ML20217Q500

Amend 127 to License NPF-39,revising Minimal Critical Power Ratio Safety Limits for Operation Cycle 8

ML20217Q500
Person / Time
Site:	Limerick
Issue date:	05/04/1998
From:	Capra R NRC (Affiliation Not Assigned)
To:
Shared Package
ML20217Q498	List: ML20217Q495 ML20217Q500 ML20217Q510
References
NUDOCS 9805080141
Download: ML20217Q500 (6)

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• Ritu u, . UNITED STATES

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1* i2 NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 3000HOO1

%.....g PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-352 LIMERICK GENERATING STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No, t27-License No. NPF-39

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Philadelphia Electric Company (the licensee) dated February 9,1998, as supplemented April 8 and 24,1998, 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 (i) that the activities a'uthorized 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; D. The issuance of this amendment will not be inimical to the common defcnse 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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PDR ADOCK 05000352  !

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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-3g is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.127 , are hereby l l

incorporated in the license. Philadelphia Electric Company shall operate the facility in I accordance with the Technical Specifications and the Environmental Protection Plan. i l

. 3. This license amendment is effective as of its date of issuance, and shall be implemented '

.'within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION l

he Robert A. Capra, Director Project Directorate I Division of Reactor Projects - 1/Il-Office of Nuclear Reactor Regulation l

Attachment:

Changes to the Technical

' Specifications l Date of Issuance: May 4, 1998 l

i ATTACHMENT TO LICENSE AMENDMENT NO.197 l FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352 l

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Replace the following pages of the Appendix A Technical Specifications with the attached page. I The revised pages are identified by Amendment number and contain vertical lines indicating the {

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Remove Ingad l 1 2-1 2-1 l 'B 2-1 B 2-1 l 6-18a 6-18a I

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L 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY' LIMITS 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' dome pressure less than 785 psig or core flow less than

10% of- rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:.

l 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%.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.

ll l1 THERMAL POWER. Hiah Pressure and Hiah Flow- {

l 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR)-shall not be less than 1.12 for j two recirculation loop operation and shall not be less than 1.14 for single recirculation loop 'operathn with the reactor _ vessel steam dome pressure greater-than 785 psig and core flw greater than 10% of rated flow.

' APPLICABILITY: OPERATIONAL CONDITIONS I and 2.

ACTION:

With MCPR less than 1.12 for'two recirculation loop operation.or. less than 1.14 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 hou'rs and comply with'the requirements of Specification 6.7.1.

REACTOR COOLANT-SYSTEM PRESSURE

2.1.3 The reactor. coolant _ system pressure, as measure'd in the reac't or vessel steam dome, shall not exceed'1325 psig. ,

j APPLICABILITY: OPERATIONAL CONDITIONS 1,-2, 3, and 4.

ACTION:-

'With the reactor coolant system pressure, as measured.in the reactor vessel

'l steam. dome, abova 1325 psig, be in at least HOT SHUTDOWN with the reactor coolant system pres:;ure 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. .

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LIMERICK - UNIT 1 2-1 Amendment No. 7,30,11f,127

2.1 SAFETY LIMITS BASES

2.0 INTRODUCTION

The fuel cladding, reactor' pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits.are established to protect the integrity of these barriors 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 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 1.12 for two recirculation loop operation and 1.14 for single recirculation loop operation. MCPR greater than 1.12 for two recirculation loop operation and 1.14 for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding 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 incre-mentally 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 3 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 signi-ficant departure from the cendition intended by design for planned operation.

The MCPR values for both dual-loop and single loop operation, listed above, are valid only for Cycle 8 operat an.

2.1.1 THERMAL POWER. Low Pressure or low Flew 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 10 lb/h, 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 grater than 28 x 10' lb/h. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly criti-cal power at this flow is approximately 3.35 MWt. With the design peaking factors, 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 psig is conservative.

LIMERICK - UNIT 1 B 2-1 Amendment No. 7,M ,111,127 l

isDMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload l cycle, or prior to any remaining portion of a reload cycle, and shall be

! documented in the CORE OPERATING LIMITS REPORT for the following:

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a. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for l Specification 3.2.1, j

b. MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1,

c. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3,

d. The MCPR(P) and MCPR(F) adjustment factors for specification 3.2.3,

e. The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4,

f. The power biased Rod Block Monitor setpoints and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.3.6,

g. The Reactor Coolant System Recirculation Flow upscale trip setpoint and allowable value for Specification 3.3.6,

h. The Recirculation MG set mechanical and electrical overspeed stop setpoints for Specification 4.4.1.1.2.

6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

a. NEDE-24011-P-A " General Electric Standard Application for Reactor Fuel" (Latest approved revision).

• l 6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.12 The CORE OPERATING LIMITS REPORT, incl Wing any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SfECIAL REPORTS 6.9.2 Special reports shall be subn.itted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

• For Cycle 8, specific documents were approved in the Safety Evaluation dated (5/4/98 ) to support License Amendment No. (127 ).

LIMERICK - UNIT 1 6-18a Amendment No. 127