7 days or be in COL.D SHUIDOWN.within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. *Ihe provisions of Sp+S"-:=h 3.0.4 are not applicable for one hour following b 4
j above 270T or prior to cooldown below 270T...
l
,u i
SURVEILLANCE REQUIREMENTS 1
3 4.1.2.4.1
'Ibe abover%44 charging pumps shall be danaastraaed OPERABM by verifying, that on recirculadon Gow, each pump dss @ a t'ischarge pressure of greater than or equ l
2410 psig when tesed pursuant to Sr+-:"=' = 4.0.5.
j 4.1.2.4.2 All charging pumps, except the above required OPERABM pump, shall be j
demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperanse of one or more o
{
RCS cold legs is less than or equal to 2707 by verifying that the control switch is in the p l
lock position.
[
l l
i i
i t
i i
i 1
s 1
i i
i A maximum of one centrifugal charging pump shall be OPERABM whenever the
{
temperature of one or more of the RCS cold legs is less than or equal to 2707.
[
1 i
i i
i NORTH ANNA -UNIT 2 W41-12 Amendmant No. 449,
- 170, i
I.
i
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION
=
LIMITING 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.*
i ACTION:
~
With less than the above required reactor coolant loops in operation, be in at i
least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
1 I
SURVEILLANCE REQUIREMENTS 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- 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.
1 i
- See Special Test Exception 3.10.4.
l NORTH ANNA - UNIT 2 3/4 4-1 i
t.
i_.
i
(
4 i
{
HOTSTANDBY LIMITING CONDITION FOR OPERATION j
3.4.1.2 a.
At least two of the reactor coolant loops listed below shall be OPERABLE:
i i
- 1. Reactor Coolant I. mop A and its associased steam generator and reactorcoolant
- pump,
[
}
j
- 2. Reactor Coolant I.oop B and its menacinewl saeam generator and reactor coolant
- Pump,
[
l
- 3. Reactor Coolant Imop C and its==aaria=wl steam generator and reactor coolant 4
- Pump, f
b.
At least one of the above coolant loops shaR be in operation.*, **
.f APPLICABILITY:
MODE 3
}
ACTION:
s
' With less than the above required loops OPERABE, restore the required loops L
OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within 3
3
[_
{
hours.
~
b.
With no coolant loop in operation, suspend au operations involving a reduction in i
baron -nation of the Reactor Coolant System and i====tiamly initiate corrective actions to return the required coolant loop to operation.
4 J
SURVEILLANCE REQUIREMENTS 1
4
{
4.4.1.2.1 Atleast the above r%C4 reactor coolant punps, if not in operation, shall be determined to be OPERABE once per 7 days by ve;ifying correct breaker aligenents and indicated poweravailability.
l 4.4.1.2.2 At least one cooling loop shall be verified to be in operation and circulating coolant l
atleast once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i, i
i j
All reactor coolant pumps may be des 4 1 for up to I bour provided (1) no operations l
{
are permitsed that would cause dilution'of the reactor coolant system baron concentration, j
and (2) core outlet temperanne is maintained at least 10*F below sannation temperature.
j The requirement to have one eaalmar loop in operation is exempted during the i.' mancel j
of the baron mixing nests as stipulated in License Condition 2.C(15Xf) and 2.C(20Xb).
i i
I NORTH ANNA - UNIT 2 V4 4-2 Amendment No. !?, !'?.
- 170,
,2 -
a e---,-,
+e.-esr.v--o-
-ve,i-..,---yv=+-=re-rr y g pr-y + er,-+
w
,-p g w i r-,
wei
I REACTOR COOLANT SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a.
At least two d the coolant loops listed below shall be OPERABG:
- 1. Reactor Coolant 1. cop A and its asaarianad seen generator and reactorcoolant
- Pump,
- 2. Reactor Coolant loop B and its menacimaad steam generator and reactor coolant Pump,*
- 3. Reactor Coolant 1. cop C and its anencinead steam generator and reactor coolant Pump,*
- 4. ResidualHeat Removal Subsysam A,**
- 5. Residual Heat Removal Subsysem B.**
b.
At least one of the above coolant loops shah be in operation."*-
APPLICABH rrY-MODES 4 and 5 ACDON:
With less than the above required loops OPERABG, iminadiately initinta a.
corrective action to return the rW-loops ao OPERABG stams as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
b.
With no coolant loop in operation, ne all operations involving a reduction in baron caarentration of the Reactor antant System and isnmediately initiate c
corrective action so return the requimd coolant loop to operation.
9 A reactor coolant pump shall not be starand with one or more of the RCS cold leg temperatures less than orequal to 270'F unless the =a~=d+ y water temperature of each j
steam generator is less than 50*F above each of the RCS cold leg temperatures.
"Ibe offsite or emergency power source ma; be inoperable in MODE 5.
"* All reactor coolant pumps and residual heat removal pumps may be de energized for up to I hour provided 1) no operations me perminnd that would cause dilution of the reactor coolant sysem baron cancentration, and 2) core outlet temperarme is maintained at least 10*F below samration temperanse.
NORTH ANNA - UNIT 2 3/44-3 Amendmant No. 449,
- 170,
I i
REACTOR COOTANT SYSTEM SAFETY AND RFT TFN VALVES - OPERATING
[
RELIEF VALVES i
LIMITING CONDITION FOR OPERATION l
3.4.3.2 Both powerw d relief valves (PORVs) and their associated block valves shall be
[
i
{
APPLICABILITY:
MODES 1,2, and 3.
a j
ACTION:
With one or both PORVs inoperable but capable of being manually cychd, within a.
I hour either restme the PORV(s) to OPERABM stams or close tbc unociated block valve (s) with power maintained so the block valve (s); otherwise, be in at least l
HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in IKYT SHUTDOWN within the l
following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
l b.
With one PORV inoperable and not capable of being manually cycled, within 1 j
hour either restore the PORV to OPERABIE stams or capable of being maanally j
cycled, or close its menaci==4 block valve and remove power from the block valve; restore the PORY to OPERABM status within the foBowing 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the fouowing 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
With both PORVs inoperable and not capable of being manually cycled, within 1 c.
hour either restore at least one PORV to OPERABE stams or capable of beirg manually cycled, or close its===aciami block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the fouowing 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
d.
With one or both block valves inoperable, within I bour restore the block valve (s) to OPERABLE status or place its associased PORV(s) in manual control Restore at least one block valve to OPERABLE stams within the next hour if both block valves are inoperable; restost the remaining inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, He provisions of S,'=ri&= tion 3.0.4 are not applicable.
e.
1 NORTH ANNA - UNIT 2 3/447a Amendunt No. 170,
e i
i i
REACIDR COOLANT SYSTEM SAFETY AND RNT INE VALVES - OPERATWG 4
RFT TNN VALVER i
SURVFTLLANCE REQUIREMENTS
)
4.4.3.2.1 In addition to the mquirements of S,M&mtion 4.0.5, each PORV shall be j
demonstrated OPERABE:
I At least once per 31 days by performing a CHANNEL FUNCDONAL TEST, a.
excluding valve operation, and.
b.
At least once per 18 months by:
1.
Operating the PORV through one complete cycle of full travel during MODES 3 or 4, and 2.
Operating the solenoid air control valves and check valves on the associated accumulators in the PORV control systems through one complete cycle of full travel, and 4
3.
Performing a CHANNEL CALIBRATION of the actuation instrumentation.
4.4.3.2.2 i
Each block valve shall be & = =
.4 OPERABM at least once per 92 days by l
operating the valve through one complete cycle of fall travel unless the block valve is closed in order to meet the requirements of ACDON b or e in Specification 3.4.3.2.
NORTH ANNA - UNIT 2 3/4 4-7b Amendment No. 170, I
i
l REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with at least 125 kw of pressurizer heaters and a water volume of less than or equal to 1240 cubic feet.
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
a.
With the pressurizer inoperable due to an inoperable emergency power supply for the pressu.'izer heaters either restore the inoperable emergency power supply within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY I
within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With the pressurizer otherwise inoperable, be in at least HOT STAN08Y with the reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
2 1
I SURVEILLANCE REQUIREMENTS l
l l
4.4.4.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i l
l NORTH ANNA - UNIT 2 3/4 4-8 Amendment No.
=_
i 1
i 1
Figure 3.4 North Anna Unit 2 a
t Reactor Coolant System Heatup Limitations 3
j Material Property Sesis Umiting Material: Lower Shell Plate Umiting ARTat 17 EFPY: 1/4-T,196 F
^
3/4-T,172 F 4
{
Healup Rates (FAw) 2500.00 m e ao
/
i J/
i j
/
li Laek Test umisl f
ff/
i 2000.00 f) 4 l
0
}
$ 1500.00
}{
A j
l f,f t
U,=:-~ _ ; : ' '
/ /
a opermison I
//
5
) /
j 1000.00
'/ /
l
/,,
Acceptable==
g operation f /
t 4
?)'
x j
Heeme neeen A Y/
(F/hrt 4
r p
f 500.00 20 go i
i 1
z 0.00 l
i O
50 100 150 200 250 300 350 400 1
)
cold Leg Tempereews (Deg. F)
North Anne Unit 2 Reecaer Coeient Syeese Heene Umlessons (Hesse Rates up to 60 F/hr) aem for wie Firet 17 EPPY (WIshout Maryne for inern_ - _. Erreral t
i 1
NORTH ANNA - UNIT 2 V4 4-27 A~h No. '". M9 t
- 170, i
i l
Figure 3.4 North Anna Unit 2
{
Reactor Coolant System Cooldown Umitations l.
1 Material Property Basis Umiting Meterial: 1.ower Shall Plate Umhing ART st 17 EFPY: 1/4-T,196 F 3/4-T,172 F 1800.00 i
I 1
l 1600.00 I
/
~
l l
k I
I j
34=.=
/
/
I' i
j 1200.00
}
/
4 i
e W^~
1
~
j 1000.00 Operation O.
j w
j 800.00
'^ '
A fL 3
ffA f a
Jfif>'
t e
/ rar i
E 600.00 s
eas m. __ _
s s
/ / r
/ r/
r gp y
}
s-
/
_A a
1
==i 20
(( [E j
400.00 =
Opersuon Z[
s g
n 1
too i
200.00 t
i i
{
0.00 1i 0
50 100 150 200 250 300 350 1
Cold Leg Temperature (Deg. F) w Aans una a neesier ca nene sveism cosmo n ummonene sc smo.n ames = =
100 F/hr) AssheeMe for the Pket 17 EPW (Whhout Mergne for Insensnonissen Erreral i
NORTH ANNA - UNIT 2 3/4 4-28 Amendment No. 9, ! 49, l
- 170,
.1
~
l 4
i PRESSURIZER i
l LIMITING CON 0! TION FOR OsERATION i
i 3.4.9.2 The pressurizer temperature shall be limited to:
i a.
A maximum heatup of 100*F or cooldown of 200*F, in any one hour i
period, and b.
A maximum spray water temperature and pressurizer temperature l
differential of 320'F.
APPLICA8!LITY: 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 1
on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the pressurizar pressure to less than j
500 psig within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
l
{
SURVEILI.ANCE REQUIREMENTS I
4.4.9.2 The pressurizer temperatures shall be deters med to be within the limits at least once per 30 minutes during systes heatup or cooldown. The
{
spray water temperature differential shalT be determined to be within the l
limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during auxiliary spray operation.
i l
i i
NORTH ANNA - UNIT 2 3/4 4-29
,-.we-v-wg-
e p-==*Wt
'-9
i J
i REAC7DR COOLANT SYSTEM i
t nW-TEMPERATURE OVERPR FMSURE PROTTCTION l
LIMITING CONDITION FOR OPERATION i
3.4.9.3 Two power-operated relief valves (PORVs) shall be OPERARM with lift somngs i
i (1) less than or equal to 415 psig whenever any RCS cold leg sernparanne is less than o 270*F, and (2) less than or equal to 375 psig whsi any RCS cold leg temperanne is less than 130*F.
{
j APPLICABHJTY-j MODE 4 when the temperature d any RCS cold leg is less than or equal to l
270'F, MODE 5, and MODE 6 when the head is on the reactor vessel and.
i the RCS is not vensed through a 2.07 square inch or larger vent.
1 4
j ACTION:
m,.
l j
With one PORV inoperable in MODE 4, restore the inoperable PORV to l
a.
l OPERABLE status within 7 days or depressuriae and vent the RCS through at least
{
a 2.07 square inch vent within the next 8 bours.
j b.-
i With one PORV inoperable in MODES 5 or 6, either (1) restore the inoperable i
PORV to OPERABE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or (2) comples depressurization and venting of the RCS through at least a 2.07 square inch vent within a total d32 l
j hours.
With both PORVs inoperabic, complete depressurization and venting of the RCS c.
j through at least a 2.07 square inch vent within 8 heurs.
}
d.
With the RCS vented per ACTIONS a, tr, orc, verify the vent pathway at least once i
per 31 days when the pathway is provided by a valve (s) that is locked, scaled, or I
j otherwise secured in the open position; otherwise, venfy the vent pathway ever 12 i
hours.
i j
In the event either the PORVs or the RCS vent (s) are used to mitigase an RCS e.
pressure transient, a Special Report shall be por 4 and submined to the
{
Commission pursuant to Specification 6.9.2 within 30 days. 'Ihe repest shall j
describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient, and any corrective action r=+as y to prevent recunence.
i f.
"Ihe provisions of Specification 3.0.4 are not applicable.
I l
1 i
1 i
i i
NOR'III ANNA -UNIT 2 3/4 4-30 Amendment No. 9, !'a.
i i
- 170,
REACTOR COOLANT SYSTEM ILW-TEMPERATURE OVERPR FRSURE PROTEcrION g
SURVEILLANCE REQUIREMENTS 4.4.9.3 Each PORV shall be demonstrated OPERABLE by:
[
Performance of a CHANNEL PUNC110NAL TESTon the PORV actuados a.
channel, but excluding valve operation, within 31 days prior to enering a condidon in which the PORV is required OPERAB12 and at least once per 31 days thereaher when the PORVis required OPERABLE.
b.
Performance of a CHANNEL CALIBRATION on the PORV acmarian channel, at least once per 18 months.
Verifying the PORY keyswitch is in the AUTO posidan and the PORV (aniarin.
l l
c.
valve is open at least onos per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpressure proaction.
d.
Testing pursuant to S : :~- +=- 4.0.5.
i
.u
\\
l 4
s NORTH ANNA - UNIT 2 3/4431 Amendment No. 170,
{
1 r-+-.
r.-
w.,
_.--m.__,
w
4 R_EACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 & 3 COMPONENTS LIMITING CONDITION FOR OPERATION 1
3/4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.1.
1 APPLICABILITY: ALL MODES ACTION:
a.
With the structural integrity of any ASME Code Class 1 component (s) i not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limits or isolate the affected component (s) prior to increasing the Reactor j
Coolant System temperature more than 50*F above the minimum temperature required by NDT considerations.
b.
With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200'F.
c.
With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or l
isolate the affected component (s) from service.
l d.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.10.1.1 In addition to the requirements of Specification 4.0.5, the l
Reactor Coolant pump flywheels shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14 Revision 1 August 1975.
4.4.10.1.2 In addition to the requirements of Specification 4.0.5, at least one third of the main member to main member welds, joining A572 material, in the steam generator supports, shall be visually examined during each 40 month inspection interval.
l NORTH ANNA - UNIT 2 3/4 4-32 Amendment No. 40 i
l l
.,m.
i i
EMERGENCY CORE COOLING SYSTEMS i
i ECCS SUBSYSTEMS -Tavr GREATER THAN 350*F i
i LIMITING CONDITION FOR OPERATION
'i l
i j
3.5.2 Two ir '+;- ='=; ECCS subsystems shall be OPERABM with each subsystees j
compnsed of:
a i
a.
One OPERABLE centrifugal charging pump, b.
One OPERABLE low head safety injection pump, i
c.
An OPERABE flow path capable of transferring fluid to the Reactor Coolant System when taking suction from the refueling water storage tank on a safety l
injection signal or from the caneninmaar.nunp when suction is transferred dunng the recirculation phase of operation.
APPLICABILITY:
MODES 1,2 and 3.
1
}
ACTION:
!j a.
With one ECCS subsystem inoperable, ressore the inoperable F4+y._ to j
OPERABG status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOTSIMIDOWN within the next 12 1
b.
In the event the ECCS is actuated and injects water inte the Reactor Coolant j
System, a Special Report shall be prepared and submined to the amminnian c
j j
pursuant tr, Saarisertion 6.9.2 within 90 days describing the circumstances of the j
actuation and the notal accumulased actuados cycles so dam. 'Ihe curent value of j
the usage factor for each a5ected safety injection nozzle shall be provided in this j
Special Report whencycr its value exceeds 0.70.
{
c.
The provisions of Sp='"=fna 3.0.4 are not applicable to 3.5.2.a and 3.5.2.b for i
l one hour following heatup above 270T or prior to cooldown below 270".
[#
SURVEILLANCE REQUIREMENTS 3
4 j
4.5.2 Each ECCS subsystem shall be demonstrated OPERABM:
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the following valves are in the indicated a.
4 l
positions with power to the valve operators removed:
5 i
)
i i
1 1
?
\\
i i
1 NORTH ANNA - UNIT 2 3/45-3 A" cad"*"' No. 449, i
- 170, j
j
.. L
4 i
j I
i EMERGENCY CORE COOLING SYSTEMS i
j SURVEILLANCE REQUIREMENTS (Continued)
Valve Number Valve Function Valve Position
]
- a. MOV-2890A
- a. LHSI to hot leg
- a. closed j;
- b. MOV-28908
- 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 j
- e. MOV-28698
- e. Ch pump to hot leg
- e. closed i
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, j
sealed, or otherwise secured in position, is in its correct position.
By a visual inspection which verifies that no loose debris (rags, c.
j trash, clothing, etc.) is present in the containment which could be j
transported to the containment sump and cause restriction of the pump suctions during LOCA conditions.
This visual inspection shall be performed:
1 1.
For all accessible areas of the containment prier to establish-l ing CONTAIMENT INTEGRITY, and i
2.
Of the areas affected within containment at the completion of
]
each containment entry when CONTAINMENT INTEGRITY is established.
l d.
At least once per 18 months by:
3 1.
A visual inspection of the containment sump and verifying that j
the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no t
evidence of structural distress or corrosion.
l At least once per 18 months, during shutdown, by:
e.
I 1.
Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.
1 2.
Verifying that each of the following pumps start automatically upon receipt of a safety injection test signal:
i a)
Centrifugal charging pump, and i
b)
Low head safety injection pump.
4 j
NORTH ANNA - UNIT 2 3/4 5-4 i
Y
= - - - - - - - -
I l
]
EMERGWNCY CORE COOLING SYSTEMS l
ECCS SUBSYSTEMS-Tave i MS MAN 350*F i
LIMITING CONDITION FOR OPERATION j
i 3.5.3 l
As a minimum, one ECCS subsysum comprised of the following shall be OPERABLE:
1 l
One OPERABM centrifhgal charging pump',
a.
b.
One OPERABG low head safety injection panp*, and 1
c.
An OPERABG flow path capable of automatically transferring fluid to the rescar j
coolant system when taking suction froen the refueling water storage tank or from j
the containmaat sump when the suction is transferred during the recirculation phase i
ofoPemtion.
s j
APPLICABILITY:
MODE 4.
j
]
ACTION:
1 With no ECCS subsystem OPERABM because of the inoperability of either the -
s.
l centrifugal charging pump or the flow path Bom the refheling water storage tank, i
restore at least one ECCS subsystem so OPERABM stams within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in i
COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
b.
With no ECCS subsysem OPERABM because of the inopambility of the low head i
safety injecolon pump, restore at least one ECCS subsysam so OPERABM status j
or maintain the Reactor Coolant System T,gless than 350*F by use of alesmaan j
heat removal methods.
c.
In the event the ECCS is actuated and injects water into the Reactor raalant i
Sysem, a Special Report shall be prepared and sahminad o the a==ineian t
c i
pursuant to SMW 6.9.2 within 90 days describing the cineumstances of the i
acmation and the total accumulased mernarian cycles to dam. 'Ibe current value of
~
j the usage factor for each affected safety injection nozzle shall be provided in this i
Special Report whenever its value exceeds 0.70.
i i
i I
i A maximum of one centrifugal charging pump and one low head safety injection pump j
shall be OPERABM whenever the temperature of one or more of the RCS cold legs is less l
than orequal to 270*F.
l 1
i I
NORTH ANNA - UNIT 2 3/45-6 Amend =ane No. ?!,19, i
- 170, i
1
i 1
i EMERGENCY CORE COOf3NG SYSTEMS i
i SURVEILLANCE REQUIREMENTS e
4.5.3.1
'Ibe ECCS subsystem shall be demonstrated OPERABLE per the applicable Survei11m-Requirements of 4.5.2.
4.5.. !
All charging pumps and safety injection pumps, except the above required OPERABLE pumps, shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperamre of one or more of the RCS cold legs is less than or equal to 270*F by verifying that the control switch
[
i is in the pull so lock position.
s i
l i
i i
i 1
4 i
1 i
i 4
i t
a
.-l 1
i I
i i
i 4
I i
i NORTH ANNA - UNIT 2 N4 5-7 Amendment No. 44,
- 170, Y
o e
EMERGENCY CORE COOLING SYSTEMS 1
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.
A contained borated water volume of at least 900 gallons.
er b.
Between 12,950 and 15,750 ppm of boron, and g
c.
A minimum solution temperature of 115'F.
l APPLICABILITY: MODES 1, 2 and 3.
l l
ACTION:
i With the boron injection tank inoperable, restore the tank to OPERABLE status t
within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to 1.77% ak/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the tank to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
t SURVEILLANCE REQUIREMENTS 4.5.4.1 The boron injection tank shall be demonstrated OPERABLE by:
a.
Verifying the contained borated water volune at least once per 7 i
- days, b.
Verifying the boron concentration of the water in the tank at least once per 7 days, and i
c.
Verifying the water temperature at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
NORTH ANNA - UNIT 2 3/4 5-8 Amendment No. 54 r
1 3
l REACTTVITY CONTROL SYSTEMS I
BASES l
1 g
BORATION SYSTEMS i
i The baron injection system ensures that negadve reacdvity conaal is available during ea mode of facility operadon, The components required so perforta this function include 1) borated j
water sources,2) charging pumps,3) separam flow paths,4) boric acid transfer pumps,5)
{
===arimwd heat tracing systems, and 6) an emergency power supply frans OPERABLE diesel j
generators.
I j
With the RCS average temperature above 200*F, a minimum of two baron injection flow j
paths are requimd to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The baration capability of either flow path is suMici,me to provide a SHUTDOWN MARGIN from *=p=~a peration conditions of 1.775 delta lvk after zenon decay o
j and cooldown to 200*F. The maximum eW baration capability requirement occurs at EOL, j
from full power TNium xenon conditions and requires 6000 gallons of 12,950 ppm barated j
water from the boric acid storage tanks or 54,200 gallons of 2300 ppm barased water from the j
refueling water storage tank.
4' With the RCS temperanse 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 ALTERA*I1ONS and positive reactivity change in the event the single injection system becomes inoperable.
The limitanon for a marimum of one connifugal charging pump so be OPERABLE and the Surveillance Requirement to verify all charging pumps except the W OPERABLE pump to j
be inoperable below 270*F provides assurance that a mass addition pressure transient can bel l
relieved by the operation of a single PORV.
i l
~Ihe boron capability required below 200*F is sufBeient to provide a SHUTDOWN j
MARGIN of 1.77% delta lvk after zenon decay and cooldown from 200'F to 140*F. This condition requires either 1378 gallons of 12,950 ppm barated water from the boric acid storage tanks or 3400 gallons of 2300 ppm borated water from the refueling water storage tank.
i I
l 4
a NORTH ANNA - UNIT 2 B3/41-3 Amendment No. 51. ?!. ! M, I
- 170, f
PEACTIVITY CONTPOL SYSTEMS BASES 1
3/4.1.2 B0 RATION SYSTEMS (Continued) i The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. The OPERABILITY of one boron injection system during REFUELING insures that this system is available for reactivity control while in MODE 6.
The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.7 and 9.0 for the solution recirculated within the I
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.
a 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'(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 analyses.
OPERABILITY of the movable control assemblier is established by observing rod motion and determining that rods are positioned within + 12 steps (indicated position) of the respective demand step counter position. The OPERABILITY of the individual rod position indication system is established by appropriate periodic CHANNEL CHECKS, CHANNEL FUNCTIONAL TESTS, and CHANNEL CALIBRATIONS. OPERABILITY of the individual rod position indicators is required to determine control rod position and thereby ensure compliance with the control rod alignment and insertion limits. The OPERABLE condition for the individual rod position indicators is defined as being capable of indicating 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 one hour stabilization in every 24 period (thermal " soak time") to allow stabilization of known thermal drift in the individual rod position indicator channels during which time the indicated rod position may vary from demand position 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 both a continuous one hour period or one consisting of several discrete intervals. During this stabilization period, greater reliance is placed upon the demand position indicators to determine rod position.
In addition, the + 24 step / hour limit is l
not applicabla when the control rod position is known tot e greater than 12 steps from the rod group step counter demand position indication. Above 50 percent of RATED THERMAL POWER, rod motion is not expected to induce thermal transients of sufficient 4
i magnitude to exceed the individual rod position indicator instrument accuracy of
+ 12 steps. Comparison of the demand position indicators to the bank insertion limits with verification of rod position by the individual rod position indicators (after 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 '>e varied within the interval of 225 to 229 steps withdrawn, inclusive. This interval permits 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 HITHDRAWN i
position within this band are administrative 1y controlled, using the rod insertion limit operator curve.
NORTH ANNA - UNIT 2 B 3/4 1-4 Amendment No.121,I29, 133
_ _ _. _ _ _ _ _ _. ~ _. _ _ _. _
i jfM REACTOR COOLANT SYSTEM 3
{
}
BASES l
j 34(1],
REACTOR COOLANT Lf0PS j
ne plant is designed to operate with all reactt r coolant loops in cperation and maintain the DNBR above design limit during all normal opersdons and andcipated transients. In MODES 1 i
and 2 with one reactorcoolant loop not in opemnon, this ci+-T ~=% requires that the plant be in atleast HOT STANDBY within I 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 l
)
In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufBeient heat removal e===MHty for removing decay heat, but single falhas canaideraticas require that at least i
two loops be OPERABLE. Thus, if the reacsor coolant loops are not OPERABIE, this
=g=r_ ='+3 requires two RHR loops so be OPERABLE. -
After the reactor has shutdown and ensemd into MODE 3 forat least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, a minimum RHR~ system flow rate of 2000 gym in MODE 5 is permited, provided there is sufBcient decay heat removal so maintain the RCS temperanse less than or equal to 1407. Since the decay heat power pr*edaa rate decreases with time after reactor shutdown, the requirements for RHR t
system decay heat zernoval also decrease. Adequase decay heat removal is provided as long as the i
reactor has been shutdown for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after enay into MODE 3 and RHR flow is
}
sufficient to maintain the RCS temperanne less than or equal to 140T. He educed flow rase providr.: additional mar 3m to vortexing at the RHR pump suction while in Mid Imop Operation.
During a readaa_ in reactor cnalant system baron cancanration the Saaameadaa 3.1.1.3.1 requirement to maintain 3000 gym flow rate provides sufBeient coolant circulation to minimiw the j
effect of a baron dilution incident and to prevent baron stratification.
i ne restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less l
than or equal to 2707 are provided to prevent RCS pressure transients, caused by energy additions l
from the secondary system which could exceed the limits of A A G to 10 CFR Part 50. The i
RCS will be gi.d against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the ir-ud.iy water emperanne of each steam l
l generator is less than 507 above each of the RCS cold leg temperstmes.
J l
The requirement to maintain the baron caarennation of an isolated loop grenser than or j
equal to the baron concentration of the operating loops ensues that no reactivity addition to the i
core could occur during sartup of an isolased loop. Verification of the boren cancennstion in an j
idle loop prior to opening the cold leg stop valve provides a reassurance of the adequacy.of the
{
boron concentration in the inalmenti loop. Operating the isolaned loop on recirculating flow for at i
least 90 minutes prior to opening its cold leg stop valve ensures adequate mixing of the coolant in j
this loop and prevents any reactivity effects due to baron concentration str'=dneariana Startup of an idle loop will inject cool water from the loop into the core. The reactivity l
transient resulting from this cool water injection is minimind by delaying isolated loop starmy i
until its temperature is within 209 of the operating loops. Making the reactor subcritical prior to
[
loop startup prevents any power spike which could result from this cool waer intfurari reactivity transient.
I i
i i;
NORTH ANNA - UNIT 2 B3/44-1 Amendment No. I'^. !"
- 170, i
i i
BASES i
3/4.4.2 AND 3/4.4.3 SAFETY AND RFr rNN VALVES
\\
The pressurizer code safety valves operam to prevent the RCS from being psessurized above its Safety Limir 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 cmy.ssure 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 -
operand relief valves (PORVs) will provide overpressure relief capability and will prevent RCS overpressurization.
i i
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 grenser than the maximurn surge rate resuking from a complete loss ofload
{
assuming no reactor trip until the first Reactor Prosective System trip a..spoint is reached (l.c., no l
credit is taken for a direct reactor trip on the loss ofload) and also assuming no operation of the -
power operased relief valves or steam dump valves.-
Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in ahdsis with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.
l The power cg.
d relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurimer code safety valves. Each PORV has a remotely operased block valve to provide a positive shutoff eaa Mity should a relief valve harame inoperable.
The OPERABILITY of the PORVs and block valves is determined on the basis of their being capable of performing the following faaaaans.
i a) Manual control of PORVs to control reactorcoolant system pressure. This is a function that may be used to mitigate certain accidents and for plant shundown.
b) Mainemining the integrity of the reactor coolant pressure boundary. 'Ihis function is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.
c) Manual control of the block valve to (1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item a, above), and (2) iso-late a PORV with excessive seat leakage (Item b, above).
NORTH ANNA -UNIT 2 B3/442 Amendment No. 434,
- 170,
i i
j 3/.L4 REACfDR COOLANT SYSTEM
]
i BASES i
)
d) Aummatic control of PORVs to control reactor coolant system pressure. This fw.;dcii reduces challenges to the code safety valves for overpressurization events.
)
e) Manual consol of a block valve to isolate a stuck open PORV.
i Survailla=* Requirements provide the assurance that the PORVs and block valves can 4
perform their functions. SMAcation 4W.2.1 addresses the PORVs and Specification 4.4.3.2.2 5
addresses the block valves. The block valves are exempt from the surveillance requirements to j
cycle the valves when they have been closed to comply with the ACITON requirements. This p:ecludes the need to cycle the valves with full system differential pressure or when mainenance l
is being Who.ed to restore an inoperable PORV to operable status.
)
Survai!1= ara Requirement 4.4.3.2.1.b.2 provides for the testing of the mechanical and 4
electrical aspects of control systems for the PORVs.
}
Testing of PORVs in HOT STANDBY or HOT SHUIDOWN is required in order to simulate the temperamre and pressure environmental effects on PORVs. Testing at CDLD j
SHUTDOWN is not considered to be a representative est for==*aaning PORV pL ance iede normalplant operating condidons.
I i
3/f,,M PRESSURIZER j
The limit on the maximum waar volume in the pressurizer assures that the parameter is j
maintamed within the normal saady stas envelope of operation assumed in the SAR. The limit i consistent with the initial SAR assumptions.1he 12-hour periodic surveillance is sufHeient to i
ensure that the parameneris restored to within its limit following expected nunsient operatio maximum water volume also ensures that a steam bubble is formed and thus hydraulically solid sysem. The requirement that a minimum number of pressurizer heaters be OPERABLE ensures that the plant will be able to establish natural cirmlah NORTH ANNA - UNIT 2 B3/44-2a Amendment No. 170,
__...__...___._______._____j
-.- - -... ~..
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i i
i BASES 3/4.4.7 CHEMISTRY 1
l The limitations on Reactor Coolant System chemistry' ensure that corrosion-j of the Reactor Coolant System is minimized and reduces the potential for j
Reactor Coolant System leakage or failure due to stress corrosion.
Maintaining i
j the chemistry within the Steady State Limits provides adequate corrosion j
protection to ensure the structural integrity of the Reactor Coolant System i
)
over the life of the plant.
The associated effects of exceeding the oxygen,
]
chloride and fluoride limits are time and temperature dependent.
Corrosion J
i studies show that operation may be continued with contaminant concentration I
levels in excess of the Steady State Limits, up to the Transient Limits, for j
the specified limited time intervals without having a significant effect on j
j the structural integrity of the Reactor Coolant System.
The time interval
)
permitting continued operation within the restrictions of the Transient Limits i
provides time for taking corrective actions to restore the contaminant con-l centrations to within the Steady State Limits.
{
The surveillance requirements provide adequate assurance that concentrations i
in excess of the limits will be detected in sufficient time to take corrective j
action.
3/4.4.8 SPECIFIC ACTIVITY j
The limitations on the specific activity of the primary coolant ensure j
that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator j
tube rupture accident in conjunction with an assumed steady state primary-to-i secondary steam generator leakage rate of 1.0 GPM.
The values for the limits i
on specific activity represent interim limits based upon a parametric evaluation i
by the NRC of typical site locations.
These values are conservative in that' j
specific site parameters of the North Anna site such as site boundary location and meteorological conditions, were not considered in this evaluation. The-i j
NRC is finalizing site specific criteria which will be used as the basis for 4
the reevaluation of the specific activity limits of this site.
This reevaluation j
may result in higher limits.
i 7
l NORTH ANNA - UNIT 2 8 3/4 4-5
BASES i
i
_ ne ACDON stamment permitting POWER OPERATION to continue forlimind time l
periods with the primary coolant's =p~ ine activity grenser than 1.0- Ci/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.41, accommodates possible iodme spiking pt===+1 which may ocicurfollowing changes in *DiERMAL POWER.l t
l Reducing T.,, to less than 500T prevents the release d activity should a steam generato tube rupture since the saturation pressure of the primary coolant is below the lift pressme of the atmospheric steam relief valves. De surveilla-requirements provide adequate assurance that l
excessive =parine activity levels in the primary caalmar will be detected in suff;cient time so take corrective action. Information obtained on indina spiking will be used to assess the parameters associatedwithspiking,hs Areductioninseqwyofisotopicanalysesfollowingpower changes may be permissible ifjustified by'the data c' tained.
b J
]
344,Q, PRRMSURE / TEMPERATURE LIMITS.
t g
Remetnr enninne Sw -..
[
All components in the Reactor Coolant Sysam are designed to withstand the effects of i
cyclic loads due no system temperature and pressure changes.Ylifse cyclic 16 ads are ins ncrmal load traaniants, reactor trips, and startup and shutdown operations. De various categor i
ofload cycles used for design purposes me provided in Section 5.2 of the UFSAR. During j
and shutdown, the rates of temperature and pressure changes me limiend so that the maximum j
sp~ iMed heatup and cooldown rates are cwinent with the design assumptions and satisfy j
stren limits forcyclic operation.
i j
During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses i
which vary from compressive at the inner wall to tensile at the outer wall. Dese thermal induced j
compressive suesses tend to allevisse the sensile snesses ladneed by the internal presswe.
Therefore, a pressme-temperstme curve based on steady stas candiriana (i.e., no thermal snes 1
l represents a lower bound of all similar curves for finia bestup rases when the inner wall of the vessel is treated a the governing location.
1 a
j De heatup analysis also covers the determinadon of pressure-temperature limitations for j
the case in which the outer wall of the vessel t=--== the conuolling location. The thermal j
gradients established during heatop produce tensile snesses at the ouer wall of the vessel. nese j
suesses are additive to the pressme induced tensile stresses which are aheady present. The thermal 1
induced snesses at the outer wall of the vessel are tensile and are 4p - =i on both the rate of heatup and the time along the heatup ramp; thi J r., a lower bound curve similar to that described j
for the beamp of the inner wall cannot be defined. caa=~-cly, for the cases in which the outer 1
wall of the vessel t+ --== the sness controlling incarina, each bestup rase ofinterest must be j
analyzed os an individual basis.
- j
- NORTH ANNA UNIT 2 B3/44-6 Amendment No. !? ! *^.
- 170, 1
l l
l l
1 l
BASES i
I i
ne heatup limit curve, Figure 3.4-2, is a composie curve which was prepared by j
determining the most conservative case, with either the inside or outside wall controlling, fo
}
heatup rate up to 60*F per hour, ne cooldown limit curves of Figure 3.4-3 are composite curves l
which were prepared based upon the same type analysis with the exception ther.the controlling j
location is always the inside wall where the cooldown thermal gradients tend to produce tensile l
stresses while producing compressive stresses at the outside wall. ne heatup and cooldown curves j
were geyed based upon the most limiting value of the predicted adjusted reference temperature i
at the end of 17 EFPY. De most recent capsule analysis results are documented in Westinghouse j
Reports WCAP-12497, January 1990. De heatup and cooldown curves are documented in l
Westinghouse Report WCAP-12503, March,1990.
ne reactor vessel materials have been tested to determine their initial RTygyr. Reactor j
operation and resultant fast neutron (E > 1 Mev) irradiarian will cause an increase in the RT yr-Nt j
An adjused reference emperature, based upon the fluence and copper content of the materialin
['
question, can be g4LT. 4 using US NRC Regulatory Guide 1.98, Revision 2. De heatup and i
{
cooldown limit curves (Figure 3.4-2 and Figure 3.4-3) include predicted adjustmens for this shift in RTyryr t the end of 17 EFPY. De reactor vessel beltline region matrial properties are lised a
l an Figures 3.4-2 and 3.4 3.
1
\\
t j
ne actual shift in the RTyryr of the vessel material will be established periodically by
{
removal and evaluation of the reactor vessel material specimens installed on the inside wall of the i
j thermal shield. The savain capsule withdrawal schedule was prepared in accordance with the 1
requirements of ASTM E-185 and is presented in the UFSAR. Regulatory Guide 1.99, Revision i
2, provides g*=aca for caleidadaa of the shift in RTyryr using measmed data. Dosunetry from j
the surveillance capsule is used to dearmine the neutron fluence to which the material specimens j
were exposed, and to support calcularianal estimates of the neunon fluence to the reactor vesset i
4~
ne pressure-temperanne limit lines shown on Figure 3.4 2 for inservice leak and
]
}
hydrostatic testing have been provided to assure compliance with the minimum emperature 4
requirements of AyA G to 10 CFR 50. De minimum temperanse for criticality =pa ihA in f
T.S. 3.1.1.5 assures compliance with the criticality limits of 10 CFR 50 Ap= ah G.
4,
]
He number of reactor vessel irradiation surveillance specimens and the frequencies for j
removing and testing these specimens are provided in the UFSAR to assure compliance with the'
[
]
requirements of Appendix H to 10 CFR Part 50.
j pn mri-l ne limitations imposed on pressurizer heatup and cooldown and spray waar emperature 1
l differential are provided to assure that the pressurimer is operated within the design criteria assumed j
for the fatigue analysis performed in accordance with the ASME Code requirementa.
l
)
NORTH ANNA - UNIT 2 B3/44-7 Amendment No. 449, J
- 170, l
i i
s I
REACIOR COOLANT SYSTEM i
I l
BASES i
tow.Tc-.y.=..m e.
a w P M n.
g De OPERABILITY of two PORVs or an RCS vent opening of greamr than 2.07 square inches ensures that the RCS will be promend fhun pressure transients which could exceed h limits of Ag-r+= t-G to 10 CFR Part 50 when one or more of the RCS cold legs are 1 ss than or equal to 2707. Either PORV has adequam relieving capability to promet the RCS fann l
owy.wsurization when the transient is limited to either (1) the start of an idle RCP with the Way water temperstare of the steam generatorless than or equal 1o 50T above the RCS cold leg temperannes or (2) the start of a charging pump and its injection into a waar-solid RCS.
l
)
Auw=narie orpassive low temperature overpressure protection (L*1DP)is %M whenever any RCS cold leg temperature is less than 2707. His temperanno is the waser emperanne swag to a metal temperanne of at least the limiting RTNDr + 50T +
instrument uncertainty. Above 2707 administrative consolis adequase protection to ensure the limits of the heatup curve (Figure 3.4-2) and the cooldown curve (Figme 3.4-3) are not violased.
The concept of requiring automatic LTOP at the lower end, and adenini,rrative consol at the upper end, of the Appendix G curves is further discussed in NRC Generic Leaer 38-11.
.j 3/4.4.10 STRUCIURALINTEGRITY 3/4.4.10.1 ASME CDDE er ASS 1. 2 and 3 COMPONENTS The inspection programs for ASME Code Class 1,2 and 3 Reactor aalant System r
components ensme 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 complianem with Section XI of the ASME Boiler and Pressure Vessel Code.
NORM ANNA - UNIT 2 B3/443 Amendment No. 449,
- 170,
a 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 will be immediately 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 j
l analysis are met.
l 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 4
l isolation valves fail to meet single failure criteria, removal of power l
to the valves is required.
]
The limits for operation with an accumulator inoperable for any reason i
except an isolation valve closed minimizes the time exposure of the l
plant to a LOCA event occurring concurrent with failure of an additional l
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-I 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 2 B 3/4 5-1 I
I EMERGRNCY CORE OfYH_rNG SYS'fEMS (ECCS) l BASES I
ECCS SUBSYSTEMS (Candn=0 With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable I
without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.
De limitation for a maximum of one centrifugal charging pump and one low head safety -
injection pump to be OPERABLE and the SurvAlt== Requirement to verify all charging pumps i
and low head safety injection pumps except the required OPERABM pump to be inoperable below 270*F provides assurance that a mass addition pressure transient can be relieved by the l
operation of a single PORV.
The Surveillance Requirements provided to ensure OPERABILITY of each component ~
I ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITYis maintained.
In the event of modifications to an ECCS subsystem that could alter the subsystem flow characteristics, a flow balance test shall be performed. The flow balance test criteria are established based on the system performance assumed in the safety analysis (minimum flow limit) and on HHSI pump runout protection (maximum flow limit). In performing the flow balance, the
{
effects of flow measurement instrument uncertainties accounting for system configuration and the variability between installed pumps must be properly considered.
Numerical acceptance criteria for the flow balance test are specified in the surveillance test procedure. These criteria are established based on the following considerations:
- 1) ne totalinjected flow to the cose (assuming spillage of the branch line with the highest flow) must meet or exceed that assumed in the safety analysis. De limiting safety analysis is the loss of coolant accident (LOCA) analysis. This criterion may vary, particularly since the inputs to the safety analysis controlled by LCD 6.9.1.7 may vary with reload cycle. The safety analysis flow requirements are thus established by the cunently applicable LOCA analysis which has demonstrated compliance with the ECCS acceptance limits of 10 CFR 50.46.
- 2) De total pumped flow must be less than the HHSI pump runout limit. This flow varies with the specific HHSI pump assumed to operate during the accident. Since the HHSI pumps also function as normal charging pumps, their characteristics, including runout limits, will vary over servicelife.
- 3) The requirements for reactor coolant pump seal injection must be met during normal operation, and the effects of seal injection during accidents must be considered in meeting constraints 1) and 2) above.
NORTH ANNA -UNIT 2 B3/45-2 Amendment No. Si, M?,"
- 170,
i j
_ ADMINISTRATIVE CONTROLS i
j SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator, Region II, within the time l
period specified for each report. These reports shall be submined pursuant to the requirement of the applicable Wh ion:
t i
- a. Inservice fa=~ 'iaa Reviews, Sg:-::9c=:en 4.0.5, shall be sp=W within 90 days j
ofcompledon.
I
- b. MODERATOR TEMPERATURE COEFFICIENT. Specification 3.1.1.4.
l
- c. Deleted.
}
- d. RADIA110N MONITORING INS 1RUMENTATION. Specification 3.3.3.1, j
Table 3.3-6, Action 35.'
i
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- e. Deleted.
b f.
LOW-TEMPERATURE OVERPRESSURE PRCyIECTION. S - -::f-:=m 3.4.9.3.
i
/
l
- g. EMERGENCY CORE COOLING SYS'IEMS. S +::P- ='+:- 3.5.2 and 3.5.3.
i
- h. SETTLEMENT OF CLASS 1 STRUCTURES. Sa~ ih ien 3.7.12.
t i.
i GROUND WATER LEVEL - SERVICE WA1ER RESERVOIR. Specificados 3.7.13.
i y, DelegagLe.
o.
a>.
.a
- k. Deleted.
^'
i' L RADIOACI1VE EFFLUENTS. As required by the ODCM.
4 j
- m. RADIO!.DGICAL ENVIRONMENTAL MONTIORINO. As requend by the j
ODCM.
- i j
- n. SEALED SOURCE CONTAMINATION. SM"- =i-= 4.7.11.1.3.
- o. REACTOR COOLANTSYSTEM STRUCTURALINTEGRITY.S =:"c=: -
i j
4.4.10. For any abnormal degradation of the struennal inegrity of the reactor veswl or the Reactor raalmar System pressure boundary deteced during the i,.J-.ance of Sa~ ih ion 4.4.10, an initial report shall be submined within 10 days after t
{
detection and a detailed report submined within 90 days after the completion of l
S +Sf- = = 4.4.10.
i
- p. CONTAINMENT STRUCIURAL INTEGRITY. Specification 4.6.1.6. For any j
abnormal degradation of the containment structure detected during the.J.mance r
)
of Sp+::~stion 4.6.1.6, an initial report shall be submined within 10 days after completion of S dh 4.6.1.6. A final report, which includes (1) a description i
j of the condition of the liner plae and concrete, (2) inspection procedure, (3) the j
tolerance on cracking, and (4) the corrective actions taken, shall be submitted j
within 90 days after the completion of Sp+-:fstion 4.6.1.6.
J J
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ii NORTH ANNA - UNIT 2 6 21
^ ~ ~ ' - ^ " ^ " ' " ' " "
i I
- 170,
!!',123
-