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License Amendment Request for Transition to Areva Atrium 10XM Fuel and Areva Safety Analysis Methodology
	ML13200A187
Person / Time
Site:	Monticello
Issue date:	07/15/2013
From:	Schimmel M; Northern States Power Co, Xcel Energy
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
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ENCLOSURES 6i 8, 10, 12, 14, 16, 18,20, 22, and 24-CONTAIN PROPRIETARY INFORMATION WITHHOLD FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390 XcI Energy Monticello Nuclear Generating Plant 2807 W County Rd 75 Monticello, MN 55362 July 15, 2013 L-MT-1 3-055 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed License No. DPR-22

Subject:

License Amendment Request for Transition to AREVA ATRIUM IOXM Fuel and AREVA Safety Analysis Methodology Pursuant to 10 CFR 50.90, the Northern States Power Company, a Minnesota Corporation (NSPM), doing business as Xcel Energy, hereby requests an amendment to the renewed operating license for Monticello Nuclear Generating Plant (MNGP). The proposed amendment involves a Fuel Transition from the Global Nuclear Fuels (GNF)

GE14 fuel design to the previously-approved AREVA ATRIUM 1OXM fuel design for use at MNGP. Specifically, NSPM proposes to revise Technical Specification (TS) 5.6.3 to add the AREVA analysis methodologies to the list of approved methods to be used in determining the core operating limits in the Core Operating Limits Report (COLR). To support this fuel transition, NSPM also proposes to revise Technical Specifications 2.1 and 4.2.1. The proposed TS 2.1 would change the steam dome pressure associated with the safety limits when using AREVA safety analysis methods, and the proposed TS 4.2.1 would add a minor clarification to the description of the fuel assemblies in the reactor core.

At this time, NSPM is requesting approval for the transition to AREVA fuel at Extended Power Uprate (EPU) conditions (i.e., 120% Original Licensed Thermal Power level) with Maximum Extended Load Line Limit Analysis (MELLLA). As discussed at a pre-application meeting with NRC Staff on February 28, 2013, implementation of this proposed amendment would preclude use of MELLLA Plus (MELLLA+) extended power-flow operating domain, which is currently under NRC review.

001

Document Control Desk Page 2 The proposed amendment involves no change to the licensed power level and no power operation outside the MELLLA power-flow operating domain. provides a description of the proposed changes, technical evaluation of the proposed changes, regulatory evaluation, and a discussion of environmental considerations. Enclosure 3 provides the existing TS pages marked-up to show the proposed changes. Enclosure 4 provides marked-up TS Bases pages for information only.

In support of the proposed TS changes, certain technical information related to the transition core design and licensing analyses, as well as information related to the AREVA analysis methodologies, has been provided in Enclosures 5 through 25 of this submittal. Enclosures 6, 8, 10, 12, 14, 16, 18, 20, 22, and 24 to this letter contain information that AREVA NP considers to be proprietary in nature and subsequently, pursuant to 10 CFR 2.390, "Public inspections, exemptions, requests for withholding,"

paragraph (a)(4), it is requested that such information be withheld from public disclosure. AREVA, as the owner of the proprietary information, has executed an affidavit provided in Enclosure 2, which identifies that the enclosed proprietary information has been handled and classified as proprietary, is customarily held in confidence, and has previously been withheld from public disclosure. AREVA requests that the enclosed proprietary information be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390.

Enclosures 7, 9, 11, 13, 15, 17, 19, 21,23, and 25 contain the redacted versions of the proprietary enclosures with the proprietary material removed, which are suitable for public disclosure.

NSPM has determined that the information for the proposed amendment does not involve a significant hazards consideration, authorize a significant change in the types or total amounts of effluent release, or result in any significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment meets the categorical exclusion requirements of 10 CFR 51.22(c)(9) and an environmental impact assessment need not be prepared.

A copy of this submittal, including the Determination of No Significant Hazards Consideration, without Enclosures 2 through 25, is being forwarded to the designated State of Minnesota official pursuant to 10 CFR 50.91(b)(1).

NSPM requests approval of this proposed amendment by January 15, 2015 to support the scheduled refueling for MNGP Cycle 28. Based on NSPM's commitment to AREVA fuel for Cycle 28 and beyond, timely approval of this amendment is essential to the startup and continued operation of MNGP. Once approved, the amendment will be implemented within 60 days after the reactor shutdown for the end of Cycle 27 refueling.

If there are any questions or if additional information is needed, please contact Glenn Adams at 612-330-6777.

Document Control Desk Page 3 Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on: July ts , 2013 Mark A. Schimmel Site Vice-President Monticello Nuclear Generating Plant Northern States Power Company-Minnesota Enclosures (25)

No. Title 1 Evaluation of Proposed Chanqe 2 AREVA Affidavits 3 TS Markups 4 TS Bases Markups Boiling Water Reactor Licensing Methodology Compendium, 5" ANP-2637 Revision 4 6 ANP-3224P Applicability of AREVA NP BWR Methods to Monticello, 7 ANP-3224NP Revision 2 8 ANP-3119P Mechanical Design Report for Monticello ATRIUM 1OXM Fuel 9 ANP-3119NP Assemblies, Revision 0 10 ANP-3092P Monticello Thermal-Hydraulic Design Report for ATRIUM I0XM 11 ANP-3092NP Fuel Assemblies, Revision 0 12 ANP-3138P Monticello Improved K-factor Model for ACE/ATRIUM 10XM 13 ANP-3138NP Critical Power Correlation, Revision 0 14 ANP-3215P Monticello Fuel Transition Cycle 28 Fuel Cycle Design 15 ANP-3215NP (EPU/MELLLA), Revision 0 16 ANP-3213P Monticello Fuel Transition Cycle 28 Reload Licensing Analysis 17 ANP-3213NP (EPU/MELLLA), Revision 1 18 ANP-321 1P Monticello LOCA Break Spectrum Analysis for ATRIUM I 0XM 19 ANP-3211NP Fuel (EPU/MELLLA), Revision 1 20 ANP-3212P Monticello LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 21 ANP-3212NP 1OXM Fuel (EPU/MELLLA), Revision 0 22 ANP-3221P Fuel Rod Thermal-Mechanical Design for Monticello ATRIUM 23 ANP-3221NP 10XM Fuel Assemblies, Cycle 28, Revision 0

Document Control Desk Page 4 24 ANP-3139P Nuclear Fuel Design Report Monticello Cycle 28 ATRIUM 1OXM 25 ANP-3139NP IFuel, Revision 1 cc: Regional Administrator, Region III, USNRC Project Manager, Monticello Nuclear Generating Plant, USNRC Resident Inspector, Monticello Nuclear Generating Plant, USNRC Minnesota Department of Commerce (w/o enclosures 2-25)

ENCLOSURE1 Evaluation of the Proposed Change License Amendment Request for Transition to AREVA ATRIUM 10XM Fuel And AREVA Safety Analysis Methodology 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Proposed Change to TS 2.1, "Safety Limits" 2.2 Proposed Change to TS 4.2.1, "Fuel Assemblies" 2.3 Proposed Change to TS 5.6.3, "Core Operating Limits Report (COLR)"

3.0 TECHNICAL EVALUATION

3.1 Design Description -ATRIUM 1OXM Fuel 3.2 Background on Reactor Core Safety Limit Changes 3.3 Current Licensing Basis 3.4 Justification for the Proposed Changes 3.5 Conclusion

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

S

6.0 REFERENCES

Page 1 of 15 NSPM Fuel Transition 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, the Northern States Power Company, a Minnesota Corporation (NSPM), doing business as Xcel Energy, hereby requests an amendment to the renewed operating license for Monticello Nuclear Generating Plant (MNGP). The proposed amendment involves a Fuel Transition from the Global Nuclear Fuels (GNF) GE14 fuel design to the previously-approved AREVA ATRIUM 1OXM fuel design for use at MNGP. Specifically, NSPM proposes to revise Technical Specification (TS) TS 5.6.3 to add the AREVA analysis methodologies to the list of approved methods to be used in determining the core operating limits in the Core Operating Limits Report (COLR). To support this fuel transition, NSPM also proposes to revise Technical Specifications 2.1 and 4.2.1.

The proposed TS 2.1 would change the steam dome pressure associated with safety limits when using AREVA safety analysis methods, and the proposed TS 4.2.1 would add a minor clarification to the description of the fuel assemblies in the reactor core.

2.0 DETAILED DESCRIPTION NSPM intends to begin using the ATRIUM 1OXM design in MNGP Cycle 28. The first reload of ATRIUM 1OXM is targeted for insertion into the core in the spring 2015 refueling outage. The ATRIUM 10XM product is a proven fuel design approved for use at two other Boiling Water Reactors (BWRs) in the United States.

At this time, NSPM is requesting approval for the transition to AREVA fuel at Extended Power Uprate (EPU) conditions with Maximum Extended Load Line Limit Analysis (MELLLA). As discussed at a pre-application meeting with NRC Staff on February 28, 2013, implementation of this proposed amendment would preclude use of MELLLA Plus (MELLLA+) extended power-flow operating domain, which is currently under NRC review.

In order to extend the use of this fuel design to MNGP and to adopt the AREVA fuel design and safety analyses, several changes to the Technical Specifications are required. The proposed changes to the TS are as follows:

2.1 Proposed Change to TS 2.1, "Reactor Core SLs" The proposed change to TS 2.1 will revise reactor core safety limits (SLs) to reduce the value of reactor steam dome pressure in TS 2.1.1.1 and 2.1.1.2 from 785 psig to a value of 586 psig.

2.2 Proposed Change to TS 4.2.1, "Reactor Core, Fuel Assemblies" The proposed change to TS 4.2.1 will insert a minor editorial change to reflect the ATRIUM 1OXM fuel assembly design feature which is better described as a "water channel", as opposed to a "water rod" of legacy fuel types.

Page 2 of 15 NSPM Fuel Transition 2.3 Proposed Change to TS 5.6.3, "Core Operating Limits Report (COLR)"

The proposed change to TS 5.6.3 will insert the titles of eighteen (18) AREVA analyses that will be used to develop core operating limits for MNGP cores loaded with AREVA ATRIUM 1OXM fuel. Technical Specification 5.6.3 addresses the analytical methods which may be used to determine input to the COLR. Currently, the MNGP specification includes GNF analytical methods. TS 5.6.3 will be revised to add appropriate NRC-approved AREVA analytical methodologies.

3.0 TECHNICAL EVALUATION

3.1 Design Description -ATRIUM 1OXM Fuel The fuel design to be introduced into MNGP in 2015 is the AREVA ATRIUM 1OXM product. This design utilizes a 10x10 array of fuel rods, with 79 full length fuel rods and 12 partial length fuel rods. The partial length fuel rods are approximately one-half the length of the full length fuel rods. The use of partial length rods improves fuel utilization in the high void upper region of the bundle and also enhances stability, pressure drop performance, and cold shutdown margin.

The ATRIUM 1OXM design does not utilize tie rods as the structural tie between the upper and lower tie plates. Instead, the design uses a central water channel having a mechanical connection to the two tie plates. The central water channel carries the mechanical loads during fuel handling. It displaces a 3x3 array of fuel rods within the bundle and serves to improve fuel economy by improving internal neutron moderation. The lower ends of the fuel rods rest on top of the lower tie plate with their lower ends laterally restrained by a spacer grid located just above the lower tie plate. No expansion springs are required on each fuel rod because a single, large reaction spring is used on the central water channel to hold the upper tie plate in the latched position. The ATRIUM 1OXM design uses a total of nine fuel rod spacers to provide lateral support for the fuel rods and to enhance thermal-hydraulic performance. The ATRIUM 1OXM design to be employed utilizes a debris resistant lower tie plate to limit introduction of foreign material into the assembly from below.

Further description of the ATRIUM 1OXM design is provided in Enclosure 8.

3.2 Background on Reactor Core Safety Limit Changes The basis for Reactor Core Safety Limits and a summary of the Pressure Regulator Failure Maximum Demand (Open) transient scenario is provided below to support the discussion of the justification of the changes associated with the AREVA safety analysis methodology in Section 3.4.1. Note that a similar discussion was provided in Reference 6.7.

Page 3 of 15 NSPM Fuel Transition 3.2.1 Reactor Core Safety Limits TS Safety Limits ensure that specified acceptable fuel design limits (SAFDLS) are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences. The Reactor Core Safety Limits are set such that fuel cladding integrity is maintained and no significant fuel damage is calculated to occur if the Safety Limits are not exceeded.

The BWR core is protected from the type of fuel failure that could occur during the Onset of Transition Boiling (OTB) by a combination of Reactor Core Safety Limits 2.1.1.1 and 2.1.1.2. Reactor Core Safety Limit 2.1.1.1 states when the reactor steam dome pressure is less than 785 psig or when core flow is less than 10 percent of rated core flow, reactor thermal power shall be less than a specified value (25 percent of Rated Thermal Power -

RTP). When reactor pressure and core flow are greater than these specified values, Reactor Core Safety Limit 2.1.1.2 prohibits operation with a MCPR Safety Limit less than the value specified to prevent fuel cladding damage that could occur when a fuel assembly experiences the OTB.

3.2.2 Pressure Regulator Failure Maximum Demand (Open) Transient Analysis As analyzed with AREVA transient analysis codes, the subject transient may result in a reactor pressure decrease below 785 psig before the reactor has been shutdown. In effect, the reactor steam dome pressure could decrease to below 785 psig while thermal power exceeds 25 percent of RTP, which would exceed the conditions in Reactor Core Safety Limit 2.1.1.1. This condition indicates that Reactor Core Safety Limit 2.1.1.1 is overly conservative with respect to this event, because during this event Critical Power Ratio (CPR) continues to increase and therefore does not threaten fuel cladding integrity. However, the pressure decrease could result in exceeding the value specified in the safety limit specification, while having no actual safety significance.

3.3 Current Licensing Basis The baseline Current Licensing Basis (CLB) supporting this License Amendment Request (LAR) is the Updated Safety Analysis Report (USAR), with special consideration for the pending licensing basis changes associated with the EPU to 120% OLTP as submitted in Reference 6.2 and subsequent supplements.

Page 4 of 15 NSPM Fuel Transition 3.4 Justification for the Proposed Changes 3.4.1 Justification for TS 2.1.1 Change The proposed change to TS 2.1.1 is provided to ensure the safety limit reflects the range of the appropriate critical power correlation to cover all applicable operating conditions and transients, including the Pressure Regulator Failure Maximum Demand (Open) transient. NSPM recently submitted a license amendment request (LAR) (Reference 6.7) to change the safety limit for GNF critical power correlation to a steam dome pressure as low as 700 psia (pounds per square inch absolute) / 686 psig (pounds per square inch gauge). Reference 6.7 provides the background and basis for the changes as they relate to GNF safety analysis methods.

The changes proposed herein for TS 2.1.1 are based on the same principles established in Reference 6.7. The values of steam dome pressure and Minimum Critical Power Ratio (MCPR) used in the proposed TS 2.1.1 are supported by the AREVA methodology described in Enclosure

6. Enclosure 6 confirms the validity of the ACE CPR correlation (for ATRIUM 1OXM fuel) and the SPCB CPR correlation (for co-resident GE14 fuel) to a steam dome pressure as low as 600 psia / 586 psig.

Thus, the proposed changes to TS 2.1.1 ensure that the core pressure and flow are within the range of validity of the specified CPR correlation during the transients.

The TS 2.1.1 changes proposed herein will have to be reconciled with the TS 2.1.1 changes proposed in Reference 6.7, depending on the timing of approval for the associated amendments. NSPM will supplement these license amendment requests as necessary.

3.4.2 Justification for TS 4.2.1 Change The proposed change to TS 4.2.1 is a minor editorial change to the fuel description. As discussed in greater detail below (section 3.4.3), the ATRIUM 1OXM fuel design (including this fuel assembly water channel annotated herein) is explicitly modeled in safety analyses and the proposed operating regime will continue to ensure that operating limits will continue to be satisfied.

3.4.3 Justification for TS 5.6.3, New COLR Methodologies The AREVA analytical methods and topical reports to be added to Technical Specification 5.6.3 are those utilized to evaluate the fuel mechanical design, along with both cycle dependent and independent safety analyses, used to establish limits identified in the COLR. Additionally, Reference 6.3 is also Page 5 of 15 NSPM Fuel Transition being added to the TS as the basis for acceptance of the ATRIUM 1OXM fuel design. Each analytical methodology being added to Technical Specification 5.6.3 has been previously reviewed and approved by the NRC.

By the terms of the renewed operating license Technical Specifications 5.6.3.b, the core operating limits must be determined using the previously-approved methods prescribed therein. Inherently, those methods were approved with certain conditions and limitations that must be satisfied in order to meet the prescribed method. A pending licensing activity described by Reference 6.9 would replace the current analysis methods used for MELLLA power-flow domain with a new method for the expanded power-flow domain known as MELLLA+. However, that method is only valid for GNF fuel types. Thereby, any core designed and loaded with AREVA fuel could not produce a valid COLR following approval of the MELLLA+

amendment.

Thus, the TS 5.6.3.b changes proposed herein will have to be reconciled with the TS 5.6.3.b changes proposed in Reference 6.9, depending on the timing of approval for the associated amendments. Should the License Amendment Request allowing use of MELLLA+ at Monticello be approved prior to the issuance of this fuel transition license amendment, NSPM will submit a supplement requesting restoration of the analysis methodologies that support core operating limits associated with MELLLA. The methods to be restored are those currently described by TS 5.6.3.b.4 and b.5, which are valid for the AREVA ATRIUM 1OXM fuel type.

3.4.4 Supplementary Analyses Disposition of the plant's postulated accidents with radiological consequences indicated that the consequences of only two accidents could be affected by the proposed change in fuel type; the LOCA and the Control Rod Drop Accident (CRDA). Note that Main Steam Line Break (MSLB) analysis is not affected because the radiological consequences of that accident are driven by reactor coolant radioactivity as opposed to the fuel isotopic inventory. Further note that the Fuel Handling Accident (FHA) with ATRIUM 1OXM fuel has been previously addressed in Reference 6.1.

Revised dose analyses using the approved Alternative Source Term methodology at EPU rated power concluded that the change in fuel type resulted in only slight variations in the radiological source term for the reactor core and a corresponding slight difference in the overall dose consequences compared to those that were summarized and provided to NRC previously in Reference 6.8. At no location did the calculated dose using ATRIUM 1OXM fuel increase more than two percent compared to the previously-reported dose (at EPU power level). Dose consequences for the LOCA and CRDA with an ATRIUM 1OXM fuel source term remained well Page 6 of 15 NSPM Fuel Transition below the regulatory limit of 10 CFR 50.67 and Regulatory Guide 1.183, as summarized below for the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR):

Fuel Transition (at EPU) Accident Dose (in Rem TEDE)

CRDA LOCA Regulatory Limit EAB 2.01 1.47 25 (6.3 for CRDA)

LPZ 0.92 1.99 25 (6.3 for CRDA)

CR 1.89 3.83 5 To bound mixed-core operation (i.e., transitions through the full AREVA core) with ATRIUM 1OXM fuel, the accident dose was calculated as the composite dose based on the highest of each calculated pathway dose from either fuel type (GEl4 or ATRIUM 1OXM). The composite bounding accident doses are those that are tabulated above, which will become the radiological results of record upon approval and implementation of this Fuel Transition license amendment.

3.4.5 Fuel Design Changes The ATRIUM IOXM design was developed using the thermal-mechanical design bases and limits outlined in Reference 6.3. Compliance with Reference 6.3 ensures the fuel design meets the fuel system damage, fuel failure, and fuel coolability criteria identified in the Standard Review Plan (Reference 6.4). The NRC reviewed and approved (per Reference 6.5) the use of Reference 6.3 for making changes and improvements to fuel designs; specifically stating such changes and improvements do not require specific NRC review and approval, provided the criteria are satisfied. The ATRIUM 1OXM design fully complies with the criteria of Reference 6.3, and therefore meets all of the required fuel licensing criteria in Reference 6.4.

The impact of the ATRIUM 1OXM design on the USAR accident analyses will be accounted for by cycle-specific reload and accident analyses.

Limiting transients from USAR Chapter 14 categories of pressure increase events, vessel water temperature decrease events, control rod withdrawal error events, core flow increase events, and increase in vessel inventory events are evaluated each cycle. Limiting analyses results for a representative transition cycle are presented in Enclosure 16.

Introduction of the ATRIUM 1OXM design fuel will not adversely impact USAR accident analyses. AREVA evaluates the control rod drop accident (USAR section 14.7.1) on a cycle-specific basis. Enclosure 16 includes a cycle-specific evaluation of the control rod drop accident for a representative Page 7 of 15 NSPM Fuel Transition transition cycle. The evaluation shows the number of rods calculated to fail in this event remains well below the value of 850 assumed in the USAR radiological evaluation of this event. The doses from the control rod drop accident remain within limits required by 10 CFR 50.67, "Accident Source Term," and Regulatory Guide 1.183 (Reference 6.6).

Regarding the LOCA analysis (USAR section 14.7.2), a baseline LOCA break spectrum analysis of ATRIUM 1OXM fuel has been performed at 120% Original License Thermal Power (OLTP) power conditions; it is included as Enclosure 18. Cycle-specific fuel design MAPLHGR (Maximum Average Planar Linear Heat Generation Rate) limits are analyzed consistent with assumptions used in the baseline Loss of Coolant Accident (LOCA) analysis. Peak cladding temperature, cladding oxidation, and hydrogen generation analyses results are included in Enclosure 20 for a representative transition cycle. The introduction of ATRIUM 1OXM fuel will not challenge the peak clad temperature, cladding oxidation, or hydrogen generation limits specified in 10 CFR 50.46 "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,"

paragraph (b). Satisfaction of the remaining two 10 CFR 50.46 acceptance criteria (coolable geometry and long-term core cooling) is also demonstrated in Enclosure 20.

The ATRIUM 1OXM design will not challenge the USAR basis of the refueling accident (USAR section 14.7.6), which has been evaluated in the Fuel Storage Changes LAR (Reference 6.1). As described therein, the fuel handling accident described (for EPU) remains bounding for ATRIUM 1OXM fuel. The doses resulting from this event will remain within the limits specified in 10 CFR 50.67.

The main steam line break accident (USAR section 14.7.3) is not affected by a change in fuel design. As stated in the USAR, no fuel failures are expected to occur as a result of this accident. The radionuclide inventory released from the primary coolant system is present in the coolant prior to the event. Therefore, the fuel design change does not alter the consequences of a main steam line break accident.

The NRC has previously reviewed and approved transitions from GE14 to ATRIUM-10 and ATRIUM 1OXM (see section 4.2 below). Previous reviews confirmed the acceptability of transitioning from GE14 to the ATRIUM design. The scope of the technical analyses provided in support of the MNGP submittal is consistent with the technical analyses provided with the precedent submittals. The MNGP fuel transition employed the AREVA methods referenced in the proposed TS change (Enclosure 2) to evaluate both the ATRIUM 1OXM fuel and the co-resident GEl4 fuel. The AREVA methodologies are applied in accordance with NRC approval for performance of design and licensing analyses for mixed cores. The Page 8 of 15 NSPM Fuel Transition thermal-hydraulic characteristics of the GE14 fuel design were explicitly accounted for and a detailed thermal-hydraulic analysis of the mixed core was performed as documented in Enclosure 10.

The GE14 lattice and fuel bundle geometry, as well as the specific enrichment and burnable poison distributions, were explicitly modeled in the neutronic analyses provided in Enclosure 14. The AREVA neutronic methods have been extensively benchmarked for cores containing GE14 fuel, including multiple MNGP cycles containing that fuel type.

The reload safety analysis documented in Enclosure 16 explicitly evaluates the transient and accident response of the mixed core. Because the GE14 fuel is explicitly modeled, the impact of the co-resident fuel on the core response to transient and accident events is accounted for. NSPM provided data to AREVA to ensure the critical power thermal limits of the co-resident fuel are appropriately calculated and monitored. In addition, NSPM provided AREVA with thermal-mechanical limits for the co-resident fuel for use in the design and safety analyses to ensure the appropriate limits are met. The core monitoring system will explicitly monitor the co-resident fuel, and will apply the specific thermal limits determined for that fuel.

3.5 Conclusion The proposed amendments are adequately justified and support the ATRIUM 1OXM fuel design and AREVA safety analysis methodologies for MNGP.

Supplementary analysis of the post-accident radiological consequences indicates that the accident dose increases for some locations, but remains well below the regulatory limit.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The following regulatory requirements relate to fuel related design and licensing criteria specified in the Standard Review Plan (Reference 6.4):

10 CFR 50, Appendix A, General Design Criteria (GDC) 10, "Reactor design,"

GDC 27, "Combined reactivity control system capability," and GDC 35, "Emergency core cooling. While Monticello is not generally licensed to the current 10 CFR 50 Appendix A GDC or the 1967 Atomic Energy Commission (AEC) proposed GDC, a comparison of the current GDC to the applicable AEC proposed GDC can be made. For the current GDC listed above, the Monticello comparative evaluation of the comparable 1967 AEC-proposed General Design Criteria is contained in Monticello USAR Appendix E.

10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors."

Page 9 of 15 NSPM Fuel Transition

10 CFR 50, Appendix K, "ECCS Evaluation Models."

With respect to methodologies suitable for use in developing the COLR, those listed in Technical Specifications must be previously approved by NRC.

4.2 Precedent A search of NRC actions on TS changes revealed the NRC has previously approved similar changes for the following plants:

Browns Ferry Nuclear Plant, Unit 1 - Issuance of Amendments Regarding the Transition to AREVA Fuel (TAC No. ME3775)(TS-473), dated July 3, 2012 (ADAMS Accession No. ML12129A149).

This amendment approved Browns Ferry transition from operations with GE14 fuel design to ATRIUM-10 fuel design at non-EPU power level. The amendment also adopted a comparable set of AREVA topical reports in the list of approved methods for determining the core operating limits. Whereas the Browns Ferry amendment provides the best precedent for the scope and content of the proposed MNGP fuel transition LAR, it differs in two notable ways: (1) Browns Ferry transitioned to ATRIUM-10, not ATRIUM 1OXM, and (2) Browns Ferry analyses were not extended to EPU power levels (105%

OLTP vs. 120% OLTP).

The Browns Ferry precedent was chosen to model the scope and content of the MNGP LAR because it was more complete and more recent than any other associated precedent.

Browns Ferry Nuclear Plant, Units 2 and 3 - Issuance of Amendments Regarding Core Operating Limits (TAC Nos. MB8433 and MB8434), dated December 30, 2003, (ADAMS Accession No. ML033650142).

These amendments represent a comparable transition as that of BFN Unit 1; however, the earlier date of approval for these Unit 2 and 3 amendments ascribes more credibility to using Unit 1 as the precedent.

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment Nos. 246 and 274 to Renewed Facility Operating Licenses Nos.

DPR-71 and DPR-62, Carolina Power & Light Company, Brunswick Steam Electric Plant, Units 1 and 2, Docket Nos. 50-325 and 50-324, March 27, 2008.

(ADAMS Accession No. ML080870546).

This amendment approved Brunswick's transition from operations with GE14 fuel design to ATRIUM-10 fuel design at EPU power level (120% OLTP). The amendment also adopted the comparable AREVA topical reports in the list of approved methods for determining the core operating limits. Whereas the Page 10 of 15 NSPM Fuel Transition Brunswick amendment provides some precedent for the scope and content of the proposed MNGP fuel transition LAR, it differs in one notable way: (1)

Brunswick transitioned to ATRIUM-1 0, not ATRIUM 10XM.

Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendments Regarding Addition of Analytical Methodology Topical Report To Technical Specification 5.6.5 (TAC Nos. ME3856 and ME 3857), dated April 8, 2011 (ADAMS Accession No. ML111010234).

This amendment approved the use of ACE/ATRIUM 1OXM Critical Power Correlation, Revision 0 (March 2010) and supported the plant's transition to ATRIUM 1OXM fuel design and associated core design methodologies. This precedent is directly applicable to this MNGP fuel transition LAR in that it provides precedent for the ATRIUM 1OXM fuel design and the associated critical power correlation requested herein.

4.3 Significant Hazards Consideration Pursuant to 10 CFR 50.90, the Northern States Power Company, a Minnesota Corporation (NSPM), doing business as Xcel Energy, hereby requests an amendment to the renewed operating license for Monticello Nuclear Generating Plant (MNGP). The proposed amendment involves a Fuel Transition from the Global Nuclear Fuels (GNF) GE14 fuel design to the previously-approved AREVA ATRIUM 1OXM fuel design for use at MNGP. Specifically, NSPM proposes to revise Technical Specification (TS) 5.6.3 to add the AREVA analysis methodologies to the list of approved methods to be used in determining the core operating limits in the Core Operating Limits Report (COLR). To support this fuel transition, NSPM also proposes to revise Technical Specifications 2.1 and 4.2.1.

The proposed TS 2.1 would change the steam dome pressure associated with the safety limits when using AREVA safety analysis methods, and the proposed TS 4.2.1 would add a minor clarification to the description of the fuel assemblies in the reactor core.

NSPM has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92(c) as discussed below:

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

Response: No Changing fuel designs and making an editorial change to TS will not increase the probability of a Loss of Coolant Accident (LOCA). The fuel cannot increase the probability of a primary coolant system breach or rupture, as there is no interaction between the fuel and the system piping. The fuel will continue to Page 11 of 15 NSPM Fuel Transition meet the 10 CFR 50.46 limits. Therefore, the consequences of a LOCA will not be increased.

The probability of a Control Rod Drop Accident (CRDA) does not increase because the ATRIUM 10XM fuel channel is mechanically compatible with the co-resident fuel and existing control blade designs. The mechanical interaction and friction forces between the ATRIUM 1OXM channel and control blades would not be higher than previous designs. In addition, routine plant testing includes confirmation of adequate control blade to control rod drive coupling.

The probability of a CRDA is not increased with the use of ATRIUM 1OXM fuel.

CRDA consequences are evaluated on a cycle-specific basis, confirming the number of calculated rod failures remains within the Updated Safety Analysis Report (USAR) design basis.

Similarly, changing the fuel design and making an editorial change to TS cannot increase the probability of an anticipated operational occurrence (AOO).

As a passive component, the fuel does not interact with plant operating or control systems. Therefore, the fuel change cannot affect the initiators of the previously evaluated AOO transient events. Thermal limits for the new fuel will be determined on a cycle-specific basis, ensuring the specified acceptable fuel design limits continue to be met. Therefore, the consequences of a previously evaluated AOO will not increase.

A disposition of the plant's postulated accidents with radiological consequences indicated that the consequences of only two accidents could be affected by the proposed change in fuel type; the LOCA and the CRDA. Revised dose analyses using the approved Alternative Source Term methodology at Extended Power Uprate (EPU) rated power concluded that the change in fuel type resulted in small variations in the radiological source term for the reactor core and a corresponding slight difference in overall dose consequences. At no location did the calculated dose increase more than two percent compared to previously-submitted radiological dose at EPU power levels. Dose consequences for the LOCA and CRDA with an ATRIUM 1OXM fuel source term remained well below the regulatory limits of 10 CFR 50.67 and Regulatory Guide 1.183.

The proposed change to Reactor Core Safety Limits involves a technical evaluation that demonstrates the range of applicability for the AREVA Critical Power Correlations will always bound the postulated pressures of plant transients using the AREVA safety analysis methodology. As a technical evaluation, this proposed change involves no physical change to a system, structure, component, or setpoint. Therefore, this proposed change in no way can affect the probability or the consequences of an accident previously evaluated.

Page 12 of 15 NSPM Fuel Transition

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

Response: No The ATRIUM 1OXM fuel product has been designed to maintain neutronic, thermal-hydraulic, and mechanical compatibility with the co-resident fuel designs currently in use at MNGP. The ATRIUM 1OXM fuel has been designed to meet fuel licensing criteria specified in NUREG-0800, "Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants." Compliance with these criteria ensures the fuel will not fail in an unexpected manner.

A change in fuel design and an editorial change to TS cannot create any new accident initiators because the fuel is a passive component having no direct influence on the performance of operating plant systems and equipment.

Hence, a fuel design change cannot create a new type of malfunction leading to a new or different kind of transient or accident. Consequently, the proposed fuel design change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Also, as a technical evaluation, the proposed change to Reactor Core Safety Limits involves no physical change to a system, structure, component, or setpoint. Therefore, this proposed change could in no way introduce a new physical interaction that would create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No The ATRIUM 1OXM fuel is designed to comply with the fuel licensing criteria specified in NUREG-0800. Cycle-specific and cycle-independent safety analyses are performed ensuring no fuel failures will occur as the result of anticipated operational occurrences, and dose consequences for accidents remain within the bounds of 10 CFR 50.67. Applicable regulatory margins and requirements are maintained.

The proposed change to Reactor Core Safety Limits is consistent with, and within the capabilities of the applicable NRC approved critical power correlation, and thus continues to ensure that valid critical power calculations are performed. No setpoints at which protective actions are initiated are altered by the proposed change. The proposed change does not alter the manner in which the safety limits are determined. This change is consistent with plant design and does not change the TS operability requirements; thus, previously evaluated accidents are not affected by this proposed change.

Page 13 of 15 NSPM Fuel Transition Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, NSPM concludes the proposed amendment does not involve a 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.

4.4 Conclusions 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.

5.0 ENVIRONMENTAL CONSIDERATION

S 10 CFR 51.22(c)(9) provides criteria for and identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment of an operating license for a facility requires no environmental assessment if the operation of the facility in accordance with the proposed amendment does not: (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and (3) result in a significant increase in individual or cumulative occupational radiation exposure. NSPM has reviewed this LAR and determined that the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment. The basis for this determination follows.

1. As demonstrated in the 10 CFR 50.92 evaluation, the proposed amendment does not involve a significant hazards consideration.

2. The proposed amendment does not result in a significant change in the types or increase in the amounts of any effluents that may be released offsite.

Implementation of the proposed project involves no new physical activity; core loading procedures and the frequency of fuel handling operations do not change. Thereby, implementing the new TS is not expected to generate any solid, gaseous, or liquid effluent that would not otherwise be generated in the course of routine reactor operations over its lifetime.

3. The proposed amendment does not result in an increase in individual or cumulative occupational radiation exposure. The radiological source term from the ATRIUM 1OXM is not significantly different from that of the legacy fuel types irradiated to licensed power levels.

Page 14 of 15 NSPM Fuel Transition

6.0 REFERENCES

6.1 NSPM letter to NRC, License Amendment Request for Fuel Storage Changes, L-MT-1 2-076, dated October 30, 2012 (ADAMS Accession No. ML12307A433).

6.2 NSPM letter to NRC, License Amendment Request: Extended Power Uprate (TAC MD9990), L-MT-08-052, dated November 5, 2008 (ADAMS Accession No. ML083230111).

6.3 ANF-89-98(P)(A) Revision 1 and Supplement 1, "Generic Mechanical Design Criteria for BWR Designs," Advanced Nuclear Fuels Corporation, dated May 1995.

6.4 NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," Section 4.2, 'Fuel System Design,' Revision 3, dated March 2007.

6.5 Letter from R.C. Jones (NRC) to R. Copeland (Siemens Power Corporation),

"Acceptance for Referencing of Topical Report ANF-89-98(P), Revision 1, "Generic Mechanical Design Criteria for BWR Fuel Designs," (TAC No. M81070)," dated April 20, 1995.

6.6 Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," U.S. Nuclear Regulatory Commission, dated July 2000.

6.7 NSPM letter to Document Control Desk (NRC), L-MT-1 3-010, "License Amendment Request: Reduce the Reactor Steam Dome Pressure Specified in the Reactor Core Safety Limits", dated March 11, 2013 (ADAMS Accession No. ML13074A811).

6.8 NSPM letter to Document Control Desk (NRC), L-MT-1 3-020, "Monticello Extended Power Uprate (EPU): Second Supplement for Gap Analysis Updates (TAC MD9990)", dated February 27, 2013 (ADAMS Accession No. ML13064A213).

6.9 NSPM letter to Document Control Desk (NRC), L-MT-1 0-003, License Amendment Request: Maximum Extended Load Line Limit Analysis Plus, dated January 21, 2010 (ADAMS Accession No. ML100280558).

Page 15 of 15

Enclosure 2 AREVA Affidavits 30 pages follow

AFFIDAVIT STATE OF WASHINGTON )

) SS.

COUNTY OF BENTON )

1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am 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 report ANP-321 I(P), Revision 1, "Monticello EPU LOCA Break Spectrum Analysis for ATRIUM M

T 1OXM Fuel," dated July 2013 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 Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) 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.

Z--

SUBSCRIBED before me this 0 day of L-*,- ,2013.

Susan K. McCoy AHNTO NOTARY PUBLIC, STATE OF WAIN MY COMMISSION EXPIRES: 1/14/2016

AFFIDAVIT STATE OF WASHINGTON )

) ss.

COUNTY OF BENTON )

1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am 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 report ANP-3224P, Revision 1, "Applicability of AREVA NP BWR Methods to Monticello," dated June 2013 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."