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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

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 Please provide a detailed description of proposed how the RAMA methodology was applied values are conservative for the intended application. to determining fluence to all RVI components. Please include descriptions The licensee states in Section 4.2.1.2 the same guidance of the various conservatisms applied to as BWR145-VIP was applied for determining conservative justify use of RAMA code for Monticello 1 4.2.1.2 4.2-7 fluence all RVI components evaluated. SLRA.

For the RVI components the licensee states that the Please provide a brief description of maximum fast neutron fluence (E >1.0 MeV) is specifically geometric modeling approach for each of 4.2.1.1 reported for the various RVI components. The statement the components of the RPV and the RVI.

and 4.2-8 and is followed with list of components and the location within Include any modeling limitations, 2 4.2.1.2 4.2-9 the component where the maximum fluence is reported. 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed) 1 Section 2.3-45 SLRA Table 2.3.3-9 of the SLRA does not include the Verify whether the fire protection 2.3.3.9 following fire protection components: components listed are within the scope of

• halon bottles SLRA in accordance with 10 CFR 54.4(a)

• sprinkler and whether they are subject to an aging

• standpipe risers management review (AMR) in

• valve body accordance with 10 CFR 54.21(a)(1). If

• Stainer housing any of the listed components are not

• Filter housing 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 subject to an AMR, the staff requests that

• seismic support for standpipes system piping the applicant provide justification for the

• Intake traveling screen/trash rack exclusion.

• floor drains for removal of fire-fighting water

• curbs and dike for oil spill confinement

• station transformer fire suppression system and components 2 Section 2.4-1 SLRA Section 2.4, Scoping and Screening Results: The staff requested that the applicant 2.4 Structures, provides the scoping and screening results of provide details of fire barrier type in-various structures within the scope of license renewal and scope building structure of the plant in subject to an AMR. Further, SLRA Table 2.4-6 provide the accordance with 10 CFR 54.4(a), and results of scoping and screening of fire barriers. However, subject to an AMR, in accordance with 10 scoping and screening results do not provide the type of CFR 54.21(a)(1). For example, the staff fire barriers present in various in-scope building structures requested that the applicant provide a list of the plant. 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 2.4-14 SLRA Table 2.4-5 of the SLRA does not include the Verify whether the fire protection fire 2.4.6 following fire protection components. barriers listed are within the scope of

• Fire wrap/electric raceway fire barrier SLRA in accordance with 10 CFR 54.4(a)

• Radiant energy shield 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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 Discuss the relevance of the approved Z1 forward into the SPEO - the Z1 alternative approval in alternative during the SPEO - given that A.2.3.4. 2016 (see ML16208A462) to use BWRVIP-48-A in lieu of the approval in ML16208A462 is only ASME Section XI is permanent and future approvals are through the 5th ISI interval, which expired 1 B.2.3.4 no longer needed 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.

SLRA Section B.2.3.4 - Operating Experience stated the Were there any minor flaws or relevant following: conditions that were specifically related to Vessel ID Attachment Welds?

2 B.2.3.4 Plant-Specific Operating Experience

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

• Owners Activity Reports (OARs) were reviewed over the Or was this just meant to be a generic last six years (2015-2021). For all minor flaws or relevant statement that the licensee looked at the conditions that required evaluation or repair/replacement OARs/.?

activities for continued service, all items were either examined and evaluated as acceptable or appropriate replacements were completed.

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

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. Describe the plan to inspect the entirety

• Xcel Energy Change Request (CR) No. 608000000142 of the CST coatings to support describes a coating inspection evaluation for condensate subsequent license renewal and storage tank (CST) T-1A. CR No. 608000000150 operation past 2030. Be prepared to describes a coating inspection evaluation for CST T-1B. discuss:

• Both CRs 608000000142 and 608000000150 describe

• When these inspections will be coating blisters and pinpoint rusting andrecommend that performed B-207 to the entirety of the CST coatings will need to be inspected

• Inspection methods 1 B.2.3.28 B-212 if life is to be extended past 2030. Acceptance criteria B.2.2.1 - Fatigue Monitoring Question LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

SLRA Section B.2.2.1 addresses Enhancement 2 1. Explain why Enhancement 2 of the regarding the parameters monitored or inspected Fatigue Monitoring AMP and SLRA Table program element of the Fatigue Monitoring Aging 4.3.1-1 do not include plant loading and Management Program (AMP). unloading transients even though USAR Section 3.2.5 indicates that MNGP Unit 1 In the enhancement, the plant procedure for the AMP will performs load-following operation. As be updated to identify and require monitoring of the 80- part of the response, clarify whether the 1 B.2.2.1 B-24 year plant design cycles, or projected cycles that are 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 effect on the fatigue analyses. If so, applicable. The plant design transients are listed in SLRA discuss why the load-following operation Table 4.3.1-1. The SLRA table does not include plant has a negligible effect on the fatigue loading and unloading transients that may be associated analyses, including the evaluation of the with load-following operation. pressure and temperature variations during load-following operation and their In comparison, Monticello USAR Section 3.2.5, which effects on fatigue.

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.

1. Clarify whether the following fatigue parameters are included in the monitoring SLRA Section B.2.2.1 addresses Enhancement 4 and trending of Enhancement 4 to ensure regarding the monitoring and trending program element that their limits are not exceeded: (1) of the Fatigue Monitoring AMP. transient cycles used in the calculations of the 80-year projected cumulative In the enhancement, the plant procedure for the AMP will usage factor (CUF) and environmental be updated to require that trending be performed to CUF (CUFen); (2) transient cycles used ensure that the fatigue parameter limits will not be in the fatigue waiver evaluation; and (3) exceeded during the subsequent period of extended transient cycles used in the CUF operation. calculations for high energy line break location postulation (e.g., break location However, the SLRA does not clearly describe which postulation based on a CUF threshold of fatigue parameters will be monitored against their limits in 0.1). If these fatigue parameters are not B-25 the enhancement. The staff needs clarification on this included in Enhancement 4, explain why 2 B.2.2.1 4.3-4 item. 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

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.

Clarifying the following Table 2 items in In SLRA Sections 3.2.2.2.1, 3.3.2.2.1 and 3.4.2.2.1, the the SLR guidance documents are applicant explained that more detailed Table 2 applied to the components that are (subsystem-level) items related to these Table 1 items evaluated in the SLRA fatigue TLAAs: (1) are not listed in the SLRA because the TLAA disposition V.D2.E-10 related to Table 1 item 3.2-1, for these Table 1 items is 10 CFR 54.21(c)(1)(i) and no 001; (2) VII.E3.A-34, VII.E3.A-62 and aging management program is credited for these items. VII.E4.A-62 related to Table 1 item 3.3-1, Examples of subsystems in the context of this discussion 002; and (3) VIII.B2.S-08 and VIII.D2.S-are the core spray system and high pressure coolant 11 related to Table 1 item 3.4-1, 001. If injection system in the ESF. so, revise the SLRA to identify these Table 2 items or similar items to clarify 3.2.2.2.1 3.2-8 However, GALL-SLR and SRP-SLR (i.e., SLR guidance the applicability of the Table 1 items and 3.3.2.2.1 3.3-21 documents) identify the following Table 2 (subsystem- metal fatigue TLAAs to specific 1 3.4.2.2.1 3.4-8 level) items in relation to the aging management and 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.

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.

1. Provide justification for why the SLRA Section 3.5.2.2.2.5 also indicates that Table 3.5-1, existing fatigue analysis for the bottom item 3.5.1-053 related to the further evaluation discussed head support skirt of the reactor pressure above is not applicable to the Monticello Nuclear vessel is not evaluated in SLRA Section Generating Plant (MNGP). The SLRA further states that 3.5.2.2.2.5. If justification cannot be the only fatigue analysis related to plant structures is the provided, revise the section to include analysis for cranes and lifting devices and for portions of the evaluation of the fatigue TLAA for the 2 3.5.2.2.2.5 3.5-35 the primary containment. In addition, the SLRA states 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

SLRA Section B.2.3.21 states [t]he MNGP Selective The staff did not identify AMRs Leaching AMP includes inspections of components made associated with copper alloys that contain of gray cast iron, ductile iron, and copper alloys (except greater than 8 percent aluminum. The for inhibited brass) that contain greater than 15 percent staff requests clarification that there are zinc or greater than 8 percent aluminum [emphasis added no copper alloy that contain greater than by staff] exposed to a raw water, closed-cycle cooling 8 percent aluminum components in 1 B.2.3.21 B-161 water, treated water, waste water, or soil environment. scope for subsequent license renewal.

The staff reviewed AR 01228174 (last page) and WO The staff did not identify AMRs 00414130 (pages 29 and 58 of 110) and noted references associated with ductile iron piping. The 2 N/A N/A to ductile iron piping. 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.

The staff did not identify AMRs associated with ductile iron piping. The The staff reviewed AR 01228174 (last page) and WO staff requests a discussion with respect 00414130 (pages 29 and 58 of 110) and noted references to if there is ductile iron piping in-scope 3 N/A N/A to ductile iron piping. for subsequent license renewal.

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 Based on recent industry operating involving the loss of function of buried gray cast iron fire experience, the staff requests a water main yard piping due to multiple factors, including discussion with respect to using the graphitic corrosion (i.e., selective leaching), reduced sample size (i.e., 3 percent with overpressuration, low cycle fatigue, and surface loads. As a maximum of 10 components) for gray 4 B.2.3.21 B-161 noted in the IN, a contributing cause to the failures of 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.

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 Based on its audit and review of the MNGP supports the exclusion of buried components that SLRA, the staff noted instances of buried are externally coated in accordance with GALL-SLR piping coating failures. The staff requests Report AMP XI.M41 for selective leaching. a discussion with respect to excluding buried components from the scope of the SLRA Section B.2.3.27, Buried and Underground Piping Selective Leaching program based on and Tanks, states [t]he most recent annual cathodic this observation.

protection system survey performed in 2021 determined that not all of the surveyed locations met the -850 mV Based on the past performance of the polarized potential criterion for buried steel cathodic protection system, the staff components[t]herefore, the cathodic protection system requests a discussion with respect to does not currently meet the acceptance criteria of NACE excluding buried components from the SP0169-2007 or NACE RP0285-2002 and is not credited scope of the Selective Leaching program 5 N/A N/A as a preventive measure at MNGP. based on cathodic protection efficacy.

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

SLRA Section 4.3.1 addresses the 80-year transient cycle 1. Clarify whether the Fatigue Monitoring projections. AMP will monitor the cycles of the following transients for the subsequent The SLRA section states that the Fatigue Monitoring period of extended operation: (1) reactor program data does not list the following transients from overpressure at 1375 psig transient; (2) 1 4.3.1 4.3-1 the USAR list: (1) reactor overpressure at 1375 psig 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) (3) rapid blowdown transient; (4) liquid rapid blowdown transient; (4) liquid poison flow at 80F poison flow at 80F transient; (5) transient; (5) operating basis earthquake (OBE) operating basis earthquake (OBE) transient; and (6) safety/relief valve actuations transient. transient; and (6) safety/relief valve actuations transient. If not, explain why The staff needs clarification as to whether the above the cycle counting of the transients is not statement in SLRA Section 4.3.2 means that these necessary to manage the aging effects of transients will not be monitored in the Fatigue Monitoring cumulative fatigue damage.

Aging Management Program (AMP).

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.

SLRA Section 4.3.1 addresses the 80-year transient cycle 1. Describe the technical basis for the projections. The SLRA section explains that the transient applicants approach that uses the most cycle projections are based on the cycle accumulation recent 10-year cycle accumulation rates rates of the most recent 10-year evaluation period up to for cycle projections but does not May 31, 2021 (i.e., the evaluation period of June 1, 2011 consider the full cycle accumulation rates through May 31. 2021). observed since the start of the operation.

However, the applicant did not clearly address why the 2. Provide clarification on the following cycle projections do not consider the full cycle items for the safety/relief valve lifts accumulation rates observed since the start of the plant transient: (1) why the cycle accumulation operation. The staff needs information on the technical rate used in cycle projections is basis of the applicants approach that uses only the most significantly lower than the full cycle recent 10-year cycle accumulation rates (e.g., the most accumulation rate observed since the recent 10 years of operation involve distinctive and stable start of the operation through May 31, cycle accumulation rates that can better represent the 2021; and (2) whether the most recent operational characteristics of the subsequent period of 10-year operation time period up to May extended operation rather than the prior cycle 31, 2021 represents the operating accumulation rates). characteristics for the subsequent period of operation in terms of cycle 2 4.3.1 4.3-1 The staff also noted that the safety/relief valve lifts 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, above, if the most recent 10-year 2021, as described in SLRA Table 4.3.1-1. In operation period does not represent the comparison, the 80-year projected cycle number of this operating characteristics for the transient is only 699. subsequent period of operation, explain why the cycle accumulation rate of the Considering that the operation of the plant started on most recent 10-year operation time September 30, 1970, the operation time period through period is used in the cycle projections the end of the cycle evaluation period (May 31, 2021) is rather than the full cycle accumulation approximately 51 years. The additional operating time rate observed since the start of the plant period following May 31, 2021 through 80 years of operation.

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.

SLRA Section 4.3.1 addresses the 80-year transient cycle 1. Clarify whether the following non-projections. Specifically, SLRA Table 4.3.1-1 describes USAR-listed transients have an impact 3 4.3.1 4.3-1 the 80-year transient cycle projections. The cycle 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 discussed in SLRA Section 4.3.2: (1) described in the following document (

Reference:

SIA sudden start transient; (2) hot standby Calculation 2100507.301, Revision 0, 80-Year Cycle with drain shutoff transient; (3) core Projections for Monticello). spray injection transient; and (4) operating basis earthquake (OBE)

SLRA Table 4.3.1-1 indicates that the following transients transient. If so, discuss the impact in are not listed in the USAR and accordingly USAR does terms of the continued validity of the not define a design cycle limit for these transients: (1) fatigue waiver evaluations for the sudden start transient; (2) hot standby with drain subsequent period of extended operation.

shutoff transient; (3) core spray injection transient; and If not, provide the technical basis for why (4) operating basis earthquake (OBE) transient. these transients do not have an impact on the fatigue waiver evaluations SLRA Table 4.3.1-1 also indicates that these transients described in SLRA Section 4.3.2.

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.

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).

1. Describe the basis for why the The staff needs to clarify the basis for why the applicant applicant did not identify a TLAA did not identify a TLAA disposition for the 80-year disposition for the 80-year transient cycle 4 4.3.1 4.3-1 transient cycle projections. 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

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

The staff needs to resolve these inconsistency regarding the TLAA dispositions. 2. Clarify whether the design limits referenced in the TLAA disposition mean In addition, the text below the paragraph heading for the transient cycles used in the fatigue TLAA disposition states that the Fatigue Monitoring AMP waiver evaluations that are described in will monitor the transient cycles which are the inputs to the SLRA Section 4.3.2. If not, provide fatigue waiver reevaluations and require action prior to additional information to clarify what the exceeding design limits that would invalidate their design limits refer to. As part of the conclusions. response, the Fatigue Monitoring AMP will include the acceptance criteria for the The staff needs clarification on whether the design limits transient cycles used in the fatigue 1 4.3.2 4.3-6 referenced in the paragraph discussed above mean the 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.

SLRA Section 4.3.2 describes the time-limited aging 1. Explain why SLRA Section 4.3.2 analysis (TLAA) on ASME Code Section III, Class 1 identifies fatigue waiver TLAAs for the fatigue waiver evaluations. SLRA Section 4.3.2 indicates instrumentation nozzles and jet pump that the following reactor pressure vessel components instrumentation nozzles even though have existing (original) fatigue waiver evaluations: (1) there is no existing fatigue waiver main closure flange, (2) head cooling spray and evaluation for these nozzles in the instrumentation nozzles; (3) vent nozzle; (4) following references for design stress instrumentation nozzles; (5) jet pump instrumentation analyses (

References:

(1) CA-68-667, nozzles; (6) intermediate range monitor/source range Revision 1, Section S14, Stress Analysis monitor (IRM/SRM) dry tube; (7) power range detector of Instrumentation Nozzles N11A-B and assembly; and (8) in-core detector assembly. N12A-B, Monticello-NSP Reactor Vessel, 3/12/1993; and (2) CA-68-668, However, the staff noted that the design stress analyses Revision 1, Section S15, Stress Analysis for the instrumentation nozzles (N11A-B and N12A-B of Jet Pump Instrumentation, Nozzles nozzles) and jet pump instrumentation nozzles (N8A-B N8A & B, Monticello-NSF Reactor nozzles) in the following references do not include a Vessel, 3/12/1993).

fatigue waiver evaluation or cumulative usage factor calculation for these nozzles (

References:

(1) CA-68-667, 2. If a fatigue waiver evaluation exists for Revision 1, Section S14, Stress Analysis of the instrumentation nozzles and jet pump Instrumentation Nozzles N11A-B and N12A-B, Monticello- instrumentation nozzles in the current NSP Reactor Vessel, 3/12/1993; and (2) CA-68-668, licensing basis of the Monticello Nuclear Revision 1, Section S15, Stress Analysis of Jet Pump Generating Plant (MNGP), provide a Instrumentation, Nozzles N8A & B, Monticello-NSF document reference for the evaluation Reactor Vessel, 3/12/1993). and discuss the technical basis for the TLAA disposition (e.g., why the fatigue Therefore, the staff needs clarification on why SLRA waiver evaluation remains valid with the Section 4.3.2 identifies fatigue waiver TLAAs for these 80-year cycle projections).

nozzles even though there is no existing fatigue wavier evaluation (e.g., evaluation per ASME Code Section III, 3. In addition, clarify whether the jet NB-3222.4(d)) for these nozzles. pump instrumentation nozzles are bounded by the steam outlet nozzle in In addition, the following reference indicates that the jet terms of CUF. If so, provide the technical pump instrumentation nozzles are exposed to the same basis of the bounding nature of the steam 2 4.3.2 4.3-6 transient of cooling from normal operating temperature as 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 instrumentation nozzles.

Associates, Calculation 2100507.309, Revision 0, Fatigue Exemption of Monticello Reactor Vessel (RPV) 4. Clarify whether the instrumentation Components, Section 4.4). SLRA Table 4.3.3-1 also nozzles are bounded by the other RPV indicates that the 80-year projected cumulative usage nozzles evaluated in SLRA Table 4.3.3-1 factor (CUF) for the steam outlet nozzle is 0.1872. in terms of CUF. If so, provide the technical basis of the bounding nature of The staff needs clarification on whether the jet pump the other RPV nozzles for the instrumentation nozzles are bounded by the steam outlet instrumentation nozzles.

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.

SLRA Section 4.3.2 describes the time-limited aging 1. Describe why the existing fatigue analysis (TLAA) on ASME Code Section III, Class 1 waiver evaluations remain valid for the fatigue waiver evaluations. subsequent period of the extended operation for the head cooling spray and SLRA Table 4.3.2-1 describes the numbers of transient instrumentation nozzles (N6A and N6B cycles used in the existing fatigue waiver evaluations and nozzles) and vent nozzle (N7 nozzle). As the 80-year projected cycles used in the fatigue waiver part of the response, clarify whether an TLAA. Specifically, SLRA Section 4.3.2 explains that the aging management program can manage applicant used the 80-year projected transient cycles to the effects of cumulative fatigue damage confirm that the existing fatigue evaluations remain valid in relation to the fatigue waiver for the subsequent period of extended operation for the evaluations for these nozzles.

following components: (1) main closure flange, (2)

IRM/SRM dry tube, (3) power range detector assembly 2. Clarify whether the following non-and (4) in-core detector assembly. USAR-listed transients have an impact on the existing fatigue wavier evaluations However, SLRA Section 4.3.2 does not clearly discuss discussed in SLRA Section 4.3.2: (1) why the existing fatigue waiver evaluations remain valid sudden start transient; (2) hot standby for the subsequent period of the extended operation for with drain shutoff transient; (3) core the head cooling spray and instrumentation nozzles (N6A spray injection transient; and (4) and N6B nozzles) and vent nozzle (N7 nozzle). operating basis earthquake (OBE) transient. If so, discuss the impact on the 3 4.3.2 4.3-6 In addition, SLRA Table 4.3.1-1 indicates that 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 If not, provide the technical basis for why accordingly USAR does not define a design cycle limit for these transients do not have an impact these transients: (1) sudden start transient; (2) hot on the validity of the fatigue waiver standby with drain shutoff transient; (3) core spray evaluations.

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.

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 1. Clarify why different cycles of the loss fluctuation cycles. For example, additional 18 cycles of of feedwater pumps transient are the transient are estimated for the power range detector estimated for the components discussed assembly, compared to those for the IRM/SRM dry tube, in the issue section as significant in the fatigue waiver evaluations for 80 years of operation. pressure fluctuation cycles in the fatigue 4 4.3.2 4.3-6 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

SLRA Section 4.3.4 addresses the time-limited aging 1. Describe how the applicant used 40-analysis (TLAA) on the fatigue of reactor pressure vessel year transient cycles and their internal (RVI) components. SLRA Section 4.3.4 explains contributions to the limiting CUF of the jet that the most significant fatigue loading occurs at the jet pump location in the determination that pump diffuser to baffle plate weld location and, therefore, the 60-year limiting CUF is conservatively this location is bounding for all other fatigue affected RVI bounding for the 80-year CUF components. calculation.

The applicant indicated that the original 40-year design 2. In related to the first request above, analysis for this limiting jet pump location estimated a 40- clarify the following items: (1) actual year cumulative usage factor (CUF) of 0.35. By the (current) cycles and 80-year projected extrapolation of this 40-year CUF, the applicant cycles of the transients used in the CUF determined that the 60-year projected CUF for the limiting calculation for the limiting jet pump jet pump location is approximately 0.5. location; (2) which transient in SLRA Table 4.3.1-1 is identical to the sudden The applicant also explained that the 60-year CUF value start of cold pump transient in the GE in the current licensing basis is conservatively bounding report; and (3) whether HPCI startup for the 80-year CUF of the limiting jet pump location. and the sudden start of cold pump transients are transients in the However, the applicant did not clearly discuss why the 60- emergency condition.

year projected CUF for the jet pump diffuser to baffle plate weld location, which is based on the extrapolation of the 3. If the 60-year CUF value (0.5) for the 40-year CUF, is bounding for the 80-year CUF calculation limiting location is not large enough to (e.g., discussion that supports the bounding nature of the bound the 80-year CUF calculation, 60-year cycles assumed in the 60-year CUF calculation in revise the 80-year CUF value and the comparison with 80-year projected cycles of relevant TLAA disposition in accordance with 10 transients). CFR 54.21©(1)(ii) or (iii).

The staff noted that the following reference describes the 4. Discuss whether the other RVI transients used in the original 40-year CUF calculation for components may have a 80-year CUF 1 4.3.4 4.3-12 the limiting jet pump location (

Reference:

General Electric 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 pump location because of additional Electric Boiling Water Reactor Jet Pumps, September transient cycles projected to be 1968). accumulated for 80 years of operation.

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.

SLRA Section 4.3.4 addresses the fatigue TLAA for the 1. Clarify why Table 2.2-9 in the GE reactor pressure vessel internal (RVI) components. SLRA report does not identify the jet pump Section 4.3.4 explains that the most significant fatigue diffuser to baffle plate weld location as loading occurs at the jet pump diffuser to baffle plate weld the most limiting CUF location for RVI location and, therefore, this location is bounding for all components. As part of the response, if other fatigue affected RVI components. Table 2.2-9 in the GE report has a minimum CUF threshold for the CUF In comparison, the following GE report addresses the listing in the table, describe such a fatigue analysis RVIs in relation to the extended power minimum CUF threshold value.

uprate (EPU) that was incorporated into the current licensing basis of the Monticello Nuclear Generating Plant 2. Clarify whether the feedwater sparger in 2013 (

Reference:

GE Report NEDO-33322, Revision 3, and jet pump diffuser to baffle plate weld Safety Analysis Report for Monticello Constant Pressure locations are bounding for the other RVI Power Uprate, October 2008). components in terms of 80-year projected 2 4.3.4 4.3-12 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 limiting CUF locations and their 80-year 60-year cumulative usage factor (CUF) of the feedwater projected CUF values based on the most sparger is estimated to be 0.32. recent CUF analyses.

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.

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

SLRA Section 4.3.5 addresses the time-limited aging 1. Resolve the apparent inconsistency analysis (TLAA) on the fatigue of ASME Code Section III, between the SIA report and SLRA Class 1 piping systems. In relation to the fatigue TLAA, Section 4.3.5 in terms of the EPU the following reference describes that cumulative fatigue condition evaluated in the fatigue TLAA usage (CUF) analysis for the Class 1 piping systems in (i.e., 549 °F versus 546 °F in the thermal detail (

Reference:

Structural Integrity Associates (SIA), transients, and 1025 psi versus 1000 psi Calculation Package 2100507.303, Revision 0, 80-year in the pressure levels).

Fatigue Analysis of Selected Class 1 Reactor Coolant Pressure Boundary (RCPB) Piping, June 24, 2022). 2. Clarify the following items: (1) whether the original condition in the SIA report Section 4.2 of the SIA report discusses the fatigue means the pre-EPU condition; (2) the analysis for the core spray line. The section indicates that relationship between the rerate condition the analysis in the report considered the original condition and the EPU condition (e.g., whether the as well as the rerate condition that bounds the rerate condition is the term used to temperature and pressure conditions of the EPU. represent the bounding condition for the EPU); and (3) whether the rerate With respect to the core spray line, Table 4 of the SIA condition is part of the EPU license report indicates that the thermal transient of the original amendment approved on December 9, condition is 546 °F to 80 °F in comparison with the 2013 (ADAMS Accession No.

thermal transient of the rerate condition, 549 °F to 80 °F. ML13316B298).

In addition, the table indicates that the original condition for pressure is 1000 psi in comparison with the rerate 3. Clarify whether the fatigue TLAA in 1 4.3.5 4.3-13 pressure of 1025 psi. SLRA Section 4.3.5 evaluates the

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions bounding pressure and temperature Based on these data, the applicant determined a pressure conditions of the EPU. If not, provide ratio of 1.025 for the implementation of the EPU (i.e., justification for why the fatigue TLAA 1025/1000). does not evaluate the bounding conditions of the EPU.

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

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

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 1. Given the relatively large reduction in cycles accumulated before piping replacement and the the bounding CUF value for the EPU condition. In addition, the SLRA section indicates recirculation and RHR piping systems that the 80-year CUF was estimated to be 0.399 in the (from 0.923 to 0.399), clarify the following reanalysis for 80 years of operation. items: (1) the operating time period for which the previous CUF value of 0.923 Given the relatively large reduction in the bounding CUF was determined (e.g., 60 years or 40 value for the recirculation and RHR piping systems (from years); (2) how the applicant removed the 0.923 to 0.399), the staff needs clarification on the conservatism from the previous CUF following items: (1) the operating time period for which the (0.923) in addition to the removal of the previous CUF value of 0.923 was determined (e.g., 60 cycles accumulated before the piping years or 40 years); (2) how the applicant removed the replacement; and (3) whether the Fatigue conservatism from the previous CUF (0.923) in addition to Monitoring AMP will monitor, if any, the removal of the cycles accumulated before the piping reduced cycles used in the 80-year CUF replacement; and (3) whether the Fatigue Monitoring analysis, as opposed to the original aging management program (AMP) will monitor, if any, design cycles, in case such reduced reduced cycles used in the 80-year CUF analysis, as cycles are assumed in the 80-year CUF 2 4.3.5 4.3-13 opposed to the original design cycles in case such 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.

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. 1. Describe more specific references to the Code provisions (e.g., paragraphs)

The staff needs more specific references to the Code that the applicant used in the fatigue 3 4.3.5 4.3-13 provisions that the applicant used in the fatigue TLAA. TLAA.

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

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 1. Describe how the 40-year cycles were range transient cycles for non-Class 1 piping systems and determined (e.g., based on piping system extrapolates the 40-yer cycles to estimate the 80-year design information, plant operation projected cycles. In turn, the 80-year cycle numbers are procedures, test requirements, UFSAR compared to the 7000 cycle limit in the implicit fatigue information and specific system-level analysis. knowledge).

However, LRA Table 4.3.6-1 does not clearly describe 2. Clarify whether the following 40-year how the 40-year cycles were determined (e.g., piping design cycles were estimated by system design information, plant operation procedures, summing up two or more design cycles test requirements, UFSAR information and specific for each of the non-Class 1 piping system-level knowledge). systems: (1) 1500 cycles for the feedwater piping; (2) 532 cycles for the SLRA Table 4.3.6-1 also includes the following 40-year nuclear boiler system; and (3) 205 cycles design cycles: (1) 1500 cycles for the feedwater piping; (2) for the reactor recirculation system. If so, 532 cycles for the nuclear boiler system; and (3) 205 describe those design cycles for each 1 4.3.6 4.3-14 cycles for the reactor recirculation system. The staff 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.

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).

1. Provide justification for why SLRA SA may need to be adjusted by a stress range reduction Section 4.3.6 does not identify the HELB factor based on the number of transient cycles that are analysis as a TLAA even though the evaluated in the implicit fatigue analysis (SLRA Section screening criteria of the HELB location 4.3.6). postulation involves the dependency on the transient cycles. If justification cannot However, SLRA Section 4.3.6 does not identify the HELB be provided, identify the HELB analysis analysis as a TLAA based on the HELB location as a TLAA and provide the disposition of postulation that involves SA and the associated cycle- the TLAA. In addition, revise the USAR 2 4.3.6 4.3-14 dependent stress range reduction factor. supplement as needed.

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

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

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 SLRA Section 4.3.7 addresses the environmentally- time periods. In addition clarify whether assisted fatigue (EAF) time-limited aging analysis (TLAA). the time periods, for which the HWC operation is not available, assume to The SLRA section indicates that the EAF screening have the NWC condition.

evaluation to determine the limiting locations (also called sentinel locations) uses bounding environmental fatigue 3. Clarify whether the maximum correction factor (Fen) based on material types. temperature referenced in relation to the Fen calculations (in the first sentence on However, the SLRA sections does not clearly describe page 4.3-19) means the maximum 1 4.3.7 4.3-19 how the applicant calculated the bounding Fen values. service temperature of each component.

SLRA Section 4.3.7 addresses the environmentally- 1. Describe the following information to assisted fatigue (EAF) time-limited aging analysis (TLAA). confirm that the screening CUFen threshold of 1.0 is low enough to be used The SLRA section indicates that the limiting locations in the screening evaluation: (1) the described in NUREG/CR-6260 for older vintage General screening CUFen value of the Electric BWR plants are evaluated in the EAF analysis. recirculation outlet nozzle and (2) the The locations in NUREG/CR-6260 include the highest screening CUFen value in the 2 4.3.7 4.3-19 recirculation outlet nozzle, as also indicated in the SLRA. EAF screening evaluation for the

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions fabrication material of the recirculation However, SLRA Section 4.3.7 does not list the outlet nozzle and the associated recirculation outlet nozzle as one of the NUREG/CR-6260 component.

locations. Instead, the SLRA section identifies the recirculation outlet nozzle location as one of the additional 2. Clarify whether the screening EAF plant-specific evaluation locations subject to EAF evaluation of the recirculation nozzle screening. The SLRA also explains that the recirculation includes both the nozzle body and the outlet nozzle is screened out because its bounding adjacent piping location (e.g., safe end or screening CUFen (also called screening Uen) is less than safe end weld). If not, explain why both the screening threshold of 1.0. the nozzle and the adjacent piping locations are not evaluated in the EAF The staff needs the following information to confirm that screening. As part of the response, the screening CUFen value of 1.0 is low enough to be describe the fabrication materials of the used in the EAF screening evaluation: (1) the screening nozzle body and the adjacent safe end CUFen value of the recirculation outlet nozzle and (2) the and weld.

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).

SLRA Section 4.3.7 addresses the environmentally- 1. Describe the following items regarding assisted fatigue (EAF) time-limited aging analysis (TLAA), the EAF screening evaluation: (1) how including the EAF screening evaluation to determine the the applicant determined thermal zones limiting EAF locations. or sections that group certain components and piping lines for proper However, the LRA does not clearly describe the following comparisons of the screening CUFen items related to the screening evaluation: (1) how the values considering the applicable applicant determined thermal zones or sections that group transient conditions; (2) whether the certain components and piping lines for proper limiting location is determined for each comparisons of the screening CUFen values considering material type (e.g., stainless steel, nickel the applicable transient conditions; (2) whether the limiting alloy and carbon steel); and (3) how the location is determined for each material type (e.g., applicant compared the highest values of 3 4.3.7 4.3-19 stainless steel, nickel alloy, and carbon/low alloy steel); 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 final limiting locations (e.g., screening the screening CUFen values to determine the final limiting process when the highest CUFen values locations (e.g., how the limiting locations were determined are close to each other in a thermal when the highest CUFen values are close to each other in zone).

a thermal zone).

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 1. Describe how the applicant removed Table 3.4.7-1. However, the SLRA does not clearly the conservatisms from the screening describe how the conservatisms were removed from the CUFen values to determine the refined screening CUFen values to determine the 80-year CUFen values listed in SLRA Table 4 4.3.7 4.3-19 projected CUFen values. 3.4.7-1.

SLRA Section 4.3.7 addresses the environmentally- 1. Clarify whether the applicant confirmed assisted fatigue (EAF) time-limited aging analysis (TLAA). that the use of the design cycles (20 cycles) of the loss of recirculation The following reference discusses the EAF analysis for pumps transient is adequate for the EAF the NUREG/CR-6260 locations applicable to the analysis. As part of the response, Monticello Nuclear Generating Plant, which is a older provide the actual cycles and 80-year vintage GE plant (

Reference:

Structural Integrity projected cycles of the transient to Associates (SIA), Calculation Package No. demonstrate the bounding or 2100507.305P, Monticello Environmentally Assisted representative nature of the design Fatigue Analysis for 80 Years, NUREG/CR-6260 cycles (i.e., 20 cycles) for the 80-year Locations, June 24, 2022). projected cycles of the transient.

Section 3.1 of the SIA report indicates that, for the loss of 2. In addition, provide justification for why recirculation pumps transient, the design cycle number SLRA Table 4.3.1-1 does not list the loss (i.e., 20 cycles) is used in the EAF analysis for 80 years of of recirculation pumps transient and operation. Based on the use of the design cycles in the related cycle information even though the EAF analysis for the recirculation and residual heat SIA report uses the transient in the EAF 5 4.3.7 4.3-19 removal (RHR) return piping, the SIA report recommends 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 revie SLRA Table 3.4.1-1 to include the the transient cycle number used in the EAF analysis. transient and the related cycle information, consistent with the existing SLRA Section 4.3.7 does not clearly indicate whether the format of the table.

applicant confirmed that the use of the design cycles (20 cycles) of the loss of recirculation pumps transient is 3. In addition, clarify whether the Fatigue adequate for the EAF analysis. Monitoring aging management program (AMP) will monitor the loss of In addition, SLRA Table 4.3.1-1, which describes the 80- recirculation pumps transient cycles for year projected cycles for fatigue analyses, does not list the subsequent period of extended the loss of recirculation pumps transient even though the operation. If not, provide justification for SIA report uses the transient in the EAF TLAA. why such monitoring of the transient is not needed for the transient cycles.

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

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 SLRA Section B.2.3.27, Buried and Underground Piping Category C). For instances where and Tanks, states the following: cathodic protection is not credited as a preventive measure, the staffs

• [t]he number of inspections for each 10-year inspection expectation would be that the applicant period, commencing 10 years prior to the start of SPEO, would state an exception and develop are based on the inspection quantities noted in NUREG- plant specific inspection quantities. The 2191, Table XI.M41-2 for Category F. staff requests a discussion with respect

• [t]he cathodic protection system does not currently meet to why Preventive Action Category F is the acceptance criteria of NACE SP0169-2007 or NACE appropriate for steel piping at Monticello, B-197 RP0285-2002 and is not credited as a preventive measure given that cathodic protection is not 1 B.2.3.27 B-198 [emphasis added by staff] at MNGP. credited as a preventive measure.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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 The table on this page includes inspection quantities for without measurement or calculation of 2 B.2.3.27 B-198 underground steel and stainless steel piping. voltage drops.

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 SLRA Section B.2.3.27 states [t]he most recent annual mixed metal environment) without cathodic protection system survey performed in 2021 confirmatory testing to verify that all determined that not all of the surveyed locations met the metals are adequately protected. In 850 mV polarized potential criterion for buried steel addition, the staff does not agree with the components. The survey also determined that not all of use of a potential of -850 mV instant on B-197 the surveyed locations met the 100 mV polarization without measurement or calculation of 3 B.2.3.27 B-198 criterion. voltage drops.

UFSAR Section 11.6, Cooling Tower System, states [a] Staff request a discussion with respect to single underground steel pipe conveys the water from if this piping is in-scope for subsequent 4 N/A N/A both pumps to two cooling towers. license renewal.

Item 3.3.1-108 is classified as N/A; however, SLRA Table 3.3.2-17 includes SLRA Table 3.3-1, Summary of Aging Management stainless steel tanks exposed to Evaluations for the Auxiliary Systems, item 3.3.1-108 concrete. Staff requests a discussion with Table 3.3 (which includes stainless steel tanks exposed to concrete) respect to why item 3.3.1-108 is 5 1 3.3-54 is classified as not applicable. classified as N/A.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions GALL Report AMP XI.M41 recommends a cathodic protection critical potential of 1,200 mV to prevent The staff requests a discussion with damage to coatings or base metals. respect to cathodic protection critical potentials and the subject observation The staff reviewed SL-008367 and noted pipe to soil from SL-008367. The staff could not potential measurements as negative as -23,921 mV with identify this GALL-SLR Report respect to a zinc reference cell between the years 1976 recommendation in current procedures or 6 N/A N/A and 2003. in any of the enhancements.

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 The staff requests a discussion with due to salt treatment during the winter months. respect to if the subject piping inspection, or any other piping inspections, have be The staff reviewed AR 01383079 and noted the coating conducted in the vicinity of where salt inspection of the buried stainless steel piping revealed treatments (i.e., chlorides) have been 7 B.2.3.27 B-205 coating failures but no signs of degradation. applied.

Please upload the original construction specification(s) for external coatings for The staff reviewed AR 01240741 and noted the following: in-scope metallic piping (carbon steel, (a) a high temperature line from the heating boiler to the gray cast iron, and stainless steel). In cold shop going into the ground was not visibly coated; addition, the staff requests a discussion and (b) site specification for coating and wrapping with respect to if there are any uncoated underground piping is only applicable to lines with a high temperature lines in-scope for 8 N/A N/A maximum fluid temperature of 160 Fahrenheit. subsequent license renewal 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 Procedure 1404-01 does not appear to interval not to exceed 5 years. The piping is pressurized to meet the pressure testing guidance 110 percent of the design pressure of any component outlined in AMP XI.M41. The staff seeks within the boundary (not to exceed the maximum a clarifying discussion with respect to if allowable test pressure of any nonisolated components) pressure testing will be used as an with test pressure being held for a continuous eight hour alternative to visual examinations (and if interval. so, whether pressure testing will be conducted in accordance with AMP 9 N/A N/A XCELMO00017-REPT-072 (Buried and Underground 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.

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

10 N/A N/A Please upload this document to the ePortal. N/A 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.

Although new backfill will be consistent The staff reviewed MPS-0984 and noted the maximum with AMP XI.M41 guidance, existing size of structural backfill shall be two inches in confined backfill does not appear to meet areas where hand tamping is required and four inches in guidance outlined in AMP XI.M41. The 11 B.2.3.27 B-199 other areas. staff requests a discussion on this topic.

The staff did not identify any enhancements related cathodic Section 4.10, Operating Experience, of XCELMO00017- protection system refurbishment or REPT-072 (Buried and Underground Piping and Tanks acceptance criteria. In addition, the staff program basis document) states MNGP will refurbish its did not identify any commitments related cathodic protection system 5 years prior to the SPEO to to cathodic protection system meet the acceptance criteria of -850 mV relative to a CSE refurbishment or acceptance criteria in (instant off), or acceptance criteria alternatives, for buried SLRA Table A-3, List of SLR 12 N/A N/A and underground steel components. 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.

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 Flow testing frequency in procedure 0268 procedure 0268, Fire Protection System Flow Test, is longer than AMP XI.M41 Revision 24, which specifies flow tests are to be recommendations. The staff requests a 13 N/A N/A performed every three years. discussion on this topic.

The staff did not identify any enhancements in SLRA Section B.2.3.27 The acceptance criteria program element of GALL-SLR associated with the acceptance criteria Report AMP XI.M41 includes recommendations related to for fire main flow tests, jock pump the following: monitoring, fire water system leak rate testing, or cathodic protection surveys.

• (g) flow tests for fire mains The staff request a discussion with

• (i) jockey pump activity respect to why there are no

• (j) fire water system leak rate testing enhancements related to these four 14 B.2.3.27 B-201 * (l and m) cathodic protection acceptance criteria recommendations from AMP XI.M41.

The corrective actions program element of GALL-SLR The staff did not identify any Report AMP XI.M41 includes recommendations related to enhancements in SLRA Section B.2.3.27 the following: associated with the corrective actions for cathodic protection survey results, B-202 * (d) cathodic protection survey results leakage during pressure tests, jock pump 15 B.2.3.27 B-203 * (e) leakage during pressure tests monitoring, or indications of cracking.

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

• (f) jockey pump monitoring The staff request a discussion with

• (g) indications of cracking respect to why there are no enhancements related to these four recommendations from AMP XI.M41.

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 The staff noted the following than -850 mV instant off, loss of material rates are recommendations from GALL SLR measured. Report Table XI-01 (cathodic protection

• [i]f a reduction in the number of inspections criteria and soil corrosivity testing) are not recommended in GALL-SLR Report, AMP XI.M41, Table included in SLRA Section A.2.2.27, XI.M41-2 is claimed based on a lack of soil corrosivity as Buried and Underground Piping and determined by soil testing, then soil testing is conducted Tanks. The staff requests a clarifying A-26 once in each 10-year period starting 10 years prior to the discussion to understand the basis for 16 A.2.2.27 A-27 subsequent period of extended operation. excluding these recommendations.

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

• 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 Describe plant-specific actions Number ML21084A164. Subsequent to submitting responding to the identified EPRI letters BWRVIP-315 for NRC review, EPRI submitted letter dated and ensuring that the applicants aging January 20, 2022, to respond to staff concerns related to management program for vessel internals the new data and its impact on the BWRVIP-315 topical accounts for the latest available 1 B.2.3.7 B-60 report (ADAMS Accession Number ML22025A113). The 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.

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 Describe licensee implementation of latest version of Regulatory Guide 1.147 incorporated by Code Case N-889 to calculate crack reference to Title 10 of the Code of Federal Regulations growth rate, given the staffs conditions 2 B.2.3.7 B-60 50.55a. on the code case.

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 Describe and justify reinspection plans 3 B.2.3.7 B-60 lower plenum components. for lower plenum components.

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 Describe how the licensee will implement 4 B.2.3.7 B-60 BWRVIP-315 contains limitations in Section 4.5.1. the identified limitations in BWRVIP-315.

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

The staff reviewed AR 01396233 and noted the following: The staff requests a discussion with 3.3-26 (a) degradation of the rx/rad waste chilled water piping is respect to if the subject operating 1 3.3.2.2.7 3.3-27 a long standing issue; (b) localized repairs are unlikely to 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) corrosion. It is not clear to the staff if this chilled water piping located in the reactor building is in OE is internal or external corrosion, or if scope for license renewal; and (d) the piping is identified the Inspection of Internal Surfaces in as carbon steel in a treated water environment. Miscellaneous Piping and Ducting Components program (which does not SLRA Section 3.3.2.2.7, Loss of Material Due to include any RIC specific enhancements)

Recurring Internal Corrosion, states [b]ased on plant- needs to be enhanced based on this OE specific OE, recurring internal corrosion is an applicable (involving a treated water environment) or effect for steel components in raw water systems that use OE referenced in SLRA Section 3.3.2.2.7 water from the Mississippi River. The Open-Cycle Cooling (involving a raw water environment).

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.

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

• 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 Explain the apparent exclusion of part, safety-related electric equipment, nonsafety-related nonsafety-related electric equipment and electric equipment whose failure under postulated certain post-accident monitoring environmental conditions could prevent satisfactory equipment and how this equipment will accomplishment of safety functions, and certain post- be considered in the EQ AMP for the 1 B.2.2.3 B-33 accident monitoring equipment. subsequent period of extended operation.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions XCELMO00017-REPT-060 notes that the EQ program health color is Red but trending as Improving. Program What is the current health color of the EQ health is expected to turn Green at the end of 2022, program? If available, please upload the 2 B.2.2.3 B-33 reflected in the 1Q2023 program health report. latest EQ health report to the ePortal.

In EWI-08.11.01, Equipment Qualification Users For environmentally qualified (EQ)

Manual, Revision 26, the applicant referenced Electric components that the applicant Power Research Institute (EPRI) Report NP-1558, A used/relied upon EPRI Report NP-1558 Review of Aging Theory and Technology, dated as the justification/basis for activation September 1980. EPRI recently updated this report (July energies for extending the qualified life of 2020) due to issues/concerns with lack of or expired EQ equipment, has the licensee reviewed technical references for certain activation energies. The this revised document to verify that their NRC staffs understanding is that this revision resulted in justification/basis for activation energies up to 30% of activation energies being removed from the remains valid for EQ components for the 3 B.2.2.3 B-33 database. requested period of operation?

Please upload a viewable version of this CD 5.11.pdf displays a validation error and is unable to be document on the ePortal for the staff to 4 B.2.2.3 B-33 viewed. review.

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

• 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 Please explain the basis for determining SLR-ISG-2021-02-MECHANICAL. Element 3 of NUREG- that not measuring hydrogen peroxide 2191 Section XI.M2 (as modified) states that Corrosive levels is an exception to the NUREG-Parameters and Water Quality are measured and 2191 guidance. If hydrogen peroxide maintained in accordance with the EPRI BWR Water monitoring is a program requirement that Chemistry Guidelines, EPRI Report 30020002623 you are taking exception to, please clarify BWRVIP-190, Revision 1 (BWRVIP-190). The NRC staff how using online noble chemistry and has not found where the EPRI Guidelines require or hydrogen water chemistry and recommend measuring hydrogen peroxide levels as part maintaining low chloride and sulfate of measuring ECP. Therefore, the determination of this as levels in the reactor coolant, are 1 B.2.3.2 B-44 an exception to the guidance is unclear to the staff. acceptable alternatives.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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 Please clarify the Monticello Nuclear analyzing additional samples and this time frame is Generating Plant (MNGP) timeframe consistent with BWRVIP-190. However, this timeframe is requirements for sampling and analyzing not stated for a parameter exceeding Action Level 2 and 3 additional samples when a parameter 2 N/A N/A in II.05. exceeds Action Level 2 and 3.

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. ((

Please clarify the following:

))

• How the MNGP requirements follow the EPRI Guidance with respect to good practice values for the concentration of This pattern is also seen with insoluble iron. (( 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 3 N/A N/A )) operation conditions.

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

((

Please explain the basis for the deviation from the EPRI guidance for the sampling frequencies of feedwater H2

)) However, concentration and control rod drive water the staff did not find a justification for the feedwater H2, or dissolved O2 concentration at power 4 N/A N/A control rod drive water dissolved O2 sampling frequency. operation conditions.

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

• Please clarify the primary parameter 6.0 of the plant strategic chemistry and section 4.4.2 of related to the mitigation of IGSCC.

the Plant Chemistry program plan it states, ((

• Please discuss how measurements of

)) From catalyst loading are incorporated into the these documents it appears that mitigation is defined implementing documents, such as solely based on (( Chemistry Limits & Sampling 5 N/A N/A )). Frequency.

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 Please clarify the frequency of measuring 6 N/A N/A frequency listed in BWRVIP-190. HWC availability.

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 (( Please explain the two different forms of

)), for consistency with the referenced what appear to be the same Table 2-11 of BWRVIP-190. However, Table 3.15 of II.05 requirements for Action Level 1 (i.e., X-Y has a similar Action Level 1 limit for dissolved oxygen in ppb in one case; <X ppb and >Y ppb in 7 N/A N/A control rod drive water written as (( )). the other case).

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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 SLRA Table 3.3.2-17, for the Standby Liquid Control solution and expected corrosion rate.

(SLC) system, states that the Water Chemistry and One- d. A description of the sodium Time Inspection programs will be used to manage the loss pentaborate solution to which the of material and long-term loss of material for carbon steel components are exposed, and how it piping in a sodium pentaborate solution environment. differs (if it differs) from the sodium Plant-specific Note 1 states that aging effects are pentaborate solution in the SLC storage managed by monitoring and controlling SLC poison tank.

storage tank treated water chemistry. Because SLC e. Given that the sodium pentaborate systems are constructed primarily from stainless steels solution in the SLC storage tank is limited SLRA (e.g., NUREG/CR-6001, ML040340671), the staff by TS 3.1.7, describe how water Table requests additional information about the carbon steel chemistry can be adjusted to manage the 8 3.3.2-17 3.3-302 components and conditions of exposure. aging effects.

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

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 Please clarify the basis for considering and the aging management program for stainless steel SCC and LOM not applicable to stainless components exposed internally to a condensate steel in the off-gas condensate, and the 3.4-91, - environment. Three of the entries list VIII.E.SP-118a and basis for considering a One-Time Table 96, three list VIII.E.SP-127a as the NUREG-2191 AMR items. Inspection unnecessary to confirm the 1 3.4.2-5 -100 The corresponding SLRA Table 1 AMR items are 3.4.1- 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.

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 Please clarify the applicability of AMR and -108 may be non-applicable, or applicable Items 3.2.1-107 and 3.2.1-108 and Table 3.2-39 components may have been omitted from the 3.X.2-X describe any components to which they 2 3.2-1 3.2-40 aging management summary tables. 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

• 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 add1 supplement.

• 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 add2 this report).

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

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 Any reason the enhancement in SLRA MNGP to use BWRVIP-321-A to maintain compliance with Section B.2.3.19 doesnt correspond with 10 CFR Part 50, Appendix H. Commitment No. 22, as well as text in 1 B.2.3.19 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

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. Reconcile the discrepancy between the SLRA and the explanation provided 1 SLRA indicates that Reference 4.7.12 was needed to 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 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 SLRA Tables 4.2.3-1 and 4.2.3-2 identify multiple welds 1&2, VLCB 1&2 and VLCB 1&2) with a lot identifier of E8018N - This includes the circumferential welds and axial welds Since the circ weld (i.e., VCBA 2&3) has

• For the axial welds - sigma initial is 12.7 degrees F a sigma initial of 0 degree F - Confirm

• For the circ weld - sigma initial is 0 degrees F that unirradiated RTndt is a measured value SLRA Reference 4.7.13 - Monticello - Pressure and

• Is yes - identify the CLB document that Temperature Limits Report (PTLR) up to 54 Effective Full has been docketed with this information.

Power Years (EFPY), Revision 1, August 2014 If not previously docketed - provide (ML14246A206) documentation supporting the material

• Identifies a component as Limiting Weld - Beltline properties for the circ weld.

• Sigma initial is 12.7 degrees F If not, provide the basis that sigma initial 2 4.2.3

• Table 8 from PTLR is 0 degree F.

With respect to USE, SLRA Section 4.2.2 states that for Upper Shelf Energy the other beltline materials lacking initial USE data, EMA was performed to evaluate the impact of revised fluence Was the data from the Monticello 120° projections and available surveillance data on EOL USE ISP(E) Surveillance reductions. Capsule (see BWRVIP-347) assessed/considered?

• SIA Calc 2100300.301P and the the tables in SLRA Section 4.2.2 If not, what is the impact of the o Identify a capsule fluence of 9E17. surveillance data from the Monticello

§ Based on fluence and references in SIA letter - This 120° ISP(E) Surveillance appears to be the 300-degree capsule Capsule on the SLRA USE analysis and the EMAs?

Discuss and provide assessment/evaluation on the impact of 4.2.2 and the 120-degree capsule on the USE 3 4.2.3 TLAA, including the EMAs, in SLRA -

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions this includes the credibility evaluation, With respect to ART, SLRA Section 4.2.3 does not when considering the 120-degree provide a discussion on surveillance data being capsule.

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 Adjusted Reference Temperature material.

• RPV material and surveillance material is heat match Any reason the SLRA doesnt (based ISP material test matrix) reflect/assessment of this surveillance data of the 30-degree and 300-degree SIA Calc 2100300.302P - references BWRVIP-135, Rev capsule? Even a comparison of Reg 4, for assessment of surveillance data - It appears only 1.99 position 1.1 versus 2.1?

the Monticello 30-degree and 300-degree capsules were included. Was the data from the Monticello 120° ISP(E) Surveillance Based on sigma delta of 8.5 degrees F - it appears the Capsule (see BWRVIP-347) from the SLRA reduced margin from credible surveillance assessed/considered?

data was used.

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 XCELMO00017-REPT-091_TLAA_Report_Rev.0

• Is the weld heat 5P6756??

• Bottom of Page 28 of 108 indicates that weld heat 4 4.2.2 5P6757 was in the last capsule pulled from Monticello in

• 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 is in the Monticello ISP(E) capsules - It the SLRA appears that this weld heat is at a This weld heat does not appear to be in the ISP material minimum in the Columbia and River Bend matrix (BWRVIP-86, Rev 1-A and BWRVIP-321-A) 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 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 Provide additional the RG 1.99 Position 1 method. For conservatism, the detail/documents/discussion related to percent drop in USE for the plates are increased by 14.77 the conservatism described in the SLRA percent which is the difference in percent decrease between the measured percent USE decrease, and the Was all available surveillance data RG 1.99 predicted percent USE decrease for the incorporated when determining this surveillance plate heat C2220. conservatism?

Based on the description - it appears the drop in USE How does this conservatism compare to was ratioed based on surveillance data - which sounds the drop in USE based on position 2.2 of 5 4.2.2 like position 2.2. of RG 1.99, Rev 2. RG 1.99, Rev 2?

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

Provide additional detail/documents/discussion related to the conservatism described in the SLRA Why are the fluences different for 1/4T Was all available surveillance data incorporated when between the SLRA Sections 4.2.1 and determining this conservatism? the values used in SLRA Sections 4.2.2 and 4.2.3 for USE and ART, 4.2.2 and How does this conservatism compare to the drop in USE respectively?

6 4.2.3 based on position 2.2 of RG 1.99, Rev 2?

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>

What is meant/intended by Technical Specification change request given the Monticello has been approved for a SLRA Section 4.2.4 states the following: PTLR??

• The P-T limit curves will be updated and a Technical Specification change request will be submitted to the NRC Is this just referring to TS 5.6.5.c (i.e.,

7 4.2.4 prior to exceeding the current 54 EFPY limit. PTLR shall be provided to the NRC..)?

Please discuss the discrepancy in the fluence values used in 0T ART values for 8 shell Course 1 Please discuss the discrepancy in the fluence values used in 1/4T ART values 9 and USE values for all the RPV materials B.2.3.1 ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Question LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

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 Explain program consistency to NUREG-Section Cleanup AMP, but instead using the ISI AMP to 2191 and explain whether any B.2.3.1 implement the NUREG-2191 BWR Water Cleanup enhancements to the ISI program is 1 B-38 program. needed.

Section As result of a self-assessment conducted in 2019, several B.2.3.1 issues were identified for program enhancement. Provide a brief discussion explaining 2 B-38 Issue # AFI-1 states, in part, Nine out of 25 AMP owners 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- will be resolved prior to the start of AMP-001. Refer to CAP-501000032327. SPEO.

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

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-Section CAP 501000032327 states that Effectiveness reviews 02, Failure to Properly Pressure Test B.2.3.1 have not been completed for AMPs contrary to guidance Reactor Vessel Flange Leak-off Lines.

Section in 4 AWI-08.11.04. It also states, Proactive OE searches Discuss whether MNGP has reviewed IN B.2.3.22 B-38 have not been performed contrary to 4 AWI-08.11.04, and 2015-04, Fatigue in Branch Connection 4 B-171 NEI-1412. Welds for site applicability.

Discuss whether there is program-tracking of license renewal commitments at MNGP. Discuss if there are programs Section CAP 501000030695 indicates that inspection plan was and procedures in place to ensure B.2.3.1 not submitted within the timeframe required by the license license renewal commitments are 5 B-38 condition, related to its License Renewal Commitment #3. completed or revised as necessary.

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

• RTmax in BWRVIP-329-A is defined as RT(unirradiated) SIA Calc 2100300.401P and SLRA

+ Shift. appear to interchangeably use the terms

• SIA Calc 2100300.401P - in the assessment for the RTmax and ART even though are plate, circ and axial welds - the EOI RTmax appears to defined different.

include the margin value from RG 1.99, Rev 2, as well.

4.2.5 and 1 4.2.6

• SLRA Section 4.2.5 and 4.2.6 states, in part: 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) calc/SLRA doesnt acknowledge or point for the RPV surface (0T) and 72 EFPY was calculated this out given that RTmax and ART and using plant-specific material chemistry (copper content, defined differently?

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 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. Confirm reference to 10 CFR 4.2.5 and 40.40(a)(z)(1) is an error in these 2 A.3.2.5 The same reference was made in SIA Calc 2100300.401P documents.

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

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 Provide discussion of discrepancy in 1 4.2.7 4.2-26 acceptability of this K on a plant-specific basis for MNGP 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 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.

Clarify the basis for the statement in the SIA calc 2100300.303 similarly indicates that a maximum SLRA related to A maximum KI of 105 KI of 105 ksiin was use based on the Ranganath ksiin was utilized per Section XI IWB-2 4.2.7 4.2-25 analysis. 3612.

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 Provide discussion on why 105 ksiin 103 ksiin. The present analysis utilizes the 105 ksiin was chosen as the maximum Kiapplied in value, which is more conservative than the 103 ksiin this SLRA when 103 ksiin was used in 3 4.2.7 4.2-25 value. 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

SLRA Section 4.2.1 has table 4.2.1.2-1 (fluence values for the core shroud)

Is the discrepancy between fluences SLRA Section 4.2.8 indicates that the max fluence for the given in 4.2.1.2-1 and the quoted section core shroud is the following: in 4.2.8 because the table is only talking about the welds?

• The fluence for the most irradiated point on the core shroud was calculated to be 5.68 x 1021 n/cm2 (E >1 Where does the value of - core shroud MeV) for 80 years. was calculated to be 5.68 x 1021 n/cm2 (E >1 MeV) for 80 years come from?

SIA Calc 2100300.402 - Page 3 of 4 - indicates that the 4.2-10, most irradiated point of the core shroud is 3.68 x 1021 Clarify the discrepancy between the 1 4.2.8 4.2-28 n/cm2 (E >1 MeV) SLRA and the supporting SIA calc?

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

Step 3 of BWRVIP25, Rev 1-A SLRA Section 4.2.9 states the following:

• An assessment was performed to confirm that the

• Previously requested - references from MNGP Core Plate Bolts can have inspections waived and BWRVIP-25 Rev 1-A have their age-related degradation managed for the o GE Hitachi Safety Communication SC SPEO for 72 EFPY. This evaluation concluded that the 11-05, "Failure to Include Seismic Input in criteria of Appendix I of BWRVIP-25, Revision 1-A to Channel Control Blade Interference justify the elimination of core plate bolt inspections at Customer Guidance." September 2011.

MNGP are satisfied. Therefore, elimination of core plate o "Failure to Include Seismic Input in bolt inspections at MNGP for the SPEO is justified. Channel-Control Blade Interference Customer Guidance," GE Hitachi Nuclear SIA Calc - 2100300.403P - Stress Relaxation of Core Energy SC 11-05, Rev.2, December 16, Plate Rim Holddown Bolts Analysis - has the detailed 2013.

evaluation.

1 4.2.9

• 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 conditions and abide by the steps from Appendix I to BWRVIP-25, Rev 1-A 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

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. If it was a TLAA for 60 years - It would be expected to still be a TLAA for 80 SLRA Section 4.2.10 - dispositions TLAA Disposition: 10 years, unless there was a revision to the CFR 54.21(c)(1)(iii) and states: CLB to remove component from the

• Aging effects of IASCC and embrittlement on the top TLAA.

guide, core shroud, and jet assembly components will be Was there a revision to the CLB to managed by the BWR Vessel Internals (B.2.3.7) AMP remove the incore instrumentation dry through the SPEO in accordance with 10 CFR tubes and guide tubes in the TLAA from 54.21(c)(1)(iii). initial license renewal?

Note the discrepancy in the components that were identified in the TLAA in the 60-year LRA and the What is the basis that the incore components in the disposition for 80-year SLRA. instrumentation dry tubes and guide 1 4.2.10 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.

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). Discuss and reconcile that the neutron fluence for the Core Support Plate The SLRA appears to rely on BWRVIP-315 for identifying exceeds the screening criterion for those components susceptible to IASCC. BWRVIP-315 is IASCC but does not appear to be 2 4.2.10 not an approved topical report. 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

Can you explain how the condition of 1 OpE OpE 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.

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 2 OpE OpE license of the plant?

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

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 4 OpE OpE changed?

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

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

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions What contributed to the RED status for the cable monitoring program during the 7 OpE OpE 2014-2015 time-frame?

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 8 this was an unintentional error.

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

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 Clarify if the above-mentioned relied upon in safety analyses or plant calculations to underground cables in the 345-kV,13.8-perform a function related to any regulated events kV, and 4.16-kV power systems are used identified by 10 CFR 54.4(a)(3). for power restoration during an SBO event. If they are, clarify why they are not SLRA Figure 2.5-1, MNGP Simplified One-Line Diagram included in the components and 2.5 Page 2.5- (For SBO Offsite Power Recovery), shows underground commodities in-scope of SLR and they 1 2.1.2 2 medium voltage cables between the 345-kV offsite power 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.

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 Clarify if the elements, resistance electric heaters as commodity groups that will meet the temperature detectors, sensors, passive component screening criterion 10 CFR thermocouples, transducers, and electric 54.21(a)(1)(i) if they have a pressure boundary function. heaters commodity groups have a pressure boundary function at MNGP, The staff did a keyword search for the above-mentioned and if the pressure boundary function for components and could not find any information about the these commodities is addressed in the 2 2.5 2.5-6 screening of these components. mechanical review.

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 Clarify why cable bus is included in the 2.5-6 is not utilized at MNGP. commodity group and passive commodity 3 2.5 2.5-1 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.

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 Discuss the aging management reviews holders insulation: Reduced electrical insulation results for the insulation material of fuse resistance due to thermal/thermoxidative degradation of holders (not part of active equipment) to organics, radiolysis, and photolysis (UV sensitive demonstrate that these fuse holders are 3.6-13 materials only) of organics; radiation-induced oxidation; in an environment that does not subject 3.6-32 moisture intrusion. them to environmental aging 3.6.2 3.6-36 mechanisms identified in NUREG-2192 4 3.0 3-11 The staff noted that the evaluation in SLRA section 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).

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 Add 1 SLRA.

Section 3.6.2.3.1, does the statement SLRA Table 3.6.2-1, Plant-Specific Note 1 states: MNGP fuse holders (not in active

1. In alignment with GALL-SLR, no AMP is required when components) insulation material and fuse holders are located in an environment that does not environment combination has no aging subject them to environmental aging mechanisms. Fuse effects requiring management mean that holder insulation material in an ALE is managed via the the fuse holders insulation material are XI.E1 AMP. MNGP fuse holders (not in active not subject to an ALE? If so, please components) insulation material and environment explain how the determination was made Table combination has no aging effects requiring management. that the insulation material are not Add 2 3.6.2-1 See SLRA Section 3.6.2.3.1 for additional information. subject to an ALE in the SLRA.

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 Add 3 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

SLRA Section 2.4.6 states, Curbs, dikes, concrete Please discuss whether there are any 1 2.4.6 2.4-15 components other than barriers are evaluated as part of 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 intended function, and if so, where are and 3.5.2-6 do not include curbs or dikes with a fire they addressed in the SLRA. In addition, barrier intended function. In addition, the staff did not discuss where curbs and dikes with other identify AMR items explicitly for curbs or dikes, regardless intended functions are addressed in the of intended function, in the SLRA. SLRA.

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 Please discuss the intent of the 2 N/A N/A additional fire damper parts. underlined statements.

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 Please clarify the functional test and with Section 19.4.1.1 of NFPA 80, 2007 Edition. visual inspection frequency of fire 3 N/A N/A Specifically, the test and inspection frequency for 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.

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 Please address the following:

fire protection component material. It is unclear whether SLRA Section B.2.3.15 includes FP guard pipe. SLRA 1. Where are FP guard pipe and rigid Tables 2.4-6 and 3.5.2-6 do not appear to include board (gypsum walls, etc.) addressed in component types of FP guard pipe or rigid board (gypsum the SLRA?

walls, etc.). It is unclear where these component types 2. What is the difference between are addressed in the SLRA. cementitious fireproofing and cementitious non-metallic fireproofing in SLRA Tables 2.4-6 and 3.5.2-6 include component types SLRA Tables 2.4-6 and 3.5.2-6?

fireproofing and non-metallic fireproofing. SLRA Table 3. What is the difference between the 3.5.2-6 indicates that the fireproofing and non-metallic two cementitious fireproofing fireproofing material is cementitious. AMR item 3.3.1-268 components in SLRA Table 3.5.2-6?

is cited for both. However, it is unclear what is the 4. What is the difference between the difference between cementitious fireproofing and two silicate thermal fiber components in cementitious non-metallic fireproofing. SLRA Table 3.5.2-6?

5. Should aluminum be added to SLRA SLRA Table 3.5.2-6 includes two cementitious Section B.2.3.15 as a fire protection 4 2.4.6 2.4-15 fireproofing component types: both citing AMR item 3.3.1- 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.

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 Please discuss whether the statement in properties, cracking/delamination, and separation. These the discussion of AMR item 3.3.1-268 in aging effects are consistent with Section 6, Fire SLRA Table 3.3-1 regarding HELB Barriers, of EPRI 3002013084, Long-Term Operations: Barriers is correct.

Subsequent License Renewal Aging Affects for Structures and Structural Components (Structural Tools), In addition, given that EPRI 3002013084 November 2018. cites loss of material, change in material properties, cracking/delamination, and The discussion of AMR item 3.3.1-268 in SLRA Table separation as aging effects for 3.3-1 states, Consistent with NUREG-2191. The Fire cementitious coatings and silicates used Protection (B.2.3.15) AMP is used to manage loss of as fireproofing/fire barriers, please material, change in material properties, cracking, discuss why not all of these aging affects 3.3-90, delamination, and separation for cementitious coating are cited for each cementitious 3.5-7, 3.5- fireproofing/fire barriers/HELB barriers exposed to air fireproofing and silicate thermal fiber 5 3.3, 3.5 96 indoor uncontrolled. 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.

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.

Please discuss whether stainless steel SLRA Section 3.5.2.1.6 does not include stainless steel should be added to SLRA Section as a material for the Fire Protection Barriers Commodity 3.5.2.1.6 and whether stainless steel fire Group. In addition, SLRA Table 3.5.2-6 only includes barrier penetration seals should be 6 3.5.2.2.2.4 3.5-34 elastomer fire barrier penetration seals. added to SLRA Table 3.5.2-6.

Work Order 00546903 states that Door-3 (TRB BLDG Please discuss whether Door 27 in 931 to Heating Boiler Room) provides fire protection, Procedure 0275-03 is a fire door. Please external flooding, security, and appendix R for safe discuss why Door 1 is inspected under shutdown considerations. Procedure 0275-03 if it is not a fire rated door.

Fire rated doors with only a fire barrier intended function are managed by the Fire Protection program per SLRA Please discuss whether fire rated doors Table 3.5.2-6. In the SLRA, it appears that doors with in SLRA Table 3.5.2-6 should include intended functions other than a fire barrier intended flood barrier and HELB barrier intended function are evaluated as part of the individual structures functions given procedures under the where they are located, and it appears the Structures Fire Protection program appear to Monitoring program is cited to manage aging effects. For address doors with these functions.

example, SLRA Tables 2.4-17 and 3.5.2-17 include doors with a flood barrier and a HELB Barrier intended functions However, if the intent was for the Fire managed by the Structures Monitoring program. Protection program to manage aging 7 N/A N/A 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 function is maintained and other functionality/operability of Appendix R Fire Doors and program(s) (e.g., Structures Monitoring HELB Doors. In addition, Procedure 0275-03 covers program) to manage aging effects to HELB doors and flooding requirements for Doors 1 and ensure intended functions other than the

27. This procedure does state that Door 1 is not a fire fire barrier intended function (e.g., flood door, however, a similar statement is not made about barrier and HELB barrier) are Door 27. It also states to contact the Flooding Engineer if maintained, then discuss how program steps to meeting flooding requirements are not met. procedures will be updated to adequately reflect this. (Breakout Question 8 is Neither Procedure 1216-01 nor Procedure 0275-03 refer similar question) 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.

Figure 5.2 of EWI-11.01.07 that identifies system Please discuss whether the staffs intended functions includes fire barrier intended functions assumption that the component types in for certain systems, for example, Intake Structure. SLRA SLRA Table 3.5.2-9 cites intended Table 3.5.2-9 does not cite a fire barrier intended function functions other than fire barrier, and for any components. The staff assumes component types SLRA Table 3.5.2-6 cites the fire barrier may be repeated in SLRA Table 3.5.2-9, for example, intended function for the same and SLRA Table 3.5.2-6. The component types in SLRA component types and that this Table 3.5.2-9 cites intended functions other than fire assumption applies to other systems barrier, and SLRA Table 3.5.2-6 cites the fire barrier identified in Figure 5.2 of EWI-11.01.07 intended function for the same component types. Is that that include a fire barrier intended 8 N/A N/A correct? 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 Please discuss why cementitious requiring management for specific material/environment fireproofing and rigid board (thermal groups includes cementitious fireproofing and rigid board insulating board) is addressed in EWI-(thermal insulating board). 11.01.07.

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.

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 Please confirm the statement in the Fire required but it just wasnt documented in the AMR. Protection program means there are no installed fireproofing that requires SLRA Table 3.5.2-6 only cites an outdoor air environment shielding. In addition, please clarify for fire rated doors and structural fire barriers (walls, whether there is fireproofing installed 9 N/A N/A ceilings, and floors). outdoors.

The Fire Protection program basis document states that Since the exception is no longer needed, the exception taken during initial license renewal related please confirm that the inspection to the testing frequency of the Cable Spreading Room frequencies stated in Revision 28 of halon system is no longer needed since GALL-SLR AMP Procedure 0328 and Ops Manual XI.M26 no longer prescribes a frequency. B.08.05-05 will continue for SPEO.

Revision 28 of Procedure 0328 states per Ops Manual Given that the frequency for visual B.08.05-05 that the system functional test is every 18 inspection only refers to headers and months, visual inspection of headers and nozzles is every nozzles, and Step 1 states all accessible 18 months, and an air flow test through headers and external surfaces (consistent with GALL-10 N/A N/A nozzles is every 3 years (these frequencies are also 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 please confirm that all accessible basis document). The Ops Manual also includes monthly external surfaces, including headers and valve position checks and halon tank weight and pressure nozzles, are visually examined every 18 checks every six months. 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.

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

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 Please discuss why the Ops Manual of the total number of penetration seals includes 10 does not state that the 20 percent of the percent of each type of penetration seal (consistent with total number of penetration seals GALL-SLR AMP XI.M26). However, Ops Manual B.08.05- includes 10 percent of each type of 05 only address the 20 percent of the total number of penetration seal.

penetration seals; it is silent on the 10 percent of each type of penetration seal. Please provide the scope and frequency for Procedures 0275-04 and 0275-05 There appears to be three procedures related to seals: and how all three procedures related to 0275-01, 0275-04, and 0275-05. Neither of these penetration seals work together with documents reference each other. In addition, Procedures regards to meeting 20 percent of the total 0275-04 and 0275-05 does not state the scope (e.g., all, number of penetration seals, including 10 12 N/A N/A 20 percent, etc.) or frequency (e.g., 24 months). percent of each type of penetration seal.

Figure 5.3 of EWI-11.01.09 include material reinforced concrete, grout.

Revision 1 of FIREPROTECT states, Grout is Given that it appears grout is a fire considered to be a part of the material constituting the protection material, please discuss barrier in which it is installed. Therefore, piping, conduit, revising the SLRA to address grout as 13 N/A N/A etc. that penetrates a barrier and sealed with grout is 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.

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 Please discuss which program will reasonable since these activities were stated to not fall execute Revision 31 of 1158-B and 14 N/A N/A under that program for SPEO. 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.

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.

Please discuss whether the program will The staff did not identify where Revision 28 of Procedure be enhanced to trend the inspection Appendix 0328 discusses trending. However, Table 1 in Section 7.0 results for fire barrier penetration seals, A and of the basis document states that Procedure 0328 will be fire barriers, fire damper assemblies, fire Appendix A-64 and updated to trend inspection results and where practical, doors, and the halon fire suppression 15 B B-109 project degradation until the next scheduled inspection. 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.

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. Please clarify the enhancement related to expanding the penetration seal The staff noted that the enhancement related to inspection scope if degradation is expanding the penetration seal inspection scope if detected appears. For instance, will the Appendix degradation is detected appears to be inconsistent in the scope be expanded if any sign of A and Fire Protection program basis document. For example, degradation is detected or is there a Appendix A-65 and Section 6.0 (description of Element 7 enhancement) and process for determining when to expand 16 B B-110 Table 1 in Section 7.0 do not include the phrase if 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.

Revision 23 of AWI-08.01.00 and Revision 1 of 0275-02 Please discuss the basis for cracks state to consider cracks greater than 0.25 inches wide in greater than 0.25 inches wide and walls, floors, and ceilings. However, these documents did provide, if available, supporting not appear to cite a reference for this crack width limit. references.

Procedure 0275-03 states that a 3/16-inch limit is used Please discuss the 3/16-inch limit being when checking for depressions or bulges in fire doors a good engineering practice and provide, 17 N/A N/A based on good engineering practice. if available, supporting references.

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

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 1. Describe how the applicant estimated stresses in the system were examined to determine an the annular number of airburst cycles allowable number of cycles for the tank of 35,000 airbursts (160 cycles/year).

(i.e., backwashing cycles) under normal and accident conditions. 2. Clarify whether the fatigue analysis needs to consider the transient cycles in SLRA Section 4.6.3 also indicates that the annular the upset and test conditions for the number of airburst cycles is conservatively estimated to condensate backwash receiving tank. If be 160 cycles/year. However, the SLRA does not clearly not, explain why the transients in the discuss how the applicant estimated the annular number upset and test conditions do not need to of airburst cycles. be considered in the fatigue analysis (e.g., the transients evaluated in the In addition, it is unclear to the staff whether this fatigue analysis are bounding for the upset and analysis needs to consider the transient cycles in the test transients, or non-applicability of upset and test conditions for the condensate backwash upset and test transients to the fatigue 1 4.6.3 4.6-5 receiving tank. 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 LRA Section LRA Page Background / Issue Discussion Question / Request Number (As applicable/needed)

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, a) Explain the apparent discrepancy 3 and MC component supports. between the referenced statement in the PBD and the SLRA B.2.3.30 and B.2.3.30; The above PBD statement appears inconsistent with GALL-SLR XI.S3 AMPs; OR provide XCELMO00017- the SLRA B.2.3.30 AMP as well as the GALL-SLR the basis of how the SLRA AMP PBD is REPT-077, AMP XI.S3 which manage the effects of all applicable in its objective consistent with the Revision 1 aging effects (which include several aging effects GALL-SLR Report AMP XI.S3, as (PBD), Section B-224; including loss of material) for Class 1, 2, 3, and MC claimed in the SLRA, and meeting the 1 3.1 PBD p4 component supports. requirements of 10 CFR 54.21(a)(3).

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 Consistency of scope of program element: requirements of 10 CFR 54.21(a)(3).

SLRA AMP includes an enhancement (commitment b) Explain why Class MC component 33(a)) to the scope of program element related to supports are not included in the evaluation of inaccessible areas, which appears to enhancement/commitment.

only address inaccessible areas of Class 1, 2 and 3 supports. c) Provide a revised enhancement/commitment that However, the scope of the SLRA also includes Class addresses the concerns raised; OR, MC components, which appears to be not included in provide an exception to the scope of the enhancement. Further, the enhancement program element with the basis of why language also does not appear to state inaccessible the AMP with exception remains B.2.3.30; Table B-225; A- areas of what components (e.g., component adequate to manage applicable aging 2 A-3 item 33 87 supports). effects.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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 a) Clarify why polymeric vibration vibration isolation elements. Also, a related isolation elements are not included in enhancement to parameters monitored or inspected the referenced enhancement, and B.2.3.30, Table B-226, A- element includes polymeric vibration isolation provide a revised enhancement if 3 A-3 item 33 88 elements necessary.

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 a) Clarify if the highlighted word support is repaired to as-new condition by .. population in the enhancement is (emphasis added). intended to be sample and accordingly provide a revised To be consistent with the GALL-SLR AMP XI.S3, the enhancement/commitment; OR justify B.2.3.30, Table B-226, A- highlighted word population in the enhancement why the enhancement need not be 4 A-3 item 33 89 should revised.

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. a) Correct the Table A-3 (item 33) implementation schedule to make There appears to be a typographical error, i.e., Table reference to the appropriate Table A-3 item A-87, A- A-3, item 33 commitment for one-time inspection is commitment regarding one-time 5 33 88 33-f) and not 33-g). inspection USAR Supplement issue: a) Provide revised language of To be consistent with GALL-SLR AMP XI.S3, SLRA referenced statements in the USAR B.2.3.30, B-224, B- Section B.2.3.30 includes appropriate enhancements supplement description in SLRA A.2.2.30, 225; (corresponds to SLR Commitments 33(f) and 33(h)) A.2.2.30 to state what the IWF AMP will Table A-3, item A-28, A- to the detection of aging effects program element do regarding one-time additional 6 33 29 that specify one-time inspection of additional 5% 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 of HSB greater than 1-inch examination for high-strength bolting (HSB) greater diameter, consistent with the than one inch in diameter, respectively. enhancements and commitments 33(f) and 33(h).

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.

a) Discuss the basis (how) for the determination for SLRA AMR item 3.5.1-068 that HS bolting is not used in AMR Item 3.5.1-068 Issue: MNGP structures or component SLRA states item 3.5.1-068, related to SCC, is not supports.

applicable because there is no high-strength (HS) steel structural bolting is not used in MNGP structures b) Clarify whether or not HS structural or component supports. bolting exists or will be used in the future at MNGP, prior to or during This appears to be contrary to the enhancement SPEO, and if not use how will it be (commitment 33(h)) related to volumetric examination ensured that it will not be used in the of HS bolting for SCC, which appears to provide an future prior to or during the SPEO.

adequate aging management of SCC consistent with GALL-SLR Report recommendations. It also appears c) Clarify and correct any potential to indicate that HS bolting exists and/or may be used discrepancy in item 3.5.1-068 with the 7 Table 3.5-1 3.5-66 in the future. 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.

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 Operating Experience for RV supports: condition of the RV supports (steel skirt From the SLRA discussion in the third bullet under and weld, anchorage, pedestal, RPV Industry Operating Experience regarding reactor stabilizer brackets, stabilizer, rods and vessel (RV) supports being not included in the IWF trusses etc). Provide recent program, it is not clear if the subject industry OE representative photos of recent material applied to MNGP and what the general material condition of these RV supports, if 8 B.2.3.30 B-227 condition of the RV supports is at MNGP. available or taken during RF031.

B-224 SLRA 3.5.2.2.2.6 states, in part: Therefore, the a) While a plant-specific program may thru B- integrity of the reactor vessel supports is assured, not be necessary, describe how the 226; and no additional aging management of reactor aging effects due to irradiation B.2.3.30; 3.5-30; vessel supports beyond the current ASME Section XI, embrittlement will be adequately 3.5.2.2.2.6, 3.5-76 Subsection IWF (B.2.3.30) AMP is necessary for managed for the RV steel support Table 3.5.2-1, thru 3.5- aging effects due to irradiation during the MNGP assembly components for the SPEO?

9 Table 3.5.2-7 83; SPEO. Provide corresponding Table 2 AMR

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 3.5-98 items, noting that it currently is not thru 3.5- Nevertheless, while a plant-specific AMP may not be included in the GALL-SLR Report.

103 necessary [note this aspect is being reviewed under TRP 76 and not TRP 43], loss of fracture toughness b) Discuss how it is included within the due to irradiation embrittlement remains an applicable scope of the program credited (e.g.,

aging effect for the RV steel supports for SLR. IWF AMP) and any related changes Although Table 2 of the PBD (XCELMO00017-REPT- that may need to be made to the AMP.

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.

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

• What spatial/structural issues exist a(2) Spatial/Structural. 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 3.4-108 SLRA Table 3.3.2-10: Fuel Pool Cooling and Cleanup -

Service through Summary of Aging Management Evaluation:

and Seal 3.4-111 Water * [Page 619/1519] For Component Type Tanks (Skimmer Surge Tanks) the Material Carbon Steel Please confirm that the Material of (with Internal Coating) did not appear in LRA Table Carbon Steel (with Internal Coating) 3.3.2-10 Auxiliary Systems - Fuel Pool Cooling and listed in SLRA Table 3.3.2-10 for Tanks Cleanup - Summary of Aging Management Evaluation (Skimmer Surge Tanks) is correct.

[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 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 a) Staff requests an explanation for Fuel Pool F/D -47B Outlet SX-2796 on FPW10B- representation of tubing to and SX-6-HK located on SLR-36257(D,3) - Tubing to SX- 2796 on drawing SLR-36257 as not 2796 is not color-coded a(2) Spatial/Structural on subject to Aging Management SLR-36257 Review.

Coord. B-8, FPC system components - Pressure

/Temperature Maximum Limitations Specified for Fuel

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 b) Staff requests an explanation for on SLR-36257(E,6) - Tubing to SX-2795 is not color- representation of tubing to and SX-coded a(2) Spatial/Structural on SLR-36257 2795 on drawing SLR-36257 as not subject to Aging Management Review.

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

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 Is not the Environment of Sodium Coating) and Elastomer: Each Material has Pentaborate Solution more appropriate 2.3-63, respectively, the Environment of Treated (Water for these two interfaces of Material with 1 2.3.3.17 3.3-299 Internal) and Treated Water (External) Environment?

Scoping/Screening Boundary Drawing SLR-36253

• Either: a(1); a(2) FUNCTIONAL; or a(3)

Standby Liquid Control System, Coordinate C-3 red appears to be appropriate. Should Line DW8-1-HS and Valve DM-56 is displayed as a(2) not this pipe segment and valve be color-2 2.3.3.17 2.3-63 Spatial/Structural green. coded as red?

a) Is this line ASME-Class 1 pipe and is it Seismic Class 1?

• If so, are its Seismic anchors up to Scoping/Screening Boundary Drawing SLR-36253 isolation valves DM-52 and AS-22 Standby Liquid Control System, Coordinate D-2: subject to AMR?

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

• If so (continued) Is line SA-3/4-JB

[Reference DBD-B.03.05 Revision 5 Section 5.2 SLC (attached to DW7 HK) qualified as (Storage) Tank, T-200; Section E Special Design Seismic II/I? OR Is it compatible with Considerations] the guidance of SLRA Section 2.1.4.2.2 b) Pipe line DW7 HK (attached to the SLC Tank) is Non-Safety Related SSCs Directly 3 2.3.3.17 2.3-63 color coded as a(2) Spatial/Structural 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.

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]

Scoping/Screening Boundary Drawing SLR-36253 Standby Liquid Control System, Coordinate E-2: b) Is this line qualified as Seismic I out to the pipe class change HC/HF? In a) The vent line on the Standby Liquid Control Tank is particular, are the necessary seismic I color coded as neither a(1) nor a(2) -- not subject to Aging and seismic II/I supports for this line Management Review (black). subject to AMR consistent with the guidance of SLRA Section 2.1.4.2.2 Non-b) Pipe line D68-3-HC is color coded as a(2) Safety Related SSCs Directly Connected Spatial/Structural. It is not clear whether the line to Safety-Related SSCs that Provide represents a Spatial leakage concern or a Structural Structural Support for the Safety-Related 4 2.3.3.17 2.3-63 concern OR BOTH. SSCs?

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

Instrument Air to Tank Level Instrumentation is not subject a) Does LT 11-45 remain to AMR (black) on the AIR side but has an a(3) ATWS functional/operable in the absence of an function (red) on the SLC side of the flag of demarcation. Instrument Air pressure supply?

Both level -transmitters LT 11-45 and LT 7449 have b) Please provide the basis for this independent Class 1E and/or D.C. power supplies (e.g., scoping/screening of the instrument air 5 2.3.3.17 2.3-63 via the EDGs.) [Reference DBD-B.03.05 Revision 5 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

A.2.2.9: Additionally, the software tool, FAC Manager', Discuss the erosion module being with the erosion module, is used to evaluate components referenced in the associated UFSAR 1 for both FAC and erosion. Supplement.

EPRI 3002005530. Recommendations for an Effective Program Against Erosive Attack (discussed in the Surry SLRA) states that a minimum safety factor should never Discuss what safety factor will be used 2 be less than 2.0. for evaluations of erosive wear.

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

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 RFO23 ISI-21-E10 FAC-07-072 shows average thickness isolation valve that also are heavier wall 4 of 0.611, whereas the nominal thickness is listed as 0.343. that may need to be looked into?

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

The SLRA states, For the components within the scope It appears there is an inconsistency in the of the MNGP Open-Cycle Cooling Water Systems AMP description of the OCCW systems in the that have internal coatings (e.g., various piping SLRA and Engineering Work Instruction components within the ESW System, etc.), the internal 11.01.12 Revision 6. A review of 1 B.2.3.11 B-82 coatings will be managed by the MNGP Internal 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, that referenced an RHR heat exchanger Heat Exchangers, and Tanks AMP (B.2.3.28). with a coating on the divider plate (page B-90).

The Open-cycle Cooling Water System Program document (EWI-11.01.12 Rev. 6) states in Section 4.3 It appears EWI 11.01.12 should be that, The MNGP Open Cycle Cooling Water (OCCW) revised.

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.

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

Item VII.C1.A-473, SRP Item 3.3.1-160 in NUREG-2191, Please describe how the Fire Water Volume 1 addresses cracking due to SCC for copper alloy System program will manage cracking of

(>15% Zn or >8% Al) piping, piping components, and heat copper alloy >15% Zn heat exchanger -

exchanger components exposed to closed-cycle cooling (diesel fire pump) tube sheet and tubes, water, raw water, and waste water managed by several and piping and piping components programs including AMP XI.M38, Inspection of Internal exposed internally to raw water.

Surfaces in Miscellaneous Piping and Ducting Specifically discuss whether surface Components. examinations will be performed or whether analyses will be performed to 1 3.3 3.3-183 AMP XI.M38 notes that periodic surface examinations are 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 detected by leakage prior to a crack aluminum components and states, Visual inspections for challenging the intended function of the leakage or surface cracks are an acceptable alternative to component, such that visual inspections conducting surface examinations to detect cracking if it would suffice.

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.

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. Please discuss whether flow blockage is an applicable aging effect requiring The Discussion for AMR Item 3.3.1-253 in SLRA Table management for the PVC valve bodies 2 3.3 3.3-201 3.3-1 states, The Inspection of Internal Surfaces in 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.

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 A-65 refueling outage prior to the SPEO.

Appendix B-116 A, SLRA Section B.2.3.16 states, The enhancements are to Please clarify the implementation Appendix 1175, be implemented no later than 6 months prior to entering schedule for the Fire Water System 3 B 1325 the SPEO. This AMP is to be implemented and its pre- 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.

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 Please confirm the staffs assumption Appendix Protection Program, January 1, 2019, ADAMS Accession that the wet pipe sprinklers will be tested A, No. ML19018A180, there are wet pipe sprinklers in the in accordance with NFPA 25, Section Appendix A-19, B- Lube Oil Storage Tank Room, Hydrogen Seal Oil Unit, 5.3.1.1.2 as recommended in Footnote 7 4 B 113 and under the turbine floor (Lube Oil Piping Sprinkler 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.

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 Please discuss why the augmented tests segments that allow water to collect. and inspections beginning 5 years prior to SPEO are not included in the However, Table A-3 in SLRA Appendix A and Section enhancement to the Parameters B.2.3.16 in SLRA Appendix B do not include this in the Monitored or Inspected program element enhancement to the Parameters Monitored or Inspected related to the augmented tests and program element related to the augmented tests and inspections.

inspections. The staff notes that the Fire Water System program basis document also does not include this in In addition, if possible, please identify Sections 4.3 and 6.0. which portions of the fire protection system are subject to augmented testing Appendix The staff notes that the table in Section B.2.3.16 of SLRA or inspection because they are normally A, A-65, B- Appendix B that provides additional detail on the dry but periodically subjected to flow and Appendix 117, B- enhancements based on GALL-SLR AMP XI.M27, Table cannot be drained or allow water to 5 B 135 XI.M27-1, includes the augmented tests and inspections 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.

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 Please clarify the trending process for be utilized to drive improvement. The staff notes that this flow testing and wall thickness appears to indicate that trending may occur outside of the measurements, including whether corrective action program. trending of inspection results is Appendix performed outside of the corrective action A, Acceptance Criteria - Clarify within the new internal program for instances where the results Appendix A-67, B- inspection procedure and relevant existing preventive are not entered into the corrective action 6 B 118 maintenance activities which inspect wall thickness that 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.

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 Please confirm how far below the frost amount of water ( 3 inches height) within the hydrant line the 3 inches of water left in the 7 N/A N/A barrel will not freeze. hydrant barrel is.

Item VII.G.A-797b, SRP Item 3.3.1-263 in NUREG-2191, Please identify which piping and piping Volume 1 addresses hardening or loss of strength, loss of components are polymeric and identify material, cracking, and flow blockage of polymeric piping, the polymeric material.

piping components, ducting, ducting components, seals exposed to air, condensation, raw water, raw water In addition, please discuss whether flow 3.3-11, (potable), treated water, waste water, underground, blockage is an applicable aging effect for 8 3.3 3.3-188 concrete, and soil managed by the Inspection of Internal 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 exposed internally to raw Components program. water.

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.

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. Please discuss whether there are 9 N/A N/A However, it is unclear what the enhancements are. In 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.

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 Appendix mechanism shall be identified and used to determine the A, most susceptible system locations for additional Appendix A-68, B- inspections, including consideration to the other unit This is an observation given that 10 B 120 systems as driven by the corrective action program. Monticello is a single unit site.

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. Please discuss whether the enhancement to the Monitoring and Table A-3 in SLRA Appendix A and Section B.2.3.16 in Trending program element in Table A-3 SLRA Appendix B do not appear to include this discussion in SLRA Appendix A and Section on impact on downstream components in the B.2.3.16 in SLRA Appendix B should enhancement to the Monitoring and Trending program include this discussion on the impact on 11 N/A N/A element. downstream components.

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 Appendix surveillances.

A, Please discuss, for consistency, whether Appendix A-65, B- This statement does not appear in the Fire Water System this statement should be included in 12 B 116 program enhancements in Table A-3 in SLRA Appendix A. Table A-3 in SLRA Appendix A.

Item VII.G.AP-76, SRP Item 3.3.1-096 in NUREG-2191, Please discuss whether flow blockage is Volume 1 addresses loss of material and flow blockage an applicable aging effect for elastomer (raw water only) for elastomer piping, piping components, hoses (pump and drain hoses) exposed 13 3.3 3.3-186 and seals exposed to air, raw water, raw water (potable), 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.

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 Please discuss whether flow blockage is raw water. an applicable aging affect for carbon steel piping and piping components, gray SLRA Table 3.3.2-9 cites AMR Item 3.3.1-064 for loss of cast iron pump casing (fire system jockey material of carbon steel piping and piping components, pump), copper alloy with greater than gray cast iron pump casing (fire system jockey pump), 15% zinc valve bodies exposed 14 3.3 3.3-187 copper alloy with greater than 15% zinc valve bodies 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.

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.

Please discuss whether AMR Item 3.3.1-SLRA Table 3.3.2-9 includes galvanized steel piping and 078 should be cited for galvanized steel piping components exposed externally to outdoor air, piping and piping components exposed carbon steel valve bodies exposed externally to indoor externally to outdoor air, carbon steel uncontrolled air, and gray cast iron valve bodies exposed valve bodies exposed externally to indoor externally to indoor uncontrolled and outdoor air. uncontrolled air, and gray cast iron valve However, AMR Item 3.3.1-078 is not cited for managing bodies exposed externally to indoor 15 3.3 loss of material of these components. uncontrolled and outdoor air.

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. Please discuss whether the cooling water jacket and lube oil heater are included in SLRA Table 3.3.2-9 includes several entries for shell side the heat exchanger component types in components, tube sheet, tube side components, and SLRA Table 3.3.2-9. In addition, discuss tubes for the diesel driven fire pump heat exchanger. The whether lubricating oil is an applicable 3.3-11, cooling water jacket heater and lube oil heater are not environment for the diesel fire pump heat 16 3.3 3.3-182 explicitly stated as a component type. exchangers.

Appendix Section B.2.3.16 in SLRA Appendix B includes an Please clarify whether the 20 percent 17 B B-115 exception to the annual main drain tests at each water- 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 refueling outage. If so, will that result in main drain test at each riser, main drain tests on 20 all riser and standpipes being inspected percent of the standpipes and risers will be performed over a certain period (i.e., 10 years)?

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.

Bolting Integrity Question LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

Section 4.2, Preventive Actions, in the basis document Bolting (XCELMO00017-REPT-073) notes that tread lubricant Integrity Basis doc containing MoS2 is not used, as described in procedure Basis pg. 8 of MMP-008, revision 12. Please clarify whether any other Document, 32 lubricants containing sulfur are currently Sections Basis Document Sections 4.2 and SLRA Section B.2.3.10 used for closure bolting.

4.2 states that procedures will be enhanced to clarify that in addition to MoS2, other lubricants containing sulfur will be If other lubricants containing sulfur were prohibited from use on pressure-retaining closure bolting. used, discuss if procedure is needed to SLRA identify those other lubricants containing Section Page B- It Is unclear to the staff whether other lubricants sulfur and how to manage the potential 1 B.2.3.10 76 containing sulfur exist in the plant currently. aging effect due to their use.

Basis doc GALL-SLR XI.M18, Element 4, Detection of Aging Please clarify whether the existing Bolting pg. 8 of Effects, stats that Closure bolting inspections includes inspection procedure includes applicable 2 Integrity 32 consideration of the guidance applicable for pressure guidance from EPRI NP-5769 and

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Basis boundary bolting in NUREG-1339 and in EPRI NP-5769. NUREG-1339 as referenced in GALL-Document, SLR XI.M18, Element 4. If not, please Section 4.2, Preventive Actions, in the basis document enhanced the procedures to incorporate (XCELMO00017-REPT-073) refers to NUREG-1339 and the applicable guidance from EPRI NP-EPRI NP-5769 as two of guidelines to prevent or mitigate 5769 and NUREG-1339 accordingly.

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.

B.2.3.29 ASME Section XI, Subsection IWE Question LRA Section LRA Page Background / Issue Discussion Question / Request Number (As applicable/needed)

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 a) Provide recent 2 program health program. Also, second bullet under Action Request reports for the ASME SC XI - IWE Examples mentions IWB, IWC, and IWD program program for verification that the owner, but not IWE. program performance was exceptional as stated in the SLRA. If health ePortal AMPS/XI.S1 Section XI IWE folder includes assessment not performed or not B-220, B- Appendix J Program Health Report and not the IWE available for IWE program, please 1 B.2.3.29 222 program health report. explain.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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 a) Provide the AR/WO that completed surfaces were completed during 2021 RFO. The the subject inspections for verification applicable plant procedure was updated in May 2021 and, referring to the AR/WO, explain to add inspection requirements for external surfaces how the inspection was completed.

of the drywell penetrations and to provide guidance for visual examination of insulated components. b) Identify and point to the specific procedure and guidance that was The staff was unable to verify the above on the revised to incorporate visual 2 B.2.3.29 B-220, ePortal. examination of insulated components.

The first bullet under Exceptions to NUREG-2191 a) Explain what is meant by the phrase states in part: The assessment concluded that the nor a fatigue waiver is required when drywell shell, non-high temperature drywell the assessment referred to in the penetrations, and penetration sleeves are subjected statement and summarized in the to a small amount of fatigue such that neither fatigue previous paragraph in the SLRA (the analysis nor a fatigue waiver is required. As such, six fatigue waiver conditions in the cracking due to cyclic loading does not require aging ASME code), and as indicated in SIA management for drywell shell, non-high temperature calculation 2100507.308, appears to be drywell penetrations, and penetration sleeves. a fatigue waiver analysis.

(emphasis added).

b) Clearly state what was done in the The same statement is also made in SLRA Section assessment and summarizing the 3.5.2.2.1.5. It is not clear what the phrase in bold results and criteria that was met to means in the context of the assessment being justify the exception; also, clearly state referred to, and it is not clear what non-high the components to which the exception temperature drywell penetrations, and penetration applies.

sleeves mean.

B.2.3.29; c) Clarify the apparently contrarian 3.5.2.2.1.5, Based on review of documents on the ePortal, the statements in the SLRA regarding staffs understanding is that the stated assessment is whether or not the aging effect of XCELMO00017- the Structural Integrity Associates calculation cracking due to cyclic loading requires REPT-076 B-214 & 2100507.308, Revision 0 (9/29/2022), Fatigue aging management and the basis for (PBD) Section B-215, Exemption of the Monticello Drywell, and noted it that determination.

3 4.3 3.5-24 performed a fatigue waiver analysis of select drywell

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions components (without CLB fatigue analysis), including d) Clarify if there are two exceptions or non-high temperature mechanical and electrical one exception being taken to the XI.S1 penetrations, in accordance with NE-2222.4(d) AMP and clearly identify each Vessel Not Requiring Analysis for Cyclic Operation exception. If there are two, explain the of the ASME code, 1974 edition and demonstrated distinction between the two exceptions.

that the fatigue waiver criteria therein were met. This Clearly identify the specific components would justify that cracking due cyclic loading does not for which the exception(s) apply.

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.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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 a) For clarity of the staffs NX-8291-2-B; Ref 9.17, Ref 9.18; Ref 9.19]. understanding, explain how the program is supplemented for aging PBD Section 4.2 Preventive Actions does not management of potential corrosion of appear to include monitoring of refueling seal drains inaccessible areas of the drywell shell for blockage and leakage. exterior.

The preventive actions program element of GALL- b) Clarify for the staffs understanding if SLR AMP XI.S1 recommends ensuring that the sand refueling seal drains (drywell to reactor pocket area drains and/or refueling seal drains are building and RPV to drywell) exist. If B.2.3.29, clear. they do exist, explain the configuration XCELMO00017- (preferably illustrated on a drawing)

REPT-076 The staff needs clarification of how the program is and, whether and how (procedure) the (PBD) Section B-214; supplemented and, to verify consistency, whether refueling seal drains are inspected for 4 4.1 PBD p8-9 refueling seal drains exist and inspected. blockage and leakage.

a) Clarify where the referenced statement is made on the Ref 9.23 PBD Section 4.2 states in part: High-strength drawing or how the determination was structural bolting is not used in the structural design of made.

the MNGP steel containment vessel [Ref 9.23].

b) Clarify the term high strength XCELMO00017- Ref 9.23 is Drawing NX-8291-76, Revision C. structural bolting in terms of REPT-076 yield/tensile strength. Also discuss if (PBD) Section The copy of the drawing on the ePortal is not clear high strength bolting will be used in the 5 4.2 PBD p10 regarding the referenced statement above. future during or prior to the SPEO.

The SLRA includes an enhancement to detection of aging effects program element (corresponds to SLR a) Specify in the Commitment 32(c)) regarding supplemental surface or enhancement/commitment the enhanced visual examinations to detect cracking of temperature threshold above which B.2.3.29, Table B-216, A- high temperature piping penetrations. defines high temperature piping 6 A-3 (item 32) 86 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.

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 a) Clarify if there are common and SCC, respectively. components subject to actions specified in both Commitments 32(c)

B.2.3.29, Table B-216, A- It is not clear if there are components that are subject and 32(d) and explain the basis for that 7 A-3 (item 32) 86 to both these commitments. determination.

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 a) Provide a revised enhancement to the detection of aging effects program The trigger specified in the GALL-SLR is the site- element in SLRA Section B.2.3.29, specific occurrence or recurrence of the stated plant- related to one-time supplemental specific OE without regard to the method, program or volumetric examination, and the process by which (how) it is identified. Contrary to associated LR Commitment 33(f) that this, the SLRA enhancement states that the triggering would make the AMP program element OE would be specific to that identified through code consistent with that in GALL-SLR AMP inspections (emphasis added), which would be an XI.S1, neutral to how (method, program unjustified exception to the GALL-SLR AMP XI.S1. or process) the triggering operating Corrosion of the containment metal surfaces that experience is identified.

originated from the inaccessible side could be identified by several other means such as during b) Additionally, also provide an repair/replacement activities, maintenance rule implementation schedule (relative to inspections and walkdowns, Appendix J activities, in the date of occurrence of the triggering B.2.3.29, Table B-217, A- addition to IWE code inspections, and should OE) for implementation of the one-time 8 A-3 (item 32) 86 therefore be neutral to program or process of its 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.

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 Specify with clarity in the DMWs associated with greater than 140oF stainless enhancement/commitment 32(f) how steel piping systems until cracking is no longer sample expansion will be conducted detected. when SCC is detected or absence of SCC cannot be confirmed as a result of There appears to be lack of clarity and context the initial sample of supplemental one-(something missing) in the above referenced time inspections.

statement regarding each additional inspection is with respect to what result in the sample inspected. b) Provide a description of the proposed one-time inspection to B.2.3.29; Table Also, the A.2.2.29 USAR supplement does not confirm the absence of SCC aging A-3 (item 32); B-217, A- appear to include a description of the proposed one- effect in the USAR supplement 9 A.2.2.29 86, A-28 time inspection to confirm the absence of SCC. description for AMP.

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 IWE Plan discusses on page 1.2-9 Category E-C, the current 3rd CISI interval, and (b) for Containment Surfaces Requiring Augmented the previous (2nd) CISI interval.

Examination.

b) Provide on ePortal, if available, It is not clear to the staff what containment surfaces representative photographs of the CISI IWE PLAN are subject to augmented examination in the current recent general material condition of the 10 (Procedure) 1.2-9 and previous CISI intervals. 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.

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 PBD Section 4.5 states that based on UT inaccessible areas of drywell shell examinations performed in 1986, 1987 and 1996, the continues to remain insignificant.

loss of material due to corrosion of inaccessible areas of the drywell is insignificant. c) Provide, if available, photographs XCELMO00017- representative of recent general REPT-076 It is not clear if additional UT examinations of the material condition of the primary (PBD) Section drywell was performed since 1996 and if so, what the containment (drywell, torus, vent 11 4.5 PBD p15 results and conclusions were. system).

SLRA Table 3.5-1, item 3.5.1-004, related to loss of a) Discuss and correct the noted material due to corrosion in inaccessible areas of inconsistencies in the referenced SLRA steel drywell shell, drywell head and embedded shell sections with regard to non-applicability and drywell shell in sand pocket areas, exposed to claim of AMR items 3.5.1-004 and air-indoor (uncontrolled) or concrete environment, 3.5.1-035 for accessible, inaccessible states that these AMRs are Not applicable. and embedded areas of drywell shell given that the component material and SLRA Table 3.5-1, item 3.5.1-035 for the same aging environment for the aging effect exists effects in accessible areas of the drywell and sand for the MNGP drywell (i.e.., to be pocket regions appropriately states the item is consistent with corresponding AMR consistent with NUREG-2191. items in SLRA Table 3.5.2-1.

Table 3.5-1, The first sentence of SLRA FE Section 3.5.2.2.1.3.1 b) Provide a discussion or statement of Table 3.5.2-1, 3.5-44, also includes item 3.5.1-004 and 3.5.1-035 as not whether there has been plant-specific 3.5.2.2.1.3.1, 3.5-78, applicable. However, this SLRA Section also OE of significant corrosion exists in 12 B.2.3.29 3.5-21 concludes: Loss of material due to corrosion 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 covered by FE 3.5.2.2.1.3.1. Also, managed by ASME SC XI - IWE (B.2.3.29) and 10 discuss OE of borated water spills and CFR 50 Appendix J (B.2.3.31) AMPs during SPEO, water ponding, if any, on containment and a separate plant-specific AMP is not required. concrete floor and how it is dispositioned on detection.

Further, SLRA Table 3.5.2-1 appropriately includes AMR line items corresponding to items 3.5.1-004 and c) Revise SLRA Table 3.5-1 and FE 3.5.1-035 since the component material and 3.5.2.2.1.3.1 accordingly.

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.

3.5-25 & SLRA 3.5.2.2.1.6 states, in part: As summarized in a) Clarify and correct the inconsistency 3.5.2.2.1.6, -26; items 3.5.1-10, 3.5.1-038, and 3.5.1-039, cracking between SLRA Section 3.5.2.2.1.6 and Table 3.5-1, 3.5-54, due to SCC is an applicable aging effect when Table 3.5-1 regarding to non-13 Table 3.5.2-1 3.5-76 stainless steel or nickel alloy components are 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 b) Clarify the apparent inconsistency in item is Not Applicable, and accordingly there are no SLRA 3.5.2.2.1.6 regarding the total corresponding AMR items in SLRA Table 3.5.2-1. The number of penetrations that will be referenced SLRA 3.5.2.2.1.6 statement appears to be sampled from for the one-time contrary to the determination in Table 3.5.-1 and inspection to confirm absence of SCC Table 3.5.2-1 with regard to AMR item 3.5.1-038.

c) Provide a discussion or statement of Additionally, SLRA 3.5.2.2.6 on page 3.5-26 states, in whether there has been plant-specific part: .. ..the IWE AMP will be enhanced to included OE of SCC for the components one-time volumetric/surface examination of 20 covered by the further evaluation.

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.

The reference further evaluations do not appear to Provide a discussion or statement of provide a discussion or statement of plant-specific OE plant-specific OE of significant 3.5.2.2.1.3.2 & of significant corrosion for the components covered by corrosion for the components covered 14 3.5.2.2.1.3.3 3.5-22 the further evaluations (FEs). by the FEs.

10 CFR Part 50, Appendix J Question LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

GALL-SLR Report Section XI.S4 Element 4-Monitoring and Trending states in part: In the case of Option B, a) Clarify if condition 4.4.2 of acceptable performance in prior tests meeting leakage EWI.08.06.02 is consistent with NEI 94-rate limits serves as a basis to adjust the testing interval. 01 for reestablishing extended testing intervals.

Section 4.4 of MNGP LLRT Extended Eligibility EWI- Determination report {EWI-08.06-02} states: For a b) Clarify if components MO-2075 and 1 08.06.02 Page 3 component to be eligible for extended test interval, two 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 interval or extended interval as Table 7.2 NOT be identified as ineligible. 4.4.2 - Component must of EWI-08.06-02 is unclear.

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.

Masonry Wall Program Question LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

Section 4.1 in the basis document (XCELMO00017-REPT-079) identifies masonry walls in several locations Masonry for the scope of the AMP, including Reactor Building, Walls Turbine Building, Plant Control and Cable Spreading Basis Basis doc Structure, Emergency Filtration Train Building, Intake Document, Pg. 7 of Structure, Emergency Diesel Generator Building/Diesel Sections 18 Oil Transfer House, Off Gas Stack, Radwaste Building, 4.1 Substation Yard, and Miscellaneous Structures Inside the Protected Area. Please identify the corresponding AMR line items for the scope of Masonry Walls SLRA SLRA Table 2 lists the items associated with the Masonry AMP in SLRA Table 3.5.2-5 Emergency Table SLRA Pg. Walls AMP. These corresponding Table 2 AMR line Filtration Train Building. Update SLRA 1 3.5.2-5 3.5-94 items are under Table 3.5.2-4 Emergency Diesel 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.

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 Masonry Basis doc procedure, but does not mention monitoring of mortar Walls Pg. 8 of joints.

Basis 18 Document, Although SLRA Section B.2.3.32 states that the AMP will Sections be enhanced to monitor and inspect for loss of material Please explain whether monitoring of 4.3 at the mortar joints and gaps between supports and mortar joints is a parameter that needs to SLRA masonry walls (page B-235), monitoring of material loss be monitored or inspected under the SLRA Page B- at mortar joints is not included in its enhancement list Masonry Walls AMP. Enhance the Section 236 and (Page B-236) and Commitment Table A-3 (Commitment procedure if necessary and update 2 B.2.3.32 A-90 # 35 on page A-90). SLRA accordingly.

GALL-SLR XI.S5, Masonry Walls, element 5, notes that Masonry inspection results are documented and compared to Please indicate where in the Walls previous inspections to identify changes or trends in the implementation procedure a comparison Basis condition of masonry walls. of inspection results with previous Document, Basis doc inspections is included. If not, update Sections Pg. 9 of Section 4.5 of the base document, Monitoring and SLRA accordingly to include an 3 4.5 18 Trending, does not explicitly mention comparing 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.

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 Masonry of the wall (i.e., acceptance by further evaluation).

Walls Basis It is unclear to staff whether section 4.7 of the base Document, Basis doc document describes acceptance by further evaluation as Please indicate that acceptance by Sections Pg. 10 of a corrective action option, and how that is discussed in further evaluation is a corrective action 4 4.7 18 the implementation procedure. option in the implementation process.

Corrosion-Structural Question LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

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.

Please clarify if any 3.5.1-98 associated SLRA Table 1 Item 3.5.1-98 identifies 3.5.1-98 as not GALL-SLR items (III.B1.1.TP-4, used, and states its component, material, and III.B1.2.TP-4, III.B1.3.TP-4, III.B2.TP-4, environment combination is addressed by item number III.B3.TP-4, III.B4.TP-4, and/or III.B5.TP-3.5.1-099. 4) are used at MNGP.

If using SLRA AMR Item 3.5.1-099 as an SLRA Table 1 Item 3.5.1-99 credits the One-Time alternative, please describe why the One-Inspection (B.2.3.20) AMP Time Inspection AMP or IWF AMP is Pages or the ASME Section XI, Subsection IWF needed and how the AMP is involved in SLRA 3.5-74 (B.2.3.30) AMP to manage loss of material of ASME Class addressing item 3.5.1-098 for Table and 3.5- 1, 2, 3, and MC aluminum and stainless steel supports component, material, and environmental 1 3.5-1 75 exposed to air and condensation. 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

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 Is there no a(2) Spatial/ Structural Management Review. SLRA Section 2.1.4.2.1 Non-Safety requirement for this pipe and valves FO-Related SSCs with Potential to Prevent Satisfactory 7-3 and FO-7-4? Please provide the accomplishment of Safety Functions appears to be of basis for not being subject to Aging 1 2.3.3.6 2.3-35 relevance for this pipe configuration. Management Review.

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

• Please provide the basis for not Aging Management Review. These components directly identifying Components Diesel Oil support the REQUIRED ACTION A.1 for LCO 3.8.3 Receiving Tank T-83; Pump P-92; and Condition A which reads Fuel oil level < 7-day supply and tank influent and effluent piping red as

> 6-day supply in storage tank. with a COMPLETION a(2) Functional and not subjecting these TIME of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. components to aging management 2 2.3.3.6 2.3-36 10CFR50 Appendix A Criterion 17 reads in part (emphasis 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.

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: (a) Does each EDG base assembly contain its own 7-day (i.e., 165 gallons)

SR 3.8.3.2 supply of lubricating oil?

This Surveillance ensures that sufficient lubricating oil (b) Is there a separate lube oil transfer inventory is available to support at least 7 days of full load system from a lube oil storage location to 3 2.3.3.6 2.3-35 operation for each EDG. The lube oil volume equivalent 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

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 Staff requests clarification for the License Renewal Note 2 does not apply. Drawing Note 2 meaning of See Note 2 for green box at 1 2.3.3.8 2.3-41 would suggest an Appendix R a(3) red Function. Valve RHRSW-68.

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

• Staff requests clarification for the 2 2.3.3.8 2.3-41 Note #2 nor the Subsequent License Renewal Notes. correct interpretation of Note 2

• The Component Types Bolting (Closure) [page Staff requests confirmation that these 576/1519] and Piping, Piping Components [pages Component Types buried in soil reflects 582/1519 & 584/1519] contain line items with an actual plant conditions. In particular, 3.3.2.1.8 Environment of Soil (External). From review of the where on the SLRA boundary drawings Table ESW SLR drawings it is not apparent where this are the buried piping and bolted 3 3.3.2-8 environment exists for the subject Component Types. 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed) a) Staff requests confirmation that the Component Types are accurately interpreted.

b) Staff request confirmation that these component types are within the domain Scoping/Screening Boundary Drawing SLR-36042 of Piping Elements and/or Piping, Reactor Building Cooling Water System [M-111]: Piping Components listed in Table 2.3.3-14 subject to AMR.

• Component Types Radiation Monitor Well? (Coord. A- Staff requests confirmation that the 5, RW-17-302) Thermowell (Coord. C-7, TE-1720) and instrumentation associated Radiation Flow Element (Coord A-5, FE-4145) are displayed on Monitor Well does not have an a(2) 1 2.3.3.14 2.3-57 this drawing. Functional (i.e., red) Intended Function.

Scoping/Screening Boundary Drawing SLR-36042-2 Reactor Building Cooling Water System [M-111-1]:

• Is the scoping methodology of SLRA a(2) Spatial/Structural: Piping inside Containment from Section 2.1.4.2.2 Non-Safety Related Penetrations X-24 & X-23. The structural supports inside SSCs Directly Connected to Safety-containment associated with this piping versus the Related SSCs that Provide Structural leakage boundary spatial concerns are two unique Support for the Safety-Related SSCs concerns and almost always are not mutually exclusive. satisfied for Containment from 2 2.3.3.14 2.3-57 However, there can be exceptions. 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

(a) Is the fact that the flow element is a non-class 1 component per the commodity group guidance of SLRA Scoping/Screening Boundary Drawing SLR-36254 section 4.3.6 ASME Section III, Class 2 Reactor Water Cleanup System and 3 and ANSI B31.1 (page 1056/1519)

(a) Coordinate C-6 displays Flow Element FE-6016 as why the component type of piping a(1) or as A(2) Functional in red for RWCU isolation elements is not listed in Table 2.3.3-15.

based on excessive flow into secondary containment -

there is no flow element listed in SLRA Table 2.3.3-15. (b) 3 ED See Note 2 is located in the vicinity of a pipe with a reducer and (b) Note 2 on drawing SLR-36254 Coord A-5; reads The blank flange? -. Is not 985 Pump Room in scope segment of these lines terminates as they enter with Clean-up Filter / Demin(s) T-202A &

the 985 Pump Room does not contain safety related B contained in the boxed area directly equipment as confirmed by walkdown documented in above the note? What is the correct ML0630504140. interpretation of this Note?

(c) Coordinate A-2; Temperature limits established for TIS12-157 & TE 12-97 satisfy an a(2)

TIS-157 and TE 12-97 (thermo well?) for pump discharge Functional need. Please justify why the (i.e., upstream) of the demineralizers, prevent damage to Piping Element (thermowell?) is not the ion exchange resin beds by ensuring temperature identified as A(2) Functional and subject 1 2.3.3.15 2.3-60 limits are not exceeded. to Aging Management Review.

Scoping/Screening Boundary Drawing SLR-36255 (a) Differential Switches dPIS 12-4-72A/B Filter/Demineralizer System: are color coded green A(2)

(a) Coordinates D-3 and D-5: Y-strainers (YS12-54A & Structural/Spatial; Is not a(2) Functional 54B) prevent Clean-up Filter / Demin resins beads from red more accurate?

entering the Reactor Coolant System (b) Assuming a(2) Functional is (b) Table 2.3.3-15 contains a line item for neither the appropriate, then provide the basis for Component Type Strainer nor Piping Element with an not including the strainer element with a Intended Function of Filter consistent with the guidance filter function subject to AMR?

2 2.3.3.15 2.3-60 of NUREG-2192 Table 2.1-5 Typical Passive

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Component-Intended Functions. (c) Differential pressure monitoring is included as an A(2) Structural /Spatial (c) Instruments dPT-12-4-69A/B at Coordinates D-5 & D-2 interaction - please justify A(2) monitor Structural/Spatial (green) versus. A(2) differential pressure, to ensure the design limits on filter/ Functional (red) demineralizer septums are not exceeded.

Scoping/Screening Boundary Drawing SLR-36255 Filter/Demineralizer System: Why are these instruments WSSLR (i.e.,

Within Scope of Subsequent License a) RWCU Instrumentation and tubing throughout drawing Renewal)? spatial interaction appears is color coded as green indicating an A(2) to be the most logical reason OR Spatial/Structural function. However, the piping and are these instruments under the EQ components that these instruments are attached to are program? - if so, they should be color 3 2.3.3.15 2.3-59 color coded as black (i.e., out of scope). coded as a(3) (red) per the Legend.

2.3.3.16 Service and Seal Water Question LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

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 Why not color coded red for (a)(2)

ESW components (piping and valves) have an a(2) Functional instead of a(2) 1 2.3.3.16 2.3-61 Functional function per Note 3. Spatial/Structural?

SLRA Section 3.3.2.1.16 contains the following:

SLRA Table 3.3.2.-16 Service and Seal Water -

Summary of Aging Management Evaluation: The SSW System components are For the valves and piping subject to aging management exposed to the following environments:

review due to a(2) Spatial/Structural as displayed on SLR-

• Air - Indoor Uncontrolled 3.3.2.1.16 36665-3 Biocide Injection System, the internal

• Condensation Service 3.4-108 environment consists of Biocide and Dispersant. SLRA

• Lubricating Oil and Seal through Table 3.3.2.-16 has no internal environment for valves

• Raw Water 2 Water 3.4-111 and piping to address these unique internal environments.

• 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

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 a) Please provide the basis for not listing b) Coord. C-1; Piping configuration from Main Condenser this Component Type in Table 2.3.4-6.

penetrations #47 and #37 to Turbine Vacuum Trip No. 1 and No. 2 - It is not clear whether these components b) Since these two one inch lines are being scoped as (2) Spatial/Structural are within the TGS integral with the overall integrity of the or CDR system. Condenser vacuum and its plate out AST function, should not these lines be color c) Coord. A-2: Subsequent License Renewal Notes: 2. coded as a(2) Functional?

Components Shown in Red for the Turbine Generator System (TGS) are in Scope for a (A)2 Functional function c) Please provide clarification of which in Support of Alternate Source Term. It is not apparent TGS components are of relevance to from review of SLR-36033 which components are of Note 2.

1 2.3.4.6 2.3-80 relevance to this Note.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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 Please provide the staff with clarification 2 2.3.4.6 2.3-80 Appear to have an a(2) Functional intended function. of these Notes.

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

• Staff requests a clarification discussion (Lubricating Oil Dump Over Flow Tanks) - It is not with Licensee of Background/Issues (a) 3 2.3.4.6 2.3-80 apparent from review of the SLR Boundary Drawings to 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.

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

• Staff requests a clarification discussion Oil Pump) listed on Table 2.3.4-6 is located within the Oil with Licensee of Background/Issues (a) 4 2.3.4.6 2.3-80 Tank (i.e., Turbine Generator Oil Reservoir). How does 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

The staff would like to discuss the AMR Item number 3.4.1-106 is copper alloy (with >15% physical configuration of all heat Zn or >8% Al) piping and piping components exposed to exchanger tubes managed by this line air and condensation, which is susceptible to stress item.

corrosion cracking and managed by the External Surfaces Monitoring of Mechanical Components Program. The Specifically, since the Ext. Surfaces discussion section of Table 3.4-1 states that this line item Program requires a minimum 20% of the Section is also applied to heat exchanger components made of surface area to be examined, discuss 3.3 the same materials that are exposed to condensation, how the visual exam will meet the 20%

Table uncontrolled indoor air, and outdoor air. minimum surface area if the heat 1 3.4-1 3.4-38 exchanger tubes in question are finned.

Water Chemistry

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

Appendix D.5 of BWRVIP-190 indicates that, ((

)) Note 7 of table 2-12 of the 2019 interim guidance of BWRVIP-190 also states, ((

• Please discuss how guidance from Appendix D.5 is incorporated into the Water Chemistry program at MNGP.

)) It is unclear to the staff Please discuss how note 7 from table 2-how guidance from Appendix D.5 and how note 7 of table 12 of the 2019 interim guidance is 2-12 is incorporated into the Water Chemistry Program at incorporated into the Water Chemistry 1 N/A N/A MNGP. Program at MNGP.

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

• Please discuss how the notes from applied to how feedwater hydrogen concentration is table 2-12 are applied in the MNGP measured at Monticello Nuclear Generating Plant approach to measuring feedwater 2 N/A N/A (MNGP). hydrogen concentration.

Question 3 of the initial breakout questions asks how the limit for reactor water insoluble iron was developed

• Please step through how the guidance because it is not listed as a diagnostic parameter for informed the limit for reactor water reactor water at power operation conditions in BWRVIP- insoluble iron at power operation 190. The response provided links the development of this conditions.

limit to feedwater iron and the staff is unclear about this

• Please clarify how this limit is related to connection and how the limit for insoluble iron was feedwater iron.

3 N/A N/A developed.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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 Please clarify how MMS availability is incorporated into plant chemistry implementing tracked and incorporated into plant 4 N/A N/A documents. chemistry implementing documents.

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 Please clarify the continuous response states that the parameters are measured measurement of HWC availability as continuously but the calculation is done daily. However, listed in table 3.16.1 in II.05.

the staff is unclear about the HWC availability frequency 5 N/A N/A as stated in II.05.

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 Please clarify how the water chemistry response it was indicated that one time inspection would AMP is being used to manage loss of be used, and water chemistry would not be adjusted to material for these components given that manage the aging effects. It is unclear to the staff how the the chemistry is not specified or adjusted Water Chemistry AMP will be used to manage loss of to prevent loss of material for carbon 6 N/A N/A material for these components. steel.

4.6.1 Fatigue of Cranes Question LRA Section LRA Page Background / Issue Discussion Question / Request Number (As applicable/needed)

SLR Page 4.6- Reactor Building and Turbine Building Cranes load Application: 2 cycle limits were projected through the SPEO in Subsection accordance with 10 CFR 54.21(c)(1)(i) in the 1 4.6.1 Fatigue of and application. Describe this TLAA dispositioning.

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

Page 14 However, Reactor Building and Turbine Building and of 15 Cranes load cycle limits were projected through the SPEO in accordance with 10 CFR 54.21(c)(1)(ii) in the Report: report.

XCELMO00017-REPT-091, Rev. 0 B.2.3.34 Inspection of Water-Control Structures Associated with Nuclear Power Plants 2.4.9 Intake Structure Question LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

1. GALL-SLR XI.S7 states that the scope of program includes sluice gates and trash racks. The staff could not Scope of program:

locate any information of the sluice gates and trash racks 1. Clarify whether MNGP has sluice in SLRA. gates and trash racks subject to aging management within the scope of SLR.

2. SLRA Section 2.4.9 states that the Intake Structures consists of an inlet channel open to the river, an 2.(a) Clarify what water-control structures uncovered, reinforced concrete forebay and a reinforced (Intake Structure, Inlet Channel, Access concrete chambered structure that encloses traveling Tunnel, Diesel Fire Pump House, East screens, various pumps and water passages. SLRA and West Service Bay, Travel Screens, Section 2.4.9 also states that in addition to the INS itself, etc.?) are included in the scope of SLR; this structure also covers the access tunnel and Diesel 2.(b) Clarify whether Inlet Channel is Fire Pump House. subject to aging management within the scope of SLR; and SLRA Section B.2.3.34 states that Intake Structure is a 2.(c) Clarify whether traveling screens only structure for water-control structure. are subject to aging management.

B.2.3.34 It is unclear to the staff what water-control structures are 3. Explain why Discharge Structure is not 2.4.9 B-243 included in the scope of SLR within the scope of SLR.

2.3.3.3 2.4-20 Table 2.3-28 It is also unclear to the staff whether the traveling screens 4. Update SLRA accordingly based on 1 2.2-1 2.2-2 are subject to the aging management. 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.

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. Parameters Monitored or Inspected:

1. Evaluate the parameters monitored or

2. AMP basis document lists parameters monitored or inspected for the ASR, and provide an inspected for Earthen Embankment Structures, Dams, enhancement if necessary.

and Canals. However, SLRA Section B.2.3.34 states that Intake Structure is a only structure for water-control 2. Clarify whether Earthen Embankment structure. AMR 3.5.1-058 in SLRA Table 3.5-1 states that Structures, Dams, and Canals are earthen water control structures, dams, embankments, applicable, and address the reservoirs, channels, inconsistency between SLRA and AMP and ponds are not credited at MNGP. basis document.

It appears there is inconsistency between SLRA and AMP 3. If Inlet Channel is in the scope, provide basis document. parameters monitored or inspected for the Inlet Channel, and evaluate whether it B.2.3.34 B-243 3. GALL-SLR XI.S7 states that parameters monitored or will be consistent with non-applicability 2 2.4.9 2.4-20 inspected for channels and canals include erosion or 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.

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.

Detection of Aging Effects:

The staff noted that an enhancement to the Structures Inspected:

Monitoring program includes provisions for increased 1. Evaluate whether an enhancement is inspection frequency based on an evaluation of the needed to include provisions for observed degradation. increased inspection frequency based on an evaluation of the observed

2. AMP basis document states that the intake structure is degradation.

inspected below the waterline using divers.

2. Is there a procedure for the divers how The staff could not locate the procedure how divers to inspect the Intake Structure below the 3 N/A N/A inspect the Intake Structure below the waterline. waterline? If not, explain it.

1. GALL-SLR XI.S7 states that degradation of piles and sheeting are accepted by engineering evaluation or Acceptance Criteria:

subject to corrective actions. 1. Describe the degradation of sheet piles if present, and provide acceptance AMP basis document does not describe acceptance criteria for sheet piles.

criteria for sheet piles listed in SLRA Table 2.4-9. The staff could not locate OE related to these sheet piles. 2. Clarify whether an enhancement to the Inspection of Water-Control Structures Table 2. AMP basis document states that quantitative second AMP is needed to include this 4 2.4-9 2.4-23 tier acceptance criteria are used to evaluate the need for 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.

Table XI- GALL-SLR Table XI-01 XI.S7 FSAR Supplement states Appendix A - USAR Supplement:

01 that the program also includes structural steel and Provide the missing information of USAR XI 01-33 structural bolting associated with water-control structures, in Appendix A.2.2.34 to be consistent 5 A.2.2.34 A-30 which is missed in SLRA Appendix A.2.2.34. with GALL-SLR recommendations.

AMR item 3.5.1-096 in SLRA Table 3.5.2-9 is Table corresponding to GALL-SLR item III.A6.TP-34, which Correct the error of GALL-SLR item 6 3.5.2-9 3.5-110 refers to AMR item 3.5-1, 071 in GALL-SLR report. reference.

3.5.2.2.2 Non-Containment Plant Structures Question LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

The SRP-SLR Section 3.5.3.2.1.2 guidance states that 1. Explain how the aging effects the reviewer ensures that the aging effects associated associated with the primary containment with the cooling system are being properly managed or ventilating and cooling system are being temperatures are being monitored to identify a problem properly managed or temperatures are with the cooling system If active cooling is relied upon to being monitored to identify a problem maintain acceptable temperatures. with the primary containment ventilating cooling system.

SLRA Section 3.5.2.2.1.2 states that the bulk drywell temperature is maintained by the primary containment 2. Provide calculations of temperature at ventilating and cooling system. It is not clear to the staff penetrations of biological shield wall for how they are properly managed or monitored. the staff to review.

SLRA Section 3.5.2.2.1.2 states that local area 3. Clarify how local temperature at temperature in the biological shield wall due to hot reactor penetration of the biological shield wall is 1 3.5.2.2.1.2 3.5-20 REC System penetrations is calculated at 179°F; less 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.

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 1. Evaluate the parameters monitored or described in SLRA. inspected for the ASR, and provide an enhancement to the Structures SLR-SLR Section 3.5.3.2.2.3, item 2 states cracking due Monitoring program regarding the to expansion from reaction with aggregates could occur in detection of ASR.

inaccessible concrete areas of Group 6 structures.

2. Explain in SLRA Section 3.5.2.2.2.3, SLRA Section 3.5.2.2.2.3, item 2 does not address how item 2 how aging effects of inaccessible 3.5.2.2.2.1 3.5-28 aging effects of inaccessible concrete areas of Group 6 concrete areas of Group 6 structures will 2 3.5.2.2.2.3 3.5-33 structures will be adequately managed. be adequately managed.

AMR 3.5-1, 046 is related to aging effect of reduction of 1. Evaluate the claim of non-applicability foundation strength and cracking due to differential of AMR item 3.5-1, 046, and provide 3.5.2.2.2.1 settlement and erosion of porous concrete sub- table 2 items.

Table 3.5- 3.5-29 foundation. The applicant claims AMR item 3.5-1, 046 to 3 1 3.5-58 be not applicable. This aging effect is related to a). 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 of the Diesel Fuel Oil Transfer House, aging effect exists if either or both conditions present. The Diesel Oil Storage Tank and and the differential settlement of the Diesel Fuel Oil Transfer Offgas Storage Building HTV exhaust House is an example. Therefore, it is applicable. pipe, and explain how their aging effects due to settlement are adequately SLRA Section 3.5.2.2.2.1, item 3, states that with the managed during the SPEO.

exception of the Diesel Fuel Oil Transfer House and Off-Gas Storage Building HTV exhaust pipe, no significant 3. Evaluate whether plant-specific settlement has been observed on any major structure and program is needed.

de-watering systems are not used. It also states that with the exception of the Diesel Fuel Oil Transfer House, 4. Update SLRA accordingly based on cracks, distortion, and increase in component stress the responses above.

levels due to settlement do not require aging management, which is not correct since they are Note: More settlement questions are applicable aging effects, and require aging management documented in the breakout questions in by the Structures Monitoring program. TRP 46, 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.

SRP-SLR Section 3.5.3.2.2.1, item 4 guidance states that a plant-specific AMP is not required for the reinforced 1. Provide ARs and WOs on the Portal, concrete exposed to flowing water if evaluation and describe OE related to leaching for determined that the observed leaching of calcium accessible concrete areas of Groups 1-5 hydroxide and carbonation in accessible areas has no and 7-9 structures at site.

impact on the intended function of the concrete structure.

2. Evaluate whether the observed FE Section 3.5.2.2.2.1, item 4 states MNGP OE does not leaching of calcium hydroxide and 4 3.5.2.2.2.1 3.5-30 indicate leaching has been observed on accessible 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 impact on the intended function of the the structure, but there is no description of the related concrete structure.

OE. The staff also could not find the OE related to leaching during the in-office OE audit.

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 1. Explain whether there are elevated temperature limits (i.e., general temperature greater than temperatures in excess of 150°F general 150 °F and local area temperature greater than 200 °F). area in Groups 1-5 concrete structures.

FE Section 3.5.2.2.2 is related to the reduction of strength 2. Explain whether there are hot pipes and modulus of concrete due to elevated temperatures causing elevated concrete temperatures in Group 1-5 concrete structures. High temperatures in at penetrations in excess of 200°F in Drywell general area and biological shield wall piping Groups 1-5 concrete structures.

penetration local areas discussed in this FE Section belong to the FE Section 3.5.2.2.1.2 for the aging effect in 3. Explain how the reduction of strength the containment. The staff is unclear whether concrete and modulus of concrete due to elevated elements in Groups 1-5 concrete structures are subject to temperatures is adequately managed.

elevated temperatures in excess of 150°F general area 4. Evaluate whether a plant-specific AMP 5 3.5.2.2.2.2 3.5-31 and 200°F local area. is needed.

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 1. Clarify the air content used for material (spalling, scaling) and cracking due to freeze- components in Group 6 structures.

thaw.

2. Discuss the Intake Structure OE CAP 501000022600 on Page 17 of 81 states that the related to freeze-thaw, and the corrective primary cause for the observed concrete matrix actions taken.

deterioration appeared to be a result of salt-infused water infiltration into the intake roof concrete aggravated by 3. Evaluated whether a plant-specific 6 3.5.2.2.2.3 3.5-31 freeze-thaw cycling. AMP is needed.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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 1. Describe the OE related to reaching in hydroxide and carbonation in accessible areas has no the Intake Structures.

impact on the intended function of the concrete structure.

2. Evaluate whether the observed During the on-site audit, the staff noted the leaching on leaching of calcium hydroxide and interior wall and ceiling of intake structure. FE Section carbonation in accessible areas has any 3.5.2.2.2.3.3 does not include the discussions of these impact on the intended function of the 7 3.5.2.2.2.3 3.5-33 OEs. Intake Structures.

AMR item 3.5.1-042 is used for managing aging effect of Loss of material (spalling, scaling) and cracking due to freeze- thaw.

Table 3.5- Provide Table 2 items for this aging 1 The staff noted that Table 3.5.2-4 Emergency Diesel effect in Emergency Diesel Generator Table 3.5-56 Generator Building does not include Table 2 items Building, or explain why it is not 8 3.5.2-4 3.5-89 associated with Table 1 item 3.5.1-042. applicable.

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.

Clarify whether Group 8 structures are Table 3.5- There appears to be inconsistency among AMR items for applicable to MNGP, and update related 9 1 Varies Group 8 Structures. SLRA Sections and Tables accordingly 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.

Provide missing Table 2 AMR items Table It appears that Table 2 AMR items are missing associated with 3.5.1-044 in SLRA Table 10 3.5.2-11 3.5-117 inaccessible concrete areas of Off-Gas Stack. 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

Clarify or reconcile any inconsistent components (e.g., refueling bellows skirt)

In the first paragraph of SLRA 3.5.2.2.1.5, the described as having CLB fatigue analysis components listed as having an existing CLB fatigue TLAAs between SLRA Sections analysis described as TLAAs in SLRA Section 4.5, and 3.5.2.2.1.5 and 4.5, and corresponding for which AMR item 3.5.1-009 is credited, appears to be Table 3.5.2-1 items which credit item partly inconsistent (e.g., refueling bellows skirt) with the 3.5.1-009. Provide conforming revision to 1 3.5.2.2.1.5 3.5-24 components described in SLRA Section 4.5 the SLRA, if necessary.

Clarify the distinction, if any, between The title of SLRA Section 4.5.1 reads Fatigue Analysis of fatigue analysis of the suppression the Suppression Chamber, Vents, Downcomers, and chamber and the torus shell which Torus Shell. It is not clear if there is a distinction between appears to be implied in the title of SLRA 2 4.5.1 4.5-1 suppression chamber and torus shell. Section 4.5.1.

a) Discuss for clarity other transients The TLAA evaluation descriptions in SLRA Sections included in the referenced TLAA 4.5.1, 4.5.2 and 4.52 generally state only the SRV lifts as evaluations, and why they were not the primary transient included in the evaluations. From a included in the SLRA description.

review of calculation 22-014 (SIA Calculation 4.5.1; 2100507.307) on the ePortal, it appears there are other b) Briefly clarify the source of the 4.5.2; 4.5-1 thru transients (e.g., seismic OBE, chugging, post chug) increase factors of 1.26 and 1.47 used 3 4.5.3 4.5-4 included in the evaluations. for power uprate and EPU, respectively.

For clarity, discuss what transients in The TLAAs in Sections 4.5.1, 4.5.2 and 4.52 are each of these TLAA evaluations will be 4.5.1; dispositioned in accordance with 10 CFR 54.21(c)(iii) and monitored by the Fatigue Monitoring 4.5.2; 4.5-1 thru the effects of fatigue will be managed by the Fatigue AMP, and how they are included and 4 4.5.3 4.5-4 Monitoring AMP (B.2.2.1). controlled within the scope of the AMP.

The TLAA evaluation for primary containment process a) Clarify the source of the maximum penetration bellows fatigue states the evaluation was 7000 cycles in the CLB for staff performed as part of ASME Class 2 and 3 and ANSI verification.

B31.1 fatigue evaluation described in Section 4.3.6.

4.5.5, 4.5-5, 4.3- b) Clarify if the term operating cycles is 5 4.3.6 4, 3-16 SLRA 4.3.6 on page 4.3-16 (last paragraph) states, in 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.

The USAR supplement description for Primary Containment Process Bellows Fatigue Analysis states a) Clarify if the reference to SLRA the evaluation was performed as part of ASME Section Section A.3.3.5 was intended to be III, Class 2 and 3 and ANSI B31.1 fatigue evaluation and A.3.3.6, and accordingly correct the is described in Section A.3.3.5. The staff noted that SLRA A.3.5.5 description, as necessary.

SLRA Section A.3.3.5 describes fatigue analysis for ASME Class 1 piping, and not for ASME Class 2 and 3 b) Provide information in SLRA A.3.3.6 and ANSI B31.1 fatigue. (to which reference appears to be made in A.3.5.5 for containment process It appears that the SLRA intended to cross reference penetration bellows) that would explicitly Section A.3.3.6 for non-Class 1 piping and not A.3.3.5. include and provides an adequate However, Section A.3.3.6 makes no mention of bellows to summary description for the TLAA of the which the summary description in A.3.5.5 applies, and subject bellows; OR, explicitly include an therefore does not appear to provide an adequate USAR adequate USAR summary description in supplement for the TLAA. A.3.5.5 itself without referencing A.3.3.6.

Also, SLRA A.3.5.5 states, in part: The bellows are c) State the transients and maximum designed for a minimum number of operating cycles over number of transient cycles for which the the design life of the plant, However, it does not state bellows were implicitly designed for, and A.3.5.5, A-51, A- what the number of cycles for which the bellows were what the corresponding estimated cycles 6 A.3.3.6 47, A-48 designed. for the SPEO.

SLRA Section 4.6.2 and A.3.6.2 do not appear to include a) State the transients / transient LCs the transients or transient load combinations (LCs) and and corresponding number of cycles corresponding number of cycles based on which the based on which the re-evaluated fatigue higher fatigue usage of 0.111 (HPCI penetration) and usage of 0.111 (HPCI penetration) and 0.343 (RCIC penetration) for 40 years were calculated 0.343 (RCIC penetration) for 40 years 4.6.2, 4.6-4, A- and projected to 80 years. were calculated and projected to 80 7 A.3.6.2 53 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 b) Provide a revised Section A.3.6.2 disposition pursuant to 10 CFR 54.21(c)(1). which includes the transients/LCs and cycles evaluated to calculate fatigue usage, and the TLAA disposition consistent with SLRA Section 4.6.2.

SLRA 4.6.2 includes a TLAA for fatigue of HPCI and RCIC Turbine Exhaust Penetrations, which are stated as Provide Table 2 AMR item(s) for Torus Attached Penetrations. From a search of the managing the fatigue aging effect for the SLRA, it appears that Table 2 AMR item is not included HPCI and RCIC Turbine Exhaust torus 8 4.6.2 4.6-4 crediting these TLAAs for these components. attached penetrations.

Irradiation-Structural Question LRA Section LRA Page Background / Issue Discussion Question / Request Number (As applicable/needed) 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, 3.5-36 and provide an overall summary thru 3.5- This is for general understanding of the evaluation presentation of the evaluations in SLRA C1 3.5.2.2.2.6 39 SLRA 3.5.2.2.2.6. 3.5.2.2.2.6.

Under subtitle Gamma Dose Biological Shield a) Provide the following information of Irradiation Evaluation, the SLRA states that the the composition of the concrete used gamma dose though 72 EFPY for the MNGP for the bioshield and RV pedestal: fine biological shield concrete was determined to be 4.85 and coarse aggregate types (including x 1010 rads, which exceeds the SRP-SLR threshold whether siliceous or calcareous); the limit of 1 x 1010 rads for radiation damage to cement type; water-cement ratio; and C2 3.5.2.2.2.6 3.5-38 concrete. 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.

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 a) Describe in sufficient technical detail 1010 rads for radiation damage to concrete, the the stated separate analysis that was SLRA further states, in part: performed of the effects of potential reduction in concrete strength due to However, a separate analysis of the potential gamma radiation and summary of the reduction in concrete strength due to gamma results that would demonstrate radiation above the recommended threshold has structural integrity is assured for all been completed for MNGP. This analysis considered relevant structural components attenuation through the concrete, and potential for (including anchorage) and a plant-radiation induced volumetric expansion (RIVE) of., specific AMP or enhancements to one as well as the impact to gamma heating or more existing AMPs is not necessary considerations. for the SPEO.

As a result, the integrity of the biological shield is b) Describe the evaluation of the assured, and no additional aging management of the increase in temperature in the concrete biological shield concrete beyond the current due to gamma dose heating effects with Structures Monitoring (B.2.3.33) AMP is necessary the results, and how the concrete for aging effects of irradiation during the SPEO. .. temperature acceptance criteria in SRP-SLR Section 3.5.2.2.1.2 or The SLRA does not appear to include information 3.5.2.2.2.2 (as supplemented by SLR-(including references) with sufficient technical detail ISG-2021-03-STRUCTURES) are met of the plant-specific separate analysis to support the with gamma heating included. Also, above conclusion. The staff needs additional clarify the operating temperature and air C3 3.5.2.2.2.6 3.5-38 information of this analysis which is a primary basis to 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.

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 a) Discuss/explain/demonstrate to what penetration local area, which experience extent thermal load and effects from the temperatures of 135°F and 179°F, respectively. inner RV air cavity have been Transware MNT-FLU-001-T-001, Rev 0) Figure 1 considered, if any, in the calculation of shows the existence of an inner and outer RV air the outer cavity temperature, cavity separated by a thermal insulation surrounding particularly on the surface of the the biological shield wall. Structural Integrity concrete.

Calculation Package Evaluation of Concrete Degradation of MNGP Bioshield states that MNGP b) Discuss/explain/demonstrate calculation results in Bioshield Brick Effectiveness whether effects of gamma dose/rays and Calculation of Local Temperature in Biological have been included in concrete thermal Shield Wall conclude that local temperatures at or calculations discussed under issue. If near pipe penetrations range from 137 to 179 oF. this has not been included, describe ANSI/ANS 6.4 states that production of secondary why not.

gamma rays just from rebars can amount to increase of twenty percent of gamma dose. It is not clear c) Discuss/explain/demonstrate whether whether effects of thermal leakage due to reactor the neutron thermalization properties of piping penetration of insulation surrounding the inner concrete may not be of concern in the cavity has been considered in estimation of RV cavity spatial distribution of secondary temperatures. It is also not clear whether the gamma-ray production that could affect aforementioned MNGP calculations considered rebar bonding within the structural 3.5.2.2.2.6; 3.5-38; gamma heating within the bioshield/pedestal concrete portion of the concrete C4 3.5.2.2.1.2 3.5-20 and whether further consideration was given to the 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.

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, Subject SIA Calculation Monticello SLR Concrete 11)

Embrittlement Assessment of Biological Shield Wall

[BSW] cites certain references that staff needs for c) Reference 5 (EPRI Report audit and verification. 3002013084 - 2018 LTO Structural Tools); Reference 14 (PM Bruck et al.

Also, the calculation refers to #3 and #4 restraints SIA Calculation and concrete ring in its Fig 1 on page 8, and it is not d) Explain #3 & #4 restraints and 2200285.301 clear of the configuration and function of these BSW concrete ring of BSW wall in Figure 1 of C5 R0 17, 8 components. subject calculation, and their function.

Assumption 1 in Section 4.0 states, in part: The a) Discuss/explain/clarify the rationale variation of gamma flux along the height of the and its adequacy for use of a generic Monticello biological shield wall [BSW] is taken from a normalized curve for variation of generic curve available in EPRI report [3002011710] gamma flux along core height for a

[4] for a 3-loop PWR. The EPRI curve, reproduced in PWR for MNGP BWR.

Figure 3(b), presents data normalized to the gamma flux at core mid-plane ,,,,height of active core region b) Clarify if gamma dose estimates at is 12 feet, Plant-specific variation of gamma top of structural concrete and top of dose along the height is not available in the design pedestal at 72 EFPY were estimated to input. verify Assumption 1 and its use in evaluation of the structural portion of Also, Section 6.0 on page 17, states in part: ., the BSW and the RV concrete pedestal and anchorage occurs in the RPV concrete pedestal provide these estimates. If not which is sufficiently remote from active core region estimated, justify why it was necessary SIA Calc such that gamma radiation levels can be assumed to verify related assumptions in the 2200285.301 less than threshold levels. Gamma dose at top of calculation and explicitly verify how C6 R0 9, 16 pedestal at 72 EFPY could be added as input to this 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. locations compare to SRP-SLR threshold limits.

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.

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 a) Explain the apparently contradictory radiolysis. .Water can also be removed from the statement regarding radiation heating in concrete by evaporation due to heat generated by the subject SIA calculation versus SIA Calc gamma radiation. .gamma radiation has a Reference 17, which is cited as the 2200285.301 greater effect on the cement paste than it has on the source of the information in the C7 R0 13, 19 aggregate materials. calculation.

Section 5.3 states, in part: The design of the a) Explain whether design check biological shield wall [BSW] was performed using the calculations were performed for normal Working Stress Method of ACI 318-63 [12], using and degraded strength concrete allowable stresses for concrete and the reinforcing (through SPEO) of the BSW for steel. Per a GE APED design specification cited in applicable controlling combination(s) for original calculations [6] (p.23, CA-68-009 Sheet 2), which the increase in allowable stress allowable stresses were increased by a factor of 1.5 by a factor of 1.5 does not apply. If so, for controlling load cases involving the jet forces. identify such controlling load combination (pointing to applicable Resulting demand-to-capacity (D/C) ratios for normal USAR section) and provide the D/C strength and degraded concrete are provided in results.

Table 5 and shown to be less than 1.0.

SIA Calc b) If not, explain/clarify why such a 2200285.301 16, 17, 7, Also, Table 4 Design Input Used in Evaluation on p7 design check is not necessary or how it C8 R0 9 states that main load combinations to include 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, or enveloped by the controlling load in part: The design calculation take loads from combination and allowable stresses for reactor building computer model which is analyzed load combination(s) where allowable using multiple load combinations [LCs}. These LCs stresses can be increased by a factor of are described as including seismic and jet forces. 1.5.

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.

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, The second paragraph on page 3.5-37 states: provide summary of results, how the Irradiation effects on the biological shield concrete, applicable acceptance criteria in SRP-the biological shield structural steel, and reactor SLR Section 3.5.2.2.2.6 were met, or vessel support structureare evaluated below. how the effects of aging would be managed for the concrete pedestal There appears to be no evaluation provided in the structure. Update the SLRA SLRA of the reinforced concrete RV support accordingly.

pedestal. It is not clear if and how the reinforced concrete pedestal, which is in the vicinity of the RV c) If not, justify why an evaluation for and would fall under the scope of the SRP-SLR irradiation effects on the pedestal 3.5-37, 3.5.2.2.2.6 FE was evaluated for applicable concrete was not necessary, or provide C9 3.5.2.2.2.6 3.5-38 irradiation effects. a sufficient evaluation.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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 3.5.2.2.2.6, description and its program elements in the ePortal a) Revise the SLRA B.2.3.33 AMP and Table 3.5-1, PBD do not appear to include reduction of strength; its PBD as appropriate to include Table 3.5.2-1, loss of mechanical properties due to irradiation (i.e., managing the aging effects/mechanism B.2.3.33 & PBD 3.5-38, radiation interactions with material and radiation- corresponding to SLRA item 3.5.1-097 XCELMO00017- 3.5-74, induced heating) as aging effects/mechanism that for which the AMP is credited in the C10 REPT-080 3.5-76, will be managed by the program. SLRA.

Discuss current general material and structural condition of biological shield wall, RV pedestal, RV support skirt and 3.5-36 RV seismic restraint based on thru 3.5- inspections performed in the past. Use C11 3.5.2.2.2.6 41 General material condition of reactor cavity area illustrative photos where available.

SLRA Table 3.5.2-1, includes a plant-specific Note 7, a) Provide a revised plant-specific Note corresponding to AMR item 3.5.1-097 for the 7 in SLRA Table 3.5.2-1 that is Biological Shield Wall which states: consistent with the evaluation in SLRA Section 3.5.2.2.2.6.

Consistent with SLR-ISG-2021-03-STRUCTURES, which allows a plant-specific AMP, or a selected AMP b) Clarify if the corresponding AMR item enhanced as necessary; the Structures Monitoring in SLRA Table 3.5.2-1 with plant-Table 3.5.2-1, 3.5-76, (B.2.3.33) AMP will be used to manage the potential specific Notes 6, 7 also include the C12 3.5.2.2.2.6 3.5-84 for reduction in strength, loss of mechanical 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 support function, and revise irradiation near the reactor vessel, as the projected accordingly.

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.

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 a) Discuss/explain/demonstrate the concrete is the maximum-power reactor statepoint discrepancy between the SLRA and condition that was determined to occur in Cycle 28. Attachment 1 to MNT-FLU-001-R-The SLRA also states that the fluence methodology 001,Rev 0. If there is no discrepancy implemented by TransWare RAMA methodology is state why.

capable of predicting specimen activities within the b) Discuss/explain/demonstrate why MNGP reactor pressure vessel and [b]ased upon different exposures resulting from these results, there is no discernable bias in the varied cycle to cycle fuel designs would computed reactor pressure vessel fluence for the not constitute a bias.

period of Cycle 1 through the end of Cycle 30 for the MNGP reactor. Attachment 1 to Transware c) What fuel assemblies are considered Proprietary MNT-FLU-001-R-001,Rev 0, however, for projected cycles beyond Cycle 28?

states that each fuel design has a different power Does this differ from the rest of the 3.5.2.2.2.6; 3.5-37 signature in the core and, therefore, results in fluence analysis in MNT-FLU-001-R-F1 4.2.1.1 4.2-3 different spatial power, [and] exposure 001?

ANSI/ANS-6.4-2006, Nuclear Analysis and Design of Discuss/explain/demonstrate whether Concrete Radiation Shielding for Nuclear Power the RAMA or any methodology was F2 3.5.2.2.2.6 Plants, widely discusses the production of secondary 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 of secondary gamma rays in areas of neutron and gamma-rays. It states that "in many concrete exposed to radiation (neutron, reactor shielding situations, the secondary gamma gamma). If not, explain why the radiation produced within the primary [bio]shield is a aforementioned secondary production more important contribution to the dose outside the of gamma rays was not pursued in shield than is the neutron radiation. It is not clear augmenting the estimation of gamma whether the RAMA methodology accounts for dose in the RV bioshield structural production of secondary gamma radiation within the concrete and other steel lined concrete bioshield/RV pedestal and if so, whether it was or that encapsulating the BWR steam applied as such to all aspects of MNG irradiated piping.

concrete (including those of radioactive waste tubes, BWR steam piping encapsulated by concrete).

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 On these pages of the referenced SLRA section, the help the staff understand the load 3.5-38 to applicant the described the configuration of the RV transfer path from the RV to the S1 3.5.2.2.2.6 3.5-39 support skirt/concrete pedestal. surrounding concrete (and vice versa).

On this page of the referenced SLRA section, the applicant states a fluence value of 3.25x1016 n/cm2 What is the approximate fluence value at a nozzle location below the beltline, but this (and corresponding dpa value) at the location is above the RV steel support skirt assembly. knuckle region of the RV support skirt The fluence value down at the RV steel support skirt assembly that may be compared to the assembly is expected to be lower, but this value is dpa value for which there is no NDT S2 3.5.2.2.2.6 3.5-39 not provided. shift per Figure 3-1 of NUREG-1509?

a) Explain/demonstrate how the MNGP RV support skirt configuration is bounded by the RV support skirt The evaluation of the RV steel support skirt assembly configuration analyzed in BWRVIP-342.

cites proprietary report BWRVIP-342. However, this report did not specifically analyze the MNGP RV b) Explain/demonstrate how the MNGP support skirt configuration. Therefore, these design basis transients and design questions pertain to plant-specific information that loads (deadweight, operational and safe would demonstrate the applicability of BWRVIP-342 shutdown earthquake, pipe S3 3.5.2.2.2.6 3.5-39 to MNGP. 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.

On this page of the referenced SLRA section, the applicant seems to be relying on the transition temperature analysis (TTA) method in NUREG-1509 Discuss/explain that the initial NDT +

as basis for the evaluation of the structural steel of shift is less than the lowest service the bioshield wall. However, the information provided temperature that the corresponding in SLRA 3.5.2.2.2.6 is not complete nor clear, or it is steel is subject to with sufficient margin.

very fragmentary. The calculation basis, This explanation and discussion, with 2200285.302.R0, is also missing some parts; it has the values from 2200285.302.R0, would most elements, but it doesnt put everything together. need to be upfront and clear in SLRA The concept of TTA is to demonstrate the following 3.5.2.2.2.6, and therefore a supplement with sufficient margin (guidance on sufficient margin to this SLRA section is called for. Also, in NUREG-1509): this section of SLRA has to clearly cite 3.5-40 to Initial NDT + shift < lowest service temperature of the 2200285.302.R0 as the basis for the S4 3.5.2.2.2.6 3.5-41 steel evaluation.

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 2200285.302.R0, Section 5.2: analyzed and results presented either in an SLRA supplement or updated Table This section of the document talks about the selection 3 of 2200285.302.R0. The lowest of +/- 72 and +/- 76 distances from the mid- service temperature of the steel at the core/midplane locations along the bioshield height corresponding location (see question S5 3.5.2.2.2.6 3.5-40 wherein the dpa exposure levels have dropped off. 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.

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- a) Explain Assumption 1 and its basis.

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 b) MNGP is a BWR-3. Explain why/how to the dpa value for MNGP at the bioshield cladding. the fluence variation curve for a PWR This assumption referenced EPRI report 3002011710 from NUREG/CR-5320 is bounding for S6 3.5.2.2.2.6 3.5-40 (2018). MNGP.

2200285.302.R0, Sections 5.3 and 6.0:

What were the design basis load 5.3 states that design basis load combinations were combinations and what was the analyzed in the finite element analyses; 6.0 presents controlling load combination? Provide stress results for the controlling load combination. the appropriate source documents in S7 3.5.2.2.2.6 3.5-40 However, specifics were not provided. the portal.

2200285.302.R0, Section 5.3 describes the stress analysis performed for the structure steel elements of Go over the stress analysis in the bioshield wall. The staff would like an explanation 2200285.302.R0, Section 5.3 to help of the stress analysis to understand the stresses in the staff understand the low stress in S8 3.5.2.2.2.6 3.5-40 the bioshield wall. the structural steel.

Discuss summary or potential docketing General catch-all to discuss portal documents that of these portal documents:

3.5-36 potentially need a summary or docketing, since these BWRVIP-342 thru 3.5- are documents that contain underlying technical 2200285.301.R0 S9 3.5.2.2.2.6 41 bases for this SLRA section. 2200285.302.R0 B.2.3.30; B-224 SLRA 3.5.2.2.2.6 under subtitle Reactor Vessel a) While a plant-specific program may 3.5.2.2.2.6, thru B- Support Steel Irradiation Evaluation, states, in part: not be necessary, describe how the Table 3.5.2-1, 226; Therefore, the integrity of the reactor vessel aging effects due to irradiation Table 3.5.2-7, 3.5-30; supports is assured, and no additional aging embrittlement will be adequately S10 B.2.3.33 3.5-76 management of reactor vessel supports beyond the managed by the IWF AMP that is

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions thru 3.5- current ASME Section XI, Subsection IWF (B.2.3.30) credited for the RV steel support 83; AMP is necessary for aging effects due to irradiation assembly components for the SPEO?

3.5-98 during the MNGP SPEO. Provide corresponding Table 2 AMR thru 3.5- items, noting that it currently is not 103 Also, SLRA 3.5.2.2.2.6 under subtitle Biological included in the GALL-SLR Report, and Shield Structural Steel Evaluation, states, in part: any related changes that may need to As a result, the integrity of the biological shield is be made to the SLRA.

assured, and no additional aging management of the biological shield beyond the current Structures b) Revise the SLRA and the PBD for Monitoring (B.2.3.33) AMP is necessary for aging the Structures Monitoring, B.2.3.33 effects due to irradiation during the MNGP SPEO. AMP to include managing the aging effects of loss of fracture toughness for Nevertheless, while a plant-specific AMP may not be the biological shield steel components necessary, loss of fracture toughness due to for which the AMP is credited. Also, irradiation embrittlement remains an applicable aging provide corresponding Table 2 AMR effect for the RV steel supports for SLR. Although items.

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.

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 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, The second paragraph on page 3.5-37 states: provide summary of results, how the Irradiation effects on the biological shield concrete, applicable acceptance criteria were the biological shield structural steel, and reactor met, or how the effects of aging would vessel support structureare evaluated below. be managed for the steel RV seismic restraint and stabilizer structure during There appears to be no evaluation in the SLRA of the the SPEO. Provide a summary steel RV seismic restraint and stabilizer structure description of the evaluation in the components (brackets, tension rods, couplings, truss SLRA, include applicable AMR items, etc.). It is not clear if and how the steel RV seismic and update the program credited to restraint and stabilizer structure, which is part of the manage the aging effects.

RV supports and would fall under the scope of the SRP-SLR 3.5.2.2.2.6 FE was evaluated for loss of c) If not, justify why an evaluation for fracture toughness due to irradiation effects. Also, it irradiation effects on the RV seismic appears that AMR items are not included in the SLRA restraint and stabilizer structure was not for managing the aging effect for these components. necessary and how the irradiation aging effects would be adequately managed S11 3.5.2.2.2.6 3.5-37 . during the SPEO.

B.2.3.33 Structures Monitoring Question LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

GALL SLR XI.S6 states that the scope of the program Scope of the Program:

B.2.3.33 B-238 includes all SCs, component supports, and structural 1. Clarify the SLR scope for these two 2.4.9 2.4-20 commodities in the scope of license renewal that are not 345kv substation houses and 115/345 kV 2.4.10 2.4-23 covered by other structural aging management programs. Control House, and address the 2.4.16 2.4-35 inconsistency.

1 2.1.4.2.1 2.1-14 1. SLR boundary drawing No. SLR-36444 lists two 345kv

Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application (SLRA) Breakout Audit Questions substation houses, one is in SLR scope, and another one 2. Clarify the SLR scope for 13.8 kV room is not. SLRA Section 2.4.10 states that the miscellaneous and address the inconsistency.

SBO yard structures include the 115/345 kV Control House, which is not in the scope of SLR identified in SLR 3. Clarify whether existing AMP program boundary drawing No. SLR-36444. It is unclear to the staff includes whip-restraints and jet whether 345kv substation house and 345 kV Control impingement shields/barriers in the House are the same structure. scope, and provide AMR items for jet impingement shields/barriers.

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 4. Clarify and include AMR items for defined in SLR boundary drawing No. SLR-36444. seismic joint fillers.

3. SLRA Section 2.1.4.2.1 states that HELB related 5. Clarify whether sump liners, tube structural components such as whip-restraints and jet tracks, and trash racks are within the impingement shields/barriers, along with the piping scope of SLR. If yes, provide AMR items supports are in the scope of SLR. The staff noted pipe for the sump liners, tube tracks, and trash whip restraints in Table 2.4-7, but the staff could not racks associated with water-control locate the jet impingement shields/barriers in SLRA table. structures.

The staff also could not locate any information of whip-restraints and jet impingement shields/barriers in 6. Provide AMR items for the Hot procedure 1385, Revision 17. Machine Shop. Also check which buildings and structural components

4. GALL-SLR XI.S6 states that seismic joint fillers are within the scope of SLR are not listed in within the scope of SLR. SLRA Appendix A.2.2.33 states these tables 1-21 of procedure 1385, and that the Structures Monitoring program includes the evaluate whether procedure 1385 needs inspection of seismic joint fillers. However, the staff could to be revised for an enhancement.

not locate AMR items associated with seismic joint fillers.

7. Explain how aging effects for the

5. GALL-SLR XI.S6 states that sump liners, tube tracks, Diesel Fire Pump House is adequately and trash racks associated with water-control structures managed, and provide AMR items.

are within the scope of SLR. However, the staff could not find the sump liners, tube tracks, and trash racks 8. Confirm and address the inconsistency associated with water-control structures in SLRA. of building names.

6. SLR boundary drawing No. SLR-36444 lists Hot 9. Based on the responses above, Machine Shop in the scope of SLR. Procedure 1385, 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 enhancements to the Structures tables do not include all the buildings and structural Monitoring program are needed, and components managed by the Structures Monitoring update SLRA if necessary.

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.

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, Parameters Monitored or Inspected:

structural sealants, and seismic joint filler, as well as their 1. Clarify whether elastomeric vibration parameters monitored or inspected in SLRA. isolators, structural sealants, and seismic joint fillers are within the scope of SLR. If

2. SRP-SLR describes the indication of Alkali-Silica yes, provide parameters monitored or Reactions (ASR) as map or patterned cracking, alkali- inspected for the elastomeric vibration silica gel exudations, surface staining, expansion causing isolators, structural sealants, and seismic structural deformation, relative movement or joint fillers, and provide AMR items.

displacement, or misalignment/distortion of attached components. AMP basis document lists parameters 2. Evaluate the parameters monitored or monitored or inspected as unique map or cracking that inspected for the ASR, and provide an 2 N/A N/A would indicate the presence of ASR. It appears that ASR 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.

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 Parameters Monitored or Inspected:

find this enhancement to the Structures Monitoring Provide the enhancement as discussed 3 N/A N/A program. in the AMP basis document.

GALL-SLR XI.S6 states that all structures are monitored on an interval not to exceed 5 years. Detection of Aging Effects:

1. Identify those normally inaccessible Procedure 1385, Revision 17 states that the inspection areas monitored on an interval of intervals for those normally inaccessible areas may exceeding five years.

exceed five years.

2. Evaluate whether the Structures It is unclear to the staff where those normally inaccessible Monitoring program has an exception to areas are located. It appears that the Structures NUREG-2191, if yes, provide justification 4 N/A N/A Monitoring program has an exception to NUREG-2191. why this exception is acceptable.

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. Detection of Aging Effects:

1. Reevaluate the procedure how to Table 19 in procedure 1385, Revision 17, states that the adequately manage aging effects of the interior of the Diesel Fire Pump House masonry block accessible interior of the Diesel Fire walls is covered with insulation. The Structures Monitoring Pump House masonry block walls program will require that the interior surfaces of the walls covered by insulation.

will be examined if exterior wall surfaces show evidence of significant aging effects. [Commitment M05006A] 2. Clarify which AMP shall be used to manage this aging effect of these It appears that the procedure considers the interior of the masonry block walls covered by Diesel Fire Pump House masonry block wall covered by insulation, and address the 5 N/A N/A insulation as inaccessible, which conflicts with GALL-SLR 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.

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.

Detection of Aging Effects:

It appears that this enhancement is for Scope of the Evaluate and Clarify which Element 6 B.2.3.33 B-240 Program. needs to be enhanced 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 Monitoring and Trending:

parameters monitored or trended. Evaluate whether this enhancement to the Inspection of Water-Control A.2.2.33 A-30 It is unclear to the staff whether the same enhancement is Structures AMP is applicable to the 7 B.2.3.34 B-244 applicable to the Structures Monitoring program. Structures Monitoring program.

AMP basis document discusses the FPL companys Monitoring and Trending:

response to RAI B.2.3.35-5 (ML18334A182) for the Evaluate whether this enhancement to Turkey Point SLRA, and the implementing procedure will the Inspection of Water-Control 8 N/A N/A be enhanced to strengthen the detail and criteria for 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 Structures Monitoring program.

involves significant corrosion related degradation. It also Monitoring and Trending:

states clarifying in the Element Monitoring and Trending Explain how the implementing procedure to the MNGP Structures Monitoring program that will be enhanced to reflect the site-inspections are to occur in no greater than 5-year specific OE.

intervals.

It appears that the Structures Monitoring program does not include this enhancement.

GALL-SLR XI.S6 describes acceptance criteria for loose bolts and nuts, structural sealants, elastomeric vibration isolation elements, and sliding surfaces. Acceptance Criteria:

Provide acceptance criteria for loose The staff reviewed the AMP basis documents, and found bolts and nuts, structural sealants, that it does not include acceptance criteria for these elastomeric vibration isolation elements, 9 N/A N/A structural components. and sliding surfaces if applicable.

Background:

Operating Experience:

The staffs OE audit identified that Action Request (AR) 1. Explain how settlement baseline was 01500214, dated 6/21/2016, discusses the settlement established.

issue for the Diesel Oil Pump House (Diesel Fuel Oil Transfer House) starting from the late 1970 and early 2. Explain how acceptance values for 1980. Recent construction placed a large concrete slab settlement were determined to ensure over the Diesel Oil Storage Tank (T-44) and large heavy I- that the fuel oil pipe stresses are within beams on the roof of the Diesel Fuel Oil Transfer House, the code allowable limit.

which added significant weight to structure and resulted in rainwater leakage through the foundation. The AR 3. Explain what corrective actions will be 01500214 also notes that the Diesel Fuel Oil Transfer taken if settlement exceeds the House structure is visibly leaning to the west, and the acceptance criteria.

settlement may be adding additional stress to the fuel oil pipes penetrating the west wall and making them 4. Update the SLRA to include the susceptible to fracture. settlement related operating experience for the Diesel Fuel Oil Transfer House, Corrective Action Program (CAP) No. 50001490358, the Diesel Oil Storage Tank (T-44), and dated 3/15/2017, indicates the NW corner of the Diesel the OGSB.

Fuel Oil Transfer House was approaching the lower limit 10 B.2.3.33 B-241 of the settlement range. The estimated elevation was 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 program can adequately manage aging range is 937.79 to 937.87 feet as described in the effects due to the settlement of the Diesel procedure 1396. Oil Pump House, the Diesel Oil Storage Tank (T-44), and OGSB to ensure that Procedure 1396 states that a CAP Action Request [NRC the fuel oil pipes, and any other impacted Commitment M05009A] will be initiated when settlement is SSCs within scope of SLR, can maintain NOT within acceptance criteria. Procedure FP-PE-RLP- their intended functions during the SPEO.

01, Revision 9 implements NRC Commitment M05009A, If needed, provide updated settlement which states Site documents that implement aging criteria with basis ( or process for management activities for license renewal will be determining conditional acceptance enhanced to ensure that an AR is prepared in accordance criteria to ensure intended function when with plant procedures whenever non-conforming the acceptance criteria is not met) conditions are found (i.e., the acceptance criteria is not against which the need for corrective met). actions are evaluated.

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

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.

Procedure 1385, Revision 17 discusses the implementation of NRC license renewal commitments M05006A, M05008A, M05009A, M05010A, M05012A, M05045A, M05046A, M05047A, M05049A, and M05050A, etc. Confirm that all the implementing procedures mentioned in these It is the staffs understanding that all the implementing commitments have been enhanced prior procedures mentioned in these commitments have been to the period of extended operation, and enhanced prior to the period of extended operation, and will continue to be implemented during 11 N/A N/A will continue to be implemented during the SPEO. the SPEO.

AMR item 3.3-1,111 claims to be not used. However, SLRA states that structural steel is addressed as part Clarify which AMR item in Section 3.5 Table of structural items in Section 3.5. It is not clear how AMR addresses the aging effect for AMR item 12 3.3-1 3.3-55 item 3.3-1, 111 is addressed in Section 3.5. 3.3-1, 111.

GALL-SLR report lists building concrete at locations of expansion and grouted anchors; grout pads for support base plates as Tables component in AMR 3.5-1, 055 to manage aging effect of Evaluate whether Table 2 items 3.5.2-4 reduction in concrete anchor capacity. associated with AMR 3.5-1, 055 in Tables and 2.5-91 However, Table 2 items associated with AMR 3.5.1-055 in 3.5.2-4 and 3.5.2-17 are correctly 13 3.5.2-17 3.5-146 Tables 3.5.2-4 and 3.5.2-17 list component of joint and 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.

Table 2 items associated with AMR item 3.5-1, 063 Table provide inconsistent list of component for concrete:

3.5-1 3.5-64 basemat, foundation. For example, Table 2 item in Table Table 3.5.2-3 lists concrete: basemat, foundation as 3.5.2-3 3.5-86 accessible, however, Table 2 items in Table 3.5.2-4 list Evaluate and clarify all Table 2 items Table concrete: basemat, foundation as both accessible and associated with AMR item 3.5-1, 063 for 14 3.5.2-4 3.5-89 inaccessible. the concrete: basemat, foundation.

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 Evaluate and clarify Table 2 items Table that Table 2 item in Table 3.5.2-3 for concrete: basemat associated with AMR 3.5.1-065 in Table 15 3.5.2-3 3.5-86 foundation (inaccessible) is missing. 3.5.2-3.

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 Table associated with AMR 3.5.1-067 for concrete: exterior walls 3.5.2-4 3.5-90 and roof (accessible) and interior walls and roof (accessible) are missing in Table 3.5.2-9, and so on. Evaluate Table 2 items associated with Table AMR 3.5.1-067 and provide missing 16 3.5.2-9 3.5-111 The applicant is requested to evaluate all the Table 2 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.

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 Explain where roofing railroad bay is Table to staff why AMR 3.5.1-072 is used for the roofing railroad located, and what is its function and 17 3.5.2-14 3.5-131 bay. aging effect.

B.2.3.32 Masonry Walls Question LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

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 1 B.2.3.32 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 LRA LRA Page Background / Issue Discussion Question / Request Number Section (As applicable/needed)

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- Provide location of end of gas extends beyond valve DM-100 and doesnt provide 10CFR54.4(a)(2) Spatial/Structural 1 2.3.3.5 2.3-33 location of scope change in piping. classification.

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 Confirm whether service water 2 2.3.3.11 2.3-48 10CFR54.4 a(2) Spatial/Structural. connections are in-scope 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 Confirm whether piping up to HK/HB within 10CFR54.4(a)(2) Spatial/Structural. However, interface or up to the Reactor Building portion of piping shows out of scope up to HK/HB interface should be in-scope per 3 2.3.4.1 2.3-67 interface. 10CFR54.4(a)(2).

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. Confirm whether CST piping is in However, same CST piping on drawing SLR-36039 (B,4) accordance with 10CFR54.4(a)(2) 4 2.3.4.1 2.3-67 shows out-of-scope. spatial/structural SLRA Section 2.3.4.2 Condensate and Feedwater: Confirm this should be drawing SLR-5 2.3.4.2 2.3-70 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 System (GEZIP) as applicable to (Page 2.3-70) lists drawing SLR-11929. However, this system.

drawing was not provided.