ML20216B658
| ML20216B658 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/11/1998 |
| From: | Kuo P NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216B662 | List: |
| References | |
| NUDOCS 9805180319 | |
| Download: ML20216B658 (5) | |
Text
pitt%qk UNITED STATES p
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,j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. maa mg
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CAROLINA POWER & LIGHT COMPANY. et al.
DOCKET NO. 50-325 BRUNSWICK STEAM Et FCTRIC PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE j
Amendment No. 194 License No. DPR-71 1.
The Nuclear Regulatory Commission (the Commission) has found that:
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The application for amendment filed by Carolina Power & Light Company (the licensee), dated February 23,1998, as supplemented by letter dated March 27, 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 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; 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 satisfed.
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. DPR-71 is hereby amended to read as follows:
9805180319 990511 PDR ADOCK 05000325 P
2-(2)
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.194, are hereby incorporated in the license. Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be l
implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION i
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ao-Tsin Kuo, eting Director roject Directorate ll-1 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technica!
Specifications D te ofissuance:
May 11, 1998 i
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ATTACHMENT TO LICENSE AMENDMENT NO.194 FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 Replace the following pages of the Appendix A Technical Specifications with the enclosed pagos. The revised areas are indicated by marginallines.
Remove Paoes insert Paoes 2-1 2-1 6-23 6-23 i
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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 800 psia or core flow less than 10% of rated flow.
APPLICABILITY:
CONDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 800 psia 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 />'.'
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THERMAL POWER (Hiah Pressure and Hiah Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.09*
I with the reactor vessel steam dome pressure greater than 800 psia and core flow greater than 10% of rated flow.
APPLICABILITY: CONDITIONS 1 and 2.
ACTION:
With MCPR less than 1.09* and the reactor vessel steam dome pressure greater I
than 800 psia a.nd 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 />.
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: CONDITIONS 1. 2. 3. and 4.
ACTION-i 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 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 />.
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- MCPR values in Technical Specification 2.1.2 are applicable only for Cycle 12 operation.
BRUNSWICK - UNIT 1 2-1 Amendment No.194
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)
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b.
The ; ore flow and core power adjustments for Specification 3.2.2.1.
c.
The MINIMUM CRITICAL POWER RATIO (MCPR) for Specifications 3.2.2.1 and 3.2.2.2.
d.
The rod block monitor upscale trip setpoint and allwable value for Specification 3.3.4.
and shall be documented in the CORE OPERATING LIMITS REPORT.
6 9.3 2 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 documents, a.
NEDE-24011-P-A. " General Electric Standard Application for Reactor Fuel" (latest approved version).
b.
The May 18. 1964 and October 22, 1984 NRC Safety Evaluation Reports for the Brunswick Reload Methodologies described in:
1.
Topical Report NF-1583.01. "A Description and Validation of Steady-State Analysis Methods for Boiling Water Reactors."
February 1983.
2.
Topical Report NF-1583.02. " Methods of RECORD." February 1983.
3.
Topical Report NF-1583.03. " Methods of PREST 0-B."
February 1983.
4.
Topical Report NF-1583.04. " Verification of CP&L Reference BWR Thennal-Hydraulic Methods Using the FIBWR Code." May 1983.
c.
The NRC Safety Evaluation for Brunswick Unit 1 Amendment No.194.
l 6.9.3.3 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.
6.9.3.4 The CORE OPERATING LIMITS REPORT. including 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.
l BRUNSWICK - UNIT 1 6-23 Amendment No.194