ML20205G372
| ML20205G372 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 03/09/1999 |
| From: | Adensam E NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20205G375 | List: |
| References | |
| NUDOCS 9904070229 | |
| Download: ML20205G372 (6) | |
Text
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UNITED STATES p
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,j j NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 20555-0001 o
PUBLIC EERVICE ELECTRIC & GAS COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-354 f
ELOFE CREEK GENERATING STATION 4
L AMENpMNT TO FACILITY OPERATING LICENSE Amendment No.117 j.
License No. NPF-57 1
1.
The Nuclear Regulatory Commission (the Commission) has found that:
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A.
The application for amendment filed by the Public Sente Electric & Gas-l Company (PSE&G) dated August 25,1998, as supplemented January 27,1999, complies with the standards and requirements of the Atomic Enorgy 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 la conformity with the appbcation, the provisions of the Act, and the rules and regulations of the Commission; C.
There is raasonable assurance:. (i) that the activities av'horized by this amerKlment can be conducted'without endangering the health ant / safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR ChUptrr I; l
D.
The issuance of this amendment wu not be inimical to the common defense and l
security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance witn W CFR Part 51 of the l
Commission's regulations and all applicable requirercents haye been satisfied.
2.'
Accordingly, the license is amended by char %es to the Techrdcal Spec.ifications as indicated in the attachmont to this licenst amendment and para graph 2 C (2) of Fa ility Operating License No. NPF-57 is hereby amended to read as follows:..
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. (2)
Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.117, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the license. PSE&G shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The lic*nse amendment is effective as of its date of issuance and shall be lmplemented within 60 days after the completion of Cycle 8.
FOR THE NUCLEAR REGULATORY COMMISSION Elinor G. Adensam, Director Project Directorate 1-2 Division of Licensing Project Managemen' Office of Nuclear Reactor Regulation Attach;nent Changes to the Technical Specifications Date of issuance: March 9, 1 W t
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ATTACHMENT TO LICENSE AMENDMENT NO.447 FACILID' OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following pages of tt.a Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remoy.e Insert 2-1 2-1 B 2-1 B 2-1 6-21 6-21 l
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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2aJ SAFETY LIMITS J
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:
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.
THERMAL POWER, Hioh Pressure and Hioh Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.09 l
with two recirculation loop operation and shall not be less than 1.11 with l
single recirculation loop operation, in_both cases with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.*
APPLICABILIIX: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With MCPR less than 1.09 with two recirculation loop operation or less than 1.11 with single recirculation-loop operation and in both cases with 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.
i REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not axceed 1325 psig.
APPLICABILITY: OPERATIONAL 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 SKUTDOWN with reactor ccolant 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.
- Values applicable to Cycle 9 operation only.
I HOPE CREEK 2-1 Amendment No. O l
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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 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 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 1.09 for two' recirculation loop operation and 1.11 for single recirculation loop operation. MCPR greater than 1.09.for twc recirculation loop operation and 1.11 for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain feel 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 i
perforations or cra : king. 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 I
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.
2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the applicable NRC-approved critical power correlation is not valid for all critical power calculations performed at reduced pressures below 785 psig or core flocs 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' 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 103 lbs/hr.
Full scale ATLAS test data taken at pressures i
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 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.
HOPE CREEK B 2-1 Amendment No.f@
l 117 1
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ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)
The analytical. methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in NEDE-24011-P-A (the latest approvod revision)*, General Electric Standard Application for Reactor Fuel (GESTAR II).
'The core operating limits shall be determined so that all applicable limits (e.g.,
fuel thermal-mechanical limits, core thermal-hydraulic limits,. ;CCS limits, nuclear limits such as shutdown margin, and transient and accioent analysis limits) of the safety analysis are met.
The CORE OPERA"ING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC 20555, with a copy to the USNRC Administrator, Region 1, within the time period specified for each report.
6.9.3 Violations of the requirements of the fire protection program described in the Final Safety Analysis Report which would have adversely affected the ability to achieve and maintain safe shatdown in the event of a fire shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC 20555, with a copy to the USNRC Administrator, Region 1, via the Licensee Event i
Report System within 30 days.
6.10 RECORD RETENTION 6.10.1 In addition to ths applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.
SPECIAL REPORTS 6.10.2 The following records shall be retained for at leacu 5 years:
a.
Records and logs of unit operation covering time interval at each p;wer level, b.
Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear sa*ety.
c.
All REPORTABLE EVENTS submitted to the Commission.
d.
Records of surveillance activities inspections, and calibrations I
required by these Technical Specifications, e.
Records of changes made to the procedures required by Specification 6.8.1.
f.
Records of radioactive shipments, g.
Records of sealed source and fission detector leak tests and results.
- For Cycle 9, as evaluated in the Safety Evaluation dated Marth 9, Iggte support License Amendment No.117 HOPE CREEK 6-21 knendment Nr. 497.117 l
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