ML20082V344
| ML20082V344 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 12/06/1983 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Georgia Power Co, Oglethorpe Power Corp, Municipal Electric Authority of Georgia, City of Dalton, GA |
| Shared Package | |
| ML20082V348 | List: |
| References | |
| TAC 52448, DPR-57-A-096 NUDOCS 8312200086 | |
| Download: ML20082V344 (12) | |
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UNITED STATES i
NUCLEAR REGULATORY COMMISSION
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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AhfNDMENT TO FACILITY OPERATING LICENSE Amendment No. 96 License No. DPR-57 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Georgia Power Company, et al.,
(the licensee) dated September 29, 1983, as supplemented October 24, 1983, and November 15, 1982, 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 forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this license amendment will not be inimical to the comon defense and recurity 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 Technical.Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DP'R-57 is hereby amended to read as follows:
8312200006 831206 PDR ADOCK 05000321 p
. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 96, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Sp2cifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Ig
/7 J.lfdAC John
. Stolz, Chief J
Op ating Reactors Branch #4 vision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: December 6,1983 9
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ATTACHMENT TO LICENSE AMENDMENT NO. 96 FACILITY OPERATING LICENSE NO. DPR-57 i
DOCKET NO. 50-321 Replace the following 'pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and j
contain a vertical line indicating the area of change.
a Remove Insert 3.11-1 3.11-1 3.11-3 3.11-3 Fig. 3.11-1 (Sheet 1)
Fig. 3.11-1 (Sheet 1)
Fig. 3.11-1 (Sheet 2)
Fig. 3.11-1 (Sheet 3)
Fig. 3.11-1 (Sheet 2)
Fig. 3.11-1 (Sheet 3)
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Fig.3.11-1(Sheet 4) 4 Fig. 3.11-1 (Sheet 5)
Fig. 3.11-4 Fig. 3.11-4 Fig. 3.11-5 Fig. 3.11-5 i
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'LFITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMEf:T5 3.11 FUEL RODS 4.11 FUEL RODS Acolicability Applicability The Limiting Conditions for Operation The Surveillance Recuirements apply to the i
associated with the fuel rods apply to parameters which monitor the fuel rod those parameters which monitor the fuel operating conditions.
rod operating conditions.
Objective Cbjective lhe Objective of the Limiting Conditions The Objective of the Surveillance for Operation is to assure the performance Requirements is to soecify. the type and of the fuel rods.
freamney of surveillance to be applied to the fuel rods.
Soecifications Soecifications A.Averace Planar Linear Heat A.Averace Planar Linear Heat.
Generation. Rate (APLHGR)
_G_eneration Rate (APLHGR)
During-power operation, the APLHGR for The APLFER for each type of fuel as a each type of fuel as a function of average function of average planar exposure shall planar exposure shall not exceed tne be determined daily during reactor limiting value shown in Figure 3.11-1, operation at 2 25% rated thermal power.
sheets 1 thru 5.
If at any time during operation it is determined by normal surveillance that the limiting value for APLFCR is being exceeded, action shall be initiated within 15 minutes to restore cceration to within the prescribed 11 nits. If the APLHGR is not returned to within the prescribed limits within two (2) hours, then reduce reactor power to less than 25% of rated thermal power within the next four (4) hours. If the limiting condition for operation is restored prior to expiration of the specified time interval, then further progression to less than 25% of rated thermal power is not recuired.
B. Linear Heat Generation Rate B. Linear Heat Generation Rate (LEGR) t LHGR)
During power operation, the LFER as a The LHGR as function of core height shall function of core height shall not exceed be checked daily during reactor operation the limiting value shown in Figure 3.11-2 at 2 25% rated thermal power.
for 7 x 7 fuel os the limiting value of 13.4 kw/ft for 8 x 8/8 x 8R fuel. If at any. time during operation it is determined by normal surveillance that the limiting value for LHGR is being eW, action shall be initiated within 15 minutes to y
restore operation to within the prescribed limits. If the Amendnent No. jii, pf,pf, h 96 3.11-1 n
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BASES FOR LIMITItG C0eOITIONS FOR CPERATION AND SURVEILLANCE REGUIREMENTS l
3.11 FUEL ROOS A
Aversoe Planar Linear Heat Generation Rate (APLFCR)
This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 50, Appendix K, even considering the postulated effects of fuel pellet densification.
The peak cladding temperature following a postulated loss-of-coolant acci-dent is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependelt second-arily on the rod to rod power distribution within an assembly. Eince expected local variations in power distribution within a fuel assembly' affect the calculated peak clad temperature by less than + 200 F relative to the peak temperature for a typical fuel design, the liliilt on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10 C.9 50, Appendix K limit. The limiting value for APUCR is shown in Figures 3.11-1., sheets 1 tnru 5.
The calculational procedure used to establish the APLHGR shown in' Figures 3.11-1, sheets 1 thru 5 is based on a loss-of-coolant accident analysis.
The analysis was performed using General Electric (GE) calculational mocels which are consistent with the recuirements of Appendix X to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1.
Differences in this analysis as compared tn previous analyses performed with Reference 1 are: (1) The analyses assumes a fuel assenoly-planar power consistent with 102% of the MAPLHGR shown in Figure 3.11.1; (2) Fission product decay is computed assuming an energy release rate of 200 MEV/ Fission; (3) Pool boiling is assumed after nucleate boiling is lost during the flow stagnation period; (4) The effects of core spray entrainment and counter-current flow limiting as described in Reference 2, are incluced in the reflooding calculations.
A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Table 1 of NEDO-21187 (3).
3.11-3 faendment No. %,
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Amnchent No. 96 FIGURE 3.11-1 (SHEET 3) 1
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