ML20134D419
| ML20134D419 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 10/17/1996 |
| From: | Reinhart M NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20134D422 | List: |
| References | |
| NUDOCS 9610220200 | |
| Download: ML20134D419 (9) | |
Text
{{#Wiki_filter:_.. m p Kro h y UNITED STATES j ,, j NUCLEAR REGULATORY COMMISSION 't WASHINGTON, D.C. 206eMm01 49.....,o CAROLINA POWER & LIGHT COMPANY. et al. i DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT. Unit 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.182 License No. DPR-71 ] 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment filed by Carolina Power & Light Company (the licensee), dated April 8,1996, as supplemented on July 30, 1996, October 4, 1996, October 8. 1996, and October 16, 1996, 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; 1 C. There is reasonable assurance (i) that the activities authorized j 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; 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; 4 and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements t have been satisfied. I 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: 4 9610220200 961017 PDR ADOCK 05000325 { P PDR
. (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.182, 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 as of the date of its issuance and shall be implemented prior to the startup of Unit 1 from the Refueling Outage 10 (BillRI). FOR THE NUCLEAR REGULATORY COMMISSION D ,O i libb W, a Mark Reinhart, Acting Dire tbr Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: October 17, 1996
ATTACHMENT TO LICENSE AMEN 0 MENT NO.182 FACILITY OPERATING LICENSE N0. DPR-71 DOCKET NO. 50-325 1 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines. Remove Paaes Insert Paaes XVI XVI l 2-1 2-1 3/4 1-20 3/4 1-20 5-1 5-1 i 6-23 6-23 6-23a i l 1
iwa ADMIN 1STRATIVE CONTROLS SECTION [BGE 6.10 RECORD RETENTION..................... 6-23a I 6.11 RADIATION PROTECTION PROGRAM.... 6-25 6.12 HIGH RADIATION AREA............. 6-25 M 0FFSITE DOSE CALCULATION MANUAL (0DCM)...... 6-26 6.14 PROCESS CONTROL PROGRAM (PCP). 6-26 M MAJOR CHANGES TO LIOUID. GASEOUS. AND SOLID WASTE TREATMENT SYSTEMS.. 6-27 1 I J t l 6 i BRUNSWICK - UNIT 1 XVI Amendment No.182 ) i i A
_A_. ._m.__.. - - ~ - ~ - - a 'q 2.1 SAFETY LIMlTS BiERMAL POWER (Low Pressure or low Flow) 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the j reactor vessel steam dome pressure less than 800 psia or core flow less than 10% of rated flow. j APPLICABJLITY: CONDITIONS 1 and 2. ACiCN: 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 hours. THERMAL POWER (Hiah Pressure and Hiah Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.10* wi; the reactor vessel steam dome pressure greater than 800 psia and core flow greater than 10% of rated flow. l APPLICABILITY: CONDITIONS 1 and 2. ACTION: ) i With MCPR less than 1.10* and the reactor vessel steam dome 3ressure greater I ) than 800 psia and core flow greater than 10% of rated flow. Je in at least HOT 3 SH'JTDOWN within 2 hours. l 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. 1 l APPLICABILITY: 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 s 1325 psig within 2 hours. I f 4
- MCPR values in Technical Specification 2.1.2 are applicable only for Cycle 11 operati'.
BRUNSWICK - UNIT 1 2-1 Amendment No.182
~ FlGURE 3.1.5-1 1 i SODIUM PENTABORATE SOLUTION VOLUME l l CONCENTRATION REQUIREMENTS t i [ I I I Sodium Pentaborate Solution Volume Concentration Requirements l 22 0 i f 21 0 w i l i - - - --.- Region of Required volume correntrsten 20.0 i ~~~~~ -- ~~~~~- ~ ~ Acceptable Opetion Range i g 19 c, ..j............ l 18.0l--~ ~ ~~. -{. 1 - ~. ~ ~ + - L- . + ~ ~ ~ ~ - + ~ - ~ ~ - ~ ~. g d 170 l l Tres vdmancertrmon range i j j ... 4.... mons. r ciar vos nna coran 4.............4.-............. 5 conca8ca o' * 'a'5' t'D run-i = 16 0 E
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-+~~-~3~~~- 10 0 2000 2500 3000 3500 4000 4500 5000 5500 Net Volurne d Solution in Tank (gals) l BRUNSWICK - UNIT 1 3/4 1-20 Amendment No.182
5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1. LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1.2-1. based on the information given in Section 2.2 of the FSAR. SITE BOUNDARY 5.1.3 The SITE BOUNDARY shall be as shown in Figure 5.1.3-1. For the purpose of effluent release calculations, the boundary for atmospheric releases is the SITE BOUNDARY and the boundary for liquid releases is the SITE BOUNDARY prior to dilution in the Atlantic Ocean. 5.2 CONTAINMENT CONFIGURATION 5.2.1 The PRIMARY CONTAINMENT is a steel-lined, reinforced concrete structure composed of a series of vertical right cylinders and truncated cones which form a drywell. This drywell is attached to a suppression chamber through a series of vents. The suppression chamber is a concrete, steel-lined pressure vessel in the shape of a torus. The primary containment has a minimum f we air volume of 288.000 cubic feet. DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for: a. Maximum internal, pressure 62 psig. b. Maximum internal temperature: drywell 300*F suppression chamber 200*F c. Maximum external pressure 2 psig. 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 560 fuel assemblies limited to the following fuel types: BP8x8R. GE8x8EB. GE8x8NB-3. and GE13. I ~ BRUNSWICK - UNIT 1 5-1 Amendment No. 182
~ ~ ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) b. The core 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 allowable 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. 1984 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. To)ical Report NF-1583.03. " Methods of PRESTO-B." Fe)ruary 1983. 4. Topical Report NF-1583.04. " Verification of CP&L Reference BWR Thermal-Hydraulic Methods Using the FIBWR Code." May 1983. c. The NRC Safeb Evaluation for Brunswick Unit 1 Amendment No. 182. I 6.9.3.3 The core operating limits shall be determined such that all a)plicable limits (e.g., fuel thermal-mechanical limits. core tiermal-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 3rovided, upon issuance for each reload cycle, to the NRC Document Control Dest with copies to the Regional Administrator and Resident Inspector. BRUNSWICK - UNIT 1 6-23 Amendment No. 182
6.10 ' RECORD RETENTION Facility records shall be retained in accordance with ANSI-N45.2.9-1974. 6.10.1 The following records shall be retained for at least five years: a. Records and logs of facility operation covering time interval at each power level. i 1 i i i BRUNSWICK - UNIT 1 6-23a Amendment No.182 l i .. _. _ -... _}}