Text

Cycle 14 Core Operating Limits Report, Supplemental Reload Licensing Report & Loss-Of-Coolant Accident Analysis Report for Operation with Extended Power Uprate
	ML021560418
Person / Time
Site:	Brunswick
Issue date:	05/31/2002
From:	O'Neil E; Carolina Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BSEP 02-0108 J11-03936SRLR, Rev 2, NEDC-31624P, Suppl 1, Rev 7
	Download: ML021560418 (66)
	v • d • e

SCP&L A Progress Energy Company MAY 3 1 2002 SERIAL: BSEP 02-0108 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 1 DOCKET NO. 50-325/LICENSE NO. DPR-71 UNIT 1 CYCLE 14 CORE OPERATING LIMITS REPORT, SUPPLEMENTAL RELOAD LICENSING REPORT, AND LOSS-OF-COOLANT ACCIDENT ANALYSIS REPORT FOR OPERATION WITH EXTENDED POWER UPRATE Ladies and Gentlemen:

The purpose of this letter is to submit the latest revision of the Core Operating Limits Report, Supplemental Reload Licensing Report, and Loss-of-Coolant Accident Analysis Report for Carolina Power & Light (CP&L) Company's Brunswick Steam Electric Plant (BSEP), Unit No. 1, for operation with extended power uprate.

Technical Specification 5.6.5.d requires that the Core Operating Limits Report, including any midcycle revisions or supplements, be provided to the NRC upon issuance. A copy of "Brunswick Unit 1, Cycle 14 Core Operating Limits Report May 2002," Revision 1, is provided in Enclosure 1. A plot of the limiting value of APLHGR, as a function of average planar exposure for each reload fuel type, is included in the enclosed Core Operating Limits Report.

A copy of "Supplemental Reload Licensing Report for Brunswick Steam Electric Plant Unit i Reload 13 Cycle 14," J 11-03936SRLR, Revision 2, Class I, dated March 2002, is provided in Enclosure 2. The most limiting and least limiting MAPLHGR values for the applicable GE13 and GE14 fuel types are provided in a table included in the Supplemental Reload Licensing Report for BSEP, Unit 1.

A copy of "Loss-of-Coolant Accident Analysis Report for Brunswick Steam Electric Plant Unit 1 Reload 13 Cycle 14," NEDC-31624P, Supplement 1, Revision 7, Class III, dated March 2002, is provided in Enclosure 3. This report contains MAPLHGR, as a function of exposure, for each lattice of the fuel designs.

These reports are applicable to operation at the maximum power level authorized by the facility operating license, as revised by License Amendment No. 222 issued May 31, 2002 (i.e., 2923 megawatts thermal). These reports supercede, in their entirety, the reports submitted by CP&L's letter dated March 22, 2002 (i.e., BSEP 02-0059) [ADAMS Accession Number ML020860125].

Brunswick Nuclear Plant PO Box 10429 Southport, NC 28461 ,1 h L

Document Control Desk BSEP 02-0108 / Page 2 Global Nuclear Fuel considers portions of the revised Loss-of-Coolant Accident Analysis Report in Enclosure 3 to be proprietary information, as indicated by the bars drawn in the margin of the report. Therefore, the document should be withheld from public disclosure in accordance with 10 CFR 9.17 and 10 CFR 2.790. An affidavit supporting the request for withholding the document is provided in Enclosure 4.

There are no regulatory commitments being made in this submittal. Please refer any questions regarding this submittal to Mr. Leonard R. Beller, Supervisor Licensing/Regulatory Programs, at (910) 457-2073.

Sincerely, Edward T. O'Neil Manager - Regulatory Affairs Brunswick Steam Electric Plant WRM/wrm

Enclosures:

1. "Brunswick Unit 1, Cycle 14 Core Operating Limits Report May 2002," Revision 1

2. "Supplemental Reload Licensing Report for Brunswick Steam Electric Plant Unit 1 Reload 13 Cycle 14," J1 1-03936SRLR, Revision 2, Class I, dated March 2002

3. "Loss-of-Coolant Accident Analysis Report for Brunswick Steam Electric Plant Unit 1 Reload 13 Cycle 14," NEDC-31624P, Supplement 1, Revision 7, Class III, dated March 2002 (Proprietary Information)

4. Global Nuclear Fuels Affidavit Regarding Withholding From Public Disclosure In Accordance With 10 CFR 9.17 and 10 CFR 2.790

Document Control Desk BSEP 02-0108 / Page 3 cc (with enclosures):

U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Luis A. Reyes, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Theodore A. Easlick, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)

ATTN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9) 11555 Rockville Pike Rockville, MD 20852-2738 cc (without enclosures):

Ms. Jo A. Sanford Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510

ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. I DOCKET NO. 50-325/LICENSE NO. DPR-71 UNIT 1 CYCLE 14 CORE OPERATING LIMITS REPORT, SUPPLEMENTAL RELOAD LICENSING REPORT, AND LOSS-OF-COOLANT ACCIDENT ANALYSIS REPORT FOR OPERATION WITH EXTENDED POWER UPRATE "Brunswick Unit 1, Cycle 14 Core Operating Limits Report May 2002,"

Revision 1

U I CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 B1 C14 Core Operating Limits Report Page 1, Revision 1 SCP&L APmogress En ry Conp*r, BRUNSWICK UNIT 1, CYCLE 14 CORE OPERATING LIMITS REPORT May 2002 Prepared By: -%-Mmu 14 Date: 65-- q--

Mourad Aissa Approved By: Date: __________ý "eorgeE. Smith Superintendent BWR Fuel Engineering I

I

CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 B1C14 Core Operating Limits Report Page 2, Revision 1 LIST OF EFFECTIVE PAGES Page(s) Revision 1-28 1 CP&L A.Progies Enewg Crmpmn

CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 B1C14 Core Operating Limits Report Page 3, Revision 1 TABLE OF CONTENTS Subject Pae Co v e r ................................................................ . ................................................ .............................. I L ist o f E ffe ctive P ag e s....................................................................................................................... 2 T ab le o f C o ntents .............................................................................................................................. 3 List o f T ab le s .................................................................................................................................... 4 List o f F igure s ................................................................................................................................... 4 Introduction and Summary ........................................................................................................... 5 Sin g le L o o p O p eratio n ...................................................................................................................... 6 Inoperable M ain Turbine Bypass System ..................................................................................... 6 AP L H GR L im its ............................................................................................................................... 7 M C P R L imits .................................................................................................................... ................ 7 RBM Rod Block Instrumentation Setpoints ................................................................................. 7 Stab ility O ption III ............................................................................................................................ 8 Refe r e n ce s ........................................................................................................................................ 9 ACP&L AProgres Enerw* Cwqmy'

CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 B1C14 Core Operating Limits Report Page 4, Revision 1 S~CAUTION References to COLR Figures or Tables should be made using titles only; figure and table numbers

[may change from cycle to cycle. ...

LIST OF TABLES Table Title Page T ab le 1: MC P R L im its ........................................................................................................................................ 10 T able 2 : RB M System Setpoints ........................................................................................................................ 11 T ab le 3 : P B D A S etp oin ts ................................................................................................................................... 12 LIST OF FIGURES Figure Title or Description Page Figure 1: APLHGR Limit Versus Average Planar Exposure ......................................................................... 13 Figure 2: APLHGR Limit Versus Average Planar Exposure ....................................................................... 14 Figure 3: APLHGR Limit Versus Average Planar Exposure ....................................................................... 15 Figure 4: APLHGR Limit Versus Average Planar Exposure ......................................................................... 16 Figure 5: APLHGR Limit Versus Average Planar Exposure ....................................................................... 17 Figure 6: APLHGR Limit Versus Average Planar Exposure ....................................................................... 18 Figure 7: APLHGR Limit Versus Average Planar Exposure ...................................................................... 19 Figure 8 : Not U sed ............................................................................................................................................... 20 Figure 9: GE 13 and GE 14 Flow-Dependent MAPLHGR Limit, MAPLHGR(F) ......................................... 21 Figure 10: GE13 and GE14 Power-Dependent MAPLHGR Limit, MAPLHGR(P) ...................................... 22 Figure 11: GE13 and GE14 Flow-Dependent MCPR Limit, MCPR(F) ......................................................... 23 Figure 12: GEl3 and GEl4 Power-Dependent MCPR Limit, MCPR(P) ....................................................... 24 Figure 13: Stability Option III Power/Flow Map: OPRM Operable, Two Loop Operation, 2923 MWt ........... 25 Figure 14: Stability Option III Power/Flow Map: OPRM Inoperable, Two Loop Operation, 2923 MWt ......... 26 Figure 15: Stability Option III Power/Flow Map: OPRM Operable, Single Loop Operation, 2923 MWt ........ 27 Figure 16: Stability Option mIPower/Flow Map: OPRM Inoperable, Single Loop Operation, 2923 MWt ...... 28 N', C A P&mL AProgrew Enermy Ccrrwi

CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 B1 C14 Core Operating Limits Report Page 5, Revision 1 Introduction and Summary This COLR revision was performed to support operation at up to 2923 MWt. The main changes are those associated with the thermal limits and Power-Flow maps. Also, the following thresholds were scaled for the higher rated thermal power: the thermal limit monitoring threshold changed from 25%

to 23% and the turbine trip scram bypass threshold changed from 30% to 26%. This report provides the values of the power distribution limits and control rod withdrawal block instrumentation setpoints for Brunswick Unit 1, Cycle 14 as required by TS 5.6.5.

OPERATING LIMIT REQUIREMENT Average Planar Linear Heat Generation Rate (APLHGR) limits TS 5.6.5.a. 1 (with associated core flow and core power adjustment factors)

Minimum Critical Power Ratio (MCPR) limits TS 5.6.5.a.2 (with associated core flow and core power adjustment factors)

Period Based Detection Algorithm (PBDA) Setpoint for Function 2.f of TS 3.3.1.1, TS 5.6.5.a.3 Oscillation Power Range Monitor (OPRM)

Allowable Values and power range setpoints for Rod Block Monitor Upscale Functions TS 5.6.5.a.4 of TS 3.3.2.1 Per TS 5.6.5.b and 5.6.5.c, these values have been determined using NRC approved methodology and are established such that all applicable limits of the plant safety analysis are met. The limits specified in this report support single loop operation (SLO) as required by TS LCO 3.4.1 and inoperable Main Turbine Bypass System as required by TS 3.7.6.

In order to support the Stability Option III with an inoperable OPRM scram function, the following is also included in this report:

OPERATING LIMIT REQUIREMENT BWROG Interim Corrective Action Stability Regions TS 3.3.1.1 LCO Condition I This report conforms to Quality Assurance requirements as specified in Reference 1.

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 11B21-0604 B1C14 Core Operating Limits Report Page 6, Revision 1 Single Loop Operation Brunswick Unit 1, Cycle 14 may operate over the entire MEOD range with Single recirculation Loop Operation (SLO) as permitted by TS 3.4.1 with applicable limits specified in the COLR for TS LCO's 3.2.1, 3.2.2 and 3.3. 1.1. The applicable limits are:

LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR) Limits: per Reference 1, the Figures 9 and 10 described in the APLHGR Limits section below include a SLO limitation of 0.8 on the MAPLHGR(F) and MAPLHGR(P) multipliers.

LCO 3.2.2, Minimum Critical Power Ratio (MCPR) Limits: per Reference 1, Table I and Figures 11 and 12, the MCPR limits presented apply to SLO without modification.

LCO 3.3.1.1, Reactor Protection System Instrumentation Function 2.b (Average Power Range Monitors Simulated Thermal Power - High) Allowable Value: per footnote b, the -AW offset value is defined in Plant procedures. The current value of 5% developed for the initial installation of Stability Option III is used for the B IC 14 COLR.

Inoperable Main Turbine Bypass System Brunswick Unit 1, Cycle 14 may operate with an inoperable Main Turbine Bypass System in accordance with TS 3.7.6 with applicable limits specified in the COLR for TS LCO 3.2.1 and 3.2.2.

Two or more bypass valves inoperable renders the System inoperable, although the Turbine Bypass Out-of-Service (TBPOOS) analysis supports operation with all bypass valves inoperable for the entire MEOD range and up to 110 IF rated equivalent feedwater temperature reduction. The system response time assumed by the safety analyses from event initiation to start of bypass valve opening is 0.10 seconds, with 80% bypass flow achieved in 0.30 seconds. The applicable limits are as follows:

LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR) Limits: in accordance with Reference 1 as shown in Figure 10, TBPOOS does not require an additional reduction in the MAPLGHR(P) limits between 23% and 26% power, as the Turbine bypass Operable and Inoperable limits are identical.

LCO 3.2.2, Minimum Critical Power Ratio (MCPR) Limits: in accordance with Reference 1, TBPOOS does not require an additional increase in the MCPR(P) multiplier between 23% and 26% power, as shown in Figure 12, as the Turbine bypass Operable and Inoperable limits are identical. TBPOOS requires increased MCPR limits, included in Table 1.

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1 B21-0604 B1C14 Core Operating Limits Report Page 7, Revision 1 APLHGR Limits The limiting APLHGR value for the most limiting lattice (excluding natural uranium) of each fuel type as a function of planar average exposure is given in Figures 1 through 7. These values were determined x,ith the SAFER/GESTR LOCA methodology described in GESTAR-I1 (Reference 2).

Figures I through 7 are to be used only when hand calculations are required as specified in the bases for TS 3.2.1. Hand calculated results may not match a POWERPLEX calculation since normal monitoring of the APLHGR limits with POWERPLEX uses the complete set of lattices for each fuel type provided in Reference 3.

The core flow and core power adjustment factors for use in TS 3.2.1 are presented in Figures 9 and

10. For any given flow/power state, the minimum of MAPLHGR(F) determined from Figure 9 and MAPLHGR(P) determined from Figure 10 is used to determine the governing limit.

MCPR Limits The ODYN OPTION A, ODYN OPTION B, and non-pressurization transient MCPR limits for use in TS 3.2.2 for each fuel type as a function of cycle average exposure are given in Table 1. These values were determined with the GEMINI methodology and GEXL-PLUS critical power correlation described in GESTAR-I (Reference 2), and are consistent with a Safety Limit MCPR of 1.12 specified by TS 2.1.1.2.

The core flow and core power adjustment factors for use in TS 3.2.2 are presented in Figures 11 and

12. For any given power/flow state, the maximum of MCPR(F) determined from Figure 11 and MCPR(P) determined from Figure 12 is used to determine the governing limit.

All MCPR limits presented in Table 1, Figure 11 and Figure 12 apply to two recirculation pump operation and SLO without modification.

RBM Rod Block Instrumentation Setpoints The nominal trip setpoints and allowable values of the control rod withdrawal block instrumentation for use in TS 3.3.2.1 (Table 3.3.2.1-1) are presented in Table 2. These values were determined consistent with the bases of the ARTS program and the determination of MCPR limits with the GEMINI methodology and GEXL-PLUS critical power correlation described in GESTAR-II (Reference 2). Reference 8 revised certain of these setpoints to reflect changes associated with the installation of the new PRNM system.

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 B1C14 Core Operating Limits Report Page 8, Revision 1 Stability Option MI Brunswick Unit 1 has implemented BWROG Long Term Stability Solution Option III (Oscillation Power Range Monitor-OPRM) as described in Referei.ce 4. Plant specific analysis incorporating the Option III hardware is described in Reference 5. Reload validation has been performed in accordance with Reference 6. The resulting stability based MCPR Operating Limit is provided for two conditions as a function of OPRM amplitude setpoint in Table 3. The reload validation calculation demonstrated that reactor stability does not produce the limiting OLMCPR for Cycle 14 as long as the selected OPRM setpoint produces values for OLMCPR(SS) and OLMCPR(2PT) which are less than the corresponding acceptance criteria. Because the acceptance criteria for OLMCPR(SS) is 1.50 and for OLMCPR(2PT) is 1.40, an OPRM setpoint (Amplitude Setpoint Sp) of 1.15 is supported for Cycle 14 without imposing any additional operational restrictions for stability protection. Therefore the OPRM PBDA setpoint limit referenced by function 2.f of Table 3.3.1.1-1 of Technical Specification 3.3.1.1 is 1.15 for Cycle 14. Per Table 3-2 of Reference 6, an Sp value of 1.15 supports selection of a Confirmation Count Setpoint Np of 16 or less.

Four Power/Flow maps for use at up to 2923 MWt (Figures 13-16) were developed based on Reference 7 to facilitate operation under Stability Option III as implemented by function 2.f of Table 3.3.1.1-1 and LCO Condition I of Technical Specification 3.3.1.1. The corresponding Reference 7 maps are simply re-formatted (no change in data) to exhibit the appropriate headers for the COLR.

All four maps illustrate the region of the power/flow map above 25% power and below 60% flow where the system is required to be enabled.

The maps supporting an operable OPRM function 2.f (Figures 13 and 15) show the same Scram Avoidance Region, which is not a licensing requirement but is an operator aid to illustrate where the OPRM system is reasonably likely to generate a scram to avoid an instability event. Figures 13 and 15 differ only in that the Figure 15 that supports SLO, indicates the maximum allowable core flow at 45 Mlbs/hr, and has the Simulated Thermal Power (STP) scram and rod block limits appropriately reduced for SLO. Note that the STP scram and rod block limits are defined in Technical Specifications, the Technical Requirements Manual, and Plant procedures, and are included in the COLR as an operator aid rather than a licensing requirement.

The maps (Figures 14 and 16) supporting an inoperable OPRM function 2.f show the BWROG-94078 Interim Corrective Actions stability regions required to support LCO Condition I.

Both figures also include a 5% Buffer Region around the Immediate Exit Region as an operator aid.

Figures 14 and 16 differ only in that the Figure 16 that supports SLO, indicates the maximum allowable core flow at 45 Mlbs/hr, and has the STP scram and rod block limits appropriately reduced for SLO.

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1 B21-0604 B1 C14 Core Operating Limits Report Page 9, Revision 1 References

1) BNP Design Calculation 1B21-0604;' "Preparation of the BIC14 Core Operating Limits Report,"

Revision 1, April 2002.

2) NEDE-2401 1-P-A; "General Electric Standard Application for Reactor Fuel," (latest approved version).

3) NEDC-31624P, "Loss-of-Coolant Accident Analysis Report for Brunswick Steam Electric Plant Unit I Reload 13 Cycle 14," Supplement 1, Revision 7, March 2002.

4) NEDO-31960-A, "BWR Owners Group Long-Term Stability Solutions Licensing Methodology,"

November 1995.

5) GE-NE-C51-00251-00-01, Revision 0, "Licensing Basis Hot Bundle Oscillation Magnitude for Brunswick I and 2," March 2001.

6) NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Application," August 1996

7) Design Calculation 0B21-1015, Revision 1, "BNP Power/Flow Maps for Stability Option III,"

February 2002.

8) Design Calculation IC5 1-0001 Revision 1, "BNP Power Range Neutron Monitoring System Setpoint Uncertainty and Scaling Calculation (1-C5 1-APRM 1 through 4 Loops and 1-C5 1 RBM-A and B Loops," July 2001.

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 Page 10, Revision 1 B1C14 Core Operating Limits Report Table 1 MCPR Limits Steady State, Non-pressurization Transient MCPR Limits Fuel Type Exposure Range: BOC - EOC GE13 and GE14 1.26 Pressurization Transient MCPR Limits, OLMCPR (100%P): Turbine Bypass System Operable Normal and Reduced Feedwater Temperature Exposure Range: Exposure Range:

MCPR Option Fuel Type BOC to EOFPC-2026 EOFPC-2026 MWd/MT to EOC MWd/MT A GE1 3 1.43 1.49 GE14 1.56 1.68 B GE13 1.38 1.41 GE14 1.45 1.51 Pressurization Transient MCPR Limits, OLMCPR (100%P): Turbine Bypass System Inoperable Normal and Reduced Feedwater Temperature MCPR Option Fuel Type BOC to EOC A GE13 1.50 GE14 1.69 B GE13 1.42 GEl4 1.52 This Table is referred to by Technical Specifications 3.2.2, 3.4.1 and 3.7.6.

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 B1C14 Core Operating Limits Report Page 11, Revision 1 Table 2 RBM System Setp6ints Setpoint Trip Setpoint Allowable Value Lower Power Setpoint (LPSP ) 27.7 < 29.0 Intermediate Power Setpoint (IPSP b) 62.7 _64.0 High Power Setpoint (HPSP ) 82.7 < 84.0 Low Trip Setpoint (LTSPc) <114.1 < 114.6 0 <108.3 _ 108.8 Intermediate Trip Setpoint (ITSP )

High Trip Setpoint (HTSPc) < 104.5 < 105.0 td2 < 2.0 seconds <2.0 seconds a RBM Operability requirements are not applicable:

(1) if MCPR Ž 1.70; or (2) if MCPR _Ž1.40 and thermal power Ž 90% Rated Thermal Power.

b Setpoints in percent of Rated Thermal Power.

C Setpoints relative to a full scale reading of 125.

For example,

114.1 means :5:114.1/125.0 of full scale.

This Table is referred to by Technical Specification 3.3,2.1 (Table 3.3.2.1-1).

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 B1C14 Core Operating Limits Report Page 12, Revision 1 Table 3 PBDA Setpoints OPRM Setpoint OLMCPR(SS) OLMCPR(2PT) 1.05 1.207 1.127 1.06 1.226 1.144 1.07 1.244 1.162 1.08 1.264 1.180 1.09 1.284 1.199 1.10 1.304 1.218 1.11 1.325 1.237 1.12 1.345 1.256 1.13 1.367 1.276 1.14 1.389 1.297 1.15 1.412 1.319 Acceptance Criteria Off-rated OLMCPR @ Rated Power 45% Flow OLMCPR PDBA Setpoint Setpoint Value Amplitude Sp 1.15 Confirmation Count Np 16 This Table is referred to by Technical Specification 3.3.1.1 (Table 3.3.1.1-1).

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 Page 13, Revision 1 B1C14 Core Operating Limits Report Figure 1 Fuel Type GEl 3-P9DTB403-5G6.0/7G5.0-1 OOT-1 46-T (GEl 3)

Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 13.0 12.0 11.0 n

10.0 CD

-1

_1 9.0 8.0 7.0 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWd/MT)

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1 B21-0604 B1C14 Core Operating Limits Report Page 14, Revision 1 Figure 2 Fuel Type GEl 3-P9DTB403-7G6.0/7G5.0-10OT-146-T (G E13)

Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 13.0 12.0 11.0 10.0 I

8.0 7.0 6.0 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWd/MT) jCP&L

Design Calculation No. 1B21-0604 CP&L Nuclear Fuels Mgmt.Safety Analysis Page 15, Revision 1 B1 C14 Core Operating Limits Report Figure 3 Fuel Type GE13-P9DTB405-5G6.0/7G5.0-1OOT-146-T (GEI3)

Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure

}

14.0 m m - in.m 1 mmml mmml mu I -

13.0 12.0 _ I_

nu M1 -----------------

This Figure is Referred To By II Technical Specification 3.2.1 I

/ Exposure Limit 11.0 (GWd/Mt) (kW/ft) 0.00 0.22 10.71 10.78 10.91

\K 1.10

-* 10.0 2.20 11.08 0 3.31 11.22 4.41 11.35 5.51 11.47

-J 6.61 11.61 0 9. 0 l 7.72 11.75 8.82 11.89 9.92 12.04 11.02 12.18 Permissible 8.0 13.78 12.17 Region of 16.53 12.02 Operation 19.29 11.82 22.05 11.58 27.56 11.17 7.0 33.07 10.60 38.58 10.18 44.09 9.82 49.60 9.49 55.12 9.19 6.0 60.63 8.81 64.74 8.51 5.0 - I I I m4in 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWd/MT)

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CP&L Nuclear Fuels Management Safety Analysis Design CaIc. No. 1B21-0604 Page 16, Revision 1 B1C14 Core Operating Limits Report Figure 4 Fuel Type GE13-P9DTB402-13G6.OI1G2.0-.10OT-146-T (GEl 3)

Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Avernge Planar Exposure 13.0 12.0 11.0 I-10.0 CD

-J 9.0 8.0 7.0 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWd/MT)

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 Page 17, Revision 1 B1C14 Core Operating Limits Report Figure 5 Fuel Type GE14-P1ODNAB416-17GZ-10OT-150-T-2496 (GE1 4)

Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 12.0 11.0 10.0 9.0 I

0, 8.0 I

-J n-7.0 6.0 5.0 4.0 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWd/MT)

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 B1C14 Core Operating Limits Report Page 18, Revision 1 Figure 6 Fuel Type GE1 4-P1 ODNAB425-16GZ-1 OOT-1 50-T-2497 (GEl 4)

Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 12.0 11.0 10.0 9.0 I

8.0 "a-7.0 6.0 5.0 4.0 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWd/MT)

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 B1C14 Core Operating Limits Report Page 19, Revision 1 Figure 7 Fuel Type GE14-P1ODNAB438-12G6.0-10,)T-150-T-2498 (GEl4)

Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 12.0 11.0 10.0 9.0 6

I 8.0 I

-J 0

7.0 6.0 5.0 4.0 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWd/MT)

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No 1B21-0604 B1C14 Core Operating Limits Report Page 20, Revision 1 I Figure 8 is Not Used

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 B1C14 Core Operating Limits Report Page 21, Revision 1 Figure 9 GE13 and GE14 Flow-Dependent MAPLHGR Limit, MAPLHGR(F) 1.10 P . .

I I This Figure is Referred To By 11-11 I I I I Two Loop Operation Limit I 1

1.05 Technical Specifications I 3.2.1 and 3.4.1 /I 1.00 Max Flow =102.5% -- _ _

0.95 107%

112%

U_ 0.90 117% -- 4 U_

a.

0.85 ____ _ - -_Single Loop Operation Limit 0

-W O 0.80

-0.75 S0.70 W 01101 00,MAPLHGR(F) =MAPFACF

MAPLHGRsTD MAPLHGRsTD = Standard MAPLHGR Limits 0.65 MAPFACF(F) = Minimum (1.0, AFWc/1 0 0 +BF)

Wc = % Rated Core Flow

,' 0.60 AF And BF Are Fuel Type Dependent Constants Given Below:

Max Core Flow 0.55 (% Rated) AF BF 102.5 0.6784 0.4861 107.0 0.6758 0.4574 0.50 112.0 0.6807 0.4214 117.0 0.6886 0.3828 0.45 0.40 I I I 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 110 Core Flow (% Rated)

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design CaIc. No. 1B21-0604 B1 C14 Core Operating Limits Report Page 22, Revision 1 Figure 10 GE13 and GE14 Power-Dependent MAPLHGR Limit, MAPLHGR (P) 1.05

ý ýI 1.00 1 I~ I This Figure is Referred To By Technical Specifications I 0.95 - 3.2.1, 3.4.1 and 3.7.6 I

0.90 0

L0 0.85 0.80 7iR Two Loop Operation Fi ur isre.T.

Tfe J S0.75 Sl p tl U

'0001ý Sin'gle Loop Operation Limit U

n 0.70

,-1000 a.

U4 0.65 I I I

"* 0.60 -Core Flow < 50% r MAPLHGR(P) = MAPFACp

MAPLHGRSTD I Turbine Bypass I MAPLHGRSTD = Standard MAPLHGR Limits 0.55 Operable or a-- t Inoperable For P < 23%:

0 No Thermal Limits Monitoring Required 0.50 For 23% < P < 26%:

For Core Flow < 50% &Turbine Bypass Operable or Inoperable r i MAPFACp = 0.567 + 0.0157 (P-26%)

For Core Flow > 50% &Turbine Bypass Operable or Inoperable 0.45 Core Flow > 50% MAPFACP = 0.433 + 0.0063 (P-26%)

00Turbine Bypass For 26% < P < 87.5%

Operable or MAPFACp = 1.0 + 0.005224 (P-100%)

0.40 Inoperable For P > 87.5%

0.35 MAPFACp = 1.0 0.30 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Power (% Rated)

CCP&L

", Jrotess

¢.etyr * .,:-

CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0604 Page 23, Revision 1 B1C14 Core Operating Limits Report Figure 11 GE13 and GE14 Flow-Dependent MCPR Limit, MCPR(F) 1.80 For Wc (% Rated Core Flow) < 40%,

MCPR(F) = (AFWc/100+BF) 1.70 For Wc (% Rated Core Flow) > 40%,

MCPR(F) = Max (1.26, AFWc/100+BF)

Max Core Flow

(% Rated) AF BF 1.60 _____102.5 0.598 1,732 107.0 -0.613 1,776 112.0 0.630 1.829 117.0 0.662 1.894

o. 1.50 0

I I I This Figure is Referred To By Max Flow = 117% Technical Specification 3.2.2, 1.40 1- i - I N

N V_

I 3.4.1 and 3.7.6 112% I 107%

102.5%

1.30 !

1.20 20 30 40 50 60 70 80 90 100 110 120 Core Flow (% Rated)

~2CP&L

CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1B21-0604 B1 C14 Core Operating Limits Report Page 24, Revision 1 Figure 12 GE13 and GE14 Power - Dependent MCPR Limit, MCPR (P) 3.80 3.70 - OLMCPR ._._

3.60 ___ ' Rated MCPR Multiplier (Kp) 3.50 7_ 'rI

__ _ _0_ I F __ __ _

3.40 I

Core Flow > 50% I 3.30 j Turbine Bypass CM Operating Limit MCPR(P) = Kp*Operating Limit MCPR(100) v 3.20 J Operable or 0.

I Inoperable I L . . . , For P < 23%.

vLO3.10 I L No Thermal Limits Monitoring Required 04 3.00 No Limits Specified V 2.90 For 23% < P < PBYPASS" Where PBYPASS z 26%

2.80 __

-j For Core Flow > 50% & Turbine Bypass Operable or Inoperable 2.70 OLMCPR(P): [ 3.15+ 0.0933(26% - P)]

0 2.60 For CoreFlow < 50% & Turbine Bypass Operable or Inoperable 2.50 \ Creo Flow < 50% OLMCPR(P) = [2.36 + 0.0700(26% - P)]

2.40 \ Turbine Bypass

_1. Operable or For 26% < P < 45% :

2.30 Inoperable Kp= 1.28 + 0.0134 (45% - P)

(L 2.20 For 45% < P < 60%:

Al 2.10 Kp = 1.15 + 0.00867 (60% - P)

C 2.00 0 For 60% < P < 87.5%

c4

0. 1.90 Kp = 1.00 + 0.00375 (100% - P) 1.80 For 87.5% <P < 100%

1.70 Kp = 1.00 + 0.0020 (100% - P) a-J 1.60 W

1.50 C.)

1.40 Cu _ _ _-_ .This Figure is Referred To By 1.30 Technical Specification 1.20 _ 3.2.2, 3.4.1, 3.7.6 1.10 i an 1 00 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 PBYPASS Power (% Rated)

SCP&L

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ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 1 DOCKET NO. 50-325/LICENSE NO. DPR-71 UNIT 1 CYCLE 14 CORE OPERATING LIMITS REPORT, SUPPLEMENTAL RELOAD LICENSING REPORT, AND LOSS-OF-COOLANT ACCIDENT ANALYSIS REPORT FOR OPERATION WITH EXTENDED POWER UPRATE "Supplemental Reload Licensing Report for Brunswick Steam Electric Plant Unit 1 Reload 13 Cycle 14,"

J1 1-03936SRLR, Revision 2, Class I, dated March 2002

GNFr Global Nuclear Fuel A Joint Venture of GE, Toshiba, & Hitachi Jll-03936SRLR Revision 2 Class I March 2002 Jll-03936SRLR, Rev. 2 Supplemental Reload Licensing Report for Brunswick Steam Electric Plant Unit 1 Reload 13 Cycle 14 (with Extended Power Uprate)

Approved: Z ... Mvi ..' 4. 11AWApproved: / . /ý G. A. Watford, Manager C. J. Paone Fuel Engineering Services Fuel Project Manager

BRUNSWICK 1 J1 1-03936SRLR Reload 13 Rev. 2 Important Notice Regarding Contents of This Report Please Read Carefully This report was prepared by Global Nuclear Fuel - Americas, LLC (GNF) solely for Carolina Power and Light Company (CP&L) for CP&L's use in defining operating limits for the Brunswick Steam Electric Plant Unit 1. The information contained in this report is believed by GNF to be an accurate and true representation of the facts known or obtained or provided to GNF at the time this report was prepared.

The only undertakings of GNF respecting information in this document are contained in the contract between CP&L and GNF for nuclear fuel and related services for the nuclear system for Brunswick Steam Electric Plant Unit 1 and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither GNF nor any of the contributors to this document makes any representation or warranty (expressed or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.

Page 2

BRUNSWICK 1 Ji 1-03936SRLR Reload 13 Rev. 2 Acknowledgement The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload Licensing Report, were performed by "Fuel Engineering Services" and "Nuclear and Safety Analysis" personnel. The Supplemental Reload Licensing Report Revision 2 was prepared by G. M.

Baka. This document has been verified by R. M. Butrovich.

Page 3

BRUNSWICK 1 J1 1-03936SRLR

~1-4 1 1 Rev. 2

%, M l.*l*LUaldt J, *.1 The basis for this report is General Electric StandardApplicationfor Reactor Fuel, NEDE-240 11-P-A 14, June 2000; and the U.S. Supplement, NEDE-2401 1-P-A-14-US, June 2000.

1. Plant-unique Items Appendix A: Analysis Conditions Appendix B: Decrease in Core Coolant Temperature Events Appendix C: Operating Flexibility Options Appendix D: Implementation of GE 14 Fuel Appendix E: Improved GE13 Thermal/Mechanical Limits Appendix F: Extended Power Uprate

2. Reload Fuel Bundles Cycle Fuel Type Loaded Number Irradiated:

GE 13-P9DTB403-5G6.0/7G5.0-1OOT- 146-T (GE 13) 12 28 GE 13-P9DTB403-7G6.0/7G5.0-1OOT- 146-T (GE 13) 12 64 GEl 3-P9DTB405-5G6.0/7G5.0- 1OOT- 146-T (GE 13) 13 52 GE 13-P9DTB402-13G6.0/1G2.0-IOOT-146-T (GE13) 13 168 New:

GE 14-P 1ODNAB438-12G6.0-10OT- 150-T-2498 (GE 14C) 14 48 GE14-P 1ODNAB425-16GZ- LOOT- 150-T-2497 (GE14C) 14 88 GE14-P 1ODNAB416-17GZ- LOOT- 150-T-2496 (GE 14C) 14 112 Total 560 Page 4

BRUNSWICK 1 JI 1-03936SRLR D I 1'2 Rev. 2 CV

_F*IU*U

3. Reference Core Loading Pattern I Nominal previous cycle core average exposure at end of cycle: 33576 MWd/MT

( 30460 MWd/ST)

Minimum previous cycle core average exposure at end of cycle 33176 MWd/MT from cold shutdown considerations: ( 30097 MWd/ST)

Assumed reload cycle core average exposure at beginning of 14350 MWd/MT cycle: (13018 MWd/ST)

Assumed reload cycle core average exposure at end of cycle 31690 MWd/MT (full power): ( 28748 MWd/ST)

Reference core loading pattern: Figure 1

4. Calculated Core Effective Multiplication and Control System Worth - No Voids, 201C Beginning of Cycle, keffective Uncontrolled 1.120 Fully controlled 0.956 Strongest control rod out 0.988 R, Maximum increase in cold core reactivity with exposure into cycle, Ak 0.000

5. Standby Liquid Control System Shutdown Capability Boron (ppm) Shutdown Margin (Ak)

(at 20'C) (at 160'C, Xenon Free) 660 0.016 1The previous cycle core average exposure at beginning of cycle is 15095 MWd/MT (13694 MWd/ST).

Page 5

BRUNSWICK 1 J1 1-03936SRLR V -1 A 1 '1 Rev. 2

6. Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parameters 2Exposure:

BOC14 to EOFPC14-2026 MWd/MT (1838 MWd/ST) with ICF Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.27 1.37 1.040 6.464 120.4 1.43 GE13 1.45 1.25 1.37 1.020 6.341 110.3 1.38 Exposure: EOFPC14-2026 MWd/MT (1838 MWd/ST) to EEOC14 with ICF Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.29 1.27 1.040 6.544 119.1 1.45 GE13 1.45 1.27 1.27 1.020 6.464 109.0 1.38 Exposure: BOC14 to EEOC14 with ICF and TBPOOS Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.28 1.27 1.040 6.524 119.2 1.46 GE13 1.45 1.26 1.27 1.020 6.406 109.4 1.40 2 End of Full Power Capability (EOFPC) is defined as end-of-cycle all rods out, 100% power/104.5% flow, and normal feedwater temperature.

Page 6

BRUNSWICK 1 JI 1-03936SRLR P~1narI I~ Rev. 2

7. Selected Margin Improvement Options Recirculation pump trip: No Rod withdrawal limiter: No Thermal power monitor: Yes Improved scram time: Yes (ODYN Option B)

Measured scram time: No Exposure dependent limits: Yes Exposure points analyzed: 2 (EOFPC14-2026 MWd/MT and EEOC14)

8. Operating Flexibility Options Single-loop operation: Yes Load line limit: Yes Extended load line limit: Yes Maximum extended load line limit: Yes Increased core flow throughout cycle: Yes Flow point analyzed: 104.5 %

Increased core flow at EOC: Yes Feedwater temperature reduction throughout cycle: Yes Temperature reduction: 110.3 0F Final feedwater temperature reduction: Yes ARTS Program: Yes Maximum extended operating domain: Yes Moisture separator reheater OOS: No Turbine bypass system OOS: Yes (credit taken for 3 of 4 valves)

Safety/relief valves OOS: Yes (credit taken for 10 of 11 valves)

ADS OOS: Yes (1 valve OOS)

Page 7

J1 1-03936SRLR BRUNSWICK 1 D 1 A 1,1"2 Rev. 2 EOC RPT OOS: No Main steam isolation valves OOS: Yes

9. Core-wide AOO Analysis Results Methods used: GEMINI; GEXL-PLUS Exposure range: BOC14 to EOFPC14-2026 MWd/MT (1838 MWd/ST) with ICF Uncorrected ACPR Event Flux QIA GE14C GE13 Fig.

(%NBR) (%NBR)

Load Reject w/o Bypass 552 127 0.31 0.26 2 Exposure range: EOFPC14-2026 MWd/MT (1838 MWd/S1) to EEOC14 with ICF Uncorrected ACPR Event Flux QIA GE14C GE13 Fig.

(%NBR) (%NBR)

Load Reject w/o Bypass 514 124 0.33 0.27 3 Exposure range: BOC14 to EEOC14 with ICF and TBPOOS Uncorrected ACPR Event Flux Q/A GE14C GE13 Fig.

(%NBR) (%NBR)

FW Controller Failure 551 127 0.34 0.28 4

10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary The rod withdrawal error (RWE) event in the maximum extended operating domain was originally analyzed in the GE BWR Licensing Report, Maximum Extended Operating Domain Analysis for Brunswick Steam Electric Plant, NEDC-31654P, February 1989. The Cycle 14 analysis resulted in a RWE ACPR of 0.11 (which is bounded by the generic ARTS ACPR of 0.14) at a rod block monitor setpoint of 108%. The MCPR for rod withdrawal error is bounded by the safety limit adjusted operating limit MCPRs in Table 10-5(a) or 10-5(b) of NEDC-31654P. In addition, the RBM System setpoints shown in Table 10-5(c) of NEDC-31654P are supported for Brunswick Unit 1 Cycle 14. The RBM operability requirements specified in Section 10.5 of NEDC-31654P have been evaluated and shown to be sufficient to ensure that the Safety Limit MCPR and cladding 1% plastic strain criteria will not be exceeded in the event of an unblocked RWE event.

Page 8

BRUNSWICK 1 J1 1-03936SRLR Pph-~nd 11 Rev. 2

11. Cycle MCPR Values 3 Safety limit: 1.12 Single loop operation safety limit: 1.14 Non-pressurization events:

Exposure range: BOC14 to EOC14 All Fuel Types Control Rod Withdrawal Error (RBM setpoint at 108%) 1.26 Loss of Feedwater Heating 4 1.26 Not limiting 5 Fuel Loading Error (mislocated)

GE14C GE13 Fuel Loading Error (misoriented) 1.24 1.25 Pressurization events:

ICF 6 Exposure range: BOC14 to EOFPC14-2026 MWd/MT (1838 MWd/ST) with Exposure point: EOFPC14-2026 MWd/MT (1838 MWd/ST)

Option A Option B GE14C GE13 GE14C GE13 Load Reject w/o Bypass 1.56 1.43 1.45 1.38 Exposure range: EOFPC14-2026 MWd/MT (1838 MWd/ST) to EEOC14 with ICF 7 Exposure point: EEOC14 Option A Option B GE14C GE13 GE14C GE13 Load Reject w/o Bypass 1.68 1.49 1.51 1.41

' The Operating Limit MCPRs for two loop operation (TLO) bound the Operating Limit MCPRs for Single Loop Operation (SLO); therefore, the Operating Limit MCPRs need not be changed for SLO.

4 See Appendix B.

5The mislocated bundle fuel loading error OLMCPR is bounded by the pressurization event OLMCPR.

6 The ICF Operating Limits for the exposure range of BOC14 to EOFPC14-2026 MWd/MT (1838 MWd/ST) bound the Operating Limits for the following domains: MELLL, ICF and FWTR, MSIVOOS and ICF.

7 The ICF Operating Limits for the exposure range of EOFPC14-2026 MWd/MT (1838 MWd /ST) to EEOC14 bound the Operating Limits for the following domains: MELLL, ICF and FWTR, MSIVOOS and ICF.

Page 9

BRUNSWICK 1 J1 1-03936SRLR Reload 13 Rev. 2 Exposure range: BOC14 to EEOC14 with ICF and TBPOOS Exposure point: EEOC14 Option A Option B GE14C GE13 GE14C GE13 FW Controller Failure 1.69 1.50 1.52 1.42

12. Overpressurization Analysis Summary Psi Pv Plant Event (psig) (psig) Response MSIV Closure (Flux Scram) 1310 1343 Figure 5

13. Loading Error Results Variable water gap misoriented bundle analysis: Yes 9 Misoriented Fuel Bundle ACPR GEl 3-P9DTB405-5G6.0/7G5.0-1 OOT- 146-T (GEl 3) 0.08 GE 13-P9DTB402-13G6.0/1 G2.0-1 0OT- 146-T (GEl 3) 0.13 GE14-P 1ODNAB416-17GZ- 1OOT- 150-T-2496 (GE 14C) 0.06 GE 14-PIODNAB425-16GZ- 10OT-150-T-2497 (GE14C) 0.12 GE 14-P 1ODNAB438-12G6.0-10OT-150-T-2498 (GE14C) 0.04

14. Control Rod Drop Analysis Results This is a banked position withdrawal sequence plant, therefore, the control rod drop accident analysis is not required. NRC approval is documented in NEDE-2401 1-P-A-US.

15. Stability Analysis Results Due to the recent Potential Reportable Condition (PRC 01-07) reported by GE on the DIVOM (Delta CPR Over Initial CPR Versus Oscillation Magnitude) slope, it is essential to confirm that the following Option III stability analysis results be applicable to Brunswick Unit 1 Cycle 14 or an interim OPRM system setpoint be used based on a validated new DIVOM slope.

8 The TBPOOS ICF Operating Limits for the exposure range of BOC14 to EEOC14 bound the Operating Limits for all domains with TBPOOS.

9 Includes a 0.02 penalty due to variable water gap R-factor uncertainty.

Page 10

BRUNSWICK 1 J11-03936SRLR P. lnn-l 1V Rev. 2 Should the Option III OPRM system be declared inoperable, the BWROG Interim Corrective Action will constitute the stability licensing basis for Brunswick Unit I Cycle 14.

Stability Option IH Brunswick Unit 1 has implemented BWROG Long Term Stability Solution Option HI (Oscillation Power Range Monitor-OPRM) as described in NEDO-31960-A, "BWR Owners Group Long-Term Stability Solutions Licensing Methodology", November 1995. Plant specific analysis incorporating the Option III hardware is described in GE-NE-C51-00251-00-01, Revision 0, "Licensing Basis Hot Bundle Oscillation Magnitude for Brunswick 1 and 2", March 2001.

Reload validation has been performed in accordance with NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Application", August 1996. The stability based MCPR Operating Limit is provided for two conditions as a function of OPRM amplitude setpoint in the following table. The two conditions evaluated are for a postulated oscillation at 45 %

core flow steady state operation (SS) and following a two recirculation pump trip (2PT) from the limiting full power operation state point. Current power and flow dependent limits provide adequate protection against violation of the Safety Limit MCPR for postulated reactor instability as long as the operating limit is greater than or equal to the specified value for the selected OPRM setpoint.

The stability-based OLMCPR was calculated for Cycle 14. The reload validation calculation demonstrated that reactor stability does not produce the limiting OLMCPR for Cycle 14 as long as the selected OPRM setpoint produces values for OLMCPR(SS) and OLMCPR(2PT) which are less than the corresponding acceptance criteria.

OPRM Setpoint OLMCPR(SS) OLMCPR(2PT) 1.05 1.207 1.127 1.06 1.226 1.144 1.07 1.244 1.162 1.08 1.264 1.180 1.09 1.284 1.199 1.10 1.304 1.218 1.11 1.325 1.237 1.12 1.345 1.256 1.13 1.367 1.276 1.14 1.389 1.297 1.15 1.412 1.319 Acceptance Off-rated OLMCPR Rated Power Criteria @ 45% Flow OLMCPR as described in SRLR Section 11 Interim Corrective Action Stability GE SIL-380 recommendations and BWROG Interim Corrective Actions (BWROG-94079) have been Page 11

BRUNSWICK I J1 1-03936SRLR Reload 13 Rev. 2 included in the Brunswick Unit 1 operating procedures. Regions of restricted operation defined in to NRC Bulletin No 88-07, Supplement 1, Power Oscillations in Boiling Water Reactors (BWRs) and expanded in BWROG-94079, are applicable to Brunswick Unit 1.

16. Loss-of-Coolant Accident Results LOCA method used: SAFER/GESTR-LOCA The SAFER/GESTR-LOCA analysis was performed for the GE14 limiting fuel type at Extended Power Uprate (EPU) conditions. The SAFER/GESTR-LOCA analysis results are presented in the Project Task Report GE-NE-A22-00113-27-01, "Brunswick Nuclear Plant Unit 1 and 2 Extended Power Uprate Task T0407: ECCS-LOCA SAFER/GESTR," Revision 0, June 2001. This analysis yielded a Licensing Basis Peak Cladding Temperature (PCT) for GE14 fuel of < 1557'F, a peak local oxidation fraction of <1% and a core-wide metal-water reaction of <0.1%. Also included in this analysis was validation of the single loop operation (SLO) MAPLHGR multiplier of 0.8.

The EPU analysis results for the limiting GE14 fuel demonstrated a -3OF impact on the Licensing Basis PCT. These results are also conservatively applicable to GEl3 fuel. These results can be applied to the GEl3 ECCS-LOCA analysis results for the current rated core power of 2558 MWt presented in NEDC 31624P, "Brunswick Units 1 and 2 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis Application to GE13 Fuel," Supplement 3, Revision 1, November 2000. This analysis yields a Licensing Basis PCT of <1707'F, a peak local oxidation fraction of <1% and a core-wide metal-water reaction of

<0.1%. The SLO MAPLHGR multiplier of 0.8 is still applicable.

The ECCS-LOCA analysis for GEl3 fuel has been reviewed in light of the proposed improved LHGR limits for GE13 fuel. From this review it was determined that the limiting ECCS results were unaffected.

Thus the referenced LOCA results are still applicable with the improved GEl 3 LHGR limits.

A review of the Brunswick Unit 1 ECCS-LOCA analyses identified errors that have not been accounted for in the reference analyses for GE13 and GEl4 fuels. The impact of the applicable error for GEl3 fuel is as follows:

10CFR50.46 Applicable Error to Brunswick Unit 1 SAFER/GESTR Reference Analysis 10CFR50.46 GEl3 Notificatio1OCFR50.46 Error Description Error Error Notifications 2001-02 Inconsistency in accounting for ECCS +100F pressure rate in OPL-4 ECCS analysis Total Licensing Basis PCT Adder +10°F As all currently known ECCS-LOCA 10CFR50.46 reportable errors have been addressed in the GE14 analysis, no additional adjustment to the Licensing Basis PCT for GE14 is required.

The most limiting and the least limiting MAPLHGRs for the GEl3 fuel based on the improved LHGR limits and for the GEl4 fuel bundles to be added in this cycle are presented in the following tables.

Page 12

BRUNSWICK 1 J1 1-03936SRLR RPlnntl 1*1 Rev. 2