ML23214A242

Breakout Questions - Aging Management Audit - Monticello Unit 1 - Subsequent License Renewal Application

ML23214A242
Person / Time
Site:	Monticello
Issue date:	05/25/2023
From:	Mary Johnson NRC/NRR/DNRL/NLRP
To:
References
Download: ML23214A242 (153)

Text

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions BREAKOUT QUESTIONS Aging Management Audit Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application February 27, 2023 - May 25, 2023

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA Section 4.2.1.2 RVI Neutron Fluence Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

4.2.1.2 4.2-7 In Section 4.2.1.2 of the SLRA the licensee discusses RAMA Fluence Methodology for determining fluence in BWR top guide and core shroud components under BWRVIP-145NP-A. The associated SE for BWRVIP-145NP-A states that

..for licensing actions provided that the calculational results are supported by sufficient justification that the proposed values are conservative for the intended application.

The licensee states in Section 4.2.1.2 the same guidance as BWR145-VIP was applied for determining conservative fluence all RVI components evaluated.

Please provide a detailed description of how the RAMA methodology was applied to determining fluence to all RVI components. Please include descriptions of the various conservatisms applied to justify use of RAMA code for Monticello SLRA.

2 4.2.1.1 and 4.2.1.2 4.2-8 and 4.2-9 For the RVI components the licensee states that the maximum fast neutron fluence (E >1.0 MeV) is specifically reported for the various RVI components. The statement is followed with list of components and the location within the component where the maximum fluence is reported.

Please provide a brief description of geometric modeling approach for each of the components of the RPV and the RVI.

Include any modeling limitations, conservatisms, or special considerations.

Section 2.3.3.9 - Fire System Section 2.4 - Scoping and Screening Results: Structures Section 2.4.6 - Fire Protection Barriers Commodity Group Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

Section 2.3.3.9 2.3-45 SLRA Table 2.3.3-9 of the SLRA does not include the following fire protection components:

• halon bottles

• sprinkler

• standpipe risers

• valve body

• Stainer housing

• Filter housing Verify whether the fire protection components listed are within the scope of SLRA in accordance with 10 CFR 54.4(a) and whether they are subject to an aging management review (AMR) in accordance with 10 CFR 54.21(a)(1). If any of the listed components are not within the scope of SLRA and are not

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions

• hanger and piping support

• seismic support for standpipes system piping

• Intake traveling screen/trash rack

• floor drains for removal of fire-fighting water

• curbs and dike for oil spill confinement

• station transformer fire suppression system and components subject to an AMR, the staff requests that the applicant provide justification for the exclusion.

2 Section 2.4 2.4-1 SLRA Section 2.4, Scoping and Screening Results:

Structures, provides the scoping and screening results of various structures within the scope of license renewal and subject to an AMR. Further, SLRA Table 2.4-6 provide the results of scoping and screening of fire barriers. However, scoping and screening results do not provide the type of fire barriers present in various in-scope building structures of the plant.

The staff requested that the applicant provide details of fire barrier type in-scope building structure of the plant in accordance with 10 CFR 54.4(a), and subject to an AMR, in accordance with 10 CFR 54.21(a)(1). For example, the staff requested that the applicant provide a list of buildings within the scope of license renewal with fire proofing material applied to some of their structural steel members or components as part of fire barriers.

3 Section 2.4.6 2.4-14 SLRA Table 2.4-5 of the SLRA does not include the following fire protection components.

• Fire wrap/electric raceway fire barrier

• Radiant energy shield Verify whether the fire protection fire barriers listed are within the scope of SLRA in accordance with 10 CFR 54.4(a) and whether they are subject to an aging management review (AMR) in accordance with 10 CFR 54.21(a)(1). If any of the listed components are not within the scope of SLRA and are not subject to an AMR, the staff requests that the applicant provide justification for the exclusion.

B.2.3.4 - BWR Vessel ID Attachment Welds Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 1

A.2.3.4.

B.2.3.4 SLRA Section A.2.3.4 states, in part, the following:

• The MNGP BWR Vessel ID Attachment Welds AMP is part of the MNGP ASME Section XI Inservice Inspection Program (Section A.2.2.1). The BWR Vessel ID Attachment Welds AMP is in accordance with approved relief request under the BWRVIP Administrative Manual and provides for condition monitoring of the BWR Vessel ID Attachment Welds.

• The AMP is updated periodically as required by 10 CFR 50.55a and the BWRVIP.

SLRA Section B.2.3.4 states the following:

• Under 10 CFR 50.55a relief, examination of BWR vessel ID attachment welds are completed

• exclusively under the guidance of the BWRVIP program documents in lieu of ASME Section XI requirements, including schedule, extent, frequency, sequence of exams, re-examinations, and additional examinations (Reference ML16208A462).

Similarly - the following documents seem to indicate that following the Z1 alternative approval in 2016 (see ML16208A462) to use BWRVIP-48-A in lieu of ASME Section XI - that future approvals are no longer needed

• EWI-08.01.02.pdf

• EWI-11.01.22, Revision 2 BWR VESSEL ID ATTACHMENT WELDS.pdf XCELMO00017-REPT-044_Rev1_BWR_Vessel_Welds.pdf

• Similarly, AMP basis document suggests that going forward into the SPEO - the Z1 alternative approval in 2016 (see ML16208A462) to use BWRVIP-48-A in lieu of ASME Section XI is permanent and future approvals are no longer needed Discuss the relevance of the approved Z1 alternative during the SPEO - given that the approval in ML16208A462 is only through the 5th ISI interval, which expired on May 31, 2022.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Excerpt from Safety Evaluation in 2016 (see ML16208A462):

Based on the information provided in the licensee's submittals, the NRC staff concludes that the alternatives proposed by the licensee will ensure that the integrity of the RVI surfaces, attachments, and core support structures is maintained with an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(z)(1),

the licensee's proposed alternative for MNGP is authorized for the fifth 10-year ISi interval, which ends on May 31, 2022.

GALL-SLR XI.M4 states, in part, the following:

• Program

Description:

The program includes inspection and flaw evaluation in accordance with the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, and the guidance in BWR Vessel and Internals Project, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines [Boiling Water Reactor Vessel and Internals Project (BWRVIP)-48-A] to provide reasonable assurance of the long-term integrity and safe operation of BWR vessel ID attachment welds.

• Parameters Monitored or Inspected: The program monitors for cracks caused by SCC, IGSCC, and cyclical loading mechanisms. Inspections performed in accordance with the guidance in BWRVIP-48-A and the requirements of the ASME Code,Section XI, Table IWB-2500-1 are used to interrogate the components for discontinuities that may indicate the presence of cracking.

2 B.2.3.4 SLRA Section B.2.3.4 - Operating Experience stated the following:

Plant-Specific Operating Experience Were there any minor flaws or relevant conditions that were specifically related to Vessel ID Attachment Welds?

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions

• Owners Activity Reports (OARs) were reviewed over the last six years (2015-2021). For all minor flaws or relevant conditions that required evaluation or repair/replacement activities for continued service, all items were either examined and evaluated as acceptable or appropriate replacements were completed.

Or was this just meant to be a generic statement that the licensee looked at the OARs/.?

B.2.3.28 - Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.3.28 B-207 to B-212 The Monticello Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks will be a new condition monitoring AMP that will manage the aging effect of loss of coating/lining integrity.

• Xcel Energy Change Request (CR) No. 608000000142 describes a coating inspection evaluation for condensate storage tank (CST) T-1A. CR No. 608000000150 describes a coating inspection evaluation for CST T-1B.

• Both CRs 608000000142 and 608000000150 describe coating blisters and pinpoint rusting andrecommend that the entirety of the CST coatings will need to be inspected if life is to be extended past 2030.

Describe the plan to inspect the entirety of the CST coatings to support subsequent license renewal and operation past 2030. Be prepared to discuss:

• When these inspections will be performed

• Inspection methods Acceptance criteria B.2.2.1 - Fatigue Monitoring Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.2.1 B-24 SLRA Section B.2.2.1 addresses Enhancement 2 regarding the parameters monitored or inspected program element of the Fatigue Monitoring Aging Management Program (AMP).

In the enhancement, the plant procedure for the AMP will be updated to identify and require monitoring of the 80-year plant design cycles, or projected cycles that are

1. Explain why Enhancement 2 of the Fatigue Monitoring AMP and SLRA Table 4.3.1-1 do not include plant loading and unloading transients even though USAR Section 3.2.5 indicates that MNGP Unit 1 performs load-following operation. As part of the response, clarify whether the load-following operation has a negligible

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions utilized as inputs to component CUFen calculations, as applicable. The plant design transients are listed in SLRA Table 4.3.1-1. The SLRA table does not include plant loading and unloading transients that may be associated with load-following operation.

In comparison, Monticello USAR Section 3.2.5, which addresses the performance characteristics of normal operation, states that variable recirculation flow control provides limited manual load-following capability for a BWR. The discussion in the USAR section indicates that the Monticello Nuclear Generating Plant (MNGP) Unit 1 performs load-following operation.

However, Enhancement 2 and SLRA Table 4.31-1 do not include monitoring of plant loading and unloading transients. The staff needs information regarding how the applicant evaluated the effects of the plant loading and unloading transients on fatigue monitoring and analyses.

effect on the fatigue analyses. If so, discuss why the load-following operation has a negligible effect on the fatigue analyses, including the evaluation of the pressure and temperature variations during load-following operation and their effects on fatigue.

2 B.2.2.1 B-25 4.3-4 SLRA Section B.2.2.1 addresses Enhancement 4 regarding the monitoring and trending program element of the Fatigue Monitoring AMP.

In the enhancement, the plant procedure for the AMP will be updated to require that trending be performed to ensure that the fatigue parameter limits will not be exceeded during the subsequent period of extended operation.

However, the SLRA does not clearly describe which fatigue parameters will be monitored against their limits in the enhancement. The staff needs clarification on this item.

1. Clarify whether the following fatigue parameters are included in the monitoring and trending of Enhancement 4 to ensure that their limits are not exceeded: (1) transient cycles used in the calculations of the 80-year projected cumulative usage factor (CUF) and environmental CUF (CUFen); (2) transient cycles used in the fatigue waiver evaluation; and (3) transient cycles used in the CUF calculations for high energy line break location postulation (e.g., break location postulation based on a CUF threshold of 0.1). If these fatigue parameters are not included in Enhancement 4, explain why the enhancement does not include these

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions parameters in the monitoring and trending of Enhancement 4.

Cumulative Fatigue Damage Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

3.2.2.2.1 3.3.2.2.1 3.4.2.2.1 3.2-8 3.3-21 3.4-8 SLRA Section 3.2.2.2.1 addresses the further evaluation of the time-limited aging analysis (TLAA) on the cumulative fatigue damage for engineering safety features (ESF) components. Similarly, SLRA Sections 3.3.2.2.1 and 3.4.2.2.1 address the TLAA on the cumulative fatigue damage for the components in the auxiliary systems and the components in the steam and power conversion systems, respectively.

These sections indicate that the metal fatigue TLAAs for the components in the ESF, auxiliary systems, and steam power conversion systems are discussed in SLRA Section 4.3. The fatigue TLAAs are also referenced in the following SLRA Table 1 (system-level) items: (1) 3.2-1, 001; (2) 3.3-1, 002; and (3) 3.4-1, 001.

In SLRA Sections 3.2.2.2.1, 3.3.2.2.1 and 3.4.2.2.1, the applicant explained that more detailed Table 2 (subsystem-level) items related to these Table 1 items are not listed in the SLRA because the TLAA disposition for these Table 1 items is 10 CFR 54.21(c)(1)(i) and no aging management program is credited for these items.

Examples of subsystems in the context of this discussion are the core spray system and high pressure coolant injection system in the ESF.

However, GALL-SLR and SRP-SLR (i.e., SLR guidance documents) identify the following Table 2 (subsystem-level) items in relation to the aging management and Clarifying the following Table 2 items in the SLR guidance documents are applied to the components that are evaluated in the SLRA fatigue TLAAs: (1)

V.D2.E-10 related to Table 1 item 3.2-1, 001; (2) VII.E3.A-34, VII.E3.A-62 and VII.E4.A-62 related to Table 1 item 3.3-1, 002; and (3) VIII.B2.S-08 and VIII.D2.S-11 related to Table 1 item 3.4-1, 001. If so, revise the SLRA to identify these Table 2 items or similar items to clarify the applicability of the Table 1 items and metal fatigue TLAAs to specific component groups.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions TLAA evaluation: (1) V.D2.E-10 related to Table 1 item 3.2-1, 001; (2) VII.E3.A-34, VII.E3.A-62 and VII.E4.A-62 related to Table 1 item 3.3-1, 002; and (3) VIII.B2.S-08 and VIII.D2.S-11 related to Table 1 item 3.4-1, 001.

The staff noted that the identification of these Table 2 items in the guidance documents is not just based on credited aging management programs but also based on the evaluation of relevant TLAAs, as indicated in the column heading, Aging Management Program (AMP)/TLAA of the sixth column of SRP-SLR Tables 3.X-1 that includes the term, TLAA.

Therefore, the staff needs to resolve this apparent between the SLRA and the SLR guidance documents, which results from the omission of the Table 2 items related to fatigue TLAAs in the SLRA.

2 3.5.2.2.2.5 3.5-35 SLRA Section 3.5.2.2.2.5 indicates that the evaluations of fatigue for component support members, anchor bolts, and welds for Groups B1.1, B1.2, and B1.3 component supports are TLAAs as defined in 10 CFR 54.3, and are addressed in SLRA Section 4.3, Metal Fatigue.

The component support groups are the following: (1)

Group B1.1: supports for ASME Code Class 1 piping and components; (2) Group B1.2: supports for ASME Class 2 and 3 piping and components; (3) Group B1.3: supports for ASME Class MC (metal containment) components.

SLRA Section 3.5.2.2.2.5 also indicates that Table 3.5-1, item 3.5.1-053 related to the further evaluation discussed above is not applicable to the Monticello Nuclear Generating Plant (MNGP). The SLRA further states that the only fatigue analysis related to plant structures is the analysis for cranes and lifting devices and for portions of the primary containment. In addition, the SLRA states

1. Provide justification for why the existing fatigue analysis for the bottom head support skirt of the reactor pressure vessel is not evaluated in SLRA Section 3.5.2.2.2.5. If justification cannot be provided, revise the section to include the evaluation of the fatigue TLAA for the bottom head support skirt.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions that the management of cumulative fatigue damage to cranes and lifting devices is addressed in SLRA item 3.3.1-001 and Section 4.6.1. The SLRA also states that the management of cumulative fatigue damage to primary containment is addressed in SLRA item 3.5.1-009 and Section 4.5.

In contrast, the staff noted that the following reference indicates that the 40-year cumulative usage factor (CUF) for the bottom head support skirt of the reactor pressure vessel is 0.2832 after excessive conservatisms were removed from the original 40-year CUF value (i.e., 0.4)

(

Reference:

GE Report NEDO-33322, Revision 3, Safety Analysis Report for Monticello Constant Pressure Power Uprate, Table 2.2-4, October 2008, ADAMS Accession No. ML083230112).

Therefore, the staff needs clarification on why the existing fatigue analysis for the bottom head support skirt, which may be a Group B1.1 support, is not evaluated in SLRA Section 3.5.2.2.2.5.

B.2.3.21 - Selective Leaching Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.3.21 B-161 SLRA Section B.2.3.21 states [t]he MNGP Selective Leaching AMP includes inspections of components made of gray cast iron, ductile iron, and copper alloys (except for inhibited brass) that contain greater than 15 percent zinc or greater than 8 percent aluminum [emphasis added by staff] exposed to a raw water, closed-cycle cooling water, treated water, waste water, or soil environment.

The staff did not identify AMRs associated with copper alloys that contain greater than 8 percent aluminum. The staff requests clarification that there are no copper alloy that contain greater than 8 percent aluminum components in scope for subsequent license renewal.

2 N/A N/A The staff reviewed AR 01228174 (last page) and WO 00414130 (pages 29 and 58 of 110) and noted references to ductile iron piping.

The staff did not identify AMRs associated with ductile iron piping. The staff requests a discussion with respect

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions to if there is ductile iron piping in-scope for subsequent license renewal.

3 N/A N/A The staff reviewed AR 01228174 (last page) and WO 00414130 (pages 29 and 58 of 110) and noted references to ductile iron piping.

The staff did not identify AMRs associated with ductile iron piping. The staff requests a discussion with respect to if there is ductile iron piping in-scope for subsequent license renewal.

4 B.2.3.21 B-161 SLRA Section B.2.3.21, Selective Leaching, states

[e]ach of the one-time and periodic inspections for the various material and environment populations comprises a 3 percent sample or a maximum of 10 components.

NUREG-2222, Disposition of Public Comments on the Draft Subsequent License Renewal Guidance Documents NUREG-2191 and NUREG-2192, provides the basis for reducing the extent of inspections for selective leaching during the subsequent period of extended operation (i.e.,

3 percent with a maximum of 10 components per GALL-SLR guidance) when compared to the extent of inspections for selective leaching during the initial period of extended operation (i.e., 20 percent with a maximum of 25 components per GALL Report, Revision 2 guidance).

Part of the basis for reducing the extent of inspections is that industry operating experience (OE) has not identified instances of loss of material due to selective leaching which had resulted in a loss of intended function for the component.

The NRC issued Information Notice (IN) 2020-04, Operating Experience Regarding Failure of Buried Fire Protection Main Yard Piping, to inform the industry of OE involving the loss of function of buried gray cast iron fire water main yard piping due to multiple factors, including graphitic corrosion (i.e., selective leaching),

overpressuration, low cycle fatigue, and surface loads. As noted in the IN, a contributing cause to the failures of Based on recent industry operating experience, the staff requests a discussion with respect to using the reduced sample size (i.e., 3 percent with a maximum of 10 components) for gray cast iron piping exposed to soil.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions buried gray cast iron piping at Surry Power Station (SPS) was the external reduction in wall thickness at several locations due to graphitic corrosion.

5 N/A N/A GALL-SLR Report AMP XI.M33, Selective Leaching, allows for the external surfaces of buried components to be excluded from the scope of the program based on the condition of external coatings and cathodic protection efficacy.

XCELMO00017-REPT-087 (Selective Leaching program basis document) states [p]lant-specific operating experience and implementation of preventive actions at MNGP supports the exclusion of buried components that are externally coated in accordance with GALL-SLR Report AMP XI.M41 for selective leaching.

SLRA Section B.2.3.27, Buried and Underground Piping and Tanks, states [t]he most recent annual cathodic protection system survey performed in 2021 determined that not all of the surveyed locations met the -850 mV polarized potential criterion for buried steel components[t]herefore, the cathodic protection system does not currently meet the acceptance criteria of NACE SP0169-2007 or NACE RP0285-2002 and is not credited as a preventive measure at MNGP.

Based on its audit and review of the SLRA, the staff noted instances of buried piping coating failures. The staff requests a discussion with respect to excluding buried components from the scope of the Selective Leaching program based on this observation.

Based on the past performance of the cathodic protection system, the staff requests a discussion with respect to excluding buried components from the scope of the Selective Leaching program based on cathodic protection efficacy.

4.3.1 Year Transient Cycle Projections Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

4.3.1 4.3-1 SLRA Section 4.3.1 addresses the 80-year transient cycle projections.

The SLRA section states that the Fatigue Monitoring program data does not list the following transients from the USAR list: (1) reactor overpressure at 1375 psig

1. Clarify whether the Fatigue Monitoring AMP will monitor the cycles of the following transients for the subsequent period of extended operation: (1) reactor overpressure at 1375 psig transient; (2) hydrostatic test to 1560 psig transient;

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions transient; (2) hydrostatic test to 1560 psig transient; (3) rapid blowdown transient; (4) liquid poison flow at 80F transient; (5) operating basis earthquake (OBE) transient; and (6) safety/relief valve actuations transient.

The staff needs clarification as to whether the above statement in SLRA Section 4.3.2 means that these transients will not be monitored in the Fatigue Monitoring Aging Management Program (AMP).

(3) rapid blowdown transient; (4) liquid poison flow at 80F transient; (5) operating basis earthquake (OBE) transient; and (6) safety/relief valve actuations transient. If not, explain why the cycle counting of the transients is not necessary to manage the aging effects of cumulative fatigue damage.

2. If these transients are rare events, clarify whether the applicant will evaluate the impact of the occurrence of these transients when these transients occur during the subsequent period of extended operation.

2 4.3.1 4.3-1 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).

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 can better represent the operational characteristics of the subsequent period of extended operation rather than the prior cycle accumulation rates).

The staff also noted that the safety/relief valve lifts

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 operation in terms of cycle calculations. For item (2) discussed

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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).

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 period up to May 31, 2021 represents the operating characteristics for the subsequent period of operation in terms of cycle calculations.

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.

3 4.3.1 4.3-1 SLRA Section 4.3.1 addresses the 80-year transient cycle projections. Specifically, SLRA Table 4.3.1-1 describes the 80-year transient cycle projections. The cycle

1. Clarify whether the following non-USAR-listed transients have an impact on the existing fatigue wavier evaluations

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions projections and the associated design input are also described in the following document (

Reference:

SIA Calculation 2100507.301, Revision 0, 80-Year Cycle Projections for Monticello).

SLRA Table 4.3.1-1 indicates that the following transients are not listed in the 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.

As discussed above, these transients do not have USAR-specified cycle limits and the 80-year projected cycles for these transients are very low (i.e., 1 cycle for each transient). Considering these unique aspects of the transients, the staff needs clarification on whether these transients have an impact on the fatigue waiver evaluations described in SLRA Section 4.3.2.

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 in terms of the continued validity of the fatigue waiver evaluations for the subsequent period of extended operation.

If not, provide the technical basis for why these transients do not have an impact on the fatigue waiver evaluations described in SLRA Section 4.3.2.

4 4.3.1 4.3-1 SLRA Section 4.3.1 addresses the 80-year transient cycle projections. However, the SLRA section does not provide a time-limited aging analysis (TLAA) disposition in accordance with 10 CFR 54.21(c)(1)(i), (ii) or (iii).

The staff needs to clarify the basis for why the applicant did not identify a TLAA disposition for the 80-year transient cycle projections.

1. Describe the basis for why the applicant did not identify a TLAA disposition for the 80-year transient cycle projections.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 4.3.2 - ASME Section III, Class 1 Fatigue Waivers Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

4.3.2 4.3-6 SLRA Section 4.3.2 addresses the time-limited aging analysis (TLAA) on ASME Code Section III, Class 1 fatigue waiver evaluations.

On SLRA page 4.3-6, the heading of the paragraph for TLAA disposition indicates that the TLAA is dispositioned in accordance with 10 CFR 54.21(c)(1)(ii) that the analysis has been projected to the end of the subsequent period of extended operation.

In contrast, the text under the paragraph heading indicates that the TLAA is dispositioned in accordance with 10 CFR 54.21(c)(1)(iii) that the effects of aging on the intended function(s) will be adequately managed for the subsequent period of extended operation. Specifically, the applicant credited the Fatigue Monitoring Aging Management Program (AMP) to manage the aging effects of fatigue for the components subject to the fatigue waiver evaluations.

The staff needs to resolve these inconsistency regarding the TLAA dispositions.

In addition, the text below the paragraph heading for TLAA disposition states that the Fatigue Monitoring AMP will monitor the transient cycles which are the inputs to the fatigue waiver reevaluations and require action prior to exceeding design limits that would invalidate their conclusions.

The staff needs clarification on whether the design limits referenced in the paragraph discussed above mean the

1. Describe which provision of 10 CFR 54.21(c)(1) is used for the disposition of the TLAA on fatigue waiver evaluations.

2. Clarify whether the design limits referenced in the TLAA disposition mean the transient cycles used in the fatigue waiver evaluations that are described in SLRA Section 4.3.2. If not, provide additional information to clarify what the design limits refer to. As part of the response, the Fatigue Monitoring AMP will include the acceptance criteria for the transient cycles used in the fatigue waiver evaluations.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions transient cycles used in the fatigue waiver evaluations that are described in SLRA Section 4.3.2.

2 4.3.2 4.3-6 SLRA Section 4.3.2 describes the time-limited aging analysis (TLAA) on ASME Code Section III, Class 1 fatigue waiver evaluations. SLRA Section 4.3.2 indicates that the following reactor pressure vessel components have existing (original) fatigue waiver evaluations: (1) main closure flange, (2) head cooling spray and instrumentation nozzles; (3) vent nozzle; (4) instrumentation nozzles; (5) jet pump instrumentation nozzles; (6) intermediate range monitor/source range monitor (IRM/SRM) dry tube; (7) power range detector assembly; and (8) in-core detector assembly.

However, the staff noted that the design stress analyses for the instrumentation nozzles (N11A-B and N12A-B nozzles) and jet pump instrumentation nozzles (N8A-B nozzles) in the following references do not include a fatigue waiver evaluation or cumulative usage factor calculation for these nozzles (

References:

(1) CA-68-667, Revision 1, Section S14, Stress Analysis of Instrumentation Nozzles N11A-B and N12A-B, Monticello-NSP Reactor Vessel, 3/12/1993; and (2) CA-68-668, Revision 1, Section S15, Stress Analysis of Jet Pump Instrumentation, Nozzles N8A & B, Monticello-NSF Reactor Vessel, 3/12/1993).

Therefore, the staff needs clarification on why SLRA Section 4.3.2 identifies fatigue waiver TLAAs for these nozzles even though there is no existing fatigue wavier evaluation (e.g., evaluation per ASME Code Section III, NB-3222.4(d)) for these nozzles.

In addition, the following reference indicates that the jet pump instrumentation nozzles are exposed to the same transient of cooling from normal operating temperature as

1. Explain why SLRA Section 4.3.2 identifies fatigue waiver TLAAs for the instrumentation nozzles and jet pump instrumentation nozzles even though there is no existing fatigue waiver evaluation for these nozzles in the following references for design stress analyses (

References:

(1) CA-68-667, Revision 1, Section S14, Stress Analysis of Instrumentation Nozzles N11A-B and N12A-B, Monticello-NSP Reactor Vessel, 3/12/1993; and (2) CA-68-668, Revision 1, Section S15, Stress Analysis of Jet Pump Instrumentation, Nozzles N8A & B, Monticello-NSF Reactor Vessel, 3/12/1993).

2. If a fatigue waiver evaluation exists for the instrumentation nozzles and jet pump instrumentation nozzles in the current licensing basis of the Monticello Nuclear Generating Plant (MNGP), provide a document reference for the evaluation and discuss the technical basis for the TLAA disposition (e.g., why the fatigue waiver evaluation remains valid with the 80-year cycle projections).

3. In addition, clarify whether the jet pump instrumentation nozzles are bounded by the steam outlet nozzle in terms of CUF. If so, provide the technical basis of the bounding nature of the steam outlet nozzle for the jet pump

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions the steam outlet nozzle (

Reference:

Structural Integrity Associates, Calculation 2100507.309, Revision 0, Fatigue Exemption of Monticello Reactor Vessel (RPV)

Components, Section 4.4). SLRA Table 4.3.3-1 also indicates that the 80-year projected cumulative usage factor (CUF) for the steam outlet nozzle is 0.1872.

The staff needs clarification on whether the jet pump instrumentation nozzles are bounded by the steam outlet nozzle in terms of CUF. The staff also needs clarification on whether the instrumentation nozzles are bounded by the other RPV nozzles (e.g., feedwater nozzles and recirculation inlet nozzles) evaluated in SLRA Table 4.3.3-1 in terms of CUF.

instrumentation nozzles.

4. Clarify whether the instrumentation nozzles are bounded by the other RPV nozzles evaluated in SLRA Table 4.3.3-1 in terms of CUF. If so, provide the technical basis of the bounding nature of the other RPV nozzles for the instrumentation nozzles.

3 4.3.2 4.3-6 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.

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

1. Describe 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). As part of the response, clarify whether an aging management program can manage the effects of cumulative fatigue damage in relation to the fatigue waiver evaluations for these nozzles.

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.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions following transients are not listed in the 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.

If not, provide the technical basis for why these transients do not have an impact on the validity of the fatigue waiver evaluations.

4 4.3.2 4.3-6 SLRA Section 4.3.2 describes the time-limited aging analysis (TLAA) on ASME Code Section III, Class 1 fatigue waiver evaluations.

Note (4) of SLRA Table 4.3.2-1 describes the significant pressure fluctuation cycles for the following components in relation to the fatigue wavier evaluations: (a) main closure flange, (b) IRM/SRM dry tube, (c) power range detector assembly and (d) in-core detector assembly.

Note (4) of SLRA Table 4.3.2-1 indicates that different cycles of the loss of feedwater pumps transient are estimated for these components as significant pressure fluctuation cycles. For example, additional 18 cycles of the transient are estimated for the power range detector assembly, compared to those for the IRM/SRM dry tube, in the fatigue waiver evaluations for 80 years of operation.

1. Clarify why different cycles of the loss of feedwater pumps transient are estimated for the components discussed in the issue section as significant pressure fluctuation cycles in the fatigue waiver evaluations.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 4.3.4 Fatigue Analysis of RPV Internals Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

4.3.4 4.3-12 SLRA Section 4.3.4 addresses the time-limited aging analysis (TLAA) on the fatigue of reactor pressure vessel internal (RVI) components. SLRA Section 4.3.4 explains that the most significant fatigue loading occurs at the jet pump diffuser to baffle plate weld location and, therefore, this location is bounding for all other fatigue affected RVI components.

The applicant indicated that the original 40-year design analysis for this limiting jet pump location estimated a 40-year cumulative usage factor (CUF) of 0.35. By the extrapolation of this 40-year CUF, the applicant determined that the 60-year projected CUF for the limiting jet pump location is approximately 0.5.

The applicant also explained that the 60-year CUF value in the current licensing basis is conservatively bounding for the 80-year CUF of the limiting jet pump location.

However, the applicant did not clearly discuss why the 60-year projected CUF for the jet pump diffuser to baffle plate weld location, which is based on the extrapolation of the 40-year CUF, is bounding for the 80-year CUF calculation (e.g., discussion that supports the bounding nature of the 60-year cycles assumed in the 60-year CUF calculation in comparison with 80-year projected cycles of relevant transients).

The staff noted that the following reference describes the transients used in the original 40-year CUF calculation for the limiting jet pump location (

Reference:

General Electric

1. Describe how the applicant used 40-year transient cycles and their contributions to the limiting CUF of the jet pump location in the determination that the 60-year limiting CUF is conservatively bounding for the 80-year CUF calculation.

2. In related to the first request above, clarify the following items: (1) actual (current) cycles and 80-year projected cycles of the transients used in the CUF calculation for the limiting jet pump location; (2) which transient in SLRA Table 4.3.1-1 is identical to the sudden start of cold pump transient in the GE report; and (3) whether HPCI startup and the sudden start of cold pump transients are transients in the emergency condition.

3. If the 60-year CUF value (0.5) for the limiting location is not large enough to bound the 80-year CUF calculation, revise the 80-year CUF value and the TLAA disposition in accordance with 10 CFR 54.21©(1)(ii) or (iii).

4. Discuss whether the other RVI components may have a 80-year CUF greater than the 60-year CUF for the jet

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Report APED-5460, Design and Performance of General Electric Boiling Water Reactor Jet Pumps, September 1968).

Specifically, Figure 4-25 in the General Electric (GE) report indicates that the following transients and their cycles were used in the 40-year CUF calculation for the limiting jet pump location: (1) HPCI [high pressure coolant injection] startup transient cycles of 16 with a CUF contribution of 0.11; (2) startup and shutdown transient cycles of 114 with a CUF contribution of 0.13; (3) sudd+E27n start of cold pump transient cycles of 5 with a CUF contribution of 0.01; and (4) design basis accident transient cycles of 1 with a CUF contribution of 0.1.

The staff needs clarification on how the 40-year transient cycles and their contribution to the limiting CUF were used in the applicants determination that the 60-year CUF is bounding for the 80-year CUF calculation for the limiting jet pump location.

pump location because of additional transient cycles projected to be accumulated for 80 years of operation.

2 4.3.4 4.3-12 SLRA Section 4.3.4 addresses the fatigue TLAA for the reactor pressure vessel internal (RVI) components. SLRA Section 4.3.4 explains that the most significant fatigue loading occurs at the jet pump diffuser to baffle plate weld location and, therefore, this location is bounding for all other fatigue affected RVI components.

In comparison, the following GE report addresses the fatigue analysis RVIs in relation to the extended power uprate (EPU) that was incorporated into the current licensing basis of the Monticello Nuclear Generating Plant in 2013 (

Reference:

GE Report NEDO-33322, Revision 3, Safety Analysis Report for Monticello Constant Pressure Power Uprate, October 2008).

1. Clarify why Table 2.2-9 in the GE report does not identify the jet pump diffuser to baffle plate weld location as the most limiting CUF location for RVI components. As part of the response, if Table 2.2-9 in the GE report has a minimum CUF threshold for the CUF listing in the table, describe such a minimum CUF threshold value.

2. Clarify whether the feedwater sparger and jet pump diffuser to baffle plate weld locations are bounding for the other RVI components in terms of 80-year projected CUF. If not, describe the potentially more

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Specifically, Table 2.2-9 in the GE report indicates that the 60-year cumulative usage factor (CUF) of the feedwater sparger is estimated to be 0.32.

However, Table 2.2-9 in the GE report does not provide a 60-year CUF value of the jet pump diffuser to baffle plate weld location that is identified as the most limiting CUF location for the RVI components.

limiting CUF locations and their 80-year projected CUF values based on the most recent CUF analyses.

4.3.5 ASME Section III, Class 1 Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

4.3.5 4.3-13 SLRA Section 4.3.5 addresses the time-limited aging analysis (TLAA) on the fatigue of ASME Code Section III, Class 1 piping systems. In relation to the fatigue TLAA, the following reference describes that cumulative fatigue usage (CUF) analysis for the Class 1 piping systems in detail (

Reference:

Structural Integrity Associates (SIA),

Calculation Package 2100507.303, Revision 0, 80-year Fatigue Analysis of Selected Class 1 Reactor Coolant Pressure Boundary (RCPB) Piping, June 24, 2022).

Section 4.2 of the SIA report discusses the fatigue analysis for the core spray line. The section indicates that the analysis in the report considered the original condition as well as the rerate condition that bounds the temperature and pressure conditions of the EPU.

With respect to the core spray line, Table 4 of the SIA report indicates that the thermal transient of the original condition is 546 °F to 80 °F in comparison with the thermal transient of the rerate condition, 549 °F to 80 °F.

In addition, the table indicates that the original condition for pressure is 1000 psi in comparison with the rerate pressure of 1025 psi.

1. Resolve the apparent inconsistency between the SIA report and SLRA Section 4.3.5 in terms of the EPU condition evaluated in the fatigue TLAA (i.e., 549 °F versus 546 °F in the thermal transients, and 1025 psi versus 1000 psi in the pressure levels).

2. Clarify the following items: (1) whether the original condition in the SIA report means the pre-EPU condition; (2) the relationship between the rerate condition and the EPU condition (e.g., whether the rerate condition is the term used to represent the bounding condition for the EPU); and (3) whether the rerate condition is part of the EPU license amendment approved on December 9, 2013 (ADAMS Accession No. ML13316B298).

3. Clarify whether the fatigue TLAA in SLRA Section 4.3.5 evaluates the

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Based on these data, the applicant determined a pressure ratio of 1.025 for the implementation of the EPU (i.e.,

1025/1000).

However, SLRA Section 4.3.5 indicates that the thermal transient from 546 °F to 80 °F, which corresponds to the original thermal transient according to the SIA report, is the basis for the EPU condition and the associated loading that are evaluated in the fatigue analysis for the core spray line.

Similarly, with respect to the residual heat removal (RHR) intertie line, Table 8 in the SIA report indicates that the first thermal transient of the original condition is 150 °F to 546 °F and the first thermal transient of the rerate condition, which bounds the EPU condition, is 150 °F to 549 °F.

In contrast, SLRA Section 4.3.5 indicates that the fatigue analysis of the RHR intertie line evaluates the thermal transient from 150 °F to 546 °F, which corresponds to the original thermal transient according to the SIA report.

Therefore, the staff needs to resolve the apparent inconsistency between the SIA report and SLRA Section 4.3.5 in terms of the EPU condition that is evaluated in the fatigue TLAA.

The staff also needs clarification on the following items:

(1) whether the original condition in the SIA report means the pre-EPU condition; (2) the relationship between the rerate condition and the EPU condition (e.g., whether the rerate condition is the term used to represent the bounding condition for the EPU); and (3) whether the rerate condition is part of the EPU license amendment bounding pressure and temperature conditions of the EPU. If not, provide justification for why the fatigue TLAA does not evaluate the bounding conditions of the EPU.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions approved on December 9, 2013 (ADAMS Accession No. ML13316B298).

2 4.3.5 4.3-13 SLRA Section 4.3.5 addresses the time-limited aging analysis (TLAA) on the fatigue of ASME Code Section III, Class 1 piping systems.

As part of the fatigue TLAA, SLRA Section 4.3.5 explains that the recirculation and residual heat removal (RHR) piping systems were reanalyzed in 2005 through 2006.

The SLRA section also indicates that the recirculation piping, including inlet nozzle safe ends and RHR supply and return lines to the containment penetrations, was replaced 1985 and that the bounding cumulative usage factor (CUF) is 0.923 at a branch line connection to the RHR intertie line.

SLRA Section 4.3,5 further explains that the applicant reanalyzed the bounding location for 80 years of operation, considering the cycles adjusted to remove the cycles accumulated before piping replacement and the EPU condition. In addition, the SLRA section indicates that the 80-year CUF was estimated to be 0.399 in the reanalysis for 80 years of operation.

Given the relatively large reduction in the bounding CUF value for the recirculation and RHR piping systems (from 0.923 to 0.399), the staff needs clarification on the following items: (1) the operating time period for which the previous CUF value of 0.923 was determined (e.g., 60 years or 40 years); (2) how the applicant removed the conservatism from the previous CUF (0.923) in addition to the removal of the cycles accumulated before the piping replacement; and (3) whether the Fatigue Monitoring aging management program (AMP) will monitor, if any, reduced cycles used in the 80-year CUF analysis, as opposed to the original design cycles in case such

1. Given the relatively large reduction in the bounding CUF value for the recirculation and RHR piping systems (from 0.923 to 0.399), clarify the following items: (1) the operating time period for which the previous CUF value of 0.923 was determined (e.g., 60 years or 40 years); (2) how the applicant removed the conservatism from the previous CUF (0.923) in addition to the removal of the cycles accumulated before the piping replacement; and (3) whether the Fatigue Monitoring AMP will monitor, if any, reduced cycles used in the 80-year CUF analysis, as opposed to the original design cycles, in case such reduced cycles are assumed in the 80-year CUF analysis.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions reduced cycles are assumed in the 80-year CUF analysis.

3 4.3.5 4.3-13 SLRA Section 4.5.3 indicates that the fatigue analyses for Class 1 components were performed in accordance with the provisions in ASME Boiler and Pressure Vessel Code,Section III, 1980 Edition with Addenda through Summer 1982.

The staff needs more specific references to the Code provisions that the applicant used in the fatigue TLAA.

1. Describe more specific references to the Code provisions (e.g., paragraphs) that the applicant used in the fatigue TLAA.

4.3.6 - ASME Section III, Class 2 and 3 and ANSI B31.1 Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

4.3.6 4.3-14 SLRA Section 4.3.6 addresses the fatigue time-limited aging analysis (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.

However, LRA Table 4.3.6-1 does not clearly describe how the 40-year cycles were determined (e.g., piping system design information, plant operation procedures, test requirements, UFSAR 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

1. Describe how the 40-year cycles were determined (e.g., based on piping system design information, plant operation procedures, test requirements, UFSAR 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, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions needs clarification on whether these cycles were estimated by summing up two or more design cycles for each of the piping systems.

2 4.3.6 4.3-14 SLRA Section 4.3.6 addresses the fatigue TLAA for ASME Code Section III Class 2 and 3 and ANSI B31.1 piping systems.

In comparison, Updated Safety Analysis Report (USAR),

Appendix I describes the current licensing basis (CLB) evaluation of the high energy line breaks outside the containment. Specifically, USAR Appendix I,Section I.3.1 describes the postulation of break locations and the screening criteria that are used to determine the break locations in the HELB analysis.

USAR Appendix I,Section I.3.1 indicates that the postulation of HELB locations is, in part, based on the allowable stress range for expansion stress (SA),

consistent with Branch Technical Position 3-3, Appendix B (ADAMS Accesso No. ML070800027).

SA may need to be adjusted by a stress range reduction factor based on the number of transient cycles that are evaluated in the implicit fatigue analysis (SLRA Section 4.3.6).

However, SLRA Section 4.3.6 does not identify the HELB analysis as a TLAA based on the HELB location postulation that involves SA and the associated cycle-dependent stress range reduction factor.

1. Provide justification for why SLRA Section 4.3.6 does not identify the HELB analysis as a TLAA even though the screening criteria of the HELB location postulation involves the dependency on the transient cycles. If justification cannot be provided, identify the HELB analysis as a TLAA and provide the disposition of the TLAA. In addition, revise the USAR supplement as needed.

4.3.7 Environmentally-Assisted Fatigue Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 1

4.3.7 4.3-19 SLRA Section 4.3.7 addresses the environmentally-assisted fatigue (EAF) time-limited aging analysis (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.

However, the SLRA sections does not clearly describe how the applicant calculated the bounding Fen values.

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.

2. Describe how the applicant considered the normal water chemistry (NWC) and hydrogen water chemistry (HWC) operations and the associated dissolved oxygen contents in the determination of the bounding Fen values. As part of the response, clarify whether the Fen values for specific periods associated with specific water chemistry operations are separately calculated for the respective time periods. In addition clarify whether the time periods, for which the HWC operation is not available, assume to have the NWC condition.

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

2 4.3.7 4.3-19 SLRA Section 4.3.7 addresses the environmentally-assisted fatigue (EAF) time-limited aging analysis (TLAA).

The SLRA section indicates that the limiting locations described in NUREG/CR-6260 for older vintage General Electric BWR plants are evaluated in the EAF analysis.

The locations in NUREG/CR-6260 include the recirculation outlet nozzle, as also indicated in the SLRA.

1. Describe the following information to confirm that the screening CUFen threshold of 1.0 is low enough to be used in the screening evaluation: (1) the screening CUFen value of the recirculation outlet nozzle and (2) the highest screening CUFen value in the EAF screening evaluation for the

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions However, 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 also 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.

The staff needs the following information to confirm that the screening CUFen value of 1.0 is low enough to be used in the EAF screening evaluation: (1) the screening CUFen value of the recirculation outlet nozzle and (2) the highest screening CUFen value in the EAF screening evaluation for the fabrication material of the recirculation outlet nozzle and the associated component.

In addition, the staff needs clarification on 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).

fabrication material of the recirculation outlet nozzle and the associated component.

2. Clarify whether the screening EAF evaluation of the recirculation nozzle includes both the nozzle body and the adjacent piping location (e.g., safe end or safe end weld). If not, explain why both the nozzle and the adjacent piping locations are not evaluated in the EAF screening. As part of the response, describe the fabrication materials of the nozzle body and the adjacent safe end and weld.

3 4.3.7 4.3-19 SLRA Section 4.3.7 addresses the environmentally-assisted fatigue (EAF) time-limited aging analysis (TLAA),

including the EAF screening evaluation to determine the limiting EAF locations.

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 location is determined for each material type (e.g.,

stainless steel, nickel alloy, and carbon/low alloy steel);

1. 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 proper comparisons of the screening CUFen values considering the applicable transient conditions; (2) whether the limiting 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

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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).

the final limiting locations (e.g., screening process when the highest CUFen values are close to each other in a thermal zone).

4 4.3.7 4.3-19 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 conservatisms associated with the screening 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. However, the SLRA does not clearly describe how the conservatisms were removed from the screening CUFen values to determine the 80-year projected CUFen values.

1. Describe how the applicant removed the conservatisms from the screening CUFen values to determine the refined CUFen values listed in SLRA Table 3.4.7-1.

5 4.3.7 4.3-19 SLRA Section 4.3.7 addresses the environmentally-assisted fatigue (EAF) time-limited aging analysis (TLAA).

The following reference discusses the EAF analysis for the NUREG/CR-6260 locations applicable to the Monticello Nuclear Generating Plant, which is a older vintage GE plant (

Reference:

Structural Integrity Associates (SIA), Calculation Package No.

2100507.305P, Monticello Environmentally Assisted Fatigue Analysis for 80 Years, NUREG/CR-6260 Locations, June 24, 2022).

Section 3.1 of the SIA report indicates that, for the loss of recirculation pumps transient, the design cycle number (i.e., 20 cycles) is used in the EAF analysis for 80 years of operation. Based on the use of the design cycles in the EAF analysis for the recirculation and residual heat removal (RHR) return piping, the SIA report recommends

1. Clarify whether the applicant confirmed that the use of the design cycles (20 cycles) of the loss of recirculation pumps transient is adequate for the EAF analysis. As part of the response, provide the actual cycles and 80-year projected cycles of the transient to demonstrate the bounding or representative nature of the design cycles (i.e., 20 cycles) for the 80-year projected cycles of the transient.

2. In addition, provide justification for why SLRA Table 4.3.1-1 does not list the loss of recirculation pumps transient and related cycle information even though the SIA report uses the transient in the EAF TLAA. If justification cannot be provided,

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions a review of operating logs to validate this assumption for the transient cycle number used in the EAF analysis.

SLRA Section 4.3.7 does not clearly indicate whether the applicant confirmed that the use of the design cycles (20 cycles) of the loss of recirculation pumps transient is adequate for the EAF analysis.

In addition, SLRA Table 4.3.1-1, which describes the 80-year projected cycles for fatigue analyses, does not list the loss of recirculation pumps transient even though the SIA report uses the transient in the EAF TLAA.

revie SLRA Table 3.4.1-1 to include the transient and the related cycle information, consistent with the existing format of the table.

3. In addition, clarify whether the Fatigue Monitoring aging management program (AMP) will monitor the loss of recirculation pumps transient cycles for the subsequent period of extended operation. If not, provide justification for why such monitoring of the transient is not needed for the transient cycles.

Buried and Underground Piping and Tanks Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.3.27 B-197 B-198 SLRA Section B.2.3.27, Buried and Underground Piping and Tanks, states the following:

• [t]he number of inspections for each 10-year inspection period, commencing 10 years prior to the start of SPEO, are based on the inspection quantities noted in NUREG-2191, Table XI.M41-2 for Category F.

• [t]he cathodic protection system does not currently meet the acceptance criteria of NACE SP0169-2007 or NACE RP0285-2002 and is not credited as a preventive measure

[emphasis added by staff] at MNGP.

Inspection quantities in GALL SLR AMP XI.M41 Preventive Action Category F are based on a cathodic protection system being credited as a preventive measure (although performance criteria are not being met to use Preventive Action Category C). For instances where cathodic protection is not credited as a preventive measure, the staffs expectation would be that the applicant would state an exception and develop plant specific inspection quantities. The staff requests a discussion with respect to why Preventive Action Category F is appropriate for steel piping at Monticello, given that cathodic protection is not credited as a preventive measure.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 2

B.2.3.27 B-198 The table on this page includes inspection quantities for underground steel and stainless steel piping.

The staff requests a discussion with respect to the cathodic protection acceptance criteria that will be used during the SPEO. The staff disagrees with the use of the 100 mV cathodic polarization acceptance criterion (in the mixed metal environment) without confirmatory testing to verify that all metals are adequately protected. In addition, the staff does not agree with the use of a potential of -850 mV instant on without measurement or calculation of voltage drops.

3 B.2.3.27 B-197 B-198 SLRA Section B.2.3.27 states [t]he most recent annual cathodic protection system survey performed in 2021 determined that not all of the surveyed locations met the 850 mV polarized potential criterion for buried steel components. The survey also determined that not all of the surveyed locations met the 100 mV polarization criterion.

The staff requests a discussion with respect to the cathodic protection acceptance criteria that will be used during the SPEO. The staff disagrees with the use of the 100 mV cathodic polarization acceptance criterion (in the mixed metal environment) without confirmatory testing to verify that all metals are adequately protected. In addition, the staff does not agree with the use of a potential of -850 mV instant on without measurement or calculation of voltage drops.

4 N/A N/A UFSAR Section 11.6, Cooling Tower System, states [a]

single underground steel pipe conveys the water from both pumps to two cooling towers.

Staff request a discussion with respect to if this piping is in-scope for subsequent license renewal.

5 Table 3.3 1

3.3-54 SLRA Table 3.3-1, Summary of Aging Management Evaluations for the Auxiliary Systems, item 3.3.1-108 (which includes stainless steel tanks exposed to concrete) is classified as not applicable.

Item 3.3.1-108 is classified as N/A; however, SLRA Table 3.3.2-17 includes stainless steel tanks exposed to concrete. Staff requests a discussion with respect to why item 3.3.1-108 is classified as N/A.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 6

N/A N/A GALL Report AMP XI.M41 recommends a cathodic protection critical potential of 1,200 mV to prevent damage to coatings or base metals.

The staff reviewed SL-008367 and noted pipe to soil potential measurements as negative as -23,921 mV with respect to a zinc reference cell between the years 1976 and 2003.

The staff requests a discussion with respect to cathodic protection critical potentials and the subject observation from SL-008367. The staff could not identify this GALL-SLR Report recommendation in current procedures or in any of the enhancements.

7 B.2.3.27 B-205 SLRA Section B.2.3.27 states [a]nnual chloride concentration samples had been increasing from 2011 to 2015. [t]he increase in chloride concentration was likely due to salt treatment during the winter months.

The staff reviewed AR 01383079 and noted the coating inspection of the buried stainless steel piping revealed coating failures but no signs of degradation.

The staff requests a discussion with respect to if the subject piping inspection, or any other piping inspections, have be conducted in the vicinity of where salt treatments (i.e., chlorides) have been applied.

8 N/A N/A The staff reviewed AR 01240741 and noted the following:

(a) a high temperature line from the heating boiler to the cold shop going into the ground was not visibly coated; and (b) site specification for coating and wrapping underground piping is only applicable to lines with a maximum fluid temperature of 160 Fahrenheit.

Please upload the original construction specification(s) for external coatings for in-scope metallic piping (carbon steel, gray cast iron, and stainless steel). In addition, the staff requests a discussion with respect to if there are any uncoated high temperature lines in-scope for subsequent license renewal 9

N/A N/A As an alternative to visual examinations of piping, GALL-SLR Report AMP XI.M41 allows the following: [a]t least 25 percent of the in-scope piping constructed from the material under consideration is pressure tested on an interval not to exceed 5 years. The piping is pressurized to 110 percent of the design pressure of any component within the boundary (not to exceed the maximum allowable test pressure of any nonisolated components) with test pressure being held for a continuous eight hour interval.

XCELMO00017-REPT-072 (Buried and Underground Procedure 1404-01 does not appear to meet the pressure testing guidance outlined in AMP XI.M41. The staff seeks a clarifying discussion with respect to if pressure testing will be used as an alternative to visual examinations (and if so, whether pressure testing will be conducted in accordance with AMP XI.M41).

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Piping and Tanks program basis document) allows for "periodic pressure testing of buried emergency service water piping." The statement references procedure 1404-01, EDG ESW Heat Exchanger Performance Testing, Revision 21.

10 N/A N/A NSP-DOL-0598, Diesel Fuel Oil Tanks Inspections, Revision 6, refences a "2003 T-44 internal inspection."

Please upload this document to the ePortal.

N/A 11 B.2.3.27 B-199 SLRA Section B.2.3.27 includes the following enhancement: [s]tate that new and replacement backfill shall meet the requirements of NACE SP0169-2007 Section 5.2.3 or NACE RP0285-2002 Section 3.6.

GALL-SLR Report AMP XI.M41 states [t]he staff considers backfill that is located within 6 inches of the component that meets ASTM D448-08 size number 67to meet the objectives of NACE SP0169-2007 and NACE RP0285-2002.

The maximum allowable backfill size per ASTM D448-08 (size number 67) is one inch.

The staff reviewed MPS-0984 and noted the maximum size of structural backfill shall be two inches in confined areas where hand tamping is required and four inches in other areas.

Although new backfill will be consistent with AMP XI.M41 guidance, existing backfill does not appear to meet guidance outlined in AMP XI.M41. The staff requests a discussion on this topic.

12 N/A N/A Section 4.10, Operating Experience, of XCELMO00017-REPT-072 (Buried and Underground Piping and Tanks program basis document) states MNGP will refurbish its cathodic protection system 5 years prior to the SPEO to meet the acceptance criteria of -850 mV relative to a CSE (instant off), or acceptance criteria alternatives, for buried and underground steel components.

The staff did not identify any enhancements related cathodic protection system refurbishment or acceptance criteria. In addition, the staff did not identify any commitments related to cathodic protection system refurbishment or acceptance criteria in SLRA Table A-3, List of SLR Commitments and Implementation

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Schedule. The staff requests a discussion on this topic.

13 N/A N/A GALL-SLR Report AMP XI.M41 states [a]ging effects associated with fire mains may be managed by either: (a) a flow test as described in Section 7.3 of NFPA 25 at a frequency of at least one test in each 1-year period; (b) monitoring the activity of the jockey pump (e.g., pump starts, run time) on an interval not to exceed 1 month; or (c) an annual system leak rate test. If the aging effects are not managed by one of these alternatives, and the extent of inspections is not based on the percentage of piping for that material type, then two additional inspections are added to the inspection quantity for that material type.

XCELMO00017-REPT-072 (Buried and Underground Piping and Tanks program basis document) discussion on periodic flow testing of buried fire main piping references procedure 0268, Fire Protection System Flow Test, Revision 24, which specifies flow tests are to be performed every three years.

Flow testing frequency in procedure 0268 is longer than AMP XI.M41 recommendations. The staff requests a discussion on this topic.

14 B.2.3.27 B-201 The acceptance criteria program element of GALL-SLR Report AMP XI.M41 includes recommendations related to the following:

• (g) flow tests for fire mains

• (i) jockey pump activity

• (j) fire water system leak rate testing

• (l and m) cathodic protection acceptance criteria The staff did not identify any enhancements in SLRA Section B.2.3.27 associated with the acceptance criteria for fire main flow tests, jock pump monitoring, fire water system leak rate testing, or cathodic protection surveys.

The staff request a discussion with respect to why there are no enhancements related to these four recommendations from AMP XI.M41.

15 B.2.3.27 B-202 B-203 The corrective actions program element of GALL-SLR Report AMP XI.M41 includes recommendations related to the following:

• (d) cathodic protection survey results

• (e) leakage during pressure tests The staff did not identify any enhancements in SLRA Section B.2.3.27 associated with the corrective actions for cathodic protection survey results, leakage during pressure tests, jock pump monitoring, or indications of cracking.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions

• (f) jockey pump monitoring

• (g) indications of cracking The staff request a discussion with respect to why there are no enhancements related to these four recommendations from AMP XI.M41.

16 A.2.2.27 A-26 A-27 GALL-SLR Report Table XI-01, FSAR Supplement Summaries for GALL-SLR Report Chapter XI Aging Management Programs, includes the following for AMP XI.M41:

• [f]or steel components, where the acceptance criteria for the effectiveness of the cathodic protection is other than -850 mV instant off, loss of material rates are measured.

• [i]f a reduction in the number of inspections recommended in GALL-SLR Report, AMP XI.M41, Table XI.M41-2 is claimed based on a lack of soil corrosivity as determined by soil testing, then soil testing is conducted once in each 10-year period starting 10 years prior to the subsequent period of extended operation.

The staff noted the following recommendations from GALL SLR Report Table XI-01 (cathodic protection criteria and soil corrosivity testing) are not included in SLRA Section A.2.2.27, Buried and Underground Piping and Tanks. The staff requests a clarifying discussion to understand the basis for excluding these recommendations.

B.2.3.7 BWR Vessel Internals Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.3.7 B-60

• On February 19, 2021, the Electric Power Research Institute (EPRI) issued a Transfer of Information Notice regarding potential non-conservatism in EPRI software.

On March 19, 2021, EPRI issued an updated Transfer of Information Notice. These documents are related to new fracture toughness data of irradiated stainless steel weld metal and are publicly available at ADAMS Accession Number ML21084A164. Subsequent to submitting BWRVIP-315 for NRC review, EPRI submitted letter dated January 20, 2022, to respond to staff concerns related to the new data and its impact on the BWRVIP-315 topical report (ADAMS Accession Number ML22025A113). The Describe plant-specific actions responding to the identified EPRI letters and ensuring that the applicants aging management program for vessel internals accounts for the latest available information.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions applicant references BWRVIP-315 for aging management of vessel internals but does not address these emerging concerns.

2 B.2.3.7 B-60 The applicant references BWRVIP-315 for aging management of vessel internals. The staff review of BWRVIP-315 is ongoing at this time. During their review, the staff noted that BWRVIP-315 references Code Case N-889 to determine irradiation assisted crack growth rate of stainless steel. This code case is conditioned in the latest version of Regulatory Guide 1.147 incorporated by reference to Title 10 of the Code of Federal Regulations 50.55a.

Describe licensee implementation of Code Case N-889 to calculate crack growth rate, given the staffs conditions on the code case.

3 B.2.3.7 B-60 The applicant references BWRVIP-315 for aging management of vessel internals. The staff review of BWRVIP-315 is ongoing at this time. The applicant states that they will follow the inspection and evaluation guidelines in BWRVIP-47-A, which states in Section 3.2.4 that baseline inspection results will be reviewed by the BWRVIP and, if deemed necessary, reinspection recommendations will be developed at a later date During their review, the staff noted that the BWRVIP has not finalized evaluation of baseline examination results and formulation of guidance regarding reinspection of lower plenum components.

Describe and justify reinspection plans for lower plenum components.

4 B.2.3.7 B-60 The applicant references BWRVIP-315 for aging management of vessel internals. The staff review of BWRVIP-315 is ongoing at this time. The staff noted that BWRVIP-315 contains limitations in Section 4.5.1.

Describe how the licensee will implement the identified limitations in BWRVIP-315.

3.3.2.2.7 Loss of Material Due to Recurring Internal Corrosion Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

3.3.2.2.7 3.3-26 3.3-27 The staff reviewed AR 01396233 and noted the following:

(a) degradation of the rx/rad waste chilled water piping is a long standing issue; (b) localized repairs are unlikely to The staff requests a discussion with respect to if the subject operating experience (OE) is recurring internal

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions be possible due to the severe level of degradation; (c) chilled water piping located in the reactor building is in scope for license renewal; and (d) the piping is identified as carbon steel in a treated water environment.

SLRA Section 3.3.2.2.7, Loss of Material Due to Recurring Internal Corrosion, states [b]ased on plant-specific OE, recurring internal corrosion is an applicable effect for steel components in raw water systems that use water from the Mississippi River. The Open-Cycle Cooling Water System (B.2.3.11) AMP, Fire Water System (B.2.3.16) AMP, and Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)

AMP are enhanced and used to manage loss of material due to the recurring internal corrosion aging effect for steel piping, piping components, tanks, and heat exchanger components exposed to raw water.

corrosion. It is not clear to the staff if this OE is internal or external corrosion, or if the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program (which does not include any RIC specific enhancements) needs to be enhanced based on this OE (involving a treated water environment) or OE referenced in SLRA Section 3.3.2.2.7 (involving a raw water environment).

B.2.2.3 Environmental Qualification of Electric Equipment Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.2.3 B-33

• In the Program Overview and Background section of XCELMO00017-REPT-060, Revision 1, the applicant notes in the first bullet of the principal objective is ensuring that safety-related electrical equipment is capable of performing its function in a harsh environment. 10 CFR 50.49 includes consideration of equipment that is important to safety which includes, in part, safety-related electric equipment, nonsafety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions, and certain post-accident monitoring equipment.

Explain the apparent exclusion of nonsafety-related electric equipment and certain post-accident monitoring equipment and how this equipment will be considered in the EQ AMP for the subsequent period of extended operation.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 2

B.2.2.3 B-33 XCELMO00017-REPT-060 notes that the EQ program health color is Red but trending as Improving. Program health is expected to turn Green at the end of 2022, reflected in the 1Q2023 program health report.

What is the current health color of the EQ program? If available, please upload the latest EQ health report to the ePortal.

3 B.2.2.3 B-33 In EWI-08.11.01, Equipment Qualification Users Manual, Revision 26, the applicant referenced Electric Power Research Institute (EPRI) Report NP-1558, A Review of Aging Theory and Technology, dated September 1980. EPRI recently updated this report (July 2020) due to issues/concerns with lack of or expired technical references for certain activation energies. The NRC staffs understanding is that this revision resulted in up to 30% of activation energies being removed from the database.

For environmentally qualified (EQ) components that the applicant used/relied upon EPRI Report NP-1558 as the justification/basis for activation energies for extending the qualified life of EQ equipment, has the licensee reviewed this revised document to verify that their justification/basis for activation energies remains valid for EQ components for the requested period of operation?

4 B.2.2.3 B-33 CD 5.11.pdf displays a validation error and is unable to be viewed.

Please upload a viewable version of this document on the ePortal for the staff to review.

Water Chemistry Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.3.2 B-44

• Section B.2.3.2 of the SLRA identifies not measuring hydrogen peroxide levels as an exception to NUREG-2191 Section XI.M2, Water Chemistry, as modified by SLR-ISG-2021-02-MECHANICAL. Element 3 of NUREG-2191 Section XI.M2 (as modified) states that Corrosive Parameters and Water Quality are measured and maintained in accordance with the EPRI BWR Water Chemistry Guidelines, EPRI Report 30020002623 BWRVIP-190, Revision 1 (BWRVIP-190). The NRC staff has not found where the EPRI Guidelines require or recommend measuring hydrogen peroxide levels as part of measuring ECP. Therefore, the determination of this as an exception to the guidance is unclear to the staff.

Please explain the basis for determining that not measuring hydrogen peroxide levels is an exception to the NUREG-2191 guidance. If hydrogen peroxide monitoring is a program requirement that you are taking exception to, please clarify how using online noble chemistry and hydrogen water chemistry and maintaining low chloride and sulfate levels in the reactor coolant, are acceptable alternatives.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 2

N/A N/A Section 2.2.3.2 of BWRVIP-190 outlines the three different action levels and the appropriate responses when each action is triggered. This includes how long it should take to collect and analyze additional samples based on which action level is achieved. The action levels and their responses are a needed requirement according to BWRVP-190. On pages 7 and 8 of the Chemistry Limits and Sampling Frequency document (II.05) plant responses to the action levels outlined in BWRVIP-190 are addressed. For a parameter exceeding Action Level 1 the document addresses the timeframe of sampling and analyzing additional samples and this time frame is consistent with BWRVIP-190. However, this timeframe is not stated for a parameter exceeding Action Level 2 and 3 in II.05.

Please clarify the Monticello Nuclear Generating Plant (MNGP) timeframe requirements for sampling and analyzing additional samples when a parameter exceeds Action Level 2 and 3.

3 N/A N/A BWRVIP-190 lists recommended chemistry parameters for reactor coolant water and their associated sampling frequencies. Many of these parameters are included in Table 3.1.1 of II.05. Most of these parameters have limits that have a guideline type of (( )). Many others reference BWRVIP-190 guidance and the associated limits for these parameters are discussed in Chapter 2 of BWRVIP-190. However, two parameters, boron and lithium concentrations, have (( )) listed in table 3.1.1 as the guidance type. ((

))

This pattern is also seen with insoluble iron. ((

))

Please clarify the following:

• How the MNGP requirements follow the EPRI Guidance with respect to good practice values for the concentration of lithium and boron in the reactor water at power operation conditions.

How the MNGP requirements follow the EPRI Guidance with respect to insoluble iron and its associated limit at power operation conditions.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 4

N/A N/A

((

)) However, the staff did not find a justification for the feedwater H2, or control rod drive water dissolved O2 sampling frequency.

Please explain the basis for the deviation from the EPRI guidance for the sampling frequencies of feedwater H2 concentration and control rod drive water dissolved O2 concentration at power operation conditions.

5 N/A N/A In section 6.1.1 of the Plant Strategic Chemistry (II.01) plan the primary parameters to monitor the effectiveness of the OLNC are listed as being ((

)). However, in Table 2-2 in Volume 2 of BWRVIP-190 for the category of OLNC plants the primary parameter should be (( )). Also, in section 6.0 of the plant strategic chemistry and section 4.4.2 of the Plant Chemistry program plan it states, ((

)) From these documents it appears that mitigation is defined solely based on ((

)).

• Please clarify the primary parameter related to the mitigation of IGSCC.

• Please discuss how measurements of catalyst loading are incorporated into the implementing documents, such as Chemistry Limits & Sampling Frequency.

6 N/A N/A Section 6.1.3.A of the Plant Strategic Chemistry plan states that Hydrogen Water Chemistry (HWC) system availability will be calculated (( )). However, in table 3.16.1 of II.05 it states that Hydrogen Availability will be measured (( )), which is consistent with the frequency listed in BWRVIP-190.

Please clarify the frequency of measuring HWC availability.

7 N/A N/A In Table 3.16.1 of II.05, the action level 1 limit for feedwater dissolved oxygen is listed as being ((

)). The NRC staff interprets this to mean Action Level 1is entered if dissolved oxygen is below ((

)), for consistency with the referenced Table 2-11 of BWRVIP-190. However, Table 3.15 of II.05 has a similar Action Level 1 limit for dissolved oxygen in control rod drive water written as (( )).

Please explain the two different forms of what appear to be the same requirements for Action Level 1 (i.e., X-Y ppb in one case; <X ppb and >Y ppb in the other case).

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 8

SLRA Table 3.3.2-17 3.3-302 SLRA Table 3.3.2-17, for the Standby Liquid Control (SLC) system, states that the Water Chemistry and One-Time Inspection programs will be used to manage the loss of material and long-term loss of material for carbon steel piping in a sodium pentaborate solution environment.

Plant-specific Note 1 states that aging effects are managed by monitoring and controlling SLC poison storage tank treated water chemistry. Because SLC systems are constructed primarily from stainless steels (e.g., NUREG/CR-6001, ML040340671), the staff requests additional information about the carbon steel components and conditions of exposure.

Please provide the following information about the carbon steel components subject to loss of material and long-term loss of material in a sodium pentaborate internal solution (SLRA Table 3.3.2-17):

a. The specific components requiring aging management.

b. Location of the components in the SLC system.

c. Amount of time the components are exposed to the sodium pentaborate solution and expected corrosion rate.

d. A description of the sodium pentaborate solution to which the components are exposed, and how it differs (if it differs) from the sodium pentaborate solution in the SLC storage tank.

e. Given that the sodium pentaborate solution in the SLC storage tank is limited by TS 3.1.7, describe how water chemistry can be adjusted to manage the aging effects.

Stress Corrosion Cracking and Loss of Material (pitting, crevice) for Stainless Steel, Nickel Alloys, and Aluminum Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

Table 3.4.2-5 3.4-91, -

96,

-100 SLRA Table 3.4.2-5 (Off-Gas - Summary of Aging Management Evaluation) includes six table entries with None as both the aging effect requiring management and the aging management program for stainless steel components exposed internally to a condensate environment. Three of the entries list VIII.E.SP-118a and three list VIII.E.SP-127a as the NUREG-2191 AMR items.

The corresponding SLRA Table 1 AMR items are 3.4.1-Please clarify the basis for considering SCC and LOM not applicable to stainless steel in the off-gas condensate, and the basis for considering a One-Time Inspection unnecessary to confirm the aging effects are not occurring.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 002 and 3.4.1-003. These referenced AMR items correspond to stress corrosion cracking (SCC) and loss of material (LOM) due to pitting or crevice corrosion.

Each of these six table entries uses Note I for NUREG-2191 consistency, meaning the aging effect for these components is not applicable. Each item also uses plant-specific footnote 1, which states that SCC and LOM are not applicable aging effects for these components because the condensate environment represents off-gas that does not have the potential to contain halides.

• The reason for considering these aging mechanisms not applicable is not clear to the NRC staff. Based on other table entries, it appears that liquid water is assumed to be present (e.g., carbon steel is managed for LOM in condensate). Therefore, the staff interprets the plant-specific footnote to mean that condensate may be present in the off-gas system but it could not contain halides.

2 Table 3.2-1 3.2-39 3.2-40 SLRA Table 3.2-1, Summary of Aging Management Evaluations for the Engineered Safety Features, states that aging management for Items 3.2.1-107 and 3.2.1-108 is consistent with NUREG-2191. These items apply to insulated stainless steel and nickel alloy piping, piping components, and tanks exposed to air or condensation.

However, neither of these items are used in any Table 3.X.2-X, and no components of this type are listed in any Table 3.X.2-X. Therefore, it appears that Items 3.2.1-107 and -108 may be non-applicable, or applicable components may have been omitted from the 3.X.2-X aging management summary tables.

Please clarify the applicability of AMR Items 3.2.1-107 and 3.2.1-108 and describe any components to which they are applied.

A.2.2.7 BWR Vessel Internals A.2.2.19 Reactor Vessel Material Surveillance

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request add1

• SLRA Section A.2.2.7 BWR Vessel Internals - provides the FSAR Supplement for the BWR Vessel Internals aging management program - Given that the guidance in BWRVIP-315 and the EPRI letter (the one we brought up with the applicant during the TRP-9 Breakout Call) are integral parts for the AMP, the staff thinks it would be appropriate for these documents to be cited in the FSAR supplement.

add2

• BWRVIP-315 is currently still under NRC review and the approved -A report is expected to be forthcoming - it may be make sense to consider capturing this aspect in the FSAR supplement as well (similar to what was done in SLRA Section A.2.2.19 Reactor Vessel Material Surveillance related to BWRVIP-321-A and the pending review of a revision to this report).

B.2.3.19 Reactor Vessel Material Surveillance Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.3.19 SLRA Section B.2.3.19 - Enhancement states :

• Implement BWRVIP-321-A, Boiling Water Reactor Vessel and Internals Project, Plan for Extension of the BWR Integrated Surveillance (ISP) Through the Second License Renewal (SLR), upon obtaining NRC approval for MNGP to use BWRVIP-321-A to maintain compliance with 10 CFR Part 50, Appendix H.

Any reason the enhancement in SLRA Section B.2.3.19 doesnt correspond with Commitment No. 22, as well as text in SLRA Section A.2.2.19

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Commitment #22 - Related to SLRA Section A.2.2.19 states:

• Implement BWRVIP-321-A and subsequent NRC approved revisions upon obtaining NRC approval for MNGP to use BWRVIP-321-A to maintain compliance with 10 CFR Part 50, Appendix H.

Given EPRI submitted an appendix (i.e., Appendix F) to revise the Irradiation Schedule in BWRVIP-321-A -

Commitment #22 and FSAR supplement would capture this revision if it is approved by the NRC. However, the actual program enhancement discussion/review doesnt.

4.2.2 RPV Materials Upper Shelf Energy (USE) Reduction Due to Neutron Embrittlement 4.2.3 Adjusted Reference Temperature (ART) for RPV Materials Due to Neutron Embrittlement 4.2.4 RPV Thermal Limit Analysis: Operating P-T Limits Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

Upper Shelf Energy SLRA Section 4.2.2 states the following:

• Reference 4.7.12 establishes the maximum allowable percent decrease in USE for 72 EFPY operation. For BWR/3-6 plate materials, the maximum allowable percent decrease is given in Reference 4.7.12.

• Reference 4.7.12 - Bounding Upper Shelf Energy Analysis for Long Term Operation, Report sponsored by EPRI, Final Report, April 2017.

Applicant indicated during audit that this reference has not been released for publication, and is not required to support SLRA conclusion that regulatory limits are still met.

SLRA indicates that Reference 4.7.12 was needed to Reconcile the discrepancy between the SLRA and the explanation provided during the audit.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions support TLAA - however, applicant explanation during audit is contrary to this the SLRA 2

4.2.3 SLRA Tables 4.2.3-1 and 4.2.3-2 identify multiple welds with a lot identifier of E8018N - This includes the circumferential welds and axial welds

• For the axial welds - sigma initial is 12.7 degrees F

• For the circ weld - sigma initial is 0 degrees F SLRA Reference 4.7.13 - Monticello - Pressure and Temperature Limits Report (PTLR) up to 54 Effective Full Power Years (EFPY), Revision 1, August 2014 (ML14246A206)

• Identifies a component as Limiting Weld - Beltline

• Sigma initial is 12.7 degrees F

• Table 8 from PTLR Confirm that the component as Limiting Weld - Beltline in Rev 1 of the PTLR is the axial welds identified in the SLRA tables 4.2.3-1 and 4.2.3-2 (i.e., VLAA 1&2, VLCB 1&2 and VLCB 1&2)

Since the circ weld (i.e., VCBA 2&3) has a sigma initial of 0 degree F - Confirm that unirradiated RTndt is a measured value

• Is yes - identify the CLB document that has been docketed with this information.

If not previously docketed - provide documentation supporting the material properties for the circ weld.

If not, provide the basis that sigma initial is 0 degree F.

3 4.2.2 and 4.2.3 With respect to USE, SLRA Section 4.2.2 states that for the other beltline materials lacking initial USE data, EMA was performed to evaluate the impact of revised fluence projections and available surveillance data on EOL USE reductions.

• SIA Calc 2100300.301P and the the tables in SLRA Section 4.2.2 o Identify a capsule fluence of 9E17.

§ Based on fluence and references in SIA letter - This appears to be the 300-degree capsule Upper Shelf Energy Was the data from the Monticello 120° ISP(E) Surveillance Capsule (see BWRVIP-347) assessed/considered?

If not, what is the impact of the surveillance data from the Monticello 120° ISP(E) Surveillance Capsule on the SLRA USE analysis and the EMAs?

Discuss and provide assessment/evaluation on the impact of the 120-degree capsule on the USE TLAA, including the EMAs, in SLRA -

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions With respect to ART, SLRA Section 4.2.3 does not provide a discussion on surveillance data being considered and /or incorporated when determining ART.

Monticello is a host plant for ISP for LR (i.e., BWRVIP-86, Rev1-A)

• Plate material C2220 appears to be controlling/limiting material.

• RPV material and surveillance material is heat match (based ISP material test matrix)

SIA Calc 2100300.302P - references BWRVIP-135, Rev 4, for assessment of surveillance data - It appears only the Monticello 30-degree and 300-degree capsules were included.

Based on sigma delta of 8.5 degrees F - it appears the from the SLRA reduced margin from credible surveillance data was used.

this includes the credibility evaluation, when considering the 120-degree capsule.

Adjusted Reference Temperature Any reason the SLRA doesnt reflect/assessment of this surveillance data of the 30-degree and 300-degree capsule? Even a comparison of Reg 1.99 position 1.1 versus 2.1?

Was the data from the Monticello 120° ISP(E) Surveillance Capsule (see BWRVIP-347) assessed/considered?

Discuss and provide assessment/evaluation on the impact of the 120-degree capsule on the ART TLAA in SLRA - this includes the credibility evaluation, when considering the 120-degree capsule What, if any, are the downstream effects to the delta RTndt and/or ART that were used for the other embrittlement TLAAs in the SLRA (e.g., SLRA sections 4.2.5 through 4.2.7) when considering the 120-degree capsule 4

4.2.2 XCELMO00017-REPT-091_TLAA_Report_Rev.0

• Bottom of Page 28 of 108 indicates that weld heat 5P6757 was in the last capsule pulled from Monticello in

• Is the weld heat 5P6756??

• If so - it doesnt look like this weld heat

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 2021 and has been incorporated into the USE analysis for the SLRA This weld heat does not appear to be in the ISP material matrix (BWRVIP-86, Rev 1-A and BWRVIP-321-A) is in the Monticello ISP(E) capsules - It appears that this weld heat is at a minimum in the Columbia and River Bend Capsules (see BWRVIP-321-NP)

• Provide discussion of the discrepancy of the weld heat Provide a discussion of how data from weld heat 5P6757 or 5P6756 was incorporated into the USE analysis for the SLRA 5

4.2.2 SLRA Section 4.2.2 states the following:

• Table 4.2.2-1 shows the predicted EOL USE values for MNGP beltline materials having initial USE data, based on the RG 1.99 Position 1 method. For conservatism, the percent drop in USE for the plates are increased by 14.77 percent which is the difference in percent decrease between the measured percent USE decrease, and the RG 1.99 predicted percent USE decrease for the surveillance plate heat C2220.

Based on the description - it appears the drop in USE was ratioed based on surveillance data - which sounds like position 2.2. of RG 1.99, Rev 2.

Provide additional detail/documents/discussion related to the conservatism described in the SLRA Was all available surveillance data incorporated when determining this conservatism?

How does this conservatism compare to the drop in USE based on position 2.2 of RG 1.99, Rev 2?

6 4.2.2 and 4.2.3 Provide additional detail/documents/discussion related to the conservatism described in the SLRA Was all available surveillance data incorporated when determining this conservatism?

How does this conservatism compare to the drop in USE based on position 2.2 of RG 1.99, Rev 2?

Why are the fluences different for 1/4T between the two tables in SIA Calc 2100300.301P?

Why are the fluences different for 1/4T between the SLRA Sections 4.2.1 and the values used in SLRA Sections 4.2.2 and 4.2.3 for USE and ART, respectively?

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Where did the EMA acceptance criteria in this calculation come from? It is different than the criteria for the corresponding EMA tables in the SLRA>

7 4.2.4 SLRA Section 4.2.4 states the following:

• The P-T limit curves will be updated and a Technical Specification change request will be submitted to the NRC prior to exceeding the current 54 EFPY limit.

What is meant/intended by Technical Specification change request given the Monticello has been approved for a PTLR??

Is this just referring to TS 5.6.5.c (i.e.,

PTLR shall be provided to the NRC..)?

8 Please discuss the discrepancy in the fluence values used in 0T ART values for shell Course 1 9

Please discuss the discrepancy in the fluence values used in 1/4T ART values and USE values for all the RPV materials B.2.3.1 ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

Section B.2.3.1 B-38 The SLRA section states that the MNGP ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD (ISI)

AMP is consistent without exception to the 10 elements of NUREG-2191,Section XI.M1, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD. The NRC staff noticed that MNGP did not provide a BWR Water Cleanup AMP, but instead using the ISI AMP to implement the NUREG-2191 BWR Water Cleanup program.

Explain program consistency to NUREG-2191 and explain whether any enhancements to the ISI program is needed.

2 Section B.2.3.1 B-38 As result of a self-assessment conducted in 2019, several issues were identified for program enhancement.

Issue # AFI-1 states, in part, Nine out of 25 AMP owners Provide a brief discussion explaining whether the issue has been resolved or

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions have not completed the job familiarization guide FL-EPE-AMP-001. Refer to CAP-501000032327.

will be resolved prior to the start of SPEO.

3 Section B.2.3.1 B-38 CAP-501000032299 indicates that Current Program Notes (Item 5-2) related to the ISI program could not be located. Enhancement needed.

Provide resolution to the CAP.

4 Section B.2.3.1 Section B.2.3.22 B-38 B-171 CAP 501000032327 states that Effectiveness reviews have not been completed for AMPs contrary to guidance in 4 AWI-08.11.04. It also states, Proactive OE searches have not been performed contrary to 4 AWI-08.11.04, and NEI-1412.

Discuss whether program effectiveness reviews have been completed, and whether proactive OE searches have been performed. In addition, discuss whether MNGP has reviewed IN 2014-02, Failure to Properly Pressure Test Reactor Vessel Flange Leak-off Lines.

Discuss whether MNGP has reviewed IN 2015-04, Fatigue in Branch Connection Welds for site applicability.

5 Section B.2.3.1 B-38 CAP 501000030695 indicates that inspection plan was not submitted within the timeframe required by the license condition, related to its License Renewal Commitment #3.

Discuss whether there is program-tracking of license renewal commitments at MNGP. Discuss if there are programs and procedures in place to ensure license renewal commitments are completed or revised as necessary.

4.2.5 RPV Circumferential Weld Examination Relief 4.2.6 RPV Axial Weld Failure Probability Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

4.2.5 and 4.2.6

• RTmax in BWRVIP-329-A is defined as RT(unirradiated)

+ Shift.

• SIA Calc 2100300.401P - in the assessment for the plate, circ and axial welds - the EOI RTmax appears to include the margin value from RG 1.99, Rev 2, as well.

• SLRA Section 4.2.5 and 4.2.6 states, in part:

SIA Calc 2100300.401P and SLRA appear to interchangeably use the terms RTmax and ART even though are defined different.

Was it intentional? If so, any reason the

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions

• The limiting maximum reference temperatures (RTMAX) for the RPV surface (0T) and 72 EFPY was calculated using plant-specific material chemistry (copper content, nickel content, chemistry factor, and RTNDT(U) (referred to as initial RTNDT)) and neutron fluence for the MNGP RPV plates and welds.

• Using plant-specific data for the RPV dimensions and limiting ARTs for the RPV plates and welds, the evaluation shows that the MNGP RPV meets the applicability criteria of BWRVIP-329-A. As such, on the technical basis of BWRVIP-329-A and as stated in the BWRVIP-329-A SER, MNGP is justified for acceptable embrittlement of RPV axial welds for up to 80 years of plant operation.

ART is defined as RT(unirradiated) + Shift + margin calc/SLRA doesnt acknowledge or point this out given that RTmax and ART and defined differently?

2 4.2.5 and A.3.2.5 SLRA Sections 4.2.5 and A.3.2.5 both state the following:

As such, on the technical basis of BWRVIP-329-A and as stated in the BWRVIP-329-A SER, MNGP is justified for request for alternative pursuant to 10 CFR 40.40(a)(z)(1) from the ASME Code,Section XI examinations for RPV circumferential weld for up to 80 years of plant operation.

The same reference was made in SIA Calc 2100300.401P Confirm reference to 10 CFR 40.40(a)(z)(1) is an error in these documents.

4.2.7 Reflood Thermal Shock Analysis of the RPV Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

4.2.7 4.2-26 SLRA Section 4.2.7 states the following:

• The critical location for the fracture mechanics analysis is at 1/4T. The peak stress intensity factor, K, at 1/4T has a value of approximately 100 ksiin. A maximum KI of 105 ksiin was utilized per Section XI IWB-3612. The acceptability of this K on a plant-specific basis for MNGP Provide discussion of discrepancy in temperature used for calculation

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions can be determined by considering a revised allowable fracture toughness applicable to the MNGP vessel for 72 EFPY. Based on a 0T ART of 197.8°F, the fracture 51oughenss KIC of 174.4°F is above the upper shelf value of 200 ksiin.

• SIA calc 2100300.303 has a similar statement to the SLRA - however a different temperature is used -

197.6°F vs 174.4°F - See page 8 of 15 of SIA calc 2100300.303 2

4.2.7 4.2-25 SLRA Section 4.2.7 states:

A maximum KI of 105 ksiin was utilized per Section XI IWB-3612.

The maximum Kiapplied in the vessel at any time during the transient is 105 ksiin, according to Reference 4.7.18.

SLRA Reference 4.7.18 - Ranganath, S., Fracture Mechanics Evaluation of a Boiling Water Reactor Vessel Following a Postulated Loss of Coolant Accident, Fifth International Conference on Structural Mechanics in Reactor Technology, Berlin, Germany, August 1979.

SIA calc 2100300.303 similarly indicates that a maximum KI of 105 ksiin was use based on the Ranganath analysis.

Clarify the basis for the statement in the SLRA related to A maximum KI of 105 ksiin was utilized per Section XI IWB-3612.

3 4.2.7 4.2-25 SLRA Section 4.2.7 states:

The maximum Kiapplied in the vessel at any time during the transient is 105 ksiin, according to Reference 4.7.18.

The LRA SER for MNGP (NUREG-1865) cites a value of 103 ksiin. The present analysis utilizes the 105 ksiin value, which is more conservative than the 103 ksiin value.

Provide discussion on why 105 ksiin was chosen as the maximum Kiapplied in this SLRA when 103 ksiin was used in the LRA?

4.2.8 Reflood Thermal Shock Analysis of the RPV Core Shroud

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

4.2.8 4.2-10, 4.2-28 SLRA Section 4.2.1 has table 4.2.1.2-1 (fluence values for the core shroud)

SLRA Section 4.2.8 indicates that the max fluence for the core shroud is the following:

• The fluence for the most irradiated point on the core shroud was calculated to be 5.68 x 1021 n/cm2 (E >1 MeV) for 80 years.

SIA Calc 2100300.402 - Page 3 of 4 - indicates that the most irradiated point of the core shroud is 3.68 x 1021 n/cm2 (E >1 MeV)

Is the discrepancy between fluences given in 4.2.1.2-1 and the quoted section in 4.2.8 because the table is only talking about the welds?

Where does the value of - core shroud was calculated to be 5.68 x 1021 n/cm2 (E >1 MeV) for 80 years come from?

Clarify the discrepancy between the SLRA and the supporting SIA calc?

4.2.9 Loss of Preload for Core Plate Rim Holddown Bolts Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

4.2.9 SLRA Section 4.2.9 states the following:

• An assessment was performed to confirm that the MNGP Core Plate Bolts can have inspections waived and have their age-related degradation managed for the SPEO for 72 EFPY. This evaluation concluded that the criteria of Appendix I of BWRVIP-25, Revision 1-A to justify the elimination of core plate bolt inspections at MNGP are satisfied. Therefore, elimination of core plate bolt inspections at MNGP for the SPEO is justified.

SIA Calc - 2100300.403P - Stress Relaxation of Core Plate Rim Holddown Bolts Analysis - has the detailed evaluation.

Step 3 of BWRVIP25, Rev 1-A

• Previously requested - references from BWRVIP-25 Rev 1-A o GE Hitachi Safety Communication SC 11-05, "Failure to Include Seismic Input in Channel Control Blade Interference Customer Guidance." September 2011.

o "Failure to Include Seismic Input in Channel-Control Blade Interference Customer Guidance," GE Hitachi Nuclear Energy SC 11-05, Rev.2, December 16, 2013.

• Provide discussion on how MGNP met

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions

• Additional detail is needed for review related to certain steps from Appendix I to BWRVIP-25, Rev 1-A conditions and abide by the recommendations of SC 11-05, including revisions.

Provide plant-specific documentation that MGNP met conditions and abide by the recommendations of SC 11-05, including revisions.

4.2.10 Susceptibility to IASCC Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

4.2.10 SLRA Section 4.2.10 states the following:

• The evaluation in the MNGP LRA identified the top guide, core shroud, and incore instrumentation dry tubes and guide tubes as being susceptible to IASCC for 60 years of operation and concluded that aging management is required through the first PEO.

Additionally, the neutron fluence values from SLRA Table 4.2.1.2.1-1 shows that the in-core instrument tubes AND guides exceed the 5E20 fluence threshold for IASCC that is discussed in SLRA section 4.2.10.

SLRA Section 4.2.10 - dispositions TLAA Disposition: 10 CFR 54.21(c)(1)(iii) and states:

• Aging effects of IASCC and embrittlement on the top guide, core shroud, and jet assembly components will be managed by the BWR Vessel Internals (B.2.3.7) AMP through the SPEO in accordance with 10 CFR 54.21(c)(1)(iii).

Note the discrepancy in the components that were identified in the TLAA in the 60-year LRA and the components in the disposition for 80-year SLRA.

If it was a TLAA for 60 years - It would be expected to still be a TLAA for 80 years, unless there was a revision to the CLB to remove component from the TLAA.

Was there a revision to the CLB to remove the incore instrumentation dry tubes and guide tubes in the TLAA from initial license renewal?

What is the basis that the incore instrumentation dry tubes and guide tubes are not in the TLAA for the SLRA?

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions I understand there is a discussion in BWRVIP-315 related to the incore instrumentation dry tubes and guide tubes -

but BWRVIP-315 is an unapproved topical report.

UFSAR Section K3.5 Irradiation Assisted Stress Corrosion Cracking (Revision 31) indicates that there is a TLAA for the incore instrumentation dry tubes and guide tubes that is incorporated into the CLB.

2 4.2.10 SLRA Table 4.2.10-1 identifies the Core Support Plate with a fluence of 1.17E21, which is greater than the screening threshold of 5E20 for IASCC identified in SLRA Section 4.2.10.

SLRA Section 4.2.10 does not appear to address this component in the TLAA Evaluation section or in the TLAA Disposition.

SLRA Section 4.2.10 - dispositions TLAA Disposition: 10 CFR 54.21(c)(1)(iii) and states:

• Aging effects of IASCC and embrittlement on the top guide, core shroud, and jet assembly components will be managed by the BWR Vessel Internals (B.2.3.7) AMP through the SPEO in accordance with 10 CFR 54.21(c)(1)(iii).

The SLRA appears to rely on BWRVIP-315 for identifying those components susceptible to IASCC. BWRVIP-315 is not an approved topical report.

Discuss and reconcile that the neutron fluence for the Core Support Plate exceeds the screening criterion for IASCC but does not appear to be included in the TLAA ?

B.2.3.38 Electrical Insulation for Inaccessible Medium-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

OpE OpE Can you explain how the condition of cables (specifically XI.E3) will be

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions monitored and if there are any trigger points (e.g., cables exposed to environmental conditions that may accelerate aging) for increasing the periodicity of monitoring/testing during the extended period of operation.

2 OpE OpE Explain the water trend fluctuation OpE for many ARs that had water entering underground vaults. Discuss any impact this has had on your underground vaulting inspection process in support of your request to renew the subsequent license of the plant?

3 OpE OpE Is there a procedure in place to deal with manholes when there is significant water intrusion from either heavy rain or rapid snow melt?

4 OpE OpE To support subsequent renewal license.

What is the process for the identification of aging of active components (i.e, relays) where surveillance and testing are not sufficient to reveal wear of components.

How often are these components changed?

5 OpE OpE Can you explain the cause for all the cable replacements found in ARs (5000001348566, 5000001352244, 5000001352311, 5000001356079)?

6 OpE OpE It was noted that there was no cable condition monitoring program coordinator/owner for some time. How is the cable aging management program being managed?

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 7

OpE OpE What contributed to the RED status for the cable monitoring program during the 2014-2015 time-frame?

8 the applicant inadvertently left out the word potentially in front of exposed to significant moisture in the scope of program program description in Section B.2.3.38. As the applicant stated that they plan to implement the program consistent with the GALL-SLR Report, as modified by SLR-ISG-2021 ELECTRICAL, please confirm whether this was an unintentional error.

2.5 - Scoping and Screening Results: Electrical And Instrumentation & Controls Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

2.5 2.1.2 Page 2.5-2 SLRA section 2.5.1.3, Elimination of Electrical and I&C Commodity Groups Not Applicable to MNGP, stated that uninsulated ground conductors are not subject to AMR because:

Uninsulated ground cables are not classified as SR nor are they relied upon for SR equipment to perform their intended function as identified in 10 CFR 54.4. Failure of an uninsulated ground conductor will not prevent the satisfactory accomplishment of any functions identified in 10 CFR 54.4(a)(1). Uninsulated ground cables are not relied upon in safety analyses or plant calculations to perform a function related to any regulated events identified by 10 CFR 54.4(a)(3).

SLRA Figure 2.5-1, MNGP Simplified One-Line Diagram (For SBO Offsite Power Recovery), shows underground medium voltage cables between the 345-kV offsite power Clarify if the above-mentioned underground cables in the 345-kV,13.8-kV, and 4.16-kV power systems are used for power restoration during an SBO event. If they are, clarify why they are not included in the components and commodities in-scope of SLR and they are not subject to AMR.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions system and the safety-related 4.16-kV buses#15 and #16 and between the 13.8-kV offsite power systems and the safety-related 4.16-kV buses #15 and 16. These underground cables are described in section 8.2, Transmission System of the MNGP Updated Safety Analysis Report (USAR)

SLRA section 2.1.2.4.5, Station Blackout (10 CFR 50.63), states:

Offsite sources identified for power restoration, and therefore in-scope for SLR, include the 345 kV, 115 kV, and 13.8 kV offsite sources. Components and commodities in-scope for SLR are those from the plant 13.8 kV and 4.16 kV busses, through and including the interconnecting transformers, disconnect switches, and busses out to and including the switchyard circuit breakers that connect to these offsite sources.

NUREG 2192, section 2.5.2.1.1, Components Within the Scope of SBO (10 CFR 50.63), discusses the offsite and onsite power systems that are relied upon to meet the requirements of the SBO Rule. NUREG 2192, section 2.5.2.1.1, stated, in part, that the offsite power system includes:

The plant system portion of the offsite power system that is used to connect the plant to the offsite power source meeting the requirements under 10 CFR 54.4(a)(3).

This path typically includes the circuit breakers that connect to the offsite system power transformers (startup transformers), the transformers themselves, the intervening overhead or underground circuits between circuit breaker and transformer and transformer and onsite electrical distribution system, and the associated control circuits and structures.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The staff noted that the above-mentioned underground cables in the 345-kV and 13.8-kV offsite power systems and the safety-related 4.16-kV onsite power systems, as shown in the SLRA Figure 2.5-1, are part of the offsite power recovery path during an SBO event. In addition, these above-mentioned underground cables are not in the listed in SLRA section 2.1.2.4.5, as components and commodities in-scope of SLR.

2 2.5 2.5-6 SLRA Table 2.5-1, Electrical and I&C Component Commodity Groups Installed at MNGP for In-Scope Systems, includes elements, resistance temperature detectors, sensors, thermocouples, transducers, and electric heaters commodity groups.

NUREG 2192, Table 2.1-6 Typical Structures, Components, and Commodity Groups, and 10 CFR 54.21(a)(1)(i) Determinations for Integrated Plant Assessment identified elements, resistance temperature detectors, sensors, thermocouples, transducers, and electric heaters as commodity groups that will meet the passive component screening criterion 10 CFR 54.21(a)(1)(i) if they have a pressure boundary function.

The staff did a keyword search for the above-mentioned components and could not find any information about the screening of these components.

Clarify if the elements, resistance temperature detectors, sensors, thermocouples, transducers, and electric heaters commodity groups have a pressure boundary function at MNGP, and if the pressure boundary function for these commodities is addressed in the mechanical review.

3 2.5 2.5-6 2.5-1 SLRA Table 2.5-1 included cable bus as a commodity group at MNGP.

SLRA section 2.5.1.2, Application of Screening Criteria 10 CFR 54.21(a)(1)(i) to Electrical and I&C Commodity Group, listed cable bus as a passive commodity meeting the screening criteria of 10 CFR 54.21(a)(1)(i). SLRA section 2.5.1.3 stated that cable bus is not utilized at MNGP.

Clarify why cable bus is included in the commodity group and passive commodity group if cable bus is not used at MNGP.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The staff noted that if cable bus is not used at MNGP, it should not be included in the list of passive commodities.

4 3.6.2 3.0 3.6-13 3.6-32 3.6-36 3-11 SLRA Table 3.6.2-1, Electrical Commodities - Summary of Aging Management Evaluation, Note 1 stated that the insulation material of the MNGP fuse holders (not in active components) insulation material has no aging effects requiring management and referenced SLRA Section 3.6.2.3.1 for additional information.

SLRA section 3.6.2.3.1, Fuse Holders, states: MNGP fuse holders (not part of active equipment): insulation material that may be subject to an [adverse localized environment] ALE that may affect insulation resistance are addressed as part of Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements. Fuse holder insulation material that is not subject to an adverse environment does not have aging effects requiring management.

SLRA Table 3.0-3, Electrical Service Environments, stated that an ALE can be due to the ALEs can caused by (1) exposure to significant moisture, or (2) exposure to heat, radiation, or moisture.

NUREG-2192 Table 3.6-1, Summary of Aging Management Programs for the Electrical Components Evaluated in Chapter VI of the GALL-SLR Report, provided the following aging effects/mechanisms on fuse holders insulation: Reduced electrical insulation resistance due to thermal/thermoxidative degradation of organics, radiolysis, and photolysis (UV sensitive materials only) of organics; radiation-induced oxidation; moisture intrusion.

The staff noted that the evaluation in SLRA section Discuss the aging management reviews results for the insulation material of fuse holders (not part of active equipment) to demonstrate that these fuse holders are in an environment that does not subject them to environmental aging mechanisms identified in NUREG-2192 for fuse holders insulation material.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 3.6.2.3.1 appear not to address aging effects on the insulation materials of the MNGP in-scope fuse holders (not part of active equipment).

Add 1 Since the applicant already determined that there are no aging effects to be managed for fuse holders insulation material, as stated in table 3.6-1 Item 3.6.1-022 and table 3.6.2-1 plant-specific Note 1, provide the evaluation that demonstrate this determination in the SLRA.

Add 2 Table 3.6.2-1 SLRA Table 3.6.2-1, Plant-Specific Note 1 states:

1. In alignment with GALL-SLR, no AMP is required when fuse holders are located in an environment that does not subject them to environmental aging mechanisms. Fuse holder insulation material in an ALE is managed via the XI.E1 AMP. MNGP fuse holders (not in active components) insulation material and environment combination has no aging effects requiring management.

See SLRA Section 3.6.2.3.1 for additional information.

Section 3.6.2.3.1, does the statement MNGP fuse holders (not in active components) insulation material and environment combination has no aging effects requiring management mean that the fuse holders insulation material are not subject to an ALE? If so, please explain how the determination was made that the insulation material are not subject to an ALE in the SLRA.

Add 3 Table 3.6-1 listed two Items 3.6.1-022 and 3.6.1-008 for fuse holders insulation material. Are both items applicable to the same insulation material? If yes, explain the difference in the evaluations provided in the discussion column. If not, clarify in the SLRA which one is applicable 2.4.6 Fire Protection Barriers Commodity Group Cracking Due to Stress Corrosion Cracking and Loss of Material Due to Pitting and Crevice Corrosion Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

2.4.6 2.4-15 SLRA Section 2.4.6 states, Curbs, dikes, concrete components other than barriers are evaluated as part of Please discuss whether there are any curbs or dikes that have a fire barrier

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions the structure where they are located. SLRA Tables 2.4-6 and 3.5.2-6 do not include curbs or dikes with a fire barrier intended function. In addition, the staff did not identify AMR items explicitly for curbs or dikes, regardless of intended function, in the SLRA.

intended function, and if so, where are they addressed in the SLRA. In addition, discuss where curbs and dikes with other intended functions are addressed in the SLRA.

2 N/A N/A Section 4.1 of Report No. XCELMO00017-REPT-065 (Fire Protection program basis document) includes a note that states, Note - some clarification of Fire Protection AMP documents may be warranted during implementation to better distinguish fire damper assemblies (housings and any portion that serves a fire barrier intended function in the closed position) from other fire barriers. In addition, Table 1 in Section 7.0 of the basis document states, Clarify that fire damper includes the housing and all parts that perform a fire barrier function in the closed position.

It is unclear to the staff if the underlined statements are indicating only the damper housing are subject to aging management review or there may be additional parts that are subject to aging management. The staff notes that SLRA Tables 2.4-6 and 3.5.2-6 include fire damper housings, however, they do not appear to include additional fire damper parts.

Please discuss the intent of the underlined statements.

3 N/A N/A GALL-SLR AMP XI.M26 recommends that fire damper assemblies be inspected for signs of corrosion and cracking at a frequency in accordance with an NRC-approved fire protection program. The staff recognizes that SLRA Table A-3 and SLRA Section B.2.3.15 include an enhancement to the procedures to inspect fire damper assemblies for corrosion and cracking.

Revision 48 of Procedure 0275-02, states that the fire damper test and inspection frequency is in accordance with Section 19.4.1.1 of NFPA 80, 2007 Edition.

Specifically, the test and inspection frequency for fire Please clarify the functional test and visual inspection frequency of fire dampers.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions dampers is every 4 years. The description of the 4-year test and inspection appears to be more functional in nature. For Procedure 0275-02, Revision 3 of PBD/AMP-013 states, A visual inspection is conducted every 24 months to verify the integrity and functionality of plant fire barrier floors, walls, structural steel coating and dampers that separate redundant trains of safe shutdown systems. However, Revision 48 of Procedure 0275-02 appears to only state perform visual inspection every 24 months of penetration fire barriers. Therefore, the functional test and visual inspection frequency of fire dampers is unclear.

4 2.4.6 2.4-15 SLRA Section 2.4.6 states, The portions of the Fire Protection Barriers Commodity Group include cable tray covers, FP guard pipe, fire damper housing, fire stop sealants (silicone, silicone foam, caulk), and cementitious (Pyrocrete walls, etc.), thermal fiber (silicates), and rigid board (gypsum walls, etc.) fireproofing. SLRA Section B.2.3.15 includes rigid board (gypsum walls, etc.) as a fire protection component material. It is unclear whether SLRA Section B.2.3.15 includes FP guard pipe. SLRA Tables 2.4-6 and 3.5.2-6 do not appear to include component types of FP guard pipe or rigid board (gypsum walls, etc.). It is unclear where these component types are addressed in the SLRA.

SLRA Tables 2.4-6 and 3.5.2-6 include component types fireproofing and non-metallic fireproofing. SLRA Table 3.5.2-6 indicates that the fireproofing and non-metallic fireproofing material is cementitious. AMR item 3.3.1-268 is cited for both. However, it is unclear what is the difference between cementitious fireproofing and cementitious non-metallic fireproofing.

SLRA Table 3.5.2-6 includes two cementitious fireproofing component types: both citing AMR item 3.3.1-Please address the following:

1. Where are FP guard pipe and rigid board (gypsum walls, etc.) addressed in the SLRA?

2. What is the difference between cementitious fireproofing and cementitious non-metallic fireproofing in SLRA Tables 2.4-6 and 3.5.2-6?

3. What is the difference between the two cementitious fireproofing components in SLRA Table 3.5.2-6?

4. What is the difference between the two silicate thermal fiber components in SLRA Table 3.5.2-6?

5. Should aluminum be added to SLRA Section B.2.3.15 as a fire protection component material?

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 268 but citing different aging effects. It is unclear what is the difference between the two cementitious fireproofing components.

SLRA Table 3.5.2-6 includes two silicate thermal fiber component types: both citing AMR Item 3.3.1-269 but citing different aging effects. It is unclear what is the difference between the two silicate thermal fiber components.

SLRA Section 3.5.2.1.6 and SLRA Table 3.5.2-6 include aluminum as a fire barrier component material, however, SLRA Section B.2.3.15 does not include aluminum as a fire protection component material.

5 3.3, 3.5 3.3-90, 3.5-7, 3.5-96 SLR-ISG-2021-02-Mechanical, Updated Aging Management Criteria for Mechanical Portions of Subsequent License Renewal Guidance (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML20181A434) added GALL-SLR Items 3.3-1, 268 and 269. The aging effects for cementitious coatings and silicates used as fireproofing/fire barriers exposed to air are loss of material, change in material properties, cracking/delamination, and separation. These aging effects are consistent with Section 6, Fire Barriers, of EPRI 3002013084, Long-Term Operations:

Subsequent License Renewal Aging Affects for Structures and Structural Components (Structural Tools),

November 2018.

The discussion of AMR item 3.3.1-268 in SLRA Table 3.3-1 states, Consistent with NUREG-2191. The Fire Protection (B.2.3.15) AMP is used to manage loss of material, change in material properties, cracking, delamination, and separation for cementitious coating fireproofing/fire barriers/HELB barriers exposed to air indoor uncontrolled.

Please discuss whether the statement in the discussion of AMR item 3.3.1-268 in SLRA Table 3.3-1 regarding HELB Barriers is correct.

In addition, given that EPRI 3002013084 cites loss of material, change in material properties, cracking/delamination, and separation as aging effects for cementitious coatings and silicates used as fireproofing/fire barriers, please discuss why not all of these aging affects are cited for each cementitious fireproofing and silicate thermal fiber component.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA Table 3.5.2-6 does not cite HELB Barrier as an intended function for any components. In addition, AMR item 3.3.1-268 was not cited for components with a HELB barrier intended function in other SLRA AMR tables.

(Breakout Question 7 addresses doors with a HELB Barrier intended function).

SLRA Table 3.5.2-6 cites AMR item 3.3.1-268 for cementitious fireproofing and cementitious non-metallic fireproofing. One line item cites only cracking, change in material properties, and delamination. The other line item cites only loss of material. Separation does not appear to have been cited for cementitious fireproofing and cementitious non-metallic fireproofing.

The discussion of AMR item 3.3.1-269 in SLRA Table 3.3-1 states, Consistent with NUREG-2191. The Fire Protection (B.2.3.15) AMP is used to manage loss of material and change in material properties of thermal fiber exposed to air indoor uncontrolled.

SLRA Table 3.5.2-6 cites AMR item 3.3.1-269 for silicate thermal fiber. One line item cites change in material properties and the other line item cites loss of material.

Cracking, delamination, and separation do not appear to have been cited for silicate thermal fiber.

SLRA Section 3.5.2.1.6 does not appear to include delamination and separation as aging effects.

The staff also notes that the Fire Protection program basis document in Section 3.3 states, MNGP fireproofing are installed in air environments and the aging effects that require management include loss of material,

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions cracking, and changes in material properties. This statement does not appear to be consistent with SLR-ISG-2021-02-Mechanical or the SLRA.

6 3.5.2.2.2.4 3.5-34 SLRA Section 3.5.2.2.2.4, states, Also, the air environment (and underground environment in manholes) for stainless steel supports or anchorage is not expected to be aggressive enough to cause cracking or localized loss of material for components (stainless steel new fuel storage racks, refueling cavity liner, component supports, anchorages, fire barrier penetration seals, insulation jacketing inside containment, aluminum insulation jacketing outside containment, and aluminum manway covers) exposed to indoor, outdoor, or underground air in the presence of wetting.

SLRA Section 3.5.2.1.6 does not include stainless steel as a material for the Fire Protection Barriers Commodity Group. In addition, SLRA Table 3.5.2-6 only includes elastomer fire barrier penetration seals.

Please discuss whether stainless steel should be added to SLRA Section 3.5.2.1.6 and whether stainless steel fire barrier penetration seals should be added to SLRA Table 3.5.2-6.

7 N/A N/A Work Order 00546903 states that Door-3 (TRB BLDG 931 to Heating Boiler Room) provides fire protection, external flooding, security, and appendix R for safe shutdown considerations.

Fire rated doors with only a fire barrier intended function are managed by the Fire Protection program per SLRA Table 3.5.2-6. In the SLRA, it appears that doors with intended functions other than a fire barrier intended function are evaluated as part of the individual structures where they are located, and it appears the Structures Monitoring program is cited to manage aging effects. For example, SLRA Tables 2.4-17 and 3.5.2-17 include doors with a flood barrier and a HELB Barrier intended functions managed by the Structures Monitoring program.

Please discuss whether Door 27 in Procedure 0275-03 is a fire door. Please discuss why Door 1 is inspected under Procedure 0275-03 if it is not a fire rated door.

Please discuss whether fire rated doors in SLRA Table 3.5.2-6 should include flood barrier and HELB barrier intended functions given procedures under the Fire Protection program appear to address doors with these functions.

However, if the intent was for the Fire Protection program to manage aging effects to ensure the fire barrier intended

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Procedure 1216-01, Fire Door Inspections, verifies functionality/operability of Appendix R Fire Doors and HELB Doors. In addition, Procedure 0275-03 covers HELB doors and flooding requirements for Doors 1 and

27. This procedure does state that Door 1 is not a fire door, however, a similar statement is not made about Door 27. It also states to contact the Flooding Engineer if steps to meeting flooding requirements are not met.

Neither Procedure 1216-01 nor Procedure 0275-03 refer to the Structures Monitoring program. In addition, the Fire Protection and Structures Monitoring programs basis documents do not state these programs credit or are credited by one another.

The staff noted a number of procedures under the Structures Monitoring program that addresses doors (e.g., Procedure 1385), however, they did not appear to address fire rated doors or reference the Fire Protection program.

It is unclear to the staff what intended functions that fire rated doors have and it is unclear what programs manage aging effects to ensure the intended functions are maintained.

function is maintained and other program(s) (e.g., Structures Monitoring program) to manage aging effects to ensure intended functions other than the fire barrier intended function (e.g., flood barrier and HELB barrier) are maintained, then discuss how program procedures will be updated to adequately reflect this. (Breakout Question 8 is similar question) 8 N/A N/A Figure 5.2 of EWI-11.01.07 that identifies system intended functions includes fire barrier intended functions for certain systems, for example, Intake Structure. SLRA Table 3.5.2-9 does not cite a fire barrier intended function for any components. The staff assumes component types may be repeated in SLRA Table 3.5.2-9, for example, and SLRA Table 3.5.2-6. The component types in SLRA Table 3.5.2-9 cites intended functions other than fire barrier, and SLRA Table 3.5.2-6 cites the fire barrier intended function for the same component types. Is that correct?

Please discuss whether the staffs assumption that the component types in SLRA Table 3.5.2-9 cites intended functions other than fire barrier, and SLRA Table 3.5.2-6 cites the fire barrier intended function for the same component types and that this assumption applies to other systems identified in Figure 5.2 of EWI-11.01.07 that include a fire barrier intended function.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Figure 5.3 of EWI-11.01.07 that identifies aging effects requiring management for specific material/environment groups includes cementitious fireproofing and rigid board (thermal insulating board).

Cementitious fireproofing and non-metallic fireproofing is managed by the Fire Protection program per SLRA Table 3.5.2-6. Rigid board (thermal insulating board) is not included in SLRA Table 3.5.2-6 (see Breakout Question 4 related to rigid board (gypsum walls, etc.)), however, rigid board (gypsum walls, etc.) is discussed in SLRA Sections 2.4.6 and B.2.3.15.

Please discuss why cementitious fireproofing and rigid board (thermal insulating board) is addressed in EWI-11.01.07.

9 N/A N/A Section 3.3 of the Fire Protection program basis document states, The MNGP Plant Structures and Commodities AMR does not specify that the installed fireproofing (indoor or outdoor) require shielding to ensure that the intended function is maintained.

The phrase does not specify could mean shielding is required but it just wasnt documented in the AMR.

SLRA Table 3.5.2-6 only cites an outdoor air environment for fire rated doors and structural fire barriers (walls, ceilings, and floors).

Please confirm the statement in the Fire Protection program means there are no installed fireproofing that requires shielding. In addition, please clarify whether there is fireproofing installed outdoors.

10 N/A N/A The Fire Protection program basis document states that the exception taken during initial license renewal related to the testing frequency of the Cable Spreading Room halon system is no longer needed since GALL-SLR AMP XI.M26 no longer prescribes a frequency.

Revision 28 of Procedure 0328 states per Ops Manual B.08.05-05 that the system functional test is every 18 months, visual inspection of headers and nozzles is every 18 months, and an air flow test through headers and nozzles is every 3 years (these frequencies are also Since the exception is no longer needed, please confirm that the inspection frequencies stated in Revision 28 of Procedure 0328 and Ops Manual B.08.05-05 will continue for SPEO.

Given that the frequency for visual inspection only refers to headers and nozzles, and Step 1 states all accessible external surfaces (consistent with GALL-SLR AMP XI.M26) are visually examined,

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions discussed in Section 3.3 of the Fire Protection program basis document). The Ops Manual also includes monthly valve position checks and halon tank weight and pressure checks every six months.

Step 1 of Procedure 0328 requires visually examination of all accessible external surfaces, including supporting, restraining, and contact surfaces of the system for any signs of degradation.

please confirm that all accessible external surfaces, including headers and nozzles, are visually examined every 18 months.

11 N/A N/A Section 4.6 of the Fire Protection program basis document states that the acceptance criteria for the Cable Spreading Room Halon System are in Table A.2-2, items F.1.e and F.1.d in Procedure 0328. However, Procedure 0328 does not appear to have a Table A.2-2.

Please provide the correct reference for the acceptance criteria associated with the Cable Spreading Room Halon System.

12 N/A N/A The staff noted that Revision 28 of Procedure 0275-01, Revision 3 of EWI-11.01.09, and the Fire Protection program basis document addresses that the 20 percent of the total number of penetration seals includes 10 percent of each type of penetration seal (consistent with GALL-SLR AMP XI.M26). However, Ops Manual B.08.05-05 only address the 20 percent of the total number of penetration seals; it is silent on the 10 percent of each type of penetration seal.

There appears to be three procedures related to seals:

0275-01, 0275-04, and 0275-05. Neither of these documents reference each other. In addition, Procedures 0275-04 and 0275-05 does not state the scope (e.g., all, 20 percent, etc.) or frequency (e.g., 24 months).

Please discuss why the Ops Manual does not state that the 20 percent of the total number of penetration seals includes 10 percent of each type of penetration seal.

Please provide the scope and frequency for Procedures 0275-04 and 0275-05 and how all three procedures related to penetration seals work together with regards to meeting 20 percent of the total number of penetration seals, including 10 percent of each type of penetration seal.

13 N/A N/A Figure 5.3 of EWI-11.01.09 include material reinforced concrete, grout.

Revision 1 of FIREPROTECT states, Grout is considered to be a part of the material constituting the barrier in which it is installed. Therefore, piping, conduit, etc. that penetrates a barrier and sealed with grout is Given that it appears grout is a fire protection material, please discuss revising the SLRA to address grout as part of the Fire Protection program.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions considered part of the barrier, but not a penetration seal.

The fire barrier inspection procedure (0275-02) requires inspection of the entire surface of each fire barrier. The grout filled annular gap around these penetrants is included within inspection under 0275-02. Procedure 0275-01 and Revision 23 of AWI-08.01.00 make the same statements, however, the cited Procedure 0275-02 does not make similar clarifying statements. With regards to grout, Procedure 0275-02 only states, Any pipe sleeves that are grouted in but the inside of the sleeve contains no seal.

In addition, the SLRA and the Fire Protection program basis document do not appear to address grout, or it being inspected as part of the fire barrier.

14 N/A N/A The Fire Protection program basis document states that the initial license renewal requirement to manage periodic testing of the diesel-driven pump and inspection of the diesel engine to ensure the fuel supply line can perform its intended function will be managed by other programs (Fire Water System, Fuel Oil Chemistry, and Selective Leaching) during SPEO.

The staff noted the following procedures related to the diesel-driven pump:

-Revision 31 of 1158-B

-Revision 59 of 0261

-Revision 4 of 0265

-Revision 32 of 4190-PM

-Revision 43 of 0192

-Revision 69 of 0266 None of these procedures are referenced in the Fire Protection program basis document which seems reasonable since these activities were stated to not fall under that program for SPEO.

Please discuss which program will execute Revision 31 of 1158-B and Revision 4 of 0265.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions None of these procedures are referenced in the Selective Leaching program basis document.

Revision 59 of Procedure 0261 and Revision 69 of Procedure 0266 are referenced in the Fire Water System program basis document.

Revision 43 of 0192 is referenced in the Fuel Oil Chemistry program basis document. Revision 32 of 4190-PM is listed on the portal under the Fuel Oil Chemistry program (and the Fire Protection program) but not referenced in the Fuel Oil Chemistry program basis document.

The staff did not identify where Revision 31 of 1158-B and Revision 4 of 0265 are now referenced.

15 Appendix A and Appendix B

A-64 and B-109 The Monitoring and Trending program element in GALL-SLR AMP XI.M26 states, The performance of the halon/CO2 fire suppression system is monitored during the periodic test to detect any degradation in the system.

These periodic tests provide data necessary for trending.

Section 3.3 of the Fire Protection program basis document states that Section 4.5 of the basis document documents that the Cable Spreading Room Halon fire suppression system procedures specify trending of inspection results of this system. However, the staff did not identify where this documentation is made in Section 4.5.

The staff did not identify where Revision 28 of Procedure 0328 discusses trending. However, Table 1 in Section 7.0 of the basis document states that Procedure 0328 will be updated to trend inspection results and where practical, project degradation until the next scheduled inspection.

Please discuss whether the program will be enhanced to trend the inspection results for fire barrier penetration seals, fire barriers, fire damper assemblies, fire doors, and the halon fire suppression system.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The enhancement related to trending inspection results appears inconsistent in the SLRA. SLRA Section B.2.3.15 states, Trend the inspection results for timely detection of aging effect so that appropriate corrective actions can be taken, while SLRA Table A-3 states, Trend the inspection results on fire barrier penetration seals, fire barriers, fire damper assemblies, and fire doors for timely detection of aging effect so that appropriate corrective actions can be taken. SLRA Section B.2.3.15 appears to include trending of all Fire Protection program inspection results, while SLRA Table A-3 appears to exclude the halon fire suppression system.

16 Appendix A and Appendix B

A-65 and B-110 The Corrective Actions program element in GALL-SLR AMP XI.M26 states, During the inspection of penetration seals, if any sign of degradation is detected within that sample, the scope of the inspection is expanded to include additional seals in accordance with the plants approved fire protection program.

The enhancement related to expanding the penetration seal inspection scope if degradation is detected appears inconsistent with GALL-SLR AMP XI.M26. SLRA Section B.2.3.15 and SLRA Table A-3 state, in part, for fire barrier penetration seals, if degradation that could result in loss of fire protection capability is detected within the inspection sample of penetration seals, that the scope of the inspection is expanded to include additional seals in accordance with the MNGPs Fire Protection AMP.

The staff noted that the enhancement related to expanding the penetration seal inspection scope if degradation is detected appears to be inconsistent in the Fire Protection program basis document. For example, Section 6.0 (description of Element 7 enhancement) and Table 1 in Section 7.0 do not include the phrase if Please clarify the enhancement related to expanding the penetration seal inspection scope if degradation is detected appears. For instance, will the scope be expanded if any sign of degradation is detected or is there a process for determining when to expand the penetration seal inspection scope.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions degradation that could result in loss of fire protection capability is detected within the inspection sample of penetration seals.

17 N/A N/A Revision 23 of AWI-08.01.00 and Revision 1 of 0275-02 state to consider cracks greater than 0.25 inches wide in walls, floors, and ceilings. However, these documents did not appear to cite a reference for this crack width limit.

Procedure 0275-03 states that a 3/16-inch limit is used when checking for depressions or bulges in fire doors based on good engineering practice.

Please discuss the basis for cracks greater than 0.25 inches wide and provide, if available, supporting references.

Please discuss the 3/16-inch limit being a good engineering practice and provide, if available, supporting references.

4.6.3 - Condensate Backwash Receiving Tank Fatigue Evaluation Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

4.6.3 4.6-5 SLRA Section 4.6.3 addresses the fatigue time-limited aging analysis (TLAA) for the condensate backwash receiving tank. The section explains that the alternating stresses in the system were examined to determine an allowable number of cycles for the tank of 35,000 airbursts (i.e., backwashing cycles) under normal and accident conditions.

SLRA Section 4.6.3 also indicates that the annular number of airburst cycles is conservatively estimated to be 160 cycles/year. However, the SLRA does not clearly discuss how the applicant estimated the annular number of airburst cycles.

In addition, it is unclear to the staff whether this fatigue analysis needs to consider the transient cycles in the upset and test conditions for the condensate backwash receiving tank.

1. Describe how the applicant estimated the annular number of airburst cycles (160 cycles/year).

2. Clarify whether the fatigue analysis needs to consider the transient cycles in the upset and test conditions for the condensate backwash receiving tank. If not, explain why the transients in the upset and test conditions do not need to be considered in the fatigue analysis (e.g., the transients evaluated in the analysis are bounding for the upset and test transients, or non-applicability of upset and test transients to the fatigue analysis).

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions B.2.3.30 - ASME Section XI, Subsection IWF Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.3.30; XCELMO00017-REPT-077, Revision 1 (PBD), Section 3.1 B-224; PBD p4 Section 3.1 of PBD states, in part: The MNGP ASME Section XI, Subsection IWF AMP [Ref. 9.3] that has the following principal objective:

• Manage the effects of loss of material for Class 1, 2, 3 and MC component supports.

The above PBD statement appears inconsistent with the SLRA B.2.3.30 AMP as well as the GALL-SLR AMP XI.S3 which manage the effects of all applicable aging effects (which include several aging effects including loss of material) for Class 1, 2, 3, and MC component supports.

a) Explain the apparent discrepancy between the referenced statement in the PBD and the SLRA B.2.3.30 and GALL-SLR XI.S3 AMPs; OR provide the basis of how the SLRA AMP PBD is in its objective consistent with the GALL-SLR Report AMP XI.S3, as claimed in the SLRA, and meeting the requirements of 10 CFR 54.21(a)(3).

2 B.2.3.30; Table A-3 item 33 B-225; A-87 Consistency of scope of program element:

SLRA AMP includes an enhancement (commitment 33(a)) to the scope of program element related to evaluation of inaccessible areas, which appears to only address inaccessible areas of Class 1, 2 and 3 supports.

However, the scope of the SLRA also includes Class MC components, which appears to be not included in the enhancement. Further, the enhancement language also does not appear to state inaccessible areas of what components (e.g., component supports).

a) Explain the apparent discrepancy between the referenced statement in the PBD and the SLRA B.2.3.30 and GALL-SLR XI.S3 AMPs; OR provide the basis of how the SLRA AMP PBD is in its objective consistent with the GALL-SLR Report AMP XI.S3, as claimed in the SLRA, and meeting the requirements of 10 CFR 54.21(a)(3).

b) Explain why Class MC component supports are not included in the enhancement/commitment.

c) Provide a revised enhancement/commitment that addresses the concerns raised; OR, provide an exception to the scope of program element with the basis of why the AMP with exception remains adequate to manage applicable aging effects.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 3

B.2.3.30, Table A-3 item 33 B-226, A-88 Consistency of Detection of Aging Effects element:

The enhancement and corresponding Commitment 33(g) to detection of aging effect element appears to include tactile inspection of only elastomeric vibration isolation elements, whereas the corresponding GALL-SLR element includes elastomeric or polymeric vibration isolation elements. Also, a related enhancement to parameters monitored or inspected element includes polymeric vibration isolation elements a) Clarify why polymeric vibration isolation elements are not included in the referenced enhancement, and provide a revised enhancement if necessary.

4 B.2.3.30, Table A-3 item 33 B-226, A-89 Consistency of monitoring and trending element:

The monitoring and trending element includes an enhancement and corresponding commitment 33(i),

which states: Revise procedures to increase or modify inspection population when a component support is repaired to as-new condition by..

(emphasis added).

To be consistent with the GALL-SLR AMP XI.S3, the highlighted word population in the enhancement should a) Clarify if the highlighted word population in the enhancement is intended to be sample and accordingly provide a revised enhancement/commitment; OR justify why the enhancement need not be revised.

5 Table A-3 item 33 A-87, A-88 Apparent typographical Error:

The implementation schedule in Table A-3 for item 33 states, in part: Start the one-time inspection in commitment 33-g) no earlier than 5 years prior to SPEO.

There appears to be a typographical error, i.e., Table A-3, item 33 commitment for one-time inspection is 33-f) and not 33-g).

a) Correct the Table A-3 (item 33) implementation schedule to make reference to the appropriate commitment regarding one-time inspection 6

B.2.3.30, A.2.2.30, Table A-3, item 33 B-224, B-225; A-28, A-29 USAR Supplement issue:

To be consistent with GALL-SLR AMP XI.S3, SLRA Section B.2.3.30 includes appropriate enhancements (corresponds to SLR Commitments 33(f) and 33(h))

to the detection of aging effects program element that specify one-time inspection of additional 5%

a) Provide revised language of referenced statements in the USAR supplement description in SLRA A.2.2.30 to state what the IWF AMP will do regarding one-time additional sample inspection and volumetric

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions sample of piping supports, and volumetric examination for high-strength bolting (HSB) greater than one inch in diameter, respectively.

However, the corresponding A.2.2.30 USAR supplement description includes language in the second paragraph that states, in part:..This AMP recommends additional inspections one-time of an additional 5 percent of the sample sizeFor high-strength bolting in sizes greater than 1-inch nominal diameter, volumetric examination

.should be performed to detect cracking in addition to VT-3 examination. (emphasis added)

The words highlighted above in the USAR summary description appears inconsistent with the language of the enhancements referenced above. An adequate USAR supplement summary description is expected to describe what the program will do and not what it recommends or what should be done.

examination of HSB greater than 1-inch diameter, consistent with the enhancements and commitments 33(f) and 33(h).

7 Table 3.5-1 3.5-66 AMR Item 3.5.1-068 Issue:

SLRA states item 3.5.1-068, related to SCC, is not applicable because there is no high-strength (HS) steel structural bolting is not used in MNGP structures or component supports.

This appears to be contrary to the enhancement (commitment 33(h)) related to volumetric examination of HS bolting for SCC, which appears to provide an adequate aging management of SCC consistent with GALL-SLR Report recommendations. It also appears to indicate that HS bolting exists and/or may be used in the future.

a) Discuss the basis (how) for the determination for SLRA AMR item 3.5.1-068 that HS bolting is not used in MNGP structures or component supports.

b) Clarify whether or not HS structural bolting exists or will be used in the future at MNGP, prior to or during SPEO, and if not use how will it be ensured that it will not be used in the future prior to or during the SPEO.

c) Clarify and correct any potential discrepancy in item 3.5.1-068 with the program enhancement (commitment

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 33(h)) and provide the basis for the determination.

8 B.2.3.30 B-227 Operating Experience for RV supports:

From the SLRA discussion in the third bullet under Industry Operating Experience regarding reactor vessel (RV) supports being not included in the IWF program, it is not clear if the subject industry OE applied to MNGP and what the general material condition of the RV supports is at MNGP.

a) Discuss the review and implications at MNGP of the referenced industry operating experience regarding inspection of RV supports.

b) State if the referenced industry OE applied to MNGP and since when has RV support inspected under the IWF program.

c) Discuss/explain in general the history of RV support inspections at MNGP, how and when these inspections are performed, and any findings of degradation from the inspections.

Illustration of inspected parts on an existing drawing(s) will be helpful if possible.

d) Discuss the general material condition of the RV supports (steel skirt and weld, anchorage, pedestal, RPV stabilizer brackets, stabilizer, rods and trusses etc). Provide recent representative photos of recent material condition of these RV supports, if available or taken during RF031.

9 B.2.3.30; 3.5.2.2.2.6, Table 3.5.2-1, Table 3.5.2-7 B-224 thru B-226; 3.5-30; 3.5-76 thru 3.5-83; SLRA 3.5.2.2.2.6 states, in part: Therefore, the integrity of the reactor vessel supports is assured, and no additional aging management of reactor vessel supports beyond the current ASME Section XI, Subsection IWF (B.2.3.30) AMP is necessary for aging effects due to irradiation during the MNGP SPEO.

a) While a plant-specific program may not be necessary, describe how the aging effects due to irradiation embrittlement will be adequately managed for the RV steel support assembly components for the SPEO?

Provide corresponding Table 2 AMR

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 3.5-98 thru 3.5-103 Nevertheless, while a plant-specific AMP may not be necessary [note this aspect is being reviewed under TRP 76 and not TRP 43], loss of fracture toughness due to irradiation embrittlement remains an applicable aging effect for the RV steel supports for SLR.

Although Table 2 of the PBD (XCELMO00017-REPT-077) on the ePortal includes loss of fracture toughness due to irradiation embrittlement among the aging effects/mechanisms managed by the program, SLRA 3.5.2.1.1, 3.5.2.1.7, Table 3.5.2-1, Table 3.5.2-7, do not include AMR items that the aging effect will be managed by the IWF AMP and SLRA B.2.3.30 does not appear to include loss of fracture toughness due to irradiation embrittlement as an aging effect that will be managed by the program.

items, noting that it currently is not included in the GALL-SLR Report.

b) Discuss how it is included within the scope of the program credited (e.g.,

IWF AMP) and any related changes that may need to be made to the AMP.

10 B-225 Element 3. discuss the difference between significant and excessive Commitment 33e 2.3.3.10 Fuel Pool Cooling and Cleanup Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

2.3.3.10 2.3-47 Scoping/Screening Boundary Drawing SLR-36257 Fuel Pool Filter/Demin. System [M-136]:

Various pressure indicators; flow elements; differential pressure indicators on this drawing are color coded as a(2) Spatial/Structural.

What spatial/structural issues exist within the Fuel Pool Filter/Demin.

Room(s) that require the demarcation of these instruments as (2) Spatial/Structural?

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 2

3.3.2.1.10 Service and Seal Water 3.4-108 through 3.4-111 SLRA Table 3.3.2-10: Fuel Pool Cooling and Cleanup -

Summary of Aging Management Evaluation:

[Page 619/1519] For Component Type Tanks (Skimmer Surge Tanks) the Material Carbon Steel (with Internal Coating) did not appear in LRA Table 3.3.2-10 Auxiliary Systems - Fuel Pool Cooling and Cleanup - Summary of Aging Management Evaluation

[pages 416/796; 417/796]. XCELMO00017-REPT-016, Revision 1 does not explain this new material type.

SLRA Section 3.3.2.1.10 contains the following excerpt:

The materials of construction for the FPC System components are:

• Carbon and Low Alloy Steel Bolting

• Carbon Steel

• Carbon Steel with Internal Coating

• Copper Alloy with Greater Than 15% Zinc

• Copper Alloy with 15% Zinc or Less

• Stainless Steel

• Stainless Steel Bolting Please confirm that the Material of Carbon Steel (with Internal Coating) listed in SLRA Table 3.3.2-10 for Tanks (Skimmer Surge Tanks) is correct.

3 2.3.3.10 2.3-47 Scoping/Screening Boundary Drawing SLR-36908 Fuel Reactor Building Sample System Flow Diagram [M-185]:

a) Coord. B-7, FPC system components -- Pressure

/Temperature Maximum Limitations Specified for Fuel Pool F/D -47B Outlet SX-2796 on FPW10B-6-HK located on SLR-36257(D,3) - Tubing to SX-2796 is not color-coded a(2) Spatial/Structural on SLR-36257 Coord. B-8, FPC system components - Pressure

/Temperature Maximum Limitations Specified for Fuel a) Staff requests an explanation for representation of tubing to and SX-2796 on drawing SLR-36257 as not subject to Aging Management Review.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Pool F/D -47A Outlet SX-2795 on FPW10A-6-HK located on SLR-36257(E,6) - Tubing to SX-2795 is not color-coded a(2) Spatial/Structural on SLR-36257 b) Staff requests an explanation for representation of tubing to and SX-2795 on drawing SLR-36257 as not subject to Aging Management Review.

2.3.3.17 Standby Liquid Control Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

2.3.3.17 2.3-63, 3.3-299 Scoping/Screening Boundary Drawing SLR-36253 Standby Liquid Control System, Coordinates C-4 and B-4

• Table 3.3.2-17: Standby Liquid Control - Summary of Aging Management Evaluation; For the Component Type Accumulator (Standby Liquid Control Accumulator) with a Material of Carbon Steel (with Internal Coating) and Elastomer: Each Material has respectively, the Environment of Treated (Water Internal) and Treated Water (External)

Is not the Environment of Sodium Pentaborate Solution more appropriate for these two interfaces of Material with Environment?

2 2.3.3.17 2.3-63 Scoping/Screening Boundary Drawing SLR-36253 Standby Liquid Control System, Coordinate C-3 Line DW8-1-HS and Valve DM-56 is displayed as a(2)

Spatial/Structural green.

• Either: a(1); a(2) FUNCTIONAL; or a(3) red appears to be appropriate. Should not this pipe segment and valve be color-coded as red?

3 2.3.3.17 2.3-63 Scoping/Screening Boundary Drawing SLR-36253 Standby Liquid Control System, Coordinate D-2:

a) The Sparger in the SLC Tank is Seismic Class II.

[Reference DBD-B.03.05 Revision 5 Section 5.2 SLC (Storage) Tank, T-200; Section E Special Design Considerations]

b) Pipe line DW7 HK (attached to the SLC Tank) is color coded as a(2) Spatial/Structural a) Is this line ASME-Class 1 pipe and is it Seismic Class 1?

• If so, are its Seismic anchors up to isolation valves DM-52 and AS-22 subject to AMR?

• If so (continued) Is line SA-3/4-JB (attached to DW7 HK) qualified as Seismic II/I? OR Is it compatible with the guidance of SLRA Section 2.1.4.2.2 Non-Safety Related SSCs Directly Connected to Safety-Related SSCs that

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Provide Structural Support for the Safety-Related SSCs (i.e. the fourth bullet) b) Please provide information as to why the sparger is designated as neither an a(2) FUNCTIONAL nor an a(2)

Spatial/Structural component.

4 2.3.3.17 2.3-63 Scoping/Screening Boundary Drawing SLR-36253 Standby Liquid Control System, Coordinate E-2:

a) The vent line on the Standby Liquid Control Tank is color coded as neither a(1) nor a(2) -- not subject to Aging Management Review (black).

b) Pipe line D68-3-HC is color coded as a(2)

Spatial/Structural. It is not clear whether the line represents a Spatial leakage concern or a Structural concern OR BOTH.

a) If the vent line should be crimped during a seismic event would the system (e.g., pumps) still be operable due to a lack of NPSH at the pumps? [Reference Information Notice No. 91-12: Potential Loss of Net Positive Suction Head (NPSH) of Standby Liquid Control System Pumps]

b) Is this line qualified as Seismic I out to the pipe class change HC/HF? In particular, are the necessary seismic I and seismic II/I supports for this line subject to AMR consistent with the guidance of 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?

5 2.3.3.17 2.3-63 Scoping/Screening Boundary Drawing SLR-36253 Standby Liquid Control System, Coordinate E-3:

Instrument Air to Tank Level Instrumentation is not subject to AMR (black) on the AIR side but has an a(3) ATWS function (red) on the SLC side of the flag of demarcation.

Both level -transmitters LT 11-45 and LT 7449 have independent Class 1E and/or D.C. power supplies (e.g.,

via the EDGs.) [Reference DBD-B.03.05 Revision 5 a) Does LT 11-45 remain functional/operable in the absence of an Instrument Air pressure supply?

b) Please provide the basis for this scoping/screening of the instrument air supply to LT 11-45.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Section 4.3.2 Electrical & USAR Section 6.6.2 Description] The instrument air for LT 11-45 has no accumulator for air storage shown on SLR-36049-13.

Flow-Accelerated Corrosion Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

A.2.2.9: Additionally, the software tool, FAC Manager',

with the erosion module, is used to evaluate components for both FAC and erosion.

Discuss the erosion module being referenced in the associated UFSAR Supplement.

2 EPRI 3002005530. Recommendations for an Effective Program Against Erosive Attack (discussed in the Surry SLRA) states that a minimum safety factor should never be less than 2.0.

Discuss what safety factor will be used for evaluations of erosive wear.

3 AR 600000229895 Proactive OE Search discusses Imperia FAC Manager Database Discuss referenced database (is it available for NRC review?)

4 RFO23 ISI-21-E10 FAC-07-072 shows average thickness of 0.611, whereas the nominal thickness is listed as 0.343.

Heavier wall pipe fitting was apparently used. Does this create a flow perturbation that can cause unanticipated downstream wall thinning? Can current inspection grid capture this effect if it is occurring? Are there other fittings between the reactor vessel and the 1st isolation valve that also are heavier wall that may need to be looked into?

B.2.3.11 - Open-Cycle Cooling Water System Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.3.11 B-82 The SLRA states, For the components within the scope of the MNGP Open-Cycle Cooling Water Systems AMP that have internal coatings (e.g., various piping components within the ESW System, etc.), the internal coatings will be managed by the MNGP Internal It appears there is an inconsistency in the description of the OCCW systems in the SLRA and Engineering Work Instruction 11.01.12 Revision 6. A review of operating experience showed a finding

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks AMP (B.2.3.28).

The Open-cycle Cooling Water System Program document (EWI-11.01.12 Rev. 6) states in Section 4.3 that, The MNGP Open Cycle Cooling Water (OCCW) systems are constructed of appropriate materials and are not lined or coated.

Page B-90 of the SLRA documents an OpE finding (501000041947), In October 2020, as part of an RHR heat exchanger inspection, several areas of the heat exchanger wall and divider plate were identified with 1/2-inch to 2-inch tubercules. The coating area under and immediately surrounding these tubercles was degraded and would easily flake away. No wall loss was observed, so pressure boundary function was not challenged. Mating surfaces were in good condition and no degradation was observed in these areas. The areas of degradation were remediated under a coating repair contingency work order.

that referenced an RHR heat exchanger with a coating on the divider plate (page B-90).

It appears EWI 11.01.12 should be revised.

3.3 Aging Management of Auxiliary Systems Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

3.3 3.3-183 Item VII.C1.A-473, SRP Item 3.3.1-160 in NUREG-2191, Volume 1 addresses cracking due to SCC for copper alloy

(>15% Zn or >8% Al) piping, piping components, and heat exchanger components exposed to closed-cycle cooling water, raw water, and waste water managed by several programs including AMP XI.M38, Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components.

AMP XI.M38 notes that periodic surface examinations are Please describe how the Fire Water System program will manage cracking of copper alloy >15% Zn heat exchanger -

(diesel fire pump) tube sheet and tubes, and piping and piping components exposed internally to raw water.

Specifically discuss whether surface examinations will be performed or whether analyses will be performed to demonstrate that surface cracks can be

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions conducted for managing cracking in stainless steel and aluminum components and states, Visual inspections for leakage or surface cracks are an acceptable alternative to conducting surface examinations to detect cracking if it has been determined that cracks will be detected prior to challenging the structural integrity or intended function of the component.

SLRA Table 3.3.2-9 states that cracking for copper alloy

>15% Zn heat exchanger - (diesel fire pump) tube sheet and tubes, and piping and piping components exposed to raw water will be managed by the Fire Water System program. The corresponding aging management review item (3.3.1-160) cites Standard Note E (consistent with GALL-SLR but different program credited) and Plant-specific Note 1 for the use of the Fire Water System program in lieu of the Open-cycle Cooling Water System program.

AMP XI.M27 does not provide additional guidance for managing cracking, whereas AMP XI.M38 does provide additional guidance for managing cracking. Section B.2.3.16 in SLRA Appendix B does not describe how the Fire Water System program inspections and testing performed in accordance with NFPA 25 will manage cracking of copper alloy >15% Zn heat exchanger -

(diesel fire pump) tube sheet and tubes, and piping and piping components exposed internally to raw water.

detected by leakage prior to a crack challenging the intended function of the component, such that visual inspections would suffice.

2 3.3 3.3-201 Item VII.G.A-787b, SRP Item 3.3.1-253 in NUREG-2191, Volume 1 addresses loss of material and flow blockage (raw water only) of PVC piping and piping components exposed to raw water, raw water (potable), and treated water managed by the Fire Water System program.

The Discussion for AMR Item 3.3.1-253 in SLRA Table 3.3-1 states, The Inspection of Internal Surfaces in Please discuss whether flow blockage is an applicable aging effect requiring management for the PVC valve bodies exposed internally to raw water.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Miscellaneous Piping and Ducting Components (B.2.3.24)

AMP and Fire Water System (B.2.3.16) AMP are used to manage loss of material of PVC piping and piping components exposed to treated water and raw water.

SLRA Table 3.3.2-9 cites AMR Item 3.3.1-253 to manage loss of material for PVC valve bodies exposed internally to raw water. However, it does not cite flow blockage as an applicable aging effect requiring management for the PVC valve bodies exposed internally to raw water.

3 Appendix A,

Appendix B

A-65 B-116

1175, 1325 The implementation schedule for the Fire Water System program in Table XI-01 of GALL-SLR states, Program is implemented and inspections or tests begin 5 years before the subsequent period of extended operation.

Inspections or tests that are to be completed prior to the subsequent period of extended operation are completed 6 months prior to the subsequent period of extended operation or no later than the last refueling outage prior to the subsequent period of extended operation.

Table A-3 in SLRA Appendix A states the implementation schedule as the following:

No later than 6 months prior to the SPEO, or no later than the last refueling outage prior to the SPEO Implement the AMP and start the pre SPEO inspections and tests no earlier than 5 years prior to the SPEO.

As written, it is unclear what is to be done no later than 6 months prior to the SPEO, or no later than the last refueling outage prior to the SPEO.

SLRA Section B.2.3.16 states, The enhancements are to be implemented no later than 6 months prior to entering the SPEO. This AMP is to be implemented and its pre-Please clarify the implementation schedule for the Fire Water System program, with enhancements.+K149

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SPEO inspections and tests begin no earlier than 5 years prior to the SPEO. The pre-SPEO inspections and tests are to be completed no later than six months prior to entering the SPEO or no later than the last refueling outage prior to the SPEO.

Section 6.0 of the Fire Water System program basis document states, Program implementation, enhancements, and pre-SPEO inspections and tests are to be completed no later than 6 months prior to the SPEO, i.e., 03/08/2030, or no later than the last refueling outage prior to the SPEO. Implement the AMP and start pre-SPEO inspections and tests no earlier than 5 years prior to the SPEO (09/08/2025).

As written, it is unclear whether the program, with enhancements, is implemented 6 months prior to the SPEO, or no later than the last refueling outage prior to the SPEO, or 5 years prior to the SPEO.

4 Appendix A,

Appendix B

A-19, B-113 Footnote 7 of Table XI.M27-1 in GALL-SLR AMP XI.M27 states, For wet pipe sprinkler systems, the subsequent license renewal application either: (1) Provides a plant-specific evaluation demonstrating that the water is not corrosive to the sprinklers (e.g., corrosion-resistant sprinklers); or (2) Proposes a one-time test of sprinklers that have been exposed to water including the sample size, sample selection criteria, and minimum time in service of tested sprinklers; or (3) Proposes to test the sprinklers in accordance with NFPA 25 Section 5.3.1.1.2.

Based on Monticello Nuclear Generating Plant, Revision 36 to Updated Safety Analysis Report, Appendix J, Fire Protection Program, January 1, 2019, ADAMS Accession No. ML19018A180, there are wet pipe sprinklers in the Lube Oil Storage Tank Room, Hydrogen Seal Oil Unit, and under the turbine floor (Lube Oil Piping Sprinkler Please confirm the staffs assumption that the wet pipe sprinklers will be tested in accordance with NFPA 25, Section 5.3.1.1.2 as recommended in Footnote 7 of Table XI.M27-1.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions System).

Section A.2.2.16 in SLRA Appendix A and Section B.2.3.16 in SLRA Appendix B do not explicitly state which option from Footnote 7 will be applied to the wet pipe sprinklers. However, the staff notes that the table in Section B.2.3.16 of SLRA Appendix B that provides additional detail on the enhancements based on GALL-SLR AMP XI.M27, Table XI.M27-1, that NFPA 25, Section 5.3.1.1.2 is cited. Therefore, the staff assumes that the wet pipe sprinklers will be tested in accordance with NFPA 25, Section 5.3.1.1.2.

5 Appendix A,

Appendix B

A-65, B-117, B-135 GALL-SLR AMP XI.M27 states that portions of water-based fire protection system components that have been wetted but are normally dry are subject to augmented testing and inspection beyond Table XI.M27-1. The augmented tests ad inspections include, In each 5-year interval, beginning 5 years prior to the subsequent period of extended operation, either conduct a flow test or flush sufficient to detect potential flow blockage, or conduct a visual inspection of 100 percent of the internal surface of piping segments that cannot be drained or piping segments that allow water to collect.

However, Table A-3 in SLRA Appendix A and Section B.2.3.16 in SLRA Appendix B do not include this in the enhancement to the Parameters Monitored or Inspected program element related to the augmented tests and inspections. The staff notes that the Fire Water System program basis document also does not include this in Sections 4.3 and 6.0.

The staff notes that the table in Section B.2.3.16 of SLRA Appendix B that provides additional detail on the enhancements based on GALL-SLR AMP XI.M27, Table XI.M27-1, includes the augmented tests and inspections Please discuss why the augmented tests and inspections beginning 5 years prior to SPEO are not included in the enhancement to the Parameters Monitored or Inspected program element related to the augmented tests and inspections.

In addition, if possible, please identify which portions of the fire protection system are subject to augmented testing or inspection because they are normally dry but periodically subjected to flow and cannot be drained or allow water to collect.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions beginning 5 years prior to SPEO, and the Fire Water System program basis document includes this information in Table 4.4-1.

6 Appendix A,

Appendix B

A-67, B-118 GALL-SLR AMP XI.M27 states, Results of flow testing (e.g., buried and underground piping, fire mains, and sprinkler), flushes, and wall thickness measurements are monitored and trended.

Table A-3 in SLRA Appendix A and Section B.2.3.16 in SLRA Appendix B include the following enhancements:

Monitoring and Trending - Results of flow testing (e.g.,

buried and underground piping, fire mains, and sprinklers/spray nozzles), flushes, and wall thickness measurements will be monitored and trended per the instructions of the specific test/inspection procedure. The staff notes that this appears to indicate that the process for trending results could vary depending on the test/inspection procedure.

Monitoring and Trending - Update spray and sprinkler system flushing procedures to enable trending of data.

Specifically, the existing flushing procedures and preventive maintenance activities will be revised to document and trend deposits (scale or foreign material) and Existing flushing procedures, as well as new flushing procedures, will include steps to compare the amount of deposits to the previous inspections results, and if the trend shows increasing deposits, then the MNGP CAP will be utilized to drive improvement. The staff notes that this appears to indicate that trending may occur outside of the corrective action program.

Acceptance Criteria - Clarify within the new internal inspection procedure and relevant existing preventive maintenance activities which inspect wall thickness that Please clarify the trending process for flow testing and wall thickness measurements, including whether trending of inspection results is performed outside of the corrective action program for instances where the results are not entered into the corrective action program.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions identified wall loss greater than the manufacturers tolerance will be entered into the MNGP CAP for engineering evaluation and trending to determine when minimum wall thickness will be reached and what corrective actions are required. The staff notes that this is unclear whether trending occurs outside of the corrective action program.

It is unclear to the NRC staff if trending of flow test and wall thickness measurement inspection results is performed outside of the corrective action program for instances where results are not entered into the corrective action program.

7 N/A N/A NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appurtenances, Section 10.4.2.1, Protection for Piping, states that the top of the pipe shall be buried not less than 12 inches below the frost line for the locality.

Section 3.3 of the Fire Water System program basis document states, The hydrant test procedure includes a step to confirm that no more than 3 inches of water remain in the barrel after a flushing and draining, otherwise the water is removed (pumped out). Since the fire mains are designed to be below the frost line and each hydrant plunger is relatively near the fire main, there is reasonable assurance that, due to heat from the earth, the limited amount of water ( 3 inches height) within the hydrant barrel will not freeze.

Please confirm how far below the frost line the 3 inches of water left in the hydrant barrel is.

8 3.3 3.3-11, 3.3-188 Item VII.G.A-797b, SRP Item 3.3.1-263 in NUREG-2191, Volume 1 addresses hardening or loss of strength, loss of material, cracking, and flow blockage of polymeric piping, piping components, ducting, ducting components, seals exposed to air, condensation, raw water, raw water (potable), treated water, waste water, underground, concrete, and soil managed by the Inspection of Internal Please identify which piping and piping components are polymeric and identify the polymeric material.

In addition, please discuss whether flow blockage is an applicable aging effect for the polymer piping and piping

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Surfaces in Miscellaneous Piping and Ducting Components program.

SLRA Section 3.3.2.1.9 identifies polymer as a material of construction for fire system components. SLRA Table 3.3.2-9 includes several line items for polymer piping and piping components. These line items cite Item VII.G.A-797b, SRP Item 3.3.1-263 for managing cracking, hardening or loss of strength, and loss of material of polymer piping and piping components exposed internally to raw water by the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program.

Flow blockage is not identified as an aging effect being managed.

The staff did not find which specific piping and piping components are polymeric and what the polymeric material is (also for line items citing Item VII.G.A-797a, SRP Item 3.3.1-263, External Surfaces Monitoring of Mechanical Components program).

The discussion of AMR Item 3.3.1-263 in SLRA Table 3.3-1 states, The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24) and External Surfaces Monitoring of Mechanical Components (B.2.3.23) AMPs are used to manage hardening or loss of strength, loss of material, and cracking of polymeric piping and piping components exposed to air indoor uncontrolled, raw water, or treated water. Flow blockage is not identified as an aging effect being managed.

components exposed internally to raw water.

9 N/A N/A Section 4.10 of the Fire Water System program basis document states, Element 10 of the MNGP Fire Water System AMP, with the enhancements included above, will be consistent without exception to NUREG-2191.

However, it is unclear what the enhancements are. In Please discuss whether there are enhancements to Element 10.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions addition, Table A-3 in SLRA Appendix A and Section B.2.3.16 in SLRA Appendix B do not appear to identify enhancements to Element 10.

10 Appendix A,

Appendix B

A-68, B-120 The staff noted that Table A-3 in SLRA Appendix A and Section B.2.3.16 in SLRA Appendix B states, in part, the following for the enhancement to the Corrective Actions program element, If a failure occurs (e.g., a through-wall leak or blockage impacting operability), the failure mechanism shall be identified and used to determine the most susceptible system locations for additional inspections, including consideration to the other unit systems as driven by the corrective action program.

This is an observation given that Monticello is a single unit site.

11 N/A N/A Section 6.0 of the Fire Water System program basis document states, in part, the following for the enhancement to Element 5, Additionally, identified deposits will be evaluated for potential impact on downstream components, such as sprinkler heads or spray nozzles.

Table A-3 in SLRA Appendix A and Section B.2.3.16 in SLRA Appendix B do not appear to include this discussion on impact on downstream components in the enhancement to the Monitoring and Trending program element.

Please discuss whether the enhancement to the Monitoring and Trending program element in Table A-3 in SLRA Appendix A and Section B.2.3.16 in SLRA Appendix B should include this discussion on the impact on downstream components.

12 Appendix A,

Appendix B

A-65, B-116 Section B.2.3.16 in SLRA Appendix B states, in part, for the enhancement to the Parameters Monitored or Inspected program element, The internal inspections will be performed during the periodic system and component surveillances.

This statement does not appear in the Fire Water System program enhancements in Table A-3 in SLRA Appendix A.

Please discuss, for consistency, whether this statement should be included in Table A-3 in SLRA Appendix A.

13 3.3 3.3-186 Item VII.G.AP-76, SRP Item 3.3.1-096 in NUREG-2191, Volume 1 addresses loss of material and flow blockage (raw water only) for elastomer piping, piping components, and seals exposed to air, raw water, raw water (potable),

Please discuss whether flow blockage is an applicable aging effect for elastomer hoses (pump and drain hoses) exposed internally to raw water.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions and treated water by the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program.

The Discussion of AMR Item 3.3.1-096 in SLRA Table 3.3-1 states, The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)

AMP is used to manage loss of material of elastomer piping, piping components, ducting and components exposed to raw water, waste water, treated water, air indoor uncontrolled, or condensation. The discussion does not address flow blockage.

SLRA Table 3.3.2-9 cites AMR Item 3.3.1-096 for loss of material of elastomer hoses (pump and drain hoses) exposed internally to raw water, however, flow blockage is not cited as an applicable aging effect for these components.

14 3.3 3.3-187 Items VII.G.A-33 and VII.G.AP-197, SRP Item 3.3.1-064 in NUREG-2191, Volume 1 address loss of material and flow blockage (raw water, raw water (potable) only) for steel and copper alloy piping and piping components exposed to raw water, treated water, and raw water (potable) managed by the Fire Water System program.

The Discussion of AMR Item 3.3.1-064 in SLRA Table 3.3-1 states, This line item is also applied to heat exchanger components. The Fire Water System (B.2.3.16) AMP is used to manage loss of material and flow blockage of steel and copper alloy piping, piping components, and heat exchanger components exposed to raw water.

SLRA Table 3.3.2-9 cites AMR Item 3.3.1-064 for loss of material of carbon steel piping and piping components, gray cast iron pump casing (fire system jockey pump),

copper alloy with greater than 15% zinc valve bodies Please discuss whether flow blockage is an applicable aging affect for carbon steel piping and piping components, gray cast iron pump casing (fire system jockey pump), copper alloy with greater than 15% zinc valve bodies exposed externally to raw water

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions exposed externally to raw water. However, flow blockage is not cited an as applicable aging effect for these components.

15 3.3 Items VII.I.A-77, SRP Item 3.3.1-078 in NUREG-2191, Volume 1 addresses loss of material of steel external surfaces exposed to indoor air uncontrolled, outdoor air, and condensation managed by the External Surfaces Monitoring of Mechanical Components program.

The Discussion of AMR Item 3.3.1-078 in SLRA Table 3.3-1 states, The External Surfaces Monitoring of Mechanical Components (B.2.3.23) AMP is used to manage loss of material of steel external surfaces exposed to air indoor uncontrolled, condensation, and air outdoor.

SLRA Table 3.3.2-9 includes galvanized steel piping and piping components exposed externally to outdoor air, carbon steel valve bodies exposed externally to indoor uncontrolled air, and gray cast iron valve bodies exposed externally to indoor uncontrolled and outdoor air.

However, AMR Item 3.3.1-078 is not cited for managing loss of material of these components.

Please discuss whether AMR Item 3.3.1-078 should be cited for galvanized steel piping and piping components exposed externally to outdoor air, carbon steel valve bodies exposed externally to indoor uncontrolled air, and gray cast iron valve bodies exposed externally to indoor uncontrolled and outdoor air.

16 3.3 3.3-11, 3.3-182 Page 27 of 100 in B.08.05-05 includes diesel fire pump cooling water jacket heater and lube oil heater.

SLRA Section 3.3.2.1.9 does not include lubricating oil as an applicable environment for fire system components.

SLRA Table 3.3.2-9 includes several entries for shell side components, tube sheet, tube side components, and tubes for the diesel driven fire pump heat exchanger. The cooling water jacket heater and lube oil heater are not explicitly stated as a component type.

Please discuss whether the cooling water jacket and lube oil heater are included in the heat exchanger component types in SLRA Table 3.3.2-9. In addition, discuss whether lubricating oil is an applicable environment for the diesel fire pump heat exchangers.

17 Appendix B

B-115 Section B.2.3.16 in SLRA Appendix B includes an exception to the annual main drain tests at each water-Please clarify whether the 20 percent sample will be at different locations each

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions based riser per NFPA 25, Section 13.2.5. Instead of a main drain test at each riser, main drain tests on 20 percent of the standpipes and risers will be performed each refueling cycle.

The staff notes that Section B.2.3.16 in SLRA Appendix B includes the following enhancement to the Monitoring and Trending program element related to sampling-based inspections: For sampling-based inspections, results will be evaluated against acceptance criteria to confirm that the sampling bases (e.g., selection, size, frequency) will maintain the components intended functions throughout the SPEO based on the projected rate and extent of degradation.

refueling outage. If so, will that result in all riser and standpipes being inspected over a certain period (i.e., 10 years)?

Bolting Integrity Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

Bolting Integrity Basis

Document, Sections 4.2 SLRA Section B.2.3.10 Basis doc pg. 8 of 32 Page B-76 Section 4.2, Preventive Actions, in the basis document (XCELMO00017-REPT-073) notes that tread lubricant containing MoS2 is not used, as described in procedure MMP-008, revision 12.

Basis Document Sections 4.2 and SLRA Section B.2.3.10 states that procedures will be enhanced to clarify that in addition to MoS2, other lubricants containing sulfur will be prohibited from use on pressure-retaining closure bolting.

It Is unclear to the staff whether other lubricants containing sulfur exist in the plant currently.

Please clarify whether any other lubricants containing sulfur are currently used for closure bolting.

If other lubricants containing sulfur were used, discuss if procedure is needed to identify those other lubricants containing sulfur and how to manage the potential aging effect due to their use.

2 Bolting Integrity Basis doc pg. 8 of 32 GALL-SLR XI.M18, Element 4, Detection of Aging Effects, stats that Closure bolting inspections includes consideration of the guidance applicable for pressure Please clarify whether the existing inspection procedure includes applicable guidance from EPRI NP-5769 and

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Basis

Document, boundary bolting in NUREG-1339 and in EPRI NP-5769.

Section 4.2, Preventive Actions, in the basis document (XCELMO00017-REPT-073) refers to NUREG-1339 and EPRI NP-5769 as two of guidelines to prevent or mitigate the degradation of bolting.

However, Section 4.4 Detection of Aging Effects in the base document doesnt explicitly mention that closure bolting inspections includes consideration of the guidance applicable for pressure boundary bolting in NUREG-1339 and in EPRI NP-5769.

NUREG-1339 as referenced in GALL-SLR XI.M18, Element 4. If not, please enhanced the procedures to incorporate the applicable guidance from EPRI NP-5769 and NUREG-1339 accordingly.

B.2.3.29 ASME Section XI, Subsection IWE Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.3.29 B-220, B-222 SLRA B.2.3.29 under Plant-Specific Operating Experience states in part: The ASME Section XI, Subsections IWB, IWC and IWD inservice inspection program health report (July 2020) was reviewed. The overall program performance was exceptional (GREEN). The program performance described is based on IWB, IWC and IWD ISI program and not a review of the IWE program health.

SLRA B.2.3.29 also talks about focused and snapshot self-assessment, but it appears to be in the context of IWB, IWC and IWD program and not the IWE program. Also, second bullet under Action Request Examples mentions IWB, IWC, and IWD program owner, but not IWE.

ePortal AMPS/XI.S1 Section XI IWE folder includes Appendix J Program Health Report and not the IWE program health report.

a) Provide recent 2 program health reports for the ASME SC XI - IWE program for verification that the program performance was exceptional as stated in the SLRA. If health assessment not performed or not available for IWE program, please explain.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 2

B.2.3.29 B-220, SLRA B.2.3.29, last bullet under subtitle Industry Operating Experience related to insulation not being removed for general visual examination of flue head and piping penetrations states: Inspection of subject surfaces were completed during 2021 RFO. The applicable plant procedure was updated in May 2021 to add inspection requirements for external surfaces of the drywell penetrations and to provide guidance for visual examination of insulated components.

The staff was unable to verify the above on the ePortal.

a) Provide the AR/WO that completed the subject inspections for verification and, referring to the AR/WO, explain how the inspection was completed.

b) Identify and point to the specific procedure and guidance that was revised to incorporate visual examination of insulated components.

3 B.2.3.29; 3.5.2.2.1.5, XCELMO00017-REPT-076 (PBD) Section 4.3 B-214 &

B-215, 3.5-24 The first bullet under Exceptions to NUREG-2191 states in part: 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.

(emphasis added).

The same statement is also made in SLRA Section 3.5.2.2.1.5. It is not clear what the phrase in bold means in the context of the assessment being referred to, and it is not clear what non-high temperature drywell penetrations, and penetration sleeves mean.

Based on review of documents on the ePortal, the staffs understanding is that the stated assessment is the Structural Integrity Associates calculation 2100507.308, Revision 0 (9/29/2022), Fatigue Exemption of the Monticello Drywell, and noted it performed a fatigue waiver analysis of select drywell a) Explain what is meant by the phrase nor a fatigue waiver is required when the assessment referred to in the statement and summarized in the previous paragraph in the SLRA (the six fatigue waiver conditions in the ASME code), and as indicated in SIA calculation 2100507.308, appears to be a fatigue waiver analysis.

b) Clearly state what was done in the assessment and summarizing the results and criteria that was met to justify the exception; also, clearly state the components to which the exception applies.

c) Clarify the apparently contrarian statements in the SLRA regarding whether or not the aging effect of cracking due to cyclic loading requires aging management and the basis for that determination.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions components (without CLB fatigue analysis), including non-high temperature mechanical and electrical penetrations, in accordance with NE-2222.4(d)

Vessel Not Requiring Analysis for Cyclic Operation of the ASME code, 1974 edition and demonstrated that the fatigue waiver criteria therein were met. This would justify that cracking due cyclic loading does not require aging management for these components consistent with the acceptance criteria in Section 3.5.2.2.1.5 of SLRA-ISG-2021-03-STRUCTURES, February 2021.

The second bullet in the SLRA exception appears to make the contrarian statement that:..Through wall cracking would be detected by the Type A ILRT.

..Thus, existing 10 CFR 50, Appendix J leak testing

[Type A] and ASME Section XI, Subsection IWE examinations at MNGP remain adequate for the drywell shell, non-high temperature drywell penetrations, and penetration sleeves without supplemental surface examination to detect cracking.

Contrary to the conclusion drawn in the previous paragraph and in the SLRA that the aging effect does not require management, this statement appears to indicate that the aging effect requires management.

The referenced SLRA statements related to the exception are also included in Section 4.3 of the PBD.

The SLRA states that the IWE AMP with enhancements is consistent with GALL-SLR AMP XI.S1 with two exceptions. It is not clear from the SLRA description if there are one or two exceptions and what the distinction is between the two, and the specific components to which each exception applies.

d) Clarify if there are two exceptions or one exception being taken to the XI.S1 AMP and clearly identify each exception. If there are two, explain the distinction between the two exceptions.

Clearly identify the specific components for which the exception(s) apply.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 4

B.2.3.29, XCELMO00017-REPT-076 (PBD) Section 4.1 B-214; PBD p8-9 Consistency of Preventive Action program element Program Basis Document (PBD) Section 4.1 states in part: The above scope is supplemented to address aging management of potential corrosion in inaccessible areas of the drywell shell exterior of the MNGP Mark I steel containment [Ref 9.5, Drawing NX-8291-2-B; Ref 9.17, Ref 9.18; Ref 9.19].

PBD Section 4.2 Preventive Actions does not appear to include monitoring of refueling seal drains for blockage and leakage.

The preventive actions program element of GALL-SLR AMP XI.S1 recommends ensuring that the sand pocket area drains and/or refueling seal drains are clear.

The staff needs clarification of how the program is supplemented and, to verify consistency, whether refueling seal drains exist and inspected.

a) For clarity of the staffs understanding, explain how the program is supplemented for aging management of potential corrosion of inaccessible areas of the drywell shell exterior.

b) Clarify for the staffs understanding if refueling seal drains (drywell to reactor building and RPV to drywell) exist. If they do exist, explain the configuration (preferably illustrated on a drawing) and, whether and how (procedure) the refueling seal drains are inspected for blockage and leakage.

5 XCELMO00017-REPT-076 (PBD) Section 4.2 PBD p10 PBD Section 4.2 states in part: High-strength structural bolting is not used in the structural design of the MNGP steel containment vessel [Ref 9.23].

Ref 9.23 is Drawing NX-8291-76, Revision C.

The copy of the drawing on the ePortal is not clear regarding the referenced statement above.

a) Clarify where the referenced statement is made on the Ref 9.23 drawing or how the determination was made.

b) Clarify the term high strength structural bolting in terms of yield/tensile strength. Also discuss if high strength bolting will be used in the future during or prior to the SPEO.

6 B.2.3.29, Table A-3 (item 32)

B-216, A-86 The SLRA includes an enhancement to detection of aging effects program element (corresponds to SLR Commitment 32(c)) regarding supplemental surface or enhanced visual examinations to detect cracking of high temperature piping penetrations.

a) Specify in the enhancement/commitment the temperature threshold above which defines high temperature piping penetrations; also, discuss its basis.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The enhancement/commitment does not specify the temperature threshold that defines high-temperature piping penetrations.

7 B.2.3.29, Table A-3 (item 32)

B-216, A-86 The SLRA includes enhancements to detection of aging effects program elements corresponding to SLR Commitments 32(c) and 32(d) related to supplemental periodic and supplemental one-time examinations to detect cracking due to cyclic loading and SCC, respectively.

It is not clear if there are components that are subject to both these commitments.

a) Clarify if there are common components subject to actions specified in both Commitments 32(c) and 32(d) and explain the basis for that determination.

8 B.2.3.29, Table A-3 (item 32)

B-217, A-86 The SLRA includes an enhancement to detection of aging effects program element (corresponds to SLR Commitment 32(e)) regarding plant-specific OE trigger-based one-time volumetric examination of containment shell metal surfaces inaccessible from one side. The enhancement states in part: Any such instance would be identified through code inspections performed since November 8, 2006 The trigger specified in the GALL-SLR is the site-specific occurrence or recurrence of the stated plant-specific OE without regard to the method, program or process by which (how) it is identified. Contrary to this, the SLRA enhancement states that the triggering OE would be specific to that identified through code inspections (emphasis added), which would be an unjustified exception to the GALL-SLR AMP XI.S1.

Corrosion of the containment metal surfaces that originated from the inaccessible side could be identified by several other means such as during repair/replacement activities, maintenance rule inspections and walkdowns, Appendix J activities, in addition to IWE code inspections, and should therefore be neutral to program or process of its a) Provide a revised enhancement to the detection of aging effects program element in SLRA Section B.2.3.29, related to one-time supplemental volumetric examination, and the associated LR Commitment 33(f) that would make the AMP program element consistent with that in GALL-SLR AMP XI.S1, neutral to how (method, program or process) the triggering operating experience is identified.

b) Additionally, also provide an implementation schedule (relative to the date of occurrence of the triggering OE) for implementation of the one-time supplemental volumetric examination.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions identification.

While the SLRA provides an implementation schedule to revise procedures for SLR Commitment 32(e), it does not appear to provide a schedule for implementing the one-time volumetric examination with respect to the date of occurrence of the triggering OE.

9 B.2.3.29; Table A-3 (item 32);

A.2.2.29 B-217, A-86, A-28 The SLRA includes an enhancement to the corrective actions program element (corresponds to SLR Commitment 32(f)) regarding corrective actions is SCC is identified. This enhancement states, in part::

This will include one additional penetration with DMWs associated with greater than 140oF stainless steel piping systems until cracking is no longer detected.

There appears to be lack of clarity and context (something missing) in the above referenced statement regarding each additional inspection is with respect to what result in the sample inspected.

Also, the A.2.2.29 USAR supplement does not appear to include a description of the proposed one-time inspection to confirm the absence of SCC.

Specify with clarity in the enhancement/commitment 32(f) how sample expansion will be conducted when SCC is detected or absence of SCC cannot be confirmed as a result of the initial sample of supplemental one-time inspections.

b) Provide a description of the proposed one-time inspection to confirm the absence of SCC aging effect in the USAR supplement description for AMP.

10 CISI IWE PLAN (Procedure) 1.2-9 The IWE Plan discusses on page 1.2-9 Category E-C, Containment Surfaces Requiring Augmented Examination.

It is not clear to the staff what containment surfaces are subject to augmented examination in the current and previous CISI intervals.

a) Discuss representative areas of the Primary Containment that have been identified for Examination Category E-C Augmented examination and the degraded condition found there for (a) the current 3rd CISI interval, and (b) for the previous (2nd) CISI interval.

b) Provide on ePortal, if available, representative photographs of the recent general material condition of the primary containment components

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions (drywell, torus, downcomers etc) in general or if available those areas that have been subjected to augmented examination.

11 XCELMO00017-REPT-076 (PBD) Section 4.5 PBD p15 PBD Section 4.5 states that based on UT examinations performed in 1986, 1987 and 1996, the loss of material due to corrosion of inaccessible areas of the drywell is insignificant.

It is not clear if additional UT examinations of the drywell was performed since 1996 and if so, what the results and conclusions were.

a) State if additional UT examinations were performed on the drywell shell since 1996.

b) If so, provide and discuss results and conclusions of these UT examinations and whether it demonstrates that the corrosion rate of inaccessible areas of drywell shell continues to remain insignificant.

c) Provide, if available, photographs representative of recent general material condition of the primary containment (drywell, torus, vent system).

12 Table 3.5-1, Table 3.5.2-1, 3.5.2.2.1.3.1, B.2.3.29 3.5-44, 3.5-78, 3.5-21 SLRA Table 3.5-1, item 3.5.1-004, related to loss of material due to corrosion in inaccessible areas of steel drywell shell, drywell head and embedded shell and drywell shell in sand pocket areas, exposed to air-indoor (uncontrolled) or concrete environment, states that these AMRs are Not applicable.

SLRA Table 3.5-1, item 3.5.1-035 for the same aging effects in accessible areas of the drywell and sand pocket regions appropriately states the item is consistent with NUREG-2191.

The first sentence of SLRA FE Section 3.5.2.2.1.3.1 also includes item 3.5.1-004 and 3.5.1-035 as not applicable. However, this SLRA Section also concludes: Loss of material due to corrosion in a) Discuss and correct the noted inconsistencies in the referenced SLRA sections with regard to non-applicability claim of AMR items 3.5.1-004 and 3.5.1-035 for accessible, inaccessible and embedded areas of drywell shell given that the component material and environment for the aging effect exists for the MNGP drywell (i.e.., to be consistent with corresponding AMR items in SLRA Table 3.5.2-1.

b) Provide a discussion or statement of whether there has been plant-specific OE of significant corrosion exists in inaccessible areas of components

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions inaccessible areas of the steel containment will be managed by ASME SC XI - IWE (B.2.3.29) and 10 CFR 50 Appendix J (B.2.3.31) AMPs during SPEO, and a separate plant-specific AMP is not required.

Further, SLRA Table 3.5.2-1 appropriately includes AMR line items corresponding to items 3.5.1-004 and 3.5.1-035 since the component material and environment for the aging effect exists at MNGP drywell..

PBD Section 4.1 Scope of program states: The above scope is supplemented to address aging management of potential corrosion of inaccessible areas of the drywell shell exterior It also states with regard to evaluation of inaccessible areas based on conditions in accessible areas, that: As part of existing program, these evaluations are performed based on inspection results [Ref 9.8, Section 4.10.20.B.1].

As noted above, the non-applicability claim SLRA Table 3.5-1, AMR item 3.5,1- 004 and 3.5.1-035 is inconsistent with the PBD and SLRA Table 3.5.2-1.

Further, the material and environment for the aging effect exists for the accessible, inaccessible and embedded MNGP drywell surfaces.

Also, SLRA Section 3.5.2.2.1.3.1 does not appear to provide a discussion of whether there has been plant-specific OE of significant corrosion, for applicable components.

covered by FE 3.5.2.2.1.3.1. Also, discuss OE of borated water spills and water ponding, if any, on containment concrete floor and how it is dispositioned on detection.

c) Revise SLRA Table 3.5-1 and FE 3.5.2.2.1.3.1 accordingly.

13 3.5.2.2.1.6, Table 3.5-1, Table 3.5.2-1 3.5-25 &

-26; 3.5-54, 3.5-76 SLRA 3.5.2.2.1.6 states, in part: As summarized in items 3.5.1-10, 3.5.1-038, and 3.5.1-039, cracking due to SCC is an applicable aging effect when stainless steel or nickel alloy components are a) Clarify and correct the inconsistency between SLRA Section 3.5.2.2.1.6 and Table 3.5-1 regarding to non-applicability of item 3.5.1-038.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions exposed to temperature in excess of 140oF.

However, SLRA Table 3.5-1, item 3.5.1-038 states the item is Not Applicable, and accordingly there are no corresponding AMR items in SLRA Table 3.5.2-1. The referenced SLRA 3.5.2.2.1.6 statement appears to be contrary to the determination in Table 3.5.-1 and Table 3.5.2-1 with regard to AMR item 3.5.1-038.

Additionally, SLRA 3.5.2.2.6 on page 3.5-26 states, in part:....the IWE AMP will be enhanced to included one-time volumetric/surface examination of 20 percent of these 24 penetration bellows (i.e., 5 inspections. However, the table on page 3.5-25 only identifies 16 penetrations equipped with SS or nickel alloy bellows.

SLRA Section 3.5.2.2.1.6 does not appear to address plant-specific OE of SCC for the applicable components.

b) Clarify the apparent inconsistency in SLRA 3.5.2.2.1.6 regarding the total number of penetrations that will be sampled from for the one-time inspection to confirm absence of SCC c) Provide a discussion or statement of whether there has been plant-specific OE of SCC for the components covered by the further evaluation.

14 3.5.2.2.1.3.2 &

3.5.2.2.1.3.3 3.5-22 The reference further evaluations do not appear to provide a discussion or statement of plant-specific OE of significant corrosion for the components covered by the further evaluations (FEs).

Provide a discussion or statement of plant-specific OE of significant corrosion for the components covered by the FEs.

10 CFR Part 50, Appendix J Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

EWI-08.06.02 Page 3 GALL-SLR Report Section XI.S4 Element 4-Monitoring and Trending states in part: In the case of Option B, acceptable performance in prior tests meeting leakage rate limits serves as a basis to adjust the testing interval.

Section 4.4 of MNGP LLRT Extended Eligibility Determination report {EWI-08.06-02} states: For a component to be eligible for extended test interval, two a) Clarify if condition 4.4.2 of EWI.08.06.02 is consistent with NEI 94-01 for reestablishing extended testing intervals.

b) Clarify if components MO-2075 and CV-3311 are to be tested at the initial

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions conditions must be satisfied: 4.4.1 - The Component must NOT be identified as ineligible. 4.4.2 - Component must have had recent test results below the admin limit.

NEI 94-01 Rev 3A section 10.2.1.2 states: The test intervals for Type B penetrations may be increased based upon completion of two consecutive periodic as-found Type B tests where results of each test are within a licensees allowable administrative limits. The Corrective Action Section 10.2.1.4 states: Once the cause determination and corrective actions have been completed, acceptable performance may be reestablished and the testing frequency returned to the extended interval in accordance with Section 10.2.1.2.

It is unclear whether Section 4.4 of EWI-08.06-02 provides adequate detail for reestablishing extended interval testing after a failed test.

interval or extended interval as Table 7.2 of EWI-08.06-02 is unclear.

Masonry Wall Program Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

Masonry Walls Basis

Document, Sections 4.1 SLRA Table 3.5.2-5 Basis doc Pg. 7 of 18 SLRA Pg.

3.5-94 Section 4.1 in the basis document (XCELMO00017-REPT-079) identifies masonry walls in several locations for the scope of the AMP, including Reactor Building, Turbine Building, Plant Control and Cable Spreading Structure, Emergency Filtration Train Building, Intake Structure, Emergency Diesel Generator Building/Diesel Oil Transfer House, Off Gas Stack, Radwaste Building, Substation Yard, and Miscellaneous Structures Inside the Protected Area.

SLRA Table 2 lists the items associated with the Masonry Walls AMP. These corresponding Table 2 AMR line items are under Table 3.5.2-4 Emergency Diesel Please identify the corresponding AMR line items for the scope of Masonry Walls AMP in SLRA Table 3.5.2-5 Emergency Filtration Train Building. Update SLRA Table 3.5.2-5 accordingly if necessary.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Generator Building, Table 3.5.2-6 Fire Protection Barriers Commodity Group, Table 3.5.2-9 Intake Structures, Table 3.5.2-10 Miscellaneous Station Blackout Yard Structures, Table 3.5.2-11 Off-Gas Stack, Table 3.5.2-13: Plant Control and Cable Spreading Structure, Table 3.5.2-14 Radioactive Waste Building, Table 3.5.2-15 Reactor Building Table, and 3.5.2-17 Turbine Building.

However, the staff did not find any line items in Table 3.5.2-5, Emergency Filtration Train Building, citing Masonry Walls AMP to manage aging effects.

2 Masonry Walls Basis

Document, Sections 4.3 SLRA Section B.2.3.32 Basis doc Pg. 8 of 18 SLRA Page B-236 and A-90 GALL-SLR XI.S5, Masonry Walls, element 3, notes that the mortar joints and gaps between the supports and masonry walls should be monitored.

Section 4.3 in the basis document lists the parameters to be monitored or inspected under the AMP, including monitoring and inspecting of gaps between the supports and masonry walls as an enhancement to MNGP procedure, but does not mention monitoring of mortar joints.

Although SLRA Section B.2.3.32 states that the AMP will be enhanced to monitor and inspect for loss of material at the mortar joints and gaps between supports and masonry walls (page B-235), monitoring of material loss at mortar joints is not included in its enhancement list (Page B-236) and Commitment Table A-3 (Commitment

1. 35 on page A-90).

Please explain whether monitoring of mortar joints is a parameter that needs to be monitored or inspected under the Masonry Walls AMP. Enhance the procedure if necessary and update SLRA accordingly.

3 Masonry Walls Basis

Document, Sections 4.5 Basis doc Pg. 9 of 18 GALL-SLR XI.S5, Masonry Walls, element 5, notes that inspection results are documented and compared to previous inspections to identify changes or trends in the condition of masonry walls.

Section 4.5 of the base document, Monitoring and Trending, does not explicitly mention comparing Please indicate where in the implementation procedure a comparison of inspection results with previous inspections is included. If not, update SLRA accordingly to include an enhancement to the AMP.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions inspection results to previous inspections.

The staff browsed Procedure 1385, Periodic Structural Inspections (Rev. 17), and did not note that inspection results should be compared to previous inspections in the procedure.

4 Masonry Walls Basis

Document, Sections 4.7 Basis doc Pg. 10 of 18 GALL-SLR XI.S5, Masonry Walls, element 7, states A corrective action option is to develop a new analysis or evaluation basis that accounts for the degraded condition of the wall (i.e., acceptance by further evaluation).

It is unclear to staff whether section 4.7 of the base document describes acceptance by further evaluation as a corrective action option, and how that is discussed in the implementation procedure.

Please indicate that acceptance by further evaluation is a corrective action option in the implementation process.

Corrosion-Structural Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

SLRA Table 3.5-1 Pages 3.5-74 and 3.5-75 SRP-SLR Table 3.5-1 Item 098 notes that stainless steel and aluminum alloy support components exposed to air with borated water leakage do not have any aging effects that require management.

SLRA Table 1 Item 3.5.1-98 identifies 3.5.1-98 as not used, and states its component, material, and environment combination is addressed by item number 3.5.1-099.

SLRA Table 1 Item 3.5.1-99 credits the One-Time Inspection (B.2.3.20) AMP or the ASME Section XI, Subsection IWF (B.2.3.30) AMP to manage loss of material of ASME Class 1, 2, 3, and MC aluminum and stainless steel supports exposed to air and condensation.

Please clarify if any 3.5.1-98 associated GALL-SLR items (III.B1.1.TP-4, III.B1.2.TP-4, III.B1.3.TP-4, III.B2.TP-4, III.B3.TP-4, III.B4.TP-4, and/or III.B5.TP-

4) are used at MNGP.

If using SLRA AMR Item 3.5.1-099 as an alternative, please describe why the One-Time Inspection AMP or IWF AMP is needed and how the AMP is involved in addressing item 3.5.1-098 for component, material, and environmental combinations.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions It is not clear to staff why SLRA AMR item 3.5.1-98 needs to be addressed by item 3.5.1-099, given that no aging needs to be managed for item 3.5.1-98.

The staff also searched SLRA Table 3.5.2 and did not find any associated GALL-SLR items (III.B1.1.TP-4, III.B1.2.TP-4, III.B1.3.TP-4, III.B2.TP-4, III.B3.TP-4, III.B4.TP-4, and/or III.B5.TP-4).

2.3.3.6 Emergency Diesel Generators Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

2.3.3.6 2.3-35 Scoping/Screening Boundary Drawing SLR-36051 Diesel Oil System (sheet 1 of 2) [M-133];

• Coordinate E-5: the piping from valve FO-7-2 to the Intake Structure wall is designated as not Subject to Aging Management Review. SLRA Section 2.1.4.2.1 Non-Safety Related SSCs with Potential to Prevent Satisfactory accomplishment of Safety Functions appears to be of relevance for this pipe configuration.

Is there no a(2) Spatial/ Structural requirement for this pipe and valves FO-7-3 and FO-7-4? Please provide the basis for not being subject to Aging Management Review.

2 2.3.3.6 2.3-36 Scoping/Screening Boundary Drawing SLR-36051-1 Diesel Oil System (sheet 2 of 2) [M-133];

• Coordinate A-6: Components Diesel Oil Receiving Tank T-83; Pump P-92; and tank influent and effluent piping.

These components are color coded black as not subject to Aging Management Review. These components directly support the REQUIRED ACTION A.1 for LCO 3.8.3 Condition A which reads Fuel oil level < 7-day supply and

> 6-day supply in storage tank. with a COMPLETION TIME of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

10CFR50 Appendix A Criterion 17 reads in part (emphasis

• Please provide the basis for not identifying Components Diesel Oil Receiving Tank T-83; Pump P-92; and tank influent and effluent piping red as a(2) Functional and not subjecting these components to aging management review.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions added):

Electric power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions. A switchyard common to both circuits is acceptable. Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, containment integrity, and other vital safety functions are maintained.

Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies.

USAR Section 8.4.1 Safeguards Emergency Diesel Generator (EDG) Systems Design Bases reads in part:

g. Each EDG shall have local fuel tanks (day tank and base tank) fed from a common diesel oil storage tank. The local tanks shall have sufficient capacity for a minimum of eight hours of full power operation of their respective unit.

The diesel oil storage tank shall provide sufficient fuel to the EDGs for at least one week of full load operation of one unit.

Also: Please to the Refer to Risk Insights for Evaluating the Monticello SLRA (Proprietary; Non-Public)

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The TS Bases for SR 3.8.3.3 reads (emphasis added):

The tests of new fuel oil prior to addition to the storage tank are a means of determining whether new fuel oil is of the appropriate grade and has not been contaminated with substances that would have an immediate detrimental impact on diesel engine combustion. If results from these tests are within acceptable limits, the fuel oil may be added to the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tank. These tests are to be conducted prior to adding the new fuel that is in the diesel oil receiving tank to the storage tank.

It is not obvious from review of the original LRA whether these components were identified as subject to AMR. The Diesel Oil Receiving Tank is neither identified/discussed in XCELMO00017-REPT-037 Revision 1 nor in DBD-B.09.08 Revision 82 on the Centrec Portal.

It appears that these components should be color coded red as a(2) Functional and subject to aging management review.

3 2.3.3.6 2.3-35 Scoping/Screening Boundary Drawing SLR-36051 Diesel Oil System (sheet 1 of 2) [M-133];

• This boundary drawing does not display an EDG lube oil transfer subsystem.

Technical Specification Bases B 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air reads in part:

SR 3.8.3.2 This Surveillance ensures that sufficient lubricating oil inventory is available to support at least 7 days of full load operation for each EDG. The lube oil volume equivalent to (a) Does each EDG base assembly contain its own 7-day (i.e., 165 gallons) supply of lubricating oil?

(b) Is there a separate lube oil transfer system from a lube oil storage location to each of the EDGs?

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions a 7-day supply is 165 gallons and is based on the EDG manufacturer's consumption values for the run time of the EDG. Implicit in this SR is the requirement to verify the capability to transfer the lube oil from its storage location to the EDG, if the EDG lube oil sump does not hold adequate inventory for 7 days of full load operation without the level reaching the manufacturer's recommended minimum level.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

2.3.3.8 Emergency Service Water Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

2.3.3.8 2.3-41 Scoping/Screening Boundary Drawing SLR-36664 RHR Service Water & Emergency Service Water Systems:

• Coordinate D-6: Component Type and Function of See Note 2 for green box at Valve RHRSW-68. Subsequent License Renewal Note 2 does not apply. Drawing Note 2 would suggest an Appendix R a(3) red Function.

Staff requests clarification for the meaning of See Note 2 for green box at Valve RHRSW-68.

2 2.3.3.8 2.3-41 Scoping/Screening Boundary Drawing SLR-36246 Residual Heat Residual Heat Removal System:

Coordinate A-5 Backup Safety Related Instrument air supply: Note 2 at Filter neither correlates to the Drawing Note #2 nor the Subsequent License Renewal Notes.

• Staff requests clarification for the correct interpretation of Note 2 3

3.3.2.1.8 Table 3.3.2-8

• The Component Types Bolting (Closure) [page 576/1519] and Piping, Piping Components [pages 582/1519 & 584/1519] contain line items with an Environment of Soil (External). From review of the ESW SLR drawings it is not apparent where this environment exists for the subject Component Types.

Staff requests confirmation that these Component Types buried in soil reflects actual plant conditions. In particular, where on the SLRA boundary drawings are the buried piping and bolted connections displayed?

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Neither review of XCELMO00017-REPT-026, Revision 1 nor review of DBD-B-08.01.04, Revision 6 on the Certrec e-portal supports this type of Environment. Neither review of USAR Section 10.4.2 nor Section 10.4.4 explains this type of Environment. It is noted that these Component Types buried in soil is Consistent with LRA Table 3.3.2-8.

2.3.3.14 Reactor Building Closed Cooling Water Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

2.3.3.14 2.3-57 Scoping/Screening Boundary Drawing SLR-36042 Reactor Building Cooling Water System [M-111]:

• Component Types Radiation Monitor Well? (Coord. A-5, RW-17-302) Thermowell (Coord. C-7, TE-1720) and Flow Element (Coord A-5, FE-4145) are displayed on this drawing.

a) Staff requests confirmation that the Component Types are accurately interpreted.

b) Staff request confirmation that these component types are within the domain of Piping Elements and/or Piping, Piping Components listed in Table 2.3.3-14 subject to AMR.

Staff requests confirmation that the instrumentation associated Radiation Monitor Well does not have an a(2)

Functional (i.e., red) Intended Function.

2 2.3.3.14 2.3-57 Scoping/Screening Boundary Drawing SLR-36042-2 Reactor Building Cooling Water System [M-111-1]:

a(2) Spatial/Structural: Piping inside Containment from Penetrations X-24 & X-23. The structural supports inside containment associated with this piping versus the leakage boundary spatial concerns are two unique concerns and almost always are not mutually exclusive.

However, there can be exceptions.

• Is the scoping methodology of 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 satisfied for Containment from Penetrations X-24, & X-23?

2.3.3.15 Reactor Water Cleanup

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

2.3.3.15 2.3-60 Scoping/Screening Boundary Drawing SLR-36254 Reactor Water Cleanup System (a) Coordinate C-6 displays Flow Element FE-6016 as a(1) or as A(2) Functional in red for RWCU isolation based on excessive flow into secondary containment -

there is no flow element listed in SLRA Table 2.3.3-15.

(b) Note 2 on drawing SLR-36254 Coord A-5; reads The in scope segment of these lines terminates as they enter the 985 Pump Room does not contain safety related equipment as confirmed by walkdown documented in ML0630504140.

(c) Coordinate A-2; Temperature limits established for TIS-157 and TE 12-97 (thermo well?) for pump discharge (i.e., upstream) of the demineralizers, prevent damage to the ion exchange resin beds by ensuring temperature limits are not exceeded.

(a) Is the fact that the flow element is a non-class 1 component per the commodity group guidance of SLRA section 4.3.6 ASME Section III, Class 2 and 3 and ANSI B31.1 (page 1056/1519) why the component type of piping elements is not listed in Table 2.3.3-15.

(b) 3 ED See Note 2 is located in the vicinity of a pipe with a reducer and blank flange? -. Is not 985 Pump Room with Clean-up Filter / Demin(s) T-202A &

B contained in the boxed area directly above the note? What is the correct interpretation of this Note?

TIS12-157 & TE 12-97 satisfy an a(2)

Functional need. Please justify why the Piping Element (thermowell?) is not identified as A(2) Functional and subject to Aging Management Review.

2 2.3.3.15 2.3-60 Scoping/Screening Boundary Drawing SLR-36255 Filter/Demineralizer System:

(a) Coordinates D-3 and D-5: Y-strainers (YS12-54A &

54B) prevent Clean-up Filter / Demin resins beads from entering the Reactor Coolant System (b) Table 2.3.3-15 contains a line item for neither the Component Type Strainer nor Piping Element with an Intended Function of Filter consistent with the guidance of NUREG-2192 Table 2.1-5 Typical Passive (a) Differential Switches dPIS 12-4-72A/B are color coded green A(2)

Structural/Spatial; Is not a(2) Functional red more accurate?

(b) Assuming a(2) Functional is appropriate, then provide the basis for not including the strainer element with a filter function subject to AMR?

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Component-Intended Functions.

(c) Instruments dPT-12-4-69A/B at Coordinates D-5 & D-2 monitor differential pressure, to ensure the design limits on filter/

demineralizer septums are not exceeded.

(c) Differential pressure monitoring is included as an A(2) Structural /Spatial interaction - please justify A(2)

Structural/Spatial (green) versus. A(2)

Functional (red) 3 2.3.3.15 2.3-59 Scoping/Screening Boundary Drawing SLR-36255 Filter/Demineralizer System:

a) RWCU Instrumentation and tubing throughout drawing is color coded as green indicating an A(2)

Spatial/Structural function. However, the piping and components that these instruments are attached to are color coded as black (i.e., out of scope).

Why are these instruments WSSLR (i.e.,

Within Scope of Subsequent License Renewal)? spatial interaction appears to be the most logical reason OR are these instruments under the EQ program? - if so, they should be color coded as a(3) (red) per the Legend.

2.3.3.16 Service and Seal Water Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

2.3.3.16 2.3-61 Scoping/Screening Boundary Drawing SLR-36041 Service Water System

• a(2) vs a(2) Functional: Coord. D-2 & E-2 Components from V-EAC-14A/B to the Circ Water Discharge Pipe (within the EFT Building and Turbine Building) - These ESW components (piping and valves) have an a(2)

Functional function per Note 3.

Why not color coded red for (a)(2)

Functional instead of a(2)

Spatial/Structural?

2 3.3.2.1.16 Service and Seal Water 3.4-108 through 3.4-111 SLRA Table 3.3.2.-16 Service and Seal Water -

Summary of Aging Management Evaluation:

For the valves and piping subject to aging management review due to a(2) Spatial/Structural as displayed on SLR-36665-3 Biocide Injection System, the internal environment consists of Biocide and Dispersant. SLRA Table 3.3.2.-16 has no internal environment for valves and piping to address these unique internal environments.

SLRA Section 3.3.2.1.16 contains the following:

The SSW System components are exposed to the following environments:

• Air - Indoor Uncontrolled

• Condensation

• Lubricating Oil

• Raw Water

• Soil

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Please provide the basis for not including an internal environment that reflects an internal environment of Biocide and Dispersant for valves and piping 2.3.4.6 Turbine Generator Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

2.3.4.6 2.3-80 Scoping/Screening Boundary Drawing SLR-36033 Main Steam [M-102]:

a) Coords. B-1/B Expansion Joints XJ-1271/-1272/-

1273/-1274 on this drawing are color coded as a(2)

Spatial/Structural. There is no Component Type for Expansion Joints in Table 2.3.4-6 Turbine Generator System Components Subject to Aging Management Review. These expansion joints are located external to Main Condenser Nozzles/penetrations 2 & 2A. These expansion joints are not addressed in XCELMO00017-REPT-032, Revision 0 b) Coord. C-1; Piping configuration from Main Condenser penetrations #47 and #37 to Turbine Vacuum Trip No. 1 and No. 2 - It is not clear whether these components being scoped as (2) Spatial/Structural are within the TGS or CDR system.

c) Coord. A-2: Subsequent License Renewal Notes: 2.

Components Shown in Red for the Turbine Generator System (TGS) are in Scope for a (A)2 Functional function in Support of Alternate Source Term. It is not apparent from review of SLR-36033 which components are of relevance to this Note.

a) Please provide the basis for not listing this Component Type in Table 2.3.4-6.

b) Since these two one inch lines are integral with the overall integrity of the Condenser vacuum and its plate out AST function, should not these lines be color coded as a(2) Functional?

c) Please provide clarification of which TGS components are of relevance to Note 2.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 2

2.3.4.6 2.3-80 Scoping/Screening Boundary Drawing SLR-36034 Turbine & Extraction Steam Sheet 1 of 2 [M-103]:

• Coord. E-3 Note 3 for Pipe OG4-16 not defined on drawing; Coord. D-3 Note 2 and Note 3 for lines V30 HB and V21-8-HB not defined on drawing. These notes Appear to have an a(2) Functional intended function.

Please provide the staff with clarification of these Notes.

3 2.3.4.6 2.3-80 Scoping/Screening Boundary Drawing SLR-36050 Turbine Oil System [M-132]:

(a) Component Type @ Coord. D-5 Vapor Extractor casing not listed in Table 2.3.4-6. Please provide basis for not listing.

(b) Table 2.3.4-6 Component Type Pump Casing (EPR Oil Pump) - EPR is not defined in the SLRA. Which pump is being implied (i.e., as displayed in the Table) on the Section 2.3.4.6 SLRA Boundary Drawings? EPR is neither defined in XCELMO00017-REPT-032, Revision 0 nor in USAR Section 11.2 Turbine-Generator System.

Please Clarify.

(c) Coordinate A-5 T-41A H2 Seal Oil Conditioner Drain Tank 22 Gal Cap Table 2.3.4-6 does not contain a Component Type line item that matches this description; Please clarify.

(d) Coord. B-4 Lube Oil Purifier Pump casing P-8A; Which Component Type line item on Table 2.3.4-6 represents this pump casing? Please Clarify.

(e) Coord. C-2 Clean Oil Overflow between the Clean Oil Storage Tank and the Dirty Lube Oil Storage Tank; Table 2.3.4-6 contains Component Type line item Tanks (Lubricating Oil Dump Over Flow Tanks) - It is not apparent from review of the SLR Boundary Drawings to

• Staff requests a clarification discussion with Licensee of Background/Issues (a) through (f).

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions which tank this line item represents. Please Clarify.

(f) Table 2.3.4-6 Component Type Pump Casing (Turbine Bearing Lift Pump) - Upon review of the SLR-Boundary Drawings the configuration of these lift pumps is not represented. Are not these pumps (casings) located at each bearing of the Turbines and contained within a separate enclosure that would self-contain any leakage from the pump casing and direct that leakage back to the main turbine oil reservoir? Please Clarify.

4 2.3.4.6 2.3-80 Scoping/Screening Boundary Drawing SLR-M8107L-087 Turbine Lube Oil Bearing [M-132]:

Coordinate C-4 reads: CONFIGURATION DURING FLUSHING WITH BOOSTER PUMP REMOVED The staff notes that it is Important to note that this Drawing does not represent normal plant operations. The Booster Pump would have been removed to support the original Turbine Oil Flush during plant construction and startup & rarely (if ever) thereafter.

(a) It is not understood how oil supply lines encased within the large trunk drain line sloping back to the oil reservoir represent an a(2) Spatial/Structural concern. Please discuss with staff.

(b) The Turning Gear Oil Pump casing is within the Oil Tank (i.e., Turbine Generator Oil Reservoir). How does the casing represent an a(2) Spatial/Structural concern?

Please discuss with staff.

(c) The DC powered Pump Casing (Emergency Bearing Oil Pump) listed on Table 2.3.4-6 is located within the Oil Tank (i.e., Turbine Generator Oil Reservoir). How does

• Staff requests a clarification discussion with Licensee of Background/Issues (a) through (e).

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions the casing represent an a(2) Spatial/Structural concern?

Please discuss with staff.

(d) Coord. B-5: Booster Pump casing is not listed on Table 2.3.4-6. Please establish basis for not including or including within the scope of components subject to AMR.

Please discuss with staff.

(e) Table 2.3.4-6 contains Component Type Pump Casing (Turb Aux Oil Pump). Does this line item represent the AC MSP pump casing @ Coord. B-4?

• If so, please establish basis for not including or including within the scope of components subject to AMR.

Please discuss with staff.

3.3 Aging Management of Auxiliary Systems Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

Section 3.3 Table 3.4-1 3.4-38 AMR Item number 3.4.1-106 is copper alloy (with >15%

Zn or >8% Al) piping and piping components exposed to air and condensation, which is susceptible to stress corrosion cracking and managed by the External Surfaces Monitoring of Mechanical Components Program. The discussion section of Table 3.4-1 states that this line item is also applied to heat exchanger components made of the same materials that are exposed to condensation, uncontrolled indoor air, and outdoor air.

The staff would like to discuss the physical configuration of all heat exchanger tubes managed by this line item.

Specifically, since the Ext. Surfaces Program requires a minimum 20% of the surface area to be examined, discuss how the visual exam will meet the 20%

minimum surface area if the heat exchanger tubes in question are finned.

Water Chemistry

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

N/A N/A Appendix D.5 of BWRVIP-190 indicates that, ((

)) Note 7 of table 2-12 of the 2019 interim guidance of BWRVIP-190 also states, ((

)) It is unclear to the staff how guidance from Appendix D.5 and how note 7 of table 2-12 is incorporated into the Water Chemistry Program at MNGP.

• Please discuss how guidance from Appendix D.5 is incorporated into the Water Chemistry program at MNGP.

Please discuss how note 7 from table 2-12 of the 2019 interim guidance is incorporated into the Water Chemistry Program at MNGP.

2 N/A N/A In the response to question 4 of the initial set of breakout questions notes 3,4 and 7 from table 2-12 of 2019 BWRVIP-190 interim guidance are highlighted in response to how Feedwater Hydrogen Concentration is monitored.

However, it is unclear to the staff how these notes are applied to how feedwater hydrogen concentration is measured at Monticello Nuclear Generating Plant (MNGP).

• Please discuss how the notes from table 2-12 are applied in the MNGP approach to measuring feedwater hydrogen concentration.

3 N/A N/A Question 3 of the initial breakout questions asks how the limit for reactor water insoluble iron was developed because it is not listed as a diagnostic parameter for reactor water at power operation conditions in BWRVIP-190. The response provided links the development of this limit to feedwater iron and the staff is unclear about this connection and how the limit for insoluble iron was developed.

• Please step through how the guidance informed the limit for reactor water insoluble iron at power operation conditions.

• Please clarify how this limit is related to feedwater iron.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 4

N/A N/A In the initial response to breakout question 5 it was indicated that coupons in the MMS system would be used to measure catalyst loading. ((

)) It is unclear to the staff how MMS availability is tracked and incorporated into plant chemistry implementing documents.

Please clarify how MMS availability is tracked and incorporated into plant chemistry implementing documents.

5 N/A N/A Question 6 of the initial breakout questions highlighted a difference in the sampling frequency for HWC availability between II.05 (continuously) and II.01 (daily). The initial response states that the parameters are measured continuously but the calculation is done daily. However, the staff is unclear about the HWC availability frequency as stated in II.05.

Please clarify the continuous measurement of HWC availability as listed in table 3.16.1 in II.05.

6 N/A N/A Question 8 of the initial breakout questions highlights carbon steel piping in a sodium pentaborate solution environment subject to loss of material. In the initial response it was indicated that one time inspection would be used, and water chemistry would not be adjusted to manage the aging effects. It is unclear to the staff how the Water Chemistry AMP will be used to manage loss of material for these components.

Please clarify how the water chemistry AMP is being used to manage loss of material for these components given that the chemistry is not specified or adjusted to prevent loss of material for carbon steel.

4.6.1 Fatigue of Cranes Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

SLR Application:

Subsection 4.6.1 Fatigue of Page 4.6-2 and Reactor Building and Turbine Building Cranes load cycle limits were projected through the SPEO in accordance with 10 CFR 54.21(c)(1)(i) in the application.

Describe this TLAA dispositioning.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Cranes.

and Report:

XCELMO00017-REPT-091, Rev. 0 Page 14 of 15 However, Reactor Building and Turbine Building Cranes load cycle limits were projected through the SPEO in accordance with 10 CFR 54.21(c)(1)(ii) in the report.

B.2.3.34 Inspection of Water-Control Structures Associated with Nuclear Power Plants 2.4.9 Intake Structure Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.3.34 2.4.9 2.3.3.3 Table 2.2-1 B-243 2.4-20 2.3-28 2.2-2

1. GALL-SLR XI.S7 states that the scope of program includes sluice gates and trash racks. The staff could not locate any information of the sluice gates and trash racks in SLRA.

2. SLRA Section 2.4.9 states that the Intake Structures consists of an inlet channel open to the river, an uncovered, reinforced concrete forebay and a reinforced concrete chambered structure that encloses traveling screens, various pumps and water passages. SLRA Section 2.4.9 also states that in addition to the INS itself, this structure also covers the access tunnel and Diesel Fire Pump House.

SLRA Section B.2.3.34 states that Intake Structure is a only structure for water-control structure.

It is unclear to the staff what water-control structures are included in the scope of SLR It is also unclear to the staff whether the traveling screens are subject to the aging management.

Scope of program:

1. Clarify whether MNGP has sluice gates and trash racks subject to aging management within the scope of SLR.

2.(a) Clarify what water-control structures (Intake Structure, Inlet Channel, Access Tunnel, Diesel Fire Pump House, East and West Service Bay, Travel Screens, etc.?) are included in the scope of SLR; 2.(b) Clarify whether Inlet Channel is subject to aging management within the scope of SLR; and 2.(c) Clarify whether traveling screens are subject to aging management.

3. Explain why Discharge Structure is not within the scope of SLR.

4. Update SLRA accordingly based on the responses above

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions

3. GALL-SLR XI.S7 includes discharge structure within the scope of SLR. SLRA Section 2.3.3.3 states that the CWT flows through the Discharge Structure to an open canal, which conveys it to the river downstream of the intake during open cycle operation. SLRA Table 2.2-1 states that Discharge Structure is not within the scope of SLR.

It is unclear to the staff whether the Discharge Structure is subject to aging management within the scope of SLR.

2 B.2.3.34 2.4.9 B-243 2.4-20

1. SRP-SLR report describes the indication of Alkali-Silica Reactions (ASR) as map or patterned cracking, alkali-silica gel exudations, surface staining, expansion causing structural deformation, relative movement or displacement, or misalignment/distortion of attached components. AMP basis document lists parameters monitored or inspected as unique map or cracking that would indicate the presence of ASR. It appears that ASR has more characteristics than ones described in the AMP basis document.

2. AMP basis document lists parameters monitored or inspected for Earthen Embankment Structures, Dams, and Canals. However, SLRA Section B.2.3.34 states that Intake Structure is a only structure for water-control structure. AMR 3.5.1-058 in SLRA Table 3.5-1 states that earthen water control structures, dams, embankments, reservoirs, channels, and ponds are not credited at MNGP.

It appears there is inconsistency between SLRA and AMP basis document.

3. GALL-SLR XI.S7 states that parameters monitored or inspected for channels and canals include erosion or Parameters Monitored or Inspected:

1. Evaluate the parameters monitored or inspected for the ASR, and provide an enhancement if necessary.

2. Clarify whether Earthen Embankment Structures, Dams, and Canals are applicable, and address the inconsistency between SLRA and AMP basis document.

3. If Inlet Channel is in the scope, provide parameters monitored or inspected for the Inlet Channel, and evaluate whether it will be consistent with non-applicability claim of AMR 3.5.1-058.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions degradation that may impose constraints on the function of the cooling system and present a potential hazard to the safety of the plant. SLRA Section 2.4.9 states the boundary for the INS includes the structure itself (including the forebay and inlet channel) and the access tunnel connecting this structure with the Turbine Building and the Diesel Fire Pump House. It appears that AMP basis document does not list parameters monitored or inspected for the channel.

3 N/A N/A

1. GALL-SLR XI.S7 states that the program includes provisions for increased inspection frequency based on an evaluation of the observed degradation.

The staff could not find any provisions for increased inspection frequency in the AMP basis document.

The staff noted that an enhancement to the Structures Monitoring program includes provisions for increased inspection frequency based on an evaluation of the observed degradation.

2. AMP basis document states that the intake structure is inspected below the waterline using divers.

The staff could not locate the procedure how divers inspect the Intake Structure below the waterline.

Detection of Aging Effects:

Inspected:

1. Evaluate whether an enhancement is needed to include provisions for increased inspection frequency based on an evaluation of the observed degradation.

2. Is there a procedure for the divers how to inspect the Intake Structure below the waterline? If not, explain it.

4 Table 2.4-9 2.4-23

1. GALL-SLR XI.S7 states that degradation of piles and sheeting are accepted by engineering evaluation or subject to corrective actions.

AMP basis document does not describe acceptance criteria for sheet piles listed in SLRA Table 2.4-9. The staff could not locate OE related to these sheet piles.

2. AMP basis document states that quantitative second tier acceptance criteria are used to evaluate the need for Acceptance Criteria:

1. Describe the degradation of sheet piles if present, and provide acceptance criteria for sheet piles.

2. Clarify whether an enhancement to the Inspection of Water-Control Structures AMP is needed to include this acceptance criteria.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions corrective action of an observed degradation. The criteria, which are identified in implementing procedure 1385, are based on industry guidelines/criteria such as ACI 349.3R-96 and ACI 201.1.

The staff could not find the above mentioned acceptance criteria in procedure 1385. The staff also noted that an enhancement to the Structures Monitoring program includes this acceptance criteria.

5 Table XI-01 A.2.2.34 XI 01-33 A-30 GALL-SLR Table XI-01 XI.S7 FSAR Supplement states that the program also includes structural steel and structural bolting associated with water-control structures, which is missed in SLRA Appendix A.2.2.34.

Appendix A - USAR Supplement:

Provide the missing information of USAR in Appendix A.2.2.34 to be consistent with GALL-SLR recommendations.

6 Table 3.5.2-9 3.5-110 AMR item 3.5.1-096 in SLRA Table 3.5.2-9 is corresponding to GALL-SLR item III.A6.TP-34, which refers to AMR item 3.5-1, 071 in GALL-SLR report.

Correct the error of GALL-SLR item reference.

3.5.2.2.2 Non-Containment Plant Structures Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

3.5.2.2.1.2 3.5-20 The SRP-SLR Section 3.5.3.2.1.2 guidance states that the reviewer ensures that the aging effects associated with the cooling system are being properly managed or temperatures are being monitored to identify a problem with the cooling system If active cooling is relied upon to maintain acceptable temperatures.

SLRA Section 3.5.2.2.1.2 states that the bulk drywell temperature is maintained by the primary containment ventilating and cooling system. It is not clear to the staff how they are properly managed or monitored.

SLRA Section 3.5.2.2.1.2 states that local area temperature in the biological shield wall due to hot reactor REC System penetrations is calculated at 179°F; less

1. Explain how the aging effects associated with the primary containment ventilating and cooling system are being properly managed or temperatures are being monitored to identify a problem with the primary containment ventilating cooling system.

2. Provide calculations of temperature at penetrations of biological shield wall for the staff to review.

3. Clarify how local temperature at penetration of the biological shield wall is adequately managed.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions than the concrete degradation threshold of 200°F. It also states that thermal insulation is credited with maintaining the temperatures in the bioshield wall below 200°F.

It is unclear to the staff how the applicant calculates the local area temperature in the biological shield wall due to hot reactor REC System penetrations.

2 3.5.2.2.2.1 3.5.2.2.2.3 3.5-28 3.5-33 SRP-SLR Sections 3.5.3.2.2.1, item 2 and 3.5.3.2.2.3, item 2 describe the indication of Alkali-Silica Reactions (ASR) as map or patterned cracking, alkali-silica gel exudations, surface staining, expansion causing structural deformation, relative movement or displacement, or misalignment/distortion of attached components.

SLRA Sections 3.5.2.2.2.1, item 2 and 3.5.2.2.2.3, item 2 state that the MNGP Structures Monitoring AMP has been refined, based on industry/fleet information, to include visual examination for unique map or cracking that would be indicative of reaction with aggregates, such as alkali-silica reaction (ASR).

It appears that ASR has more characteristics than ones described in SLRA.

SLR-SLR Section 3.5.3.2.2.3, item 2 states cracking due to expansion from reaction with aggregates could occur in inaccessible concrete areas of Group 6 structures.

SLRA Section 3.5.2.2.2.3, item 2 does not address how aging effects of inaccessible concrete areas of Group 6 structures will be adequately managed.

1. Evaluate the parameters monitored or inspected for the ASR, and provide an enhancement to the Structures Monitoring program regarding the detection of ASR.

2. Explain in SLRA Section 3.5.2.2.2.3, item 2 how aging effects of inaccessible concrete areas of Group 6 structures will be adequately managed.

3 3.5.2.2.2.1 Table 3.5-1 3.5-29 3.5-58 AMR 3.5-1, 046 is related to aging effect of reduction of foundation strength and cracking due to differential settlement and erosion of porous concrete sub-foundation. The applicant claims AMR item 3.5-1, 046 to be not applicable. This aging effect is related to a).

1. Evaluate the claim of non-applicability of AMR item 3.5-1, 046, and provide table 2 items.

2. Describe OE related to the settlements

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions differential settlement, b) porous sub-foundation. This aging effect exists if either or both conditions present. The differential settlement of the Diesel Fuel Oil Transfer House is an example. Therefore, it is applicable.

SLRA Section 3.5.2.2.2.1, item 3, states that with the exception of the Diesel Fuel Oil Transfer House and Off-Gas Storage Building HTV exhaust pipe, no significant settlement has been observed on any major structure and de-watering systems are not used. It also states that with the exception of the Diesel Fuel Oil Transfer House, cracks, distortion, and increase in component stress levels due to settlement do not require aging management, which is not correct since they are applicable aging effects, and require aging management by the Structures Monitoring program.

SLRA Section 3.5.2.2.2.1, item 3, indirectly acknowledged significant settlements of the Diesel Fuel Oil Transfer House and Off-Gas Storage Building HTV exhaust pipe, but does not include the information of settlements and their aging management.

The procedure 1396 provides the instructions and baselines to check for settling of the Diesel Oil Transfer House, T-44 (Diesel Oil Storage Tank), and the Offgas Storage Building HTV exhaust pipe.

of the Diesel Fuel Oil Transfer House, Diesel Oil Storage Tank and and the Offgas Storage Building HTV exhaust pipe, and explain how their aging effects due to settlement are adequately managed during the SPEO.

3. Evaluate whether plant-specific program is needed.

4. Update SLRA accordingly based on the responses above.

Note: More settlement questions are documented in the breakout questions in TRP 46, Structures Monitoring program.

4 3.5.2.2.2.1 3.5-30 SRP-SLR Section 3.5.3.2.2.1, item 4 guidance states that a plant-specific AMP is not required for the reinforced concrete exposed to flowing water if evaluation determined that the observed leaching of calcium hydroxide and carbonation in accessible areas has no impact on the intended function of the concrete structure.

FE Section 3.5.2.2.2.1, item 4 states MNGP OE does not indicate leaching has been observed on accessible

1. Provide ARs and WOs on the Portal, and describe OE related to leaching for accessible concrete areas of Groups 1-5 and 7-9 structures at site.

2. Evaluate whether the observed leaching of calcium hydroxide and carbonation in accessible areas has an

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions concrete areas that would impact intended functions of the structure, but there is no description of the related OE. The staff also could not find the OE related to leaching during the in-office OE audit.

impact on the intended function of the concrete structure.

5 3.5.2.2.2.2 3.5-31 SPR-SLR Section 3.5.3.2.2.2 states a plant-specific evaluation should be performed if any portion of the concrete Groups 1-5 structures exceeds specified temperature limits (i.e., general temperature greater than 150 °F and local area temperature greater than 200 °F).

FE Section 3.5.2.2.2 is related to the reduction of strength and modulus of concrete due to elevated temperatures in Group 1-5 concrete structures. High temperatures in Drywell general area and biological shield wall piping penetration local areas discussed in this FE Section belong to the FE Section 3.5.2.2.1.2 for the aging effect in the containment. The staff is unclear whether concrete elements in Groups 1-5 concrete structures are subject to elevated temperatures in excess of 150°F general area and 200°F local area.

1. Explain whether there are elevated temperatures in excess of 150°F general area in Groups 1-5 concrete structures.

2. Explain whether there are hot pipes causing elevated concrete temperatures at penetrations in excess of 200°F in Groups 1-5 concrete structures.

3. Explain how the reduction of strength and modulus of concrete due to elevated temperatures is adequately managed.

4. Evaluate whether a plant-specific AMP is needed.

6 3.5.2.2.2.3 3.5-31 SRP-SLR Section 3.5.3.2.2.3, item 1 guidance states a plant-specific program is not necessary if the concrete was constructed with air content of 3 to 8 percent.

However, FE Section 3.5.2.2.2.3.1 does not include the air content used for the concrete components in Group 6 Structures, and OE related to aging effect of loss of material (spalling, scaling) and cracking due to freeze-thaw.

CAP 501000022600 on Page 17 of 81 states that the primary cause for the observed concrete matrix deterioration appeared to be a result of salt-infused water infiltration into the intake roof concrete aggravated by freeze-thaw cycling.

1. Clarify the air content used for components in Group 6 structures.

2. Discuss the Intake Structure OE related to freeze-thaw, and the corrective actions taken.

3. Evaluated whether a plant-specific AMP is needed.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 7

3.5.2.2.2.3 3.5-33 SRP-SLR Section 3.5.3.2.2.3, item 3 guidance states that a plant-specific AMP is not required for the reinforced concrete exposed to flowing water if evaluation determined that the observed leaching of calcium hydroxide and carbonation in accessible areas has no impact on the intended function of the concrete structure.

During the on-site audit, the staff noted the leaching on interior wall and ceiling of intake structure. FE Section 3.5.2.2.2.3.3 does not include the discussions of these OEs.

1. Describe the OE related to reaching in the Intake Structures.

2. Evaluate whether the observed leaching of calcium hydroxide and carbonation in accessible areas has any impact on the intended function of the Intake Structures.

8 Table 3.5-1 Table 3.5.2-4 3.5-56 3.5-89 AMR item 3.5.1-042 is used for managing aging effect of Loss of material (spalling, scaling) and cracking due to freeze-thaw.

The staff noted that Table 3.5.2-4 Emergency Diesel Generator Building does not include Table 2 items associated with Table 1 item 3.5.1-042.

Provide Table 2 items for this aging effect in Emergency Diesel Generator Building, or explain why it is not applicable.

9 Table 3.5-1 Varies Table 3.5-1 states Group 7 structures are not applicable to MNGP for AMR items 3.5.1-042, 043, and 044.

However, Table 3.5-1 also states Group 7 and group 8 structures are not applicable to MNGP for AMR item 3.5.1-047, 052, 063, 064, 065, 066, and 067.

There appears to be inconsistency among AMR items for Group 8 Structures.

Clarify whether Group 8 structures are applicable to MNGP, and update related SLRA Sections and Tables accordingly 10 Table 3.5.2-11 3.5-117 Table 2 item components in SLRA Table 3.5.2-11 only list accessible concrete associated with Table 1 item 3.5.1-044. Table 1 AMR item in SLRA Table 3.5-1 includes component of all groups for all concrete.

It appears that Table 2 AMR items are missing inaccessible concrete areas of Off-Gas Stack.

Provide missing Table 2 AMR items associated with 3.5.1-044 in SLRA Table 3.5.2-11.

Containment Liner Plate, Metal Containments, and Penetrations Fatigue Analyses, TLAA 4.6

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Fatigue Analyses of High Pressure Coolant Injection and Reactor Core isolation Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

3.5.2.2.1.5 3.5-24 In the first paragraph of SLRA 3.5.2.2.1.5, the components listed as having an existing CLB fatigue analysis described as TLAAs in SLRA Section 4.5, and for which AMR item 3.5.1-009 is credited, appears to be partly inconsistent (e.g., refueling bellows skirt) with the components described in SLRA Section 4.5 Clarify or reconcile any inconsistent components (e.g., refueling bellows skirt) described as having CLB fatigue analysis TLAAs between SLRA Sections 3.5.2.2.1.5 and 4.5, and corresponding Table 3.5.2-1 items which credit item 3.5.1-009. Provide conforming revision to the SLRA, if necessary.

2 4.5.1 4.5-1 The title of SLRA Section 4.5.1 reads Fatigue Analysis of the Suppression Chamber, Vents, Downcomers, and Torus Shell. It is not clear if there is a distinction between suppression chamber and torus shell.

Clarify the distinction, if any, between fatigue analysis of the suppression chamber and the torus shell which appears to be implied in the title of SLRA Section 4.5.1.

3 4.5.1; 4.5.2; 4.5.3 4.5-1 thru 4.5-4 The TLAA evaluation descriptions in SLRA Sections 4.5.1, 4.5.2 and 4.52 generally state only the SRV lifts as the primary transient included in the evaluations. From a review of calculation 22-014 (SIA Calculation 2100507.307) on the ePortal, it appears there are other transients (e.g., seismic OBE, chugging, post chug) included in the evaluations.

a) Discuss for clarity other transients included in the referenced TLAA evaluations, and why they were not included in the SLRA description.

b) Briefly clarify the source of the increase factors of 1.26 and 1.47 used for power uprate and EPU, respectively.

4 4.5.1; 4.5.2; 4.5.3 4.5-1 thru 4.5-4 The TLAAs in Sections 4.5.1, 4.5.2 and 4.52 are dispositioned in accordance with 10 CFR 54.21(c)(iii) and the effects of fatigue will be managed by the Fatigue Monitoring AMP (B.2.2.1).

For clarity, discuss what transients in each of these TLAA evaluations will be monitored by the Fatigue Monitoring AMP, and how they are included and controlled within the scope of the AMP.

5 4.5.5, 4.3.6 4.5-5, 4.3-4, 3-16 The TLAA evaluation for primary containment process penetration bellows fatigue states the evaluation was performed as part of ASME Class 2 and 3 and ANSI B31.1 fatigue evaluation described in Section 4.3.6.

SLRA 4.3.6 on page 4.3-16 (last paragraph) states, in a) Clarify the source of the maximum 7000 cycles in the CLB for staff verification.

b) Clarify if the term operating cycles is synonymous with thermal cycles.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions part: These containment penetration process bellows have been designed for a maximum of 7000 operating cycles.

It is not clear what the source of the 7000 operating cycles is within the CLB, and if operating cycles are the same as thermal cycles which is the transient being evaluated in the TLAA.

6 A.3.5.5, A.3.3.6 A-51, A-47, A-48 The USAR supplement description for Primary Containment Process Bellows Fatigue Analysis states the evaluation was performed as part of ASME Section III, Class 2 and 3 and ANSI B31.1 fatigue evaluation and is described in Section A.3.3.5. The staff noted that SLRA Section A.3.3.5 describes fatigue analysis for ASME Class 1 piping, and not for ASME Class 2 and 3 and ANSI B31.1 fatigue.

It appears that the SLRA intended to cross reference Section A.3.3.6 for non-Class 1 piping and not A.3.3.5.

However, Section A.3.3.6 makes no mention of bellows to which the summary description in A.3.5.5 applies, and therefore does not appear to provide an adequate USAR supplement for the TLAA.

Also, SLRA A.3.5.5 states, in part: The bellows are designed for a minimum number of operating cycles over the design life of the plant, However, it does not state what the number of cycles for which the bellows were designed.

a) Clarify if the reference to SLRA Section A.3.3.5 was intended to be A.3.3.6, and accordingly correct the SLRA A.3.5.5 description, as necessary.

b) Provide information in SLRA A.3.3.6 (to which reference appears to be made in A.3.5.5 for containment process penetration bellows) that would explicitly include and provides an adequate summary description for the TLAA of the subject bellows; OR, explicitly include an adequate USAR summary description in A.3.5.5 itself without referencing A.3.3.6.

c) State the transients and maximum number of transient cycles for which the bellows were implicitly designed for, and what the corresponding estimated cycles for the SPEO.

7 4.6.2, A.3.6.2 4.6-4, A-53 SLRA Section 4.6.2 and A.3.6.2 do not appear to include the transients or transient load combinations (LCs) and corresponding number of cycles based on which the higher fatigue usage of 0.111 (HPCI penetration) and 0.343 (RCIC penetration) for 40 years were calculated and projected to 80 years.

a) State the transients / transient LCs and corresponding number of cycles based on which the re-evaluated fatigue usage of 0.111 (HPCI penetration) and 0.343 (RCIC penetration) for 40 years were calculated and projected to 80

years,

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Also, the USAR supplement description for HPCI and RCIC turbine exhaust penetrations do not include a TLAA disposition pursuant to 10 CFR 54.21(c)(1).

b) Provide a revised Section A.3.6.2 which includes the transients/LCs and cycles evaluated to calculate fatigue usage, and the TLAA disposition consistent with SLRA Section 4.6.2.

8 4.6.2 4.6-4 SLRA 4.6.2 includes a TLAA for fatigue of HPCI and RCIC Turbine Exhaust Penetrations, which are stated as Torus Attached Penetrations. From a search of the SLRA, it appears that Table 2 AMR item is not included crediting these TLAAs for these components.

Provide Table 2 AMR item(s) for managing the fatigue aging effect for the HPCI and RCIC Turbine Exhaust torus attached penetrations.

Irradiation-Structural Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request C1 3.5.2.2.2.6 3.5-36 thru 3.5-39 This is for general understanding of the evaluation SLRA 3.5.2.2.2.6.

a) For clarity of the staffs understanding, explain in sufficient detail using illustrative drawings or diagrams, the general arrangement and configuration of the structures and components that fall under the scope of the further evaluation for irradiation effects in SLRA 3.5.2.2.2.6. Also, point to critical areas that were evaluated, including relative to the core belt-line, and provide an overall summary presentation of the evaluations in SLRA 3.5.2.2.2.6.

C2 3.5.2.2.2.6 3.5-38 Under subtitle Gamma Dose Biological Shield Irradiation Evaluation, the SLRA states that the gamma dose though 72 EFPY for the MNGP biological shield concrete was determined to be 4.85 x 1010 rads, which exceeds the SRP-SLR threshold limit of 1 x 1010 rads for radiation damage to concrete.

a) Provide the following information of the composition of the concrete used for the bioshield and RV pedestal: fine and coarse aggregate types (including whether siliceous or calcareous); the cement type; water-cement ratio; and the design compressive strength.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Since gamma rays can break anisotropic chemical bonds in concrete such as a covalent bond, and can reduce water content by radiolysis and evaporation, the staff needs information on the composition of concrete used for the bioshield and RV pedestal to further assess the risk. This information is not included in the SLRA section.

C3 3.5.2.2.2.6 3.5-38 Under subtitle Gamma Dose Biological Shield Irradiation Evaluation, noting that the gamma dose though 72 EFPY for the MNGP biological shield concrete was determined to be 4.85 x 1010 rads, which exceed the SRP-SLR threshold limit of 1 x 1010 rads for radiation damage to concrete, the SLRA further states, in part:

However, a separate analysis of the potential reduction in concrete strength due to gamma radiation above the recommended threshold has been completed for MNGP. This analysis considered attenuation through the concrete, and potential for radiation induced volumetric expansion (RIVE) of.,

as well as the impact to gamma heating considerations.

As a result, the integrity of the biological shield is assured, and no additional aging management of the biological shield concrete beyond the current Structures Monitoring (B.2.3.33) AMP is necessary for aging effects of irradiation during the SPEO...

The SLRA does not appear to include information (including references) with sufficient technical detail of the plant-specific separate analysis to support the above conclusion. The staff needs additional information of this analysis which is a primary basis to a) Describe in sufficient technical detail the stated separate analysis that was performed of the effects of potential reduction in concrete strength due to gamma radiation and summary of the results that would demonstrate structural integrity is assured for all relevant structural components (including anchorage) and a plant-specific AMP or enhancements to one or more existing AMPs is not necessary for the SPEO.

b) Describe the evaluation of the increase in temperature in the concrete due to gamma dose heating effects with the results, and how the concrete temperature acceptance criteria in SRP-SLR Section 3.5.2.2.1.2 or 3.5.2.2.2.2 (as supplemented by SLR-ISG-2021-03-STRUCTURES) are met with gamma heating included. Also, clarify the operating temperature and air gap in the reactor cavity.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions make its regulatory finding.

Additionally, the aging effects for SRP-SLR AMR Item 3.5-1, 097 includes radiation interactions with material and radiation-induced heating. It is not clear if and how the effects of gamma heating was included in concrete temperature assessment and what the results were.

C4 3.5.2.2.2.6; 3.5.2.2.1.2 3.5-38; 3.5-20 The SLRA states that [t]he average air temperature inside the drywell during normal plant operation is limited to 135°F. It also states that the [p]lant areas that bound high temperature considerations are the drywell general area and biological shield wall piping penetration local area, which experience temperatures of 135°F and 179°F, respectively.

Transware MNT-FLU-001-T-001, Rev 0) Figure 1 shows the existence of an inner and outer RV air cavity separated by a thermal insulation surrounding the biological shield wall. Structural Integrity Calculation Package Evaluation of Concrete Degradation of MNGP Bioshield states that MNGP calculation results in Bioshield Brick Effectiveness and Calculation of Local Temperature in Biological Shield Wall conclude that local temperatures at or near pipe penetrations range from 137 to 179 oF.

ANSI/ANS 6.4 states that production of secondary gamma rays just from rebars can amount to increase of twenty percent of gamma dose. It is not clear whether effects of thermal leakage due to reactor piping penetration of insulation surrounding the inner cavity has been considered in estimation of RV cavity temperatures. It is also not clear whether the aforementioned MNGP calculations considered gamma heating within the bioshield/pedestal concrete and whether further consideration was given to the a) Discuss/explain/demonstrate to what extent thermal load and effects from the inner RV air cavity have been considered, if any, in the calculation of the outer cavity temperature, particularly on the surface of the concrete.

b) Discuss/explain/demonstrate whether effects of gamma dose/rays have been included in concrete thermal calculations discussed under issue. If this has not been included, describe why not.

c) Discuss/explain/demonstrate whether the neutron thermalization properties of concrete may not be of concern in the spatial distribution of secondary gamma-ray production that could affect rebar bonding within the structural portion of the concrete bioshield/pedestal.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions production of secondary gamma rays potentially increasing temperature of the concrete locally beyond the SRP-SLR limit of 200 oF.

C5 SIA Calculation 2200285.301 R0 17, 8 Subject SIA Calculation Monticello SLR Concrete Embrittlement Assessment of Biological Shield Wall

[BSW] cites certain references that staff needs for audit and verification.

Also, the calculation refers to #3 and #4 restraints and concrete ring in its Fig 1 on page 8, and it is not clear of the configuration and function of these BSW components.

Provide on ePortal the following:

a) Original design calculations for bioshield wall (CA-68-009 R0, CA 860 R0), and RV Pedestal b) SI File No. 2200285.202P (Refs. 7,

11) c) Reference 5 (EPRI Report 3002013084 - 2018 LTO Structural Tools); Reference 14 (PM Bruck et al.

d) Explain #3 & #4 restraints and concrete ring of BSW wall in Figure 1 of subject calculation, and their function.

C6 SIA Calc 2200285.301 R0 9, 16 Assumption 1 in Section 4.0 states, in part: The variation of gamma flux along the height of the Monticello biological shield wall [BSW] is taken from a generic curve available in EPRI report [3002011710]

[4] for a 3-loop PWR. The EPRI curve, reproduced in Figure 3(b), presents data normalized to the gamma flux at core mid-plane,,,,height of active core region is 12 feet, Plant-specific variation of gamma dose along the height is not available in the design input.

Also, Section 6.0 on page 17, states in part:., the anchorage occurs in the RPV concrete pedestal which is sufficiently remote from active core region such that gamma radiation levels can be assumed less than threshold levels. Gamma dose at top of pedestal at 72 EFPY could be added as input to this a) Discuss/explain/clarify the rationale and its adequacy for use of a generic normalized curve for variation of gamma flux along core height for a PWR for MNGP BWR.

b) Clarify if gamma dose estimates at top of structural concrete and top of pedestal at 72 EFPY were estimated to verify Assumption 1 and its use in evaluation of the structural portion of BSW and the RV concrete pedestal and provide these estimates. If not estimated, justify why it was necessary to verify related assumptions in the calculation and explicitly verify how gamma dose estimates at these

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions evaluation in order to verify the assumption.

It is not clear if gamma dose estimates at top of structural concrete and top of pedestal at 72 EFPY was estimated to verify the assumption(s) referenced in the paragraphs above.

locations compare to SRP-SLR threshold limits.

C7 SIA Calc 2200285.301 R0 13, 19 Section 5.2 states on p13: The literature indicates little to no effect due to gamma heating [17], thus effects of temperature on the biological shield wall can be neglected. To the contrary, Reference [17]

on p19 which is NUREG/CR-7171, states in Section 8.2.1 Temperature effects, that Gamma and neutron radiation can produce elevated temperatures and thermal gradients in concrete. Further Section 10 of the NUREG/CR states, in part: Water in concrete can be decomposed by gamma rays by a process of radiolysis..Water can also be removed from the concrete by evaporation due to heat generated by gamma radiation..gamma radiation has a greater effect on the cement paste than it has on the aggregate materials.

a) Explain the apparently contradictory statement regarding radiation heating in the subject SIA calculation versus Reference 17, which is cited as the source of the information in the calculation.

C8 SIA Calc 2200285.301 R0 16, 17, 7, 9

Section 5.3 states, in part: The design of the biological shield wall [BSW] was performed using the Working Stress Method of ACI 318-63 [12], using allowable stresses for concrete and the reinforcing steel. Per a GE APED design specification cited in original calculations [6] (p.23, CA-68-009 Sheet 2),

allowable stresses were increased by a factor of 1.5 for controlling load cases involving the jet forces.

Resulting demand-to-capacity (D/C) ratios for normal strength and degraded concrete are provided in Table 5 and shown to be less than 1.0.

Also, Table 4 Design Input Used in Evaluation on p7 states that main load combinations to include a) Explain whether design check calculations were performed for normal and degraded strength concrete (through SPEO) of the BSW for applicable controlling combination(s) for which the increase in allowable stress by a factor of 1.5 does not apply. If so, identify such controlling load combination (pointing to applicable USAR section) and provide the D/C results.

b) If not, explain/clarify why such a design check is not necessary or how it is bounded to assure intended function

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Seismic, Jet Force. Also, assumption 4 on p9 states, in part: The design calculation take loads from reactor building computer model which is analyzed using multiple load combinations [LCs}. These LCs are described as including seismic and jet forces.

It is not clear whether calculations were performed for controlling combination(s) for which normal code allowable stress apply (i.e., the increase in allowable stress by a factor of 1.5 is not allowed (e.g., LC 2a:

D+R+E in USAR Section 12.2.1.4) and whether corresponding acceptance criteria were met.

or enveloped by the controlling load combination and allowable stresses for load combination(s) where allowable stresses can be increased by a factor of 1.5.

C9 3.5.2.2.2.6 3.5-37, 3.5-38 The second paragraph on page 3.5-37 states:

Irradiation effects on the biological shield concrete, the biological shield structural steel, and reactor vessel support structureare evaluated below.

There appears to be no evaluation provided in the SLRA of the reinforced concrete RV support pedestal. It is not clear if and how the reinforced concrete pedestal, which is in the vicinity of the RV and would fall under the scope of the SRP-SLR 3.5.2.2.2.6 FE was evaluated for applicable irradiation effects.

a) Clarify if the reinforced concrete pedestal structure was evaluated for irradiation effects, noting that the gamma dose estimate on the bioshield wall exceeded the SRP-SLR threshold limit and the estimate of gamma dose at the pedestal structure is not provided in the SLRA.

b) If so, explain how it was evaluated, provide summary of results, how the applicable acceptance criteria in SRP-SLR Section 3.5.2.2.2.6 were met, or how the effects of aging would be managed for the concrete pedestal structure. Update the SLRA accordingly.

c) If not, justify why an evaluation for irradiation effects on the pedestal concrete was not necessary, or provide a sufficient evaluation.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions C10 3.5.2.2.2.6, Table 3.5-1, Table 3.5.2-1, B.2.3.33 & PBD XCELMO00017-REPT-080 3.5-38, 3.5-74, 3.5-76, Under subtitle Gamma Dose Biological Shield Irradiation Evaluation, the SLRA states, in part:

As a result, the integrity of the biological shield is assured, and no additional aging management of the biological shield concrete beyond the current Structures Monitoring (B.2.3.33) AMP is necessary for aging effects of irradiation during the SPEO.

SLRA Table 3.5-1 states that item 3.5.1-097 is applicable and Consistent with NUREG-2191, and includes two (2) corresponding AMR items in Table 3.5.2-1 crediting the Structures Monitoring Program to manage aging effects of irradiation of concrete.

However, the credited SLRA B.2.3.30 AMP description and its program elements in the ePortal PBD do not appear to include reduction of strength; loss of mechanical properties due to irradiation (i.e.,

radiation interactions with material and radiation-induced heating) as aging effects/mechanism that will be managed by the program.

a) Revise the SLRA B.2.3.33 AMP and its PBD as appropriate to include managing the aging effects/mechanism corresponding to SLRA item 3.5.1-097 for which the AMP is credited in the SLRA.

C11 3.5.2.2.2.6 3.5-36 thru 3.5-41 General material condition of reactor cavity area Discuss current general material and structural condition of biological shield wall, RV pedestal, RV support skirt and RV seismic restraint based on inspections performed in the past. Use illustrative photos where available.

C12 Table 3.5.2-1, 3.5.2.2.2.6 3.5-76, 3.5-84 SLRA Table 3.5.2-1, includes a plant-specific Note 7, corresponding to AMR item 3.5.1-097 for the Biological Shield Wall which states:

Consistent with SLR-ISG-2021-03-STRUCTURES, which allows a plant-specific AMP, or a selected AMP enhanced as necessary; the Structures Monitoring (B.2.3.33) AMP will be used to manage the potential for reduction in strength, loss of mechanical a) Provide a revised plant-specific Note 7 in SLRA Table 3.5.2-1 that is consistent with the evaluation in SLRA Section 3.5.2.2.2.6.

b) Clarify if the corresponding AMR item in SLRA Table 3.5.2-1 with plant-specific Notes 6, 7 also include the biological shield wall with a structural

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions properties, or cracking of the biological shield due to irradiation near the reactor vessel, as the projected values for neutron and gamma radiation incident on the shield wall are less than the threshold values of 1x1019 n/cm2 and 1x1010 rads, respectively.

(emphasis added)

Based on the evaluation in SLRA Section 3.5.2.2.2.6, the highlighted statement is incorrect for gamma radiation.

Also, the corresponding Table 3.5.2-1 AMR item is only for the portion of the biological shield wall that has a shielding function, but the portion with a structural support function does not appear to be included.

support function, and revise accordingly.

F1 3.5.2.2.2.6; 4.2.1.1 3.5-37 4.2-3 The SLRA states The neutron source that was used to calculate the neutron fluence, as well as the gamma dose at 72 EFPY, for the biological shield concrete is the maximum-power reactor statepoint condition that was determined to occur in Cycle 28.

The SLRA also states that the fluence methodology implemented by TransWare RAMA methodology is capable of predicting specimen activities within the MNGP reactor pressure vessel and [b]ased upon these results, there is no discernable bias in the computed reactor pressure vessel fluence for the period of Cycle 1 through the end of Cycle 30 for the MNGP reactor. Attachment 1 to Transware Proprietary MNT-FLU-001-R-001,Rev 0, however, states that each fuel design has a different power signature in the core and, therefore, results in different spatial power, [and] exposure a) Discuss/explain/demonstrate the discrepancy between the SLRA and to MNT-FLU-001-R-001,Rev 0. If there is no discrepancy state why.

b) Discuss/explain/demonstrate why different exposures resulting from varied cycle to cycle fuel designs would not constitute a bias.

c) What fuel assemblies are considered for projected cycles beyond Cycle 28?

Does this differ from the rest of the fluence analysis in MNT-FLU-001-R-001?

F2 3.5.2.2.2.6 ANSI/ANS-6.4-2006, Nuclear Analysis and Design of Concrete Radiation Shielding for Nuclear Power Plants, widely discusses the production of secondary Discuss/explain/demonstrate whether the RAMA or any methodology was used for the production and estimation

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions gamma rays when there is an exposure of steel to neutron and gamma-rays. It states that "in many reactor shielding situations, the secondary gamma radiation produced within the primary [bio]shield is a more important contribution to the dose outside the shield than is the neutron radiation. It is not clear whether the RAMA methodology accounts for production of secondary gamma radiation within the bioshield/RV pedestal and if so, whether it was applied as such to all aspects of MNG irradiated concrete (including those of radioactive waste tubes, BWR steam piping encapsulated by concrete).

of secondary gamma rays in areas of concrete exposed to radiation (neutron, gamma). If not, explain why the aforementioned secondary production of gamma rays was not pursued in augmenting the estimation of gamma dose in the RV bioshield structural concrete and other steel lined concrete or that encapsulating the BWR steam piping.

S1 3.5.2.2.2.6 3.5-38 to 3.5-39 On these pages of the referenced SLRA section, the applicant the described the configuration of the RV support skirt/concrete pedestal.

Go over the configuration of the entire support system for the RV (all the steel and all the concrete components),

showing the appropriate drawings to help the staff understand the load transfer path from the RV to the surrounding concrete (and vice versa).

S2 3.5.2.2.2.6 3.5-39 On this page of the referenced SLRA section, the applicant states a fluence value of 3.25x1016 n/cm2 at a nozzle location below the beltline, but this location is above the RV steel support skirt assembly.

The fluence value down at the RV steel support skirt assembly is expected to be lower, but this value is not provided.

What is the approximate fluence value (and corresponding dpa value) at the knuckle region of the RV support skirt assembly that may be compared to the dpa value for which there is no NDT shift per Figure 3-1 of NUREG-1509?

S3 3.5.2.2.2.6 3.5-39 The evaluation of the RV steel support skirt assembly cites proprietary report BWRVIP-342. However, this report did not specifically analyze the MNGP RV support skirt configuration. Therefore, these questions pertain to plant-specific information that would demonstrate the applicability of BWRVIP-342 to MNGP.

a) Explain/demonstrate how the MNGP RV support skirt configuration is bounded by the RV support skirt configuration analyzed in BWRVIP-342.

b) Explain/demonstrate how the MNGP design basis transients and design loads (deadweight, operational and safe shutdown earthquake, pipe rupture/blowdown) are bounded by the

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions those analyzed in BWRVIP-342.

Provide the appropriate source documents in the portal.

c) Provide the LST values at the RV support skirt of MNGP (at the knuckle region and down at the cylindrical skirt region) and explain how the values are bounded by the LST value used in BWRVIP-342. Provide the appropriate source documents in the portal.

d) SLRA 3.5.2.2.2.6 provided a maximum initial NDT value of 40°F for the MNGP RV bottom head. However, no initial NDT values for the MNGP RV support skirt itself (cylindrical portion) and associated welds were provided.

Provide initial NDT values for the MNGP RV support skirt and its associated welds. What materials (i.e.,

SA-###) are the MNGP RV bottom head and the cylindrical support skirt made of? Provide the appropriate source documents in the portal.

e) The second to the last paragraph of the RV steel skirt evaluation concluded that the EPRI document (i.e., BWRVIP-342) is applicable to MNGP. However, that conclusion was based on fluence levels alone. Somehow parts a, b, c, and d would need to be brought into the argument, i.e., how plant-specific MNGP parameters show that MNGP is bounded by the evaluations in

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions BWRVIP-342, which are based on the two main evaluation methods recommended in NUREG-1509.

S4 3.5.2.2.2.6 3.5-40 to 3.5-41 On this page of the referenced SLRA section, the applicant seems to be relying on the transition temperature analysis (TTA) method in NUREG-1509 as basis for the evaluation of the structural steel of the bioshield wall. However, the information provided in SLRA 3.5.2.2.2.6 is not complete nor clear, or it is very fragmentary. The calculation basis, 2200285.302.R0, is also missing some parts; it has most elements, but it doesnt put everything together.

The concept of TTA is to demonstrate the following with sufficient margin (guidance on sufficient margin in NUREG-1509):

Initial NDT + shift < lowest service temperature of the steel Discuss/explain that the initial NDT +

shift is less than the lowest service temperature that the corresponding steel is subject to with sufficient margin.

This explanation and discussion, with the values from 2200285.302.R0, would need to be upfront and clear in SLRA 3.5.2.2.2.6, and therefore a supplement to this SLRA section is called for. Also, this section of SLRA has to clearly cite 2200285.302.R0 as the basis for the evaluation.

S5 3.5.2.2.2.6 3.5-40 2200285.302.R0, Section 5.2:

This section of the document talks about the selection of +/- 72 and +/- 76 distances from the mid-core/midplane locations along the bioshield height wherein the dpa exposure levels have dropped off.

Why not analyze the height/location at bioshield inner diameter where the peak 72 EFPY dpa level of 2.07x10-3 dpa occurs? According to Table 1 of 2200285.302.R0 this location is close to mid-plane of active core region. Dont the structural steel columns run the whole active core height? And, therefore, there is structural steel at this location of peak dpa level? If there is structural steel there, that location with peak dpa level would need to be analyzed and results presented either in an SLRA supplement or updated Table 3 of 2200285.302.R0. The lowest service temperature of the steel at the corresponding location (see question related to TTA method above) would

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions also have to be shown in Table 3 or presented in an SLRA supplement.

S6 3.5.2.2.2.6 3.5-40 On the tech basis for the bioshield steel evaluation, non-proprietary 2200285.302.R0, Section 4.0:

a) Assumption 1 explains assumptions between dpa and fluence.

b) Assumption 2 states there is no available plant-specific variation of fluence along height of the MNGP bioshield wall. Therefore, the fluence variation curve in NUREG/CR-5320 for a PWR was used, normalized to the dpa value for MNGP at the bioshield cladding.

This assumption referenced EPRI report 3002011710 (2018).

a) Explain Assumption 1 and its basis.

b) MNGP is a BWR-3. Explain why/how the fluence variation curve for a PWR from NUREG/CR-5320 is bounding for MNGP.

S7 3.5.2.2.2.6 3.5-40 2200285.302.R0, Sections 5.3 and 6.0:

5.3 states that design basis load combinations were analyzed in the finite element analyses; 6.0 presents stress results for the controlling load combination.

However, specifics were not provided.

What were the design basis load combinations and what was the controlling load combination? Provide the appropriate source documents in the portal.

S8 3.5.2.2.2.6 3.5-40 2200285.302.R0, Section 5.3 describes the stress analysis performed for the structure steel elements of the bioshield wall. The staff would like an explanation of the stress analysis to understand the stresses in the bioshield wall.

Go over the stress analysis in 2200285.302.R0, Section 5.3 to help the staff understand the low stress in the structural steel.

S9 3.5.2.2.2.6 3.5-36 thru 3.5-41 General catch-all to discuss portal documents that potentially need a summary or docketing, since these are documents that contain underlying technical bases for this SLRA section.

Discuss summary or potential docketing of these portal documents:

BWRVIP-342 2200285.301.R0 2200285.302.R0 S10 B.2.3.30; 3.5.2.2.2.6, Table 3.5.2-1, Table 3.5.2-7, B.2.3.33 B-224 thru B-226; 3.5-30; 3.5-76 SLRA 3.5.2.2.2.6 under subtitle Reactor Vessel Support Steel Irradiation Evaluation, states, in part:

Therefore, the integrity of the reactor vessel supports is assured, and no additional aging management of reactor vessel supports beyond the a) While a plant-specific program may not be necessary, describe how the aging effects due to irradiation embrittlement will be adequately managed by the IWF AMP that is

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions thru 3.5-83; 3.5-98 thru 3.5-103 current ASME Section XI, Subsection IWF (B.2.3.30)

AMP is necessary for aging effects due to irradiation during the MNGP SPEO.

Also, SLRA 3.5.2.2.2.6 under subtitle Biological Shield Structural Steel Evaluation, states, in part:

As a result, the integrity of the biological shield is assured, and no additional aging management of the biological shield beyond the current Structures Monitoring (B.2.3.33) AMP is necessary for aging effects due to irradiation during the MNGP SPEO.

Nevertheless, while a plant-specific AMP may not be necessary, loss of fracture toughness due to irradiation embrittlement remains an applicable aging effect for the RV steel supports for SLR. Although Table 2 of the IWF PBD (XCELMO00017-REPT-077) on the ePortal includes loss of fracture toughness due to irradiation embrittlement among the aging effects/mechanisms managed by the program, SLRA Sections 3.5.2.1.1, 3.5.2.1.7, Table 3.5.2-1, and Table 3.5.2-7, do not include AMR items that the aging effect will be managed by the IWF AMP and SLRA B.2.3.30 does not appear to include loss of fracture toughness due to irradiation embrittlement as an aging effect that will be managed by the program.

Further, the SLRA B.2.3.33 Structures Monitoring AMP and its PBD program elements do not included loss of fracture toughness due to irradiation embrittlement as aging effect/mechanism for the Biological Shield Structural Steel that is credited to be managed by the program. Also, the SLRA Table 3.5.2-1 do not include corresponding AMR items.

credited for the RV steel support assembly components for the SPEO?

Provide corresponding Table 2 AMR items, noting that it currently is not included in the GALL-SLR Report, and any related changes that may need to be made to the SLRA.

b) Revise the SLRA and the PBD for the Structures Monitoring, B.2.3.33 AMP to include managing the aging effects of loss of fracture toughness for the biological shield steel components for which the AMP is credited. Also, provide corresponding Table 2 AMR items.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions S11 3.5.2.2.2.6 3.5-37 The second paragraph on page 3.5-37 states:

Irradiation effects on the biological shield concrete, the biological shield structural steel, and reactor vessel support structureare evaluated below.

There appears to be no evaluation in the SLRA of the steel RV seismic restraint and stabilizer structure components (brackets, tension rods, couplings, truss etc.). It is not clear if and how the steel RV seismic restraint and stabilizer structure, which is part of the RV supports and would fall under the scope of the SRP-SLR 3.5.2.2.2.6 FE was evaluated for loss of fracture toughness due to irradiation effects. Also, it appears that AMR items are not included in the SLRA for managing the aging effect for these components.

a) Clarify if the steel RV seismic restraint and stabilizer structure components (brackets, tension rods, couplings, truss etc.) were evaluated for loss of fracture toughness due to irradiation embrittlement effects.

b) If so, explain how it was evaluated, provide summary of results, how the applicable acceptance criteria were met, or how the effects of aging would be managed for the steel RV seismic restraint and stabilizer structure during the SPEO. Provide a summary description of the evaluation in the SLRA, include applicable AMR items, and update the program credited to manage the aging effects.

c) If not, justify why an evaluation for irradiation effects on the RV seismic restraint and stabilizer structure was not necessary and how the irradiation aging effects would be adequately managed during the SPEO.

B.2.3.33 Structures Monitoring Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.3.33 2.4.9 2.4.10 2.4.16 2.1.4.2.1 B-238 2.4-20 2.4-23 2.4-35 2.1-14 GALL SLR XI.S6 states that the scope of the program includes all SCs, component supports, and structural commodities in the scope of license renewal that are not covered by other structural aging management programs.

1. SLR boundary drawing No. SLR-36444 lists two 345kv Scope of the Program:

1. Clarify the SLR scope for these two 345kv substation houses and 115/345 kV Control House, and address the inconsistency.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions substation houses, one is in SLR scope, and another one is not. SLRA Section 2.4.10 states that the miscellaneous SBO yard structures include the 115/345 kV Control House, which is not in the scope of SLR identified in SLR boundary drawing No. SLR-36444. It is unclear to the staff whether 345kv substation house and 345 kV Control House are the same structure.

2. SLRA Section 2.4.16 includes 13.8 kV room within the scope of SLR, but it is not included in the scope of SLR as defined in SLR boundary drawing No. SLR-36444.

3. SLRA Section 2.1.4.2.1 states that HELB related structural components such as whip-restraints and jet impingement shields/barriers, along with the piping supports are in the scope of SLR. The staff noted pipe whip restraints in Table 2.4-7, but the staff could not locate the jet impingement shields/barriers in SLRA table.

The staff also could not locate any information of whip-restraints and jet impingement shields/barriers in procedure 1385, Revision 17.

4. GALL-SLR XI.S6 states that seismic joint fillers are within the scope of SLR. SLRA Appendix A.2.2.33 states that the Structures Monitoring program includes the inspection of seismic joint fillers. However, the staff could not locate AMR items associated with seismic joint fillers.

5. GALL-SLR XI.S6 states that sump liners, tube tracks, and trash racks associated with water-control structures are within the scope of SLR. However, the staff could not find the sump liners, tube tracks, and trash racks associated with water-control structures in SLRA.

6. SLR boundary drawing No. SLR-36444 lists Hot Machine Shop in the scope of SLR. Procedure 1385,

2. Clarify the SLR scope for 13.8 kV room and address the inconsistency.

3. Clarify whether existing AMP program includes whip-restraints and jet impingement shields/barriers in the scope, and provide AMR items for jet impingement shields/barriers.

4. Clarify and include AMR items for seismic joint fillers.

5. Clarify whether sump liners, tube tracks, and trash racks are within the scope of SLR. If yes, provide AMR items for the sump liners, tube tracks, and trash racks associated with water-control structures.

6. Provide AMR items for the Hot Machine Shop. Also check which buildings and structural components within the scope of SLR are not listed in these tables 1-21 of procedure 1385, and evaluate whether procedure 1385 needs to be revised for an enhancement.

7. Explain how aging effects for the Diesel Fire Pump House is adequately managed, and provide AMR items.

8. Confirm and address the inconsistency of building names.

9. Based on the responses above, evaluate whether additional

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions revision 17 lists Tables 1 through 21. However, these tables do not include all the buildings and structural components managed by the Structures Monitoring program. For example, Hot machine Shop is not listed in these tables. The staff also could not find any information of Hot Machine Shop in SLRA.

7. SLR boundary drawing No. SLR-36444 lists Diesel Fire Pump House within the scope of SLR. SLRA Section 2.4.9 states In addition to the INS itself, this structure also covers the access tunnel and Diesel Fire Pump House. It is not clear to the staff whether Diesel Fire Pump House is a part of Intake Structure, the staff could not locate the information how the Diesel Fire Pump House is adequately managed.

8. SLRA B.2.3.33 discusses the plant-specific operating experience for the Diesel Oil Pump House, which its name was used by ARs. SLRA Section 2.4.3 discusses the Diesel Fuel Oil Transfer House. During the on-site audit, the staff confirmed they are the same building.

enhancements to the Structures Monitoring program are needed, and update SLRA if necessary.

2 N/A N/A

1. GALL-SLR XI.S6 states that elastomeric vibration isolators, structural sealants, and seismic joint fillers are monitored for cracking, loss of material, and hardening.

The staff could not locate elastomeric vibration isolators, structural sealants, and seismic joint filler, as well as their parameters monitored or inspected in SLRA.

2. SRP-SLR describes the indication of Alkali-Silica Reactions (ASR) as map or patterned cracking, alkali-silica gel exudations, surface staining, expansion causing structural deformation, relative movement or displacement, or misalignment/distortion of attached components. AMP basis document lists parameters monitored or inspected as unique map or cracking that would indicate the presence of ASR. It appears that ASR Parameters Monitored or Inspected:

1. Clarify whether elastomeric vibration isolators, structural sealants, and seismic joint fillers are within the scope of SLR. If yes, provide parameters monitored or inspected for the elastomeric vibration isolators, structural sealants, and seismic joint fillers, and provide AMR items.

2. Evaluate the parameters monitored or inspected for the ASR, and provide an enhancement if necessary.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions has more characteristics than ones described in the AMP basis document.

3 N/A N/A AMP basis document discusses the FRP companys response to RAI B.2.3.35-7 (ML18334A182) for the Turkey Point SLRA, and intends to enhance the Element Parameters Monitored or Inspected to the MNGP Structures Monitoring program to ensure that concrete is monitored for the aging effect of increase in porosity and permeability and loss of strength due to leaching of calcium hydroxide and carbonation. The staff could not find this enhancement to the Structures Monitoring program.

Parameters Monitored or Inspected:

Provide the enhancement as discussed in the AMP basis document.

4 N/A N/A GALL-SLR XI.S6 states that all structures are monitored on an interval not to exceed 5 years.

Procedure 1385, Revision 17 states that the inspection intervals for those normally inaccessible areas may exceed five years.

It is unclear to the staff where those normally inaccessible areas are located. It appears that the Structures Monitoring program has an exception to NUREG-2191.

Detection of Aging Effects:

1. Identify those normally inaccessible areas monitored on an interval of exceeding five years.

2. Evaluate whether the Structures Monitoring program has an exception to NUREG-2191, if yes, provide justification why this exception is acceptable.

5 N/A N/A GALL-SLR report Table IX.B, Use of terms for structures and components, defined areas covered or obstructed by insulation and protective coatings as accessible.

Table 19 in procedure 1385, Revision 17, states that the interior of the Diesel Fire Pump House masonry block walls is covered with insulation. The Structures Monitoring program will require that the interior surfaces of the walls will be examined if exterior wall surfaces show evidence of significant aging effects. [Commitment M05006A]

It appears that the procedure considers the interior of the Diesel Fire Pump House masonry block wall covered by insulation as inaccessible, which conflicts with GALL-SLR Detection of Aging Effects:

1. Reevaluate the procedure how to adequately manage aging effects of the accessible interior of the Diesel Fire Pump House masonry block walls covered by insulation.

2. Clarify which AMP shall be used to manage this aging effect of these masonry block walls covered by insulation, and address the inconsistency.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions report recommendations.

It is also not clear to the staff why the Structures Monitoring program instead of Masonry Walls program is used to manage aging effect of these interior masonry walls, which conflicts with the information provided in SLRA.

6 B.2.3.33 B-240 SLRA B.2.3.33 provides an enhancement to Detection of Aging Effects to explicitly include inspection of the following components and commodities, such as expansion plugs, fuel storage racks(new fuel), and so on.

It appears that this enhancement is for Scope of the Program.

Detection of Aging Effects:

Evaluate and Clarify which Element needs to be enhanced 7

A.2.2.33 B.2.3.34 A-30 B-244 SLRA Appendix A.2.2.33 states that quantitative results (measurements) and qualitative information from periodic inspections are trended with sufficient detail, such as photographs and surveys for the type, severity, extent, and progression of degradation, to ensure that corrective actions can be taken prior to a loss of intended function.

The staff could not locate the above-mentioned information in procedure 1385. The staff noted that the Inspection of Water-Control Structures AMP provides an enhancement to include trending of quantitative measurements and qualitative information for findings exceeding the acceptance criteria for all applicable parameters monitored or trended.

It is unclear to the staff whether the same enhancement is applicable to the Structures Monitoring program.

Monitoring and Trending:

Evaluate whether this enhancement to the Inspection of Water-Control Structures AMP is applicable to the Structures Monitoring program.

8 N/A N/A AMP basis document discusses the FPL companys response to RAI B.2.3.35-5 (ML18334A182) for the Turkey Point SLRA, and the implementing procedure will be enhanced to strengthen the detail and criteria for Monitoring and Trending:

Evaluate whether this enhancement to the Inspection of Water-Control Structures AMP is applicable to the

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions engineering evaluation performed when a deficiency involves significant corrosion related degradation. It also states clarifying in the Element Monitoring and Trending to the MNGP Structures Monitoring program that inspections are to occur in no greater than 5-year intervals.

It appears that the Structures Monitoring program does not include this enhancement.

Structures Monitoring program.

Monitoring and Trending:

Explain how the implementing procedure will be enhanced to reflect the site-specific OE.

9 N/A N/A GALL-SLR XI.S6 describes acceptance criteria for loose bolts and nuts, structural sealants, elastomeric vibration isolation elements, and sliding surfaces.

The staff reviewed the AMP basis documents, and found that it does not include acceptance criteria for these structural components.

Acceptance Criteria:

Provide acceptance criteria for loose bolts and nuts, structural sealants, elastomeric vibration isolation elements, and sliding surfaces if applicable.

10 B.2.3.33 B-241

Background:

The staffs OE audit identified that Action Request (AR) 01500214, dated 6/21/2016, discusses the settlement issue for the Diesel Oil Pump House (Diesel Fuel Oil Transfer House) starting from the late 1970 and early 1980. Recent construction placed a large concrete slab over the Diesel Oil Storage Tank (T-44) and large heavy I-beams on the roof of the Diesel Fuel Oil Transfer House, which added significant weight to structure and resulted in rainwater leakage through the foundation. The AR 01500214 also notes that the Diesel Fuel Oil Transfer House structure is visibly leaning to the west, and the settlement may be adding additional stress to the fuel oil pipes penetrating the west wall and making them susceptible to fracture.

Corrective Action Program (CAP) No. 50001490358, dated 3/15/2017, indicates the NW corner of the Diesel Fuel Oil Transfer House was approaching the lower limit of the settlement range. The estimated elevation was Operating Experience:

1. Explain how settlement baseline was established.

2. Explain how acceptance values for settlement were determined to ensure that the fuel oil pipe stresses are within the code allowable limit.

3. Explain what corrective actions will be taken if settlement exceeds the acceptance criteria.

4. Update the SLRA to include the settlement related operating experience for the Diesel Fuel Oil Transfer House, the Diesel Oil Storage Tank (T-44), and the OGSB.

5. Explain how the Structures Monitoring

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions found to be 937.80 feet, and its settlement acceptance range is 937.79 to 937.87 feet as described in the procedure 1396.

Procedure 1396 states that a CAP Action Request [NRC Commitment M05009A] will be initiated when settlement is NOT within acceptance criteria. Procedure FP-PE-RLP-01, Revision 9 implements NRC Commitment M05009A, which states Site documents that implement aging management activities for license renewal will be enhanced to ensure that an AR is prepared in accordance with plant procedures whenever non-conforming conditions are found (i.e., the acceptance criteria is not met).

The NRC staff conducted on-site audit on operating experience related to settlement of the Diesel Fuel Oil Transfer House, the Diesel Oil Storage Tank (T-44), and the Offgas Storage Building HTV exhaust pipe (OGSB).

The NRC staff reviewed the settlement data and its trend over time provided by the applicant, and found that the Diesel Fuel Oil Transfer House has about 5.4 inch of significant baseline differential settlement between NE/SE corners and NW/SW corners, and the projected settlement of the Diesel Fuel Oil Transfer House may exceed acceptance criteria during the SPEO.

Issues:

It is unclear to the NRC staff how settlement baseline was established, and how acceptance values for settlement were determined to ensure that the fuel oil pipe stresses are within the code allowable limit. It is also unclear to the NRC staff what corrective actions will be taken if settlement exceeds the acceptance criteria.

It is unclear what evaluation methodology will be used to program can adequately manage aging effects due to the settlement of the Diesel Oil Pump House, the Diesel Oil Storage Tank (T-44), and OGSB to ensure that the fuel oil pipes, and any other impacted SSCs within scope of SLR, can maintain their intended functions during the SPEO.

If needed, provide updated settlement criteria with basis ( or process for determining conditional acceptance criteria to ensure intended function when the acceptance criteria is not met) against which the need for corrective actions are evaluated.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions implement the settlement requirements for the Structures Monitoring program such that aging effects due to the settlement will be adequately managed through the end of the SPEO if future settlement exceeds the acceptance criteria specified in the procedure 1396.

Based on the operating experience it is unclear whether settlements are expected to exceed or will exceed the settlement acceptance criteria during the SPEO.

Therefore, program elements (e.g., acceptance criteria, corrective actions or evaluation methodology to change the acceptance criteria) related to settlement need to be described, modified or enhanced to demonstrate that the aging management program will be adequate to manage the aging effect during the SPEO.

11 N/A N/A Procedure 1385, Revision 17 discusses the implementation of NRC license renewal commitments M05006A, M05008A, M05009A, M05010A, M05012A, M05045A, M05046A, M05047A, M05049A, and M05050A, etc.

It is the staffs understanding that all the implementing procedures mentioned in these commitments have been enhanced prior to the period of extended operation, and will continue to be implemented during the SPEO.

Confirm that all the implementing procedures mentioned in these commitments have been enhanced prior to the period of extended operation, and will continue to be implemented during the SPEO.

12 Table 3.3-1 3.3-55 AMR item 3.3-1,111 claims to be not used. However, SLRA states that structural steel is addressed as part of structural items in Section 3.5. It is not clear how AMR item 3.3-1, 111 is addressed in Section 3.5.

Clarify which AMR item in Section 3.5 addresses the aging effect for AMR item 3.3-1, 111.

13 Tables 3.5.2-4 and 3.5.2-17 2.5-91 3.5-146 GALL-SLR report lists building concrete at locations of expansion and grouted anchors; grout pads for support base plates as component in AMR 3.5-1, 055 to manage aging effect of reduction in concrete anchor capacity.

However, Table 2 items associated with AMR 3.5.1-055 in Tables 3.5.2-4 and 3.5.2-17 list component of joint and Evaluate whether Table 2 items associated with AMR 3.5-1, 055 in Tables 3.5.2-4 and 3.5.2-17 are correctly documented.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions penetration seals with flood barrier and HELB barrier intended function. The staff also noted that aging effect of loss of sealing for the joint and penetration seals associated with other AMR items is managed by the Structures Monitoring program.

It is unclear to the staff whether aging effect of reduction in concrete anchor capacity is applicable to joint and penetration seals.

14 Table 3.5-1 Table 3.5.2-3 Table 3.5.2-4 3.5-64 3.5-86 3.5-89 Table 2 items associated with AMR item 3.5-1, 063 provide inconsistent list of component for concrete:

basemat, foundation. For example, Table 2 item in Table 3.5.2-3 lists concrete: basemat, foundation as accessible, however, Table 2 items in Table 3.5.2-4 list concrete: basemat, foundation as both accessible and inaccessible.

Evaluate and clarify all Table 2 items associated with AMR item 3.5-1, 063 for the concrete: basemat, foundation.

15 Table 3.5.2-3 3.5-86 Table 2 item associated with AMR 3.5.1-065 list concrete:

basemat foundation as accessible in Table 3.5.2-3, but Table 2 items in other tables list concrete: basemat foundation as both accessible and accessible. It appears that Table 2 item in Table 3.5.2-3 for concrete: basemat foundation (inaccessible) is missing.

Evaluate and clarify Table 2 items associated with AMR 3.5.1-065 in Table 3.5.2-3.

16 Table 3.5.2-4 Table 3.5.2-9 3.5-90 3.5-111 AMR item 3.5.1-067 in Table 3.5-1 includes Groups 1-5, 7, 9: Concrete: interior; above-grade exterior, Groups 1-3, 5, 7 concrete:

below-grade exterior; foundation, Group 6: concrete: all.

It appears that some Table 2 items associated with AMR 3.5.1-067 are missing. For example, Table 2 item associated with AMR 3.5.1-067 for concrete: basemat, foundation is missing in Table 3.5.2-4; Table 2 items associated with AMR 3.5.1-067 for concrete: exterior walls and roof (accessible) and interior walls and roof (accessible) are missing in Table 3.5.2-9, and so on.

The applicant is requested to evaluate all the Table 2 Evaluate Table 2 items associated with AMR 3.5.1-067 and provide missing Table 2 items.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions items associated with AMR item 3.5.1-067 to ensure that relevant Table 2 items are present in the SLRA.

17 Table 3.5.2-14 3.5-131 Table 2 item associated with AMR 3.5.1-072 lists roofing railroad bay as component in Table 3.5.2-14. AMR 3.5.1-072 is for seals and gasket moisture barriers, it is unclear to staff why AMR 3.5.1-072 is used for the roofing railroad bay.

Explain where roofing railroad bay is located, and what is its function and aging effect.

B.2.3.32 Masonry Walls Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.3.32 During breakout session on TRP 45 Question 1, the applicant responded that no masonry walls included in the scope of license renewal for the Emergency Filtration Train Building (EFB), and the AMP basis document will be revised during the implementation.

SLRA Section B.2.3.32, Masonry Walls, includes an enhancement (Enhancement

1) to the Scope of Program program element which update implementing procedure to include the inspection of masonry walls in the EFB and Radwaste Building. Since there is no masonry wall within the scope of the EFB as stated by the applicant in breakout, Enhancement 1 of the AMP and related commitment in SLRA Table A-3 also need to be amended accordingly.

2.3.3.5 Demineralized Water 2.3.3.11 Heating and Ventilation 2.3.4.1 Condensate Storage 2.3.4.2 Condensate and Feedwater

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Question Number LRA Section LRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

2.3.3.5 2.3-33 Scoping/Screening Boundary Drawing SLR-36159 Demineralizer System:

In SLR-36159 (C,1) line shows partially in-scope and partially out of scope. Piping connection from DWS to Off-gas extends beyond valve DM-100 and doesnt provide location of scope change in piping.

Provide location of end of 10CFR54.4(a)(2) Spatial/Structural classification.

2 2.3.3.11 2.3-48 Scoping/Screening Boundary Drawing SLR-36261 Heating and Ventilation System:

On SLR-36261, Service water connections (E,4) to AC-CHILLER (V-CH-1) show as out of scope. However, SLR-36041 (B,3) shows same piping as in-scope of 10CFR54.4 a(2) Spatial/Structural.

Confirm whether service water connections are in-scope 3

2.3.4.1 2.3-67 Scoping/Screening Boundary Drawing SLR-36039 Condensate Storage System:

Drawing SLR-36039 (B,3) shows HPCI pump return line (SC16-10-HB) from NH-36250 connecting drawing as within 10CFR54.4(a)(2) Spatial/Structural. However, portion of piping shows out of scope up to HK/HB interface.

Confirm whether piping up to HK/HB interface or up to the Reactor Building interface should be in-scope per 10CFR54.4(a)(2).

4 2.3.4.1 2.3-67 Scoping/Screening Boundary Drawing SLR-36247 Condensate Storage System:

Drawing SLR-36247 shows Condensate Storage (TEST) piping as GREEN for CST/RHR interface (B,5) at connection to drawing NH-36039 for CST (T-1A). CST piping shows GREEN and connecting RHR piping is RED.

However, same CST piping on drawing SLR-36039 (B,4) shows out-of-scope.

Confirm whether CST piping is in accordance with 10CFR54.4(a)(2) spatial/structural 5

2.3.4.2 2.3-70 SLRA Section 2.3.4.2 Condensate and Feedwater:

Confirm this should be drawing SLR-119259 Zinc Injection Passivation

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Section 2.3.4.2 Condensate and Feedwater of the SLRA (Page 2.3-70) lists drawing SLR-11929. However, this drawing was not provided.

System (GEZIP) as applicable to system.