ML19289F395
| ML19289F395 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 04/27/1979 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Virginia Electric & Power Co (VEPCO) |
| Shared Package | |
| ML19289F396 | List: |
| References | |
| NPF-04-A-010 NUDOCS 7906070319 | |
| Download: ML19289F395 (15) | |
Text
8 UNITED STATES
[#
.k NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 Q
T of ]
VIRGINIA ELECTRIC AhD POWER C0FPANY DOCKET NO. 50-338 NORTH ANNA POWER STATION, UNIT N0.1 FACILITY OPERATING LICENSE License No. NPF-4 Amendment No.10 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The issuance of this license amend'..ent to Virginia Electric and Power Company complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the license, as amended, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulationst 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 Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Appendix A Technical Specifications as indicated in the attachment to this license amendment. Facility Operating License No. NPF-4 is hereby amended to read as follows:
7 9 0 6 0 7 0 Slot 2231 085 (2 ) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.10, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date o'f issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Domenic B. Vassallo, Assistant Director for Light Water Reactors Division of Project Management Date of Issuance: APR 2 71979
Enclosure:
Revised pages to Appendix A Technical Specifications 2231 086
ATTACHMENT TO LICENSE AMENDMENT N0.10 FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. 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.
Paaes IV 3/4 3-56 (added) 3/4 3-57 (added) 3/4 4-31 g731 qg7 3/4 4-31a (added)
I Uu/
B 3/4 3-3
IfiDEX LIMITING CONDITIONS FOR OPERATION AllD SURVEILLANCE REQUIREMENTS SECTION Pace 3/4.0 APPLICABILITY.............................................
3/40-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T
> 200*F..........................
3/41-1 avg Shutdown Margin - T 1 200 F..........................
3/4 1-3 avg Boron Dilution-Reactor Cool ant Flow..................... 3/4 1-4 Boron Dilution-Valve Position...........................
3/4 1-5 Moderator Temperatu re Coeffi cient....................... 3/41-6 Minimum Temperatu re for Cri ti cali ty..................... 3/4 1-7 3/4.1.2 BORATION SYSTEMS Fl ow Pa th s - S h u tdown................................... 3/4 1-8 Flow Paths - Operating..................................
3/4 1-9 Charging Pump - Shutdown................................
3/4 1-11 Charging Pumps - Operating..............................
3/4 1-12 Boric Acid Transfer Pumps - Shutdown....................
3/4 1-13 Boric Acid Trans fer Pumps - Operating................... 3/4 1-14 Borated Water Sources - Shutdown........................ 3/4 1-15 Borated Water Sources - 0perating....................... 3/4 1-16 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height............................................
3/4 1-18 Pesi tion Indi cator Cnannel s-Operati ng................... 3/4 1-21 Posi tion Indi cator Channel s-Shutdown.................... 3/4 1-22 Rod Drop Time...........................................
3/4 1-23 S h u tdown Ro d In s e rti o n Li mi t............................ 3/4 1-24 Control Rod Inserti on Limi ts............................ 3/4 1-25 Pa rt length Rod Insertion Limi ts........................ 3/4 1-23 2231 088 NORTH ANNA - UllIT_l-III -
O
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION Pace 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 Ax i a l Fl u x D i f f ere nc e..................................
3/4 2-1 3/4.2.2 Heat Fl ux Hot Channel Factor...........................
3/4 2-5 3/4.2.3 Nuclear Enthalpy Hot Channel Factor....................
3/4 2-9 3/4.2.4 Quadrant Power Tilt Ratio..............................
3/4 2-12 3/4.2.5 DNB Parameters.........................................
3/4 2-14 3/4.2.6 Axial Power Distributi on...............................
3/4 2-16 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION....................
3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION......................................
3/4 3-15 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring...................................
3/4 3-35 Movable Incore Detectors...............................
3/4 3-39 Seismic Instrumentation................................
3/4 3-40 Meteorological Instrumentation.........................
3/4 3-43 Auxiliary Shutdown Panel Monitoring Instrumentation....
3/4 3-46 Post-Accident Instrumentation..........................
3/4 3-49 Fire Detection Instrumentation.........................
3/4 3-52 Axial' Power Distribution Monitori ng System.............
3/4 3-54 Loose Parts Monitoring System..........................
3/4 3-56 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS Normal 0peration.......................................
3/4 4-1 Isolated Loop..........................................
3/4 4-4 Isolated Loop Startup..............
3/4 4-5 NORTH ANNA - UNIT 1 IV '
Amendment No. 3,
I INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued) h-ii R
3 a.
If the absolute value of is greater than 203, another R
J map shall be completed to verify the new R.
If the second map j
shows the first to be in error, the first map s_ hall be dis-regarded.
If the second map confims the new R, four more maps j
(including rodded configurations _ allowed by the insertion limits) will be completed so that a new R 3 and oj can be defined from the six new maps.
4.3.3.8.2 The APDMS shall be demonstrated OPERASLE:
a.
By performance of a CHANNEL FUNCTIONAL TEST within 7 days prior to its use and at least once per 31 days thereafter when used for monitoring F (Z).
j b.
At least once per 18 months, by performance of a CHANNEL CALIBRATION.
2231 090 NORTH ANNA - UNIT 1 3/4 3-55 Amendment No. 3, 5
INSTRUMENTATION LOOSE PARTS MONITORING SYSTEMS LIMITING CONDITIONS FOR OPERATION 3.3.3.9 The loose parts monitoring system instrumentation identified in Table 3.3-12 shall ba OPERABLE.
APPLICABILITY: MODES 1, 2 and 3.
ACTION If all channels of one or more collection regions are inoperable, restore the instrument (s) to OPERABLE status within 30 days or, in lieu of any report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channels to OPERABLE status.
SURVEILLANCE REOUIREMENTS 4.3.3.9 Each channel of the loose parts monitoring system identified in Table 3.3-12 shall be demonstrated OPERABLE by the performance of:
a.
A CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
A CHANNEL FUNCTIONAL TEST at least once per 31 days.
c.
A CHANNEL CALIBRATION at least once per 18 months.
2231 091 NORTH ANNA
, UNIT 1 3/4 3-56 Amendment No.10
TABLE 3.3-12, LOOSE PARTS MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE 1.
Steam Generator Transducers 1/ steam generator 2.
Reactor Vessel Flange Transducers 1/2 3.
Reactor Vessel Lower Planum Transducers 1/2 2231 092 NORTH ANNA - UNIT 1 3/4 3-57 Amendment No. 10
, u REACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 & 3 COMPONENTS LIMITING CONDITION FOR OPERATION 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.
APPLICABILITY: ALL MODES.
ACTION:
a.
With the structural integrity of any ASME Code Class 1 com-ponent(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 more than 50'F above the minimum temperature required by MDT considerations.
b.
With the structural integrity of any ASME Code Class 2 com-ponent(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 com-ponent(s) not conforming to the above requirements, restore the structural integrity of the affected ccmponent(s) to within its limit or isolate the affected component (s) from service.
d.
With any RCP shaft deflection indication greater than 20 mils, the reactor shall be placed in at least HOT STANr8Y within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the affected RCP(s) tripped and then affected flow straightener plate (s) ultrasonically examined.
e.
The provisions of Specification 3.0.4 are not applicable.
2231 093 NORTH ANNA - UNIT 1 3/4-4-31 Amendment No. 3,
REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS 4.4.10.1 In addition to the requirements of Specification 4.0.5,1) che 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 and 2) the flow straighteners in each steam generator-to-RCP elbow shall be ultrasonically examined whenever a RCP shaft deflection of greater than 20 mils is indicated and at least once per 18 months.
2231 094 NORTH ANNA - UNIT 1 3/4'4-31a Amendment No. 10
REACTOR C0OLANT SYSTEM STRUCTURAL IN'EGRITY STEAM GENERATOR SUPPORTS LIMITING CONDITION FOR OPERATION 3.4.10.2 The temperature of the steam generator supports shall be l
maintained:
a.
> 225 F for A572 material monitored at a middle level corner during operation and at a top level corner during heatup of the supports.
b.
< 355'F at the monitored top le;91 corner.
c.
> 85'F for A36 material monitorea at a bottom level corner during heatup.
APPLICABILITY: With pressurizer pressure > 1000 psig.
ACTION:
With the temperature of any steam generator support outside the above limits, restore the temperature to within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be below 1000 psig within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.10.2.1 The steam generator support temperatures for A572 material shall be verified to be within the specified limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.10.2.2 The steam generator suppcrt temperatures for A36 material shall be verified to be within the specified limit prior to exceeding a pressurizer pressure of > 1000 psig.
4.4.19.2.3 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.
2231 095 NORTH ANNA - UNIT 1 3/4 4-32 Amendment No. ;f, 3
INSTRUMENTATION BASES 3/4.3.3.6 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.
3/4.3.3.7 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the poten-tial for damage to safety related equipment and is an integral element in the overall facility fire protection program.
In the event that a portion of the fire detection instrumentation is ir. operable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.
3/4.3.3.8 AXIAL F0WER DISTRIBUTION MONITORING SYSTEM (APDMS)
OPERABILITY of the APDMS ensures that sufficient capability is available for the measurement of the neutron flux spatial distribution within the reactor core. This capability is required to (1) monitor the core flux patterns that are representative of the peak core power density, and (2) limit the core average axial power profile such that the total power peaking factor F is maintained within acceptable q
limits.
3/4.3.3.9 LOOSE PARTS MONITORING SYSTEM OPERABILITY of the Loose Parts Monitoring System provides assurance that loose parts within the RCS will be detected.
This capability is designed tn 'asure that loose parts will not collect and create undesirable flow bhxages.
2231 096 NORTH ANNA - UNIT 1 B 3/4 3-3
. Amendment No. 3,