Text

Request for License Amendments Regarding Linear Heat Generation Rate and Core Operating Limits Report References for Areva Np Fuel
	ML070300570
Person / Time
Site:	Brunswick
Issue date:	01/22/2007
From:	Scarola J; Progress Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BSEP 06-0129, TSC-2006-06
	Download: ML070300570 (119)
	v • d • e

SProgress Energy James Scarola Vc rsdn Brunswick Nuclear Plant Progress Energy Carolinas, Inc.

JAN 2 22007 SERIAL: BSEP 06-0 129 10 CER 50.90 TSC-2006-06 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324/License Nos. DPR-71 and DPR-62 Request for License Amendments Regarding Linear Heat Generation Rate and Core Operating Limits Report References for ARE VA NP Fuel Ladies and Gentlemen:

In accordance with the Code of Federal Regulations, Title 10, Part 50.90, Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc.,

is requesting a revision to the Technical Specifications (TSs) for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The proposed license amendments revise:

(1) Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," to add references to ARE VA NP analytical methods that will be used to determine core operating limits, and (2) add a new Technical Specification 3.2.3, "Linear Heat Generation Rate (LHGR)," add a new definition to Technical Specification 1.1 for LHGR, and revise Technical Specifications 3.4.1 and 3.7.6 to incorporate restrictions on LHGR when in single recirculation loop operation or with an inoperable Turbine Bypass System. An evaluation of the proposed license amendments is provided in Enclosure 1.

CP&L has evaluated the proposed change in accordance with 10 CER 50.9 1(a)(1), using the criteria in 10 CER 50.92(c), and determined that this change involves no significant hazards considerations.

CP&L is providing, in accordance with 10 CER 50.9 1(b), a copy of the proposed license amendment to the designated representative for the State of North Carolina.

CP&L requests approval of the proposed amendments by January 30, 2008. Once approved, the Unit 1 amendment shall be implemented prior to start-up from the 2008 Unit 1 refueling outage and the Unit 2 amendment shall be implemented prior to start-up from the 2009 Unit 2 refueling outage.

There are no regulatory commitments associated with this submittal. Please refer any questions regarding this submittal to Mr. Randy C. Ivey, Manager - Support Services, at (910) 457-2447.

P.O.Boo10429 Southport, NC28461Cc T> 910.457.3698 F> 910.457.2803

Document Control Desk BSEP 06-0 129 / Page 2 I declare, under penalty of perjury, that the foregoing is true and correct. Executed on January 22, 2007.

Sincerely, 7James Scarola WRM/wrm

Enclosures:

1. Evaluation of Proposed License Amendment Requests

2. ARE VA NP BWR Approved Topical Reports for Brunswick Units 1 and 2 Technfical Specification COLR References

3. GNIF BWR Approved Topical Reports for Brunswick Units 1 and 2 Technical Specification COLRZ References

4. Marked-up Technical Specification Pages - Unit 1

5. Typed Technical Specification Pages - Unit 1

6. Typed Technical Specification Pages - Unit 2

7. Marked-up Technical Specification Bases Pages - Unit 1 (For Information Only)

Docume--it Control Desk BSEP 06-0129 / Page 3 cc (with enclosures):

U. S. Nuclear Regulatory Commission, Region II ATTN: Dr. William D. Travers, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Eugene M. DiPaolo, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)

ATTN: Mr. Stewart N. Bailey (Mail Stop OWEN 8B1) 11555 Rockville Pike Rockville, MD 20852-2738 Ms. Jo A. Sanford Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-05 10 Ms. Beverly 0. Hall, Section Chief Radiation Protection Section, Division of Environmental Health North Carolina Department of Environment and Natural Resources 3825 Bar-rett Drive Raleigh, NC 27609-7221

BSEP 06-0 129 Enclosure I Page 1 of 12 Evaluation of Proposed License Amendment Requests

Subject:

Request for License Amendments Regarding Linear Heat Generation Rate Requirements and Core Operating Limits Report References for ARE VA NP Fuel 1.0 Description This letter is a request by Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., to amend the Technical Specifications of Renewed Operating Licenses DPR-71 and DPR-62 for the Brunswick Steam Electric Plant (BSEP),

Unit Nos. 1 and 2.

The proposed license amendments will revise: (1) Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," to add references to ARE VA NP analytical methods that will be used to determine core operating limits, and (2) add a new Technical Specification 3.2.3, "Linear Heat Generation Rate (LHGR)," add a new definition to Technical Specification 1.1 'for LHGR, and revise Technical Specifications 3.4.1 and 3.7.6 to incorporate restrictions on LHGR when in single recirculation loop operation or with an inoperable Turbine Bypass System.

These proposed Technical Specification changes are needed to support the transition to ARE VA NP fuel, and to ARE VA NP core design and analysis services. CP&L is planning to begin using AREVA NP fuel starting with the BSEP, Unit 1 refueling outage scheduled for March 2008 and the BSEP, Unit 2 refueling outage scheduled for February 2009.

2.0 Proposed Changes Core Operating Limits Report Methodologies Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," item b, is being revised as shown below. The specific changes to Technical Specification 5.6.5, item b, are shown in bold typeface.

b. 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:

1. NEDE-2401 1-P-A, General Electric Standard Application for Reactor Fuel.

2. XN-NF-81 -58(P)(A), RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model.

BSEP 06-0129 Enclosure 1 Page 2 of 12

3. XN-NF-85-67(P)(A), Generic Mechanical Design for Exxon Nuclear Jet Pump BWR. Reload Fuel.

4. EMF-85-74(P) Supplement 1(P)(A) and Supplement 2(P)(A),

RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model.

5. ANF-89-98(P)(A), Generic Mechanical Design Criteria for BWR Fuel Designs.

6. XN-NF-80-19(P)(A) Volume 1, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis.

7. XN-NF-80-19(P)(A) Volume 4, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads.

8. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2.

9. XN-NF-80-19(P)(A) Volume 3, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description.

10. XN-NF-84-105(P)(A) Volume 1, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis.

11. ANF-524(P)(A), ANF Critical Power Methodology for Boiling Water Reactors.

12. ANF-913(P)(A) Volume 1, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses.

13. ANF-1358(P)(A), The Loss of Feedwater Heating Transient in Boiling Water Reactors

14. EM[F-2209(P)(A), SPCB Critical Power Correlation.

15. EMF-2245(P)(A), Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel.

16. EMF-2361(P)(A), EXEM BWR-2000 ECCS Evaluation Model.

BSEP 06-0129 Enclosure 1 Page 3 of 12

17. EMF-2292(P)(A), ATRIUMTM-10: Appendix K Spray Heat Transfer Coefficients.

18. EMiF-CC-074(P)(A) Volume 4, BWR Stability Analysis:

Assessment of STAIF with Input from MICROBURN-B2.

19. NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications.

Linear Heat Generation Rate First, a new definition for LHGR is being added to Technical Specification 1.1, "Definitions."

LINEAR HEAT The LHGR shall be the heat generation rate per GENERATION RATE (LHGR) unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

Second, a new Technical Specification 3.2.3, "Linear Heat Generation Rate (LHGR)," is being added, as shown below.

3.2.3 LINEAR BEAT GENERATION RATE (LHGR)

LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY: THERMAL POWER Ž:23% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.1 Restore LHGR(s) to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months limits, within limits.

B. Required Action and B. 1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion POWER to Time not met. < 23% RTP.

BSEP 06-0 129 Enclosure 1 Page 4 of 12 SURVEILLANCE REQUIREMENTS__________

SURVEILLANCE FREQUENCY SR 3.2. 1.1 Verify all LHGRs are less than or Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after equal to the limits specified in the >Ž 23% RTP COLR.

AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter Third, Technical Specification 3.4.1, "Recirculation Loops Operating," is being revised as shown below. The specific changes to Technical Specification 3.4.1 are shown in bold typeface.

3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched recirculation pump speeds shall be in operation, OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR;

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR),"

single loop operation limits specified in the COLR; and

c. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," single loop operation limits specified in the COLR; and

d. LCO 3.3. 1.1, "Reactor Protection System (RPS) Instrumentation,"

Function 2.b (Average Power Range Monitors Simulated Thermal Power-High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation.

BSEP 06-0129 Enclosure 1 Page 5 of 12 Fourth, Technical Specification 3.7.6, "The Main Turbine Bypass System," is being revised as shown below. The specific changes to Technical Specification 3.7.6 are shown in bold typeface.

3.7.6 The Main Turbine Bypass System LCO 3.7.6 The Main Turbine Bypass System shall be OPERABLE.

OR The following limits are made applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR; and

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR),"

limits for an inoperable Main Turbine Bypass System, as specified in the COLR; and

c. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR.

Fifth, Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," item a, is being revised as shown below. The specific changes to Technical Specification 5.6.5, item a, are shown in bold typeface.

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1;

2. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.2;

3. The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.3;

BSEP 06-0 129 Enclosure 1 Page 6 of 12

4. The period based detection algorithm (PBDA) setpoint for Function 2.f, Oscillation Power Range Monitor (OPRM)

Upscale, for Specification 3.3. 1.1; and

5. The Allowable Values and power range setpoints for Rod Block Monitor Upscale Functions for Specification 3.3.2. 1. contains a marked-up version of the Unit 1 Technical Specifications showing the proposed changes. Since the affected Technical Specification Sections (i.e., 1.1, 3.2.3, 3.4. 1, 3.7.6, and 5.6.5 for Unit 1 and Unit 2 are identical, only the mark-up for Unit I is provided.

Enclosures 5 and 6 provide typed versions of the affected Unit 1 and Unit 2 Technical Specification pages, respectively. These typed Technical Specification pages are to be used for issuance of the proposed amendments.

CP&L will make supporting changes to the Technical Specification Bases in accordance with Technical Specification 5.5.10, "Technical Specifications (TS) Bases Control Program." provides a markup of the affected Technical Specification Bases pages for Unit 1. These pages are being submitted for information only and do not require issuance by the NRC.

3.0 Background On March 10, 2006, CP&L held a public meeting with the NRC to discuss plans to change the Brunswick Unit 1 and 2 reload fuel supplier from Global Nuclear Fuels - Americas (GNF-A) to AREVA NP. A summary of this meeting was issued by the NRC on April 27, 2006 (i.e., Reference 1). As presented during the public meeting, CP&L plans included submittal of at least two separate license amendment applications to request the Technical Specification changes needed to support loading and use of the ATRllJMTM-10 fuel assemblies.

This letter provides the proposed Technical Specification changes necessary to incorporate the AREVA NP analytical methods, previously approved by the NRC, that will be used to establish core operating limits. Under a separate submittal (i.e., Reference 2), CP&L has submitted proposed Technical Specification changes needed to support the receipt and storage of new AREVA NP fuel bundles. If changes to the Minimum Critical Power Ratio (MCPR) Safety Limit (SL) values are required to support operation with the ATRIUMTM_ 10 fuel, CP&L anticipates submittal of those MCPR SL changes in July 2007.

CP&L plans to begin using ARE VA NP fuel beginning with the March 2008 Unit I refueling outage; therefore, CP&L requests that these Technical Specification changes be issued by January 30, 2008. Once approved, the Unit 1 amendment shall be implemented prior to start-up from the 2008 Unit 1 refueling outage and the Unit 2 amendment shall be implemented prior to start-up from the 2009 Unit 2 refueling outage.

BSEP 06-0129 Enclosure 1 Page 7 of 12 4.0 Technical Analysis Core Operating Limits Report Methodologies Currently, Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," item b, identifies the NRC reviewed and approved analytical methods to be used to determine core operating limits. A revision to Technical Specification 5.6.5.b is needed to reference the ARE VA NP analytical methods that will be used in upcoming fuel cycles to determine core operating limits.

Each methodology reference being added is being cited consistent with the format established in TSTF-363, Revise Topical Report References in ITS 5.6.5, COLR, which was approved by the NRC on April 13, 2000. TSTF-363 has subsequently been incorporated into the latest NRC-approved version of the Standard Technical Specifications for BWRI4 plants (i.e., Revision 3.1 of NUREG-1433). NUREG-1433, Revision 3.1, states that: (1) Technical Specification 5.6.5.b should identify the topical report(s) by number and title or identify the staff Safety Evaluation Report for a plant-specific methodology by NRC letter and date, and (2) the COLR will contain the complete identification for each Technical Specification referenced topical report(s) used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).

Brunswick Technical Specification 5.6.5.b contains an existing reference to the GNF-A topical report NEDE-240111-P-A, General Electric StandardApplicationfor Reactor Fuel.

This analytical methodology reference is being retained because some GNEF-A-derived core operating limits will continue to be applicable to the co-resident GNF-A fuel. However, to maintain consistency with the form-at established in Revision 3 of NJIREG-1433, the existing phrase "(latest approved version)" is being removed from the Technical Specification 5.6.5.b reference to NEDE-24011I-P-A.

Brunswick Technical Specification 5.6.5.b also contains an existing reference to the GNF-A topical report NEDE-32906P-A, TRACG Application for Anticipated Abnormal Operational Occurrences (A00) Transient Analyses. The TRACG analytical methodology was added to the Brunswick Technical Specifications through Amendments 231 (i.e., Reference 3) and 262 (i.e., Reference 4) for Units 1 and 2, respectively. At the time the TRACG analytical methodology was added to the Brunswick Technical Specifications, the TRACG method had not yet been incorporated into the GNF-A topical report NEDE-2401 I-P-A. The TRACG analytical methodology has subsequently been added to GNTF-A topical report NEDE-2401 I-P-A; therefore, the separate Technical Specification 5.6.5.b reference to GNIF topical report NEDE-32906P-A is no longer necessary and is being removed.

Finally, a reference to the GNF-A topical report NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications, is being added to Technical Specification 5.6.5.b. Although NEDO-32465-A has been incorporated into the

BSEP 06-0 129 Enclosure 1 Page 8 of 12 compendium of methods contained in GNF-A topical report NIEDE-2401 I1-P-A, the NEDO-32465-A topical report will be used by AREVA NP as part of their detection and suppression analyses. provides a list of the NRC-approved ARE VA NP topical report references being added to Technical Specification 5.6.5.b and the associated Technical Specification Limiting Condition for Operations for the core limits or setpoints for which the methodology is used. provides a list showing the applicability of the retained GNF-A topical report NEDE-24011-P-A reference to Technical Specification 5.6.5.a core operating limits. Details regarding the GNE-A methodologies for calculating core limits are available in NEDE-2401 I1-P-A- 15.

Linear Heat Generation Rate The Brunswick Technical Specifications currently do not include LHGR as a required power distribution limit. The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location and is an important fuel design parameter. Limits are calculated for LHGR to ensure that the cladding circumferential plastic strain will not exceed one percent and that the fuel centerline will not melt. Thus, limits on the LHGR in the COLR are specified to ensure the fuel design limits are not exceeded anywhere in the core during normal operation, including anticipated operational occurrences.

For GNF-A fuel, the average planar linear heat generation rate (APLHGR) limits, in addition to protecting against excess peak clad temperature, include the exposure dependent thermal-mechanical limits for the fuel rods and thereby ensure that the LHGR limits are not exceeded.

During the upcoming BSEP mixed core cycles, the APLHGR limits that were developed by GNE-A for their fuel types will continue to be the applicable limits and will ensure that the LHGR limits for GNF-A fuel are not exceeded.

To support the use of ARE VA NP fuel, a new Technical Specification 3.2.3 is being proposed to add the LHGR limit as a separate limit. For ARE VA NP fuel, both the APLHGR and the LHGR limits will be specified in the COLR.

Currently, Technical Specification 3.4.1, Recirculation Loop Operating, and Technical Specification 3.7.6, Main Turbine Bypass System, contain equipment out-of-service provisions for adjusting APLHGR and MCPR limits to ensure thermal margins are maintained if certain equipment is not in service. The specific adjusted APLHGR and MCPR limits for these conditions are currently documented in the COLR. As part of the proposed amendments, a similar provision is being added to Technical Specification 3.4.1 and Technical Specification 3.7.6 to apply an adjustment to the LHGR limit when a single reactor recirculation loop or the main turbine bypass system is inoperable. The specific adjusted LHGR limits for these conditions will also be documented in the COLR.

BSEP 06-0 129 Enclosure 1 Page 9 of 12 5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., is submitting proposed amendments to support transition to AREVA NP fuel and core design and analysis services for the Brunswick Steam Electric Plant (BSEP),

Units I and 2. The proposed amendments: (1) revise Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," to add references to AREVA NP analytical methods that will be used to determine core operating limits, and (2) add a new Technical Specification 3.2.3, 'Linear Heat Generation Rate (LHGR)," add a new definition to Technical Specification 1.1 for LHGR, and revise Technical Specifications 3.4.1 and 3.7.6 to incorporate restrictions on LHGR when in single recirculation loop operation or with an inoperable Turbine Bypass System. CP&L has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CER 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed amendments revise the list of NRC-approved analytical methods used to establish core operating limits. Core operating limits are established to ensure that fuel design limits are not exceeded during operating transients or accidents. The analytical methods used to determine core operating limits are those methods that have previously been found acceptable by the NRC and are required to be listed in the Technical Specification section governing the Core Operating Limits Report. The application of these NRC-approved analytical methods will continue to ensure that acceptable operating limits are established and applied to operation of the reactor core.

The proposed amendments will add a new Technical Specification 3.2.3, "Linear Heat Generation Rate (LHGR)," for fuel bundles, add a new definition to Technical Specification 1.1 for LHGR, and revise Technical Specifications 3.4.1 and 3.7.6 to incorporate restrictions on LHGR when in single recirculation loop operation or with an inoperable Turbine Bypass System. These LHGR limits will be established using NRC-approved analytical methods to ensure that fuel performance during normal, transient, and accident conditions is acceptable.

Based on the above, the proposed amendments do not involve an increase in the probability or consequences of an accident previously evaluated.

BSEP 06-0129 Enclosure 1 Page 10 of 12

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No As previously stated, the proposed amendments support transition from Global Nuclear Fuels - Americas (GNF-A) fuel and core design and analysis services to ARE VA NP fuel and core design and analysis services. The AREVA NP fuel assemblies which will be used in the BSEP Unit 1 and 2 cores will be similar in design to the GNF-A fuel that will be co-resident in the cores. The BSEP, Unit 1 and 2 cores in which this fuel will operate will be designed to meet all applicable design and licensing criteria.

Adherence to these design and licensing criteria will not introduce any new modes of operation or introduce any new accident precursors, and thus will preclude the introduction of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The proposed amendments will continue to require that core operating limits be determined using NRC-approved analytical methods. Acceptable fuel performance is obtained by ensuring that the peak cladding temperature (PCT) during a postulated design basis loss-of-coolant accident (LOCA) is maintained less than the limits specified in 10 CER 50.46, and that the core remains in a coolable geometry following a postulated design basis LOCA. The proposed amendments ensure that adequate margin will continue to be maintained to the 2200 degree PCT limit of 10 CFR 50.46, and the use of NRC-approved analytical methods will continue to ensure acceptable fuel performance during normal operations, as well as during transient and accident conditions. Therefore, the proposed amendments do not involve a significant reduction in a margin of safety.

Based on the above, CP&L concludes that the proposed amendments present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Reg uirements/Criteri a 10 CER 50.36, "Technical Specifications," provides the regulatory requirements for the content required in a licensee's Technical Specifications. Criterion 2 of 10 CFR 50.36(c)(2)(ii) requires a limiting condition for operation to be established for a process variable, design feature, or operating restriction that is an initial condition of a design

BSEP 06-0 129 Enclosure 1 Page 11 of 12 basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water power reactors," establishes the acceptance criteria for the design basis LOCA.

Paragraph (b)(1) requires the calculated maximum fuel element cladding temperature (i.e.,

PCT) to not exceed 22000 FE The use of NRC-approved analytical methods to determine core operating limits will continue to ensure that fuel performance during normal, transient, and accident conditions complies with these requirements. Specific APLHGR limits will be determined and documented in the COLR to ensure compliance with 10 CER 50.46(b)(1).

LHGR is being added to the Technical Specifications to support Criterion 2 of 10 CFR 50.36(c)(2)(ii).

6.0 Environmental Considerations A review has determined that the proposed amendments would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendments do not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CER 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendments.

7.0 References

1. "Summary of March 10, 2006, Meeting with Carolina Power & Light Company (CP&L),

Regarding the Fuel Transition Process and Licensing Submittals for Brunswick Steam Electric Plant, Units 1 and 2," dated April 27, 2006, ADAMS Accession Number MLO6 1000574.

2. Letter from Mr. James Scarola (CP&L) to the Document Control Desk (NRC), "Request for License Amendments Regarding Fuel Design and Storage Requirements for AREVA NP Fuel," dated January 22, 2007.

3. Letter from Ms. Brenda L. Mozafari (NRC) to Mr. C. J. Gannon (CP&L), "Issuance of Amendment Re: Minimum Critical Power Ratio Safety Limit (TAC No. MC 1249),"

dated March 26, 2004, ADAMS Accession Number ML040900042.

BSEP 06-0129 Enclosure 1 Page 12 of 12

4. Letter from Ms. Brenda L. Mozafari (NRC) to Mr. C. J. Gannon (CP&L), "Issuance of Amendment on Addition of TRACG Methodology for Determining Core Operating Limits (TAC No. MC4385)," dated March 4, 2005, ADAMS Accession Number ML050470131.

BSEP 06-0 129 Enclosure 2 Page 1 of 4 AREVA NP BWR Approved Topical Reports for Brunswick Units 1 and 2 Technical Specification COLR References Report Applicable LCO Justification XN-NF-81-58(P)(A) Revision 2 and 3.2.1 /ATRllJMTM-10 Provides an analytical capability to predict Supplements 1 and 2, RODEX2 Fuel Rod 3.2.2 / ATRULJMTM- 10, GE 14 BWR fuel thermal and mechanical conditions Thermal-MechanicalResponse Evaluation 3.2.3 / ATRIUMTM- 10 for normal core operation and to establish Model. initial conditions for power ramping, non-LOCA and LOCA analyses.

XN-NF-85-67(P)(A) Revision 1, Generic 3.2.3 / ATRfLJMTM- 10 Describes the process used to develop linear Mechanical Designfor Exxon Nuclear Jet heat generation rates for fuel designs.

Pump BWR Reload Fuel.

EMI-85-74(P) Revision 0 3.2.3 / ATRJUMTM_ 10 Extends the exposure limit of the RODEX2A Supplement 1(P)(A) and code which is a version of RODEX2 that Supplement 2(P)(A), RODEX2A (BWR) includes a fission gas release model specific Fuel Rod Thermal-MechanicalEvaluation to BWR fuel designs.

Model.

AN-F-89-98(P)(A) Revision 1 and 3.2.3 / ATRfUMTM- 10 Establishes a set of design criteria which Supplement 1, Generic Mechanical Design assures that BWR fuel will perform Criteriafor BWR Fuel Designs. ____________________satisfactorily throughout its lifetime.

BSEP 06-0 129 Enclosure 2 Page 2 of 4 Report Applicable LCO Justification X.N-NF-80-19(P)(A) Volume 1 and 3.2.1 / ATRITJMTM-10 Development of BWR core analysis Supplements 1 and 2, Exxon Nuclear 3.2.2 / ATRTUMTM-10, GE14 methodology which comprises codes for fuel Methodology for Boiling Water Reactors - 3.2.3 / ATRIUMTM-10 neutronic parameters and assembly bumnup Neutronic Methods for Design and Analysis. 3.3.2. 1, Table 3.3.2. 1-1 /calculations, reactor core simulation diffusion ATRI1JlVTM-10, GE14 theory calculations, core and channel hydrodynamic stability predictions, and producing input for nuclear plant transients.

Subsequently approved codes or methodologies have superseded portions of this report. Applicable portions include control rod drop accident, and methodology to determine neutronic reactivity parameters, void reactivity, Doppler reactivity, scram reactivity, delayed neutron fraction, and prompt neutron lifetime.

XN-NF-80-19(P)(A) Volume 4 Revision 1, 3.2.1 /ATRIT1MTM-10 Summarizes the types of BWR licensing Exxon Nuclear Methodology for Boiling 3.2.2 / ATR~iJvLTN4M-10, GE14 analyses performed, identifies the Water Reactors: Application of the ENC 3.2.3 / ATRIUMTM-10 methodologies used.

Methodology to BWR Reloads.

EMF-2158(P)(A) Revision 0, Siemens 3.2.2 I ATRRTN/TM-10, GE14 Describes the reactor core simulator code Power CorporationMethodology for 3.2.3 I ATRHUMTM-10 MICROBURN-B2 and the lattice physics Boiling Water Reactors: Evaluation and 3.3.2. 1, Table 3.3.2. 1-1 Icode CASMO-4.

LValidation of CASMO-4/MICROBURN-B2. ATRJUMTM-10, GEJ4

BSEP 06-0 129 Enclosure 2 Page 3 of 4 Report Applicable LCO Justification XN-NE-80-19(P)(A) Volume 3 Revision 2, 3.2.2 / ATRIUMTM-10, GE14 Provides overall methodology for determining Exxon Nuclear Methodology for Boiling a MCPR operating limit.

Water Reactors, THERMEX. Thermnal Limits Methodology Summary Description. ___________ ________

XN-NE-84-105(P)(A) Volume 1 and 3.2.2 / ATRIUMTM-10, GE14 Provides a capability to perform analyses of Volume 1 Supplements 1 and 2, transient heat transfer behavior in BWR XCOBRA-T: A Computer Code for BWR assemblies.

Transient Thennal-Hydraulic Core Analysis._____________________

ANF-524(P)(A) Revision 2 and 3.2.2 / ATR~IUMTM-10, GE14 Provides a methodology for the determination Supplements 1 and 2, ANF Critical Power of thermal margins, specifically the MCPR Methodologyfor Boiling Water Reactors. safety limit.

ANEF-913(P)(A) Volume I Revision 1 and 3.2.2 / ATRRLTMTM-10, GE14 Provides a computer program for analyzing Volume 1, Supplements 2, 3, and 4, BWR system transients.

COTRANSA2: A Computer Programfor Boiling Water Reactor TransientAnalyses. ____________________

ANF-1358(P)(A) Revision 3, The Loss of 3.2.2 / ATRTUMTM-10, GE14 Presents a generic methodology for evaluating FeedwaterHeating Transient in Boiling the loss of feedwater heating event.

Water Reactors.

EMIF-2209(P)(A) Revision 2, SPCB Critical 3.2.2 / ATRIUMTM-10, GE14 Presents the critical power correlation for use Power Correlation. with the ATRIUTMTM-10 and GE14 fuel

___ __ ___ ____ ___ ___ ___ ____ ___ ___ ___ designs.

BSEP 06-0 129 Enclosure 2 Page 4 of 4 Report Applicable LCO Justification EMIF-2245(P)(A) Revision 0, Application of 3.2.2 / GE14 Provides direct and indirect approaches to Siemens Power Corporation'sCritical develop parameters necessary to appropriately Power Correlationsto Co-Resident Fuel. model co-resident fuel with an approved critical power correlation.

EME-2361(P)(A) Revision 0, EXEM 3.2.1 / ATRIUMTM-10 Describes an evaluation model methodology BWR-2000 ECCS Evaluation Model. for licensing analyses of postulated LOCAs in jet pump BWRs. The methodology was developed to comply with 10 CER 50.46 and 10 CFR 50 Appendix K criteria.

EMF-2292(P)(A) Revision 0, 3.2.1 / ATRIUMTM-10 Provides measured cladding temperatures ATRIUMiTm-1O: Appendix K Spray Heat from spray heat transfer tests to justify the use Transfer Coefficients. of Appendix K coefficients for ATRIUTMTM-l10fuel LOCA analyses.

EMF-CC-074(P)(A) Volume 4 Revision 0, 3.3. 1.1 / ATRIUMNTM-10, GE14 Describes methodology for stability analysis BWR Stability Analysis - Assessment of with input from the MICROBIURN-B2 reactor STAIF with Input from MICROBURN-B2. core simulator.

NEDO-32465-A, Reactor Stability Detect 3.3. 1.1 / ATRIUMTM-10, GE14 Describes the bases for period-based detection and Suppress Solutions Licensing Basis algorithm setpoint for the OPRMs.

LMethodology for Reload Applications. I_________________ I_____________I_____

BSEP 06-0 129 Enclosure 3 Page 1 of 2 GNE BWR Approved Topical Reports for Brunswick Units 1 and 2 Technical Specification COLR References Applicable LCO /Assembly Types COLR Specified Limit NEDE-24011 -P-A Applicability 3.2.1 / GE14 Average Planar Linear Heat Generation The NEDE-2401 1-P-A APLHGR limits are applied to Rate (APLHGR) off-rated conditions using power- and flow-dependent factors, MAPFACP and MAPFACf.

NIEDE-24011I-P-A APLHGR limits are applied to the following Equipment Out-of-Service (EOOS) conditions:

S LO(TS53.4. 1)

" Turbine Bypass OGS (TS 3.'7.6) 3.2.2 /GE14 Minimum Critical Power Ratio (MCPR) NEDE-24011I-P-A will not be applied to establish MCPR operating limits.

Operating Limit MCPR for rated and off-rated power and flow conditions and EOOS operation will be established for GE 14 fuel using ARE VA-approved methods identified in Enclosure 2.

3.2.3 / GE14 Linear Heat Generation Rate (LHGR) NEDE-240 11-P-A will not be applied to establish off-rated or flow-dependent or HOOS corrections to LHGR limits.

BSEP 06-0 129 Enclosure 3 Page 2 of 2 Applicable LCO / Assembly Types COLR Specified Limit I-NF-DE-24011I-P-A Applicability 3.3.2. 1, Table 3.3.2. 1-1 IGE 14 Rod Block Monitor (RBM) setpoints N-EDE-2401 I-P-A will not be applied to establish RBM and applicable reactor thermal power setpoints and applicable reactor thermal power ranges.

ranges The RBM setpoints and applicable thermal power ranges will be established using the ARE VA approved methods identified in Enclosure 2.

3.3. 1.1, Table 3.3. 1.1-1 IGE 14 Oscillation Power Range Monitor NTEDE-24011-P-A will not be applied to establish the (OPRM) period-based detection OPRMI setpoints. The OPRM setpoints will be algorithm (PBDA) setpoint limits established using ARE VA NP approved methods identified in Enclosure 2.

BSEP 06-0 129 Enclosure 4 Markup of Technical Specification Pages - Unit I

Definitions 1.1 1.1 Definitions LEAKAGE b. Unidentified LEAKAGE (continued)

All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or L vessel wall.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of TEST all required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power ratio (CPR)

RATIO (MCPR) that exists in the core. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

(continued)

Brunswick Unit 1114 1.1-4 Amendment No. e

INSERT 1.1 DEFINITIONS LINEAR HEAT The LHGR shall be the heat generation rate GENERATION RATE (LHGR) per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LHGR 3.2.3 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY: THERMAL POWER Ž:23% RTP.

ACTIONS__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A. 1 Restore LHGR(s) to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months limits, within limits.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion POWER to < 23% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.2. 1.1 Verify all LHGRs are less than or equal to the limits Once within specified in the COLR. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

Ž!23% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter Brunswick Unit 132-AmnetNo 3.2-4 Amendment No.

Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched recirculation pump speeds shall be in operation, OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR; b.LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," single1 GNERT-ý loop operation limits specified in the COLR; (0 LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation,"

Function 2.b (Average Power Range Monitors Simulated Thermal Power-High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation.

APPLICABILITY: MODES 1 and 2.

ACTIONS_____________

CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A. 1 Satisfy the requirements of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months not met. the LCO.

(continued)

Brunswick Unit 1 3.4-1 BrunwickUni No.& I 1 34-1Amendment

INSERT LCQ 3.4.1 C. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," single ioop operation limits specified in the COLR; and

Main Turbine Bypass System 3.7.6 3.7 PLANT SYSTEMS 3.7.6 The Main Turbine Bypass System LCO 3.7.6 The Main Turbine Bypass System shall be OPERABLE.

OR The following limits are made applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR;GZ an inoperable Main Turbine Bypass System, as specified in the APPLICABILITY: THERMAL POWER Ž!23% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A. 1 Satisfy the requirements of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months not met. the LCO.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion Time POWER to < 23% RTP.

not met.

Brunswick Unit 1 3.7-20 BrunwickUnit1No.g I 3.-20Amendment

INSERT LCO 3.7.6

c. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR.

Reporting Requirements 51.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1;

2. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.2; The period based detection algorithm (PBDA) setpoint for

~y Function 2.f, Oscillation Power Range Monitor (OPRM) Upscale, for Specification 3.3.1.1; and .

The Allowable Values and power range setpoints for Rod Block

~ Monitor Upscale Functions for Specification 3.3.2.1.

b. 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:

1. NEDE-2401 1-PA"General Electric Standard Application for Reactor Fuel" Ia;4stýa oedvsioný)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued)

Brunswick Unit 1 5.0-20 Amendment No.0

INSERT TS 5.6.5.a

3. The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.3; INSERT TS 5.6.5.b

2. XN-NF-81-58(P)(A), RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model.

3. XN-NF-85-67(P)(A), Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel.

4. EMF-85-74(P) Supplement 1(P)(A) and Supplement 2(P)(A),

RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model.

5. ANF-89-98(P)(A), Generic Mechanical Design Criteria for BWR Fuel Designs.

6. XN-NF-80-19(P)(A) Volume 1, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis.

7. XN-NF-80-19(P)(A) Volume 4, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENG Methodology to BWR Reloads.

8. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/M ICROBU RN-B2.

9. XN-NF-80-19(P)(A) Volume 3, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description.

10. XN-NF-84-1 05(P)(A) Volume 1, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis.

11. ANF-524(P)(A), ANF Critical Power Methodology for Boiling Water Reactors.

12. ANF-91 3(P)(A) Volume 1, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses.

INSERT TS 5.6.5.b (Continued)

13. ANF-1358(P)(A), The Loss of Feedwater Heating Transient in Boiling Water Reactors.

14. EMF-2209(P)(A), SPCB Critical Power Correlation.

15. EMF-2245(P)(A), Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel.

16. EMF-2361(P)(A), EXEM BWR-2000 ECCS Evaluation Model.

17. EMF-2292(P)(A), ATRI UM TM-10: Appendix K Spray Heat Transfer Coefficients.

18. EMF-CC-074(P)(A) Volume 4, BWR Stability Analysis:

Assessment of STAIF with Input from MICROBURN-B32.

19. NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications.

BSEP 06-0 129 Enclosure 5 Typed Technical Specification Pages - Unit 1

Definitions 1.1 1.1 Definitions LEAKAGE b. Unidentified LEAKAGE (continued)

All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LINEAR HEAT The LHGR shall be the heat generation rate per unit length of GENERATION RATE (LHGR) fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of TEST all required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power ratio (CPR)

RATIO (MCPR) that exists in the core. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

(continued)

Brunswick Unit 1 1.1-4 Bruswik Uit 1.-4Amendment No. I

LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY: THERMAL POWER Ž!23% RTP.

ACTIONS_______ ___

CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.1 Restore LHGR(s) to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months limits, within limits.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion POWER to < 23% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all LHGRs are less than or equal to the limits Once within specified in the COLR. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

Ž:23% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter Brunswick Unit 1 3.2-4 Bruswik Uit 3.-4Amendment No. I

Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched recirculation pump speeds shall be in operation, OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR;

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR;

c. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," single loop operation limits specified in the COLR; and

d. LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation,"

Function 2.b (Average Power Range Monitors Simulated Thermal Power-High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation.

APPLICABILITY: MODES 1 and 2.

ACTIONS__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A. 1 Satisfy the requirements of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months not met. the LCO.

(continued)

Brunswick Unit 1341AmnetNo 3.4-1 Amendment No. I

Main Turbine Bypass System 3.7.6 3.7 PLANT SYSTEMS 3.7.6 The Main Turbine Bypass System LCO 3.7.6 The Main Turbine Bypass System shall be OPERABLE.

OR The following limits are made applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR; I

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR; and

c. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR.

APPLICABILITY: THERMAL POWER Ž!23% RTP.

ACTIONS__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Satisfy the requirements of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months not met. the LCO.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion Time POWER to < 23% RTP.

not met.

Brunswick Unit 1 3.7-20 Bruswik Uit 3.-20Amendment No. I

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2. 1;

2. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.2;

3. The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.3;

4. The period based detection algorithm (PBDA) setpoint for Function 2.f, Oscillation Power Range Monitor (OPRM) Upscale, for Specification 3.3.1.1; and

5. The Allowable Values and power range setpoints for Rod Block Monitor Upscale Functions for Specification 3.3.2.1.

b. 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:

1. NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel."

2. XN-NF-8 1-58(P)(A), RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model.

3. XN-NF-85-67(P)(A), Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel.

4. EMF-85-74(P) Supplement 1(P)(A) and Supplement 2(P)(A),

RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model.

5. ANF-89-98(P)(A), Generic Mechanical Design Criteria for BWR Fuel Designs.

(continued)

Brunswick Unit 1 5.0-20 Bruswik Uit 5.-20Amendment No. I

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

6. XN-NF-80-1 9(P)(A) Volume 1, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis.

7. XN-NF-80-19(P)(A) Volume 4, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads.

8. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B32.

9. XN-NF-80-19(P)(A) Volume 3, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description.

10. XN-NF-84-105(P)(A) Volume 1, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis.

11. ANF-524(P)(A), ANF Critical Power Methodology for Boiling Water Reactors.

12. ANF-913(P)(A) Volume 1, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses.

13. ANF-1358(P)(A), The Loss of Feedwater Heating Transient in Boiling Water Reactors.

14. EMF-2209(P)(A), SPCB Critical Power Correlation.

15. EMF-2245(P)(A), Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel.

16. EMF-2361(P)(A), EXEM BWR-2000 ECCS Evaluation Model.

17. EMF-2292(P)(A), ATRI UM TM -10: Appendix K Spray Heat Transfer Coefficients.

18. EMF-CC-074(P)(A) Volume 4, BWR Stability Analysis -

Assessment of STAlE with Input from MICROBURN-B32.

19. NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications.

(continued)

Brunswick Unit 1 5.0-21 Bruswik Uit 5.-21Amendment No. I

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SIDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Brunswick Unit 1 5.0-22 Bruswik Uit 5.-22Amendment No. I

High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area As provided in paragraph 20.1601 (c) of 10 CIFR Part 20, the following controls shall be applied to high radiation areas in place of the, controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20:

5.7.1 High Radiation Areas with Dose Rates not exceeding 1.0 rem/hour (at 30 centimeters from the radiation sources or from any surface Penetrated by the radiation)

a. Each accessible entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b. Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall:

1. Possess a radiation monitoring device that continuously displays radiation dose rates in the area ("radiation monitoring and indicating device"); or

2. Possess a radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached ("alarming dosimeter"), with an appropriate alarm setpoint; or

3. Possess a radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area; or (continued)

Brunswick Unit 1 502 5.0-23 mnmn No.

Amendment o I

High Radiation Area 5.7 5.7 High Radiation Area 5.7.1 Hicih Radiation Areas with Dose Rates not exceeding 1.0 rem/hour (at 30 centimeters from the radiation sources or from any surface penetrated by the radiation) (continued)

4. Possess a self-reading dosimeter and be under the surveillance, as specified in the RWP or equivalent, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area; or

5. Be under the surveillance, as specified in the RWP or equivalent, of an individual at the work site, qualified in radiation protection procedures, equipped with a radiation monitoring and indicating device who is responsible for controlling personnel radiation exposure within the area.

e. Except for individuals qualified in radiation protection procedures, entry into such areas shall be made only after dose rates in the area have been established and entry personnel are knowledgeable of them.

5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour (at 30 centimeters from the radiation source or from any surface Penetrated by the radiation), but less than 500 rads/hour (at 1 meter from the radiation source or from any surface Penetrated by the radiation)

a. Each accessible entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked door, gate, or guard that prevents unauthorized entry, and in addition:

1. All such door and gate keys shall be maintained under the administrative control of the shift superintendent or the radiation control supervisor or designated representative; and

2. Doors and gates shall remain locked or guarded except during periods of personnel or equipment entry or exit.

(continued)

Brunswick Unit 1 5.0-24 Bruswik Uit 5.-24Amendment No. I

High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour (at 30 centimeters from the radiation source or from any surface penetrated by the radiation), but less than 500 rads/hour (at 1 meter from the radiation source or from any surface penetrated by the radiation) (continued)

b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall:

1. Possess an alarming dosimeter with an appropriate alarm setpoint; or

2. Possess a radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area; or

3. Possess a direct-reading dosimeter and be under the surveillance, as specified in the RWP or equivalent, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area; or

4. Be under the surveillance, as specified in the RWP or equivalent, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring and indicating device who is responsible for controlling personnel exposure within the area; or (continued)

Brunswick Unit 1 5.0-25 Bruswik Uit 5.-25Amendment No. I

High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour (at 30 centimeters from the radiation source or from any surface Penetrated by the radiation), but less than 500 rads/hour (at 1 meter from the radiation source or from any surface penetrated by the radiation) (continued)

5. Possess a radiation monitoring and indicating device in those cases where the options of Specifications 5.7.2.d.2, 5.7.2.d.3, and 5.7.2.d.4. above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle.

e. Except for individuals qualified in radiation protection procedures, entry into such areas shall be made only after dose rates in the area have been established and entry personnel are knowledgeable of them.

f. Such individual areas that are within a larger area that is controlled as a high radiation area, where no enclosure exists for purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, but shall be barricaded and conspicuously posted as a high radiation area, and a conspicuous, clearly visible flashing light shall be activated at the area as a warning device.

Brunswick Unit 1 502 5.0-26 mnmn No.

Amendment o I

BSEP 06-0 129 Enclosure 6 Typed Technical Specification Pages - Unit 2

Definitions 1.1 1.1 Definitions LEAKAGE b. Unidentified LEAKAGE (continued)

All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LINEAR HEAT The LHGR shall be the heat generation rate per unit length of GENERATION RATE (LHGR) fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power ratio (CPR)

RATIO (MCPR) that exists in the core. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

(continued)

Brunswick Unit 2 1.1-4 Bruswik Uit 1.-4Amendment No. I

LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY: THERMAL POWER Ž!23% RTP.

ACTIONS_______ ___

CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.1 Restore LHGR(s) to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months limits, within limits.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion POWER to < 23% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all LHGRs are less than or equal to the limits Once within specified in the COLR. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

Ž!23% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter Brunswick Unit 2 3.2-4 Uit 3.-4Amendment

~ruswik No. I

Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched recirculation pump speeds shall be in operation, OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR;

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR;

c. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," single loop operation limits specified in the COLR; and

d. LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation,"

Function 2.b (Average Power Range Monitors Simulated Thermal Power-High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Satisfy the requirements of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months not met. the LCO.

(continued)

Brunswick Unit 234-AmnetNo 3.4-1 Amendment No. I

Main Turbine Bypass System 3.7.6 3.7 PLANT SYSTEMS 3.7.6 The Main Turbine Bypass System LCO 3.7.6 The Main Turbine Bypass System shall be OPERABLE.

OR The following limits are made applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR; I

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR; and

c. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR.

APPLICABILITY: THERMAL POWER Ž!23% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Satisfy the requirements of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months not met. the LCO.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion Time POWER to < 23% RTP.

not met.

Brunswick Unit 2 372 3.7-20 mnmn No.