ML20196D650
| ML20196D650 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/05/1988 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20196D654 | List: |
| References | |
| NPF-68-A-014 NUDOCS 8812090154 | |
| Download: ML20196D650 (8) | |
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wash WGTON, D. C. 20665
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i GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATI0ft MUNICIPAL ELECTIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA V0GTLE ELECTRIC GENERATING PLANT, UNIT 1 A11ENDMENT TO FACILITY OPERATING LICENSE Amendment No. 14 License No. NPF-68 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the Vogtle Electric Generating Plant, Unit 1 (the f acility) Facility Operating License No. NPF-68 filed by the Georgia Power Company acting for itself, Oglethorpe Power Corpo-ration, Municipal Electric Authority of Georgia, and City of D<on, Georgia, (the licensees) dated August 12, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set 1
4 th in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as ameilad, the provisions of the Act, and the rules and regulations of the Comission; l
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 regulations set forth in 10 CFR Chapter I; D.
The issuance of this license amendment will not be inimical to the comon defense and security or to the health and safety of the public; i
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 i
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2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-68 is hereby dmended to read as follows!
(2) Technical Specif' ions and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 14, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license.
GPC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
David B. Matthews, Director Project Directorate II-3 r
Division of Reactor Projects - I/II i
Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
December 5, 1988 t
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Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-68 is hereby amended to read as follows:
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.
14, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. GPC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Da B.
tatthews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
December 5, 1988 I
1
A ATTACHMENT TO LICENSE AMENDMENT NO. 14 FACILITY OPERATING LICENSE NO. NPF-68 DOCKET NO. 50-424 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 areas of change. The corresponding overleaf pages are also provided to maintain document completeness.
Amended Page Overleaf Page 3/4 5-4 3/4 5-3 8 3/4 5-2 B 3/4 5-1 h
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EMERGENCY CORE COOLING SYSTEMS s
3/4.5.2 ECCS SUBSYSTEMS - T,y GREA1ER THAN OR EQUAL TO 350*F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:
a.
One OPERABLE centrifugal charging pump, b.
One OPERABLE Safety Injection pump, c.
One OPERABLE RHR heat exchanger, d.
An OPERABLE flow path capable of taking suction from the refueling water storage tank on a Safety Injection signal and semi-automatically transferring suction to the containment emergency sump during the recirculation phase of operation.
APPLICABILITY:
H00ES 1, 2, and 3.*
ACTION:
a.
Witt one ECCS subsystem inoserable, restere the inoperable subsystem to CPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.8.2 witnin 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date.
The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
- The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into Mode 3 for the Safety Injection Pumps declared inoperable pursuant to Specification 3.5.3.2 provided the Safety Injection Pumps are restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever occurs first.
V0GTLE - UNIT 1 3/4 5-3
EMERGENCY CORE COOLING SYSTEMS SVRVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
a.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the following valves i
are in the indicated positions with power lockout switches in the lockout position:
t Valve Number Valve Function Valve Position HV-8835 SI Pump Cold Leg. Inj.
OPEN h/-8840 RHR Pump Hot Leg. Inj.
CLOSED HV-8813 SI Pump Mini. Flow Isol.
OPEN HV-8806 SI Pump Suction from RWST OPEN HV-8802A, B SI Pump Hot Leg Inj.
CLOSED HV-8809A, B RHR Pump Cold Leg Inj.
OPEN*
l b.
At least once per 31 days by:
1)
Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high points, and 2)
Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
c.
By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the Containment Emergency Sump and cause restriction of the pump suctions during LOCA conditions.
This visual inspection shall be performed:
1)
For all accessible areas of the containment prior to establish-ing CONTAINHENT INTEGRITY, and 2)
Of the areas affected within containment at the completion of I
each containment entry when CONTAINMENT INTEGRITY is established, d.
At least once per 18 months by:
1)
Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System by e'suring that:
a)
With a simulated or actual Reactor f ~ ant System pressure signal greater than or equal to 377 p,') the interlocks prevent the valves from being opened, and b)
With a simulated or actual Reactor Coolant System pressure j
signal less than or equal to 750 psig the interlocks will L
cause the valves to automatically close.
2)
A visual inspection of the Containment Emergency Sump and verify-l ing that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.
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- Either valve may be realigned in MODE 3 for testing pursuant to 1
Specification 4.4.6.2.2 V0GTLE - UNIT 1 3/4 5-4 Amendment No. 14 i
3/4.5 EMERGENCY CORE COOLING !. STEMS BASES 3/4.5.1 ACCUf10LATORS The OPERABILITY of each Reactor Coolant System (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 analysis are met.
The accumulator power operated isolation valves are considered to be "operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met.
In addition, as these accumulator isolation valves f ail to meet single failure criteria, removal of power to the valves is required.
The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures.
If a closed isolation valve cannot be 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 sufficient 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 postulated 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.
With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable witnout single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.
I V0GTLE - UNIT 1 8 3/4 5-1
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EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued)
The limitation for all safety injection pumps to be inoperable below 350'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.
The Surveillance Requirements provided to ensure OPERABILITY of ea' h component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.
Survoi11ance Requirements for throttle valve position stops and flow balance tessing provide assurance that proper ECCS flows will be maintained in the event of a LOCA.
Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to:
(1) prevent total pump flow from exceeding runout conditions when the system is in its mintnum resistance configuration, (2) provide the proper flow split between inje: tion points in accordance with the assumptions used in the ECCS-LOC /i analyses, (3) provide an acceptable level <f total ECCS flow to all injection ooints equal to or above that assumed it the ECCS-LOCA analyses and (4) to ensure that centrifugal charging pump injection flow which is directed through the seal injection path is less than or equal to the amount assumed in the safety analysis.
The surveillance requirer ents for leakage testing of ECCS check valves ensure that a failure of one valle will not cause an intersystem LOCA.
In MODE 3, with either HV-8809 A or 3 closed for ECCS check valve leak testing, adequate ECCS flow for core cooling in the event of a LOCA is assured.
3/4.5.4 REFUELING WATER STORAGE TANK The OPERABILITY of the Refueling Water Storage Tank (RWST) as part of the ECCS ensures that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS cooldown.
RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant l
accident, or a steam line rupture.
The limits on RWST minimum volume and boron concentration ensure that
- 1) sufficient water is available within containment to permit recirculation cooling flow to the core, 2) the r:4ctor will remain subcritical in the cold condition following a s'nall LOCA or steamline break, assuming complete mixing of the RWST, RCS, and ECCS tater volumes with all control rods inserted except the most reactive control assembly (ARI-1), and 3) the reactor will remain subcritical in the cold condition following a large break LOCA (break flow
> 3.0 FT2) assuming complete mixing of the RWST, RCS, ECCS water and other sources of water that may eventually reside in the sump, post-LOCA with all control rods assumed to be out.
The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.0 and 10.5 for the solution recirculated 4
within containment after a LOCA.
This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.
V0GTLE - UNIT 1 B 3/4 5-2 Amendment No. 14 I