ML20129F536

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Proposed Tech Specs,Revising Minumum Critical Power Ratio Safety Limit Ref in TS 2.1.2 from 1.07 to 1.10 for Unit 1 Cycle 11 Operation & Revises TS 5.3.1 to Reflect New Fuel Type Being Added to Unit 1 Reactor Core
ML20129F536
Person / Time
Site: Brunswick Duke Energy icon.png
Issue date: 10/16/1996
From:
CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20129F523 List:
References
NUDOCS 9610290154
Download: ML20129F536 (9)
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Text

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INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.10 RECORD RETENTION. . .. . .. .. . . . . 6-23a I 6.11 RADIATION PROTECTION PROGRAM. . .. . . 6-25 6.12 HIGH RADIATION AREA. . . 6-25 6.13 0FFSITE DOSE CALCULATION MANUAL (ODCM). . . . 6-26 6.14 PROCESS CONTROL PROGRAM (PCP).. .. . .. . . . 6-26 6.15 MAJOR CHANGES TO LIQUID. GASEOUS. AND SOLID WASTE TREATMENT SYSTEMS. . . . .. . 6-27 i

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l 9610290154 961016 PDR ADOCK 05000325 P PDR BRUNSWICK - UNIT 1 XVI Amendment No.

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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.

t 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 />.

THERMAL POWER (Hiah Pressure and Hiah Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.10* 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 1.10* 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 SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

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.

APPLICABILITY: CONDITIONS 1. 2. 3. and 4.

ACTION:

l 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 11 operation.

BRUNSWICK - UNIT 1 2-1 Amendment No.

5.0 DESIGN FEATURES 5.1 SITF, 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 free I

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 L FUEL ASSEMBLIES l l 5.3.1 The reactor core shall contain 560 fuel assemblies limited to the following fuel types: BP8x8R. GE8x8EB. GE8x8NB-3, and GE13. l l

BRUNSWICK - UNIT 1 5-1 Amendment No.

', ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) i

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 I and 3.2.2.2. l 1
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 1 described in the following documents.

a. NEDE-24011-P-A, " General Electric Standard Application for Reactor l Fuel" (latest approved version).

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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 PREST 0-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 Safety Evaluation for Brunswick Unit 1 Amendment No. 182. l 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 incluaing any mid-cycle revisions or supplements shall be 3rovided, upon issuance for each reload cycle, to the NRC Document Control Desc with copies to the Regional Administrator and Resident Inspector.

BRUNSWICK - UNIT 1 6-23 Amendment No.

ADMINISTRATIVr' CONTROLS 6.10 RECORD GETENTION 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.

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1 BRUNSWICK - UNIT 1 6-23a Amendment No. i

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ENCLOSURE 3

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BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO.1 1 NRC DOCKET NO. 50-325 I OPERATING LICENSE NO. DPR-71 SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENT FUEL CYCLE 11 RELOAD LICENSING ,

(NRC TAC NO. M95263) )

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MARKED-UP TECHNICAL SPECIFICATION PAGES - UNIT 1 I

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3 . ,

l INDEX a .

ADMINISTRATIVE CONTROLS

, SECTION PACE

. 6.10 RECORD RETENTION.......................................... -23a. f

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j .. 6.11 RADIATION PROTECTION PR0 CRAM.............................. 6-25 3

} 6.12 HICH RADIATION AREA....................................... 6-25 4

, 'l 6.13 0FFSITE DOSE CALCULATION MANUAL (ODCM).................... 6-26 ( 'l l

i' 6. 14 PROCESS CONTROL PROGRAM (PCP)..................'........... 6-26 l i 6.15 MAJOR CHANCES TO LIQUID, CASEOUS, AND ,

i SOLID WASTE TREATMENT SYSTEMS.......................... 6-27 4

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i BRUNS'JICK - UNIT L XVI Amendment No. M s

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2.0 SAFETY LIMITS AND LIMITINC SAFETY SYSTEM SETTINGS l

2.1 SAFETY LIMITS THERMAL POWER (Low Pressure or Low Flow) i 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 I 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 />.

  • THERMAL POWER (High Pressure and High Flow) g 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than +-e+

with the reactor vessel steam dome pressure greater than 800 psia and core flow greater than 10% of rated flow.

APPLICABILITY: CONDITIONS I and 2.

ACTION: g l.lD With MCPR less than &-6+ and the reactor vessel steam dome pressure greater l

! than 800 psia 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 />. .

L REACTOR COOLANT SYSTEM PRESSURE '

2.1.3 The reactor coolant system pressure, as measured in the reactor vessel  !

l steam dome, shall not exceed 1325 psig.

i APPLICABILITY: CONDITIONS 1, 2, 3, and 4.

i l 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 < 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 Techntent SpeciRcaten '2..t.2. are appucable l only & Cycle 11 opercMon.

l BRUNSWICK - UNIT 1 2-1 Amendment N o . .Mt--

  • e ADMINISTRATIVE CONTROLS CORE OPERATINC LIMITS REPORT (Continued)
b. The core flow and core power adjustments for Specification 3.2.2.1 4
c. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specifications 3.2.2.1 and 3.2.2.2. {'k
d. The rod block monitor upscale trip setpoint and allowable value for Specification 3.3.4.

/[

and shall be documented in the CORE OPERATINC LIMITS REPORT.

6.9.3.2 The analytical methods used to determine the core operat,ing limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.

a. NEDE-240ll-P-A, "Ceneral 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. TopicaljReport NF-1583.02, " Methods of RECORD," February 1983.
3. Topical Report NF-1583.03, " Methods of PRESTO-B," Februa ry 1983.

4 Topical Report NF-1583.04, " Verification of CP&L Reference BWR Thermal-Hydraulic Methods Using the FIBWR Code " May 1983.

C.he NR0 Sdety Evalunden krErunswick linit t Amendme,d No.182..

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.

6.10 RECORD RETENTION Facility records shall be retained in accordance with ANSI-N45.2.9-1974.

if biovE 7"o 6.10.1 The following records shall be retained for at least five years:

PAGE" 4,g3 a. Records and logs of facility operation covering time-interval at each power level.

BRUNSWICK - UNIT 1 6-23 Amendment N o . 14+T- -

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