BSEP 10-0052, Request for License Amendments - Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, Core Operating Limits Report

Request for License Amendments - Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, Core Operating Limits Report

ML101310388
Person / Time
Site:	Brunswick
Issue date:	04/29/2010
From:	Annacone M Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ANP-10298PA, BSEP 10-0052, TSC-2010-01
Download: ML101310388 (25)

Text

Progress Energy Michael J. Annacone ViceliheJA cnPresident Brunswick Nuclear Plant APR 2 9 2010 SERIAL: BSEP 10-0052 10 CFR 50.90 TSC-2010-01 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Request for License Amendments - Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)"

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 for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The proposed license amendments revise Technical Specification 5.6.5.b by adding AREVA Topical Report ANP-10298PA, A CE/ATRIUM 1OXM CriticalPower Correlation,Revision 0, March 2010, to the list of analytical methods that have been reviewed and approved by the NRC for determining core operating limits. An evaluation of the proposed license amendments is provided in Enclosure 1.

CP&L has evaluated the proposed change in accordance with 10 CFR 50.91 (a)(1), using the criteria in 10 CFR 50.92(c), and determined that this change involves no significant hazards considerations.

In accordance with 10 CFR 50.91(b), CP&L is providing 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 March 4, 2011, in order to support reactor start-up following the Unit 2 refueling outage, which is currently scheduled to begin on March 5, 2011. Once approved, the Unit 2 amendment shall be implemented prior to start-up from the 2011 Unit 2 refueling outage and the Unit 1 amendment shall be implemented prior to start-up from the 2012 Unit 1 refueling outage.

No regulatory commitments are contained in this submittal. Please refer any questions regardingthis submittal to Ms. Annette Pope, Supervisor - Licensing/Regulatory Programs, at (910) 457-2184.

Progress Energy Carolinas, Inc.

P.. Box 10429 A* 01 Southport, NC 28461 T> 910.457.3698

Document Control Desk BSEP 10-0052 / Page 2 I declare, under penalty of perjury, that the foregoing is true and correct. Executed on April 29, 2010.

Sincerely, J. Annacone WRM/wrm

Enclosures:

1. Evaluation of License Amendment Request

2. Marked-up Technical Specification Pages - Unit 1

3. Typed Technical Specification Pages - Unit 1

4. Typed Technical Specification Pages - Unit 2

5. Marked-up Technical Specification Bases Pages - Unit 1 (For information only)

Document Control Desk BSEP 10-0052 / Page 3 cc (with enclosures):

U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Luis A. Reyes, Regional Administrator 245 Peachtree Center Avenue, NE, Suite 1200 Atlanta, GA 30303-1257 U. S. Nuclear Regulatory Commission ATTN: Mr. Philip B. O'Bryan, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)

ATTN: Mrs. Farideh E. Saba (Mail Stop OWFN 8G9A) 11555 Rockville Pike Rockville, MD 20852-2738 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Mr. W. Lee Cox, III, Section Chief Radiation Protection Section North Carolina Department of Environment and Natural Resources 1645 Mail Service Center Raleigh, NC 27699-1645

BSEP 10-0052 Enclosure 1 Page 1 of 12 Evaluation of Proposed License Amendment Request

Subject:

Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)"

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 for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The proposed license amendments revise Technical Specification (TS) 5.6.5.b by adding AREVA Topical Report ANP-1 0298PA, ACE/ATRIUM IOXM CriticalPower Correlation,Revision 0, March 2010, to the list of analytical methods that have been reviewed and approved by the NRC for determining core operating limits.

2.0 Proposed Change The proposed amendments will add AREVA Topical Report ANP-10298PA, ACE/ATRIUM 10XM CriticalPower Correlation,Revision 0, March 2010, to the list of analytical methods specified in Technical Specification 5.6.5.b that have been reviewed and approved by the NRC for determining core operating limits.

For convenience, Enclosure 2 contains a marked-up version of the Unit 1 Technical Specifications showing the proposed changes. Since Technical Specification 5.6.5.b for Unit 1 and Unit 2 is identical, only the mark-up for Unit 1 is provided. Enclosures 3 and 4 provide typed versions of the Unit 1 and Unit 2 Technical Specifications, respectively. These typed Technical Specification pages are to be used for issuance of the proposed amendments.

In addition, in support of this proposed Technical Specification change, Technical Specification Bases Section 2.1.1 will be revised to reflect application of AREVA Topical Report ANP-10298PA, ACE/ATRIUM ]OXM CriticalPower Correlation. The Technical Specification Bases changes are provided in Enclosure 5, for information only, and do not require NRC approval.

3.0 Background Core operating limits are established each operating cycle in accordance with Technical Specification 3.2, "Power Distribution Limits" and Technical Specification 5.6.5, "Core Operating Limits Report (COLR)." These operating limits ensure that the fuel design limits are not exceeded during any conditions of normal operation and in the event of any Anticipated Operational Occurrence (AOO). The methods used to determine the

BSEP 10-0052 Enclosure 1 Page 2 of 12 operating limits are those previously found acceptable by the NRC and listed in Technical Specification 5.6.5.b.

On March 27, 2008 (i.e., ADAMS Accession Number ML080870478), in response to CP&L's application dated January 22, 2007 (i.e., ADAMS Accession Number ML070300570), the NRC issued License Amendments 246 and 274 for BSEP, Units 1 and 2, respectively, revising the Technical Specifications to support use of AREVA fuel and core design methodologies. Beginning with the Cycle 17 reactor core for BSEP, Unit 1 and Cycle 19 core for BSEP, Unit 2, CP&L began using AREVA fuel and core design methodologies to determine core operating limits.

On August 19, 2009 (i.e., ADAMS Accession Number ML092321080), CP&L presented plans to use the ATRIUM IOXM (A OXM) fuel design at BSEP to the NRC. As presented on August 19, 2009, use of the A OXM fuel design includes the addition of the ACE/A OXM Critical Power Correlation to the list of analytical methods specified in Technical Specification 5.6.5.b that have been reviewed and approved by the NRC for determining core operating limits.

By letter dated March 11, 2010 (i.e., ADAMS Accession Number ML100670225), the NRC staff found that Topical Report ANP-10298P, Revision 0, ACE/ATRIUM ]OXM CriticalPower Correlation,is acceptable for referencing in licensing applications for boiling water reactors to the extent specified and under the limitations delineated in the topical report and the NRC's final safety evaluation. By letter dated April 23, 2010, AREVA transmitted to the NRC the accepted version of the topical report, ANP- 10298PA, A CE/ATRIUM I OXM CriticalPower Correlation,Revision 0, March 2010.

By letter dated April 15, 2010, AREVA provided to the NRC the results of evaluations performed for the A1OXM fuel design to demonstrate compliance with NRC approved fuel licensing criteria defined in ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic MechanicalDesign Criteriafor BWR Fuel Designs,Advanced Nuclear Fuels Corporation, May 1995. These results are presented in ANP-2899P, Revision 0, Fuel Design Evaluationfor ATRIUMTM IOXMBWR Reload Fuel, provided as an enclosure to the April 15, 2010 AREVA letter. With ANF-89-98(P)(A) Revision 1 and Supplement 1, the NRC approved a set of generic acceptance criteria to be satisfied by AREVA for new BWR fuel designs. In accordance with the process described in ANF-89-98(P)(A)

Revision 1 and Supplement 1,' new fuel designs or fuel design changes satisfying the ANF-89-98(P)(A) acceptance criteria do not require explicit staff review and approval (i.e., satisfaction of the acceptance criteria is sufficient for approval by reference to the acceptance criteria).

In a separate license amendment request, submitted by letter dated April 29, 2010, CP&L has proposed another revision to TS 5.6.5.b to incorporate an unrelated AREVA topical

BSEP 10-0052 Enclosure 1 Page 3 of 12 topical report (i.e., AREVA Topical Report BAW-1 0247PA, Realistic Thermal-MechanicalFuel Rod Methodologyfor Boiling Water Reactors, Revision 0, April 2008).

The proposed addition of AREVA Topical Reports ANP- 10298PA, Revision 0, and BAW-10247PA, Revision 0, are independent of each other.

Technical Analysis System Description/ApplicableSafety Analysis For each operating cycle, the core design is evaluated to ensure that greater than 99.9 percent of the fuel rods in the reactor core avoid transition boiling during plant operation, if the safety limit is not exceeded. The derivation of MCPR Safety Limit values in the Technical Specifications, using the NRC-accepted methods described in the topical reports specified in Technical Specification 5.6.5.b, ensures the MCPR Safety Limit is not exceeded during all modes of plant operation and anticipated operational occurrences.

Currently, reactor core critical power limits are determined using the analytical methodology described in Topical Reports EMF-2209(P)(A), SPCB CriticalPower Correlation;EMF-2245(P)(A), Application of Siemens Power Corporations Critical Power Correlationsto Co-ResidentFuel; and ANF-524(P)(A), ANF CriticalPower Methodology for Boiling Water Reactors. These topical reports have been previously accepted by the NRC and are listed in Technical Specification 5.6.5.b as methodologies that may be used to determine core operating limits for BSEP, Units 1 and 2. The proposed amendments will add AREVA Topical Report ANP-10298PA, ACE/ATRIUM 1OXM CriticalPower Correlation,Revision 0, March 2010, to the list of analytical methods specified in Technical Specification 5.6.5.b that have been reviewed and approved by the NRC for determining core operating limits.

Upon approval of this license amendment application and incorporation of Topical Report ANP- 10298PA, Revision 0, into the BSEP Unit 1 and 2 Technical Specifications, CP&L will implement the analytical methods described in the report, and in conformance with the limitations described in the topical report and the NRC's safety evaluation.

Conformance with Methodology and Safety Evaluation Limitations The methodology described in ANF-524(P)(A), ANF CriticalPower Methodologyfor Boiling Water Reactors, uses Monte Carlo evaluations to demonstrate that at least 99.9 percent of the fuel rods in the core will not experience dry-out during normal operation and anticipated operational occurrences.

By letter dated February 14, 2008 (i.e., ADAMS Accession Number ML080520349),

CP&L confirmed the power distribution uncertainties used by AREVA's ANF-524(P)(A)

BSEP 10-0052 Enclosure 1 Page 4 of 12 MCPR Safety Limit methodology were applicable to BSEP. The ACE/A1OXM critical power correlation has no impact on power distribution uncertainty, therefore, this confirmation remains applicable.

The ANF-524(P)(A) methodology is modified slightly for use with the ACE correlation form due to the channel integration process used with the ACE correlation. The modifications concern the treatment of channel bow variation along the length of the fuel bundle. The detailed explanation and justification for the modifications were provided by AREVA in response to NRC staff Request for Additional Information (RAI) 18 in ANP-10249PA, ACE/ATRIUM-10 CriticalPower Correlation,Revision 0. In the Safety Evaluation of ANP-10249PA, the NRC staff reviewed the revised methodology and concluded that this treatment of channel bow remains conservative. The ACE/ATRIUM- 10 and the ACE/AI OXM correlations share the same basic correlation form. As described in Section 5.9 of ANP-10298PA, the ACE/A1OXM correlation implements the same channel bow treatment as the ACE/ATRIUM- 10 correlation.

Therefore, consistent with the approved methodology, CP&L's implementation of the analytical methods described in Topical Report ANP-10298PA, Revision 0, will include these modifications.

Analyses to determine whether a change to the Technical Specification MCPR Safety Limit will be required, with implementation of the ACE/AI OXM correlation methodology, have not been completed. If analyses indicate that a change is required, the Technical Specification change will be separately requested. The BSEP Unit 2 Cycle 20-MCPR Safety Limit analyses are planned to be complete in August 2010; therefore, the request for BSEP Unit 2 would be made at that time, if required. CP&L expects these analyses will indicate a Technical Specification change is required, based on the conservatism inherent in the treatment of channel bow by the ACE/A1OXM correlation methodology. An example of this conservatism is available, for information, in Section 4.1 and Table 4-1 of AREVA document ANP-10307P, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0. AREVA transmitted this document to the NRC by letter dated October 14, 2009 (i.e., ADAMS Accession Number ML092920444).

The NRC identified two limitations and conditions on use of the ACE/AI OXM correlation. These limitations and conditions, and the demonstration that BSEP complies with the conditions, follow.

Limitation and Condition 1:

"Since ACE/ATRIUM-1OXM was developed from test assemblies designed to simulate ACE/ATRIUM-I OXM fuel, the methodology may only be used to perform evaluations for fuel of that type without further justification."

BSEP 10-0052 Enclosure 1 Page 5 of 12 The first BSEP Unit 2 core to use A10XM fuel will be BSEP Unit 2 Cycle 20. The BSEP Unit 2 Cycle 20 core is expected to consist of approximately 226 fresh AI OXM fuel assemblies, 238 once burned ATRIUM-10 fuel assemblies, and 96 twice burned GEl4 fuel assemblies loaded in or adjacent to the control cells on the core periphery.

The first BSEP Unit 1 core to use A1OXM fuel will be BSEP Unit 1 Cycle 19. The BSEP Unit 1 Cycle 19 core will contain the third reload of AREVA fuel on this unit and, therefore, is not expected to contain any GE1 4 fuel.

While permitted by the ANP- 10298PA methodology, CP&L will not apply the ACE/A OXM critical power correlation to co-resident GE 14 fuel using the EMF-2245(P)(A) methodology, as described by ANP-10298PA. CP&L will continue to apply the SPCB correlation to this fuel design.

CP&L will not apply the ACE/AI OXM correlation to AREVA fuel other than the A1OXM fuel design.

Limitation and Condition 2:

"ACE/ATRIUM- OXM should not be used outside its range of applicability defined by the range of the test data from which it was developed and the additional justifications provided by AREVA in this submittal. This range is listed in Table 2-1 of Reference 1." Reference 1 is ANP-10298P.

Table 2-1 of ANP-1 0298PA identifies the ACE/A 10XM correlation range of applicability for conditions of mass flow rate, pressure, inlet subcooling, and design local peaking. The restrictions on these conditions are implemented in AREVA engineering guidelines. The restrictions on range of applicability for mass flow rate, pressure, and inlet subcooling are also implemented in AREVA engineering computer codes, which include the BSEP POWERPLEX-II1 core monitoring system. The restriction on design local peaking is also implemented in AREVA automation tools.

ATRIUM-] OXM Fuel Design The NRC approved the use of AREVA fuel and core design methodologies to determine BSEP core operating limits with the issuance of License Amendments 246 and 274 for BSEP, Units 1 and 2, respectively. AREVA licensing topical report ANF-89-98(P)(A).

Revision 1 and Supplement 1 is one of these NRC-approved methodologies.

ANF-89-98(P)(A) Revision 1 and Supplement 1, as clarified by a Siemens Power Corporation letter dated October 12, 1999 (i.e., Reference 1) and an NRC letter dated May 31, 2000 (i.e., Reference 2) requires that a summary of the evaluation of the A1OXM design against the NRC-approved generic design criteria be provided to the NRC for information. AREVA provided this evaluation to the NRC for information by letter dated April 15, 2010, which transmitted AREVA document ANP-2899P, Revision 0, Fuel

BSEP 10-0052 Enclosure 1 Page 6 of 12 Design Evaluationfor A TRIUMTM 1 OXM B WR Reload Fuel. In accordance with the process described in ANF-89-98(P)(A) Revision 1 and Supplement 1, new fuel designs or fuel design changes satisfying the ANF-89-98(P)(A) design criteria do not require explicit NRC review and approval (i.e., satisfaction of the design criteria is sufficient for approval by reference to the criteria).

ANP-2899P identifies fuel design criteria, specified in ANF-89-98(P)(A) Revision 1 and Supplement 1, which are evaluated on a cycle specific basis. Reports summarizing the results of analyses performed to demonstrate BSEP compliance with the cycle specific criteria are provided by AREVA to CP&L as part of the normal reload licensing document package. This type of information is not available until later in the reload licensing process. Consistent with the process described in ANF-89-98(P)(A) Revision 1 and Supplement 1 (as clarified by References 1 and 2), CP&L will provide the reports produced for the BSEP Unit 2 Cycle 20 reload to the NRC for information. The reports will be provided in supplemental letters as they are completed during the reload licensing process, on the schedule presented below.

Report Schedule for Transmitting to NRC Fuel Cycle Design Report August 2010 Thermal-Hydraulic Design Report August 2010 LOCA Analysis Reports September 2010 Mechanical Design Reports November 2010 Reload Safety Analysis Report November 2010 ANP-2899P also identifies fuel design criteria, specified in ANF-89-98(P)(A) Revision 1 and Supplement 1, that are evaluated on a plant-specific basis. These criteria address thermal hydraulic compatibility, fuel lift-off, structural deformation and LOCA performance.

The key differences in system configuration between BSEP Unit 1 and Unit 2 are in the core inlet region and the Turbine Bypass System. The orifice diameter in Unit 2 is smaller than Unit 1, 2.09 inches compared to 2.43 inches, and the Turbine Bypass System for Unit 2 has 10 valves whereas Unit 1 has 4 valves. Differences in neutronic design and operation are minimal since both units operate on 24 month fuel cycles.

Based on the minimal differences between Units 1 and 2, CP&L will include, for information, the Thermal-Hydraulic Design and Reload Safety Analysis Reports with our transmittal of the Core Operating Limits Report prior to startup from the first Unit 1 refueling outage that loads AREVA A1OXM fuel into the'reactor core. These reports will summarize compliance with plant and cycle specific ANF-89-98(P)(A) Revision 1 and Supplement 1 fuel design criteria for the first Unit 1 cycle that uses A1OXM fuel.

Additional information supporting evaluation of plant specific thermal-hydraulic

BSEP 10-0052 Enclosure 1 Page 7 of 12 compatibility, fuel lift off, structural deformation and LOCA performance criteria for Unit 1 is provided below.

Section 4.1.1 of ANP-2899P presents the results of an example thermal hydraulic compatibility analysis to demonstrate that the Al 0XM fuel is compatible with the ATRIUM-10 fuel design for an example BWR/4 core. The example BWR/4 core used for this analysis is BSEP Unit 1; however, the Thermal-Hydraulic Design Report prepared for AI OXM fuel introduction into BSEP Unit 1 will also be provided to the NRC for information as described above.

The plant-specific fuel lift off criteria are evaluated for A OXM as described in Section 3.3.8 of ANP-2899P. These criteria are satisfied for the ATRIUM-10 fuel design currently operating in both BSEP Unit 1 and Unit 2. A1OXM fuel lift margin is greater than ATRIUM- 10 based on fuel mass and pressure drop differences presented in Tables 2.1 and 4.2 of ANP-2899P; therefore, plant-specific fuel lift criteria remain satisfied for both BSEP Unit 1 and Unit 2 with A0OXM fuel.

The plant-specific structural deformation fuel design criteria are evaluated for AI OXM as described in Section 3.4.4 of ANP-2899P. The structural deformation fuel design criteria were confirmed for the ATRIUM-10 fuel design currently operating in both BSEP Unit 1 and Unit 2 based on BSEP core support plate motions calculated using ATRIUM-10 fuel channel properties and an assembly weight slightly greater than the ATRIUM- 10 fuel design. These BSEP core support plate motions are not impacted by use of A1OXM fuel, because the A0OXM fuel design uses channels of the same design and material (i.e.,

Zircaloy-4) as the ATRIUM-10 fuel design, and the A1OXM assembly weight is consistent with that used to calculate the BSEP core support plate motions. AREVA will evaluate the consequences of BSEP core support plate motion on Al0XM fuel. The results of this AREVA evaluation will be summarized in the Brunswick Unit 2 Cycle 20 Mechanical Design Report, which will be provided to the NRC for information as described above. These results will be applicable to both BSEP Unit 1 and Unit 2, because the BSEP Unit 1 and Unit 2 core support plate motions are the same.

The plant-specific LOCA performance criteria are evaluated for A10XM as described in ANP-2899P. Break spectrum and heat-up LOCA analyses are performed as part of the reload analyses for the first reload using a new fuel design, and only heat-up analyses are performed for any new lattice designs introduced thereafter. Heat-up analyses performed for new lattice designs typically confirm the applicability of the initial LOCA analyses.

The bounding BSEP unit is determined and analyzed as part of the initial reload LOCA analyses, so that the analyses will be applicable to both BSEP Unit 1 and Unit 2. The results of the initial reload evaluation will be summarized in the Brunswick LOCA Reports, which will be provided to the NRC for information as described above.

BSEP 10-0052 Enclosure 1 Page 8 of 12 5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration The proposed change will add, to Technical Specification 5.6.5.b, an additional topical report describing an NRC reviewed and approved analytical method for determining core operating limits. The new analytical method, which is described in AREVA Topical Report ANP-10298PA, ACE/ATRIUM ]OXM CriticalPower Correlation,Revision 0, March 2010, provides a new correlation for predicting the critical power for boiling water reactors containing A1OXM fuel. 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 CFR 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 probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The proposed amendments add an additional analytical methodology to the list of NRC-approved analytical methods identified in Technical Specification 5.6.5.b that can be used to establish core operating limits. The change does not require any physical plant modifications, physically affect any plant components, or entail changes in plant operation. Since no individual precursors of an accident are affected, the proposed amendments do not increase the probability of a previously analyzed event.

The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. The proposed amendments add an additional analytical methodology to the list of NRC-approved analytical methods used to establish core operating limits. The addition of the topical report to Technical Specification 5.6.5.b will allow a new analytical methodology to be used to determine critical power ratio limits. Minimum Critical Power Ratio (MCPR) Safety Limit values, which are defined in Technical Specification 2.1.1.2, are calculated to ensure that greater than 99.9 percent of the fuel rods in the reactor core avoid transition boiling during plant operation, if the safety limit is not exceeded. The derivation of MCPR Safety Limit values in the Technical Specifications, using these NRC-accepted methods, will continue to ensure the MCPR Safety Limit is not exceeded during all modes of plant operation and anticipated operational occurrences. The addition of the analytical methodology described in Topical Report ANP-1 0298PA to Technical Specification 5.6.5.b does not alter the assumptions of accident analyses or the Technical Specification Bases.

BSEP 10-0052 Enclosure 1 Page 9 of 12 Based on the above, the proposed amendments do not increase the consequences of a previously analyzed accident.

Therefore, the proposed amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No Creation of the possibility of a new or different kind of accident requires creating one or more new accident precursors: New accident precursors may be created by modifications of plant configuration, including changes in allowable modes of operation. The proposed amendments do not involve any plant configuration modifications or changes to allowable modes of operation. The proposed Topical Report addition to Technical Specification 5.6.5.b provides an analytical methodology for determining core critical power limits that ensures no new accident precursors are created. Therefore, the proposed amendments do not create the possibility 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 add an additional analytical methodology to the list of NRC-approved analytical methods identified in Technical Specification 5.6.5.b that can be used to establish core operating limits. This addition to Technical Specification 5.6.5.b will allow a new NRC-accepted analytical methodology to be used to determine critical power ratio limits. The MCPR Safety Limit provides a margin of safety by ensuring that at least 99.9 percent of the fuel rods do not experience transition boiling during normal operation and anticipated operational occurrences if the MCPR Safety Limit is not exceeded. The proposed change will ensure the current level of fuel protection is maintained by continuing to ensure that the fuel design safety criterion (i.e., that no more than 0.1 percent of the rods are expected to be in boiling transition if the MCPR Safety Limit is not exceeded) is met.

Therefore, the proposed amendments do not result in a significant reduction in the margin of safety.

BSEP 10-0052 Enclosure 1 Page 10 of 12 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 Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.

CP&L has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the Technical Specifications, and do not affect conformance with any General Design Criterion (GDC) differently than described in, the Updated Final Safety Analysis Report (UFSAR).

10 CFR 50.36(c)(5) states that the Technical Specifications will include administrative controls that address the provisions relating to organization and management, procedures, record keeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. The Core Operating Limits Report (COLR) is required as a part of the reporting requirements specified in the Brunswick Technical Specifications Administrative Controls section. The Technical Specifications require the core operating limits to be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and to be documented in the COLR. In addition, it requires the analytical methods used to determine the core operating limits to be those that have been previously reviewed and approved by the NRC, and specifically to be those described in Technical Specification 5.6.5.b. The proposed amendments ensure that these requirements are met.

Generic Letter (GL) 88-16, Removal of Cycle-Specific Parametersfrom Technical Specifications, provided guidance on relocating numerical values in Technical Specifications and the referencing of associated methodology topical reports in the Administrative Controls Technical Specifications and the COLR. However, in a recent letter to the Technical Specification Task Force (TSTF) (i.e., Reference 3), the NRC has informed the TSTF that relaxation of Technical Specification Topical Report documentation from the guidance provided in GL 88-16 and Technical Specification Task Force (TSTF) Traveler TSTF-3 63, Revise Topical Report References in ITS 5.6.5, COLR, is no longer appropriate. Accordingly, the proposed addition of the AREVA Topical Report ANP-10298PA methodology reference to Technical Specification 5.6.5.b includes the information necessary to identify the specific revision of the methodology (i.e., the revision number and issuance date).

10 CFR 50, Appendix A, General Design Criterion (GDC) 10 requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any

BSEP 10-0052 Enclosure 1 Page 11 of 12 condition of normal operation, including the effects of anticipated operational occurrences.

To ensure compliance with GDC 10, CP&L performs plant-specific critical power limit analyses using NRC-approved methodologies. The MCPR Safety Limit ensures that sufficient conservatism exists in the operating limit MCPR such that, in the event of an anticipated operational occurrence, there is a reasonable expectation that at least 99.9 percent of the fuel rods in the core will avoid boiling transition for the power distribution within the core including all uncertainties.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 Environmental Considerations A review has determined that the proposed amendments are administrative in nature and do not change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, and do not change an inspection or surveillance requirement. 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 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendments.

7.0 References

1. Letter from James F. Mallay (Siemens Power Corporation) to the NRC Document Control Desk, "Revisions to Attachment 1 of Letter NRC:99:030, Request for Concurrence on SER Clarifications," dated October 12, 1999.

2. Letter from Stuart Richards (NRC) to James F. Mallay (Siemens Power Corporation), "Siemens Power Corporation Re: Request for Concurrence on Safety Evaluation Report Clarifications (TAC No. MA6160)," dated May 31, 2000.

BSEP 10-0052 Enclosure 1 Page 12 of 12

3. Letter from Stacey L. Rosenberg (NRC) to the Technical Specification Task Force, "Technical Specification Task Force (TSTF) Traveler 363, Revision 0,

'Revise Topical Report References in ITS 5.6.5, COLR'," dated November 2, 2009.

BSEP 10-0052 Enclosure 2 Marked-up Technical Specification Pages - Unit 1

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

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. 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.

20. ANP1O298PA, ACE! 16. EMF-2361(P)(A), EXEM BWR-2000 ECCS Evaluation Model.

ATRIUM 1OXM Cr itical Power Correlation, 17. EMF-2292(P)(A), ATRIUMTM-10: Appendix K Spray Heat Transfer Revision 0, March 2010. Coefficients.

18. EMF-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.

(continued)

Brunswick Unit 1 5.0-21 Amendment No. 246 I

BSEP 10-0052 Enclosure 3 Typed Technical Specification Pages - Unit 1

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

20. ANP-10298PA, ACE/ATRIUM 10XM Critical Power Correlation, Revision 0, March 2010.

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.

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 Amendment No. I

BSEP 10-0052 Enclosure 4 Typed Technical Specification Page - Unit 2

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

20. ANP-10298PA, ACE/ATRIUM 1OXM Critical Power Correlation, Revision 0, March 2010.

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.

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 2 5.0-22 Amendment No. I

BSEP 10-0052 Enclosure 5 Marked-up Technical Specification Bases Pages - Unit 1 (For information only)

Reactor Core SLs B 2.1.1 BASES BACKGROUND Operation above the boundary of the nucleate boiling regime could result (continued) in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.

Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

The reactor vessel water level SL ensures that adequate core cooling capability is maintained during all MODES of reactor operation.

Establishment of Emergency Core Cooling System initiation setpoints higher than this safety limit provides margin to ensure the safety limit will not be reached or exceeded such that fuel damage would occur.

APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and AOOs. The reactor core SLs are established to preclude violation of the fuel design criterion that an MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.

The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in combination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR limit.

2.1.1.1 Fuel Cladding Inti correlations are > 0.09LE+/-6 The SRPG8 critical power RO~- a ~d r ri cal ýwercalculationss 2

at pressures >-Z74--4 psia and bundle mass fluxes : *7-F4-6 Ibm/hr-ft Refeee*le4). Fol ation I r ssures or low flows, another basis i used, as follows: > 600 Since the pressure dro i e ypass region is essentially all elevation head, the core pressure drop at low power and flows will always be > 4.5 psi. Analyses show that with a bundle flow of References 1 and 4 28 x 103 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi., Thus, the bundle flow with a 4.5 psi driving head will be > 28 x 103 lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia (0 psig) to 800 psia (785 psig) indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER > 46% RTP.

Thus, a THERMAL POWER limit of 23% RTP for reactor pressure

< 785 psig is conservative.

(continued)

Brunswick Unit 1 B 2.1.1-2 Revision No. 58 1

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2 MCPR SAFETY ANALYSES (continued) The fuel cladding integrity SL is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved AREVA critical power correlations. One specific uncertainty included in the SL is the uncertainty inherent in the critical power correlation. References 1, 2, 3 and 4 - ap41- describe the uncertainties and methodology used to determine the MCPR SL.

2.1.1.3 Reactor Vessel Water Level During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active irradiated fuel to provide core cooling capability. In conjunction with LCOs, the limiting safety system settings, defined in LCO 3.3.1.1 as the Allowable Values, establish the threshold for protective system action to prevent exceeding acceptable limits, including this reactor vessel water level SL, during Design Basis Accidents. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

(continued)

Brunswick Unit 1 B 2.1.1-3 Revision No. 58 I

Reactor Core SLs B 2.1.1 BASES SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to prevent the release of radioactive materials to the environs.

SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential VIOLATIONS for radioactive releases in excess of 10 CFR 50.67, "Accident Source e(Re-4). Therefore, it is required to insert all insertable I and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The E2 hour Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

REFERENCES 1. EMF-2209(P)(A), "SPCB Critical Power Correlation," (as identified in the COLR).

2. EMF-2245(P)(A), "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," (as identified in the COLR).

3. ANF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors," (as identified in the COLR).

10 IF 50--.67.

ANP-10298PA, "ACE/ATRIUM 10XM Critical Power Correlation," (as identified in the COLR).

10 CFR 50.67.

Brunswick Unit 1 B 2.1.1-4 Revision No. 58 I