BSEP 10-0133, Response to Additional Information Request Supporting License Amendment Requests for Addition of Analytical Methodology Topical Reports to Technical Specification 5.6.5 (NRC TAC Nos. ME3856, ME3857, ME3858, and ME3859)

Response to Additional Information Request Supporting License Amendment Requests for Addition of Analytical Methodology Topical Reports to Technical Specification 5.6.5 (NRC TAC Nos. ME3856, ME3857, ME3858, and ME3859)

ML103330242
Person / Time
Site:	Brunswick
Issue date:	11/18/2010
From:	Jefferson W Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 10-0133, TAC ME3856, TAC ME3857, TAC ME3858, TAC ME3859, TSC-2010-01, TSC-2010-02
Download: ML103330242 (51)

Text

Progress Energy NOV` 182010 10 CFR 50.90 TSC-2010-01 & TSC-2010-02 SERIAL: BSEP 10-0133 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 Response to Additional Information Request Supporting License Amendment Requests for Addition of Analytical Methodology Topical Reports to Technical Specification 5.6.5 (NRC TAC Nos. ME3856, ME3857, ME3858, and ME3859)

References:

1. Letter from Michael J. Annacone to the U.S. Nuclear Regulatory Commission (Serial: BSEP 10-0052), "Request for License Amendments - Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," dated April 29, 2010 (ADAMS Accession Number ML101310388)

2. Letter from Michael J. Annacone to the U.S. Nuclear Regulatory Commission (Serial:

BSEP 10-0057), "Request for License Amendments - Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," dated April 29, 2010 (ADAMS Accession Number ML101340389)

3. Letter from Farideh E. Saba (USNRC) to Michael J. Annacone (CP&L),

"Brunswick Steam Electric Plant, Units 1 and 2 - Request for Additional Information Regarding License Amendment Requests for Addition of Analytical Methodology Topical Reports to TS 5.6.5, Core Operating Limits Report (TAC Nos. ME3856, ME3857, ME3858, and ME3859), dated October 28, 2010 (ADAMS Accession Number ML102910167)

Ladies and Gentlemen:

By letters dated April 29, 2010 (i.e., References 1 and 2), Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., requested license amendments to revise the Technical Specifications (TS) for the Brunswick Steam Progress Energy Carolinas, Inc.

Brunswick Nuclear Plant PO Box 10429 Southport, NC 28461 C)

Document Control Desk BSEP 10-0133 / Page 2 Electric Plant (BSEP), Unit Nos. 1 and 2. The proposed license amendments revise Technical Specification 5.6.5.b by adding: (1) AREVA Topical Report ANP- 10298PA, A CE/A TRIUM 1OXM CriticalPower Correlation,Revision 0, March 2010, and (2) AREVA Topical Report BAW- 10247PA, Realistic Thermal-MechanicalFuel Rod Methodologyfor Boiling Water Reactors, Revision 0, April 2008, to the list of analytical methods that have been reviewed and approved by the NRC for determining core operating limits.

By letter dated October 28, 2010, the NRC requested additional information regarding the referenced license amendment requests. The enclosure to this letter provides the requested information.

No regulatory commitments are contained in this letter. Please refer any questions regarding this submittal to Mr. Lee Grzeck, Acting Supervisor - Licensing/Regulatory Programs, at (910) 457-2487.

I declare, under penalty of perjury, that the foregoing is true and correct. Executed on November 18, 2010.

Wlliam erso~n, Direct - Site Operations Brunswick Steam Electric Plant WRM/wrm

Enclosures:

1. Response to Request for Additional Information (Proprietary Information -

Withhold from Public Disclosure in Accordance With 10 CFR 2.390)

2. AREVA Affidavit Regarding Withholding Enclosure 1 from Public Disclosure

3. Response to Request for Additional Information (Non-Proprietary Version)

Document Control Desk BSEP 10-0133 / Page 3 cc (with Enclosures 1, 2, and 3):

U. S. Nuclear Regulatory Commission, Region I1 ATTN: Mr. Luis A. Reyes, Regional Administrator 245 Peachtree Center Ave, 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 cc (with Enclosures 2 and 3):

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 10-0133 Enclosure 2 AREVA Affidavit Regarding Withholding Enclosure 1 from Public Disclosure

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. 1am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in the attachment to Progress Energy letter BSEP 10-0133 with subject entitled "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, Response to Additional Information Request Supporting License Amendment Requests for Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5 (NRC TAC Nos. ME3856, ME3857, ME3858, and ME3859) dated November 2010 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Documentis considered proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available,

on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this /

day of _______________) _,2010.

Kathleen Ann Bennett NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 8/31/11 Reg. # 110864

BSEP 10-0133 Enclosure 3 Page 1 of 44 Response to Request for Additional Information (Non-Proprietary Version)

By letters dated April 29, 2010, Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., requested license amendments to revise the Technical Specifications (TS) 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: (1)AREVA Topical Report ANP-10298PA, ACE/ATRIUM IOXM Critical Power Correlation,Revision 0, March 2010, and (2) AREVA Topical Report BAW-10247PA, Realistic Thermal-Mechanical Fuel Rod Methodologyfor Boiling Water Reactors, Revision 0, April 2008, to the list of analytical methods that have been reviewed and approved by the NRC for determining core operating limits. On October 28, 2010, via a letter, the NRC requested additional information regarding the referenced license amendment requests. The requested information is provided below.

NRC Question No. I BSEP 10-0052, Page 5 of 12, Limitation and Condition I The licensee has indicated that "CP&L will not apply the ACE/A IOXM correlation to AREVA fuel other than the AI OXM fuel design."

Please explain the procedure and methods that are in place at BSEP to support this statement, as well as how this restriction is implemented for transients and accident analyses for the BSEP Unit 1 and 2 cores containing different types of fuel.

Response No. I BSEP procedures require review of each core reload in accordance with 10 CFR 50.59. This review confirms core operating limits specified in Technical Specifications (TS) 5.6.5.a, including minimum critical power ratio (MCPR) limits based on transient and accident analyses performed for the core reload, are determined using analytical methods listed in TS 5.6.5.b that have been reviewed and approved by the NRC. These analytical methods include restrictions on their use. BSEP procedures governing creation of the Core Operating Limits Report (COLR) for each core reload also require these core operating limits be determined with the NRC-approved methodologies listed in TS 5.6.5.b. BSEP procedures governing the creation of the core monitoring system data bank implement fuel design-specific critical power ratio (CPR) correlation restrictions for core monitoring. These BSEP procedures and administrative methods implement the restrictions on application of CPR correlations to the different types of fuel in the BSEP Unit 1 and 2 cores.

AREVA engineering guidelines implement the restriction that the ACE/ATRIUM 1OXM correlation is used only for the ATRIUM 1OXM (A1OXM) fuel design unless further justification is provided, consistent with the Safety Evaluation Report (SER) restriction presented in ANP-1 0298PA, Revision 0, ACE/ATRIUM JOXM CriticalPower Correlation. AREVA

BSEP 10-0133 Enclosure 3 Page 2 of 44 engineering guidelines similarly implement the EMF-2209(P)(A), SPCB CriticalPower Correlation,restriction that the SPCB correlation is used only for ATRIUM-9B and ATRIUM- 10 (AlO) AREVA fuel designs. These correlations may be applied to previously exposed co-resident fuel (i.e., non AREVA fuel) in accordance with EMF-2245(P)(A), Application of Siemens Power Corporations CriticalPower Correlationsto Co-Resident Fuel, and the applicable approved correlation Licensing Topical Report. AREVA and BSEP apply the ACE/ATRIUM 1OXM CPR correlation to AI OXM fuel and the SPCB correlation to A1O and co-resident Brunswick GE14 fuel.

NRC Question No. 2 BSEP 10-0052, Page 5 of 12, Limitation and Condition 2 Please explain how the restrictions on the range of conditions for mass flow rate, pressure, inlet subcooling, and design local peaking for the ACE/ATRIUM IOXM critical power ratio correlation are implemented in the AREVA engineering guidelines and in the BSEP core monitoring system.

Response No. 2 Flow rate, pressure, and inlet subcooling limitations are programmatically handled by the ACELIB software library in compliance with the safety evaluation and methodology description of ANP-10298PA. Error messages are generated in the calculation when conditions fall outside the range of applicability according to the limitations set forth in ANP-10298PA. Each computer code using the ACE/ATRIUM I OXM correlation uses this library. The ACELIB software library is used directly by MICROBURN-B2 (MB2) as part of the BSEP core monitoring system and by AREVA engineering analysis software.

Design local peaking is controlled administratively during the core and assembly design. The automated documentation processes include a check of each assembly design to ensure that the limit on design local peaking is satisfied. In the safety limit calculation, when a perturbation causes the local peaking to exceed the design local peaking limit, additional uncertainty is assigned. This additional uncertainty is identified in ANP-10298PA.

NRC Question No. 3 ANP-2899, Part Length Rods (PLRs)

Given that the BSEP Unit 2 core will have three different types of fuel assemblies with PLRs of different lengths, please address the following:

(a) What is the possibility of any of these PLRs being located in a limiting position?

(b) Will any of these rods undergo boiling transition/dryout during normal operating conditions or during transients and accident conditions?

(c) How are the PLRs treated during the critical power ratio (CPR) calculations?

BSEP 10-0133 Enclosure 3 Page 3 of 44 Response No. 3 (a) The part length rods are neither more or less limiting than the full length rods. Whether they are in a limiting position depends on the core and fuel assembly design and operation. Should conditions be present that indicate that dryout will occur on a part length rod, it will be the limiting rod and the critical power of the assembly will be determined accordingly.

(b) The COLR limits determined using analytical methods listed in TS 5.6.5.b that have been reviewed and approved by the NRC assure that more than 99.9% of the rods avoid boiling transition during transients. More than 99.9% of the rods also avoid boiling transition during normal operating conditions, because margin to boiling transition is greater under normal operating conditions than during transient conditions.

While the occurrence of boiling transition may not occur in some accident scenarios, the avoidance of boiling transition is not one of the acceptance criteria for accidents. Boiling transition will likely occur in some accidents. For example, during a loss of cooling accident, boiling transition is expected for most of the rods, including part length rods, in the hot assembly.

(c) Each of the correlations (i.e., SPCB and ACE/ATRIUM IOXM) treats the partial length rod no differently than a full length rod. It has an additive constant and each correlation considers if this is the limiting rod in the fuel assembly.

Ii II NRC Question No. 4 BSEP 10-0052, Gadolinia Rods Please explain how the gadolinia (U0 2-Gd 2O 3) rods are treated during the application of the ACE/ATRIUM 1OXM critical power correlation for MCPR calculations.

Response No. 4 There is no difference between the way that the CPR is calculated if the rod contains gadolinia or if the rod does not contain gadolinia. The appropriate heat generation rate is determined by the

BSEP 10-0133 Enclosure 3 Page 4 of 44 MB2 core simulator for each rod used in the determination of the CPR. All fueled rods are considered in the CPR calculation.

NRC Question No. 5 BSEP 10-0052, Page 4 of 12 The licensee stated that "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 detailed explanations and justification for these modifications were provided by AREVA in response to NRC Request for Additional Information (RAI) No. 18 in ANP- I0249(P)(A),

"ACE ATRIUM-10 Critical Power Correlation," Revision 0.

The licensee further stated that "consistent with the approved methodology, CP&L's implementation of the analytical methods described in Topical Report ANP-10298(P)(A),

Revision 0 (ACE/ATRIUM I OXM Critical Power Correlation), will include these modifications."

Please explain the potential impact of these modifications on the implementation of the ACE/ATRIUM IOXM critical power correlation for the ATRIUMI OXM fuel at BSEP.

Response No. 5 The determination of the safety limit minimum critical power ratio using the ANF-524(P)(A) methodology, as modified by Request for Additional Information (RAI) 18 in ANP-10249PA, when applying the ACE correlation form makes use of a ((

as described in the response to RAI 18 in ANP-10249PA.

The ANP-10298PA ACE/ATRIUM 1OXM critical power correlation differs from the ANP- 10249PA ACE/ATRIUM- 10 critical power correlation ((

T]

The impact of the modifications is that they ((

11

BSEP 10-0133 Enclosure 3 Page 5 of 44 NRC Question No. 6 BSEP 10-0052, Page 4 of 12 The licensee stated that "analyses to determine whether a change to the TS MCPR safety limit will be required, with the implementation of the ACE/ATRIUM 1OXM correlation methodology, have not been completed. If analyses indicate that a change is required, the Technical Specification change will be separately requested."

Please confirm, with supporting analyses documents, whether or not a change to the TS MCPR limit is required.

Response No. 6 A change to the TS MCPR limit is not required for Brunswick Unit 2 Cycle 20. The supporting analysis document is ANP-2956(P), Revision 0, Brunswick Unit 2 Cycle 20 Reload Safety Analysis, which has been provided to the NRC, for information, by CP&L letter B SEP 10-0126, dated November 9, 2010.

NRC Question No. 7 BSEP 10-0052, Page 7 of 12, ANP-2936(P)

(a) Please provide a detailed calculation to show hydraulic compatibility as it relates to the performance of the three fuel designs in the BSEP Unit 2 core.

(b) Considering the fact that the proposed BSEP Unit 2 core for the next few cycles will be a mixed core, please demonstrate, with supporting analyses, how the thermal-hydraulic design criteria per the Standard Review Plan (NUREG-0800), Section 4.4, and Table 3.1 of ANP-2936(P), are satisfied for the BSEP Unit 2 transition core configurations.

Response No. 7 (a) Results of the thermal hydraulic compatibility analysis for Brunswick Unit 2 are presented in ANP-2936(P), Revision 0, Brunswick Unit 2 Thermal-HydraulicDesign Report for ATRIUM lOXM Fuel Assemblies. Analysis results are presented for four core configurations including the core configuration for Brunswick Unit 2 Cycle 19 (A 10 and GE14 fuel) and Brunswick Unit 2 Cycle 20 (AIOXM,A1O and GE14 fuel), which is also identified as the "First transition" core in the report. Results for the first transition core presented in Tables 3.5 through 3.8 demonstrate that the thermal-hydraulic compatibility criteria are met for the Brunswick Unit 2 Cycle 20 core loading with the three fuel designs. The Brunswick Unit 2 thermal hydraulic calculations that support ANP-2936(P) were provided to the NRC via the following: Letter, R. L. Gardner (AREVA) to H. D.

Cruz (USNRC), NRC: 10:092, InformationalTransmittal ofATRIUM 10XM Supporting Calculationsfor Audit, dated October 13, 2010.

BSEP 10-0133 Enclosure 3 Page 6 of 44 (b) Demonstration that the thermal design criteria are met as Brunswick transitions to a full core of A10XM fuel is accomplished in four ways:

5. Results are presented in ANP-2936(P). Tables 3.7 through 3.8 of ANP-2936(P) present results for Brunswick Unit 2 as the core transitions from the current Brunswick Unit 2 Cycle 19 core configuration to a full core of A1OXM fuel. The results show:

IIII

6. ((

7. Plant- and fuel-specific analyses are performed to demonstrate that the fuel centerline, LOCA, and seismic criteria are met. The fuel centerline and seismic analyses are presented in the mechanical design reports (i.e., ANP-2950P, Revision 0, ATRIUM IOXM Fuel Rod Thermal and Mechanical Evaluation for Brunswick Unit 2 Cycle 20 Reload BRK2-20 and ANP-2948P, Revision 0,, Mechanical Design Report for Brunswick ATRIUM 1OXM Fuel Assemblies), which have been provided to the NRC, for information, via CP&L letters BSEP 10-0118 and BSEP 10-0126, dated October 12, 2010, and November 9, 2010, respectively. The LOCA analysis reports (i.e., ANP-2941(P), Revision 0, Brunswick Units I and 2 LOCA Break Spectrum Analysis for ATRIUM I OXM Fuel and ANP-2943(P), Revision 0, Brunswick Units I and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 1OXM Fuel), which have been provided to the NRC, for information, via CP&L letter BSEP 10-0112, dated September 30, 2010, describe the results of the Brunswick A IOXM LOCA analyses and include a discussion of the applicability of the analysis to mixed and full A IOXM core configurations.

8. Cycle-specific analyses are performed for the control rod drop accident and over pressurization evaluations. Analyses of the limiting abnormal operational occurrences (AOOs) are performed each cycle to establish operating limits to ensure the adequate thermal margin performance of the fuel. In addition, cycle-specific stability analyses are performed. In all of these cycle-specific evaluations, each fuel type in the core is explicitly modeled thereby accounting for any mixed core impact.

The results of these analyses are presented in ANP-2956(P), Revision 0, Brunswick Unit 2 Cycle 20 Reload Safety Analysis, which has been provided to the NRC, for information, via CP&L letter BSEP 10-0126, dated November 9, 2010.

BSEP 10-0133 Enclosure 3 Page 7 of 44 NRC Question No. 8 BSEP 10-0052, ANP-2936(P)

General Design Criteria (GDC) 10 of Appendix A to Title 10 of the Code of FederalRegulations (10 CFR) Part 50, 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 (SAFDL) are not exceeded during normal operation, including the effects of abnormal operational occurrences (AOOs).

In addition, acceptance criterion 1.b of the Standard Review Plan, Section 4.4, requires that at least 99.9 percent of the fuel rods in the core will not experience boiling transition during normal operation' or A0Os.

Please describe, with supporting analyses, how the ACE/ATRIUM 1OXM correlation is used to demonstrate that this requirement is met.

Response No. 8 Thermal margin performance of each fuel design in the core is determined using an applicable critical power correlation. The ACE/ATRIUM IOXM critical power correlation is used for AI OXM fuel to determine the core MCPR safety limit (i.e., MCPR at which 99.9% of the rods in the core would be expected not to experience boiling transition) and the change in CPR (ACPR) for AI OXM fuel during the limiting AOOs. These results are used to establish MCPR operating limits for AIOXM fuel. Since the MCPR safety limit evaluation is a core wide evaluation, the analysis applies the applicable critical power correlation to other fuel types in the core (i.e., the SPCB correlation is utilized for the Brunswick GEl4 and AIO fuel).

The A1OXM critical power correlation will also be used to monitor the A1OXM fuel to ensure that the AIOXM MCPR limits are not violated. The SPCB correlation is used to monitor the GE14 and AlO fuel.

A description of the analyses performed to establish the Brunswick Unit 2 Cycle 20 AI OXM MCPR limits is presented in ANP-2956(P), Revision 0, Brunswick Unit 2 Cycle 20 Reload Safety Analysis, which has been provided to the NRC, for information, via CP&L letter BSEP 10-0126, dated November 9, 2010.

NRC Question No. 9 ANP-2936(P)

The proposed Cycle 19 BSEP Unit 2 core with AREVA ATRIUM- 10, AREVA ATRIUM 10 XM, and GE14 fuel designs will constitute a "mixed core."

BSEP 10-0133 Enclosure 3 Page 8 of 44 Please provide details of the potential impact of the mixed core on the CPR calculations, accounting for the differences in mechanical, thermal, and hydraulic characteristics of the three fuel designs in the transition core at BSEP Unit 2.

Response No. 9 Tables 3.7 and 3.8 of ANP-2936(P), Revision 0, Brunswick Unit 2 Thermal-HydraulicDesign Reportfor ATRIUM lOXMFuel Assemblies, include CPR results for high power and average power assemblies for mixed core configurations as Brunswick Unit 2 transitions from the Brunswick Unit 2 Cycle 19 core configuration to a full core of A1OXM fuel. The core configurations used in the analyses are presented in Table 3.4 of ANP-2936(P).

In the AREVA thermal-hydraulic methodology, each fuel type is explicitly modeled. As a result, the impacts of the differences in mechanical design on geometry and loss coefficients are explicitly accounted for. The critical power performance of each fuel type is also explicitly modeled using the applicable critical power correlation for each fuel design. The ACE/ATRIUM I 0XM critical power correlation is used for AI OXM fuel. The SPCB critical power correlation (i.e., EMF-2209(P)(A), SPCB CriticalPower Correlation)is used for A10 fuel and GE14 fuel. Application of the SPCB correlation to GE14 fuel is based on the indirect process described in EMF-2245(P)(A), Application of Siemens Power Corporation'sCritical Power Correlationsto Co-Resident Fuel.

NRC Question No. 10 ANP-2936(P)

GDC 12 of 10 CFR Part 50, Appendix A requires suppression of reactor power oscillations so that SAFDLs are not exceeded.

Please demonstrate, with supporting analyses and calculations, how thermal-hydraulic and neutronic stability of the mixed core is maintained at BSEP throughout the upcoming and following cycles of operation.

Response No. 10 BSEP Units 1 and 2 are detect and suppress Option III plants that utilize a power range neutron monitoring (PRNM)-based system compliant with NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodologyfor Reload Applications. AREVA performs cycle-specific calculations to determine the required Option III system setpoint(s) necessary to ensure that the MCPR Safety Limit is not exceeded. This includes the calculation of the DIVOM (Delta over Initial MCPR Versus Oscillation Magnitude) as well as delta-MCPR response to a two recirculation pump trip (2RPT) event using the cycle-specific licensing basis core. The operating limit minimum critical power ratio (OLMCPR) versus setpoint is provided for both a steady-state and the 2RPT events as described in NEDO-32465-A. The methodology used in

BSEP 10-0133 Enclosure 3 Page 9 of 44 this analysis remains the same for AlOXM as that used for the A10 fuel, with the addition of the ACE correlation for AIOXM fuel.

For times in which the primary OPRM system is not available, regions are defined on the power-flow map in accordance with the Backup Stability Protection (BSP) described in OG02-0119-260, Backup Stability Protection(BSP)for Inoperable Option III Solution. These regions have been defined conservatively for both BSEP units and are confirmed on a cycle-specific basis.

Cycle-specific stability analyses are performed based upon the actual licensing basis core design, which explicitly includes co-resident fuel designs. Both the Oscillation Power Range Monitor (OPRM) setpoints versus OLMCPR and the backup stability regions for the initial introduction ofA1OXM are presented in ANP-2956(P), Revision 0, Brunswick Unit 2 Cycle 20 Reload Safety Analysis, which has been provided to the NRC, for information, via CP&L letter BSEP 10-0 126, dated November 9, 2010. These results are based upon explicit modeling of the BSEP Unit 2 Cycle 20 core, including the A1OXM and all co-resident fuel types. Following cycles of operation are similarly analyzed.

NRC Question No. 11 BSEP 10-0057, ANP-2899(P), Section 3.2.2 Section 3.2.2 of ANP-2899(P) indicates that "AREVA evaluates fuel centerline temperature using RODEX4 for both normal operating conditions and AOOs. Fuel failure from the overheating of the fuel pellets is not allowed. The melting point of the fuel includes adjustments for gadolinia content."

Assuming degraded thermal conductivity, and with a lower melting point of the Gd 2 O 3 -UO2 mixture, please describe what adjustments are made in the Gadolinia rods to prevent failure of the Gadolinia rod from melting. In addition, please address whether there is any restriction on the linear heat generation rate limit for the Gadolinia rods during normal operation and A0Os.

Response No. 11 The RODEX4 code takes into account thermal conductivity degradation with dependencies on both burnup and gadolinia content. As a further conservatism, the methodology considers a penalty on melting temperature as a function of gadolinia content. These features of the code and methodology were reviewed and approved as part of the RODEX4 topical report, BAW-10247PA Revision 0.

The urania and gadolinia fuel within a fuel assembly are analyzed and monitored to the same linear heat generation rate (LHGR) limit. Plant- and cycle-specific analyses are performed on the specified nuclear design to ensure that the fuel melting temperature criterion along with other fuel rod design criteria are satisfied. If during the design process a gadolinia rod is found to fail the design criteria, then a change is made to the nuclear design to lower the local peaking of the

BSEP 10-0133 Enclosure 3 Page 10 of 44 gadolinia rod until the design criteria are satisfied. Typically, the change is accomplished by lowering the enrichment content of the affected gadolinia fuel rod. Alternately, the LHGR limit to be applied to both uranium and gadolinia rods may be reduced until the design criteria are satisfied. In either case, the result is a design that meets the thermal-mechanical criteria using LHGR limits that will be applied during subsequent operation of the fuel.

In the case of an LHGR-limited fuel assembly, the fuel assembly power will be restricted by the LHGR of the highest powered fuel rod in the fuel assembly. Normally, a urania fuel rod will have the lowest margin to the LHGR limit within the fuel assembly because of the aforementioned effects on conductivity and melting temperature. It is also normal in the enrichment distribution design to include enrichment reductions in the gadolinia fuel. However, it is possible for a gadolinia fuel rod to become LHGR-limiting within the fuel assembly at some point in a cycle. For example, a fuel rod with low gadolinia concentration and higher enrichment could achieve a higher power level at intermediate exposure levels. Such a case is explicitly included as part of the RODEX4 analysis, because the cycle-specific enrichment and gadolinia design is used in producing the power history inputs to the analysis.

NRC Question No. 12 BSEP 10-0052, BSEP 10-0057, Section 5.1 In Section 5.1, "No Significant Hazards Consideration", the licensee stated that "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."

Given the fact that the proposed amendments will ultimately enable BSEP Units 1 and 2 to transition to a new AREVA fuel, please justify or cb'rrect if necessary the statements that "the change does not require any plant modifications or physically affect any plant components."

Response No. 12 The changes evaluated by the referenced "No Significant Hazards Consideration" are the license amendment changes requested by CP&L letters BSEP 10-0052 and BSEP 10-0057 dated April 29, 2010. These changes are limited solely to changes in analytical methodology, which do not physically impact the plant. CP&L letters BSEP 10-0052 and BSEP 10-0057 do not request NRC review or approval of the A0OXM fuel design, because this fuel design is separately approved upon providing to the NRC, for information, summary confirmations of compliance with the NRC approved fuel licensing criteria defined in ANF-89-98(P)(A), Revision I and Supplement 1, Generic Mechanical Design Criteriafor BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995. BSEP uses the plant modification change control process to document each core reload, and the A1OXM fuel design change will be implemented under this process in accordance with 10 CFR 50.59. The "No Significant Hazards Consideration" evaluations provided by CP&L letters BSEP 10-0052 and BSEP 10-0057, as revised by CP&L letter BSEP 10-0083 dated July 22, 2010, do not consider a fuel design change, because a fuel

BSEP 10-0133 Enclosure 3 Page I I of 44 design change is outside the scope of the license amendments requested by CP&L letters BSEP 10-0052 and BSEP 10-0057.

NRC Question No. 13 BSEP 10-0052, ANP-2899(P)

Section 4.1.4, Rod Bow, of ANP-2899P states that "at higher exposures, a CPR penalty is determined as a function of exposure and fractional rod rod spacing closure."

Please explain how CPR penalties are evaluated for higher exposures for ATRIUM IOXM fuel.

Response No. 13 As identified in section 3.3.5 of ANP-2899P, Revision 0, Fuel Design Evaluationfor ATRIUM IOXMBWR Reload Fuel, rod bow is calculated using the approved model in XN-75-32(P)(A),

Supplements I through 4, ComputationalProcedurefor Evaluating Fuel Rod Bowing. This approved model provides the method of calculating gap closure as a function of exposure.

II NRC Question No. 14 ANP-2899(P)

Please provide supporting analyses and calculations for Tables 4.2 through 4.5 of ANP-2899(P).

Response No. 14 The analysis results presented in Tables 4.2 through 4.5 of ANP-2899P are sample calculations showing the thermal-hydraulic compatibility between the A] OXM and A1O fuel designs. As indicated in CP&L letter BSEP 10-0052 dated April 29, 2010, the BWR/4 example used in the thermal-hydraulic evaluation is for Brunswick Unit 1. The intent of the results presented is to provide an example of the thermal-hydraulic design criteria evaluation.

The thermal-hydraulic calculations that support ANP-2899P were provided to the NRC via the following: Letter, R. L. Gardner (AREVA) to H. D. Cruz (USNRC), NRC: 10:092, Informational TransmittalofA TRIUM 1OXM Supporting Calculationsfor Audit, dated October 13, 2010.

BSEP 10-0133 Enclosure 3 Page 12 of 44 NRC Question No. 15 ANP-2899(P), Section 5.0, Nuclear Design Evaluation Nuclear fuel and core analyses for ATRIUM IOXM are performed using the NRC-approved XN-NF-80-19(P)(A) and EMF-2158(P)(A) methodology to assure that the new assembly and/or design features meet the nuclear design criteria established for the fuel and core.

ANP-2899(P), Section 5 0, states that the accuracy of the above methodology has been demonstrated for ATRIUM-10 fuel assembly design by comparison to measurements taken at operating reactors for many years at many reactors.

Accordingly, please address the following:

(a) The ATRIUM I OXM fuel design involves changes in fuel density, part length rods, active fuel length, diameter and length of the fuel pellet, and cladding outer diameter from the ATRIUM 10 fuel design. Please explain how these changes are accounted for in the methodology and justify the statement that "these changes in fuel design have no impact on the applicability of the methodology."

(b) Approval ofEMF-2158(P), Revision 0, was subjected to six conditions as stipulated in the safety evaluation (Section 6.0) attached to the letter from the NRC (Stephen Dembek) to Siemens Power Corporation (James Mallay), dated October 18, 1999. Please describe how these conditions are satisfied where the methodology of EMF-2158(P) is applied to ATRIUM IOXM fuel design.

Response No. 15 (a) Parameters such as fuel density, part length rods, active fuel length, fuel pellet diameter, and fuel cladding diameter are all inputs to the methodology. The methodology explicitly accounts for such changes in design parameters. The changes in these parameters for the A1OXM fuel are insignificant relative to the changes that have been included in the validation of the methodology that demonstrate the methodology's capability to evaluate these parameters. Fuel designs including 7x7, 8x8, 9x9 and l0xl0 with corresponding changes in pellet and cladding diameters were presented in the topical report, EMF-2158(P)(A), Siemens Power CorporationMethodologyfor Boiling Water Reactors:

Evaluation and Validation of CASMO-4/MICROBURN-B2.

(b) There are six restrictions on the application of the CASMO-4/MICROBURN-B2 code system imposed by the NRC during their review ofEMF-2158(P). They are listed below with the demonstration of compliance for BSEP Unit 2 Cycle 20:

1. The CASMO-4/MICROBURN-B2 code system shall be applied in a manner that predicted results are within the range of the validation criteria and measurement uncertainties presented in EMF-2158(P).

BSEP 10-0133 Enclosure 3 Page 13 of 44 The application of the CASMO-4/MICROBURN-B2 code system for BSEP involves the following models:

• Control blade B-10 depletion

• Explicit neutronic treatment of the spacer grids

" Explicit water rod hydraulics

• Explicit neutronic treatment of the plenum above the PLFRs The application of these models has been evaluated for the BSEP benchmark and AREVA cycles relative to the EMF-2158(P)(A) statistics. The results of the statistical analysis demonstrated that these models meet the approved methodology validation criteria and measurement uncertainties presented in EMF-2158(P)(A).

2. The CASMO-4/MICROBURN-B2 code system shall be validated for analyses of any new fuel design which departs from current orthogonal lattice designs and/or exceed gadolinia and U-235 enrichment limits.

The A0OXM fuel has an orthogonal lattice design and the nuclear design does not exceed the gadolinia concentration or U-235 enrichment limits.

3. The CASMO-4/MICROBURN-B2 code system shall only be used for BWR licensing analyses and BWR core monitoring applications.

BSEP is a General Electric BWR-4; therefore, BSEP analyses performed with this code system meet this SER restriction.

4. The review of the CASMO-4/MICBROBURN-B2 code system should not be construed as a generic review of the CASMO-4 or MICROBURN-B2 computer codes.

CASMO-4 and MICROBURN-B2 are being used in a manner consistent with the original approval.

5. The CASMO-4/MICROBURN-B2 code system is approved as a replacement for the CASMO-3/MICROBURN-B code system used in NRC-approved AREVA BWR licensing methodology and in AREVA BWR core monitoring applications. Such replacements shall be evaluated to ensure that each affected methodology continues to comply with its SER restrictions and/or conditions.

CASMO-4/MICROBURN-B2 does not violate the SER restrictions of separate methodologies used in the licensing analyses.

BSEP 10-0133 Enclosure 3 Page 14 of 44

6. AREVA shall notify any customer who proposes to use the CASMO-4/

MICROBURN-B2 code system independent of any AREVA fuel contract that conditions 1 through 4 above must be met. AREVA's notification shall provide positive evidence to the NRC that each customer has been informed by AREVA of the applicable conditions for using the code system.

This licensing analysis is being performed by AREVA through a fuel contract between AREVA and CP&L. CP&L has been provided a copy of the topical report which includes the SER and all clarifications. AREVA has conducted training sessions with CP&L personnel to assure that they are familiar with the application of the methodology.

NRC Question No. 16 BSEP 10-0057, BAW-10247(P)(A)

NRC approval of topical report BAW-1 0247(P) was subject to compliance with five limitations and conditions which are listed in the safety evaluation for the topical report attached to the letter from the NRC (Ho Nieh) to AREVA Nuclear Power (Ronnie Gardner), dated February 12, 2008.

Accordingly, please address the following:

(a) Provide documentation for how BSEP is planning to comply with Limitation Number I regarding the fission gas release (FGR) model in BAW-10247(P)(A).

(b) In response to Limitation and Condition 3, the licensee indicated that the "BSEP core operating limits analyses performed with RODEX4 will not use the hydrogen pickup model within RODEX4." The NRC staff requests that the licensee explain the details of the model and its usage to assess the hydrogen pick up at BSEP.

(c) In response to Limitation and Condition 4, the licensee indicated that the "BSEP core operating limits analyses performed with the BAW-10247PA, Revision 0 methodology will use the specific values of the equation constants and tuning parameters derived in Topical Report BAW-10247(P), Revision 0 (as updated by RAI responses)." Justify the applicability of the validation process, the specific values of the equation constants, and tuning parameters derived in BAW-10247(P) to the ATRIUM 1OXM fuel design.

(d) The licensee response to Limitation and Condition 5 of topical report BAW- 10247(P) indicated that "where we have plant specific measurements indicating abnormal crud, our analyses for the plant will be based on the plant specific data" (Page 4 of 4 of the proprietary file "Initial RODEX4 Draft Comments.doc," attached to the email dated November 2, 2007, from J. Holm (AREVA) to Holly Cruz (NRC)). Provide the plant specific data for abnormal crud or corrosion layer BSEP.

BSEP 10-0133 Enclosure 3 Page 15 of 44 (e) Page 7 of 18 of BSEP 10-0057 indicates that the BSEP inspections of "irradiated GE14 fuel found to be clean with no evidence of tenacious crud." The licensee further stated that "low crud levels were similarly noted during recently completed inspections of irradiated ATRIUM-10 fuel operated in BSEP Unit 1, Cycle 17." Provide supporting documents to show that the ATRIUM fuel designs will not have design basis crud formation at BSEP.

Response No. 16 (a) Limitation Number I precludes the use of a grain size input larger than 20 micrometers 3-D in the case when the as-manufactured grain size exceeds 20 micrometers 3-D.

AREVA has proceduralized the inputs to RODEX4 by code automation. The code automation prepares the input file for RODEX4 and includes logic to ensure the correct input of the grain size. An excerpt from a RODEX4 input file used in the fuel rod analyses for the BSEP A1OXM design is provided below.

[II

_] The code automation and inputs are prepared and verified according to the AREVA Quality Assurance Program.

(b) No hydrogen pickup model was applied to the AIOXM fuel at BSEP. An approved hydrogen concentration limit is currently not required in the licensing bases for the plant.

(c) BAW- I 0247PA, Revision 0, Realistic Thermal-MechanicalFuel Rod Methodologyfor Boiling Water Reactors, was written to specifically cover the AREVA fuel manufactured in Richland, Washington. Section 2.2.4 of BAW-10247PA identifies the validation ranges and Section 4.4 provides additional information on the validation ranges of the models.

The model calibrations were made to generically cover the fuel pellets along with specific calibration to the AREVA BWRA1O fuel rod cladding. The A1OXM fuel rod design makes use of the standard AREVA BWR [I

)) The fuel rod cladding for the A1OXM design is manufactured using the same process and product specifications as used for the A1O fuel design. The fuel pellets for the A1OXM design are manufactured in Richland by the same [I )) used for a number of years in the production of the A10 fuel design. Thus, the model parameters are directly applicable to the A1OXM fuel rod design.

BSEP 10-0133 Enclosure 3 Page 16 of 44 (d) Please see the RAI 16(e) response below regarding crud levels at the BSEP units. The data summarized in the 16(e) response indicate normal crud and normal corrosion layers.

(e) The fuel rod thermal-mechanical analyses with RODEX4 make use of the corrosion model fitting parameter and uncertainty range as given in the Response 17c of the first round of RAI questions for BAW-10247PA, Revision 0. The corrosion model parameter and uncertainty are based on the liftoff data provided in the response, which are from plants with normal, low levels of crud in addition to oxidation. A poolside examination of Al0 fuel was conducted in March 2010 in BSEP Unit 1 during the Cycle 18 refueling outage. The maximum mid-span measurements are plotted in Figures 16.1 and 16.2 below for comparison to the baseline data used to derive the corrosion model parameter and uncertainties, as presented in BAW-10247PA, Revision 0.

BSEP 10-0133 Enclosure 3 Page 17 of 44

((

II

(( 1I

BSEP 10-0133 Enclosure 3 Page 18 of 44

[I II II II II NRC Question No. 17 BSEP 10-0057, EMF-2158(P)(A)

(a) The licensee used topical report EMF-2158(P)(A) to calculate the radial and axial power distribution measurement uncertainties listed on Pages 3 and 4 of 18 of BSEP 10-0057, Enclosure 1. Please provide detailed analyses, calculations, and the database information used to establish the uncertainties listed in the two tables (unnumbered) on Page 4 of 18 of Enclosure 1 of BSEP 10-0057, and as illustrated in Figures 1, 2, and 3 on Pages 11 through 13 of Enclosure I of BSEP 10-0057.

(b) Please discuss the applicability of the BSEP TIP measurement process and database to the ATRIUM IOXM fuel design.

BSEP 10-0133 Enclosure 3 Page 19 of 44 Response No. 17 Response 17 (a)

As stated on Pages 9-1 of Topical Report EMF-2158(P)(A), AREVA methodology for measuring the power distribution in a BWR reactor and the procedure by which the uncertainty associated with the measurement of a BWR power distribution would be determined was originally described in XN-NF-80-19(P)(A), Vol. 1, Supplement 3 & 4 (November 1990), Benchmark Results for the CASMO-3G/MICROBURN-B CalculationMethodology. As stated on Page 9-3 of EMF-2158(P)(A), XN-NF-80-19(P)(A) provides the detailed derivation of the constitutive relationships for the uncertainty components. Section 5 of XN-NF-80-19(P)(A), Volume 1, Supplement 3, hereafter XN-NF-80-19 in this response, and Section 9 of EMF-2158(P)(A),

hereafter EMF-2158 in this response, together provide a very detailed description of the analyses and calculations performed to determine the TIP uncertainty components. For this reason, the requested description of the detailed analyses and calculations performed to determine the TIP uncertainty components listed in the two tables (i.e., unnumbered) on Page 4 of 18 of Enclosure I of BSEP 10-0057 is provided in Table 17.1 below, via reference to these NRC approved topical reports.

Figures 1, 2, and 3 on Pages 11 through 13 of Enclosure I of CP&L letter BSEP 10-0057 dated April 29, 2010, present 177 database points. Each database point is calculated using a Traversing Incore Probe (TIP) flux map consisting of measurements obtained from 21 axial levels at 31 radial core locations. Except for the size of the data population, the detailed equations provided by reference in Table 17.1 below are the same for both the database points and the final uncertainty components. Each database point is based on 651 local TIP readings (i.e., 21 times 3 1) obtained at a core operating state characterized by its core power, core void fraction, and core power to flow ratio, whereas the TIP uncertainty components are based on 115,227 local TIP readings (i.e., 651 local TIP readings times 177 flux maps).

The database information used to establish the uncertainties listed in CP&L letter BSEP 10-0057 consists of the 177 TIP flux maps and the corresponding MICROBURN-B2 (MB2) calculated TIP readings. The 177 TIP flux maps were obtained from BSEP Units I and 2 core operating states from March of 2000 through February of 2010, from cores consisting entirely of 9x9 GEI 3 fuel, mixed cores of GE13 and lOxi0 GE14 fuel, GE14 fuel alone, and mixed cores of GE14 and lOxIO ATRIUM-10 fuel.

The BSEP MB2 benchmark completed by AREVA to incorporate explicit water rod, part length fuel rod (PLFR) plenum, and spacer model options was not available at the time the TIP statistics presented in CP&L letter BSEP 10-0057 were calculated; however, none of these changes materially affect MB2 calculated TIP distributions. CP&L has since recalculated TIP statistics based on the latest AREVA benchmark. The database values presented in CP&L letter BSEP 10-0057 that are dependent on MB2 calculated TIP response (i.e., see Note 2 below Table 17.1) are plotted against the values calculated based on the latest AREVA benchmark in Figure 17.1. As shown on Figure 17.1, the linear regression slope and intercept for each component are essentially 1 and 0, respectively. This result demonstrates the explicit water rod,

BSEP 10-0133 Enclosure 3 Page 20 of 44 PLFR plenum and spacer model options have no material impact on the TIP statistics, or the database trends presented in Figures 1,2 and 3 on Pages 11 through 13 of Enclosure I of CP&L letter BSEP 10-0057.

TIP uncertainty component values and trends based on the latest MB2 benchmark (i.e.,

incorporating explicit water rod, PLFR plenum, and spacer model options) are provided as Table 17.2 and Figures 17.2, 17.3, and 17.4. The database consisting of the 177 values for each of the eight uncertainty components tabulated in CP&L letter BSEP 10-0057, recalculated based on this latest benchmark, is provided as Table 17.3.

The proprietary D-Lattice uncertainty component values identified in Sections 9.4 and 9.5 of EMF-2158(P)(A) bound the BSEP-specific uncertainty component values shown in Table 17.2.

This result demonstrates uncertainties applied by the BAW-1 0247PA methodology and determined in accordance with the EMF-2158(P)(A) methodology are applicable to BSEP Units 1 and 2.

Response 17 (b)

Evaluation of the BSEP TIP database for previous cycles including both units has demonstrated that the uncertainties documented in EMF-2158(P)(A) for D-Lattice plants remain conservative.

None of the features of the ATRIUM 1OXM design will have any impact on the accuracy of the methodology to predict TIP responses.

Table 17.1: TIP Uncertainty Component Detailed Analyses and Calculations Component Detailed Analyses and Calculations 6

T mijk This component is nodal TIP measurement uncertainty. XN-NF-80-19, page 141 and following and EMF-2158, page 9-4 and following detail the calculation of this component.

Note the nomenclature for TIP measurement uncertainty (nodal, radial and planar) is 6T m in EMF-2158 and 6TIP in XN-NF-80-19.

Note also that EMF-2158 and XN-NF-80-19 define the formula for various uncertainty components in terms of a relative difference variable (i.e., "d"), which is not explicitly defined by a formula. For two compared values (e.g., a and b),

d is simply (a-b)/a.

5Tmij This component is radial TIP measurement uncertainty. XN-NF-80-19, page 141 and following, and EMF-2158, page 9-5 detail the calculation of this component.

BSEP 10-0133 Enclosure 3 Page 21 of 44 Table 17. 1: TIP Uncertainty Comnonent Detailed Analyses and Calculations Component Detailed Analyses and Calculations 6Tmplanar This component is planar TIP measurement uncertainty (i.e., see also Note I below this Table 17.1). EMF-2158, page 9-5 details the calculation of this component. The paragraph below Equation 5.24 on XN-NF-80-19, page 142 further details the calculation of this component, and also generally describes the calculation of a planar deviation.

6 T'ijk This component is the nodal deviation between measured TIP readings and the corresponding TIP readings calculated by MB2. EMF-2158, Equation 9-20 details the calculation of this component.

Note all the values for this component (i.e., the nodal, radial and planar values) include both TIP measurement uncertainty and MB2 calculated TIP uncertainty.

6 T'ij This component is the radial deviation between measured TIP readings and the corresponding TIP readings calculated by MB2. EMF-2158, Equation 9-22 details the calculation of this component.

(6T'planar This component is the planar deviation between measured TIP readings and the corresponding TIP readings calculated by MB2. An explicit equation for this component is not provided by EMF-2158 or XN-NF-80-19; however, (6T'planar is calculated using the same equation as that for 6 T'ijk, after normalization of the data by plane (i.e., see also Note 1 below this Table 17.1).

6 Sijk This component is nodal synthesis procedure uncertainty. Synthesis procedure uncertainty is described in the last paragraph on page 143 of XN-NF-80-19.

XN-NF-80-19, Equation 5.25 details the calculation of this component. See also Note 2 below this Table 17.1.

6 Sij This component is radial synthesis procedure uncertainty. Synthesis procedure uncertainty is described in the last paragraph on page 143 of XN-NF-80-19.

XN-NF-80-19, Equation 5.26 details the calculation of this component. See also Note 2 below this Table 17.1.

6 Tijk This component is net nodal calculated TIP distribution uncertainty. It is determined by ((

R]

BSEP 10-0133 Enclosure 3 Page 22 of 44 Table 17.1: TIP Uncertainty Component Detailed Analyses and Calculations Component Detailed Analyses and Calculations

• Tij This component is net radial calculated TIP distribution uncertainty. It is determined by 8

Tplanar This component is net planar calculated TIP distribution uncertainty. It is determined by I.

Response 17(a) and Table 17.1 Notes:

1. The first paragraph of XN-NF-80-19, Section 5.2 provides additional information on application of planar uncertainties in XN-NF-80-19, Equations 5.1 and 5.2, which is also applicable to EMF-2158, Equations 9-5 and 9-12. The difference between a planar and nodal deviation is that the data to be compared are normalized to I by axial plane before calculation of a planar deviation. Otherwise, planar and nodal deviations are calculated using the same formula.

2. H I.

BSEP 10-0133 Enclosure 3 Page 23 of 44 Table 17.2: Updated TIP Component Uncertainties Component BSEP 10-0057 Value Latest EI'AF-2158 Benchmark MI32 Value Value (D -Lattice)

(5T m ijk 1.90% 1.90% [H

( Tmij 1.25% 1.25% 1]

(STmplanar 1.97% 1.97% [I 6T'ijk 4.47% 4.44% [1 2.07% 2.07% ((

(T'planar 2.58% 2.58% (( II (SSij 0.22% 0.21% ((

6 Sijk 1.79% 1.68% III I]

Calculation method [I

[1 ii-8Tij provided in BSEP 10-0057 instead of H1 ii-(5Tpianar proprietary value I- -Th Table 17.3: TIP Uncertainty Component Database Core Power/

Unit Date MWth Void Flow 6Tmij 6T m ijk 6Tm_p 6sij 6Sijk I 3/29/00 2553 0.501 37.7 0.82% 1.34% 1.35% 0.13% 1.55%

5/9/00 2560 0.491 37.5 1.14% 1.50% 1.50% 0.11% 1.62%

6/21/00 2560 0.498 38.8 1.03% 1.48% 1.49% 0.12% 1.58%

7/31/00 2558 0.469 34.7 1.07% 1.45% 1.45% 0.16% 1.69%

9/11/00 2555 0.477 36.2 1.08% 1.54% 1.54% 0.18% 1.72%

10/23/00 2555 0.487 37.1 0.84% 1.29% 1.30% 0.18% 1.69%

12/4/00 2555 0.450 34.0 0.89% 1.42% 1.43% 0.14% 1.69%

1/15/01 2555 0.466 35.7 0.57% 1.47% 1.48% 0.20% 1.90%

2/26/01 2554 0.468 35.3 1.02% 1.38% 1.38% 0.14% 1.85%

4/2/01 2555 0.481 36.8 0.59% 0.98% 1.00% 0.15% 1.70%

5/14/01 2563 0.480 37.6 0.80% 1.39% 1.41% 0.10% 1.66%

6/25/01 2559 0.461 36.9 0.82% 1.31% 1.41% 0.13% 1.57%

8/6/01 2555 0.436 36.0 0.70% 1.23% 1.28% 0.12% 1.51%

9/17/01 2554 0.404 33.6 1.29% 1.68% 1.76% 0.15% 1.52%

10/29/01 2552 0.431 34.7 0.97% 1.36% 1.36% 0.17% 1.47%

BSEP 10-0133 Enclosure 3 Page 24 of 44 Table 17.3: TIP Uncertainty Component Database Core Power!

Unit Date MWth Void Flow WT Wo5Tik 6T m~p 6Sij 6Sijk 1 12/12/01 2554 0.405 33.6 1.11% 1.37% 1.37% 0.15% 1.34%

1 1/14/02 2529 0.385 31.5 0.61% 1.30% 1.30% 0.15% 1.62%

1 6/10/02 2734 0.508 35.2 1.27% 2.13% 2.33% 0.28% 1.76%

1 6/25/02 2759 0.510 35.7 1.65% 2.38% 2.51% 0.24% 1.75%

1 8/7/02 2757 0.515 37.1 1.54% 2.04% 2.24% 0.25% 1.79%

1 9/19/02 2754 0.5 13 37.2 1.78% 2.22% 2.22% 0.25% 1.83%

1 10/4/02 2750 0.487 35.3 1.43% 1.90% 2.08% 0.24% 1.91%

1 10/16/02 2754 0.483 35.4 1.05% 1.65% 2.04% 0.25% 1.88%

1 11/20/02 2754 0.487 36.0 1.43% 1.97% 2.24% 0.23% 1.76%

1 12/17/02 2755 0.492 35.8 1.32% 1.92% 2.16% 0.24% 1.69%

1 2/4/03 2753 0.514 37.6 1.23% 1.60% 1.60% 0.25% 1.79%

1 3/6/03 2748 0.515 37.9 1.84% 2.22% 2.55% 0.30% 1.67%

1 4/9/03 2758 0.510 37.6 2.10% 2.59% 2.92% 0.25% 1.59%

1 5/15/03 2739 0.482 36.0 2.19% 2.65% 2.94% 0.26% 1.57%

1 6/18/03 2739 0.466 35.0 2.13% 2.68% 2.92% 0.26% 1.54%

1 7/23/03 2740 0.469 37.0 1.95% 2.47% 2.62% 0.19% 1.66%

1 8/27/03 2742 0.468 37.7 2.05% 2.68% 2.78% 0.22% 1.54%

1 10/2/03 2739 0.469 37.2 1.21% 2.17% 2.32% 0.22% 1.46%

1 11/5/03 2734 0.434 34.3 2.04% 2.79% 2.84% 0.20% 1.47%

1 12/18/03 2723 0.417 34.0 2.18% 2.81% 2.86% 0.20% 1.49%

1 1/27/04 2716 0.368 33.8 1.25% 1.92% 2.01% 0.18% 2.26%

1 5/19/04 2919 0.489 37.4 1.10% 1.95% 2.00% 0.23% 1.66%

1 6/23/04 2920 0.494 37.3 0.81% 1.84% 1.84% 0.24% 1.85%

1 8/3/04 2918 0.488 36.9 1.10% 2.04% 2.04% 0.28% 1.97%

1 9/9/04 2925 0.493 37.0 0.90% 1.79% 1.82% 0.22% 2.03%

1 10/6/04 2923 0.493 38.1 0.90% 1.67% 1.71% 0.23% 2.03%

1 11/17/04 2926 0.494 37.9 1.44% 2.06% 2.06% 0.23% 1.93%

1 12/15/04 2921 0.492 37.6 1.16% 1.86% 1.86% 0.30% 2.08%

1 1/19/05 2919 0.485 36.8 1.08% 2.14% 2.14% 0.23% 2.17%

1 2/24/05 2922 0.476 37.1 1.37% 2.36% 2.62% 0.23% 2.22%

1 3/30/05 2926 0.470 37.1 1.65% 2.51% 2.51% 0.34% 2.21%

1 4/28/05 2907 0.461 37.1 0.93% 1.92% 1.92% 0.25% 2.15%

BSEP 10-0133 Enclosure 3 Page 25 of 44 Table 17.3: TIP Uncertainty Component Database Core Power/

m ijk Unit Date MWth Void Flow 6TTmi Mij 6TT 6Tm_p 5Sij 8Sijk 6/8/05 2923 0.482 36.8 0.95% 1.69% 1.72% 0.25% 2.19%

7/27/05 2922 0.488 37.2 1.02% 1.91% 2.07% 0.21% 2.25%

1 8/17/05 2918 0.484 37.2 1.33% 2.04% 2.12% 0.23% 1.93%

.1 9/21/05 2919 0.466 36.9 1.01% 1.73% 1.79% 0.24% 1.89%

I 10/26/05 2921 0.495 36.9 1.34% 2.19% 2.20% 0.20% 2.16%

11/29/05 2917 0.468 36.8 1.06% 2.26% 2.41% 0.20% 2.11%

1/4/06 2922 0.477 37.1 0.87% 1.63% 1.91% 0.16% 1.60%

2/8/06 2907 0.447 36.1 0.82% 1.55% 1.74% 0.17% 1.86%

5/16/06 2913 0.507 37.2 0.84% 1.18% 1.40% 0.18% 1.19%

6/22/06 2925 0.508 37.9 1.23% 1.50% 1.50% 0.14% 1.04%

7/25/06 2921 0.509 38.0 1.37% 1.78% 1.78% 0.15% 1.07%

8/31/06 2921 0.506 38.0 1.25% 1.94% 1.99% 0.22% 1.15%

10/3/06 2920 0.501 36.7 1.13% 1.46% 1.46% 0.20% 1.12%

11/11/06 2920 0.499 36.8 1.39% 1.91% 1.92% 0.18% 1.05%

12/14/06 2913 0.496 37.5 1.13% 1.74% 1.74% 0.23% 1.22%

1/17/07 2921 0.494 37.3 1.60% 1.89% 1.89% 0.20% 1.20%

2/21/07 2915 0.494 37.3 1.21% 1.69% 1.70% 0.16% 1.19%

1 3/28/07 2914 0.494 1.90%

37.4 1.39% 1.87% 0.18% 1.26%

1 5/2/07 2917 0.482 1.42%

36.7 1.72% 1.76% 0.17% 1.18%

1 6/7/07 2920 0.479 36.9 1.66% 1.97% 2.10% 0.16% 1.06%

I 6/19/07 2917 0.473 37.4 1.81% 2.09% 2.20% 0.13% 1.10%

7/24/07 2918 0.466 37.2 1.48% 1.93% 1.97% 0.15% 1.14%

8/28/07 2920 0.452 36.9 1.79% 2.11% 2.18% 0.14% 1.38%

10/2/07 2918 0.474 36.8 2.71% 3.20% 3.20% 0.17% 1.29%

11/7/07 2916 0.465 37.1 2.41% 2.93% 2.92% 0.18% 1.36%

11/14/07 2917 0.460 36.9 2.44% 2.88% 2.88% 0.20% 1.44%

12/19/07 2910 0.484 36.2 2.56% 2.88% 2.89% 0.14% 1.15%

1/15/08 2890 0.484 35.9 2.67% 3.04% 3.08% 0.12% 1.09%

2/13/08 2870 0.446 35.8 2.42% 2.85% 2.86% 0.20% 1.40%

2/20/08 2853 0.436 35.5 2.06% 2.56% 2.57% 0.18% 1.42%

5/3/08 2914 0.502 36.5 0.92% 1.59% 1.66% 0.26% 1.48%

6/5/08 2927 0.490 37.3 0.98% 1.66% 1.72% 0.29% 1.37%

BSEP 10-0133 Enclosure 3 Page 26 of 44 Table 17.3: TIP Uncertainty Component Database Core Power/

Unit Date MWth Void Flow 6T m ij 6T m ijk 6Tnp 6Sij 6Sijk 1 7/10/08 2919 0.484 37.3 0.98% 1.64% 1.71% 0.24% 1.34%

1 8/13/08 2915 0.487 36.8 1.24% 1.90% 1.94% 0.25% 1.55%

1 9/17/08 2923 0.482 36.7 1.21% 2.08% 2.21% 0.18% 1.62%

1 10/22/08 2924 0.480 37.3 1.00% 1.92% 1.92% 0.27% 1.70%

1 12/10/08 2917 0.484 37.4 0.97% 1.69% 1.72% 0.22% 1.73%

1 1/15/09' 2920 0.481 37.8 0.95% 1.84% 2.00% 0.25% 1.83%

1 2/18/09 2917 0.482 37.4 1.01% 2.03% 2.54% 0.26% 1.97%

1 3/25/09 2917 0.487 37.2 1.32% 1.90% 1.90% 0.28% 2.06%

1 4/28/09 2917 0.494 37.8 1.12% 1.76% 1.76% 0.28% 2.12%

1 6/5/09 2915 0.477 37.3 0.81% 1.70% 1.87% 0.27% 2.08%

1 7/16/09 2918 0.477 37.6 1.02% 2.17% 2.43% 0.34% 1.89%

1 8/12/09 2916 0.470 36.8 1.02% 1.92% 2.21% 0.28% 1.75%

1 9/16/09 2918 0.475 37.0 1.23% 1.94% 2.30% 0.31% 1.54%

1 11/4/09 2917 0.456 36.9 1.19% 1.78% 2.20% 0.23% 1.41%

1 12/9/09 2921 0.469 36.8 1.08% 2.03% 2.17% 0.19% 1.21%

1 1/13/10 2915 0.473 36.6 1.15% 1.76% 1.79% 0.17% 1.09%

1 2/17/10 2844 0.424 35.5 1.35% 1.92% 1.99% 0.20% 1.33%

2 4/4/01 2550 0.488 33.5 1.41% 2.15% 2.18% 0.19% 1.61%

2 5/8/01 2558 0.479 33.5 1.58% 2.29% 2.29% 0.18% 1.58%

2 6/19/01 2558 0.485 34.9 1.10% 1.99% 2.03% 0.18% 1.63%

2 7/31/01 2557 0.488 34.9 1.08% 1.96% 2.02% 0.26% 1.65%

2 9/11/01 2551 0.464 32.6 1.42% 2.25% 2.27% 0.24% 1.84%

2 10/23/01 2562 0.469 34.4 1.28% 2.23% 2.26% 0.23% 1.81%

2 12/4/01 2554 0.481 37.2 1.25% 2.21% 2.23% 0.25% 1.85%

2 1/15/02 2558 0.468 33.9 1.08% 2.13% 2.14% 0.31% 1.95%

2 2/26/02 2557 0.468 35.4 1.04% 2.25% 2.28% 0.19% 1.72%

2 4/9/02 2556 0.474 37.7 0.88% 1.83% 1.84% 0.20% 1.63%

2 5/21/02 2561 0.476 37.8 0.89% 1.94% 1.98% 0.23% 1.58%

2 7/2/02 2561 0.466 35.0 1.05% 1.88% 1.95% 0.21% 1.53%

2 8/13/02 2553 0.439 33.2 0.98% 1.94% 1.96% 0.18% 1.59%

2 9/25/02 2555 0.453 37.2 0.89% 1.72% 1.72% 0.21% 1.63%

2 11/11/02 2559 0.410 34.2 1.10% 1.91% 1.91% 0.22% 1.73%

BSEP 10-0133 Enclosure 3 Page 27 of 44 Table 17.3: TIP Uncertainty Component Database Core Power!

Unit Date MWth Void Flow 6Tmij 6T m ijk 6T m~p 6Si6 6Sijk 2 12/18/02 2557 0.403 34.1 1.01% 1.94% 1.95% 0.19% 1.79%

2 1/30/03 2549 0.376 33.1 1.03% 1.88% 1.88% 0.17% 1.94%

2 5/21/03 2825 0.492 37.4 0.85% 1.98% 2.07% 0.25% 1.55%

2 7/2/03 2817 0.484 37.8 1.10% 2.05% 2.05% 0.21% 1.60%

2 8/13/03 2826 0.487 38.5 1.10% 2.06% 2.06% 0.26% 1.53%

2 9/15/03 2823 0.490 37.6 1.00% 2.03% 2.08% 0.27% 1.52%

2 10/22/03 2820 0.488 37.2 0.90% 1.99% 2.01% 0.26% 1.59%

2 11/26/03 2830 0.489 36.6 0.98% 2.22% 2.22% 0.28% 1.68%

2 1/9/04 2823 0.492 36.6 2.04% 3.24% 3.40% 0.30% 1.99%

2 2/12/04 2822 0.483 37.1 1.18% 1.93% 2.07% 0.33% 2.56%

2 3/10/04 2820 0.486 38.3 1.07% 1.83% 1.89% 0.32% 2.40%

2 4/14/04 2817 0.484 37.6 0.89% 1.85% 1.93% 0.33% 2.45%

2 5/20/04 2827 0.474 37.0 1.04% 1.89% 1.89% 0.28% 2.22%

2 6/18/04 2818 0.465 36.1 0.79% 1.67% 1.68% 0.28% 2.14%

2 7/21/04 2827 0.464 36.3 1.06% 1.94% 1.95% 0.32% 2.28%

2 8/25/04 2806 0.462 36.7 1.28% 2.02% 2.07% 0.23% 2.30%

2 10/5/04 2815 0.448 36.5 1.42% 2.18% 2.20% 0.24% 2.16%

2 11/3/04 2828 0.469 36.6 1.22% 1.94% 1.98% 0.20% 1.99%

2 1/12/05 2805 0.423 34.9 1.36% 2.56% 2.57% 0.18% 2.14%

2 5/25/05 2921 0.506 37.8 0.89% 1.33% 1.50% 0.18% 1.23%

2 7/7/05 2928 0.493 37.7 0.97% 1.46% 1.46% 0.18% 1.18%

2 8/3/05 2913 0.493 37.8 0.85% 1.30% 1.30% 0.15% 1.11%

2 9/7/05 2918 0.487 36.7 1.06% 1.74% 1.76% 0.20% 1.29%

2 10/12/05 2928 0.489 37.6 0.67% 1.41% 1.41% 0.19% 1.22%

2 12/21/05 2913 0.494 37.2 0.87% 1.30% 1.30% 0.22% 1.31%

2 1/25/06 2915 0.487 37.8 0.75% 1.66% 1.66% 0.20% 1.43%

2 3/3/06 2923 0.489 37.5 1.18% 1.67% 1.67% 0.24% 1.44%

2 4/4/06 2915 0.482 37.8 1.13% 1.80% 1.80% 0.22% 1.36%

2 5/10/06 2917 0.481 37.9 1.19% 1.73% 1.73% 0.19% 1.25%

2 6/14/06 2888 0.498 38.1 1.28% 1.76% 1.76% 0.20% 1.38%

2 7/19/06 2922 0.487 37.7 1.33% 1.82% 1.82% 0.19% 1.12%

2 8/30/06 2915 0.494 37.7 0.83% 1.50% 1.57% 0.19% 1.17%

BSEP 10-0133 Enclosure 3 Page 28 of 44 Table 17.3: TIP Uncertainty Component Database Core Power!

Unit Date MWth Void Flow 6T Mij 5T m ijk 6T m _p BSij 6Sijk 2 10/4/06 2923 0.479 36.8 1.57% 1.98% 2.08% 0.20% 1.23%

2 11/1/06 2920 0.508 37.6 1.14% 1.79% 1.79% 0.15% 1.06%

2 12/6/06 2917 0.494 37.2 0.95% 1.84% 1.88% 0.15% 1.03%

2 1/10/07 2915 0.487 36.9 1.03% 1.47% 1.55% 0.17% 1.22%

2 2/15/07 2907 0.458 36.4 1.11% 1.80% 1.81% 0.16% 1.53%

2 4/23/07 2914 0.518 37.5 0.98% 1.35% 1.54% 0.15% 1.78%

2 5/30/07 2926 0.515 37.8 1.00% 1.39% 1.50% 0.14% 1.73%

2 7/3/07 2915 0.483 37.6 0.75% 1.54% 1.59% 0.19% 1.65%

2 8/9/07 2912 0.480 37.1 1.18% 1.72% 1.80% 0.21% 1.72%

2 9/12/07 2915 0.480 37.9 1.13% 1.56% 1.57%. 0.13% 1.56%

2 10/17/07 2911 0.491 37.4 0.87% 1.27% 1.30% 0.19% 1.61%

2 11/20/07 2906 0.498 36.6 0.90% 1.59% 1.59% 0.18% 1.58%

2 12/26/07 2917 0.496 37.6 1.23% 1.69% 1.69% 0.15% 1.47%

2 2/1/08 2917 0.506 37.5 0.74% 1.21% 1.29% 0.22% 1.77%

2 3/12/08 2918 0.507 38.0 0.76% 1.17% 1.35% 0.20% 1.88%

2 4/9/08 2922 0.494 37.1 1.102% 1.50% 1.50% 0.20% 1.63%

2 5/14/08, 2915 0.492 37.8 0.88% 1.25% 1.25% 0.20% 1.67%

2 6/26/08 2914 0.476 37.5 0.84% 1.17% 1.20% 0.18% 1.56%

2 7/23/08 2913 0.466 37.3 0.80% 1.18% 1.27% 0.17% 1.49%

2 8/27/08 2919 0.453 37.3 0.70% 1.37% 1.53% 0.15% 1.78%

2 10/8/08 2917 0.469 37.2 0.66% 1.20% 1.30% 0.18% 1.70%

2 11/21/08 2917 0.493 37.4 0.63% 1.32% 1.44% 0.16% 1.53%

2 12/30/08 2905 0.469 37.6 0.62% 1.37% 1.51% 0.16% 1.62%

2 1/28/09 2921 0.472 37.3 0.58% 1.05% 1.22% 0.10% 1.54%

2 2/19/09 2906 0.466 36.7 0.89% 1.31% 1.61% 0.12% 1.64%

2 5/5/09 2902 0.498 36.8 1.01% 1.77% 1.81% 0.24% 2.03%

2 6/17/09 2930 0.491 37.3 0.85% 1.58% 1.63% 0.22% 1.86%

2 7/22/09 2912 0.487 37.5 0.83% 1.64% 1.65% 0.20% 1.87%

2 8/26/09 2915 0.493 37.5 0.70% 1.74% 1.77% 0.20% 1.88%

2 10/14/09 2915 0.488 37.0 0.70% 1.65% 1.71% 0.20% 1.87%

2 11/18/09 2921 0.494 37.2 1.04% 1.60% 1.62% 0.20% 1.35%

2 12/16/09 2926 0.491 37.1 0.89% 1.59% 1.62% 0.24% 1.98%

BSEP 10-0133 Enclosure 3 Page 29 of 44 Table 17.3: TIP Uncertainty Component Database Core Power/

Unit Date MWth Void Flow 6T m ij 5T m ijk 6Tm p 6Sij 6Sijk 2 1/20/10 2916 0.485 37.6 1.03% 2.04% 2.06% 0.16% 1.91%

2 2/24/10 2920 0.487 37.7 1.00% 1.93% 1.95% 0.19% 1.72%

Table 17.3: TIP Uncertainty Component Database Core Power/

Unit Date MWth Void Flow 5T'ij 6T'ijk 6T'_p I 3/29/00 2553 0.501 37.7 2.18% 3.77% 2.59%

I 5/9/00 2560 0.491 37.5 2.49% 3.76% 2.87%

6/21/00 2560 0.498 38.8 2.31% 3.37% 2.69%

7/31/00 2558 0.469 34.7 2.43% 3.89% 2.82%

9/11/00 2555 0.477 36.2 2.20% 3.95% 2.55%

10/23/00 2555 0.487 37.1 2.05% 4.91% 2.58%

12/4/00 2555 0.450 34.0 2.01% 6.31% 2.32%

1/15/01 2555 0.466 35.7 2.20% 6.16% 2.59%

1 2/26/01 2554 0.468 35.3 2.55% 6.16% 2.65%

4/2/01 2555 0.481 36.8 2.18% 4.79% 2.29%

I 5/14/01 2563 0.480 37.6 2.32% 4.21% 2.55%

1 6/25/01 2559 0.461 36.9 2.35% 4.00% 2.59%

8/6/01 2555 0.436 36.0 2.13% 3.84% 2.44%

9/17/01 2554 0.404 33.6 2.08% 3.69% 2.45%

I 10/29/01 2552 0.431 34.7 1.43% 3.10% 1.95%

12/12/01 2554 0.405 33.6 1.36% 2.72% 1.76%

1/14/02 2529 0.385 31.5 1.32% 2.86% 1.92%

6/10/02 2734 0.508 35.2 1.71% 3.99% 2.27%

6/25/02 2759 0.510 35.7 1.81% 3.91% 2.40%

8/7/02 2757 0.515 37.1 1.78% 3.71% 2.20%

9/19/02 2754 0.513 37.2 1.72% 4.02% 2.17%

10/4/02 2750 0.487 35.3 1.93% 4.33% 2.37%

10/16/02 2754 0.483 35.4 1.95% 4.64% 2.35%

11/20/02 2754 0.487 36.0 1.91% 4.64% 2.30%

BSEP 10-0133 Enclosure 3 Page 30 of 44 Table 17.3: TIP Uncertainty Component Database Core Power/

Unit Date MWth Void Flow 6T'ij 6T'ijk 6T'_p I 12/17/02 2755 0.492 35.8 1.99% 4.80% 2.42%

I 2/4/03 2753 0.514 37.6 2.39% 4.16% 2.48%

1 3/6/03 2748 0.515 37.9 2.20% 4.07% 2.58%

1 4/9/03 2758 0.510 37.6 2.57% 3.82% 2.94%

I 5/15/03 2739 0.482 36.0 2.67% 3.80% 3.07%

1 6/18/03 2739 0.466 35.0 2.34% 3.64% 2.84%

7/23/03 2740 0.469 37.0 2.57% 4.24% 2.99%

1 8/27/03 2742 0.468 37.7 2.41% 4.540/ 3.00%

1 10/2/03 2739 0.469 37.2 2.41% 4.41% 3.07%

1 11/5/03 2734 0.434 34.3 2.41% 4.34% 3.13%

1 12/18/03 2723 0.417 34.0 2.39% 4.14% 3.10%

1 1/27/04 2716 0.368 33.8 2.30% 6.74% 3.35%

1 5/19/04 2919 0.489 37.4 1.51% 3.39% 2.17%

1 6/23/04 2920 0.494 37.3 1.51% 3.52% 2.47%

1 8/3/04 2918 0.488 36.9 1.46% 3.44% 2.35%

9/9/04 2925 0.493 37.0 1.39% 3.59% 2.12%

10/6/04 2923 0.493 38.1 1.44% 3.44% 2.07%

1 11/17/04 2926 0.494 37.9 1.38% 3.11% 1.99%

1 12/15/04 2921 0.492 37.6 1.56% 3.35% 2.14%

1 1/19/05 2919 0.485 36.8 1.56% 3.69% 2.34%

1 2/24/05 2922 0.476 37.1 1.82% 4.29% 2.52%

1 3/30/05 2926 0.470 37.1 1.81% 4.92% 2.44%

1 4/28/05 2907 0.461 37.1 1.47% 4.49% 2.21%

1 6/8/05 2923 0.482 36.8 1.40% 3.95% 2.05%

1 7/27/05 2922 0.488 37.2 1.59% 4.31% 2.17%

1 8/17/05 2918 0.484 37.2 1.50% 4.39% 2.08%

1 9/21/05 2919 0.466 36.9 1.39% 5.15% 2.14%

1 10/26/05 2921 0.495 36.9 1.83% 7.71% 2.52%

1 11/29/05 2917 0.468 36.8 1.93% 5.93% 2.90%

1/4/06 2922 0.477 37.1 1.82% 6.57% 2.26%

2/8/06 2907 0.447 36.1 1.61% 7.27% 2.16%

5/16/06 2913 0.507 37.2 1.53% 3.20% 1.83%

BSEP 10-0133 Enclosure 3 Page 31 of 44 Table 17.3: TIP Uncertainty Component Database Core Power/

Unit Date MWth Void Flow 6T'ij 6T'ijk 6T'_p 6/22/06 2925 0.508 37.9 1.50% 3.29% 1.77%

7/25/06 2921 0.509 38.0 1.42% 2.98% 1.74%

8/31/06 2921 0.506 38.0 1.64% 2.79% 1.93%

10/3/06 2920 0.501 36.7 1.42% 2.55% 1.74%

11/11/06 2920 0.499 36.8 1.59% 2.47% 1.83%

12/14/06 2913 0.496 37.5 1.29% 2.50% 1.69%

1/17/07 2921 0.494 37.3 2.09% 3.18% 2.29%

2/21/07 2915 0.494 37.3 2.18% 3.54% 2.38%

1 3/28/07 2914 0.494 37.4 2.08% 3.66% 2.35%

l 5/2/07 2917 0.482 36.7 1.89% 3.43% 2.15%

6/7/07 2920 0.479 36.9 2.27% 3.37% 2.56%

l 6/19/07 2917 0.473 37.4 2.40% 3.42% 2.71%

7/24/07 2918 0.466 37.2 2.01% 4.05% 2.53%

I 8/28/07 2920 0.452 36.9 2.16% 4.83% 2.72%

1 10/2/07 2918 .0.474 36.8 2.44% 4.93% 3.09%

11/7/07 2916 0.465 37.1 2.31% 4.74% 2.90%

11/14/07 2917 0.460 36.9 2.45% 4.89% 2.97%

I 12/19/07 2910 0.484 36.2 2.53% 5.75% 2.79%

1 1/15/08 2890 0.484 35.9 2.67% 5.96% 2.84%

1 2/13/08 2870 0.446 35.8 2.44% 7.05% 2.69%

2/20/08 2853 0.436 35.5 2.39% 7.52% 2.69%

5/3/08. 2914 0.502 36.5 2.42% 4.15% 3.21%

6/5/08 2927 0.490 37.3 2.51% 3.84% 3.17%

7/10/08 2919 0.484 37.3 2.37% 3.64% 3.00%

8/13/08 2915 0.487 36.8 2.34% 3.71% 2.87%

9/17/08 2923 0.482 36.7 2.46% 3.77% 2.99%

10/22/08 2924 0.480 37.3 2.28% 3.76% 2.81%

12/10/08 2917 0.484 37.4 2.53% 4.02% 3.04%

1/15/09 2920 0.481 37.8 2.40% 3.91% 2.95%

2/18/09 2917 0.482 37.4 3.01% 4.39% 3.50%

3/25/09 2917 0.487 37.2 2.75% 4.15% 3.23%

4/28/09 2917 0.494 37.8 2.78% 4.20% 3.27%

BSEP 10-0133 Enclosure 3 Page 32 of 44 Table 17.3: TIP Uncertainty Component Database Core Power/

Unit Date MWth Void Flow 6T'ij 6T'ijk 6T'_p 1 6/5/09 2915 0.477 37.3 2.74% 4.08% 3.24%

1 7/16/09 2918 0.477 37.6 2.94% 4.35% 3.42%

1 8/12/09 2916 0.470 36.8 3.05% 4.61% 3.59%

1 9/16/09 2918 0.475 37.0 2.79% 4.60% 3.32%

1 11/4/09 2917 0.456 36.9 2.32% 4.70% 2.87%

1 12/9/09 2921 0.469 36.8 2.63% 4.45% 2.98%

1 1/13/10 2915 0.473 36.6 2.68% 4.52% 3.03%

1 2/17/10 2844 0.424 35.5 2.98% 6.50% 3.55%

2 4/4/01 2550 0.488 33.5 2.58% 4.19% 2.98%

2 5/8/01 2558 0.479 33.5 2.55% 3.91% 3.00%

2 6/19/01 2558 0.485 34.9 2.24% 3.52% 2.68%

2 7/31/01 2557 0.488 34.9 2.20% 3.36% 2.71%

2 9/11/01 2551 0.464 32.6 2.11% 3.66% 2.60%

2 10/23/01 2562 0.469 34.4 2.02% 3.92% 2.54%

2 12/4/01 2554 0.481 37.2 1.90% 4.10% 2.48%

2 1/15/02 2558 0.468 33.9 1.88% 4.49% 2.48%

2 2/26/02 2557 0.468 35.4 2.03% 3.96% 2.69%

2 4/9/02 2556 0.474 37.7 1.93% 3.44% 2.58%

2 5/21/02 2561 0.476 37.8 2.21% 3.66% 2.88%

2 7/2/02 2561 0.466 35.0 2.22% 3.77% 2.83%

2 8/13/02 2553 0.439 33.2 2.28% 4.20% 2.86%

2 9/25/02 2555 0.453 37.2 2.22% 4.40% 2.86%

2 11/11/02 2559 0.410 34.2 2.28% 4.29% 2.89%

2 12/18/02 2557 0.403 34.1 2.10% 4.14% 2.73%

2 1/30/03 2549 0.376 33.1 2.01% 4.32% 2.64%

2 5/21/03 2825 0.492 37.4 1.47% 3.96% 2.39%

2 7/2/03 2817 0.484 37.8 1.31% 4.17% 2.18%

2 8/13/03 2826 0.487 38.5 1.49% 3.95% 2.21%

2 9/15/03 2823 0.490 37.6 1.63% 4.40% 2.44%

2 10/22/03 2820 0.488 37.2 1.64% 4.33% 2.35%

2 11/26/03 2830 0.489 36.6 1.86% 3.82% 2.54%

2 1/9/04 2823 0.492 36.6 2.84% 4.52% 3.54%

BSEP 10-0133 Enclosure 3 Page 33 of 44 Table 17.3: TIP Uncertainty Component Database Core Power/

Unit Date MWth Void Flow 6T'ij 8T'ijk 6T'_p 2 2/12/04 2822 0.483 37.1 1.84% 4.77% 2.99%

2 3/10/04 2820 0.486 38.3 1.86% 4.51% 2.97%

2 4/14/04 2817 0.484 37.6 1.98% 4.47% 3.13%

2 5/20/04 2827 0.474 37.0 2.32% 4.44% 3.37%

2 6/18/04 2818 0.465 36.1 2.09% 4.87% 3.20%

2 7/21/04 2827 0.464 36.3 2.08% 5.25% 3.31%

2 8/25/04 2806 0.462 36.7 2.29% 5.93% 3.30%

2 10/5/04 2815 0.448 36.5 2.25% 6.22% 3.33%

2 11/3/04 2828 0.469 36.6 2.04% 6.72% 2.93%

2 1/12/05 2805 0.423 34.9 2.36% 8.84% 3.57%

2 5/25/05 2921 0.506 37.8 1.51% 2.73% 1.89%

2 7/7/05 2928 0.493 37.7 1.61% 2.61% 1.86%

2 8/3/05 2913 0.493 37.8 1.55% 2.51% 1.86%

2 9/7/05 2918 0.487 36.7 1.75% 2.92% 1.96%

2 10/12/05 2928 0.489 37.6 1.64% 3.11% 1.89%

2 12/21/05 2913 0.494 37.2 1.73% 4.54% 2.08%

2 1/25/06 2915 0.487 37.8 1.77% 4.47% 2.19%

2 3/3/06 2923 0.489 37.5 2.13% 5.25% 2.52%

2 4/4/06 2915 0.482 37.8 2.14% 5.04% 2.64%

2 5/10/06 2917 0.481 37.9 2.12% 4.55% 2.61%

2 6/14/06 2888 0.498 38.1 2.73% 4.09% 3.18%

2 7/19/06 2922 0.487 37.7 2.76% 3.73% 3.15%

2 8/30/06 2915 0.494 37.7 2.62% 4.97% 3.02%

2 10/4/06 2923 0.479 36.8 2.86% 6.33% 3.27%

2 11/1/06 2920 0.508 37.6 2.97% 7.92% 3.28%

2 12/6/06 2917 0.494 37.2 2.73% 7.22% 3.10%

2 1/10/07 2915 0.487 36.9 2.26% 7.30% 2.68%

2 2/15/07 2907 0.458 36.4 1.84% 7.23% 2.39%

2 4/23/07 2914 0.518 37.5 1.60% 5.95% 1.90%

2 5/30/07 2926 0.515 37.8 1.50% 5.67% 1.76%

2 7/3/07 2915 0.483 37.6 1.40% 3.96% 1.76%

2 8/9/07 2912 0.480 37.1 1.43% 3.64% 1.78%

BSEP 10-0133 Enclosure 3 Page 34 of 44 Table 17.3:. TIP Uncertainty Component Database Core Power!

Unit Date MWth Void Flow 8T'ij 5T'ijk 6T'_p 2 9/12/07 2915 0.480 37.9 1.39% 3.34% 1.69%

2 10/17/07 2911 0.491 37.4 1.31% 3.53% 1.6.4%

2 11/20/07 2906 0.498 36.6 1.42% 3.94% 1.85%

2 12/26/07 2917 0.496 37.6 1.44% 2.75% 1.80%

2 2/1/08 2917 0.506 37.5 1.47% 2.86% 1.71%

2 3/12/08 2918 0.507 38.0 1.45% 2.94% 1.71%

2 4/9/08 2922 0.494 37.1 1.43% 2.57% 1.78%

2 5/14/08 2915 0.492 37.8 1.51% 2.92% 1.73%

2 6/26/08 2914 0.476 37.5 1.38% 3.66% 1.85%

2 7/23/08 2913 0.466 37.3 1.39% 4.17% 1.95%

2 8/27/08 2919 0.453 37.3 1.53% 4.96% 2.22%

2 10/8/08 2917 0.469 37.2 1.54% 4.16% 2.11%

2 11/21/08 2917 0.493 37.4 1.69% 3.60% 2.15%

2 12/30/08 2905 0.469 37.6 2.05% 3.36% 2.43%

2 1/28/09 2921 0.472 37.3 1.86% 3.02% 2.15%

2 2/19/09 2906 0.466 36.7 2.02% 3.45% 2.31%

2 5/5/09 2902 0.498 36.8 1.77% 4.23% 2.42%

2 6/17/09 2930 0.491 37.3 1.90% 3.91% 2.49%

2 7/22/09 2912 0.487 37.5 1.81% 3.77% 2.40%

2 8/26/09 2915 0.493 37.5 1.87% 3.93% 2.55%

2 10/14/09 2915 0.488 37.0 1.82% 3.75% 2.40%

2 11/18/09 2921 0.494 37.2 1.82% 3.36% 2.20%

2 12/16/09 2926 0.491 37.1 1.70% 3.77% 2.13%

2 1/20/10 2916 0.485 37.6 1.82% 3.52% 2.37%

2 2/24/10 2920 0.487 37.7 1.74% 3.29% 2.27%

BSEP 10-0133 Enclosure 3 Page 35 of 44 0.10 0.09 0.08 0.07 0.06 Qi 0) 0.05 0

t-.

V) 0.04 0n 0.03 0.02 0.01 0.00 0.00 0.01 0.02 0.03 0.04 0.05 0.06 0.07 0.08 0.09 0.10 Values Recalculated From Latest MB2 Benchmark 0 6T'ij D 6T'ijk \ 6T'_planar -y=x

- Linear (6T'ij) - Linear (6T'ijk) - Linear (6T'_planar)

Figure 17.1: Impact of MB2 Explicit Water Channel/Rod, PLFR Plenum and Spacer Options on BSEP TIP Statistics

BSEP 10-0133 Enclosure 3 Page 36 of 44 0.10 0.09 0.08 0.07 0.06 0.05 F- 0.04 0.03 0.02 0.01 0.00 30 31 32 33 34 35 36 37 38 39 40 Core Power/Flow (MWth-hr/Mlb)

• BSEP Database - - BSEP Specific 0.10 0.09 .................... .................... ................... ..................... ...................................

0.08 ................... .................... ...................

0 0 0.07 .................... . .... * .......... .................... .................... ................... ..... ............ T * .................... .................... ...................

0.06 .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ! ........

i...........

• *... .t ....... .......

i ..............  :...................

0.05 . 7.......... .* .................. . . . .........

F- 0.04 0.03 .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... ... ...................... ....................

0.02 i J

• JJ 0.01 0.00 0.34 0.36 0.38 0.40 0.42 0.44 0.46 0.48 0.50 0.52 0.54 Core Void Fraction I

• BSEP Database - - BSEP Specific I 0.10 .......................

0.09 ...................... ................. ........... ...................  :....................... 1...................... 1....................

0.08 ...................... ...............

0.07 ...................... ............... ...................... ....................... .......................

...................... ............................... *4.

0.06 T'`0.05

~~......

~~~.

........7.....................*..

- ~...*

. . . . . . . . . . }"

i.......

• .... i

........... ........... ......... ........ I..... ......

00.04 ...

0.03

........ ....... iiiiiiiil 0.02 0.01 0.00 2500 2550 2600 2650 2700 2750 2800 2850 2900 2950 Core MWth I

• BSEP Database - - BSEP Specific Figure 17.2: BSEP 6T'ijk Gamma TIP Database (Explicit Water Channel/Rod, PLFR Plenum and Spacer Options)

BSEP 10-0133 Enclosure 3 Page 37 of 44 0.06 0.05 0.04 0.03 0.02 0.01 0.00 30 31 32 33 34 35 36 37 38 39 40 Core Power/Flow (MWth-hr/Mlb)

• BSEP Database -- BSEP Specific 0.06 0.05 0.04 0.03 10 0.02 0.01 0.00 0.34 0.36 0.38 0.40 0.42 0.44 0.46 0.48 0.50 0.52 0.54 Core Void Fraction I BSEP Database - - BSEP Specific '

0.06 -....

0.05 .........

0.04 .........

0.03 .........

F-)

0.02 0.01 .........

0.00 -

2500 2550 2600 2650 2700 2750 2800 2850 2900 2950 Core MWth I

• BSEP Database - - BSEP Specific I Figure 17.3: BSEP 6T'ij Gamma TIP Database (Explicit Water Channel/Rod, PLFR Plenum and Spacer Options)

BSEP 10-0133 Enclosure 3 Page 38 of 44 0.06 0.05 0.04 a 0.03 3 0.02 F--

10 0.01 0.00 30 31 32 33 34 35 36-- 37 38 39 40 Core Power/Flow (MWth-hr/Mlb)

I

• BSEP Database - - BSEP SpecificI 0.06 0.05 0.04 0.03

. 0.02 F--

10 0.01 0.00 0.34 0.36 0.38 0.40 0.42 0.44 0.46 0.48 0.50 0.52 0.54 Core Void Fraction 1 BSEP Database - - BSEP Specific I

0. 06 0.05 ...................

0 .0 4 ....................... ....

0.0 .......

0 r-- 0 .0 1 ......................

,a0.01 0.00 2500 2550 2600 2650 2700 2750 2800 2850 2900 2950 Core MWth I # BSEP Database - - BSEP Specific I Figure 17.4: BSEP 6T'planar Gamma TIP Database (Explicit Water Channel/Rod, PLFR Plenum and Spacer Options)

BSEP 10-0133 Enclosure 3 Page 39 of 44 0.040 10 0.035 U,

PO 0.030 0.025

• 0.020 0.015 o 0.010

• 0.005

.,0.000 0) 0 0.01 0.02 0.03 0.04 0.05 0.06 0.07 0.08 0.09 0.1 Measured vs Calculated TIP Deviation (6Tij, 6Tijk)

I

• 6Sij *5Sijk Figure 17.5: BSEP 6S Versus 6T'

BSEP 10-0133 Enclosure 3 Page 40 of 44 NRC Question No. 18 ANP-2920(P), Brunswick Unit 2 Cycle 20 Fuel Cycle Design In Section 2.0 of ANP-2920(P), the licensee, stated that "beyond the full power capability, the cycle has been designed to achieve 38 gigawatts days additional energy via Constant Pressure Power Coastdown operation."

Please briefly explain the process regarding constant pressure power coastdown operation.

Response No. 18 During coastdown operation, the reduction in core power also means a reduction in steam flow with a corresponding reduction in the steam line pressure drop. When operating on equalizer header pressure control, maintaining a constant pressure at the header will result in continually reducing dome pressure during coastdown due to the reduction in steam flow. The alternative is to improve plant thermal efficiency by maintaining a constant dome pressure during coastdown.

The statement quoted in this RAI is made in ANP-2920(P), since the statepoints defined by the depletion steps are used to define the restarts used in the later licensing analyses, which are performed to bound both constant dome pressure and constant header pressure coastdown operation.

NRC Question No. 19 ANP-2920(P), Brunswick Unit 2 Cycle 20 Fuel Cycle Design Section 3.1 of ANP-2920(P) states that elevation views of the fresh reload fuel design axial enrichment and Gadolinia distributions are shown in Appendix B, Figures B.1 and B.2.

Please explain the naming convention that identifies the enrichment and Gadolinia distributions.

In addition, please provide the range of Gadolinia enrichments for each fuel design that will be used during the upcoming operating cycles for BSEP.

Response No. 19 The lattice naming convention includes three different character groups. The first group identifies the fuel product line and lattice geometry. The second grouping specifies lattice average enrichment. The 'L' in the second group simply specifies that this is a lattice value. The third and final group describes the gadolinia loading by number of rods and concentration. If no

'V' is provided in this group then all gadolinia rods have the specified concentration. If a 'V' is listed, then the lattice contains more than one gadolinia concentration and the highest concentration is provided.

BSEP 10-0133 Enclosure 3 Page 41 of 44

[I

_ll Examples:

XMLCB- (( 11 - (( 11 I st group: 'XMLCB' specifies this is an ATRIUM 1OXM product line bottom geometry lattice 2nd group: [H 11 specifies this is a lattice with (( 11 U-235 by weight average enrichment 3rd group: 11 specifies the lattice contains ((

II XMLCT- (( ]I-I[ II I st group: 'XMLCT' specifies this is an ATRIUM 1OXM product line top geometry lattice 2nd group: [I )) specifies this is a lattice with (( _)) U-235 by weight average enrichment 3rd group: II 11 specifies the lattice contains ((

II Bundle identifications are similar except that no geometry specification is provided in the first character group (i.e., XMLC for ATRIUM I OXM) and a 'B' is provided in the second character group as a bundle designator.

Projected future operating cycles (i.e., Table 19.1) show the as-expected variations in enrichment and gadolinia loading for future operating cycles. These are provided as representative of expected future loadings and are not limiting values. Enrichment within a lattice is limited by a 5% U-235 pellet limit [

11

BSEP 10-0133 Enclosure 3 Page 42 of 44 Table 19.1: Planned and Projected ATRIUM 1OXM Core Loading for BSEP

((

NRC Question No. 20 Impact on the Spent Fuel Pool ATRIUM IOXM fuel has a larger outer diameter, slightly longer active fuel length, larger pellet diameter, higher pellet density, and higher uranium weight than those of'resident fuel types at BSEP. Accordingly, please determine whether a spent fuel pool criticality analysis is necessary to accommodate the new fuel in the pool. Provide details of the analysis.

Response No. 20 AREVA has completed a BSEP-specific A0OXM spent fuel pool criticality safety analysis, which CP&L will implement in accordance with IOCFR50.59 to support introduction ofAl OXM. The analysis models the A 1OXM design, including the larger A IOXM fuel rod outer diameter, slightly longer active fuel length, larger pellet diameter, higher pellet density, and higher uranium weight relative to the A10 fuel design. Additional analysis details are summarized in Figure 20.1 for information.

In CP&L letters BSEP 10-0052 and BSEP 10-0057 dated April 29, 2010, the Company identified reports that would be provided to the NRC, for information, consistent with the process

BSEP 10-0133 Enclosure 3 Page 43 of 44 described in ANF-89-98(P)(A), Revision I and Supplement 1, Generic MechanicalDesign Criteriafor BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995. A spent fuel pool criticality analysis report is not included among the identified reports, because spent fuel pool criticality is not an ANF-89-98(P)(A) design criteria, i.e., execution of the NRC approved ANF-89-98(P)(A) process approves a new fuel design for use without provision of spent fuel pool criticality analysis summaries to the NRC.

BSEP 10-0133 Enclosure 3 Page 44 of 44 SeS Figure 20.1: BSEP A10XM Spent Fuel Pool Criticality Safety Analysis Overview