ML20005E322
| ML20005E322 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 12/29/1989 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20005E323 | List: |
| References | |
| NUDOCS 9001050067 | |
| Download: ML20005E322 (7) | |
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UNITED STATES i
NUCLEAR REGULATCRY COMMISSION
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WAsHINoToN, o. C. 2066s 5
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r GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION 3
MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA i
l CITY OF DALTON, GEORGIA i
00CKET NO. 50-366 i
EDWIN !. HATCH NUCLEAR PLANT UNIT NO. 2 l
AMENDMENT TO FACILITY OPERATING LICENSE j
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Amendment No.105 License No. NPF-5 1.
The Nuclear Regulatory Commission (the Commission) has found that:
7 A.
The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 2 (the facility) Facility Operating License No. NPF-5 filed by Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton., Georgia (the licensee) dated September 18, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, 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 (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter 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 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, f.Obb0 6
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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. NPF-5 is hereby amended to read as follows:
i Technical Specifications 4
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.105, are hereby incorporated in the license.
The license-Tha11 operate the facility in accordance with the Technical Specif tions.
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This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
t FOR THE NUCLEAR REGULATORY COMMISSION M k#
David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: December 29, 1989 t
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V ATTACHMENT TO LICENSE AMENDMENT NO.105 i
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FACILITY OPERATING LICENSE NO. NPF+$
DOCKET NO. 50-366 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.
Remove PaQes Insert Pages 2-1 2-1 I
B 2-1 B 2-1
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B 2-4 B 2-4 B 3/4 2-3 B 3/4 2-3 i
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2.0 SAFETY LIMITS AND LIMITING $AFETY SYSTEM $ETTING$
2.1 SAFETY LIMITS TWERMAL POWER (Low Pressure or Low Flow) 2.1.1 THERMAL POWER shall not exceed 25*4 of RATED THERMAL POWER w the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flew.
APPLICABILITY: CONDITIONS I and 2.
ACTION:
With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10*4 of rated flow, be in at least HOT $HUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
THERMAL POWER (Hich Pressure and High Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.04 for two-loop rectreviation or 1.05 for single-loop recirculation l
operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10*4 of rated flow.
AoPLICABILITY:
CONDITIONS 1 AND 2.
ACTION:
With MCPR less than 1.04 for two-loop recirculation or 1.05 for single-loop l
l recirculation operation and the reactor vessel steam come pressure creater than 785 osig and core flow greater than 10*4 of rated flow, be
.n at least HOT $HUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam come, shall not exceed 1325 psig.
APPLICABILITY: CONDITIONS 1, 2, 3 and 4.
ACTION:
With the reactor coolant system pressure, as measured in the reactor vessel steam dome, aeove 1325 osig, be in at least HOT SHUTDOWN with reactor coolant system pressure s 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
HAICH-UNIT 2 2-1 Amendment No. 105
1 2.1 SAFETY LIMITS I
EA$ES 2.0 The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs.
i of these barriers during normal plant operations and anticipated tran sients.
The fuel cladding integrity Safety limit is set such that no fuel damage is calculated to occur if the limit is not violated.
Because fuel damage is not directly observable, a step-back approach is used to establish a $afety Limit such that the MCPR is not less than 1.04 for two-loop operation and 1.05 for single-loop operation.
These limits represent a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical carriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.
Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.
Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings.
While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.
Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.
These conditions represent a significant departure from the condition intended by design for planned operation.
4 The evaluations which justify normal operation, abnormal transient, and accident analyses for two-loop operation are discussed in detail in Reference 3.
Evaluation for single-loop operation demonstrates that two-loop transient i
and accident tralysts are more liniting than single-loop, Reference 4 2.1.1 THERWAL p0WER (Low Pressure or Low Flow)
The use of the GEXL correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means.
This is done by establishing a limiting condition on core THERMAL POWER with the following basis.
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi.
Analyses show that with a bundle flow of 28 x 108 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 10' lbs/hr.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking f actors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.
Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 osig is conservative.
HATCH - UNIT 2 B 2-1 Amendment No.105
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- e Bases Table B 2.1.2-1 UNCERTAINTIES U$ED IN THE 0ETERMINATION L
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0F THE FUEL CLADDING $AFETY LIMIT
- Standard t
l Quantity Deviation L
.(% of point)
Feedwater Flow 1.76 Feecwater Temperature' O.76 Reactor Pressure 0.5
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Core Inlet Temperature 0.2 o
I5 Core Total Flow 2.5 Channel Flow Area 3.0 i
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Friction Factor Multiplier 10.0 Channel Friction Factor Multiplier 5,0 l'
fr TIP Readings 8.7
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R Factor 1.5 Critical Pcwer 3.6 t
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- The uncertainty analysis used to establish the core wide Safety Limit MCPR is' based on the assumption of quadrant power symmetry for the reactor core.
HATCH - UNIT 2 B 2-4 Amendment No.105
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POWER DISTRIBUTION LIMITS BASES g
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N 3/4.2.2 APRM SETPOINTS This section deleted.
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L 3/4.2.3 MINIMUM CRITICAL POWER RATIO p
Jit The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.04 for two-loop operation and 1.05 l
for single-loop operation, and an analysis of abnormal operational transients as described in References 1 and 3.
For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting as given in Specification ~ 2.2.1.
To assure that the fuel cladding integrity Safety Limits are not exceeded
'during any anticipated abnormal operational transient, the most limiting transients-have been analyzed to-determine which results in the largest reduction in CRITICAL POWER RATIO (CPR).
The type of transients evaluated were loss of flow, increase in peessure and power, positive reactivity insertion, and coolant temperature decrease.
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HATCH - UNIT 2 B 3/4 2-3 Amendment No. 105 u