ML20134M349
| ML20134M349 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 02/12/1997 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Philadelphia Electric Co |
| Shared Package | |
| ML20134M351 | List: |
| References | |
| NPF-85-A-087 NUDOCS 9702200208 | |
| Download: ML20134M349 (5) | |
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4 UNITED STATES j
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 200 6 4001 1
PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-353 LIMERICK GENERATING STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.87 License No. NPF-85 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Philadelphia Electric Company (the licensee) dated December 6, 1996, as supplement by letter dated January 15, and 28, 1997, 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 Commission; C.
There is reasonable assurance (i) that the activities authorized by 4
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; 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.
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i 9702200208 970212 PDR ADOCK 05000353 P
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. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-85 is hereby amended to read as follows:
Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised i
through Amendment No. 87, are hereby incorporated into this license.
Philadelphia Electric Company shall operate the facility in accordance i
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.
FOR THE NUCLEAR REGULATORY COMISSION L
Jphn F. Stolz, Director Pto ect Directorate I-Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: February 12, 1997
1 ATTACHMENT TO LICENSE AMENDMENT NO.87 FACILITY OPERATING LICENSE NO. NPF-85 I
DOCKET NO. 50-353
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Replace the following pages of the Appendix A Technical Specifications with the attached page. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert 2-1 2-1 B 2-1 B 2-1 1
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4 2
A
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1 a
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1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS j
l THERMAL POWER. Low Pressure or low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam done pressure less than 785 psig or core flow less than 10% of rated i
- flow, i
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam done pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of l
Specification 6.7.1.
THERMAL POWER. Hioh Pressure and Hiah Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR).shall not be less than 1.11 for two l
recirculation loop operation and shall not be less than 1.12 for single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With MCPR less than 1.11 for two recirculation loop operation or less than 1.12 l
for single recirculation loop operation and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.
APPLICABILITY: OPERATION CONDITIONS 1, 2, 3, and 4.
ACTION:
With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
LIMERICK - UNIT 2 2-1 Amendment No. H,83, 87
2.1 SAFETY LIMITS BASES 2.0 INTR 000CTION I
The fuel cladding,-reactor pressure vessel and primary system piping are the principle barriers to the release of radioactive materials to the environs.
Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit i
is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.11 for two recirculation loop operation and 1.12 for single recirculation loop j
operation. MCPR greater than 1.11 for two recirculation loop operation and 1.12 for single recirculation loop operation represents a conservative margin relative i
to the conditions required to maintain fuel cladding integrity. The fuel cladding i
is one of the physical barriers 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 4
i cladding perforations, however, can result from thermal stresses which occur from i
reactor operation significantly above design conditions and the Limiting Safety 1
System Settings. While fission product migration from cladding perforation is j
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 j
conditions represent a significant departure from the condition intended by design i
for planned operation. The MCPR values for both dual-loop and single loop operation, j
listed above, are valid only for-Cycle 5 operation.
j 4
2.1.1 THERMAL POWER. Low Pressure or low Flow i
The use of the (GEXL) correlation is not valid for all critical power i
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 i
means. This is done by establishing a limiting condition on core THERMAL POWER 1
with the following basis. Since the pressure drop in the bypass region is i
essentially all elevation head, the core pressure drop at low power and flows will i
always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 1
10* lb/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' lb/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 1
i flow is approximately 3.35 MWt. With the design peaking factors, this corresponds 1
to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER i
limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
LIMERICK - UNIT 2 B 2-1 Amendment No. H, 83, 87
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