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Attachments 9a, B, 10a, B, 11a, 12a and 13a and 13b, Including ANP-3933NP, Revision 0, Monticello ATWS-I Evaluation for Atrium 11 Fuel, Affidavit
	ML21211A596
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Site:	Monticello
Issue date:	06/30/2021
From:	; Framatome, Xcel Energy
To:	; Office of Nuclear Reactor Regulation
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ML21211A594	List: L-MT-21-044, License Amendment Request: Application to Adopt Advanced Framatome Methodologies; ML21211A596;
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	L-MT-21-044 ANP-3933NP, Rev 0
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ENCLOSURE ATTACHMENT 9a MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST APPLICATION TO ADOPT ADVANCED FRAMATOME METHODOLOGIES ANP-3933NP REPORT, REVISION 0 MONTICELLO ATWS-I EVALUATION FOR ATRIUM 11 FUEL JUNE 2021 (38 pages follow)

Controlled Document ANP-3933NP Monticello ATWS-I Evaluation Revision 0 for ATRIUM 11 Fuel June 2021

0414-12-F04 (Rev. 004, 04/27/2020)

Controlled Document

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page i Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 All Initial Issue

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page ii Contents Page

1.0 INTRODUCTION

............................................................................................... 1-1 2.0 RAMONA5-FA ATWS-I METHODOLOGY ........................................................ 2-1 3.0 SCENARIO IDENTIFICATION .......................................................................... 3-1 3.1 Turbine Trip With Bypass........................................................................ 3-1 3.2 Two Recirculation Pump Trip .................................................................. 3-2 4.0 ATRIUM 11 CPROM CORRELATION............................................................... 4-1 4.1 CPROM Correlation for ATRIUM 11 ....................................................... 4-1 4.2 ATRIUM 11 CPROM Bounds of Applicability ........................................ 4-15 5.0 DISCUSSION OF RESULTS ............................................................................. 5-1 5.1 Turbine Trip With Bypass........................................................................ 5-1 5.2 Two Recirculation Pump Trip .................................................................. 5-1 6.0 COMPLIANCE WITH LIMITATIONS AND CONDITIONS ................................. 6-1

7.0 CONCLUSION

S ................................................................................................ 7-1

8.0 REFERENCES

.................................................................................................. 8-1 List of Tables Table 4-1 [ ] .......................................................... 4-2 Table 4-2 Statistics [ ]............................................... 4-7 Table 4-3 Statistics [ ] .............................................. 4-7 Table 4-4 Statistics [ ] ................................... 4-8 Table 4-5 Statistics [ ] ........................................... 4-8 Table 4-6 Statistics [ ] .......................................... 4-12 Table 4-7 Statistics [ ].......................................... 4-13 Table 4-8 Statistics [ ] ............................... 4-13 Table 4-9 Statistics [ ] ...................................... 4-13 Table 5-1 Monticello EFW ATRIUM 11 TTWB ATWS-I Calculations ........................... 5-2 Table 5-2 Monticello EFW ATRIUM 11 2RPT ATWS-I Calculations ............................ 5-2

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page iii List of Figures Figure 4-1 Calculated versus measured critical power, [ ] ....................... 4-3 Figure 4-2 [ ] ....................................................... 4-4 Figure 4-3 [ ] ...................................................... 4-4 Figure 4-4 [ ]............................................ 4-5 Figure 4-5 [ ] ....................................... 4-5 Figure 4-6 [ ] .............................................................. 4-6 Figure 4-7 Calculated versus measured critical power, [ ] ..................... 4-9 Figure 4-8 [ ] .................................................. 4-10 Figure 4-9 [ ] .................................................. 4-10 Figure 4-10 [ ] ..................................... 4-11 Figure 4-11 [ ] ................................. 4-11 Figure 4-12 [ ] ........................................................ 4-12 Figure 4-13 Calculated versus measured critical power ............................................ 4-14 Figure 5-1 ATRIUM 11 [ ] TTWB Core Power Response .................................................................................................. 5-3 Figure 5-2 ATRIUM 11 [ ] TTWB Core Inlet Subcooling Response................................................................................ 5-4 Figure 5-3 ATRIUM 11 [ ] TTWB Core Inlet Flow Response .......................................................................................... 5-5 Figure 5-4 ATRIUM 11 [ ] TTWB Water Level Response .................................................................................................. 5-6 Figure 5-5 ATRIUM 11 [ ] TTWB Limiting Channel Inlet Flow Results ........................................................................ 5-7 Figure 5-6 ATRIUM 11 [ ] TTWB Limiting Channel PCT Response ............................................................................ 5-8

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page iv Nomenclature Acronym Definition 2RPT Two Recirculation Pump Trip ATWS Anticipated Transient Without Scram ATWS-I Anticipated Transient Without Scram With Instability BOC Beginning of Cycle CPR Critical Power Ratio CPROM Critical Power Reduced Order Model ECPR Experimental Critical Power Ratio EFW Extended Flow Window EOC End of Cycle EOI Emergency Operating Instruction EOP Emergency Operating Procedure EPU Extended Power Uprate L&C Limitation and Condition LHGR Linear Heat Generation Rate PCT Peak Clad Temperature PHE Peak Hot Excess TR Topical Report TTWB Turbine Trip With Bypass

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 1-1

1.0 INTRODUCTION

This document presents the results of calculated BWR instability transients that are not terminated by scram and thus power and flow oscillations are allowed to grow to large amplitudes. This class of transients is referred to as Anticipated Transients Without Scram with Instability (ATWS-I). The code used for simulating this event is RAMONA5-FA, which was developed and approved in Reference 1.

The scope of this analysis covers the Monticello EPU/EFW operating domain using an equilibrium ATRIUM 11 core.

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 2-1 2.0 RAMONA5-FA ATWS-I METHODOLOGY The Framatome generic methodology for the evaluation of Anticipated Transients Without Scram with Instability (ATWS-I) is presented in Reference 1. This report documents a RAMONA5-FA based method for evaluating the fuel specific portion of the event. This method is intended to cover the initial ATWS-I event through the time that operator actions suppress core oscillations. This method is not intended for use in evaluating containment effects or over pressurization during the event.

A key improvement of the RAMONA5-FA code is the addition of [

] This model is described in Section 5.5.5 of Reference 1, and the benchmarking of the ATRIUM 10XM fuel type with this new correlation is given in Appendix A of Reference 1. This correlation was extended to ATRIUM 11 using the procedure provided in Section A.4 of Reference 1, with the results of the benchmarking given in Section 4.0 of this report.

Analyses performed here are in compliance with the calculation procedure defined in Section 8.0 of Reference 1. Acceptance requires the limiting peak clad temperature results remain below 2200 °F (1204 °C). Section 6.0 of this report also provides additional descriptions of how each Limitation and Condition for the topical report was implemented.

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 3-1 3.0 SCENARIO IDENTIFICATION 3.1 Turbine Trip With Bypass The first event, the TTWB, is initiated by a turbine trip. The recirculation pumps are tripped by an ATWS high pressure pump trip. Core flow drops rapidly to natural circulation which drives core power lower. As the core settles at natural circulation, the feedwater temperature begins to decrease due to the loss of extraction steam. The feedwater temperature decrease causes core power to rise. If no action is taken, then core power will rise to the point that it crosses the instability boundary. Without the ability to scram, and if mitigating operator action is sufficiently delayed, oscillations will begin to grow and will reach large magnitudes, accompanied by reverse flow at the inlet of the hot channels. Once the oscillations reach sufficient magnitude, cyclical dryout and rewetting of the cladding surface will begin. If the oscillations continue to grow, they can reach sufficient magnitude that rewetting is no longer possible and large excursions of clad temperature that could potentially challenge acceptance criteria will occur.

In order to prevent large amplitude oscillations plant Emergency Operating Instructions (EOI) or Emergency Operating Procedures (EOP) instruct the operators to reduce water level upon recognition of an ATWS scenario. This reduction in water level has two primary effects. The first is to reduce core flow by reducing the density head on the downcomer side of the loop. The resulting reduction in core flow also serves to reduce core power. At this point in the transient scenario, the cold feedwater is spraying through a steam environment and falling a significant distance until it reaches the water level. Passing through the steam environment heats the feedwater to the point that it approaches saturation. This significantly reduces the core inlet subcooling, which not only directly adds a stabilizing force to the system, but also reduces core power further stabilizing the system. The suppression of oscillations and reduction in power terminate any temperature excursion in the fuel and ends the fuel-specific portion of the ATWS-I event.

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 3-2 For this event, the rate of feedwater temperature decrease and the time of operator action are of primary importance. As such, the values used for these two input assumptions are the same as the current Monticello ATWS-I analysis of record. The feedwater temperature decrease is assumed to start 10 seconds after the valve closure.

Temperature is then assumed to decrease with a 30 second time constant until the final temperature is reached. The time critical operator action for water level reduction is assumed to begin at 90 seconds, consistent with the current analysis of record for Monticello.

3.2 Two Recirculation Pump Trip The two recirculation pump trip event (2RPT) evolves similarly to the TTWB event with two exceptions, the event initiation and the feedwater temperature excursion. Since the 2RPT event does not involve a turbine isolation, the turbine remains online and extraction steam to the feedwater heaters is maintained so the feedwater temperature remains significantly higher than the TTWB event. Because of this, the power excursion in the 2RPT event is mild which results in a less severe event for the same operator intervention times. One difference between the two transient scenarios is that the TTWB event should generate an automatic scram very early in the event leading to earlier identification of the ATWS scenario by the operators. There may not be a scram signal at the initiation of the 2RPT which means the event may be allowed to progress further due to a delayed identification of ATWS by the operator. As such, the 2RPT may become limiting if operator action in the TTWB event occurs fast enough to suppress oscillations prior to dryout. For the evaluations presented in this report, a stable limit cycle without sustained dryout is demonstrated prior to the time of operator action assumed in the current analysis of record for Monticello.

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-1 4.0 ATRIUM 11 CPROM CORRELATION A new dryout correlation was presented in Appendix A of Reference 1. This correlation, named Critical Power Reduced Order Model (CPROM), has been developed based on Framatome correlation development guidelines similar to dryout licensing correlations such as ACE. The CPROM correlation range of applicability is wide [ ] making it well-suited to fitting into transient models of post-dryout that include cyclical dryout and rewetting with possible failure to rewet. CPROM is an integral part of the RAMONA5-FA transient model described in Section 5.5.5 of Reference 1.

4.1 CPROM Correlation for ATRIUM 11

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-2 Table 4-1 [ ]

[

]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-3

[

]

Figure 4-1 Calculated versus measured critical power, [ ]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-4 Figure 4-2 [ ]

Figure 4-3 [ ]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-5 Figure 4-4 [ ]

Figure 4-5 [ ]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-6 Figure 4-6 [ ]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-7 Table 4-2 Statistics [ ]

Table 4-3 Statistics [ ]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-8 Table 4-4 Statistics [ ]

Table 4-5 Statistics [ ]

[ ]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-9 Figure 4-7 Calculated versus measured critical power, [ ]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-10 Figure 4-8 [ ]

Figure 4-9 [ ]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-11 Figure 4-10 [ ]

Figure 4-11 [ ]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-12 Figure 4-12 [ ]

Table 4-6 Statistics [ ]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-13 Table 4-7 Statistics [ ]

Table 4-8 Statistics [ ]

Table 4-9 Statistics [ ]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-14 The figure below shows comparison of the calculated and measured critical power for the combined data set. The mean critical power ratio is [ ] and the standard deviation of the calculated versus measured critical power for the entire database is

[ ] and the number of data points is [ ].

Figure 4-13 Calculated versus measured critical power

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 4-15 4.2 ATRIUM 11 CPROM Bounds of Applicability The bounds of applicability are determined by the data available to benchmark the correlation. For ATRIUM 11, the [

]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 5-1 5.0 DISCUSSION OF RESULTS 5.1 Turbine Trip With Bypass For this evaluation, [ ] are analyzed consistent with the procedure identified in Section 8.0 of Reference 1. An equilibrium ATRIUM 11 core was utilized.

The events were initiated from rated power at the EFW boundary. As discussed in Section 6.0, in order to ensure that these results will be bounding of future cycles, [

]

Table 5-1 gives the TTWB results from the Monticello EFW ATRIUM 11 analysis including the results of the sensitivity analyses. For each case (assuming operator intervention to lower the water level after [ ]), the peak clad temperature (PCT) is provided. Relevant system and limiting channel plots for the limiting case can be found in Figure 5-1 through Figure 5.6.

5.2 Two Recirculation Pump Trip Table 5-2 gives the 2RPT results from the Monticello EFW ATRIUM 11 analyses. The base statepoints analyzed were chosen consistent with those analyzed for the TTWB event. For each case it was assumed that no operator intervention to lower the water level occurred. The results show [

]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 5-2 Table 5-1 Monticello EFW ATRIUM 11 TTWB ATWS-I Calculations Table 5-2 Monticello EFW ATRIUM 11 2RPT ATWS-I Calculations

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 5-3 Figure 5-1 ATRIUM 11 [ ] TTWB Core Power Response

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 5-4 Figure 5-2 ATRIUM 11 [ ] TTWB Core Inlet Subcooling Response

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 5-5 Figure 5-3 ATRIUM 11 [ ] TTWB Core Inlet Flow Response

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 5-6 Figure 5-4 ATRIUM 11 [ ] TTWB Water Level Response

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 5-7 Figure 5-5 ATRIUM 11 [ ] TTWB Limiting Channel Inlet Flow Results

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 5-8 Figure 5-6 ATRIUM 11 [ ] TTWB Limiting Channel PCT Response

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 6-1 6.0 COMPLIANCE WITH LIMITATIONS AND CONDITIONS The SE for the Reference 1 topical report lists seven limitations and conditions that applications of the methodology must satisfy. The following section discusses how the calculations in this package comply with those limitations and conditions.

Limitation and Condition 1 The gap conductance sensitivity shall be repeated or otherwise justified for transitions to new fuel designs.

A gap conductance sensitivity study was performed for the limiting statepoint, [

]. The gap conductance was varied by

[ ]. This sensitivity showed that the PCT for the limiting channel can vary between [

]. This shows the overall impact is [ ] within the available margin for ATRIUM 11 fuel at Monticello.

Limitation and Condition 2 If the acceptance criteria for the first paragraph in Step 3 of Section 8.0 of the TR are met, additional justification must still be provided to demonstrate adequate margin in operator action timing for variations in neutron kinetics response from specific core designs. This justification may be provided by following Steps 3.a through 3.c, as amended by the response to RAI 15, or providing an alternative justification on a plant-specific basis.

In order to ensure that this calculation will bound future core designs, [

]

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 6-2

[

]

ensure that it will remain bounding of future cycles.

Limitation and Condition 3 Plant-specific evaluations that are intended to be bounding of all core designs must be confirmed to provide reasonable assurance that neutron kinetics characteristics such as possible differences in dominant oscillation modes or the potential for multiple oscillation modes to be active simultaneously are bounded by the analysis of record.

As discussed in the compliance to L&C 2, [

] providing for a bounding calculation regardless of oscillation mode.

Limitation and Condition 4 Due to the unique neutron kinetics characteristics associated with transition cycles, all transition cycles must be dispositioned in a manner consistent with Limitations and Conditions #2 and #3.

As discussed in the compliance to L&C 2, [

]. These effects introduce significant conservatism and will bound the effects associated with transition cycles.

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 6-3 Limitation and Condition 5 The ATWS-I analysis must be performed for both the TTWB and 2RPT events during the initial implementation of this methodology, to confirm which event is limiting. Subsequent evaluations may only consider the event determined to be limiting, except which changes are made to the plant design or operation that may affect stability behavior during ATWS, such as: turbine bypass capability, fraction of steam-driven feedwater pumps, and changes expected to significantly increase core inlet subcooling during ATWS events.

Analyses were performed for both TTWB and 2RPT events and the results are reported in Section 2. These results demonstrate that [ ].

Limitation and Condition 6 The steam line and valve modeling options shall be confirmed to accurately capture the expected plant-specific system performance during ATWS-I events.

The steam line and valve models were built based on plant geometry and setpoints as supplied by Xcel Energy. This allows for an accurate representation of the evolution of the event. A review of the system transient plots show reasonable behavior during the event.

Limitation and Condition 7 Plant-specific applications must justify that the selected settings and modeling options are appropriate, including core and vessel nodalization, time step control parameters, and noise parameters. In particular, the modeling should be reasonably consistent with both the characteristics of the plant in question and the validation basis for the RAMONA5-FA ATWS-I methodology as discussed in this SE.

All nodalization is consistent with the nodalization of the benchmarks and sample problem from the original Reference 1 topical report. In addition, the time step control and noise parameters were identical to those used in the topical report. A nodalization study was performed to demonstrate that the selected nodalization is reasonably representative of the plant.

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 7-1

7.0 CONCLUSION

S The results of the Monticello EPU/EFW ATRIUM 11 ATWS-I analyses have been presented. Both the TTWB and the 2RPT event have been simulated. The results of these analyses demonstrate that the limiting peak clad temperatures remain below the 2200 °F (1204 °C) temperature criteria.

Controlled Document Framatome Inc. - ANP-3933NP Revision 0 Monticello ATWS-I Evaluation for ATRIUM 11 Fuel Page 8-1

8.0 REFERENCES

1. ANP-10346P-A Revision 0, ATWS-I Analysis Methodology for BWRs Using RAMONA5-FA, Framatome Inc., October 2019.

ENCLOSURE ATTACHMENT 9b MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST APPLICATION TO ADOPT ADVANCED FRAMATOME METHODOLOGIES AFFIDAVIT FOR ANP-3933P REPORT, REVISION 0 MONTICELLO ATWS-I EVALUATION FOR ATRIUM 11 FUEL JUNE 2021 (3 pages follow)

AFFIDAVIT

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

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

3. I am familiar with the Framatome information contained in the report ANP-3933P Revision 0, "Monticello ATWS-I Evaluation for ATRIUM 11 Fuel," dated June 2021 and referred to herein as "Document." Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome 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 Framatome 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 Framatome to determine whether information should be classified as proprietary:

(a) The information reveals details of Framatome'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 Framatome.

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

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

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

8. Framatome 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.

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

Executed on: June 7, 2021 Alan B. Meginnis Jm~---- p

ENCLOSURE ATTACHMENT 10a MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST APPLICATION TO ADOPT ADVANCED FRAMATOME METHODOLOGIES ANP-3929NP REPORT, REVISION 0 MONTICELLO ATRIUM 11 CONTROL ROD DROP ACCIDENT ANALYSES WITH THE AURORA-B CRDA METHODOLOGY JUNE 2021 (41 pages follow)

Controlled Document ANP-3929NP Monticello ATRIUM 11 Control Rod Revision 0 Drop Accident Analyses with the AURORA-B CRDA Methodology June 2021

0414-12-F04 (Rev. 004, 04/27/2020)

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page i Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 All Initial Issue

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page ii Contents Page

1.0 INTRODUCTION

............................................................................................... 1-1 2.0 REGULATORY BASIS ...................................................................................... 2-1 3.0 INITIAL METHODOLOGY DEMONSTRATION ................................................. 3-1 3.1 Initial Conditions ..................................................................................... 3-1 3.2 Group Pull Sequence .............................................................................. 3-1 3.3 Inoperable Control Rod Positions ........................................................... 3-2 3.4 Time in Cycle .......................................................................................... 3-4 3.5 Group Critical Position ............................................................................ 3-4 3.6 Determination of Static Control Rod Worth ............................................. 3-6 3.7 Transient Evaluation ............................................................................... 3-8 4.0 EVALUATION AGAINST CLADDING FAILURE CRITERIA .............................. 4-1 4.1 High Temperature Cladding Failure ........................................................ 4-1 4.2 PCMI Cladding Failure ............................................................................ 4-3 4.3 Molten Fuel Cladding Failure Threshold ................................................. 4-4 5.0 RADIOLOGICAL CONSEQUENCES ................................................................ 5-1 6.0 SYSTEM PRESSURE AND CPR ...................................................................... 6-1 7.0 CORE COOLABILITY........................................................................................ 7-1 8.0 LIMITATIONS AND CONDITIONS .................................................................... 8-1

9.0 REFERENCES

.................................................................................................. 9-1 APPENDIX A EVALUATION THRESHOLD DETERMINATION .............................. A-1

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page iii List of Tables Table 3.1 Group Pull Sequences............................................................................... 3-1 Table 3.2 Near Critical Range ................................................................................... 3-4 Table 3.3 Group Out Eigenvalues ............................................................................. 3-5 Table 3.4 Selected Rods with Inoperable Control Rods 2nd Group ........................... 3-6 Table 3.5 Selected Rods with Inoperable Control Rods 3rd and 4th Group ................ 3-7 Table 3.6 Initial Conditions ........................................................................................ 3-8 Table 3.7 Maximum Prompt Enthalpy Rise and Total Enthalpy 2nd Group .............. 3-10 Table 3.8 Maximum Prompt Enthalpy Rise and Total Enthalpy 3rd Group............... 3-11 Table 4.1 Assemblies with Fuel Rod High Temperature Failures .............................. 4-1 Table 5.1 Transient Fission Gas Release Fractions .................................................. 5-3 Table 5.2 Total Fission Gas Release Fractions ......................................................... 5-4 Table 5.3 Fuel Rod Failures ...................................................................................... 5-4 Table 7.1 Peak Radial Average Fuel Enthalpy .......................................................... 7-1 Table A.1 Example Evaluation Boundary Characteristics .......................................... A-3

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page iv List of Figures Figure 3.1 Inoperable Rods for 1st Through 4th Group Pulls ....................................... 3-3 Figure 3.2 Map of Assemblies Evaluated for Drop of Control Rod [ ] ............... 3-9 Figure 4.1 Total Enthalpy Versus High Temperature Cladding Failure Threshold All Drops ................................................................................... 4-2 Figure 4.2 High Temperature Nominal and High Burnup for Drops [ ]

and [ ] .............................................................................................. 4-2 Figure 4.3 Minimum Failure Threshold Based on EOC Hydrogen Content ................ 4-3 Figure 4.4 PCMI Cladding Failure Results.................................................................. 4-4 Figure A.1 Data for Evaluation Boundary .................................................................... A-1 Figure A.2 Example Evaluation Boundary for the Monticello ATRIUM 11 Core .......... A-2

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page v Nomenclature Acronym Definition ASME American Society of Mechanical Engineers AST alternate source term BOC beginning of cycle BPWS banked position withdrawal sequence BWR boiling water reactor CFR code of federal regulations CHF critical heat flux CPR critical power ratio CRDA control rod drop accident CWSR cold work stress relief aka SRA CZP cold zero power EFP end of full power EOC end of cycle GDC general design criteria GMUL gas multiplier LBRF licensing basis release fraction LTR licensing topical report LWR light water reactor MOC middle of cycle corresponding to peak reactivity NRC Nuclear Regulatory Commission, U.S.

PCMI pellet clad mechanical interaction RIA reactivity insertion accident RPS reactor protection system SE safety evaluation SER safety evaluation report SRA stress relief annealed SRP standard review plan SSRF steady state release fraction TFGR transient fission gas release USAR updated safety analysis report H transient change in enthalpy H_p prompt enthalpy rise H_tot total enthalpy rise

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 1-1

1.0 INTRODUCTION

The Framatome AURORA-B CRDA methodology has been used to evaluate the Monticello ATRIUM 11 equilibrium fuel cycle (Reference 1). The methodology includes the use of a nodal three-dimensional kinetics solution with both thermal-hydraulic (T-H) and fuel temperature feedback. These models provide more precise localized neutronic and thermal conditions than previous methods to show compliance with regulatory guidance criteria as presented in Regulatory Guide 1.236 (Reference 3). The report summarizes the application of the AURORA-B CRDA methodology (Reference 4) on the Monticello ATRIUM 11 equilibrium cycle.

The control rod drop calculations were performed with the AURORA-B CRDA methodology. All startup sequences were evaluated and no fuel rod failures were identified through middle of cycle. Evaluations of the drops at the end of full power and end of cycle identified potential fuel rod failures in one startup sequence.

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 2-1 2.0 REGULATORY BASIS The current regulatory basis for the acceptance criteria for the Monticello licensing is fuel failure at 170 cal/g and violent expulsion of fuel at 280 cal/g consistent with Reference 5 (SRP 15.4.9, Revision 2). This demonstration evaluation using the methodology of Reference 4 is applied using the criteria of RG 1.236 (Reference 3). It is anticipated that the acceptance and failure criteria of RG 1.236 will be applied in future applications of this methodology at Monticello.

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 3-1 3.0 INITIAL METHODOLOGY DEMONSTRATION The initial application of the AURORA-B CRDA methodology involves sensitivity studies and determination of an evaluation boundary. The determination of the evaluation boundary provided in Appendix A is a demonstration of the process discussed in Reference 4 for Monticello with ATRIUM 11 fuel. This same process will be followed for transition cores.

3.1 Initial Conditions Sensitivity studies are performed [

]

3.2 Group Pull Sequence All allowed pull orders are evaluated such that each control rod group, with the exception of groups 5 and 6, is pulled as the second group as indicated in Table 3.1. The third and fourth groups are assumed to be banked. It is assumed that the first and second groups selected for withdrawal are completely withdrawn prior to pulling control rods in the third group. For clarification since both the first and second groups must be out before the third group, both pull sequences A1234 and A2134 have the same starting control rod pattern for the third group.

Therefore the sequences A1234 and A2143 also cover sequences A2134 and A1243.

Table 3.1 Group Pull Sequences Analyzed Groups for both A and B Group Numbers Sequences 1st and 2nd Groups (1,2), (2,1), (3,4), (4,3) 3rd and 4th Groups (3,4), (4,3), (1,2), (2,1)

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 3-2 3.3 Inoperable Control Rod Positions A maximum of 8 inoperable control rods are allowed for this plant with up to three inoperable per group. The size of the Monticello core does not accommodate more than two inoperable control rods per group per sequence without having the inoperable control rods infringe upon the quarter of the core where the rod drop analysis is conducted. Having inoperable control rods in the quarter of the core where the rod drop analysis is conducted prevents the power in the core from peaking properly in the quarter of interest and has the potential to dampen the severity of rod drop analysis results. Thus, two inoperable control rods were selected for each of the first four groups withdrawn in each sequence. The assignment of inoperable control rods adhered to the separation criteria on group bases.

Given the uniform core configuration for the equilibrium cycle, the prior analyses in Reference 4 being limited by drops based on inoperable rod configurations, and that investigations showed drops without inoperable rods are overall non-limiting in this analysis, only drops with inoperable control rod configurations were fully evaluated. The selected inoperable control rods for drops in the second, third, and fourth groups are identified in Figure 3.1. For each set of inoperable rods, all control rods in the next group are dropped to evaluate the impact of the inoperable rods. Therefore the position of the inoperable control rods was evaluated based on dropping all rods.

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 3-3 Figure 3.1 Inoperable Rods for 1st Through 4th Group Pulls

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 3-4 3.4 Time in Cycle

[

]

3.5 Group Critical Position The first step is to evaluate the end of group or bank position k-effective values to determine where criticality is anticipated to occur for the given control rod withdrawal sequence. The near critical range, determined per Section 7.4.1 of Reference 4, is given in Table 3.2. The calculated k-effective values at the end of groups 1 through 4 for the A and B sequence withdrawals are given in Table 3.3. [

]

Table 3.2 Near Critical Range

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 3-5 Table 3.3 Group Out Eigenvalues

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 3-6 3.6 Determination of Static Control Rod Worth Based on the results of the group worth, static control rod worths were then determined. From the static rod drops, [

] The selected second group rods for transient evaluation are given in Table 3.4 and the third group rods are given in Table 3.5.

Table 3.4 Selected Rods with Inoperable Control Rods 2nd Group

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 3-7 Table 3.5 Selected Rods with Inoperable Control Rods 3rd and 4th Group

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 3-8 3.7 Transient Evaluation The evaluation of each rod drop is performed with the AURORA-B system. The initial pre-rod drop state point is established with the MICROBURN-B2 core simulator. The initial conditions used for the transient calculation are identified in Table 3.6.

Table 3.6 Initial Conditions The channel grouping with a [ ] is used for this analysis.

(Figure 3.2 illustrates the assemblies evaluated for the drop of control rod [ ] .) Once the channel grouping is defined, the power history information is processed to obtain the fuel rod characteristics for use in the RODEX-4 fuel rod mechanical models.

The maximum prompt enthalpy increase for the peak fuel rod and the maximum total enthalpy reported include the application of the uncertainty multiplier of [ ] on the enthalpy increase.

The prompt enthalpy increase along with total enthalpy for the second group drops is given in Table 3.7. Likewise Table 3.8 contains the prompt enthalpy increase and total enthalpy for third group control rods (the third group bounded the fourth group.) Although there are high worth banked drops, the actual nodal enthalpy increase is small for the BOC banked drops compared to drops later in cycle with a top peaked power shape.

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 3-9 Figure 3.2 Map of Assemblies Evaluated for Drop of Control Rod [ ]

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 3-10 Table 3.7 Maximum Prompt Enthalpy Rise and Total Enthalpy 2nd Group

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 3-11 Table 3.8 Maximum Prompt Enthalpy Rise and Total Enthalpy 3rd Group

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 4-1 4.0 EVALUATION AGAINST CLADDING FAILURE CRITERIA 4.1 High Temperature Cladding Failure

[

]

Table 4.1 Assemblies with Fuel Rod High Temperature Failures

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 4-2 Figure 4.1 Total Enthalpy Versus High Temperature Cladding Failure Threshold All Drops Figure 4.2 High Temperature Nominal and High Burnup for Drops

[ ] and [ ]

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 4-3 4.2 PCMI Cladding Failure The ATRIUM 11 fuel is clad with stress relief annealed (SRA) Zircaly-2 cladding. (Framatome uses the term Cold Work Stress Relieved CWSR to refer to SRA material.) Therefore, the SRA low temperature failure threshold is applied.

To establish the minimum failure threshold, the maximum fuel rod nodal hydrogen at end of cycle was tabulated for each assembly using the hydrogen model of Reference 8 [

]

Figure 4.3 Minimum Failure Threshold Based on EOC Hydrogen Content

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 4-4 Since 150 cal/g is the maximum of the failure threshold curve, [

] The rod drops are evaluated with assumed inoperable control rods. [

]

Figure 4.4 PCMI Cladding Failure Results 4.3 Molten Fuel Cladding Failure Threshold

[ ]

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 5-1 5.0 RADIOLOGICAL CONSEQUENCES The dose consequences for the CRDA determined for Monticello are summarized in the USAR.

The licensing basis does evaluation based on ATRIUM 10XM fuel determined that 1,206 (850 8x8 equivalent from the USAR) fuel rods could fail for Monticello. Since the actual number of ATRIUM 11 allowed fuel rod failures has not been determined at this time, it is assumed that the allowed number of failures will be similar to that of ATRIUM 10XM. Therefore, demonstrating that there is significantly less than 1,206 fuel rod failures will confirm that the radiological consequences are bounded by those given in the USAR.

Evaluation of dose consequences for fuel rod failures Since fuel rod failures had been identified, revised release fractions or total release fraction (TOTR) are determined using the Licensing Basis Release Fractions (LBRF) from RG 1.183 conservatively applied as the steady state release fractions (SSRF) with the transient fission gas release (TFGR) as described in RG 1.236. A ratio of the new TOTR to the LBRF used in the original licensing basis is then generated following the method provided in ANP-10333PA.

The transient release terms, from Reference RG 1.236, expressed as a fraction are:

Peak Pellet BU < 50 GWd/MTU:

[(0.26 ) 13]

= 0 100 Peak Pellet BU 50 GWd/MTU:

[(0.26 ) 5]

= 0 100 The total fuel rod release fraction TOTR is dependent on the nuclide group and the enthalpy dependent TFGR average over the 24 nodes of fuel for a full length fuel rod. (If the failure were in a different length fuel rod, the number of axial nodes would be adjusted accordingly.)

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 5-2 Burnup < 50:

[(0.26 ) 13]

= + 24 100 Burnup 50:

[(0.26 ) 5]

= + 24 100

Where,

H fuel enthalpy increase (cal/g)

SSRF is the steady state release fraction

Three multipliers (GMUL) are established in DG 1327 (Reference 10) to be applied to the above TFGR term:

Group GMUL Applied to Stable long lived isotopes (e.g., Kr-85) 1.0 Kr-85 Cs-134 and Cs-137 1.414 Alkali Metals Short-lived radioactive isotopes 0.333 Iodines, nobles, halogens (i.e., I, Xe and Kr noble gases except Kr-85)

As noted above, the LBRF are utilized for Monticello as the SSRF for the respective groups.

For this analysis of the ATRIUM 11 fuel, a maximum of [ ] fuel rods for any control rod drop case exceeded one or more failure criteria. Based on the enthalpy increase, the enthalpy dependent release terms were determined for the fuel rods. The transient fission gas release fractions are provided in Table 5.1 based on nodal values of the peak fuel rod enthalpy increase. [

] The total release fraction and the ratio to the licensing bases release fraction are provided in Table 5.2.

The actual number of fuel rod failures is provided in Table 5.3. [

]

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 5-3

[ ] ; therefore this event remains within the current evaluated dose consequences for Monticello.

Table 5.1 Transient Fission Gas Release Fractions

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 5-4 Table 5.2 Total Fission Gas Release Fractions Table 5.3 Fuel Rod Failures

[ ]

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 6-1 6.0 SYSTEM PRESSURE AND CPR The impact of the CRDA on system pressure was addressed in Reference 4 and does not cause stresses to exceed Emergency Condition (Service Level C), as defined in Section III of the ASME Boiler and Pressure Vessel code. This generic evaluation on the impact of CRDA on system pressure remains applicable for Monticello.

The CPR response was evaluated in Section 7.7 of Reference 4 and results in a conclusion that the CRDA in the power range is [

]

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 7-1 7.0 CORE COOLABILITY Two criteria are identified in Reference 3 for allowable limits with respect to core coolability.

Peak radial average enthalpy < 230 cal/g

The peak fuel temperature in the outer 90 percent of the pellets volume must remain below incipient fuel melting conditions

[

]

Table 7.1 Peak Radial Average Fuel Enthalpy

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 8-1 8.0 LIMITATIONS AND CONDITIONS The SER for the Reference methodology included a number of limitations and conditions. Some of the conditions are from the base AURORA-B AOO methodology (Reference 6) and additions specific to the CRDA are included. The number of the limitations and conditions below is consistent with that found in the AURORA-B CRDA SER.

1. AURORA-B may not be used to perform analyses that result in one or more of its CCDs (S-RELAP5, MB2-K, MICROBURN-B2, RODEX4) operating outside the limits of approval specified in their respective TRs, SEs, and plant-specific license amendment requests (LARs). In the case of MB2-K, MB2-K is subject to the same limitations and conditions as MICROBURN-B2. (This is Conditions 1 of the SE for the base AURORA-B TR. It remains applicable to CRDA analyses for BWRs/2-6.)

This condition is met for application of the AURORA-B CRDA methodology to Monticello.

14. The scope of the NRC staffs approval of AURORA-B does not include the ABWR design. (This is condition 14 of the SE for the base AURORA-B TR. It remains applicable to CRDA analyses for BWRs/2-6.)

This condition is met for Monticello since it is a BWR/3.

20. The implementation of any new methodology within the AURORA-B EM (i.e.,

replacement of an existing CCD) is not acceptable unless the AURORA-B EM with the new methodology incorporated into it has received NRC review and approval. An existing NRC-approved methodology cannot be implemented within the AURORA-B EM without NRC review of the updated EM. (This is a revised version of Condition 20 of the SE for the base AURORA-B TR, rewritten to be specific to the CRDA application. It remains applicable to CRDA analyses for BWRs/2-6.)

The evaluation model will be implemented for Monticello as described in the base AURORA-B and AURORA-B CRDA Topical Reports. No CCD as described in the TR are replaced and therefore the intent of this condition is met.

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 8-2

21. NRC-approved changes that revise or extend the capabilities of the individual CCDs comprising the AURORA-B EM may not be incorporated into the EM without prior NRC approval. (This is Condition 21 of the SE for the base AURORA-B TR. It remains applicable to CRDA analyses for BWRs/2-6)

[

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22. As discussed in Section 3.3.1.5 and Section 4.0 of Reference 6 (the SE for the base AURORA-B TR), the SPCB and ACE CPR correlations for the ATRIUM-10 and ATRIUM-10XM fuels, respectively, are approved for use with the AURORA-B EM.

Other CPR correlations (existing and new) that would be used with the AURORA-B EM must be reviewed and approved by the NRC or must be developed with an NRC-approved approach such as that described in EMF-2245(P)(A), Revision 0, Application of Siemens Power Corporations Critical Power Correlations to Co-Resident Fuel. Furthermore, if transient thermal-hydraulic simulations are performed in the process of applying AREVA CPR correlations to co-resident fuel, these calculations should use the AURORA-B methodology. (This is Condition 22 of the SE for the base AURORA-B TR. It remains applicable to at-power CRDA analyses for BWRs/2-6.)

This condition is met within ANP-10333PA for at power evaluations. The ACE ATRIUM 11 Correlation has been reviewed and approved by the NRC (Reference 7).

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 8-3

23. Except when prohibited elsewhere, the AURORA-B EM may be used with new or revised fuel designs without prior NRC approval provided that the new or revised fuel designs are substantially similar to those fuel designs already approved for use in the AURORA-B EM (i.e., thermal energy is conducted through a cylindrical fuel pellet surrounded by metal cladding, flow in the fuel channels develops into a predominantly vertical annular flow regime, etc.). New fuel designs exhibiting a large deviation from these behaviors will require NRC review and approval prior to their implementation in AURORA-B. (This is Condition 23 of the SE for the base AURORA-B TR. It remains applicable to CRDA analyses for BWRs/2-6.)

This condition is met as ATRIUM 11 does exhibit the structural similarities described in the restriction.

24. Changes may be made to the AURORA-B EM in the [

] areas discussed in Section 4.0 of Reference 6 (the SE for the base AURORA-B TR) without prior NRC approval. (This is Condition 24 of the SE for the base AURORA-B TR. It remains applicable to CRDA analyses for BWRs/2-6.)

This condition is met through the use of the Framatome software development procedures.

25. The parallelization of individual CCDs may be performed without prior NRC approval as discussed in Section 4.0 of Reference 6 (the SE for the base AURORA-B TR).

(This is Condition 25 of the SE for the base AURORA-B TR. It remains applicable to CRDA analyses for BWRs/2-6.)

No confirmation is required for this condition.

26. AREVA must continue to use existing regulatory processes for any code modifications made in the [

] areas discussed in Section 4.0 of

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 8-4 Reference 6 (the SE for the base AURORA-B TR). (This is Condition 26 of the SE for the base AURORA-B TF. It remains applicable to CRDA analyses for BWRs/2-6.)

This condition is met through the use of the Framatome software development procedures which include 10CFR50.59 licensing considerations.

27. The control rod model at each location in the core used for CRDA analyses with the AURORA-B EM shall use a control rod geometry and composition that is verified to bound the control rod worth for the physical control rod used in the location, for all axial elevations.

[

] Therefore, this condition is met.

28. Licensees utilizing AURORA-B to perform CRDA analyses using the methodology described in this TR shall confirm that the recommended maximum rod velocity of 3.11 ft/s is conservative for their control rods.

The licensee has confirmed that this condition is met for the control rods at Monticello.

29. If the check to verify that the total enthalpy is limiting at 10 percent core flow CZP conditions by [

] fails, AREVA shall perform a more comprehensive evaluation to verify that they have identified the limiting initial conditions for that plant. This evaluation should consider a range of flow values and corresponding plant-specific minimum temperatures that is sufficiently broad to clearly identify the combination of initial conditions which maximizes the total enthalpy for the limiting rod.

[

]

Monticello with ATRIUM 11 fuel for determining the total enthalpy.

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 8-5

30. When individual control rods are evaluated using the CRDA analysis methodology, if necessary, alternate distributions of inoperable rods should be utilized to ensure inclusion of at least one evaluation within each group of 4 quadrant symmetric control rods that maximizes the change in face- and/or diagonally-adjacent uncontrolled cells as a results of the candidate control rod withdrawal.

[

]

31. The evaluation boundary curve used to determine candidate control rods for further evaluation based on their static rod worths must be verified to bound the following local characteristics of the fuel being evaluated: design pin peaking factors, fuel assembly design, location in or adjacent to the outermost ring of control rods, and average burnup for the 16 fuel assemblies surrounding the rod of interest.

This condition is addressed in Appendix A for this demonstration analysis.

32. If the highest worth rod at a given core statepoint results in a total enthalpy that is higher than the minimum high temperature failure threshold (i.e., lowest threshold for all rod internal pressures), additional rods must be considered for evaluation. This may be done by evaluating the next highest worth rods at the core statepoint of interest until the minimum high temperature failure threshold is met, or by using an approach analogous to the evaluation boundary curve for the PCMI failure threshold (as subject to condition 29).

The highest control rod worth did result in a total enthalpy which exceeded the minimum high temperature failure threshold. Therefore, additional control rods were evaluated to address this condition (see Section 4.1).

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page 8-6

33. If the methodology described in ANP-10333 is used to analyze the CRDA event with a fuel assembly design that has a different fuel rod geometry and/or manufacturing tolerances than the one used as a basis for the sensitivity study on gap width, the sensitivity study shall be repeated for the new fuel assembly design, using bounding values consistent with the uncertainty range for [

] limiting increase in the peak total enthalpy, the total uncertainty shall be increased accordingly for total enthalpies calculated based on the new fuel assembly design.

The ATRIUM 11 product line requires an evaluation of the gap sensitivity study. The sensitivity studies were performed with a bounding value for the uncertainty range of [

]. The resulting increase in peak total enthalpy [ ]

34. The uncertainty designate in the CRDA TR of [

] for the enthalpy rises calculated using the CRDA analysis methodology may not be reduced without prior NRC approval.

The uncertainty of [ ] percent is used in this evaluation.

Other conditions:

In Section 3.0 of the SER, the NRC discussed the evolving CRDA criteria at the time of approval of ANP-10333. The final criteria (RG 1.236) must be reviewed to confirm that any changes relative to DG-1327 other than clarifications, editorial changes, or numeric thresholds must be evaluated.

[

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9.0 REFERENCES

1. ANP-3881P, Revision 0, Monticello ATRIUM 11 Equilibrium Cycle Fuel Cycle Design Report, Framatome, November 2020.

2. NUREG-0800, Section 15.4.9, Revision 3, SPECTRUM OF ROD DROP ACCIDENTS (BWR). Standard Review Plan: LWR Edition, US NRC: Washington, DC. March 2007.

3. RG 1.236, Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents, June 2020. (NRC ADAMS ML20055F490)

4. ANP-10333P-A, Revision 0, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA), Framatome, March 2018.

5. NUREG-0800, Section 15.2.9, Revision 2, SPECTRUM OF ROD DROP ADDICENTS (BWR). Standard Review Plan: LWR Edition, US NRC: Washington, DC. July 1981.

6. ANP-10300P-A, Revision 1, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Transient and Accident Scenarios, Framatome, January 2018.

7. ANP-10335P-A, Revision 0, ACE/ATRIUM 11 Critical Power Correlation Topical Report, Framatome, May 2018.

8. BAW-10247PA, Revision 0, Supplement 1P-A Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors Supplement 1:

Qualification of RODEX4 for Recrystallized Zircaloy-2 Cladding, AREVA, April 2017.

9. ANP-3924P, Revision 0, Applicability of Framatome BWR Methods to Monticello with ATRIUM 11 Fuel, Framatome Inc., June 2021.

10. DG-1327, Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents, July 2019 (NRC ADAMS ML18302A106).

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page A-1 Appendix A Evaluation Threshold Determination Limitation and Condition 31 states:

31. The evaluation boundary curve used to determine candidate control rods for further evaluation based on their static rod worths must be verified to bound the following local characteristics of the fuel being evaluated: design pin peaking factors, fuel assembly design, location in or adjacent to the outermost ring of control rods, and average burnup for the 16 fuel assemblies surrounding the rod of interest.

The process to generate an evaluation threshold is demonstrated based upon the process described in the response to RAI-5 in ANP-10333Q1P (included in Reference 4). The peak fuel rod enthalpy rise was elevated using a multiplication factor of [ ] to double the uncertainty.

The elevated enthalpy rise values were then tabulated against the static control rod worth.

Figure A.1 Data for Evaluation Boundary

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page A-2 Figure A.2 Example Evaluation Boundary for the Monticello ATRIUM 11 Core Based on the example evaluation boundary in Figure A.2 created using the data in Figure A.1, any case below and to the right of the boundary should be investigated. When using an evaluation boundary line, consideration of high burnup assemblies with lower PCMI threshold limits is required and they should be conservatively assessed.

The local characteristics of fuel used to establish the evaluation boundary with respect to design fuel rod peaking factors, fuel assembly design, core location, and the average burnup of the 16 assemblies around the dropped rod have been are provided in Table A.1. This table is for use with future core licensing to confirm the applicability of the evaluation boundary curve.

Controlled Document Framatome Inc. ANP-3929NP Revision 0 Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology Page A-3 Table A.1 Example Evaluation Boundary Characteristics

[

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ENCLOSURE ATTACHMENT 10b MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST APPLICATION TO ADOPT ADVANCED FRAMATOME METHODOLOGIES AFFIDAVIT FOR ANP-3929P REPORT, REVISION 0 MONTICELLO ATRIUM 11 CONTROL ROD DROP ACCIDENT ANALYSES WITH THE AURORA-B CRDA METHODOLOGY JUNE 2021 (3 pages follow)

AFFIDAVIT

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

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

3. I am familiar with the Framatome information contained in the report ANP-3929P Revision 0, "Monticello ATRIUM 11 Control Rod Drop Accident Analyses with the AURORA-B CRDA Methodology," dated June 2021 and referred to herein as "Document."

Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.