L-MT-14-088, Areva Atrium 10XM Fuel Transition License Amendment Request Supplement

Areva Atrium 10XM Fuel Transition License Amendment Request Supplement

ML14323A026
Person / Time
Site:	Monticello
Issue date:	11/11/2014
From:	Fili K Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-14-088, TAC ME2479
Download: ML14323A026 (37)

Text

ENCLOSURE 4 CONTAINS PROPRIETARY INFORMATION WITHHOLD FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390 Monticello Nuclear Generating Plant 2807 W County Rd 75 Monticello, MN 55362 November 11, 2014 L-MT-14-088 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 AREVA ATRIUM 10XM Fuel Transition License Amendment Request Supplement (TAC MF2479)

References:

1) Letter from Mark A. Schimmel (NSPM), to Document Control Desk (NRC), License Amendment Request for Transition to AREVA ATRIUM 10XM Fuel and AREVA Safety Analysis Methodology, L-MT-13-055, dated July 15, 2013. (ADAMS Accession No. ML13200A185)

2) Letter from John C. Grubb (NSPM), to Document Control Desk (NRC),

License Amendment Request: Reduce the Reactor Steam Dome Pressure Specified in the Reactor Core Safety Limits, L-MT-13-010, dated March 11, 2013. (ADAMS Accession No. ML13074A811)

3) AREVA Topical Report, ACE/ATRIUM 10XM Critical Power Correlation, ANP-10298P-A, Revision 1, dated March 2014.

In Reference 1, Northern States Power Company, a Minnesota corporation (NSPM),

doing business as Xcel Energy, requested approval of an amendment to the Monticello Nuclear Generating Plant (MNGP) Renewed Operating License (OL) and Technical Specifications (TS). The proposed change would allow reactor operation with AREVA ATRIUM 10XM fuel and allow use of AREVA safety analysis methods for the development of core operating limits.

The purpose of this supplement is to propose non-technical updates to the previously proposed TSs 2.1.1 and 5.6.3 changes described in Reference 1. These changes are necessary to accommodate related NRC licensing activities that have transpired during the timeframe of the NRC review of Reference 1. The proposed update to TS 2.1.1 is a restructuring of the steam dome pressure safety limit TS to accommodate the original change proposed in Reference 1 and a similar change that was proposed to the same TS in Reference 2. The proposed update to TS 5.6.3 would revise the list of approved

Document Control Desk Page 2 methodologies to adopt Revision 1 of the AREVA ACE Critical Power Correlation, which was recently approved by NRC in Reference 3.

This supplement also provides a summary of the corrections that have been made to the Loss of Coolant Accident (LOCA) analysis that supported the original license amendment request (LAR) (Reference 1). These corrections were identified in the vendors corrective action program. The corrections to the AREVA LOCA analysis are being reported herein to inform the NRC of the changes. As described in Enclosure 4, these changes are small in magnitude and minimal in their effect. Since these changes do not affect current MNGP operation (i.e., MNGP is not operating with AREVA fuel),

these changes are not currently subject to 10 CFR 50.46 reporting.

The changes proposed herein were discussed with NRC Staff at a public meeting on October 1, 2014. provides justification for the proposed revisions to TSs 2.1.1 and 5.6.3. provides revised markups to TSs 2.1.1 and 5.6.3.

To support the proposed revisions to Safety Limits 2.1.1, Enclosure 3 provides revised markups to TS Bases 2.1.1. These markups to TS Bases section 2.1.1 supersede the markups to section 2.1.1 that were provided in the Reference 1 LAR. These markups are provided for information only. provides the proprietary corrections to AREVA LOCA analysis results that were provided in the Reference 1 LAR. provides the non-proprietary corrections to AREVA LOCA analysis results that were provided in the Reference 1 LAR. The nonproprietary report is being provided based on the NRCs expectation that the submitter of the proprietary information should provide, if possible, a nonproprietary version of the document with brackets showing where the proprietary information has been deleted. provides an affidavit executed to support withholding Enclosure 4 from public disclosure. Information in Enclosure 4 contains proprietary information as defined by 10 CFR 2.390. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4). Accordingly, NSPM respectfully requests that the AREVA proprietary information in Enclosure 4 be withheld from public disclosure in accordance with 10 CFR 2.390(a)4, as authorized by 10 CFR 9.17(a)4.

Correspondence with respect to the copyright or proprietary aspects of the AREVA information in Enclosure 4 or the supporting AREVA affidavit in Enclosure 6 should be addressed to Mr. Alan Meginnis, Manager - Product Licensing, AREVA Inc.,

2101 Horn Rapids Road, Richland, Washington 99354.

Desk The changes proposed herein do not affect the conclusions of the Significant Hazards Consideration and the Environmental Consideration evaluations provided in the Reference 1 LAR.

In accordance with 10 CFR 50.91 (b), a copy of this application supplement is being provided to the designated Minnesota Official without enclosures.

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

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: November 11, 2014 Site Vice-President Monticello Nuclear Generating Plant Northern States Power Company-Minnesota Enclosures (6) cc: Administrator, Region Ill, USNRC Project Manager, Monticello Nuclear Generating Plant, USNRC Resident Inspector, Monticello Nuclear Generating Plant, USNRC Minnesota Department of Commerce (w/o enclosures)

L-MT-14-088 License Amendment Request Supplement Technical Specification Update For AREVA ATRIUM 10XM Fuel Transition 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 TS 2.1.1, Reactor Core SLs (Safety Limits) 2.2 TS 5.6.3, Core Operating Limits Report (COLR) 3.0 EVALUATION OF PROPOSED CHANGES 3.1 TS 2.1.1, Reactor Core SLs (Safety Limits) 3.2 TS 5.6.3, Core Operating Limits Report (COLR)

4.0 REGULATORY EVALUATION

AND ENVIRONMENTAL EVALUATION

5.0 REFERENCES

Page 1 of 5

L-MT-14-088 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, submitted Reference 5.1 to request approval of an amendment to the Monticello Nuclear Generating Plant (MNGP) Renewed Operating License (OL) and Technical Specifications (TS). The changes proposed in Reference 5.1 would allow reactor operation with AREVA ATRIUM 10XM fuel and allow use of AREVA safety analysis methods for the development of core operating limits. To accomplish said changes, Reference 5.1 included:

• Markup of TS 2.1.1 to change the value of steam dome pressure to align the safety limit with the proposed AREVA safety analysis methodology.

• Markup of TS 5.6.3 to include Revision 0 of the AREVA ACE Critical Power Correlation.

The purpose of this supplement is to propose non-technical updates to the markups of TSs 2.1.1 and 5.6.3 that were proposed in Reference 5.1. These changes are necessary to accommodate related NRC licensing activities that have transpired during the timeframe of the NRC review of Reference 5.1. The proposed update to TS 2.1.1 is a restructuring of the steam dome pressure safety limit TS to accommodate the original change proposed in Reference 5.1 and a similar change that was proposed to the same TS in Reference 5.2. Also, the proposed update to TS 5.6.3 would revise the list of approved methodologies to adopt Revision 1 of the AREVA ACE Critical Power Correlation, which was recently approved by NRC in Reference 5.3.

2.0 DETAILED DESCRIPTION In order to reconcile the TSs proposed in Reference 5.1 with the licensing changes approved by Reference 5.3 and the pending licensing changes proposed by Reference 5.2, several updates to the Technical Specification markups are required. The proposed changes are as follows:

2.1 TS 2.1.1, Reactor Core SLs (Safety Limits)

In Reference 5.2, NSPM provided a markup of TS 2.1.1 to reduce the value of reactor steam dome pressure in TS 2.1.1.1 and 2.1.1.2 from 785 pounds per square inch gauge (psig) to a value of 686 psig. This value of reactor steam dome pressure is appropriate for the General Electric - Hitachi (GEH) safety analysis methodology.

Subsequently, in Reference 5.1, NSPM provided a markup of TS 2.1.1 to reduce the value of reactor steam dome pressure in TS 2.1.1.1 and 2.1.1.2 Page 2 of 5

L-MT-14-088 from 785 psig to an even lower value of 586 psig, a value that is appropriate for the AREVA safety analysis methodology. A value below the GEH-methodology limit of 686 psig is necessary for AREVA safety analysis methodology because AREVA found that the steam dome pressure for one particular transient can drop below 686 psig when using AREVA transient methods. This is described further in Reference 5.1.

Thus, two pending submittals are competing for the same value.

Unfortunately, neither of the proposed values would accurately cover current operations (using GEH methodology) as well as the future fuel transition operations (using AREVA methodology). To accommodate this condition, NSPM is proposing to restructure TS 2.1.1 to require the use of the GEH safety limit when operating under GEH safety analysis methods and use of the AREVA safety limit when operating under the AREVA safety analysis methods. The method used for safety analysis (whether GEH or AREVA) is established during the core reload safety analysis process that precedes any particular fuel cycle. In that same reload process, the appropriate operating limits for that analysis methodology are provided in the Core Operating Limits Report (COLR).

The changes are manifest in the markup provided in Enclosure 2 to this supplement.

2.2 TS 5.6.3, Core Operating Limits Report (COLR)

In Reference 5.1, NSPM proposed to add Revision 0 of the AREVA ACE Critical Power Correlation to the list of approved safety analysis methods used for developing core operating limits. At the time that the license amendment request (LAR) was submitted, Revision 1 had not yet been approved, so the LAR requested Revision 0 with one small modification related to the definition of the radial peaking function used to calculate the assembly K-factor. This correction is incorporated into the approved Revision 1, such that there is no difference between the submitted methodology and the revised method.

Thus, the proposed revision to TS 5.6.3 involves a change from ACE Revision 0, dated March 2010, to ACE Revision 1, dated March 2014. This change is manifest in item number 20 in the list of analysis methodologies (see Insert #1 of Enclosure 2 to this supplement).

Page 3 of 5

L-MT-14-088 3.0 EVALUATION OF PROPOSED CHANGES 3.1 TS 2.1.1, Reactor Core SLs (Safety Limits)

The proposed change to TS 2.1.1 is justified on the following bases:

a. The proposed change to TS 2.1.1 is non-technical in nature. It retains the safety limit values that were previously provided and justified in the respective submittals (References 5.1 and 5.2).

b. The proposed change uses an Improved Standard Technical Specification logical connector (Reference 5.4, Section 1.2) to correctly distinguish which safety limits apply when using the respective safety analysis methodology. The indentations of the markup put a clear OR operator between the GEH methods and the AREVA methods to indicate that either condition will impose the prescribed action. This structure was important to establish because the sub-items include other logical operators (or, and) that would have otherwise disturbed the intended logic of the specification.

c. The proposed markup was drafted so that it may be implemented before, after, or concurrent with the pending amendment for Reference 5.2. No further supplement would be required.

d. There is precedent in pressurized water reactor technical specifications for including several safety limits in one specification, each one related to a different safety analysis methodology. Reference 5.5.

3.2 TS 5.6.3, Core Operating Limits Report (COLR)

The proposed change to TS 5.6.3 is justified on the following bases:

a. AREVA ACE Revision 1 (Reference 5.3) incorporates the K-factor correction that had previously been proposed as an adjunct to Revision

0. Thus, adopting Revision 1 provides for a more consolidated licensing basis.

b. The NRC stated that ACE Revision 1 was acceptable for referencing in licensing applications for nuclear power plants to the extent specified and under the limitations specified in the topical report and the enclosed final safety evaluation.

c. NSPM has confirmed that the conditions of applicability for the use of ACE Revision 1 apply to MNGP.

4.0 REGULATORY EVALUATION

AND ENVIRONMENTAL EVALUATION The proposed changes are intended to facilitate approval and implementation of the AREVA fuel transition with consideration of other licensing activities discussed above. These changes are non-technical in nature, and only affect those Page 4 of 5

L-MT-14-088 Technical Specifications that had been previously annotated for change with the MNGP transition to AREVA Fuel (Reference 5.1). The TS revisions proposed herein are supported by the Regulatory Evaluation and Environmental Considerations provided in Reference 5.1. Further, the original Significant Hazards Consideration bounds the changes proposed herein.

5.0 REFERENCES

5.1 Letter from M A Schimmel (NSPM), to Document Control Desk (NRC), License Amendment Request for Transition to AREVA ATRIUM 10XM Fuel and AREVA Safety Analysis Methodology, L MT-13-055, dated July 15, 2013. (ADAMS Accession No. ML13200A185) 5.2 Letter from John C. Grubb (NSPM), to Document Control Desk (NRC), License Amendment Request: Reduce the Reactor Steam Dome Pressure Specified in the Reactor Core Safety Limits, L-MT-13-010, dated March 11, 2013. (ADAMS Accession No. ML13074A811) 5.3 AREVA Topical Report, ACE/ATRIUM 10XM Critical Power Correlation, ANP-10298P-A, Revision 1, dated March 2014.

5.4 MNGP Technical Specifications, Amendment 181.

5.5 NRC letter to Arkansas Nuclear One, Arkansas Nuclear One, Unit 1 - Issuance of Amendment RE: Revision to Technical Specification 2.1.1.1, Reactor Core Safety Limits (TAC No. MF2277). (ADAMS Accession No. ML14169A475)

Page 5 of 5

L-MT-14-088 Enclosure 2 Marked-Up Technical Specification Pages Note: These markups are made to the latest revision of the approved and amended MNGP Technical Specifications (Amendment 181). Thus, the insert to TS Section 5.6.3 now starts at number 5 (whereas the original markup started at 6).

4 pages follow 2.0-1 5.6-2 Insert (2 pages)

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow

< 10% rated core flow:

THERMAL POWER shall be 25% RTP.

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10% rated core flow:

MCPR shall be 1.15 for two recirculation loop operation or 1.15 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be 1332 psig.

2.2 SL VIOLATIONS With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

(GEH methods),

OR With the reactor steam dome pressure < 586 psig or core flow < 10%

rated core flow (AREVA methods), then (GEH methods),

OR With the reactor steam dome pressure 586 psig and core flow 10%

rated core flow (AREVA methods), then Monticello 2.0-1 Amendment No. 146, 165

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT (COLR) (continued)

4. Control Rod Block Instrumentation Allowable Value for the Table 3.3.2.1-1 Rod Block Monitor Functions 1.a, 1.b, and 1.c and associated Applicability RTP levels;

5. Reactor Protection System Instrumentation Delta W value for Table 3.3.1.1-1, Function 2.b, APRM Simulated Thermal Power - High, Note b; and

6. The Manual Backup Stability Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region II), the modified APRM Simulated Thermal Power - High setpoints used in the OPRM (Function 2.f), Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel"; and

2. (Not Used.)

3. (Not Used.)

Insert new items 4. NEDO-33075-A, Revision 6, "General Electric Boiling Water Reactor shown in Insert #1 Detect and Suppress Solution - Confirmation Density," January 2008.

The COLR will contain the complete identification for each of the Technical Specification referenced topical reports used to prepare the COLR (i.e.,

report number, title, revision, date, and any supplements).

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

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

Monticello 5.6-2 Amendment No. 146, 159, 175, 180

Insert #1 to TS 5.6.3.b Page 1 of 2

5. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, March 1984.

6. EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, February 1998.

7. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, May 1995.

8. XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors -Neutronic Methods for Design and Analysis, March 1983.

9. XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, June 1986.

10. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, October 1999.

11. XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, January 1987.

12. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, February 1987.

13. ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, August 1990.

14. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, September 2009.

15. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, August 2000.

16. EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, May, 2001.

17. EMF-2292(P)(A) Revision 0, ATRIUMTM-10: Appendix K Spray Heat Transfer Coefficients, September 2000.

18. EMF-CC-074(P)(A) Volume 4 Revision 0, BWR Stability Analysis - Assessment of STAIF with Input from MICROBURN-B2, August 2000.

Insert #1 to TS 5.6.3.b Page 2 of 2

19. BAW-10247P-A Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, February 2008.

20. ANP-10298P-A Revision 1, ACE/ATRIUM 10XM Critical Power Correlation, March 2014.

21. ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, June 2011.

22. BAW-10255PA Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, May 2008.

L-MT-14-088 Enclosure 3 Marked-Up Technical Specification Bases Pages Note: These markups are made to the latest revision of the approved and amended MNGP Technical Specifications Bases. They supersede the markups to TS Bases Section 2.1.1 that were provided in the License Amendment Request submitted July 15, 2013. These markups are constructed so as to not incorporate or overlap with the TS markups provided for the License Amendment Request submitted March 11, 2013.

4 pages follow B2.1.1-2 B2.1.1-3 B2.1.1-4 B2.1.1-5

Reactor Core SLs B 2.1.1 BASES BACKGROUND (continued) reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

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

Establishment of Emergency Core Cooling System initiation setpoints higher than this SL provides margin such that the SL will not be reached or exceeded.

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

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

Specifications 2.1.1.1 and 2.1.1.2 are each written to describe a Safety Limit that is appropriate for the respective safety analysis method being used to operate the reactor. This TS construction was approved to provide flexibility during the MNGP transition from operating under the General Electric - Hitachi (GEH) safety analysis methods to the AREVA safety analysis methods. Separate SLs were required because no single value of steam dome pressure would accurately cover the applicability range of GEH methodology as well as the applicability range of AREVA methodology during fuel transition operations. To accommodate this transition, TS 2.1.1.1 and 2.1.1.2 are structured to require the use of the GEH safety limit when operating under GEH safety analysis methods and use of the AREVA safety limit when operating under the AREVA safety analysis methods. The method used for safety analysis (whether GEH or AREVA) is established during the core reload safety analysis process that precedes any particular fuel cycle. In that same process, the appropriate operating limits for that analysis methodology are provided in the Core Operating Limits Report (COLR).

Monticello B 2.1.1-2 Revision No. 4

Reactor Core SLs B 2.1.1 APPLICABLE SAFETY ANALYSES (continued) 2.1.1.1 Fuel Cladding Integrity AREVA critical power correlations (ACE and SPCB) are applicable at pressures > 586 psig and core flows > 10% of rated flow. AREVA correlations are used for cores analyzed with AREVA safety analysis methods, with the ACE correlation used for AREVA fuel and the SPCB correlation used for co-resident fuel.

GE critical power correlations are applicable for all critical power calculations at pressures 785 psig and core flows 10% of rated flow.

For operation at low pressures or low flows, another basis is used, as follows:

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be > 4.56 psi. Analyses (Ref. 2) show that with a bundle flow of 28 x 103 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.56 psi driving head will be > 28 x 103 lb/hr. Full scale ATLAS test data taken at pressures from 0 psig to 785 psig indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER

> 50 % RTP. Thus, a THERMAL POWER limit of 25% RTP for reactor pressure < 785 psig or < 10% core flow is conservative.

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

Monticello B 2.1.1-3 Revision No. 4

Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued)

The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power.

For operating cycles using AREVA safety analysis methods, the probability of the occurrence of boiling transition is determined using the approved AREVA correlations. For such operating cycles, References 8, 9, 10, and 11 describe the uncertainties and methodologies used in determining the MCPR SL.

For operating cycles using GEH safety analysis methods, Tthe probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity SL calculation are given in Reference 2. Reference 3 includes a tabulation of the uncertainties used in the determination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis.

2.1.1.3 Reactor Vessel Water Level During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active irradiated fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

Monticello B 2.1.1-4 Revision No. 4

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

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

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

SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential for VIOLATIONS radioactive releases in excess of 10 CFR 50.67, Accident source term, limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

REFERENCES 1. USAR, Section 1.2.2.

2. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (revision specified in Specification 5.6.3).

3. NEDE-31152P, "General Electric Fuel Bundle Designs," Revision 8, April 2001.

4. 10 CFR 50.67.

5. Reserved.

6. Reserved.

7. Reserved.

8. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.

9. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporations Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.

10. ANP-10298P-A Revision 1, ACE/ATRIUM 10XM Critical Power Correlation, AREVA, March 2014.

11. ANP-10307PA, Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.

Monticello B 2.1.1-5 Revision No. 4

L-MT-14-088 Enclosure 5 AREVA Report FS1-001-8933 Non-Proprietary Error Evaluation for Monticello LOCA Break Spectrum Analyses Revision 1.0 14 pages follow

IDENTIFICATION REVISION FS1-0018933 1.0 AREVA Front End BG Fuel BU TOTAL NUMBER OF PAGES: 14 Error Evaluation for Monticello LOCA Break Spectrum Analyses (Non-Proprietary Version)

ADDITIONAL INFORMATION:

This is the Non-Proprietary Version of FS1-0018932.

The Source Reference Record for this document is FS1-0019021.

PROJECT Monticello-1 (USA010) DISTRIBUTION TO PURPOSE OF DISTRIBUTION HANDLING Restricted AREVA CATEGORY DTR - Data Report STATUS This document is electronically approved. Records regarding the signatures are stored in the Fuel BU Document Database. Any attempt to modify this file may subject employees to civil and criminal penalties. EDM Object Id: F - Release date (YYYY/MM/DD) :  [Western European Time]

RoleROLES This Name text shall NAMES no be visible - Adjust frames overDate DATES to ensure signature ORGANIZATIONS (YYYY/MM/DD) block will Organization completely cover this text SIGNATURES

ULWHU ,1*+$0-HUU\  $5(9$,QF

5HYLHZHU AUTHOR 0817($18,RQ INGHAM Jerry  $5(9$,QF

$SSURYHU 0(*,11,6$ODQ  $5(9$,QF

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APPROVER SCHNEPP Robert; MEGINNIS Alan This text shall no be visible - Adjust frames over to ensure signature block will completely cover this text Exportkennzeichnung AL: 0E001 ECCN: 0E001 RELEASE DATA: Die mit "AL ungleich N" gekennzeichneten Güter unterliegen bei der Ausfuhr aus der EU bzw.

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reviewed or approved by the following regions: Germany: N

CW01L Rev. 4.1 - 10/02/14

N° FS1-0018933 Rev. 1.0 Error Evaluation for Monticello LOCA Break Handling: Restricted AREVA Page 2/14 Spectrum Analyses (Non-Proprietary Version)

REVISIONS REVISION DATE EXPLANATORY NOTES 1.0 See 1st page New document release date AREVA - Fuel BU This document is subject to the restrictions set forth on the first or title page

N° FS1-0018933 Rev. 1.0 Error Evaluation for Monticello LOCA Break Handling: Restricted AREVA Page 3/14 Spectrum Analyses (Non-Proprietary Version)

TABLE OF CONTENTS

1. INTRODUCTION .............................................................................................................................. 5

2. ERRORS IN LOCA BREAK SPECTRUM REPORT ......................................................................... 5

3. ERRORS IN LOCA BREAK SPECTRUM ANALYSES ..................................................................... 5

4. EVALUATION OF ERRORS IN LOCA ANALYSES SUPPORTING THE FUEL TRANSITION........ 7

5. LOCA ANALYSES SUPPORTING THE OPERATION OF AREVA FUEL AT EFW ......................... 8 LIST OF TABLES Table 1: Change in PCT for ATRIUM' 10XM Fuel Supported by EXEM BWR-2000 Methodology...........9 LIST OF APPENDICES Appendix A: Error Evaluation for Break Spectrum (Reference 1) .............................................................10 Appendix B: Error Evaluation for Responses to SRXB RAI-2 (Reference 4) ...........................................13 REFERENCES

1. ANP-3211(P) Revision 1, Monticello EPU LOCA Break Spectrum Analysis for ATRIUM' 10XM Fuel, AREVA NP, July 2013.

2. ANP-3212(P) Revision 0, Monticello EPU LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM' 10XM Fuel, AREVA NP, May 2013.

3. License Amendment Request for Transition to AREVA ATRIUM 10XM Fuel and AREVA Safety Analysis Methodology, July 15, 2013, MNGP L-MT-13-055, ML13200A185.

4. ANP-3286P Revision 0, Responses to RAI from SRXB on MNGP Transition to AREVA Fuel, AREVA Inc., January 2014.

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N° FS1-0018933 Rev. 1.0 Error Evaluation for Monticello LOCA Break Handling: Restricted AREVA Page 4/14 Spectrum Analyses (Non-Proprietary Version)

5. Letter from Xcel Energy to USNRC, AREVA ATRIUM 10XM Fuel Transition - Response to Request for Additional Information (TAC MF2479), January 31, 2014, MNGP L-MT-14-003, ML14035A298.

6. FS1-0015976 Revision 1, Notification of Error in AREVA Report ANP-3211(P) Revision 1 Under NRC Review for Monticello Fuel Transition, April 8, 2014.

7. EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001.

8. ANP-3295P Revision 2, Monticello Licensing Analysis For EFW (EPU/MELLLA+), AREVA Inc.,

September 2014.

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N° FS1-0018933 Rev. 1.0 Error Evaluation for Monticello LOCA Break Handling: Restricted AREVA Page 5/14 Spectrum Analyses (Non-Proprietary Version)

1. Introduction AREVA LOCA analyses for Monticello are documented in References 1 and 2. Xcel Energy submitted these reports to the NRC in support of the transition to AREVA fuel and safety analysis methods (Reference 3). The NRC requested additional information and AREVA responses were reported in Reference 4, which Xcel Energy provided to the NRC in Reference 5. Three of the SRXB RAIs were associated with the AREVA LOCA analyses; RAI-1, RAI-2 and RAI-3. The LOCA analyses summarized in References 1, 2, and 4 were produced under AREVAs 10 CFR 50 Appendix B Quality Assurance Program and continue to be subject to the AREVA corrective action program.

2. Errors in LOCA Break Spectrum Report AREVA has identified several errors in the Monticello LOCA report (Reference 1). These are cases where there is no error in the analyses, but the wrong value was reported.

x Condition Report 2014-1564 identified that the maximum planar average metal water reactions (MWR) were incorrectly reported in Table 6.1 of Reference 1. The maximum planar average MWR was 1.17% but it was incorrectly reported as 1.12%. This error was previously reported in Reference 6 which Xcel Energy provided to the NRC in Reference 5.

x Condition Report 2014-6600 identified that two PCTs are incorrectly reported in Reference 1. [

]

The impact of errors in the LOCA break spectrum analyses are evaluated below and compared to the correct reporting of the analyses described above.

Reference 2 presents the results of exposure dependent MAPLHGR analyses. These analyses continue to be subject to the AREVA corrective action program, but at this time there are no known errors.

3. Errors in LOCA Break Spectrum Analyses AREVA has identified several errors in the Monticello LOCA analyses reported in References 1 and 4.

These errors have been investigated in the AREVA corrective action program. The NRC was informed of these LOCA errors during a meeting in Rockville Maryland on October 1, 2014. The appropriate method for reporting the impact of these errors was discussed. Since the AREVA LOCA analyses are not the analysis of record for Monticello at this time, 10 CFR 50.46 is not technically applicable. However, since the NRC review of these LOCA analyses is still open, it is in the interest of all parties that the NRC receives an evaluation of the errors as soon as practical. Accepting a 50.46 style of evaluation based on the analyses performed in support of the AREVA corrective action program would enable the NRC to AREVA - Fuel BU This document is subject to the restrictions set forth on the first or title page

N° FS1-0018933 Rev. 1.0 Error Evaluation for Monticello LOCA Break Handling: Restricted AREVA Page 6/14 Spectrum Analyses (Non-Proprietary Version) receive the evaluation of the errors sooner. The NRC reviewer indicated that a 50.46 style of reporting is acceptable as long as it is provided before the end of the fuel transition review. The reported PCTs are estimates consistent with requirements of 10 CFR 50.46.

There are three errors associated with [

]

Reference 7 describes the NRC approved AREVA BWR LOCA methodology. [

]

x Condition Report 2014-3971 identified [

] This error did not impact the limiting break conditions.

x Condition Report 2014-4011 identified [

] This error did not impact the limiting break conditions.

x Condition Report 2014-4024 identified [

]

However, this error did not impact the time used in any of the calculations.

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N° FS1-0018933 Rev. 1.0 Error Evaluation for Monticello LOCA Break Handling: Restricted AREVA Page 7/14 Spectrum Analyses (Non-Proprietary Version) x Condition Report 2014-5101 identified [

] Therefore, this error did not impact the limiting break conditions.

x Condition Report 2014-5818 identified [

]

At the time of the writing of this document there are two additional potential errors in the discovery stage of AREVAs corrective action program. These potential errors are not expected to impact the current limiting break conditions. Until the NRC review of these LOCA analyses is complete, they will be subject to the AREVA corrective action program and errors will be evaluated and reported in a similar 50.46 style report. After the NRC review of these LOCA analyses is complete, they will be subject to the normal reporting specified in 50.46.

4. Evaluation of Errors in LOCA Analyses Supporting the Fuel Transition A typical 50.46 report would evaluate the impact of errors on the limiting PCT reported in the exposure dependent MAPLHGR analysis (Reference 2). Since these errors impact the PCTs reported for non-limiting break conditions in reports currently under review by the NRC, the impact of the errors is being evaluated for the PCTs reported in References 1 and 4. The impact of correcting each of these errors for the limiting break conditions is summarized in Table 1. The computer code versions used to evaluate the LOCA errors identified above are fully qualified per AREVA QA requirements for codes used to perform licensing calculations.

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In Reference 1, the errors impact the PCTs presented in Tables 6.3 and 6.4. The tables in Appendix A compare the reported PCTs with the corrected PCTs. In Table 6.3, the reported PCTs are presented on the left side of the column and the corrected PCTs are presented on the right side of the column. In Table 6.4, the reported PCTs are presented on the top of the row and the corrected PCTs are presented on the bottom of the row.

In Reference 4, the errors impact the PCTs presented in response to SRXB RAI-2. The response to SRXB RAI-2 presents PCTs in two unnumbered tables. The tables in Appendix B compare the reported PCTs with the corrected PCTs. In both tables, the reported PCTs are presented on the left side of the column and the corrected PCT are presented on the right side of the column.

The evaluations performed in support of the corrective action program indicate [

]. These analyses continue to be subject to the AREVA corrective action program.

5. LOCA Analyses Supporting the Operation of AREVA Fuel at EFW In support of operating AREVA fuel at Extended Flow Window (EFW) conditions, additional LOCA break spectrum analyses [ ]. The results of these analyses were presented in Section 6.1 of Reference 8. The analysis errors described above were identified and corrected under AREVAs 10 CFR 50 Appendix B Quality Assurance Program in support of Reference 8. The two additional potential errors will be addressed under the AREVA corrective action program.

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Table 1: Change in PCT for ATRIUM' 10XM* Fuel Supported by EXEM BWR-2000 Methodology

• ATRIUM is a trademark of AREVA Inc.

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Appendix A: Error Evaluation for Break Spectrum (Reference 1)

Table 6.3: TLO Recirculation Line Break Spectrum Results for 102% Power [ ] SF-LPCI

[ * ]

[ ]

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Appendix A: Error Evaluation for Break Spectrum (Reference 1)

(Continued)

Table 6.3: TLO Recirculation Line Break Spectrum Results for 102% Power [ ] SF-LPCI (Continued)

[ * ]

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Appendix A: Error Evaluation for Break Spectrum (Reference 1)

(Continued)

Table 6.4: Summary of TLO Recirculation Line Break Results Highest PCT Cases

[ * ]

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Appendix B: Error Evaluation for Responses to SRXB RAI-2 (Reference 4)

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Appendix B: Error Evaluation for Responses to SRXB RAI-2 (Reference 4)

(Continued)

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L-MT-14-088 Enclosure 6 AREVA Affidavit 3 pages follow

AFFIDAVIT STATE OF WASHINGTON )

) ss.

COUNTY OF BENTON )

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

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

3. I am familiar with the AREVA information contained in the Document FS1-0018932, Revision 1, "Error Evaluation for Monticello LOCA Break Spectrum Analyses,"

dated October, 2014 and referred to herein as "Document." Information contained in this Document has been classified by AREVA as proprietary in accordance with the policies established by AREVA 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 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 to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA'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.

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

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

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'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 only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA 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 ~efore me this J_ t/ ~

day ofO ~\, b~ , 2014.

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