ML20034F429
| ML20034F429 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 02/19/1993 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | Georgia Power Co, Oglethorpe Power Corp, Municipal Electric Authority of Georgia, City of Dalton, GA |
| Shared Package | |
| ML20034F430 | List: |
| References | |
| GL-90-006, NPF-68-A-056, NPF-81-A-035 NUDOCS 9303030136 | |
| Download: ML20034F429 (9) | |
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UNITED sT ATEs i '-
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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA i
CITY OF DALTON. GEORGIA f
i V0GTLE ELECTRIC GENERATING PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No. 56 i
License No. NPF-68 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Vogtle Electric Generating i
Plant, Unit 1 (the facility) Facility Operating License No NPF-68 filed by the Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated November 18, 1991, as supplemented March 2, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
i and the Commission's rules and regulations as set forth in 10 CFR j
Chapter I; B.
The facility will operate it, conformity with the application, the provisions of the Act, and the rules and regulations of the
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Commission; C.
There is reasonable assurance (i) that the activities authorized by.
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this amendment can be conducted without endangering the health and l
safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; j
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 j
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.
1 9303030136 930219 -
DR ADOCK 05000424 PDR
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Accordingly, the license is hereby amended by page changes to the l
Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-68 i
l is hereby amended to read as follows:
1 Technical Specifications and Environmental Protection Plan l
l The Technical Specifications contained in Appendix A, as revised through l
Amendment No. 56, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated l
into this license.
GPC shall operate the facility in accordance with the i
s Technical Specifications and the Environmental Protection Plan.
l 3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION j
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r ht/L David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-I/II l
Office of Nuclear Reactor Regulation l
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Attachment:
i Technical Specification Changes l
i Date of Issuance:
February 19, 1993
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E GEORGIA POWER COMPANY j
i OGLETHORPE POWER CORPORATION
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MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA j
i CITY OF DALTON. GEORGIA i
V0GTLE ELECTRIC GENERATING PLANT. UNIT 2 1
-l AMENDMENT TO FACILITY OPERATING LICENSE
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Amendment No. 35 License No. NPF-81 1.
The Nuclear Regulatory Commission (the Commission) has found'that:
i A.
The application for amendment to the Vogtle Electric Generating Plant, Unit 2 (the facility) Facility Operating License No. NPF-81 i
filed by the Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City l
of Dalton, Georgia (the licensees), dated November 18, 1991, as i
supplemented March 2, 1992, complies with the standards and l
requirements of the Atomic Energy Act of 1954, as amended.(the Act),
and the Commission's rules and regulations as set forth in 10'CFR i
Chapter I; j
4 B.
The facility will operate in conformity with the application, the 1
provisions of the Act, and the rules and regulations of the Commission-1 C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and l
safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in '10 CFR Chapter I, 1
d 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 I
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.
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l l 2.
Accordingly, the license is hereby amended by page changes.to the i
Technical Specifications as indicated in the attachment to this license i
amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-81 is hereby amended to read as follows:
i Technical Soecifications and Environmental Protection Plan
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The Technical Specifications contained in Appendix A, as revised through l
Amendment No.
35, 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 i
'echnical Specifications and the Environmental Protection Plan.
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3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/
David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-I/II
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Office of Nuc} ear Reactor Regulation 1
i Attachment Technical Spacification j
j Changes l
Date of. Issuance:
February 19, 1993 l
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l ATTACHMENT -T0 LICENSE AMENDMENT NO.56 I
FACIllTY OPERATING LICENSE NO. NPF-68 DOCKET NO. 50-424 AND
.l TO LICENSE AMENDMENT NO. 35
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FACILITY OPERATING tlCENSE NO. NPF-81 l
l DOCKET NO. 50-425 Replace the following pages of the Appendix "A" Technical Specifications with i
the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
I Remove Paoes Insert Paoes l
3/4 4-10 3/4 4-10 l
B 3/4 4-3 B 3/4 4-3 I
1 3/4 4-34 3/4 4-34
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B 3/4 4-16 B 3/4 4-16 i
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i PEACTOR COOLANT SYSTEM i
3/4.4.4 RELIEF VALVES
'I LIMITING CONDITION FOR OPERATION l
3.4.4 Both power-operated relief valves (PORVs) and their associated block l
l valves shall be OPERABLE.
i APPLICABILITY: MODES 1, 2, and 3.
l ACTION:
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a.
With one or both PORV(s) inoperable, because of excessive seat l
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leakage, within I hour either restore the PORV(s) to OPERABLE status j
or close the associated block valve (s) with power maintained to the l
1' 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 SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
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b.
With one or both PORV(s) inoperable due to causes other than exces-
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sive seat leakage, within I hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve, and 1.
With only one PORV OPERABLE, restore at least a total cf two l
PORVs 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 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 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 />, or f
t 2.
With both PORVs inoperable, restore at least one PORV to l
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OPERABLE status 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 within the i
next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 />.
c.
With one or both block valve (s) inoperable, within I hour restore the block valve (s) to OPERABLE status or place its associated PORV(s) in i
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manual control.
Restore at least one block valve to OPERABLE status within the next hour if both block valves are inoperable; restore any
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j 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 />, i
d.
The provisions of Specification 3.0.4 are not applicable.
t SURVElltANCE RE0VIREMENTS j
4.4.4.1 Each PORV shall be demonstrated OPERABLE at least once per 18 months by:
a.
Operating the valve through one complete cycle of full travel, and b.
Performing a CHANNEL CALIBRATION.
j V0GTLE UNITS - 1 & 2 3/4 4-10 Amendment No.56 (Unit 1)
Amendment No. 35 (Unit 2)
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BASES 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves.
Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.
The PORVs are equipped l
with automatic actuation circuitry and manual control capability. No credit is i
taken for accident mitigation by automatic PORV operation in the analyses for MODE 1, 2, and 3 transients. The PORV(s) are considered OPERABLE in either the j
manual or automatic mode. The automatic mode is the preferred configuration, j
since pressure relieving capability is provided without reliance on operator j
l action.
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i 3/4.4.5 STEAM GENERATORS
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The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-i 1
tained.
The program for inservice inspection of steam generator tubes is l
based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice in-spection of steam generator tubing is essential in order to maintain surveil-lance of the conditions of the tubes in the event that there is evidence of l
mechanical damage or progressive degradation due to design, manufacturing i
errors, or inservice conditions tnat lead to corrosion.
Inservice inspection l
of steam generator tubing also provides a means of characterizing the nature l
l and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in j
negligible corrosion of the steam generator tubes.
If the secondary coolant i
i chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The exter,t of cracking during l
j plant operation would be limited by the limitation of steam generator tube i
leakage between the Reactor Coolant System and the Secondary Coolant System
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(primary-to-secondary leakage - 500 gallons per day per steam generator).
Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.
Operating plants have i
demonstrated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam genera-tor blowdown.
Leakage in excess of this limit will require plant shutdown and j
an unscheduled inspection, during which the leaking tubes will be located and l
pl ugged.
Wastage-type defects are unlikely with proper chemistry treatment of the
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secondary coolant.
However, even if a defect should develop in service, it i
1 will be found during scheduled inservice steam generator tube examinations.
Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness.
Steam generator tube inspections of operating plants have demonstrated the capability to j
reliably detect degradation that has penetrated 20% of the original tube wall thickness.
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V0GTLE UNITS - 1 & 2 B 3/4 4-3 Amendment No. 56 (Unit 1)
Amendment No. 35 (Unit 2) j
COLD OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION j
3.4.9.3 At least one of the following groups of Cold Overpressure Protection Devices shall be OPERABLE when the reactor coolant system (RCS) is not depres-surized through a vent path capable of relieving at least 670 gpm water flow at 470 psig.
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i a.
Two power-operated relief valves (PORVs) with lift settings which vary l
i with RCS temperature and which do not exceed the limits established in Figure 3.4-4a (Unit 1), Figure 3.4-4b (Unit 2), or l
t b.
Two residual heat removal (RHR) suction relief valves each with a l
setpoint of 450.psig 3%, or
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c.
One RHR SRV and or.2 PORV with setpoints as described above.
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APPLICABillTY: MODES 4, 5, and 6 with the reactor vessel head on.
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ACTION:
j a.
In MODE 4, with only one PORV or one RHR SRV OPERABLE, restore one l
a additional valve to OPERABLE status within the next 7 days or depres-l surize and vent the RCS, as specified in 3.4.9.3 above, within the-next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
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b.
In MODES 5 and 6, with only one PORV or one RHR SRV OPERABLE, restore one additional valve to OPERABLE status within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or i
depressurize and vent the RCS, as.specified in 3.4.9.3 above, within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
c.
In MODES 4, 5, or 6 with none of the PORVs or RHR SRVs OPERABLE, 3
depressurize and vent the RCS as specified in Specification 3.4.9.3 above, within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
d.
In the event that the PORVs and/or RHR SRVs or the RCS vent (3) are used to mitigate an RCS pressure transient, a Special Report shall be i
prepared and submitted to the Commission pursuant to Specifica-tion 6.8.2 within 30 days.
The report shall describe the circumstances initiating the transient, the effect of the PORVs, the RHR suction relief valves or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence, i
e.
The provisions of Specification 3.0.4 are not applicable.
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V0GTLE UNITS - 1 & 2 3/4 4-34 Amend.nent No. 56 (Unit 1)
Amendment No. 35 (Unit 2) t
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BASES PRESSURE / TEMPERATURE LIMITS (Continued)
Although the pressurizer operates in temperature ranges above those for which j
there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatique analysis performed in accordance 3
with the ASME Code requirements.
COLD OVERPRESSURE PROTECTION SYSTEMS I
The OPERABILITY of two PORVs, two RHR suction relief valves, a PORV and RHR SRV, l
or an RCS vent capable of relieving at least 670 gpm water flow at 470 psig ensures that the RCS will be protected from pressure transients which could exceed the limits I
of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 350*F.
The PORVs have adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either:
(1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to i
50*F above the RCS cold leg temperatures, or (2) the start of all three charging pumps ;
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and subsequent injection into a water-sclid RCS. The RHR SRVs have adequate relieving i capability to protect the RCS from overpressurization when the transient is limited to :
c either:
(1) the start of an idle RCP with the secondary to primary water temperature j
4 difference of the steam generator less than or equal to 25*F at an RCS temperature of j
350*F and varies linearly to 50*F at an RCS temperature of 200*F or less, or (2) the start of all three charging pumps and subsequent injection into a water-solid RCS. A combination of a PORV and an RHR SRV also provides overpressure protection for the RCS.
i The Maximum Allowed PORV Setpoint for the Cold Overpressure Protection System j
(COPS) is derived by analysis which models the performance of the COPS assuming i
various mass input and heat input transients. Operation with a PORV Setpoint less j
than or equal to the maximum Setpoint ensures that the nominal 13 EFPY for Unit I and 16 EFPY for Unit 2 Appendix G reactor vessel NDT limits criteria will not be violated j
with consideration for a maximum pressure overshoot beyond the PORV setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. To ensure that mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of l
all safety injection pumps while in MODES 4, 5, and 6 with the reactor vessel-head in-l stalled and disallow start of an RCP if secondary temperature is more than 50*F above j
primary temperature. Additional temperature limitations are placed on the starting of a Reactor Coolant Pump in Specification 3.4.1.3.
These limitations assure that'the RHR system remains within its ASME design limits when the RHR relief valves are used to prevent RCS overpressurization.
The Maximum Allowed PORV Setpoint for the COPS will be updated based on the j
l results of examinations if reactor vessel material irradiation surveillance' specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 16.3-3 cf the VEGP FSAR.
3/4.4.10 STRUCTURAL INTEGRiiv i
The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 g
components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.
These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where i
specific written relief.has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
V0GTLE UNITS - 1 & 2 B 3/4 4-16 Amendment No. 56 (Unit 1)
Amendment No. 35 (Unit 2)
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