L-MT-14-044, ANP-3135NP, Applicability of Areva BWR Methods to Extended Flow Window for Monticello.

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ANP-3135NP, Applicability of Areva BWR Methods to Extended Flow Window for Monticello.
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L-MT- 14-044 4 Enclosure 14 AREVA Report ANP-3135NP (Non-Proprietary)

Applicability of AREVA BWR Methods to Extended Flow Window for Monticello Revision 0 April 2014 25 pages follow

Controlled Document ANP-3135NP Revision 0 Applicability of AREVA BWR Methods to Extended Flow Window for Monticello April 2014 A

AREVA Inc. AR EVA

Controlled Document AREVA Inc.

ANP-3135NP Revision 0 Applicability of AREVA BWR Methods to Extended Flow Window for Monticello

Controlled Document AREVA Inc.

ANP-3135NP Revision 0 Copyright © 2014 AREVA Inc.

All Rights Reserved

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page i Nature of Changes Item Page Description and Justification

1. All This is the initial issue AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page ii Contents 1 .0 In tro d uctio n ................................................................................................................................... 1 2.0 Overview ....................................................................................................................................... 3 2.1 SER Restriction by Topical Report ................................................................................. 7 2.2 EMF-3028P-A Volume 2 Revision 4, "RAMONA5-FA: A Computer Program for BWR Transient Analysis in the Time Domain Volume 2: Theory Manual",

AREVA NP, March 2013 ............................................................................................... 8 2.3 ANP-10262PA Revision 0, "Enhanced Option III Long Term Stability Solution," AREVA NP, May 2008 ................................................................................... 9 2.4 ANP-10298PA Revision 0 Supplement 1P Revision 0 "Improved K-factor Model for ACE/ATRIUM 1OXM Critical Power Correlation," AREVA NP, Decem ber 2011 ............................................................................................................... 11 3.0 Therm al Hydraulics ..................................................................................................................... 12 4.0 AREVA CHF/CPR Correlations ............................................................................................... 12 5.0 Safety Lim it MCPR ....................................................................................................................... 12 6.0 Mechanical Lim its Methodology .............................................................................................. 13 7.0 Core Neutronics .......................................................................................................................... 14 7.1 Bypass Voiding ................................................................................................................ 14 8.0 Transient Analysis ....................................................................................................................... 16 9.0 LOCA Analysis ............................................................................................................................ 16 10.0 Stability Analysis ......................................................................................................................... 17 10.1 Linear Stability ................................................................................................................. 17 10 .2 D IVO M ............................................................................................................................ 17 10 .3 A T W S-I ............................................................................................................................ 17 11.0 Sum m ary .................................................................................................................................... 18 12.0 References ................................................................................................................................. 19 Tables Table 1 AREVA Licensing Topical Reports Addressed in Reference 1 ................................................. 5 Figures Figure 1 Monticello Power Flow Operating Map with the MELLLA+ (EFW ) .......................................... 2 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page iii Nomenclature AOO anticipated operational occurrence APRM average power range monitor ATWS-I anticipated transient without scram instability BWR boiling water reactor CFR Code of Federal Regulations CHF critical heat flux CPR critical power ratio DIVOM delta-over-initial CPR versus oscillation magnitude ECCS emergency core cooling system EO-1i1 Enhanced Option III EPU extended power uprate EFW extended flow window GNF Global Nuclear Fuels LAR License Amendment Request LHGR linear heat generation rate LOCA loss-of-coolant accident LPRM local power range monitor MAPLHGR maximum average planar linear heat generation rate MELLLA maximum extended load line limit analysis MELLLA+ maximum extended load line limit analysis plus MCPR minimum critical power ratio NRC Nuclear Regulatory Commission, U.S.

NSSS nuclear steam supply system OEM original equipment manufacturer OLMCPR operating limit minimum critical power ratio OPRM oscillation power range monitor PBDA period based detection algorithm RAI request for additional information SER Safety Evaluation Report SLMCPR safety limit minimum critical power ratio SPT stability protection trip AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 1 1.0 Introduction This document reviews the AREVA Inc. licensing methodologies to demonstrate that they are applicable to operation of the Monticello Nuclear Generating Plant including Extended Power Uprate (EPU) conditions as well as the Extended Flow Window (EFW). EPU conditions refer to a power uprate to 120%

of the originally licensed rated thermal power. The applicability of the AREVA Inc. licensing methodologies for EPU was addressed in Reference 1. EFW is the AREVA term used to denote the expanded power-flow domain also known as Maximum Extended Load Line Limit Analysis Plus (MELLLA+) as described by Figure 1.

AREVA Inc.

Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 2 C ore Flow (% of Rate( d) 0 10 20 30 40 50 60 70 80 90 100 110 120 120 10l0% CLT'P = 200 Nff 110- Power Flow - - - -- - - - - - - - - - - - - - - -- - -- - - -- - - -

A: 51.-9" 34.2'1% 100".OLTP = 1670bnXt IAL1A BonaI B: 20.8%3991% V Core Flown%

100 = 57.6 Nl~fbnIbr I- 04 100 209

- 439.9% I1 II I .10 2000 X ---

D- 100.0% 99-0% P IGOý0. 7749 'W - O)

P=

90 -- E 100.0% 100.0% --- wliere: PCULP and W=%Core Fl-----

F. 93.3% 1000%-1 G: 37.%. 100.0%' F 80 H: 20.8% 100.0% 1,oLB~ i 1500 1: 83.3% 105.0% '<______

70 J: 37.5% 111.4% L---------- - ---- L peboandan-is deted is: - -- L---

K: 100.0% 105.0% NI

(~4 d P =(22.191 + (0.89714W) - (0.001 19051W~))-1.208 0 60 L I100T.0% 80.0% r. 1 I - hr=OLTP and W=Core Flow M: 82.5% 57.411 0.

N: 70.2% 57.4% I 50 - ------------------ ---- 1-------- - --- 1000 S 3 Flow 'Reewn -~------

0

0. I-w 40 0

U 30 B

-f--

I I

500 T ----1 ---- ----

20 10 0 5 I . , , , 20 0 0 5 !10 '15 20 25 30 35 40 45 50 55 60 65 Core Flow (Mlbm/hr)

Figure 1 Monticello Power Flow Operating Map with the MELLLA+ (EFW)

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 3 2.0 Overview The Safety Evaluation Report (SER) restrictions associated with the AREVA BWR topical reports determined to be applicable for EPU conditions for Monticello were summarized in Section 2.1 of Reference 1. These topical reports are listed in Table 1.

The purpose of this report is to review the SER restrictions and range of applicability associated with the AREVA BWR topical reports determined to be applicable for the core and reactor conditions experienced under EFW conditions and to determine any challenges to the validity of the models. This review identified that for the AREVA BWR topical reports listed in Table 1, there are no SER restrictions on the parameters most impacted by operation in the EFW domain: steam flow, feedwater flow, jet pump M-ratio, and core average void fraction. Additionally, the range of conditions associated with the EFW domain fall within the qualified range of applicability for the applicable AREVA topical reports.

When the reactor power is increased and/or the core flow is decreased, the resultant impact on operating margin is mitigated to a large extent by a decrease in the limiting assembly radial power factor. This decrease in the limiting assembly radial power factor is necessary since the thermal operating limits (MCPR, MAPLHGR and LHGR) that restrict assembly power are dependent on the limiting assembly power but are fairly insensitive to the core thermal power. From this fundamental constraint the following observations may be made about the EFW operating conditions:

1. The reduction in the hot assembly radial peaking factor leads to a more uniform radial power distribution and consequently a more uniform core flow distribution. The net result being less flow starvation of the hottest assemblies.
2. With the flatter radial power distribution, more assemblies and fuel rods are near thermal limits.
3. From a system perspective, there will be higher steam flow and feedwater flow rates at core flows where the core power was previously constrained by the MELLLA operating boundary.
4. With an increase in the average assembly power in the reactor at core flows where the core power was previously constrained by the MELLLA operating boundary, the core pressure drop will increase slightly resulting in a decrease in the jet pump M-ratio for a given core flow rate.
5. Core average and core exit void fractions will increase.
6. Core inlet enthalpy will decrease (core inlet subcooling will increase).

Based on these fundamental characteristics of EFW operation, each of the major analysis domains (thermal-hydraulics, core neutronics, transient analysis, LOCA and stability) are assessed to determine any challenges to EFW application.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 4 The AREVA methodologies are characterized by technically rigorous treatment of phenomena and are very well benchmarked (>100 cycles of operation plus gamma scan data for ATRIUMTM-10*).

The similarity of operating conditions between current and EFW conditions assures that the methods used to compute reactivity and power distributions remain valid. Furthermore, the characteristics computed by the steady-state core simulator and used in safety analyses remain valid.

  • ATRIUM is a trademark of AREVA Inc.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 5 Table 1 AREVA Licensing Topical Reports Addressed in Reference 1 Document Number Document Title XN-NF-79-56(P)(A) "Gadolinia Fuel Properties for LWR Fuel Safety Evaluation" Revision 1 and Supplement 1 XN-75-32(P)(A) "Computational Procedure for Evaluating Fuel Rod Bowing" Supplements 1 through 4 (Base document not approved.)

XN-NF-81-58(P)(A) "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Revision 2 and Model" Supplements 1 and 2 XN-NF-81-51(P)(A) "LOCA-Seismic Structural Response of an Exxon Nuclear Company BWR Jet Pump Fuel Assembly" XN-NF-85-74(P)(A) "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Revision 0 Model" XN-NF-85-92(P)(A) "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results" ANF-89-98(P)(A) Revision 1 "Generic Mechanical Design Criteria for BWR Fuel Designs" and Supplement 1 ANF-90-82(P)(A) Revision 1 "Application of ANF Design Methodology for Fuel Assembly Reconstitution" EMF-85-74(P) Revision 0 "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Supplement I(P)(A) and Model" Supplement 2(P)(A)

EMF-93-177(P)(A) "Mechanical Design for BWR Fuel Channels" Revision 1 BAW-1 0247PA Revision 0 "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors" XN-NF-80-19(P)(A) "Exxon Nuclear Methodology for Boiling Water Reactors -

Volume 1 and Supplements Neutronic Methods for Design and Analysis" 1 and 2 XN-NF-80-19(P)(A) "Exxon Nuclear Methodology for Boiling Water Reactors:

Volume 4 Revision 1 Application of the ENC Methodology to BWR Reloads" AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 6 Document Number Document Title EMF-CC-074(P)(A) "STAIF - A Computer Program for BWR Stability Analysis in the Volume 1 Frequency Domain" and Volume 2 "STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain -

Code Qualification Report" EMF-2158(P)(A) Revision 0 "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/

MICROBURN-B2" EMF-CC-074(P)(A) "BWR Stability Analysis Assessment of STAIF with Input from Volume 4, Revision 0 MICROBURN-B2" BAW-10255PA Revision 2 "Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code" XN-NF-79-59(P)(A) "Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies" XN-NF-80-19(P)(A) "Exxon Nuclear Methodology for Boiling Water Reactors, Volume 3 Revision 2 THERMEX: Thermal Limits Methodology Summary Description" ANF-524(P)(A) Revision 2 "ANF Critical Power Methodology for Boiling Water Reactors" and Supplements 1 and 2 EMF-2245(P)(A) Revision 0 "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel" EMF-2209(P)(A) Revision 3 "SPCB Critical Power Correlation" ANP-10249PA Revision 1 "ACE/ATRIUM-10 Critical Power Correlation" ANP-10298PA Revision 0 "ACE/ATRIUM 1OXM Critical Power Correlation" ANP-10307PA Revision 0 "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors" XN-CC-33(A) Revision 1 "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option Users Manual" XN-NF-80-19(P)(A) "Exxon Nuclear Methodology for Boiling Water Reactors: EXEM Volumes 2, 2A, 2B and 2C BWR ECCS Evaluation Model" XN-NF-82-07(P)(A) "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Revision 1 Model" XN-NF-84-105(P)(A) "XCOBRA-T: A Computer Code for BWR Transient Volume 1 and Volume 1 Thermal-Hydraulic Core Analysis" Supplements 1 and 2 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 7 Document Number Document Title ANF-913(P)(A) Volume 1 "COTRANSA2: A Computer Program for Boiling Water Reactor Revision 1 and Volume 1 Transient Analyses" Supplements 2, 3 and 4 ANF-91-048(P)(A) "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model" ANF-91-048(P)(A) "BWR Jet Pump Model Revision for RELAX" Supplements 1 and 2 EMF-2292(P)(A) Revision 0 "ATRIUMm-10: Appendix K Spray Heat Transfer Coefficients" EMF-2361(P)(A) Revision 0 "EXEM BWR-2000 ECCS Evaluation Model" 2.1 SER Restriction by Topical Report The NRC has approved the following additional licensing topical reports that describe the methods and assumptions used by AREVA to demonstrate the adequacy of BWR fuel system design for Monticello when operating in EFW conditions. These reports address mechanical design criteria, required mechanical and thermal conditions, nuclear design criteria and required fuel and thermal conditions used in licensing analyses, thermal and hydraulic criteria and thermal conditions used in steady-state and transient licensing analyses, AOO and accident analyses. The purpose of each topical report and the restrictions that have been placed on the methods presented are described in the following pages.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 8 2.2 EMF-3028P-A Volume 2 Revision 4, "RAMONA5-FA: A Computer Programfor BWR TransientAnalysis in the Time Domain Volume 2: Theory Manual", AREVA NP, March 2013.

Purpose:

The RAMONA5-FA code is used to calculate delta-over-initial CPR versus oscillation magnitude (DIVOM) slopes in the AREVA stability methodology, Reference 2. Limitations and Conditions 2 and 3 of Reference 2 impose an interim 10 percent penalty on the DIVOM slopes calculated by the RAMONA5-FA methodology in EFW domains until the NRC performed a full code review including constitutive relations, numerics, neutronic methods and benchmarks. The purpose of this report was to support the NRC full code review of RAMONA5-FA.

SER Restrictions: None.

Implementation of SER Restrictions: None.

Observations: This SER removes the interim 10 percent penalty on DIVOM slopes from BAW-10255PA, Rev. 2 safety evaluation in EFW domains.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 9 2.3 ANP-10262PA Revision 0, "EnhancedOption IIl Long Term Stability Solution," AREVA NP, May 2008.

Purpose:

Document and justify the applicability of the Enhanced Option III (EO-Ill) Long Term Stability Solution methodology to the EPU/EFW operating domain.

SER Restrictions:

The NRC staff has concluded that this topical report is acceptable with the limitations and conditions as follows:

1. The NRC staff has not reviewed the hardware and software implementation of EO-Ill Long Term Stability Solution because it will be plant specific. AREVA has stated that implementation is not part of the generic EO-Ill Long Term Stability Solution, even though the EO-Ill Long Term Stability Solution implements an additional scram function (the channel-stability exclusion region) not present in the original Option III platforms. Plant implementations, including those using any original Option III platform, will require plant-specific reviews.
2. The original Option III is already approved for plant operation up to 20 percent EPU. The EO-Ill Long Term Stability Solution is an extension of Option III, where the DIVOM correlation is guaranteed to be well-behaved by the channel-stability exclusion region. Thus, the EO-Ill Long Term Stability Solution is, in essence, an Option III implementation with the added channel-stability exclusion region scram. Therefore, the NRC staff finds that EO-Ill is a technically acceptable methodology for any reactor operating up to 20 percent EPU conditions.
3. The confirmation analyses documented in Section 5 of TR ANP-10262(P), Revision 0, and the response to the NRC staff RAI, indicate that the EO-Ill Long Term Stability Solution methodology provides significant protection against MCPR criteria violations during anticipated instability events even under high-power-density conditions, including EPU and MELLLA+. Under all analyzed conditions, the loss of MCPR margin induced by the instability event is compensated by the gain in MCPR margin induced by the reduction in flow, so that the net MCPR margin is positive. Based on this analysis, the NRC staff finds that the EO-Ill Long Term Stability Solution is a technically acceptable methodology for any reactor operating up to MELLLA+ conditions. Extension of operating domains beyond MELLLA+ have not been considered by the NRC staff and will require a re-evaluation of the EO-Ill Long Term Stability Solution scram effectiveness by the NRC staff.
4. Operation with feedwater heaters out of service (FWHOOS) is not anticipated in EFW like MELLLA+; therefore TR ANP-10262(P), Revision 0, specifies the use of equilibrium feedwater conditions. If a plant-specific application of the EO-Ill Long Term Stability Solution allows for a FWHOOS condition, two Stability Protection Trip (SPT) regions will have to be calculated, with and without FWHOOS. TSs must enforce the change of SPT region settings when the FWHOOS AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 10 condition is declared. Alternatively, a plant-specific application may choose to implement the more conservative of the two SPT regions.

5. The EO-Ill Long Term Stability Solution does not provide an integrated backup stability solution if the primary stability protection system is declared inoperable. Instead, it provides an example of TSs and rationale for their applicability in the responses to the NRC staff RAI. Therefore, the NRC staff review of the stability related TS requirements and/or a different backup stability solution must be performed on a plant-specific basis.
6. Plant-specific applications will include the specification of the backup stability protection. One possible EO-Ill Long Term Stability Solution backup stability protection is the SPT, which provides an automated scram upon entry on the channel-stability exclusion region. The SPT is an acceptable backup stability protection solution for up to 120 days (typical TS range) if the primary stability protection system is declared inoperable. However, the SPT scram region must include the natural circulation line. To be an acceptable backup stability solution, the SPT must include the following scram conditions: (1) recirculation pumps are tripped, or (2) the power flow inside the channel-stability exclusion region. Either of the two conditions should result in scram. In addition to the SPT, administrative interim corrective actions must be enforced with cycle-specific regions. The NRC staff finds that a SPT implemented with the above conditions would provide an acceptable backup stability implementation for up to 120 days under MELLLA+ conditions because it provides protection for the most likely scenarios where large amplitude oscillations could occur.
7. Plant specific applications will include an evaluation of the uncertainty induced by the presence of bypass voids on the OPRM and APRM readings. OPRM uncertainties will result in a set-down of the OPRM PBDA setpoint. APRM uncertainties will be applied to the SPT exclusion region.

Implementation of SER Restrictions: The SER restrictions define requirements to be addressed for the initial licensing application of EO-Ill at each plant. For Monticello;

1. The implementation of EO-Ill for Monticello will require review by the NRC.
2. Operation only up to 20% EPU conditions is in the Monticello licensing basis.
3. Monticello EFW is not an extension beyond MELLLA+ and will not require a resubmittal of the methodology.
4. Operation with FWHOOS is not included in the Monticello licensing basis.
5. The implementation of EO-Ill for Monticello will require review by the NRC.
6. The requirements for SPT to be an acceptable backup stability will be satisfied for Monticello in proposed revisions to TS 3.3.1.1.
7. This is addressed in ANP-3295P (the Licensing Analysis report which supports the EFW LAR).

Observations: The NRC stated that the EO-Ill methodology is technically acceptable for the EPU/MELLLA+

domain. Monticello EFW is the first application of this AREVA topical report.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 11 2.4 ANP-10298PA Revision 0 Supplement IP Revision 0 "Improved K-factor Model for ACE/ATRIUM IOXM CriticalPower Correlation,"AREVA NP, December 2011*.

Purpose:

The K-factor methodology in Section 3.9 of Reference 3 was modified in response to deficiencies found in the axial averaging process and the additive constants were revised as a result of the change to the K-factor model.

SER Restrictions:

1. The ACE/ATRIUM 1OXM methodology may only be used to perform evaluations of AREVA ATRIUM 1OXM fuel design. The ACE/ATRIUM 1OXM correlation may also be used to evaluate the performance of the co-resident fuel in mixed cores as discussed in Section 3.4 of the SE.
2. ACE/ATRIUM 1OXM correlation shall not be used outside the range of applicability defined by the range of test data prescribed in Table 2-1 of Reference 2 of reference 4.

Note: Reference 2 of restriction 2 is ANP-10298PA, Revision 0 (Reference 5).

Implementation of SER Restrictions:

1. SER restriction 1 is implemented in engineering guidelines.
2. SER restriction 2 is directly implemented in the engineering software implementing the ACE correlation, ACELIB, for mass flow, pressure, and inlet subcooling. The correlation bounds are monitored and when appropriate NRC approved actions are taken by the code when the operating conditions fall outside the correlation bounds. For maximum local peaking, the restriction is implemented via neutronics bundle design guidelines.

Observations: This topical report was not generically approved when Reference 1 was issued. Since that time the final Safety Evaluation (Reference 4) was provided.

  • This report has been submitted for approval and is expected to be approved prior to application of this method at Monticello, but the approved version of the Supplement is not available at this time. A final SE has been received from the NRC.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 12 3.0 Thermal Hydraulics AREVA assembly thermal-hydraulic methods are qualified and validated against full-scale heated bundle tests in the KATHY test facility in Karlstein, Germany. The KATHY tests are used to characterize the assembly two-phase pressure drop and CHF performance. This allows the hydraulic models to be verified for AREVA fuel designs over a wide range of hydraulic conditions prototypic of reactor conditions.

The standard matrix of test conditions for KATHY is compared to reactor conditions in Figure 2-1 of Reference 1. This figure illustrates that the test conditions bound typical EFW assembly conditions. The data is based upon the projected EFW operating conditions for the Monticello reactor. Figure 2-1 of Reference 1 also shows that the key physical phenomena (e.g. fluid quality and assembly flows) for EFW conditions are equivalent to current reactor experience.

This similarity of assembly conditions is further enforced in AREVA analysis methodologies by the imposition of SPCB and ACE correlation limits and, therefore, both current core designs and EFW core designs must remain within the same parameter space. Since the bundle operating conditions for EFW are within the envelope of hydraulic test data used for model qualification and operating experience, the hydraulic models used to compute the core flow distribution and local void content remain valid for EFW conditions.

4.0 AREVA CHFICPR Correlations All AREVA CHF and CPR correlations are approved by the NRC staff to be applicable over specified ranges of assembly operating conditions. The NRC staff also approved specific corrective actions when the computed conditions fall outside of the approved range to assure that conservative calculations are obtained. For both EFW and pre-EFW conditions, some analyses can predict assembly conditions to be outside the approved range of specified conditions for the CHF correlations. Consequently, the AREVA licensing methods are programmed to determine whether the computed assembly conditions fall outside of the approved range of applicability for the CHF correlation and impose approved corrective actions as appropriate to conservatively assess the critical power margin for the assembly.

5.0 Safety Limit MCPR The safety limit MCPR (SLMCPR) methodology is used to determine the Technical Specification SLMCPR value that ensures that 99.9% of the fuel rods are expected to avoid boiling transition during normal reactor operation and anticipated operation occurrences. In support of the Monticello fuel transition LAR, the SLMCPR methodology being applied for Monticello is described in References 1 and

6. This methodology remains applicable for the EFW operating conditions.

The SLMCPR is determined by statistically combining calculation uncertainties and plant measurement uncertainties that are associated with the calculation of MCPR. The thermal hydraulic, neutronic, and critical power correlation methodologies are used in the calculation of MCPR. The applicability of these methodologies for EFW conditions is discussed in sections 3.0, 4.0 and 7.0 of this report.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 13 AREVA calculates the SLMCPR on a cycle-specific basis to protect all allowed reactor operating conditions. The analysis incorporates the cycle-specific fuel and core designs. The initial MCPR distribution of the core is a major factor affecting how many rods are predicted to be in boiling transition.

The MCPR distribution of the core depends on the neutronic design of the reload fuel and the fuel assembly power distributions in the core. AREVA SLMCPR methodology specifies that analyses be performed with a design basis power distribution that "... conservatively represents expected reactor operating states which could both exist at the MCPR operating limit and produce a MCPR equal to the MCPR safety limit during an anticipated operational occurrence." (Reference 6, Section 3.3.2).

Commitments resulting from the application of GNF SLMCPR methodology to MELLLA+ are summarized in Reference 7 (Attachment 3, Section 2.2.1 and Appendix B Limitations and Conditions 12.6). These have been reviewed for applicability to the AREVA methodology and implemented as appropriate. AREVA performs SLMCPR calculations for the maximum and minimum licensing core flow at 100% power. In support of the application of AREVA methods for Monticello EFW, SLMCPR calculations were also performed for the minimum core flow on the EFW boundary, Point M on Figure 1. The Monticello specific results are summarized in the Licensing Analysis report (Reference 8) submitted as part of the EFW License Amendment Request (LAR).

6.0 Mechanical Limits Methodology The LHGR limit is established to support plant operation while satisfying the fuel mechanical design criteria. The methodology for performing the fuel rod evaluation is described in Reference 9. Fuel rod design criteria evaluated by the methodology are contained in References 9 and 10. The evaluation addresses operation at EFW because the power history inputs are directly obtained for the projected EFW operating conditions for the Monticello reactor. This operation remains within the range of applicability, in terms of LHGR and burnup, of the methodology while meeting the design limits for the fuel.

Fuel rod power histories are generated as part of the methodology for equilibrium cycle conditions as well as cycle-specific operation. A comprehensive number of uncertainties are taken into account in the categories of operating power uncertainties, code model parameter uncertainties, and fuel manufacturing tolerances. In addition, adjustments are made to the power history inputs for possible differences in planned versus actual operation. Upper limits on the analysis results are obtained for comparison to the design limits for fuel melt, cladding strain, rod internal pressure and other topics as described by the design criteria.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 14 Since the power history inputs, which include LHGR, fast neutron flux, reactor coolant pressure and reactor coolant temperature, are used as input to the analysis, the results explicitly account for conditions at EFW such as higher coolant voiding and offsets in axial power and neutron fast flux. The resulting LHGR limit is used to monitor the fuel so it is maintained within the same maximum allowable steady-state power envelope as analyzed. The mechanical methods remain valid for EFW conditions.

7.0 Core Neutronics The AREVA neutronic methodologies are characterized by technically rigorous treatment of phenomena and are very well benchmarked (>100 cycles of operation plus gamma scan data for ATRIUM-10).

Recent operating experience is tabulated in Table 2-1 of Reference 1. This table presents the reactor operating conditions and in particular the average and hot assembly powers for both US and European applications. As can be seen from this information, the average and peak bundle powers in this experience base exceed that associated with the Monticello EFW application.

The increased steam flow from increased core power, whether at rated core flows or reduced core flows in the EFW, comes from increased power in normally lower power assemblies in the core, operating at higher power levels. For EFW operation, the high powered assemblies will be subject to the same LHGR, MAPLHGR, MCPR, and cold shutdown margin limits and restrictions, as high powered assemblies in pre-EFW cores.

The similarity of operating conditions between current and EFW conditions assures that the neutronic methods used to compute the nodal reactivity and power distributions remain valid. Furthermore, the neutronic characteristics computed by the steady-state simulator and used in safety analysis remain valid.

7.1 Bypass Voiding The steady-state core simulator, MICROBURN-B2, explicitly models the assembly specific flow paths in through the orifice and out through the lower tie-plate flow holes and the channel seals. A [

] through the core support plate and up through the space between the fuel channels. The flow through the water rods/channel is modeled as [

]. The numerical solution for the individual flow paths is computed based on a general parallel channel hydraulic solution that imposes a constant pressure drop across the core fuel assemblies and the bypass region. This solution scheme incorporates [

] that is dependent on the [ ].

The level of bypass boiling for a given state-point is a direct result of the hydraulic solution. The potential for boiling increases as the power/flow ratio increases or the inlet sub-cooling decreases. The MICROBURN-B2 core simulator [ ] to estimate the potential for localized bypass boiling. The capability of the model to predict localized bypass boiling is AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 15 demonstrated in Figure A-39 of Reference 1 for a Monticello case at the worst condition of the design step through at rated thermal power.

CASMO-4 has the capability to specify the density of the moderator in the bypass and in-channel water rods, [

1. Minimal bypass voiding is encountered during full power, steady-state EFW operation so there is no impact on steady-state analyses.

Monticello bypass voiding at 100% power (120% of the original licensed thermal power) was assessed for various core flows. There was minimal localized bypass boiling in the EFW range at 100% power level.

As the flow is reduced along the 100% power line into the EFW domain, the decrease in flow is compensated by increased sub-cooling which more than compensates for the decrease in flow. When flow is further reduced along the highest rod line, boiling in the bypass is calculated to begin. This is in the area of stability concerns where the boiling in the bypass is modeled explicitly. For normal operation at 100% power, boiling in the bypass is not expected to occur, so there is no impact on the lattice local peaking or the LPRM response.

The MICROBURN-B2 state-point specific solution for bypass flow rate and [ ] is then used as initial conditions in the transient analyses. When the reactor operates on high rod-lines at low flow conditions, the in-channel pressure drop decreases to a point where a solid column of water cannot be supported in the bypass region, and voiding occurs in the core bypass. For these conditions (in the region of core stability concerns) the neutronic feedback of bypass voiding [

Bypass voiding is of greatest concern for stability analysis due to its direct impact on the fuel channel flow rates and the axial power distributions. The reduced density head in the core bypass due to boiling results in a higher bypass flow rate and consequently a lower hot channel flow rate. This lower hot channel flow rate results in a more bottom-peaked power distribution due to lower reactivity in the top of the core caused by boiling in the bypass region and higher exit voids. The lower flow and the more bottom-peaked power distribution destabilize the core through higher channel decay ratios. AREVA stability methods directly model these phenomena to assure that the core stability is accurately predicted.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 16 8.0 Transient Analysis Transient analyses are performed to evaluate the system and hot assembly response during potentially limiting events. The AREVA transient analysis methodology applied for the Monticello fuel transition is summarized in Reference 1. This methodology remains applicable for the EFW operating conditions.

In support of the Monticello fuel transition LAR, Reference 1 (particularly Appendices C.1, D, E.3.1, E.3.3, and F) provided additional information on the AREVA transient methodology. From these qualifications and the observation that the nodal hydraulic conditions during EFW are expected to be within the current operating experience, the transient analysis methods remain valid for EFW reactor conditions.

Commitments resulting from the application of GNF transient methodology to MELLLA+ are summarized in Reference 7 (Attachment 3, Section 9.1 and Appendix B Limitations and Conditions 9.9 -9.11). These have been reviewed for applicability to the AREVA methodology and implemented as appropriate. In support of the application of AREVA methods for Monticello EFW, transient analyses were performed to support operation anywhere in the EPU/EFW operating domain, Figure 1. This included transient analyses initiated at the minimum core flow at 100% core power, Point L on Figure 1, and at the minimum core flow on the MELLLA+ boundary, Point M on Figure 1. The results are used to establish thermal limits provided in the Licensing Analysis report (Reference 8) submitted as part of the EFW LAR.

9.0 LOCA Analysis LOCA analyses are performed to establish MAPLHGR limits which assure 10 CFR 50.46 acceptance criteria would be satisfied during the unlikely event of a LOCA. In support of the Monticello fuel transition LAR, the AREVA LOCA analysis methodology applied for the fuel transition is summarized in References 1 and 11. This methodology remains applicable for the EFW operating conditions.

In support of the Monticello fuel transition LAR, Reference 1 (particularly Section 2-34 and Appendix E.3.2) provided additional information on the AREVA LOCA methodology. From these qualifications and the observation that the nodal hydraulic conditions during EFW are expected to be within the current operating experience, the LOCA analysis methods remain valid for EFW reactor conditions.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 17 Commitments resulting from the application of GNF LOCA methodology to MELLLA+ are summarized in Reference 7 (Attachment 3, Section 9.1 and Appendix B Limitations and Conditions 9.7 - 9.8 and 12.10 -

12.14). These have been reviewed for applicability to the AREVA methodology and implemented as appropriate.

In support of the Monticello fuel transition, LOCA break spectrum analysis were performed for the minimum and maximum core flow at 100% core power, Points L and K on Figure 1. In support of the application of AREVA methods for Monticello EFW, LOCA analyses were performed for the minimum core flow on the MELLLA+ boundary, Point M on Figure 1. The analysis considered the same full range of break sizes, break location, break type, and ECCS single failures that were evaluated in support of the fuel transition. The additional EFW LOCA results are summarized in the Licensing Analysis report (Reference 8) submitted as part of the EFW LAR.

10.0 Stability Analysis 10.1 Linear Stability The flatter radial power profile characteristic of MELLLA+ core designs will tend to decrease the first azimuthal eigenvalue separation and result in slightly higher regional decay ratios. These effects are computed by STAIF as it directly computes the channel, global, and regional decay ratio and does not rely on a correlation to protect the regional mode.

STAIF has been benchmarked against full assembly tests (in KATHY facility) to validate the channel hydraulics from a decay ratio of approximately 0.4 to limit cycles. These tests or benchmarks exceed the bounds of allowed operation. These benchmarks include prototypical ATRIUM-10 assemblies. From a reactor perspective, STAIF is benchmarked to both global and regional reactor data as late as 1998, and, therefore, includes current reactor cycle and fuel design elements. This strong benchmarking qualification and the direct computation of the regional mode assure that the impact of the MELLLA+ core designs are reflected in the stability analysis. The Monticello results are summarized in the Licensing Analysis report (Reference 8) submitted as part of the EFW LAR.

10.2 DIVOM RAMONA5-FA has been generically approved for MELLLA+ operation in support of the Enhanced Option III Long Term Stability Solution (References 12 and 13). The Monticello results are summarized in the Licensing Analysis report (Reference 8) submitted as part of the EFW LAR.

10.3 ATWS-I A Monticello specific assessment of the ATRIUM 1OXM fuel for the EFW using the NRC approved NSSS vendor methodology is outside the scope of this document and has been submitted separately (References 14 and 15).

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 18 11.0 Summary This review concluded that there are no SER restrictions on AREVA methodology that are impacted by EFW. Since the EFW core and assembly conditions for Monticello are equivalent to core and assembly conditions of other plants for which the methodology was benchmarked, the AREVA methodology (including uncertainties) remains applicable for EFW conditions at Monticello.

More specifically:

a) The steady state and transient neutronics and thermal-hydraulic analytical methods and code systems supporting EFW are within NRC approved applicability ranges because the conditions for EFW application are equivalent to existing core and assembly conditions in other plants for which the AREVA methodology was benchmarked.

b) The calculational and measurement uncertainties applied in EFW applications are valid because the conditions for EFW application are equivalent to existing core and assembly conditions for which the AREVA methodology was benchmarked.

c) The assessment database and uncertainty of models used to simulate the plant response at EFW conditions are equivalent to core and assembly conditions for which the AREVA methodology was benchmarked.

AREVA Inc.

    • Document Applicability of AREVA BWR ANP-3135NP Methods to Extended Flow Revision 0 Window for Monticello Page 19 12.0 References
1. ANP-3224P Revision 2, Applicability of AREVA NP BWR Methods to Monticello, AREVA NP, June 2013.
2. BAW-1 0255PA Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP, May 2008.
3. ANP-10249PA Revision 1 "ACE/ATRIUM-10 Critical Power Correlation," AREVA NP, September 2009.
4. Letter, Sher Bahadur (U.S. Nuclear Regulatory Commission) to P. Salas (AREVA), "FINAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR TOPICAL REPORT ANP-10298PA, REVISION 0, SUPPLEMENT 1P, REVISION 0, "IMPROVED K-FACTOR MODEL FOR ACE/ATRIUM 1OXM CRITICAL POWER CORRELATION" (TAC NO.

ME7963)", March 31, 2014.

5. ANP-10298PA, Revision 0, ACE/ATRIUM IOXM CriticalPower Correlation,AREVA NP, March 2010.
6. ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.
7. Letter, T.J. O'Connor (MNGP) to U.S. Nuclear Regulatory Commission, "License Amendment Request: Maximum Extended Load Line Limit Analysis Plus," L-MT-10-003, January 21, 2010 (Attachment 3, NEDC-33435P Revision 1, DRF 0000-0089-5121, Safety Analysis Report for Monticello Maximum Extended Load Line Limit Analysis Plus, GE Hitachi Nuclear Energy, December 2009 ). (NRC Accession Number ML100280557)
8. ANP-3295P Revision 0, Monticello Licensing Analysis for EFW (EPU/MELLLA+), AREVA, April 2014.
9. BAW-1 0247PA Revision 0, Realistic Thermal-MechanicalFuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008.
10. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteriafor BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.

M T

11. ANP-3211 (P) Revision 1, Monticello EPU LOCA Break Spectrum Analysis for A TRIUM IOXM Fuel, AREVA NP, July 2013.
12. EMF-3028P-A Volume 2 Revision 4, RAMONA5-FA: A Computer Programfor BWR Transient Analysis in the Time Domain Volume 2: Theory Manual, AREVA NP, March 2013.
13. ANP-10262PA, Revision 0, Enhanced Option Ill Long Term Stability Solution, AREVA NP, May 2008.
14. ANP-3274P Revision 0, Analytical Methods for Monticello ATWS-I, AREVA, April 2014.
15. ANP-3284P Revision 0, Results of Analysis and Benchmarking of Methods for Monticello ATWS-I, AREVA, April 2014.

AREVA Inc.

L-MT-14-044 5 Enclosure 15 GEH Affidavit 3 pages follow

GE-Hitachi Nuclear Energy Americas, LLC AFFIDAVIT I, James F. Harrison, state as follows:

(1) I am Vice President, Fuel Licensing, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas LLC (GEM), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in Attachment I of Enclosure I to Xcel Energy, Monticello Nuclear Generating Plant, letter L-MT-14-044 to the Document Control Desk, (US NRC), Re: License Amendment Request-for AREVA Extended Flow Window.

The GEH proprietary information in Attachment 1 of Enclosure 1 to L-MT-14-044 is identified by double square brackets. ((This sentence is an example. 131)) The superscript notation (3) refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F2.d 871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA. 704 F2.d 1280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;
b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
c. Information which reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;
d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

L-MT-14-044 Attachment 1 of Enclosure 1 Affidavit Page I of 3

GE-Hitachi Nuclear Energy Americas, LLC The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a and (4)b above.

(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEI, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GEH.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains the details regarding stability evaluations performed by GEH for the BWR.

The development of the GEH stability solutions for the BWR was achieved at a significant cost to GEH. The development of the evaluation process along with the interpretation and application of the results is derived firom the extensive experience database that constitutes a major GEH asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

L-MT-14-044 Attachment I of Enclosure 1 Affidavit Page 2 of 3

GE-Hitachi Nuclear Energy Americas, LLC The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the infornation were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

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

Executed on this 6th day of August 2014.

James F. Harrison Vice President, Fuel Licensing Regulatory Affairs GE-Hitachi Nuclear Energy Americas, LLC 3901 Castle Hayne Road Wilmington, NC 28401 james.harrison@ge.com L-MT- 14-044 Attachment 1 of Enclosure 1 Affidavit Page 3 of 3

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