L-MT-23-034, Subsequent License Renewal Application Response to Request for Additional Information Set 1

Subsequent License Renewal Application Response to Request for Additional Information Set 1

ML23227A175
Person / Time
Site:	Monticello
Issue date:	08/15/2023
From:	Hafen S Northern States Power Company, Minnesota, Xcel Energy Inc
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-MT-23-034
Download: ML23227A175 (1)

Text

Xcel Energy* 2807 West County Road 75 Monticello, MN 55362 August 15, 2023 L-MT-23-034 10 CFR 54.17 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed Facility Operating License No. DPR-22 Subsequent License Renewal Application Response to Request for Additional Information Set 1

References:

1) Letter from Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy to Document Control Desk, "Monticello Nuclear Generating Plant Docket No. 50-263, Renewal License Number DPR-22 Application for Subsequent Renewal Operating License" dated January 9, 2023, ML23009A353

2) Email from the NRC to Northern States Power Company, A Minnesota corporation (NSPM), d/b/a Xcel Energy, "Monticello SLRA - Request for Additional Information - Set l" dated July 19, 2023, ML23200A350 and ML23200A351 Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy hereafter "NSPM", is submitting responses to requests for additional information (RAls) to the Subsequent License Renewal Application, listed in Reference 1.

On July 19, 2023, the NRC issued RAls Set 1 (Reference 2) to NSPM. The RAI responses are provided in the following enclosures to this letter. Any RAI responses that required revisions to the SLRA are marked up in accordance with the next paragraph. Previous changes incorporated as a result of Supplements are provided by bold, black font and noted in the response.

In the enclosures, changes are described along with the affected section(s) and page number(s) of the docketed SLRA (Reference 1) where the changes are to apply. For clarity, revisions to the SLRA are provided with deleted text by strikethrough and inserted text by bold red underline.

Document Control Desk L-MT-23-034 Page 2 Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

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

Executed on August /~ , 2023.

Shaw af Site Vice Pr 1dent, Monticello Nuclear Generating Plant Northern States Power Company - Minnesota cc: Administrator, Region 111, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce

Document Control Desk L-MT-23-034 Page 3 Enclosures Index Enclosure Subject No.

01 RAI 3.5.2.2.1.5-1 02 RAI 4.3.1-1 03 RAI 4.3.2-1 04 RAI 4.3.6-1 05 RAI 2.3.3.14-1 06 RAI 4.3.7-1 07 RAI 4.3.7-2 08 RAI 4.3.7-3 09 RAI 4.3.7-4 10 RAI B.2.2.1-1 1

RAI 3.5.2.2.1.5-1

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 1 Page 1 of 14 RAI 3.5.2.2.1.5-1 Regulatory Basis:

Title 10 of the Code of Federal Regulations (CFR) Section 54.21(a)(3) requires the applicant to demonstrate that the effects of aging for structures and components in the scope of license renewal and subject to aging management review (AMR) will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis (CLB) for the period of extended operation.

As described in the SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21(a)(3) by referencing the GALL-SLR Report when evaluation of the matter in the GALL-SLR Report applies to the plant. SRP-SLR Section 3.1.2.4 states: If the applicant identifies an exception to any of the program elements of the cited GALL-SLR Report AMP, the SLRA AMP should include a basis demonstrating how the criteria of 10 CFR 54.21(a)(3) would still be met. The reviewer should then confirm that the SLRA AMP with all exceptions would satisfy the criteria of 10 CFR 54.21(a)(3).

The parameters monitored or inspected program element of GALL-SLR AMP XI.S1 ASME Section XI, Subsection IWE states, in part: Steel, stainless steel (SS), and dissimilar metal weld pressure-retaining components that are subject to cyclic loading but have no CLB fatigue analysis (i.e., components covered by SRP-SLR Table 3.5-1, items 27 and 40, and corresponding GALL-SLR items; as applicable), are monitored for cracking.

The detection of aging effects program element of GALL-SLR AMP XI.S1 states, in part:

"The requirements of ASME Code Section XI, Subsection IWE and 10 CFR 50.55a are supplemented to perform surface examination (or other applicable technique) in addition to visual examinations, to detect cracking in steel, SS, and dissimilar metal weld pressure-retaining components that are subject to cyclic loading but have no CLB fatigue analysis (i.e., components covered by SRP-SLR Table 3.5-1, items 27 and 40, and corresponding GALL-SLR items; as applicable to the plant)."

SRP-SLR Section 3.5.2.2.1.5 Cumulative Fatigue Damage, as modified by SLR-ISG-2021 STRUCTURES (ADAMS Accession No. ML2018A381), provides guidance for further evaluation of cumulative fatigue damage of containment pressure-retaining boundary components subject to cyclic loading but have no CLB fatigue analysis, and states in part:

"For the above-stated containment pressure-retaining components (corresponding to Table 3.5-1, Items 027 and 040) subject to cyclic loading for which no CLB fatigue analysis exists at the time of an SLRA submittal, a plant-specific further evaluation may be performed to demonstrate that cracking due to cyclic loading is an aging effect that does not require aging management for the component. As one acceptable approach, the aging effect does not require aging management actions if the further evaluation demonstrates that the six criteria for cyclic loading in paragraph NE-3222.4(d) (NE-3221.5(d) in 1980 and later code editions), Analysis for Cyclic Operation, Vessels Not Requiring Analysis for Cyclic Service, of ASME Code,Section III, Division 1 (1974 edition or later edition incorporated by reference in 10 CFR 50.55a(a)(i)),

that provide for a waiver from detailed fatigue analysis are satisfied for applicable component materials through the end of the subsequent period of extended operation. The option to

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 1 Page 2 of 14 perform a fatigue waiver analysis to address the aging effect of cracking due to cyclic loading, for specific containment metallic components, is in lieu of performing supplemental surface examinations."

Background:

SLRA Section B.2.3.29 takes exception to the GALL-SLR AMP XI.S1 recommendations (stated in the regulatory basis) to monitor for cracking through supplemental surface examinations, and in the first bullet under Exceptions to NUREG-2191 states in part:

".An assessment was performed to address the following design inputs for components materials comprising the primary containment that could be subject to cyclic loading but have no CLB fatigue analysis: (1) Atmospheric-to-operating pressure cycle; (2) Normal operation pressure fluctuation; (3)Temperature difference - startup and shutdown; (4) Temperature difference - normal operation; (5) Temperature difference - dissimilar metals; and (6)

Mechanical loads.

The assessment concluded that the drywell shell, non-high temperature drywell penetrations, and penetration sleeves are subjected to a small amount of fatigue such that neither fatigue analysis nor a fatigue waiver is required. As such, cracking due to cyclic loading does not require aging management for drywell shell, non-high temperature drywell penetrations, and penetration sleeves.

MNGP does not monitor for cracking utilizing supplemental surface examinations except at accessible portions of certain steel and stainless-steel penetrations associated with high temperature systems. "

SLRA Section 3.5.2.2.1.5 Cumulative Fatigue Damage references the criteria in SRP-SLR Section 3.5.2.2.1.5, as modified by SLR-ISG-2021-03-STRUCTURES. SLRA Section 3.5.2.2.1.5 states, in part:

"An assessment was performed which concluded that the drywell shell and non-high temperature drywell penetrations, and penetration sleeves are subjected to a small amount of fatigue, therefore, fatigue analysis, or a fatigue waiver, for the drywell shell and drywell penetrations is not required. This assessment did not include drywell penetration bellows, which have fatigue analysis, and adapters of high temperature drywell mechanical penetrations."

Request:

1. For each of the drywell components for which the assessment was performed and credited to justify the exception(s) to GALL-SLR AMP XI.S1 taken in SLRA Section B.2.3.29 AMP, describe in sufficient technical detail the assessment, including cyclic loads and material inputs used, code edition, provisions, and criteria used/met with summary of results, that would demonstrate the acceptance criteria in SRP-SLR Section 3.5.2.2.1.5 (as modified by SLR-ISG-2021-03-STRUCTURES) were met to conclude that the aging effect of cracking due to cyclic loading does not require management.

2. Update the SLRA as necessary.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 1 Page 3 of 14 Response to RAI 3.5.2.2.1.5-1 Italicized text in the Background discussion above was replaced in SLRA Section B.2.3.29 with phrasing that better aligns with discussion of fatigue waiver analysis presented in Appendix A to SLR-ISG-2021-03-STRUCTURES (refer to Enclosure 27 of Reference 1); however, the corresponding summary in SLRA Section 3.5.2.2.1.5 was not updated. Discussion in SLRA Section 3.5.2.2.1.5 and data in Table 3.5-1 are being revised to (a) provide a clarified summary of the fatigue waiver analysis and (b) describe the evaluation in sufficient technical detail per numbered item 1 in the Request above.

References:

1. L-MT-23-025, Monticello Nuclear Generating Plant, Docket No. 50-263, Renewed Facility Operating License No. DPR-22, Subsequent License Renewal Application Supplement 2, ML23177A218.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 1 Page 4 of 14 Associated SLRA Revisions:

SLRA Section 3.5.2.2.1.5 (third and subsequent paragraphs on page 3.5-24) is revised as follows:

As part of license renewal activities, an assessment fatigue waiver analysis was performed which concludeddemonstrating that if the Monticello drywell (primary containment),

including mechanical and electrical penetrations, had been designed to the 1974 Edition of ASME Code,Section III, it would have met the criteria for fatigue exemption specified by Subsection NE-3222.4(d) Vessel Not Requiring Analysis for Cyclic Operation. This analysis concluded that the drywell shell, and non-high temperature drywell penetrations, and penetration sleeves are subjected to a small and acceptable amount of fatigue, therefore, such that detailed fatigue analysis, or a fatigue waiver, for the drywell shell and drywell penetrations is not required. As such, cracking due to cyclic loading is not an aging effect requiring management for the drywell shell, non-high temperature drywell penetrations, and penetration sleeves. The fatigue waiver analysis for the drywell and its penetrations is summarized at the end of this section.

The assessmentanalysis did not include drywell penetration bellows, which have fatigue analysis, and penetration adapters of high temperature drywell mechanical penetrations tabulated in Section 3.5.2.2.1.6. As summarized in items 3.5.1-027, and 3.5.1-040, cracking due to cyclic loading is not an aging effect requiring management for the drywell shell, non-high temperature drywell penetrations and penetration sleeves. Cracking due to cyclic loading for portions of high-temperature piping penetrations that are not pressurized during local leak rate testing and do not have a CLB fatigue analysis will be managed by the ASME Subsection XI, Subsection IWE (B.2.3.29) AMP, including an enhancement to inspect accessible portions for cracking, and 10 CFR Part 50, Appendix J (B.2.3.31) AMP, respectively, during the SPEO.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 1 Page 5 of 14 Fatigue Waiver Analysis for the Monticello Drywell 1.0 Introduction 1.1 For drywell components that do not have a fatigue analysis performed per the original plant design specifications and construction codes, MNGP has performed an evaluation to show that the drywell would have met the criteria for a fatigue waiver. This evaluation consists of a fatigue waiver analysis based on transient cycle limits (as described in NUREG-2191,Section X.M1), and is a technically acceptable approach to analytically address the effects of cyclic loading (fatigue aging effects) for containment metallic pressure-retaining boundary components (as described in SLR-ISG-2021-03-STRUCTURES, Appendix A).

2.0 Summary 2.1 The MNGP drywell shell, drywell penetrations, and penetration sleeves were determined not to have an existing CLB fatigue analysis and therefore have no fatigue TLAAs. In addition, non-piping penetrations (CRD hatch, equipment hatch, personnel airlocks, electrical penetrations, and seismic restraint inspection ports) are considered not to have a CLB fatigue evaluation.

The Monticello primary containment (i.e., the drywell and its penetrations) was designed to the 1965 Edition of ASME Section III, Subsection B, which does not require a fatigue analysis or exemption/waiver. As part of SLRA development, MNGP has documented a fatigue waiver analysis for the SPEO to demonstrate that if the Monticello drywell, including mechanical and electrical penetrations, had been designed to the 1974 Edition of ASME Section III, then it would have met the six criteria in NE-3222.4(d) Vessel Not Requiring Analysis for Cyclic Operation.

Methodology, temperature-dependent material properties, and alternating stress intensity values used in this analysis are based on the 1974 Code edition.

For establishing Class MC boundaries, guidance is taken from the 1974 Edition of the ASME Code with Addenda through Winter 1975. The limits of Class MC for some example penetrations are depicted in Section III, Figure NE-1120-1, which indicates that penetration nozzles and end caps can be Class MC but process pipe and flued heads are not Class MC.

The analysis justifies that the drywell and the portions of its penetrations within the scope of the evaluation (summarized below) meet the fatigue waiver criteria specified in NE-3222.4(d) of Section III of the ASME Code and can be considered exempt from detailed fatigue analysis. The six criteria address the following design inputs:

(1) Atmospheric-to-operating pressure cycles, (2) Normal operation pressure fluctuations, (3) Temperature difference - startup and shutdown, (4) Temperature difference - normal operation, (5) Temperature difference - dissimilar materials, and (6) Mechanical loads.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 1 Page 6 of 14 2.2 This analysis does not include penetration X-3, construction manway, because it has been seal welded and encased in concrete on its inside and outside surfaces.

2.3 The following drywell components are not included in this analysis because they are already designed to Class MC:

attachments to the drywell shell and ring girder due to support modifications drywell floor modifications penetrations X-53A & B and X-54A & B 2.4 This analysis excludes any portions of penetrations that are affected by process fluid to the extent that temperature exceeds the drywell design temperature of 281°F.

2.5 For penetrations with hot process pipe, the penetration nozzle is welded to a flued head as shown in USAR Figure 5.2-6. Based on this configuration within the penetration assembly, only the penetration nozzle (including bellows) is Class MC.

The penetration nozzle is remote from the process piping and is expected to be at or below 281°F. As a result, the Class MC portion of a penetration with hot process pipe is covered by this analysis.

2.6 Several penetration assemblies include caps made of material that is not included in the design stress intensity (Sm) table for ferritic steels. For this evaluation, this material is considered adjacent material and not Class MC material. The following penetration assemblies are in this category, and are included in this analysis:

• X-15 clean up return SA-283 grade C or SA-420 grade WFL1

• X-100E neutron monitoring signal SA-283 grade C or SA-420 grade WFL1

• X-101C recirc pump power SA-283 grade C or SA-420 grade WFL1

• X-102 indication & control SA-283 grade C or SA-420 grade WFL1

• X-104E CRD rod position indicator SA-283 grade C or SA-420 grade WFL1 2.7 The drywell is made of SA-516 grade 70 carbon steel (CS) plate. The Class MC portions of CS penetrations are made of this material and SA-333 grade 1. In addition, some penetrations contain stainless steel (SS) portions made of SA-240 grade 304 and SA-312 grade TP304. The temperature-dependent material properties tabulated in ASME Section III, Appendix I, are identical for both SS materials. The analysis is simplified to perform an evaluation of the six criteria on a material basis instead of on an individual penetration basis.

2.8 Conservative, bounding number of cycles for 80 years were used with specific values as follows, with further explanations provided in subsequent sections:

Number of full pressure cycles: 452 (see Section 3.1)

Number of significant pressure fluctuations: 17 (see Section 3.2)

Number of startup/shutdown cycles: 289 (per SLRA Table 4.3.1-1)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 1 Page 7 of 14 Number of significant temperature fluctuations: 600 (see Section 3.4)

Number of significant load fluctuations: 9340 (see Section 3.6) 3.0 Evaluation The six subsections below present (paraphrased) each of the six criteria for a fatigue waiver specified in the ASME Code for Class MC components and a summary of the associated analysis performed for the drywell and the portions of its penetrations within the scope of the evaluation.

The following nomenclature is used:

inst = instantaneous coefficient of thermal expansion 1 = inst for material 1 of a material discontinuity 2 = inst for material 2 of a material discontinuity E = elastic modulus at Tave E1 = E for material 1 of a material discontinuity E2 = E for material 2 of a material discontinuity P = pressure Sa = alternating stress intensity on applicable fatigue curve at specified number of cycles Sm = design stress intensity S = alternating stress intensity at 106 cycles T = temperature Tave = average temperature for either adjacent points or two materials, as applicable 3.1 From ASME Section III, NE-3222.4(d)(1): Atmospheric-to-operating pressure cycles:

The number of pressure cycles from atmospheric to operating pressure and back (including startup/shutdown) does not exceed the number of cycles on the fatigue curve at an alternating stress intensity (Sa) of 3Sm at operating temperature.

Evaluation: NE-3222.4(d)(1):

The number of pressure cycles over the full range from zero to normal operating pressure is 452. This is the sum of the cycle limits (listed in USAR Table 4.2-1) for startup/shutdown, design hydro test at 1250 psig, hydrostatic test to 1560 psig, and loss of feedwater pumps. These are events during which reactor pressure varies over the full range (or nearly so for loss of feedwater pumps). This value is appropriate since it is reasonable to expect the drywell to be pressurized in most cases when the reactor is pressurized. Sm is the allowable stress intensity value at a given temperature from ASME Section III, Appendix I, Tables I-1.1 and I-1.2. Sa corresponds to the allowable amplitude of the alternating stress component and is plotted against the number of cycles on the fatigue curves found in ASME Section III, Appendix I, Figures I-9.1 and I-9.2.

At 452 cycles and Sm values taken at 281°F, the resulting fatigue curve cycles at 3Sm are as follows:

SA-516 Gr. 70 (Sm= 21.4 ksi) = 1983 cycles SA-333 Gr. 1 (Sm = 17.8 ksi) = 3560 cycles SA-240 Gr. 304 (Sm= 20.0 ksi) = 9341 cycles

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 1 Page 8 of 14 Therefore, resulting cycles based on the fatigue curve for each material is greater than the expected 452 operating pressure cycles for 80 years and the criterion is met.

3.2 From ASME Section III, NE-3222.4(d)(2): Normal operation pressure fluctuations:

The pressure range during normal operation does not exceed (1/3)(design pressure)(Sa/Sm), where Sa is taken from the fatigue curve at a number of cycles equal to the number of significant pressure fluctuations. Significant pressure fluctuations are those with a pressure change greater than (1/3)(design pressure)(S/Sm), where S is taken from the fatigue curve at 106 cycles.

Evaluation: NE-3222.4(d)(2):

Operating pressure range = -0.1 to 1 psig Design pressure range = -2 to 62 psig Per NE-3222.4(d)(2) significant pressure fluctuations are those for which the excursion exceeds the quantity:

Design Pressure

• 1/3

• S/Sm, where S = the value of Sa obtained from the applicable design fatigue curve for 106 cycles.

At 106 cycles, Sa = 12.5 ksi for CS.

For SA-516 Gr. 70, Sm = 21.4 ksi.

Therefore, for SA-516 Gr. 70, significant pressure fluctuations are those for which the excursion exceeds the quantity:

0.064*12.5/(3*21.4) = 0.012 ksi For SA-333 Gr. 1, Sm= 17.8 ksi.

Therefore, for SA-333 Gr. 1, significant pressure fluctuations are those for which the excursion exceeds the quantity:

0.064*12.5/(3*17.8) = 0.015 ksi At 106 cycles, Sa= 26 ksi for SS.

For SA-240 Gr. 304, Sm= 20.0 ksi.

Therefore, for SA-240 Gr. 304, significant pressure fluctuations are those for which the excursion exceeds the quantity:

0.064*26/(3*20.0) = 0.028 ksi These pressure values define what is a significant pressure fluctuation.

The number of significant pressure fluctuations during normal operation is assumed to be 17, corresponding to one prior to plant operation and one every 5 years for 80 years. This is conservative given that the design basis accident (DBA) is a postulated event that is not expected to occur, and leak rate testing of the entire containment is only required every 10 years.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 1 Page 9 of 14 At 17 pressure fluctuations; from ASME Section III, Appendix I, Figure I-9.1, the corresponding Sa for CS is 445 ksi (which includes SA-516 and SA-333).

At 17 pressure fluctuations; from ASME Section III, Appendix I, Figure I-9.2, the corresponding Sa for SS is 507 ksi (which includes SA-240).

The resulting limit on the normal operation pressure range:

For SA-516 Gr. 70: 0.064*445/(3*21.4) = 0.443 ksi For SA-333 Gr. 1: 0.064*445/(3*17.8) = 0.533 ksi For SA-240 Gr. 304: 0.064*507/(3*20.0) = 0.541 ksi These values are significantly greater than the normal operation pressure fluctuation range of 1.1 psig; as a result, the normal operation pressure fluctuation meets the requirements of criterion (2).

3.3 From ASME Section III, NE-3222.4(d)(3): Temperature difference - startup and shutdown:

The temperature difference (T) between any two adjacent points during normal operation, startup, and shutdown does not exceed Sa/(2Einst), where Sa is taken from the fatigue curve for the specified number of startup/shutdown cycles, and E and inst are taken at the average temperature Tave for the two points.

Evaluation: NE-3222.4(d)(3):

The expected number of startup-shutdown cycles is 289 in which the drywell and penetrations experience the maximum temperature difference between any two adjacent points occurring during Startup and Shutdowns. This is the bounding value used in SLRA Table 4.3.1-1.

Maximum Operating Temperature for the vessel steel is 281°F (from the original specification for the drywell)

The minimum temperature while the plant is shut down is assumed to be 40°F, which is conservative compared with a more typical value of 70°F in that it yields a higher temperature range during startup/shutdown.

Thus, the maximum operating temperature difference is 281°F - 40°F = 241°F.

Normal operating average T: (281 + 70)/2 = 175.5°F (70°F is used in the averaging instead of 40°F to yield conservatively high Einst). Parameter values at Tave from ASME Section III, Appendix I, Tables I-5.0 and I-6.0:

inst (CS)= 6.56E-06 in/in/°F inst (SS)= 9.45E-06 in/in/°F E (CS)= 27738 ksi E (SS)= 27813 ksi At 289 cycles; from ASME Section III, Appendix I, Figure I-9.1, Sa = 132.6 ksi (CS).

Therefore, for the CS components, the limit on the temperature difference is 132.6/(2*27738*6.56E-06) = 364°F.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 1 Page 10 of 14 At 289 cycles: from ASME Section III, Appendix I, Figure I-9.2, Sa = 163.5 ksi (SS).

Therefore, for the SS components, the limit on the temperature difference is 163.5/(2*27813*9.45E-06) = 311°F These temperature limits are both significantly greater than the maximum expected temperature difference of 241°F considered for startup and shutdowns.

These results indicate that the containment could experience shutdown and startup temperature differences (between any two adjacent points) of 364°F (CS) and 311°F (SS}, 289 times over 80 years without exceeding the NE-3222.4(d)(3) threshold. In reality, experiencing the NE-3222.4(d)(3) startup and shutdown temperature difference thresholds stated above for 289 cycles is not credible given that the maximum operating temperature difference of 241°F.

3.4 From ASME Section III, NE-3222.4(d)(4): Temperature difference - normal operation:

The range of T between any two adjacent points during normal operation does not exceed Sa/(2Einst), where Sa is taken from the fatigue curve for the specified number of significant temperature fluctuations. Significant temperature fluctuations are those with a range of temperature difference greater than S/(2Einst), where S is taken from the fatigue curve at 106 cycles.

Evaluation: NE-3222.4(d)(4):

The number of significant temperature fluctuations is assumed to be 600; this is chosen as a bounding high value for which the applicable criteria are met without additional detailed analysis. This is reasonable given that this evaluation excludes locations for which T exceeds 281°F as per Section 2.4, and therefore includes only locations that are essentially unaffected by process fluid transients.

The expected maximum temperature-difference fluctuation equals the difference between maximum operating temperature (281°F) and the minimum operating temperature (assumed to be 40°F), which equals 241°F, and is conservative.

The range of T between adjacent points is assumed to be 1/2 the maximum theoretical value of (281 - 40) - (40 - 281) = 482°F, or 482/2 = 241°F. This is conservative given that (a) no thermal transients are defined for the drywell, and (b) the Class MC portion of penetrations with hot process piping is remote from the process piping as discussed in Section 2.5.

As stated in Section 3.3, Tave = 175.5°F. Parameter values at Tave from ASME Section III, Appendix I, Tables I-5.0 and I-6.0:

inst (CS)= 6.56 E-06 in/in/°F inst (SS)= 9.45 E-06 in/in/°F E (CS)= 27738 ksi E (SS)= 27813 ksi At 600 cycles: from ASME Section III, Appendix I, Figure I-9.1, Sa = 98.7 ksi (CS).

Therefore, the limit on the temperature-difference for CS is:

98.7/(2*27738*6.56E-06) = 271°F.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 1 Page 11 of 14 At 600 cycles, from ASME Section III, Appendix I, Figure I-9.2, Sa = 128.3 ksi (SS).

Therefore, the limit on the temperature-difference for SS is:

128.3/(2*27813*9.45E-06) = 244°F.

These temperature-difference limits between adjacent points are greater than the expected temperature-difference fluctuation of 241°F between adjacent points.

These results indicate that the drywell and penetrations could experience normal operation temperature differences (between any two adjacent points) of 271°F (CS) and 244°F (SS), 600 times over 80 years without exceeding the NE-3222.4(d)(4) threshold.

3.5 From ASME Section III, NE-3222.4(d)(5): Temperature difference - dissimilar materials:

At dissimilar material discontinuities, the range of temperature change does not exceed Sa/[2(E11 - E22)], where Sa is taken from the fatigue curve for the specified number of significant temperature fluctuations, E1 and E2 are the moduli of elasticity, and 1 and 2 are inst for each material at the average temperature.

Significant temperature fluctuations are those with a range of temperature change greater than S/[2(E11 - E22)], where S is taken from the fatigue curve at 106 cycles.

Evaluation: NE-3222.4(d)(5):

As stated in Section 2.7, materials used in Class MC portions of CS penetrations are SA-516 grade 70 and SA-333 grade 1; both are steels with carbon content of 0.3% or less. Adjacent parts include those made of SS(plug, cap, radiation shield, flued head, leak chase channel, welding neck flange, or blind flange) or steels with carbon content of greater than 0.3% (flued head or cap).

As stated in Section 3.3, Tave = 175.5°F. Parameter values at Tave from ASME Section III, Appendix I, Tables I-5.0 and I-6.0:

inst (CS)= 6.56 E-06 in/in/°F inst (SS)= 9.45 E-06 in/in/°F E (CS, C 0.3% or less)= 27738 ksi E (CS, C >0.3%) = 29575 ksi E (SS)= 27813 ksi At 106 cycles, Sa = 12.5 ksi for CS.

For the Class MC-to-SS interface, a significant T fluctuation is:

12.5/[2*(27813*9.45E 27738*6.56E-06)] = 77°F.

The expected number of transients in which these dissimilar metal discontinuities experience the maximum temperature fluctuations at this interface is 600, consistent with the basis provided in the previous section.

The expected maximum temperature-difference fluctuation equals the difference between maximum operating temperature (281°F) and the minimum operating temperature (assumed to be 40°F), which equals 241°F, and is conservative.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 1 Page 12 of 14 At 600 cycles: from ASME Section III, Appendix I, Figure I-9.1, Sa = 98.7 ksi (CS).

Therefore, the limit on the maximum temperature-difference for the Class MC-to-SS interface is:

98.7/[2*(27813*9.45E 27738*6.56E-06)] = 610°F The temperature-difference limit for the Class MC-to-SS interface between adjacent points is greater than the expected temperature fluctuation of 241°F. This result indicates that portions of the drywell and penetrations with these dissimilar metals could experience maximum temperature fluctuations stated above 600 times over 80 years without exceeding the NE-3222.4(d)(5) threshold. Realistically, this NE-3222.4(d)(5) threshold is not credible given that the maximum operating temperature differences between adjacent dissimilar materials is only 241°F.

For the Class MC-to-CS, C >0.3% interface, a significant T fluctuation is:

12.5/[2*(29575*6.56E 27738*6.56E-06)] = 519°F.

As stated above, the maximum T range between adjacent points is only 241°F; thus, these dissimilar metal discontinuities will not experience a significant T fluctuation and further evaluation is not required.

3.6 From ASME Section III, NE-3222.4(d)(6): Mechanical loads:

The full range of stress due to mechanical loads (excluding pressure but including pipe reactions) does not exceed the Sa value for the total number of significant load fluctuations. Significant load fluctuations are those with a range of stress greater than the Sa value from the fatigue curve at 106 cycles.

Evaluation: NE-3222.4(d)(6):

This criterion addresses mechanical reactions on the containment and not thermal or pressure transients. The number of significant load fluctuations is based on an analysis of torus penetrations X-221 (HPCI Turbine Exhaust) and X-212 (RCIC Turbine Exhaust) and is conservatively estimated as the sum of cycles for the load combinations tabulated below:

Load combination design cycles SRVm + OBE 1000 SRVm 290 SRVs 3380 TOTAL 4670 x2 for 80 years 9340 At 106 cycles, Sa= 12.5 ksi for CS and Sa= 26 ksi for SS. The mechanical load stress range limits at 9340 cycles are as follows:

From ASME Section III, Appendix I, Figure I-9.1, Sa = 38.9 ksi (CS)

From ASME Section III, Appendix I, Figure I-9.2, Sa = 60.0 ksi (SS)

The range of mechanical load stress is assumed to be 1.5Sm for the material being analyzed, which is conservative given that it is the maximum permitted primary

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 1 Page 13 of 14 membrane plus bending stress intensity and one half the maximum permitted primary plus secondary stress intensity range.

The corresponding mechanical load stresses are SA-516 Gr. 70 CS: 1.5*23.3 ksi= 34.95 ksi SA-333 Gr. 1 CS: 1.5*18.3 ksi= 27.45 ksi SA-240 Gr. 304 SS: 1.5*20.0 ksi= 30 ksi Since the mechanical load stress is less than the corresponding Sa for the mechanical load stress range limits, the mechanical load criterion is satisfied.

4.0 Conclusion 4.1 As summarized above, all six criteria for a fatigue waiver specified in the ASME Code for Class MC components are satisfied. In accordance with the rules of NE-3222.4(d) of Section III of the ASME Code, the drywell and the portions of its penetrations within the scope of this evaluation (see Section 2.0) can be considered exempt from detailed fatigue analysis.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 Enclosure 01 Page 14 of 14 SLRA Table 3.5-1 on page 3.5-55 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Item Aging Management Further Evaluation Component Requiring Discussion Number Program Recommended Management 3.5.1-040 Unbraced Cracking due to cyclic XI.S1, Yes (SRP-SLR Not applicable.

downcomers, steel loading (CLB fatigue "ASME Section XI, Section 3.5.2.2.1.5) elements: vent analysis does not Subsection IWE" This item number is not applicable to the header; downcomers exist) MNGP Mark I steel containment. This item number is applicable only to BWR Mark III containments.

Further evaluation is documented in Section 3.5.2.2.1.5.

2 RAI 4.3.1-1

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 2 Page 1 of 3 RAI 4.3.1-1 Regulatory Basis:

Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background:

SLRA Section 4.3.1 addresses the 80-year transient cycle projections. The SLRA section explains that the transient cycle projections are based on the cycle accumulation rates of the most recent 10-year evaluation period up to May 31, 2021 (i.e., the evaluation period of June 1, 2011 through May 31. 2021).

Issue:

However, the applicant did not clearly address why the cycle projections do not consider the full cycle accumulation rates observed since the start of the plant operation. The staff needs information on the technical basis of the applicants approach that uses only the most recent 10-year cycle accumulation rates (e.g., the most recent 10 years of operation involve distinctive and stable cycle accumulation rates that are a better representation of the operational characteristics of the subsequent period of extended operation than the prior cycle accumulation rates would be).

The staff also noted that the safety/relief valve lifts transient has a total cycle number of 619 as of May 31, 2021, as described in SLRA Table 4.3.1-1. In comparison, the 80-year projected cycle number of this transient is only 699.

Considering that the operation of the plant started on September 30, 1970, the operation time period through the end of the cycle evaluation period (May 31, 2021) is approximately 51 years.

The additional operating time period following May 31, 2021 through 80 years of operation is approximately 29 years (i.e., 80 - 51 years).

Based on the cycle numbers and operating time periods discussed above, the cycle accumulation rate of the safety/relief valve lifts transient for cycle projections is approximately 2.8 cycles/year (i.e., (699 - 619 cycles)/29 years) for the time period after May 31, 2021. In comparison, the previous full cycle accumulation rate since the start of the operation through May 31, 2021 is approximately 12.1 cycles/year (i.e., 619 cycles/51 years),

which is significantly greater than the cycle accumulation rate used in the cycle projections (2.8 cycles/year).

Therefore, the staff needs clarification on the following items for the safety/relief valve lifts transient: (1) why the cycle accumulation rate used in cycle projections (2.8 cycles/year) is significantly lower than the full cycle accumulation rate (12.1 cycles/year) observed since the start of the operation through May 31, 2021; and (2) whether the most recent 10-year operation

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 2 Page 2 of 3 period up to May 31, 2021 represents the operating characteristics for the subsequent period of extended operation in terms of cycle calculations.

Requests:

1. Describe the technical basis for the applicants approach that uses the most recent 10-year cycle accumulation rates for cycle projections but does not consider the full cycle accumulation rates observed since the start of the operation.

2. Provide clarification on the following items for the safety/relief valve lifts transient: (1) why the cycle accumulation rate used in cycle projections is significantly lower than the full cycle accumulation rate observed since the start of the operation through May 31, 2021; and (2) whether the most recent 10-year operation time period up to May 31, 2021 represents the operating characteristics for the subsequent period of extended operation in terms of cycle calculations. For item (2) discussed above, if the most recent 10-year operation period does not represent the operating characteristics for the subsequent period of operation, explain why the cycle accumulation rate of the most recent 10-year operation time period is used in the cycle projections rather than the full cycle accumulation rate observed since the start of the plant operation. Update the SLRA if needed.

Response to RAI 4.3.1-1:

1. The most recent 10 years of operation have cycle accumulation rates that provide the most accurate projections of future cycle accumulation rates. The testing of Main Steam Safety/Relief Valves (S/RV) license amendment request (LAR) and safety evaluation from 2012 (ML12034A020 and ML12185A216, respectively) significantly changed S/RV, Emergency Core Cooling System, and Low-Low set valves surveillance testing requirements which affect the number of expected accumulated transients and therefore accumulation rates prior to this date are not representative of how the plant is currently operated. The Boiling Water Reactor Owners' Group (BWROG) Evaluation of NUREG-0737, "Clarification of TMI Action Plan Requirements," Item ll.K.3.16, "Reduction of Challenges and Failures of Relief Valves," recommended that the number of S/RV openings be reduced as much as possible. NUREG-1482, Rev. 1 "Guidelines for lnservice Testing at Nuclear Power Plants," paragraph 4.3.2.1 states in recent years, the NRC staff has received numerous requests for relief and/or Technical Specification (TS) changes related to the stroke testing requirements for BWR dual-function main steam S/RVs. Both Appendix I to the ASME OM Code and the plant-specific TSs require stroke testing of S/RVs after they are reinstalled following maintenance activities. Several licensees have determined that in situ testing of the S/RVs with reactor pressure can contribute to undesirable seat leakage of the valves during subsequent plant operation and have received approval to perform testing at a laboratory facility coupled with in situ tests and other verifications of actuation systems as an alternative to the testing required by the OM Code and TSs. NUREG-0123, "Standard Technical Specifications for General Electric Boiling Water Reactors," and NUREG-0626, "Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in GE-Designed Operating Plants and Near-Term Operating License Applications," also recommend reducing the number of challenges to the S/RVs. This LAR modified the TS surveillance requirements by providing an alternative methodology where a series of overlapping tests are utilized

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 2 Page 3 of 3 to demonstrate required functioning of the S/RVs, in lieu of manually actuating the S/RVs with steam during plant startup. Due to this change in the way the plant is operated, the most recent 10 years of operation was used for all projected transients to determine the most realistic expected accumulation rates through the subsequent period of extended operation (SPEO).

2. (1) The accumulation of S/RV lifts is significantly lower (as noted, about 3/year rather than 12/year) due to the NRC-approved LAR (ML12185A216) described in the response to question 1 of this RAI. Accumulation of S/RV cycles earlier in plant life was mostly due to online testing, which is no longer required by TS. Capability of S/RV actuation are now demonstrated by manually stroking the valve actuator during each refueling outage (or maintenance outage if an S/RV is replaced) without lifting the main valve disc off the seat and by crediting a series of overlapping tests.

(2) The rate of cycle accumulation based on the most recent 10-year rates is both representative of how MNGP is currently operated and the expected rate of accumulation through the end of the SPEO. The frequency of testing S/RVs is based on the approved Inservice Testing Program.

Associated SLRA Revisions:

There are no SLRA Revisions associated with this response.

3 RAI 4.3.2-1

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 3 Page 1 of 2 RAI 4.3.2-1:

Regulatory Basis:

Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background:

SLRA Section 4.3.2 describes the time-limited aging analysis (TLAA) on ASME Code Section III, Class 1 fatigue waiver evaluations. SLRA Table 4.3.2-1 describes the numbers of transient cycles used in the existing fatigue waiver evaluations and the 80-year projected cycles used in the fatigue waiver TLAA. Specifically, SLRA Section 4.3.2 explains that the applicant used the 80-year projected transient cycles to confirm that the existing fatigue evaluations remain valid for the subsequent period of extended operation for the following components: (1) main closure flange, (2) IRM/SRM dry tube, (3) power range detector assembly and (4) in-core detector assembly.

Issue:

However, SLRA Section 4.3.2 does not clearly discuss why the existing fatigue waiver evaluations remain valid for the subsequent period of the extended operation for the head cooling spray and instrumentation nozzles (N6A and N6B nozzles) and vent nozzle (N7 nozzle).

In addition, SLRA Table 4.3.1-1 indicates that the following transients are not listed in the updated safety analysis report (USAR) and, accordingly, USAR does not define a design cycle limit for these transients: (1) sudden start transient; (2) hot standby with drain shutoff transient; (3) core spray injection transient; and (4) operating basis earthquake (OBE) transient.

SLRA Table 4.3.1-1 also indicates that these transients have not occurred during the plant operation (as of May 31, 2021) and each of these transients is estimated to have one projected cycle for 80 years of operation. The staff needs clarification on whether these non-USAR-listed transients have an impact on the validity of the fatigue wavier evaluations discussed in SLRA Section 4.3.2.

Request:

1. Clarify the following: (1) whether the existing fatigue wavier evaluations for the head cooling spray and instrumentation nozzles and vent nozzle are based on the original design transient cycles described in SLRA Table 4.3.1-1; and (2) how the Fatigue Monitoring AMP can manage the effects of cumulative fatigue damage in relation to the fatigue waiver evaluations for these nozzles.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 3 Page 2 of 2

2. Clarify whether the following non-USAR-listed transients have an impact on the existing fatigue wavier evaluations discussed in SLRA Section 4.3.2: (1) sudden start transient; (2) hot standby with drain shutoff transient; (3) core spray injection transient; and (4) operating basis earthquake (OBE) transient. If so, discuss the impact on the validity of the fatigue waiver evaluations. If not, provide the technical basis for why these transients do not have an impact on the validity of the fatigue waiver evaluations in SLRA Section 4.3.2.

Response to RAI 4.3.2-1:

1. The existing fatigue waiver evaluations for the head cooling spray and instrumentation nozzles and vent nozzle are described in the original GE RPV stress report.

The stress reports determined no specific transient has been specified for these nozzles and they are, therefore, under the action of normal operating transients and rapid cool down. These transients have been analyzed for in other areas of the vessel and on nozzles with larger openings. By comparison with other nozzles having more severe shocks due to direct flows, it can be concluded that provisions for primary plus secondary stresses have been met.

(1) Therefore, the components that bounded the head cooling spray and instrumentation nozzles and vent nozzle in the original analysis did use the original design transient cycles described in SLRA Table 4.3.2-1.

(2) The design limits for the transients listed in Table 4.3.2-1 that are tracked by the Fatigue Monitoring (B.2.2.1) AMP are provided in Table 4.3.1-1. This information was added to the SLRA by Enclosure 09 to Reference 1.

2. (1) sudden start transient; (2) hot standby with drain shutoff transient; and (3) core spray injection transient are emergency or faulted conditions that do not have an effect on the fatigue waiver evaluations because they do not occur as part of the normal operation of the plant. Additionally, while (4) operating basis earthquake (OBE) is an upset condition, it is not expected to occur even though other analyses conservatively use 1 in CUF calculations. These locations are also in the top head of the RPV and the subject thermal transients do not affect locations in the top head. This information was included in the application by Enclosure 10b to Reference 1.

References:

1. L-MT-23-031, Monticello Nuclear Generating Plant, Docket No. 50-263, Renewed Facility Operating License No. DPR-22, Subsequent License Renewal Application Supplement 4 and Response to Request for Confirmation of Information - Set 1, ML23199A154.

Associated SLRA Revisions:

There are no SLRA Revisions associated with this response.

4 RAI 4.3.6-1

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 4 Page 1 of 2 RAI 4.3.6-1:

Regulatory Basis:

Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background:

SLRA Section 4.3.6 addresses the fatigue TLAA for ASME Code Section III Class 2 and 3 and ANSI B31.1 piping systems. Specifically, SLRA Table 4.3.6-1 describes the 40-year full range transient cycles for non-Class 1 piping systems and extrapolates the 40-yer cycles to estimate the 80-year projected cycles. In turn, the 80-year cycle numbers are compared to the 7000 cycle limit in the implicit fatigue analysis.

Issue:

However, LRA Table 4.3.6-1 does not clearly describe how the applicant determined the 40-year cycles that were used in the 80-year cycle projections for non-Class 1 fatigue analysis (e.g., based on piping system design specification and information, plant operation procedures, test requirements, USAR information and specific system-level knowledge).

SLRA Table 4.3.6-1 also includes the following 40-year design cycles: (1) 1500 cycles for the feedwater piping; (2) 532 cycles for the nuclear boiler system; and (3) 205 cycles for the reactor recirculation system. The staff needs clarification on whether these cycles were estimated by summing up two or more design cycles for each of the piping systems.

Request:

1. Clarify how the applicant determined the 40-year cycles that were used in the 80-year cycle projections for non-Class 1 fatigue analysis (e.g., based on piping system design specification and information, plant operation procedures, test requirements, USAR information and specific system-level knowledge)

2. Clarify whether the following 40-year design cycles were estimated by summing up two or more design cycles for each of the non-Class 1 piping systems: (1) 1500 cycles for the feedwater piping; (2) 532 cycles for the nuclear boiler system; and (3) 205 cycles for the reactor recirculation system. If so, describe those design cycles for each piping system.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 4 Page 2 of 2 Response to RAI 4.3.6-1:

1. The 40-year cycles that were used in the 80-year cycle projections for non-Class 1 fatigue analyses were determined by using the limiting number of design cycles which was the reactor feedwater nozzle fatigue analyses. This value was doubled to account for 80 years of operation. This provides a conservative approximation of the actual number of thermal cycles experienced by the piping system, as documented in Section 4.3.4.2 of NUREG-1865 (ML063050414).

2. The 40-year design cycles are from the original Reactor Pressure Vessel specification, included in Appendix H, Reactor Pressure Vessel Design Summary Report of the Updated Safety Analysis Report.

a. Feedwater piping is based on the feedwater nozzle cycles and is assumed to experience 1500 cycles from a temperature of 100oF to 546oF to 100oF and then from 260oF to 376oF (100oF to 546oF to 100oF being one full temperature cycle).

b. Nuclear boiler system is based on steam outlet nozzle cycles and is assumed to experience 532 cycles from a temperature of 100oF to 546oF, 531 cycles from 546oF to 100oF and 1 cycle from 546oF to 100oF.

c. Recirculation system is based on the recirculation inlet nozzle and is assumed to experience 200 cycles from a temperature of 100oF to 546oF and from 546oF to 90oF. The Reactor Recirculation Outlet nozzle included 5 additional cycles from 546oF to 130oF.

Associated SLRA Revisions:

There are no SLRA Revisions associated with this response.

5 RAI 2.3.3.14-1

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 5 Page 1 of 3 RAI 2.3.3.14-1 Regulatory Requirements:

10 CFR 54.4 Scope reads in part:

(a) Plant systems, structures, and components within the scope of this part are--

(2) All nonsafety-related systems, structures, and components whose failure could prevent satisfactory accomplishment of any of the functions identified in paragraphs (a)(1)(i), (ii), or (iii) of this section Guidance Documents:

NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, NEI 95-10, Industry Guideline For Implementing The Requirements of 10 CFR Part 54 -

The License Renewal Rule; Revision 6, Appendix F Issue:

SLRA Section 2.1.4.2.2, Non-Safety Related SSCs Directly Connected to Safety-Related SSCs that Provide Structural Support for the Safety-Related SSCs states in part:

"For NSR SSCs directly connected to SR SSCs, the in-scope boundary for SLR extends into the NSR portion of the piping and supports up to and including the first equivalent anchor beyond the safety/non-safety interface."

SLRA Section 2.3.3.14, Reactor Building Closed Cooling Water describes the results of scoping and screening of the Reactor Building Closed Cooling Water.

NUREG-2192, Section 2.1.3.1.2 Nonsafety-Related states in part :

"In order to comply, in part, with the requirements of 10 CFR 54.4(a)(2), all applicants must include in scope all nonsafety-related piping attached directly to safety-related piping (within the scope of SLR) up to a defined anchor point consistent with the plant CLB. This anchor point may be served by a true anchor (a device or structure which ensures forces and moments are restrained in three (3) orthogonal directions) or an equivalent anchor, such as a large piece of plant equipment (e.g., a heat exchanger,) determined by an evaluation of the plant-specific piping design (i.e., design documentation, such as piping stress analysis for the facility). Applicants should be able to define an equivalent anchor consistent with their CLB (e.g., described in the UFSAR or other CLB documentation), which is being credited for the 10 CFR 54.4(a)(2) evaluation, and be able to describe the SCs that are part of the nonsafety-related piping segment boundary up to and including the anchor point within the scope of SLR."

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 5 Page 2 of 3 NEI 95-10, Revision 6, Appendix F [page F-4] Section 2, Discussion states in part:

"When demonstrating that failures of non safety-related SSCs would not adversely impact on the ability to maintain intended functions, a distinction must be made between non safety-related SSCs that are connected to safety-related SSCs and those that are not connected to safety-related SSCs. For a non safety-related SSC that is connected to a safety-related SSC, the non safety-related SSC should be included within the scope of license renewal up to the first seismic anchor past the safety/non-safety interface."

Scoping/Screening Boundary Drawing SLR-36042-2 Reactor Building Cooling Water System

[M-111-1] displays piping inside Containment from Penetrations X-24(Coordinate C-8) & X-23 (Coordinate C-3) that is color coded as a(2) Spatial/Structural . The structural supports inside containment associated with this piping versus the leakage boundary spatial concerns are two unique concerns and typically not mutually exclusive.

Request:

It is not clear from the staffs review of this boundary drawing whether the guidance of NUREG-2192 Section 2.1.3.1.2 and NEI 95-10 Appendix F has been met. Staff requests the applicant to confirm the NSR piping connected to Penetrations X-24 & X-23(and/or other similar as applicable) is seismically supported consistent with the guidance of NUREG-2192 and NEI 95-10, Revision 6 as determined by an evaluation of the plant-specific piping.

Response to RAI 2.3.3.14-1:

The guidance from NUREG-2192 Revision 0, Section 2.1.3.1.2 and NEI 95-10 Appendix F has been met as shown in SLRA Section 2.1.4.2.2 (page 2.1-16), which states:

Section 4 of Appendix F of NEI 95-10 states that for NSR SSCs that are directly connected to SR SSCs (typically piping systems), the NSR piping and supports, up to and including the first equivalent anchor beyond the safety/non-safety interface, are within the scope of SLR per 10 CFR 54.4(a)(2).

NEI 95-10 Appendix F Section 4.4 states that an alternative to specifically identifying a seismic anchor or series of equivalent anchors that support the SR/NSR piping interface is to include enough of the NSR piping run to ensure these anchors are included and thereby ensure the piping and anchor intended functions are maintained.

In Section 4.4.d, one acceptable option is to identify NSR piping runs that are connected at both ends to SR piping and include the entire run of NSR piping. It has for the basis that all NSR piping between the ends of the SR piping is conservatively included in the scope of license renewal ensuring the analysis endpoint is enveloped.

The NSR piping connected to Penetrations X-24 and X-23 does not require seismic supports to be identified on the SLRBD to be consistent with the guidance in NUREG-2192 Section 2.1.3.1.2 and NEI 95-10 Appendix F. As shown on boundary drawing SLR-36042-2, all NSR piping within containment connected to Penetrations X-24 and X-23 is highlighted green to identify that it is all in scope per 10 CFR 54.4(a)(2) for spatial/structural impacts of the the NSR components on SR

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 5 Page 3 of 3 components. In addition, per SLRA Section 2.1.4.2.3, all NSR piping supports for non-seismic or seismic II/I piping systems with a potential for spatial interaction with SR SSCs, are included within the scope of SLR. These supports are addressed in a commodity fashion within the civil/structural area review.

Associated SLRA Revisions:

No SLRA changes have been identified as a result of this response.

6 RAI 4.3.7-1

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 6 Page 1 of 2 RAI 4.3.7-1 Regulatory Basis Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background

SLRA Section 4.3.7 addresses the EAF TLAA, including the EAF screening evaluation to determine the limiting EAF locations.

SLRA Section 4.3.7 indicates that, after the screening evaluation, the applicant removed the conservatism associated with the screening environmental cumulative usage factor (CUFen) values in more detailed EAF evaluation to determine the refined CUFen values for 80 years of operation, as described in SLRA Table 3.4.7-1.

Issue However, the SLRA does not clearly describe how the conservatism was removed from the screening CUFen values to determine the 80-year projected CUFen values.

Request Describe how the applicant refined the screening CUFen values to determine the CUFen values listed in SLRA Table 3.4.7-1 and the technical basis of the refinement. If the refinement of the screening CUFen values relied on code provisions such as ASME Code provisions, describe specific references to the code provisions.

Response to RAI 4.3.7-1 The following steps were taken, as applicable, to reduce conservatism for screening CUFen values listed in SLRA Table 4.3.7-1.

Screening Fen values used were bounding values for screening purposes. Refined Fen values were calculated, consistent with NUREG/CR-6909, Revision 1 (ML16319A004),

in some cases based on load set temperatures and / or strain rate.

NUREG/CR-6909 Revision 1 fatigue curves were used in lieu of ASME design curves. This results in a decrease in CUF values for carbon and low-alloy steel locations. For stainless steel and nickel-based alloy locations, there is generally an increase in CUF values using the NUREG/CR-6909 Revision 1 fatigue curve, which can vary depending on where on the curve most of the alternating stress and allowable cycle values fall.

Load set pairs were added to account for varying proportions of cycle accumulation.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 6 Page 2 of 2 Since operating basis earthquake (OBE) is an independent transient, where OBE was combined with transients such as Heatup and Cooldown, the OBE load set was combined with normal operating conditions.

Sm (the temperature-dependent design stress intensity) averaging per Note 1 of ASME Section III, 1980 Edition with Addenda through Summer 1982, Figure NB-3222-1 was used in the calculation of Ke (strain concentration factor) where applicable.

Associated SLRA Revisions:

There are no SLRA Revisions associated with this response.

7 RAI 4.3.7-2

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 7 Page 1 of 2 RAI 4.3.7-2 Regulatory Basis Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background

SLRA Section 4.3.7 addresses the environmentally-assisted fatigue (EAF) TLAA. The SLRA section indicates that the EAF screening evaluation to determine the limiting locations (also called sentinel locations) uses bounding environmental fatigue correction factor (Fen) based on material types.

Issue However, SLRA Section 4.3.7 not clearly describe how the applicant calculated the bounding Fen values.

Request

1. Describe how the applicant calculated the bounding Fen values in terms of determining the (1) strain rate, (2) sulfur content for carbon and low alloy steels and (3) dissolved oxygen in the reactor coolant as the input to the Fen calculations. As part of the response, provide the technical basis for the bounding nature of the Fen values.

2. Clarify whether the maximum temperature referenced in relation to the Fen calculations (in the first sentence on SLRA page 4.3-19) means the maximum service temperature of each component.

Response to RAI 4.3.7-2

1. For EAF screening, bounding Fen values were calculated using the equations in NUREG/CR-6909, Revision 1 (ML16319A004) as follows:

1) The lowest (bounding) strain rate value was used for each material type.

2) The maximum (bounding) S*, sulfur content, value of 3.47 (Sulfur > 0.015 wt. %)

was used for carbon and low alloy steels.

3) For carbon and low alloy steels, dissolved oxygen (DO) values for each chemistry regime (Normal Water Chemistry, Hydrogen Water Chemistry alone, and On-Line Noble Chemistry) in each DO zone were used. Within each DO zone, Fen was calculated as an average weighted by time in each chemistry regime.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 7 Page 2 of 2 For stainless steels (SS), the bounding O* (dissolved oxygen) value of 0.29 was used for all chemistry regimes, based on hydrogen water chemistry.

For nickel-based alloys (NBA), the O* value associated with each chemistry regime was used to calculate Fen, which was then averaged based on time in each chemistry regime.

2. The maximum temperature referenced in Enclosure 16 of Reference 1 is the EPU design temperature of 549 F, which was used as the service temperature for each component as a bounding temperature.

References:

1. L-MT-23-025, Monticello Nuclear Generating Plant, Docket No. 50-263, Renewed Facility Operating License No. DPR-22, Subsequent License Renewal Application Supplement 2, ML23177A218.

Associated SLRA Revisions:

There are no SLRA Revisions associated with this response.

8 RAI 4.3.7-3

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 8 Page 1 of 2 RAI 4.3.7-3 Regulatory Basis Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background

SLRA Section 4.3.7 addresses the EAF TLAA. The SLRA section indicates that the limiting locations described in NUREG/CR-6260 for older vintage General Electric boiling water reactor (BWR) plants are evaluated in the EAF analysis. The locations in NUREG/CR-6260 include the recirculation outlet nozzle, as indicated in the SLRA.

Issue SLRA Section 4.3.7 does not list the recirculation outlet nozzle as one of the NUREG/CR-6260 locations. Instead, the SLRA section identifies the recirculation outlet nozzle location as one of the additional plant-specific evaluation locations subject to EAF screening. The SLRA explains that the recirculation outlet nozzle is screened out because its bounding screening CUFen (also called screening Uen) is less than the screening threshold of 1.0.

However, it is not clear to staff whether the screening evaluation of the recirculation outlet nozzle includes both the nozzle body and the adjacent piping location (e.g., safe end or safe end weld). In addition, the SLRA does not clearly discuss whether the recirculation outlet nozzle body is bounded by the recirculation inlet nozzle body or other reactor pressure vessel nozzle bodies.

Request

1. Clarify whether the screening evaluation of the recirculation outlet nozzle includes both the nozzle body and the adjacent piping location (e.g., safe end or safe end weld). If not, provide justification for why the screening evaluation for EAF does not include both the nozzle body and the adjacent piping location (e.g., safe end or safe end weld). As part of the response, describe the fabrication materials of the nozzle body and the adjacent safe end and weld as baseline information.

2. Clarify whether the recirculation outlet nozzle body is bounded by the recirculation inlet nozzle body or other reactor pressure vessel nozzle bodies in the EAF analysis and the basis of the determination. If not, provide the following information for the recirculation outlet nozzle body, consistent with the information in SLRA Table 4.3.7-1: (1) 80-year projected CUF; (2) Fen; and (3) 80-projected CUFen.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 8 Page 2 of 2 Response to RAI 4.3.7-3

1. The screening evaluation of the recirculation outlet nozzle does not include both the nozzle body and adjacent piping location. The screening evaluation included only the stainless steel safe end to pipe weld. This is because the safe end to pipe weld was determined to be the bounding location for the recirculation outlet nozzle as shown in the original Reactor Pressure Vessel (RPV) stress report and verified during the evaluations for the SLRA. The nozzle body is fabricated from low alloy steel, whereas the safe end and piping are fabricated from stainless steel. The bounding Fen for the carbon steel nozzle body is 8.508 and the bounding Fen for the stainless steel safe end and adjacent piping is 8.346. The original vessel stress report usage at the bounding safe end to pipe weld is larger than the nozzle body by more than a factor of 5. Since the bounding Fen values are very similar and the bounding location for the recirculation outlet nozzle is the stainless steel safe end to pipe weld location, EAF was not calculated for both locations.

2. The recirculation outlet nozzle body is bounded by the recirculation inlet nozzle body. In the original RPV stress report only the bounding usage was reported at the stainless steel safe end. For the outlet nozzle body, based on the allowable cycles of 2500 shown for the carbon steel nozzle body and the original cycles of 400 evaluated, the original outlet nozzle body usage is 400/2500 = 0.16. This is less than the value of 0.2 for the recirculation inlet nozzle body.

Associated SLRA Revisions:

There are no SLRA Revisions associated with this response.

9 RAI 4.3.7-4

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 9 Page 1 of 2 RAI 4.3.7-4 Regulatory Basis Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background

SLRA Section 4.3.7 addresses the EAF TLAA, including the EAF screening evaluation to determine the limiting EAF locations (also called sentinel EAF locations).

Issue However, the LRA does not clearly describe the following items related to the screening evaluation: (1) how the applicant determined thermal zones or sections that group certain components and piping lines for proper comparisons of the screening CUFen values considering the applicable transient conditions; (2) whether the limiting EAF location is determined for each material type (e.g., stainless steel, nickel alloy, and carbon/low alloy steel); and (3) how the applicant compared the highest values of the screening CUFen values to determine the final limiting locations (e.g., how the limiting locations were determined when the highest CUFen values are close to each other in a thermal zone).

Request Describe the following items regarding the EAF screening evaluation: (1) how the applicant determined thermal zones or sections that group certain components and piping lines for adequate comparisons of the screening CUFen values considering the applicable transient conditions; (2) whether the limiting EAF location is determined for each material type (e.g.,

stainless steel, nickel alloy and carbon steel); and (3) how the applicant compared the highest values of the screening CUFen values to determine the final limiting locations (e.g., screening process when the highest CUFen values are close to each other in a thermal zone). As part of the response, discuss the technical basis of the applicants approach for the items discussed above.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 9 Page 2 of 2 Response to RAI 4.3.7-4 The EAF screening approach is based on the concept of Sentinel Locations where a limited number of locations will experience the highest environmental fatigue usage. Therefore, monitoring the Sentinel Locations effectively manages EAF for all locations. This extends the concept that was used in NUREG/CR-6260 (ML11356A333) and adds a semi-quantitative ranking system to demonstrate that each plant component exposed to reactor coolant having a fatigue analysis can be represented by at least one Sentinel Location, or if the CUFen value is below 1.0, does not need to be represented by a Sentinel Location. One of the necessary aspects of the approach is to ensure screening is performed on a consistent basis.

The EAF screening evaluation considered the following in determining thermal zones:

(1) A thermal zone is defined as a collection of piping and/or vessel components which are evaluated to the same group of thermal transients during plant operations. Within a thermal zone, thermal shocks and thermal bending stresses vary depending only on the materials, geometry, and location of the component in the system. Therefore, the EAF screening evaluation established thermal zones based on the set of contributing design transients.

(2) While not all thermal zones necessarily contain multiple material types, if the thermal zone does contain multiple material types, the limiting location is determined for each material type to ensure both CUF and Fen values are considered in determining CUFen values.

(3) Within each material type in a thermal zone, the location with the highest bounding CUFen is selected; the location with the second highest CUFen is also selected if both of the following are true:

The top CUFen 1.0 and The top two CUFen values are within a factor of 25%

Associated SLRA Revisions:

There are no SLRA Revisions associated with this response.

0 RAI B.2.2.1-1

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 0 Page 1 of 3 RAI B.2.2.1-1 Regulatory Basis Pursuant to 10 CFR 54.21(a)(3), the SLRA must demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the subsequent period of extended operation.

Background

In its supplement, enclosure 18 dated June 26, 2023 (ADAMS Accession No. ML23177A218),

the applicant indicated that flexible power operation at MNGP started in 2019 and that the flexible power operation includes reducing power to 80 percent, allowing for windmills to operate when wind generation is predicted to be greater than demand. The applicant also explained that the flexible power operations and load-following changes in reactor power have minor impact on temperature (less than 50 °F) and pressure and have negligible impact on fatigue analyses.

Issue In the supplement, the applicant did not clearly discuss the basis for the determination that the flexible power operations and the associated load-following changes have negligible impact on fatigue analyses and, therefore, cycle counting is not needed.

Request Provide the following information to clarify the technical basis of the applicants determination that the flexible power operations have negligible impact on fatigue analyses: (1) pressure variation in the flexible power operation; (2) whether the temperature and pressure variations associated with the flexible power operation result in cyclic stresses below the fatigue endurance limit for the reactor coolant pressure boundary components and piping; and (3) if the answer to item (2) above is no, additional information to demonstrate that the flexible operation cycles have negligible impact on the cumulative usage factor (CUF) of reactor coolant pressure boundary components and piping (e.g., bounding CUF contribution of a flexible operation cycle and the total number of flexible operation cycles for 80 years of operation to confirm that the total CUF contributions are negligible).

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 0 Page 2 of 3 Response to RAI B.2.2.1-1 MNGP is primarily run as a baseload unit at 100% power. Flexible (flex) power operation at MNGP started in 2019. Flexible power operation includes reducing power to 80%, allowing for windmills to operate when wind generation is predicted to be greater than demand. Flexible power operations and load-following changes in reactor power have minor impact on temperature (< 50oF) and pressure and, as a result, have negligible impact on fatigue analyses.

Reactor pressure is controlled to a reactor pressure versus reactor power curve within +/- 10 psig.

1) The reactor pressure versus reactor power curve increases 10 psig from 75% to 100%

power, which results in a 980 psig minimum and 1010 psig maximum pressure. The maximum pressure change during flexible power operation is approximately 30 psig.

2) Pressure changes of up to 30 psig correspond to temperature changes of up to 3.6oF (544oF to 547.6oF). These pressure and temperature changes result in cyclic stresses that are below the fatigue endurance limit (< 50oF) for the reactor coolant pressure boundary components and piping.

3) Because the pressure and temperature changes result in cyclic stresses that are below the fatigue endurance limit, flexible operation cycles have negligible impact on the cumulative usage factor (CUF) of reactor coolant pressure boundary components and piping (e.g., bounding CUF contribution of a flexible operation cycle and the total number of flexible operation cycles for 80 years of operation to confirm that the total CUF contributions are negligible).

The SLRA revision included in this RAI response includes changes made in Enclosure 18 of Reference 1 (shown in bold black font).

References:

1. L-MT-23-025, Monticello Nuclear Generating Plant, Docket No. 50-263, Renewed Facility Operating License No. DPR-22, Subsequent License Renewal Application Supplement 2, ML23177A218.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-034 0 Page 3 of 3 Associated SLRA Revisions:

SLRA Section B.2.2.1 on page B-23 is revised to include additional detail about flexible power operations as follows:

B.2.2.1 Fatigue Monitoring The MNGP Fatigue Monitoring AMP is an existing preventive AMP that manages fatigue damage of RPV components, RCPB piping components, and other components. This AMP provides an acceptable basis for managing fatigue of components that are subject to fatigue or cycle-based TLAAs or other analyses that assess fatigue or cyclical loading.

The Fatigue Monitoring AMP monitors and tracks the number of critical thermal, pressure, and seismic transients to ensure that the CUF and CUFen for each analyzed component does not exceed the applicable limit through the SPEO. The program monitors and tracks the number of thermal and pressure transients as specified in USAR Table 4.2-1.

Load-following operation is a design option, as described in USAR Sections 3.2.5 and 3.3.3.2.2. MNGP is primarily run as a baseload unit at 100% power. Flexible (flex) power operation at MNGP started in 2019.

Flexible power operation includes reducing power to 80%, allowing for windmills to operate when wind generation is predicted to be greater than demand. Flexible power operations and load-following changes in reactor power have minor impact on temperature (<50F) and pressure and have negligible impact on fatigue analyses. Pressure changes up to approximately 30 psig occur during flex power operations. This corresponds to temperature changes of up to 3.6oF (544oF to 547.6oF).

These pressure and temperature changes result in cyclic stresses that are below the fatigue endurance limit for the reactor coolant pressure boundary components and piping, and so have negligible impact on CUF of components and do not require counting by the Fatigue Monitoring AMP.