ML24156A154

Breakout Questions - Aging Management Audit - Perry Unit 1 - License Renewal Application

ML24156A154
Person / Time
Site:	Perry
Issue date:	06/04/2024
From:	Vaughn Thomas NRC/NRR/DNRL/NLRP
To:
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ML24156A152	List: ML24156A154
References
EPID L-2023-RNW-0002
Download: ML24156A154 (157)

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BREAKOUT QUESTIONS Aging Management Audit Perry Nuclear Power Plant, Unit 1 License Renewal Application November 20, 2023 - April 19, 2024

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Sections A.1.3 and B.2.3 ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program TRP-1 Question Number LRA Section LRA Page Background / Issue Discussion Question /

Request 1

A.1.3 and B.2.3 A-9, B-19 LRA states that periodic self-assessments of the ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD program are performed to identify the areas that need improvement to maintain the quality performance of the program. The staff performed a search of the portal but did not find any related procedures that specify such periodic self-assessments.

Provide (to the portal) site procedures which specify and dictate periodic self-assessments of Perrys lnservice Inspection program. Provide the most recent self-assessment as an example including a summarize of strengths and weaknesses.

Provide brief discussion that illustrates program improvement achieved from the self-assessment.

2 Perry Portal File No.

ATA-2020-14736 Perry Portal File Title Pre-NRC Inspection Self-Assessment on lnservice Inspection Activities -

1R18 A 4th Interval program update review identified that "Several components were identified to have been inadvertently deleted from the Inservice Examination Plan (ISEP) during Revision 22 but were still scheduled in the ISI scheduling database (Iddeal).

It stated that the issue has since been resolved.

Discuss how differences between ISEP and Iddeal are reconciled. What is the frequency of such program review?

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 3

Perry Portal File No.

ATA-2020-14736 Perry Portal File Title Pre-NRC Inspection Self-Assessment on lnservice Inspection Activities -

1R18 Page 5 of the assessment noted program weaknesses. It stated that Weaknesses in vendor UT performance specifically for reactor vessel/specialty examinations.

Discuss whether and how the weaknesses have been addressed.

4 Perry Portal File No.

ATA-2020-14736 Perry Portal File Title Pre-NRC Inspection Self-Assessment on lnservice Inspection Activities -

1R18 The assessment stated that ATA-2020-15085 was generated to monitor an URI related to ten head penetrations potentially needed to be re-added to the Inservice Inspection program for examination.

Provide ATA-2020-15085 and its related URI to the portal. Discuss the status of the URI.

5 Perry Portal File No.

SN-SA-2019-1319 Perry Portal File Title Pre-NRC Inspection Self-Assessment on lnservice Inspection Activities -

1R17 Page 7 of the assessment stated that several closed work orders contained deficiencies.

Specifically, ASME/safety related parts were not recorded in the work order in accordance with NOP-WM-4300, Section 4.8.14. It further stated that Associated ATAs -2019-2180 and 2019-2181 were generated to perform extent of condition reviews.

Summarize the extent of condition reviews associated with ATA -2019-2180 and explain whether and how the deficiency has since been corrected.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 6

Perry Portal File No.

CR-2015-10559 Perry Portal File Title Classification of RPV Leak Detection Line The Operating Experience section on the portal noted that a CR was generated to address a question from an NRC inspection.

The question was regarding Perrys classification of its RPV Leak Detection Line. No further details were provided on the portal.

Provide a full copy of CR-2015-10559. Discuss classification of Perrys RPV Leak Detection Line. Explain why it is not classified as ASME Code Class

1. Explain whether it should be in the scope of the Inservice Inspection program or in the scope of One Time Inspection of ASME Code Class 1 Small-Bore Piping program.

LRA Section: B.2.27 Internal Coatings/Linings for In Scope Piping, Piping Components, Heat Exchangers, and Tanks Program (TRP 12)

Question Number LRA Section LRA Page Background / Issue Discussion Question /

Request 1

B.2.27 B-84 to B-86 The Program Description for this AMP requires, in part, that:

a) Inspections of coatings/linings will be performed for signs of coating failures.

b) When acceptance criteria are not met, physical testing will be performed (where possible) in conjunction with repair or replacement of coating/lining.

The maximum interval of subsequent coating inspections will be consistent with Table 4a of GALL Report AMP XI.M42 in LR-ISG-2013-01.

The Operating Experience section in this AMP describes the inspection and repair of the internal coating of a Division 2 Fuel Oil Tank in 2010. A Condition Report uploaded to the Portal clarifies that this tank was Standby Diesel Generator Fuel Oil Tank PY-1R45A0002B, that the original internal coating was Rustoleum, and that repair of defective areas can be done with Carboline 187.

A Plant Administrative Procedure uploaded to the Portal The staff requests a discussion of the following:

1. Describe the inspection interval and inspection history for tank PY-1R45A0002B subsequent to the 2010 inspection described in the Operating Experience section of this AMP.

2. Discuss the timing and results of any recent inspections of the internal surfaces of the following tanks. For this discussion, consider operating experience from a period of at least ten years, with

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions identifies the following steel tanks which are also internally coated with Rustoleum.

PY-1R45A0002A(B) Division 1 and 2 Fuel Oil Storage Tanks PY-1R45A0003(B) Division 1 and 2 Diesel Generator Fuel Oil Day Tanks PY-1R45A0004 HPCS Fuel Oil Storage Tank PY-1R45A0005 HPCS Fuel Oil Day Tank the period starting a minimum of ten years prior to the Perry LRA submittal.

PY-1R45A0002A PY-1R45A0003(B)

PY-1R45A0004 PY-1R45A0005 LRA Section: B.2.30 - Monitoring of Neutron Absorbing Materials Other Than Boraflex Program (TRP 15)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

B.2.30 B-90 Section B.2.30 states that The Monitoring of Neutron-Absorbing Materials other than Boraflex Program is an existing condition monitoring program. It also states that the condition monitoring program consists of periodic in-situ testing and inspections.

However, in its response to Generic Letter 2016-01, PNPP states that in-situ testing is not performed at PNPP. Therefore, the characterization of this program as an existing program is unclear.

Please provide the most recent in-situ testing results.

Please provide a planned in-situ testing schedule for the Boral Panels.

LRA Section: B.2.42 - Selective Leaching Program (TRP 34)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

N/A N/A N/A The staff requests an operating experience search using the keyword graphiti during the breakout to capture graphitic corrosion, graphitization, and graphitisation.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 2

B.2.42 B-117 LRA Section B.2.42, Selective Leaching Program, states

[t]he program for selective leaching of materials will ensure the integrity of the components made of ductile iron.

UFSAR Chapter 9, page 9A.5-35 states [t]here are also limited portions of fiberglass reinforced plastic and cement lined ductile iron mains. UFSAR Chapter 9, page 9.4-134 references a ductile iron casing with respect to the brine recirculation pumps.

The staff did not identify any aging management review (AMR) line items associated with ductile iron components in the LRA.

Based on its review of the LRA and UFSAR, the staff requests a discussion with respect to whether there are ductile iron components in-scope for license renewal.

3 B.2.42 B-119 The LRA Section B.2.42 operating experience discussion states [a]

single instance of through-wall coating damage to underground cast iron pipe with confirmed evidence of leaching (not in-scope piping) has occurred at PNPP.

Report M14339, BETA Laboratory Failure Analysis Report:

Buried Fire Protection Piping, notes through-wall leaks due to external dealloying corrosion for buried ductile iron piping.

Based on this operating experience, the staff requests a discussion with respect to why one-time inspections are appropriate for in-scope buried components susceptible to selective leaching (i.e., gray cast iron and potentially ductile iron).

4 B.2.42 B-119 LRA Section B.2.42 includes one exception which states

[m]aterials exposed to contaminated fuel oil and water-contaminated lube oil are managed under the XI.M39, Lubricating Oil Analysis and XI.M30, Fuel Oil Chemistry programs NUREG-2221, Technical Bases for Changes in the Subsequent License Renewal Guidance Documents NUREG-2191 and NUREG-2192, states the following with respect to deleting the water-contaminated lubricating oil environment from GALL-SLR Report AMP XI.M33: [t]he staff concluded that there is reasonable assurance that water contaminated lubricating oil would not result in loss of material due to selective leaching sufficient to result in a loss of intended function because it is unlikely that the predominate locations where water could accumulate would be constructed of a susceptible material (e.g.,

tanks, strainers).

The staff requests a discussion with respect to the basis for this exception, it appears to be based on subsequent license renewal guidance.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: B.2.8 - Buried and Underground Piping and Tanks Program (TRP 35)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

N/A N/A CR-2014-10690 notes an underground fire protection pipe break.

CR-2015-05016 references buried air-line leaks.

CR-2019-02683, CR-2020-01366, and CR-2020-08181 reference underground leaks.

The staff requests a discussion with respect to whether the subject break and leaks were due to age-related degradation. If so, the staff requests a discussion on additional details such as piping material, internal vs. external corrosion, degradation mechanism, etc.

2 N/A N/A CR-2013-18952 notes degraded and disbonded coatings on a portion of buried piping in the condensate transfer and storage system.

The staff requests a discussion with respect to if this is representative of the condition of other buried piping in-scope for license renewal.

3 B.2.8 B-34 Enhancement No. 1 to LRA Section B.2.8, Buried and Underground Piping and Tanks Program, states [i]dentify the systems containing buried or underground piping within the scope of license renewal.

The scope of program outlined in NOP-ER-2007, Underground Piping and Tanks Integrity Program, includes piping that is safety-related, contains licensed material, or contains environmentally hazardous material.

The staff requests a discussion with respect to the intent of the subject enhancement. The current program outlined in NOP-ER-2007 does not appear to account for all piping in-scope for license renewal.

It is unclear to the staff why the enhancement does not identify specific systems (e.g., fire protection) not currently included in the program.

4 B.2.8 B-34 B-35 Enhancement Nos. 2 through 5 to LRA Section B.2.8, Buried and Underground Piping and Tanks Program, state the following:

Describe opportunistic inspections.

Describe cathodic protection.

Describe fire jockey pump monitoring.

Describe the directed underground and buried The staff requests a discussion with respect to the intent of these enhancements. It is difficult for the staff to establish intent given the lack of specificity.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions inspections.

5 A.1.8 A-15 LRA Section A.1.8, Buried and Underground Piping and Tanks Program, states [t]he acceptance criteria for the effectiveness of the cathodic protection is less than or equal to

-850 mV (-690 mV for Test Well 50). PTI-R35-P0002, Annual Structure to Soil Potential Surveys, Section 5.7, Acceptance Criteria, specifies measurements are compared to measure ON potentials.

The staff requests a discussion with respect to (a) why instant-off measurements are not specified; and (b) the basis for using an alternative acceptance criterion at Test Well 50.

6 N/A N/A GALL Report AMP XI.M41, Buried and Underground Piping and Tanks, recommends a limiting critical potential of -1,200 mV to prevent damage to coatings or base metals.

The staff requests a discussion with respect to where the limiting critical potential is specified in current procedures.

7 Table 3.3.1 3.3-127 LRA Table 3.3.1, item 3.3.1-107 states [t]here are no nickel alloy components subject to aging management in the Auxiliary systems that are exposed to soil.

UFSAR Section 9.0, page 9A.5-35 states [u]nderground pipe is mainly unlined, unwrapped nickel copper alloy steel.

The staff requests a discussion with respect to how this material, if in-scope for license renewal, is accounted for in the LRA.

8 Table 3.3.1 3.3-128 LRA Table 3.3.1, item 3.3.1-109x states [n]ot applicable.

There are no underground (in vaults or tunnels outside of buildings) aluminum, copper alloy, stainless steel, or nickel alloy piping components subject to aging management in the Auxiliary systems. LRA Table 3.3.2-55, Service Water System, uses item 3.3.1-109x in two locations.

The staff requests a discussion with respect to the discrepancy between the two LRA Tables.

9 N/A N/A LRPY-AMP-XI.M41, Aging Management Program Evaluation Results - XI.M41 Buried and Underground Piping and Tanks Program, Section 2.2.2 states buried steel components are provided with protective coatings The staff requests a discussion with respect to if other metallic materials exposed to soil (i.e., cast iron and stainless steel) are externally coated.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 10 N/A N/A LRPY-AMP-XI.M41, Section 2.7, Corrective Actions, does not include the prescriptive corrective actions in LR-ISG-2015-01, Changes to Buried and Underground Piping and Tank Recommendations, and there is no enhancement related to this.

The staff requests a discussion with respect to this topic.

11 N/A N/A For parameters monitored or inspected, NOP-ER-2007-01, As-Found Buried/Underground Piping Examination Report, and NOP-ER-2007, Underground Piping and Tanks Integrity Program, do not appear to address aging effects for buried fiberglass piping.

The staff requests a discussion with respect to managing the effects of aging for buried fiberglass piping in the fire protection system.

12 N/A N/A PTI-R45-P5199, Diesel Generator Fuel Oil Storage Tank Cleaning and Inspection, discusses performing internal visual examinations of the tanks.

The staff requests a discussion with respect to how external corrosion of the subject tanks will be managed during the period of extended operation.

13 B.2.8 B-33 LRA Section B.2.8 states [i]f damage to the protective coatings is found and the piping surface is exposed, the pipe will be inspected for loss of material due to general, pitting, crevice, or microbiologically influenced corrosion.

Cracking of buried stainless steel piping will be managed using the Buried and Underground Piping and Tanks Program in the fire protection system but not in the condensate transfer and storage system.

The staff requests a discussion with respect to the following: (a) why LRA Section B.2.8 does not state that piping will be inspected for cracking due to stress corrosion cracking if damage to protective coatings is found; (b) whether this implies that buried stainless steel piping is not coated; and (c) why cracking is being managed in the fire protection system but not in the condensate transfer and storage system.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 14 N/A N/A It is unclear to the staff where acceptance criteria (a), (b),

(c), (d), (f), and (i) from LR-ISG-2015-01 are accounted for in current procedures.

Acceptance criteria (l) related to cathodic protection acceptance criteria will be covered in Breakout Question No. 5.

The staff requests a discussion with respect to this topic.

15 A.1.8 A-15 LRA Section A.1.8 states [a]nnual cathodic protection surveys are conducted. The acceptance criteria for the effectiveness of the cathodic protection is less than or equal to

-850 mV (-690 mV for Test Well 50). Where the acceptance criteria is not met, loss of material rates are measured

[emphasis added by staff].

SRP-LR Table 3.0-1, FSAR Supplement for AMP XI.M41, states [f]or steel components, where the acceptance criteria for the effectiveness of the cathodic protection isother than -

850 mV instant off, loss of material rates are measured.

LRA Section A.1.8 states loss of material rates are measured if acceptance criteria are not met; however, the FSAR supplement in SRP-LR states loss of material rates are measured if alternative cathodic protection acceptance criteria are being used. The staff requests a discussion with respect to this apparent discrepancy.

LRA Section: B.2.36 - One-Time Inspection of ASME Code Class 1 Small Bore Piping Program (TRP 36)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

A.1.36 and B.2.36 A-36, B-102 LRA states that One-time inspections will be completed within the six year period prior to the period of extended operation and no later than six months prior to the period of extended operation.

Perrys current license expires on midnight of 11/07/2026. Six-months prior, would place the expected completion of the One-time Inspection of ASME Code Class 1 Small-Bore Piping program at Perry, prior to 05/07/2026. Based on past refueling outages, the expected dates of the next two refueling outages at Perry would be around April of 2025, and April of 2027.

Since, there is not much time left to implement the one-time small-bore inspection program prior to the period of extended operations, provide Perrys expected schedule for the completion of this program, and if necessary, revise the LRA accordingly.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: B.2.25 - Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (TRP 39)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

B.2.25 B-82 LRA Section B.2.25, Internal Surfaces in Miscellaneous Piping and Ducting Components Program, states [i]nternal OE identified instances of leakage from a sink drain line likely caused by corrosive chemicals being drained in the sink. The damaged hard pipe was replaced with new plastic pipe[o]ther OE identified instances of internal corrosion associated with out of scope components that would not be managed under the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program.

None of the internal operating experience reviewed were identified as consequential to plant operation.

The staff request a discussion with respect to the following: (a) whether corrosive chemicals have been used in other in-scope piping; and (b) why the internal corrosion associated with out-of-scope components is not representative of the condition of in-scope components.

2 Table 3.3.2-24 3.3-299 LRA Table 3.3.2-24, Fire Protection System, states gray cast iron with internal coating/lining exposed to raw water will be managed for loss of material using the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program. The AMR item cites Table 1 item 3.3.1-139.

Table 1 item 3.3.1-139 is referenced in thirteen Table 2 items, all of which cite the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program (except for in one instance as noted in the Background/Issue column). The staff requests a clarifying discussion with respect to why the Internal Surfaces in Miscellaneous Piping and Ducting Components program is cited in lieu of the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program for this AMR item.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 3

B.2.25 B-80 LRA Section B.2.25 states [l]icense renewal in-scope polymers have no identified aging effects requiring management.

GALL AMP XI.M38 states the program manages the effects of aging of polymer materials in all environments to which these materials are exposed.

LRA Table 3.3.1, item 3.3.1-119 (which is applicable to PVC),

states [t]he sight glass in the Fire Water System is not PVC but is a transparent polymer. The safety related instrument air tanks (condensate receiver / separator) are not specifically addressed.

Eight AMR line items cite plant-specific note 402 which states

[t]he materials are polyvinylidene fluoride (PVDF) and polyvinyl chloride (PVC).

The staff requests a discussion on the following based on its observations related to polymeric materials:

Are the safety related instrument air tanks (condensate receiver /

separator) constructed from PVC? These items are linked to item 3.3.1-119.

For the eight AMR line item that cite plant-specific note 402, it is unclear which ones are PVC and which ones are PVDF.

A discussion on the basis for why non-PVC polymers cite no aging effects requiring management (AERM). GALL Rev. 2 addresses PVC but provides minimal guidance on other polymeric materials.

GALL-SLR guidance does cite aging effects requiring management for polymeric materials exposed to air, treated water, and raw water.

Table 3.3.2-15e Table 3.3.2-24 3.3-255 3.3-293 3.3-295 LRA Tables 3.3.2-15e and 3.3.2-24 state stainless steel flexible connections, flexible hoses, and mufflers exposed to diesel exhaust will be managed for cracking using the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program.

The staff requests a discussion with the respect to the type of inspection that will be performed. AMP XI.M38, Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components contained in LR-ISG-2012-02 does not provide specific guidance.

However, the GALL-SLR version of AMP XI.M38 provides various

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions options such as surface examinations, VT-1 inspections, and detecting staining on the external surfaces of diesel exhaust piping.

LRA Table 3.5.2 Bulk Commodities (TRP 79)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

3.5 3.5-163 Row 316 of table 3.5.2-4 cites generic note E and states that the Structures Monitoring Program will manage the aging effects of change in material properties and cracking for tefzel ties made of elastomers. However, GALL recommends that for VII.F1.AP-102 the External Surfaces Monitoring of Mechanical Components program to be used. The basis for choosing the Structures Monitoring program is unclear to the staff.

Please provide the basis for choosing the Structures Monitoring Program as opposed to the External Surfaces Monitoring of Mechanical Components to manage the aging effects of change in material properties and cracking for tefzel ties.

LRA Section: B.2.44 - Water Chemistry Program (TRP 2)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

N/A N/A Page 1 of the Reactor Water Specification in REC-0104, Chemistry Specifications, lists the sampling frequency for reactor water conductivity as ((

)). According to the 2019 BWR Water Chemistry interim guidance, the sampling interval for conductivity in all modes should be (( )). The specifications for feedwater follow a similar pattern, except the sampling intervals are listed as (( )) Both parameters are ((

)). According to BWRVIP-190, ((

.)) So, the sampling interval Please clarify the sampling interval for reactor water and feedwater conductivity.

Please describe the basis for any deviations from the EPRI water chemistry guidance for these parameters.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions is unclear to the staff, and it is unclear whether any deviations to the guidance have been documented for these parameters.

2 N/A N/A On page 1 of the Feedwater Specifications in REC-0104 it lists the sampling frequency for reactor water dissolved oxygen as (( )).

According to the 2019 BWR Water Chemistry interim guidance, the sampling interval for dissolved oxygen in modes 1-3 should be (( )). This parameter is a

((

)). According to BWRVIP-190, ((

)) So, the sampling interval is unclear to the staff, and it is unclear whether there are any deviations to the guidance for these parameters.

Please clarify the sampling interval for feedwater dissolved oxygen.

Please describe the basis for any deviations from the EPRI water chemistry guidance for this parameter.

3 N/A N/A The 2019 BWR Water Chemistry Interim guidance includes sampling frequency for electrochemical potential (ECP) for reactor water at power operation. However, REC-0104 does not appear to address sampling frequency for ECP.

Please clarify the measurement frequency for reactor water ECP.

4 N/A N/A For chlorides and sulfates for reactor water at hot standby REC-0104 states that the sampling frequency is (( )).

However, the 2019 Water Chemistry interim guidance states that the sampling frequency should be every (( )). For chlorides and sulfates for reactor water at hot shutdown REC-0104 states that the sampling interval is daily. However, the 2019 Water Chemistry Interim Guidance states that,

((

))

Please describe the basis for the apparent deviations in sampling frequency for reactor water chlorides and sulfates at Hot Standby and Hot Shutdown conditions.

5 N/A N/A PYPM-OC-0001, Strategic Chemistry Plan, states that there is no analyzer for dissolved hydrogen monitoring for feedwater. It states that the hydrogen injection rate is logged daily and a hydrogen benchmark test which developed a correlation between injection rate and hydrogen concentration. It also states that grab sample analysis serves Please describe how PNPPs feedwater hydrogen sampling frequency meets the sampling frequency outlined in the 2019 BWR Water Chemistry Interim Guidance.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions as a check for concentration. REC-0104 also states that feedwater hydrogen is sampled quarterly. However, the 2019 Water Chemistry Interim guidance states that feedwater dissolved hydrogen should be measured ((.))

6 N/A N/A CR-2018-05909 and CR-2021-02014 describe instances of OLNC platinum precipitation. The 2018 CR states that OLNC injection ports were clogged with platinum. CR-2021-02014 also indicates that there was a blockage due to platinum buildup. However, it is unclear to the staff how the consequences of these blockages on the system are assessed.

Please discuss the following about these events:

a. Given that platinum precipitated near the injection site, how was the effect of not delivering the platinum to the intended locations evaluated?

b. What conclusions were drawn about the effect of these precipitation events on aging effects (e.g., cracking of the components meant to be protected by OLNC additions)?

c. What changes, if any, were made to the program to prevent recurrence?

7 3.5 3.5-131 In row 56 of table 3.5.2-4 the LRA states that cracking in aluminum component and piping supports will be managed by Water Chemistry and one-time inspection and cites generic note E. The table also cites table 1 item 3.3.1,136 from LR-ISG 2012- 02 and GALL item number VII.G.A-412. However, item 3.3.1,136 from LR-ISG 2012-02 refers to Steel, stainless steel, or aluminum fire water storage tanks exposed to air-indoor uncontrolled, air-outdoor, condensation, moist air, raw water, treated water, and GALL item VII.G.A-412 refers to, Fire water storage tanks.

Given that the GALL item cited is specifically for tanks to be managed by the Fire Water System Program, please discuss the basis of selection of this item for components and piping supports to be managed by the Water Chemistry and One-Time Inspection programs.

8 3.5 3.5-86 Row 140 in table 3.5.2-1 of the LRA refers to stainless steel Upper containment pool gate steel exposed to treated water. It states that Water Chemistry and One-Time inspection will be used to manage the aging effect of loss of material and cites generic note E. However, the configuration of gate steel and Please clarify the following:

a. The specific component that the Upper containment pool gate steel refers to.

b. The configuration of the

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions its exposure to treated water is unclear to the staff.

components exposure to treated water.

c. How Water Chemistry and One-Time inspection will be used to manage loss of material.

9 3.5 3.5-87 Row 141 in table 3.5.2-1 of the LRA refers to Upper containment pool gates that are composed of aluminum exposed to treated water. It states that Water Chemistry and One-Time inspection will be used to manage the aging effect of cracking and cites generic note H. However, the design of this component and the nature of its exposure to treated water is unclear to the staff.

Please clarify the following:

a. The design of the Upper containment pool gates.

b. The configuration of the components exposure to treated water.

c. How Water Chemistry and One-Time inspection will be used to manage cracking.

LRA Section: 4.1.1 - Identification of TLAAs (TRP 58)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.1 The following are TLAAs - included in LRA Section 4.6 4.6.4 Fatigue Analysis - Earthquake Cyclic Loading 4.6.11 - Allowable Stress Analysis of BOP ASME Code Class 1, 2 and 3 Components 4.6.12 - RPV Annealing Based on the information in the LRA - it is not clear how these analyses meet the six criteria of a TLAA as defined in 10 CFR 54.3 It appears these topics are covered directly in other TLAAs in the LRA (e.g., LRA Section 4.2 - RPV embrittlement and LRA Section 4.3 - Metal Fatigue) - and are not standalone TLAAs.

Provide a discussion of how each criterion for TLAA in 10 CFR 54.3 is satisfied for each of these topics/analyses.

Discuss whether these are analyses are addressed by other sections in the LRA.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 2

4.1 LRA Section 4.1 and LRPY-TLAA-001 indicate that CLB documents and Design Basis documentation was searched to identify potential TLAAs. The documents that were searched were provided in the referenced documents.

It is not clear how the applicant performed this search of the documents and the licensees disposition of analyses that were identified in this search.

For example - were there certain targeted keywords used by the applicant to perform an informed search or was brute force used by reviewing every single page of the CLB documents and Design Basis documentation.

Additionally, LRA Table 4.1-1 indicates with regard to the applicability to Perry for several potential TLAAs that No (No TLAA Identified) - Given the lack of documentation on how CLB documents were searched/reviewed it is not clear based on its review during the audit of LRA Section 4.1 and LRPY-TLAA-001 how the applicant determined that these potential TLAAs are not applicable to Perry.

Provide a discussion of how the CLB documents and Design Basis documentation was searched.

If it was with the use of key words -

specify the terms that were used and discuss how they were selected to provide assurance that all TLAAs were identified Provide documentation of the assessment/review for results that were identified via the search of the CLB documents and Design Basis documentation that were not TLAAs.

In particular, which criterion/criteria for a TLAA was not met for potential TLAAs.

Provide the documentation/assessment of how the applicant determined these potential TLAAs identified in NRC guidance were not TLAAs (e.g., no analysis found, analysis exists but does not meet all six criteria).

In particular - since ISI plans/IST plans were not part of the documentation that was reviewed -

Explain how the applicant determined that Inservice flaw growth analyses that demonstrate structure stability for 40 years is not applicable to Perry.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 3

4.1 LRA Section 4.1 and LRPY-TLAA-001 indicate that CLB documents and Design Basis documentation was searched to identify potential TLAAs UFSAR Facility Operating License; EPRI BWRVIP documents (incorporated by reference in the CLB);

NRC Safety Evaluation Reports (SERs); and, 10 CFR 50.12 Exemption Requests.

The staff noted that the applicants list of documentation that was reviewed did not include calculations and design reports referenced in the USAR and FOL.

It is not clear whether the applicant reviewed the supporting calculations and design reports to determine whether there is a TLAA for the current licensing basis. If this review was not performed, it is not clear how the applicant has assurance that a thorough review to identify TLAAs in accordance with 10 CFR Part 54 has been done.

Discuss whether the calculations and design reports referenced in the USAR and FOL were reviewed as part of the process for identifying TLAAs for the LRA.

Provide the supporting documentation/report of this review that was performed for calculations and design reports referenced in the USAR and Facility Operating License If a review of the calculations and design reports referenced in the USAR and Facility Operating License was not performed - Discuss how the applicant has assurance a sufficient/thorough review has been completed to identify TLAAs in accordance with 10 CFR Part 54 LRA Section: 3.2.2.5 - Loss of Material due to General Corrosion and Fouling that Leads to Corrosion (TRP 76)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

3.2.2.2.5 Table 3.2.1 3.2-12 3.2-17 LRA Section 3.2.2.2.5, related to general corrosion and fouling of containment spray system nozzle and flow orifices, states that this item is not used because there are no steel spray system flow orifices or nozzles subject to aging management in an internal, uncontrolled air environment in engineered safety features (ESF) systems at PNPP. It also identifies the orifices and spray nozzles as stainless steel. The corresponding Table 1 and GALL items are 3.2.1-6 and V.D2.EP-113.

Please discuss the evaluation in LRA Section 3.2.2.2.5 and Table 1 Item 3.2.1-6 with respect to the considerations about fouling from periodic wetting and drying described in NUREG-2221 and SRP-SLR Section 3.2.2.2.3.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions The staff notes that the aging effect for this topic was clarified in the discussion of changes for subsequent license renewal in NUREG-2221, Table 2-1, page 2-191, Item V.D2.EP-113a/113b. The flow blockage concern is related to fouling from corrosion products generated in components upstream of the nozzles and orifices due to periodic wetting and drying, and it therefore can occur in stainless steel nozzles and orifices. As a result, SRP-SLR Section 3.2.2.2.3 has the following considerations for this aging effect:

(a) The applicant identifies those portions of the system that are normally dry but subject to periodic wetting; (b) plant-specific procedures exist to drain the normally dry portions that have been wetted during normal plant operation or inadvertently; (c) the plant-specific configuration of the drains and piping allow sufficient draining to empty the normally dry pipe; (d) plant-specific OE has not revealed loss of material or flow blockage due to fouling; and (e) a one-time inspection is conducted to verify that loss of material or flow blockage due to fouling has not occurred.

These considerations are based in part on operating experience described in NRC Information Notice 2013-06 and not specific to a subsequent period of extended operation.

LRA Section: 3.3 - Aging Management of Auxiliary Systems (TRP 92)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

3.3 3.3-134 AMR item 3.3.1-119 in Revision 2 of NUREG-1801 indicates there are no aging effects/mechanisms requiring management by an AMP for PVC piping, piping components, and piping elements exposed to air with borated water leakage, indoor uncontrolled air, condensation (internal), and waste water.

Please discuss the following:

What polymer materials the sight glass and tank (condensate receiver/separator) are made from.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions The Discussion of AMR item 3.3.1-119 in LRA Table 3.3.1 states, The sight glass in the Fire Water System is not PVC, but a transparent polymer.

LRA Table 3.3.2-24 cites AMR item 3.3.1-119 for polymer sight glasses with leakage boundary and pressure boundary intended functions exposed externally to indoor uncontrolled air and internally by raw water. The associated notes are standard note F, Material not in NUREG-1801 for this component, and plant-specific note 310 that states, Based on plant operating experience, there are no aging effects requiring management for the polymer Fire Protection water sight glasses, or Safety Related Instrument Air condensate tank. Polymers are not expected to experience aging effects unless exposed to elevated temperatures or radiation levels capable of attacking the specific chemical composition. The sight glass is not PVC but is a transparent polymer. These components are exposed to indoor air externally, and to condensation or fire water internally. These environments do not include elevated temperatures or radiation levels.

LRA Table 3.3.2-52 cites AMR item 3.3.1-119 for polymer tank (condensate receiver/separator) with a leakage boundary intended function exposed externally to indoor uncontrolled air and internally by condensation. The associated notes are standard note C, Component is different, but consistent with NUREG-1801 item for material, environment, and aging effect.

AMP is consistent with NUREG-1801 AMP, and plant-specific note 310 (see above).

However, it is unclear what the transparent polymer materials for the sight glass and tank (condensate receiver/separator) are. Also, it is unclear why standard note C was used for the tank (condensate receiver/separator). Even though the component is different, standard note F would be consistent with the treatment for the sight glass.

Why standard note C was used for the tank (condensate receiver/separator).

Please discuss if it was considered that Revision 2 of NUREG-1801appears to indicate that rigid polymers should be visually inspected, and AMR items for managing the effects of aging of polymeric piping and piping components were added to GALL-SLR.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions The staff notes that Revision 2 of NUREG-1801 identifies AMPs for managing the effects of aging for polymeric materials, such as External Surfaces Monitoring of Mechanical Components (XI.M36) and Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components. While there are no AMR items, the Acceptance Criteria program element for AMPs XI.M36 and XI.M38 state, For rigid polymers, surface changes affecting performance, such as erosion, cracking, crazing, checking, and chalking, are subject to further investigation.

In addition, the staff notes that Volume 1 of NUREG-2191 (GALL-SLR) includes AMR item 3.3.1-263 and for managing aging of polymeric piping and piping components exposed to air, condensation, raw water, raw water (potable), treated water, waste water, underground, concrete, and soil. LRA Section 2.1.3 indicates that NUREG-2191 was treated as operating experience that may be relevant.

2 3.4 3.4-76 AMR item 3.4.1-57 in Revision 2 of NUREG-1801 indicates there are no aging effects/mechanisms requiring management by an AMP for PVC piping, piping components, and piping elements exposed to air with borated water leakage, indoor uncontrolled air, and condensation (internal).

LRA Table 3.4.2-11 cites AMR item 3.4.1-57 for polymer piping, sight glass (body), strainer body, and valve body exposed externally to indoor uncontrolled air. The associated notes are standard note A (Consistent with NUREG-1801 item for component, material, environment, and aging effect. AMP is consistent with NUREG-1801 AMP) and plant-specific note 402 that states, Based on plant operating experience, there are no aging effects requiring management for the Service Water and Emergency Service Water Chlorination system polymer components in a treated water or air - indoor uncontrolled environment. The materials are polyvinylidene Please discuss the following:

Was the intent to cite AMR item 3.4.1-57 for the PDVF components?

Please discuss the use of standard notes F and G in these instances.

Like Breakout Question 1, please discuss if it was considered that Revision 2 of NUREG-1801 appears to indicate that rigid polymers should be visually inspected, and AMR items for managing the effects of aging of polymeric piping and piping components

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions fluoride (PVDF) and polyvinyl chloride (PVC), and the treated water environment is sodium hypochlorite at a concentration of 12% to 15%. This material is not expected to experience aging effects unless exposed to elevated temperatures or radiation levels capable of attacking the specific chemical composition.

The material in these environments is not expected to experience significant aging effects due to elevated temperatures or radiation levels.

The Discussion for AMR item 3.4.1-57 in LRA Table 3.4.1 states, in part, Consistent with NUREG-1801. Polyvinylidene fluoride (PVDF) is considered an equivalent material for this comparison for components in the Emergency Service Water Chlorination system. PVC piping, piping components and piping elements in the Steam and Power Conversion systems subject to aging management exposed to an external air-indoor, uncontrolled environment is aligned to this item.

LRA Table 3.4.2-11 includes polymer piping, sight glass (body), strainer body, and valve body exposed internally to treated water. However, no NUREG-1801 or Table 1 items are cited. The associated notes are standard notes F (Material not in NUREG-1801 for this component) and G (Environment not in NUREG-1801 for this component and material), and plant-specific note 402 (see above). Based on plant-specific note 402, the staff understands that these polymers are PVDF.

Therefore, AMR item 3.4.1-57 is not cited for PVDF in Table 3.4.2-11. It isnt clear if the intent was to cite AMR item 3.4.1-57 for the PVDF components. In addition, it is unclear why standard note F was cited for the polymer piping, sight glass (body), and valve body, but standard note G was cited for the polymer strainer body. Each of these components are polymer and the environment is treated water.

were added to GALL-SLR.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 3

3.5 3.5 -158 LRA Table 3.5.2-4 includes porcelain sanitary fixtures inside the control room exposed externally to indoor uncontrolled air and raw water. However, no NUREG-1801 or Table 1 items are cited. The associated notes are standard note H (Aging effect not in NUREG-1801 for this component, material, and environment combination) and plant-specific note 520 that states, Based on Industry OE, porcelain in waste water and air has no aging effect in waste water or air. PNPP internal OE supports this conclusion. However, as previously noted, raw water, not waste water, is cited.

Please discuss if the environments cited and plant-specific note 520 are correct.

LRA Section: 3.0 - Aging Management Review Results (TRP 98)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

3.2.2.2.3.2 3.2.2.2.6 3.3.2.2.3 3,3,2,2,5 3.4.2.2.2 3.4.2.2.3 Tables 3.2.1 3.3.1 3.4.1 3.2-11 3.2-13 3.3-66 3.3-67 3.4-12 3.4-13 Further evaluation sections 3.2.2.2.3.2,3.2.2.2.6, 3.3.2.2.3, 3.3.2.2.5, 3.4.2.2.2 and 3.4.2.2.3 are associated with Table 1 items 3.2.1-4, 3.2.1-7, 3.3.1-4, 3.3.1-6, 3.4.1-2, and 3.4.1-3.

These Table 1 items are identified in the LRA as not applicable based on no stainless steel components being exposed to outdoor air o based on the components being insulated and managed by another Table 1 item. This aging management approach is consistent with the GALL report; however, the guidance for the corresponding further evaluation sections in subsequent license renewal (SLR) was changed to account for circumstances in which the actual environment may differ from the nominal environment. For example, indoor components may be exposed to air recently introduced from outdoors. According to SLR guidance, the plant-specific operating experience is reviewed, and a one-time inspection is performed to confirm the absence of the effect (SRP-SLR and GALL-SLR and SRP-SLR Supplemental Staff Guidance, March 2016, ML16041A090, Item C.)

In addition, the SLR guidance includes nickel alloys in the The staff requests the following:

a. Please discuss how these further evaluations address the possibility that the air environment includes air recently introduced into buildings that may contain contaminants that could concentrate and cause loss of material (due to pitting and crevice corrosion) or stress corrosion cracking.

b. Please discuss the applicability and evaluation for loss of material due to pitting and crevice corrosion for any nickel alloys exposed to outdoor air in the engineered safety features systems, auxiliary systems, and steam and power conversion systems.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions corresponding further evaluation sections for loss of material due to pitting or crevice corrosion. The staff was not able to determine if the components addressed by these further evaluation sections were evaluated to determine if they are exposed to outdoor air or another source of moisture or contaminants that could cause loss of material or stress corrosion cracking.

2 Table 3.3.2-56 Table 3.3.1 Page 3.3-443 Page 3.3-135 Table 3.3.2-56, Standby Liquid Control System, Rows 29 and 31, notes that the external surface of the stainless steel Standby Liquid Control (SLC) System tank is exposed to uncontrolled indoor air. Based on GALL report item VII.J.AP-17 and Table 3.3-1 item 3.3.1-120, no aging effects requiring management are identified and no AMPs are proposed.

Although this aging management approach is consistent with the GALL report, there is operational experience with aging effects in a stainless steel SLC tank due to an unexpected source of contaminants (Licensee Event Report 254-2006-004, Through-wall Leak in Standby Liquid Control Tank Due to the Original Construction Use of Grout with Leachable Halogens, 12/11/2006, ML063530355.)

The staff was not able to determine if the PNPP tank design and location could potentially expose it to conditions more corrosive than the nominal environment of uncontrolled indoor air.

Please describe the materials in contact with the exterior of the SLC tank and how the tank exterior was evaluated for loss of material and stress corrosion cracking due to unexpected exposure conditions (e.g.,

contaminants, moisture, temperature).

3 Table 1 items 3.1.1-107, 3.2.1-063, 3.3.1-120, 3.4.1-058 3.1-44, 3.2-40, 3.3-35, 3.4-43 Table 1 items 3.1.1-107, 3.2.1-063, 3.3.1-120, 3.4.1-058 are consistent with the GALL report in identifying no aging effects requiring management and no AMPs. However, it is not clear to that staff if the components aligned to these items for indoor air have been evaluated to determine if the air has recently been introduced from outdoors and may contain contaminants or moisture that could cause loss of material or stress corrosion cracking.

For example, item 3.4.1-58 states stainless steel piping, piping components and piping elements subject to aging Please discuss how the components aligned to these AMR items have been evaluated to determine if the air environment includes air recently introduced into buildings that may contain contaminants that could concentrate and cause loss of material or stress corrosion cracking.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions management in the Div. 1 & 2 Standby Diesel Generator Exhaust, Intake and Crankcase system and exposed to air-indoor, uncontrolled are aligned with this item. There are no aging effects requiring management and no AMPs associated with this item, which is consistent with the GALL report.

However, based on information from another application (e.g.,

River Bend RAI response ML18130A935), components in the diesel generator building may be exposed to air recently introduced into the building.

LRA Section: B.2.19 - Fatigue Monitoring Program (TRP 55)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

B.2.19 B-66 LRA Section B.2.19 addresses Enhancement 1 regarding the scope of the program program element of the Fatigue Monitoring Aging Management Program (AMP).

Enhancement 1 states that the station implementing procedures will be updated to clarify the scope of the Fatigue Monitoring Program. However, Enhancement 1 does not clearly describe what aspects of the program scope will be clarified in the implementing procedures.

Discuss what aspects of the program scope will be clarified in the implementing procedures when Enhancement 1 is implemented.

2 B.2.19 B-66 Regulatory Issue Summary (RIS) 2008-30, Fatigue Analysis of Nuclear Power Plant Components, addresses a potential concern regarding fatigue usage calculations that considers components response to a step change in temperature. The RIS indicates that the concern involves an input in which only one value of stress is used for the evaluation of the actual plant transients. The RIS indicates that the detailed stress analysis requires consideration of six stress components, as discussed in ASME Code,Section III, Subsection NB, Subarticle NB-3200.

In comparison, the operating experience section in LRA Section B.2.19 evaluates the operating experience related to Discuss the operating experience evaluation regarding the concern addressed in RIS 2008-30 to confirm that the Fatigue Monitoring AMP does not have such a concern.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions the Fatigue Monitoring aging management program (AMP).

However, the operating experience section does not include the evaluation regarding RIS 2008-30.

The staff needs to confirm that the Fatigue Monitoring AMP does not have the concern addressed in RIS 2008-30.

LRA Section: 4.3.1-Class 1 Fatigue (TRP 60.1)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.3.1 4.3.3 4.3-1 4.3-11 The Perry operating experience (OE) search results for the keyword environmental address the fatigue analysis of the HPCS (high pressure core spray) piping system that is described in CR-2023-03889.

The OE search results indicate that Calculation B13-027 describes the fatigue analysis for HPCS node 27 (weld between the HPCS piping and core spray nozzles safe end extension).

The OE search results also indicate that Calculation B13-027 does not address a fatigue analysis for the following HPCS locations: (1) safe end to safe end extension weld; and (2) safe end to core spray nozzle weld.

The staff needs clarification as to why the two weld locations above are not included in the fatigue analysis regarding cumulative usage factor (CUF) and environmentally adjusted CUF (CUFen).

1. Discuss why the two weld locations discussed in the background/issue section are not included in the fatigue analysis.

If HPCS node 27 is bounding for the two locations in terms of CUF and CUFen, discuss the technical basis for the bounding nature of node 27 in comparison with the two weld locations. As part of the response, clarify the material of fabrication of these two welds.

2 4.3.1 4.3-3 Section 5.2 of SIA 2001140.301, Rev. 3 (in Calculation B13-025, Rev. 1) indicates that the 60-year transient cycle projections and the associated cumulative usage factor (CUF) calculations use the long-term cycle accumulation rate and short-term cycle accumulation rate.

Specifically, the SIA report indicates that, except for the

1. Describe what component the STUDS_SUBSQNT location refers to in relation to the cycle projection and CUF calculation.

2. Describe the transients that are applicable to the cycle

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions STUDS_SUBSQNT location, the other locations use the weighting factors of 1 and 3 for the long-term cycle accumulation rate and short-term cycle accumulation rate (10-year period up to April 13, 2021), respectively.

In comparison, Section 5.2 of SIA 2001140.301, Rev. 3 indicates that, for the STUDS_SUBSQNT location, the long-term cycle accumulation rate is not used and that only the short-term cycle accumulation rate is used. The short-term cycle accumulation rate for the STUDS_SUBSQNT location is based on the cycles for the most recent 3 years up to April 13, 2021.

The staff needs clarification regarding the STUDS_SUBSQNT location and why only the most recent 3 years of cycle accumulation (up to April 14, 2021) is used the 60-year cycle projections.

In addition, the staff noted that the SIA report addresses that STUDS_PRIOR location that uses both long-term and short-term cycle accumulation rates. The staff needs to clarify the differences between the STUDS_SUBSQNT and STUDS_PRIOR locations in terms of the cycle projection approach.

The staff also noted that Table 5 of the SIA report indicates that the CUF value of the STUDS_SUBSQNT location increased from 0 to 0.43 for the time period from October 1, 2016 to April 13, 2021 (approximately 4.5 year period). The staff needs to clarify why this location involved a rapid increase in CUF and how the applicant will ensure that the CUF for this location does not exceed the design limit (1.0) for the period of extended operation.

projections and CUF calculation for the STUDS_SUBSQNT location.

3. Clarify why the 60-year cycle projection and CUF calculation for STUDS_SUBSQNT location use only the short-term cycle accumulation data (i.e.,

most recent 3 years of cycle data up to April 13, 2021). As part of the discussion, explain why the most recent 3 years of cycle data can represent the future cycle accumulation for the period of extended operation.

4. Discuss why the STUDS_PRIOR location uses both the long-term and short-term (10-year) cycle accumulation rates in the cycle projections, as opposed to the STUDS_SUBSQNT location.

As part of the discussion, compare the overall cycle accumulation rates for the future cycle projections between these two locations.

5. Explain why the STUDS_SUBSQNT location involved a rapid increase in CUF and how the applicant will ensure that the CUF for this location does not exceed the design limit (1.0)

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions for the period of extended operation.

LRA Section: 4.3.2 - Non-Class 1 Fatigue (TRP 60.2)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.3.2 4.3-12 LRA Section 4.3.2 addresses the fatigue time-limited aging analysis (TLAA) for the non-Class 1 piping systems (i.e.,

ASME Section III Class 2 and 3 and ANSI B31.1 piping systems).

The staff noted that the fatigue TLAA relies on the implicit fatigue analysis provisions in ASME Code,Section III, NC-3611 and ND-3611. These provisions allow no reduction in the allowable stress range for thermal expansion stresses if the number of equivalent full temperature cycles does not exceed 7000 cycles.

LRA Section 4.3.2 indicates that the number of cycles for each non-Class 1 piping system was considered to confirm that the number of 60-year cycles does not exceed 7000 cycles in the implicit fatigue analysis.

However, LRA Table 4.3.2 does not clearly describe how the 60-year cycles were determined (e.g., based on piping system design information, plant operation procedures, test requirements, UFSAR information and specific system-level knowledge).

Describe how the 60-year cycles were determined (e.g., based on piping system design information, plant operation procedures, test requirements, UFSAR information and specific system-level knowledge).

2 4.3.2 4.3-12 LRA Section 4.3.2 addresses the implicit fatigue analysis for the non-Class 1 piping systems. The provisions in ASME Code Section III, NC-3611 and ND-3611 allows that, if the number of equivalent full temperature cycles does not exceed 7000 cycles, no reduction is applied to the allowable stress range for thermal expansion stresses.

1. Describe the number of 60- year equivalent full temperature cycles for the non-Class 1 piping systems connected to the Class 1 piping systems. As part of the response, clarify the following: (1) whether all

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section 4.3.2 indicates that portions of some of the non-Class 1 piping systems, such as residual heat removal and high pressure core spray piping systems, were designed in accordance with ASME Section III, Class 2 or 3, or ANSI B31.1 requirements. The LRA section also explains that these piping systems are attached to ASME Section III, Class 1 piping and are affected by the same transients as the Class 1 systems. LRA Section 4.3.2 indicates that, in the implicit fatigue analysis, the number of 60- year cycles for the non-Class 1 piping systems connected to Class 1 piping systems is based on the transient cycles listed in LRA Table 4.3-1.

However, the LRA does not provide the number of projected 60-year equivalent full temperature cycles determined in the implicit fatigue analysis for the piping systems connected to Class 1 piping systems.

the transient cycles listed in LRA Table 4.3-1 are applicable to the estimation of the equivalent full temperature cycles in the implicit fatigue analysis; and (2) whether Transient 29, SRV [safety relief valve] Actuation (Acoustic Wave) is included in the cycle estimation for the implicit fatigue analysis.

3 4.3.2 4.3-12 LRA Section 4.3.2 addresses the implicit fatigue analysis for the non-Class 1 piping systems. The LRA section does not clearly describe the number of projected 60-year equivalent full temperature cycles determined in the implicit fatigue analysis for the following non-Class 1 piping systems and their technical basis: (1) control and computer room humidification; (2) reactor plant sampling; (3) fire protection; (4) auxiliary steam and drains; (5) hydrogen chemistry system; (6) post accident sampling; (7) Div. 1 & 2 standby diesel generator exhaust, intake, and crankcase; and (8) emergency diesel generator.

1. Describe the number of projected 60-year equivalent full temperature cycles for the following non-Class 1 piping systems and their technical basis: (1) control and computer room humidification; (2) reactor plant sampling; (3) fire protection; (4) auxiliary steam and drains; (5) hydrogen chemistry system; (6) post accident sampling; (7)

Div. 1 & 2 standby diesel generator exhaust, intake, and crankcase; and (8) emergency diesel generator.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: 4.3.3 - Environmental Fatigue (TRP 60.3)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.3.3 4.3-15 LRA Section 4.3.3 addresses the environmental fatigue time-limited aging analysis (TLAA) fatigue. Environmental fatigue is also called environmentally assisted fatigue (EAF).

The LRA section indicates that the applicant performed an EAF screening evaluation to determine the EAF locations that may be more limiting than the locations addressed in NUREG/CR-6260. The LRA section also explains the applicant used the guidance in NUREG/CR-6909, Revision 0 in the EAF screening evaluation.

The staff noted that Revision 1 of NUREG/CF-6909 provides the more recent guidance for determining the environmental effect on fatigue in comparison with Revision 0 of NUREG/CR-6909.

However, LRA Section 4.3.3 does not clearly discuss whether the applicant compared the EAF screening results in the LRA with those based on Revision 1 of NUREG/CR-6909 (i.e., the latest guidance for EAF analysis).

1. Clarify whether the applicant compared the EAF screening results in the LRA with those based on Revision 1 of NUREG/CR-6909 (i.e., the latest guidance for EAF analysis). If not, discuss why such comparison is not necessary.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 2

4.3.3 4.3-17 LRA Table 4.3-5 describes the limiting EAF locations, and their 60-year projected environmentally adjusted cumulative usage factor (CUFen) values. LRA Section 4.3.3 indicates that the EAF analysis includes the locations of NUREG/CR-6260 and other component locations that may be more limiting than the NUREG/CR-6260 locations in terms of EAF.

However, LRA Table 4.3-5 does not clearly describe which locations are the plant-specific EAF locations that may be more limiting than the NUREG/CR-6260 locations.

1. Clarify whether LRA Table 4.3-5 includes the plant-specific limiting EAF locations that may be more limiting than the NUREG/CR-6260 locations. If so, identify the plant-specific limiting locations. If not, provide justification for why LRA Table 4.3-5 does not describe plant-specific limiting EAF locations other than the NUREG/CR-6260 locations. If such justification cannot be provided, discuss a plan to revise LRA Table 4.3-5 to include the plant-specific limiting EAF locations and their 60-year projected CUFen values. their 60-year projected CUFen values.

3 4.3.3 4.3-15 4.3-16 LRA Section 4.3.3 addresses the EAF TLAA, including the EAF screening evaluation to determine the limiting EAF locations.

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

1. Describe the following items regarding the EAF screening evaluation: (1) whether the screening evaluation was performed for each material type; (2) how the applicant determined thermal zones or sections that group certain components and piping lines for proper comparisons of the screening CUFen values considering the applicable transient conditions; and (3) how the applicant compared the highest values of the screening CUFen values to determine the final limiting locations (e.g., screening process when the highest CUFen values are close to each other in a thermal zone).

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 4

4.3.3 4.3-16 LRA Section 4.3.3 addresses the EAF TLAA fatigue, including the EAF screening evaluation to determine the limiting EAF locations.

LRA Section 4.3.3 indicates that, after the screening evaluation, the applicant performed more detailed EAF evaluation to remove some conservatisms and determine the refined CUFen values for 60 years of operation, as described in LRA Table 4.3-5. However, the LRA does not clearly describe how the conservatisms associated with the screening CUFen values were removed to refine the 60- year projected CUFen values.

1. Describe how the applicant performed the detailed evaluation to determine the refined CUFen values listed in LRA Table 4.3-5. As part of the discussion, explain the following:

(a) how the applicant removed the conservatisms associated with the screening CUFen values; (b) how the applicant determined the strain rates for the CUFen calculations in the screening evaluation and detailed evaluation respectively; and (c) how the applicant determined the component temperatures for the CUFen calculations in the screening evaluation and detailed evaluation respectively.

5 Section 4.3.3 Table 4.3-5 4.3-16 Table 125 of SIA 1300310.303, Rev. 1 (in Calculation B13-039, Rev. 0) indicates that the 60-year screening CUFen for the feedwater nozzle (stainless steel location) is 1.06. In comparison, LRA Table 4.3-5 indicates that the refined 60-year CUFen for the component location is 0.951. The staff needs clarification on how the applicant refined the screening CUFen value (1.06) to determine the 60-year CUFen value (0.951) in LRA Table 4.3-5.

1. Discuss how the applicant refined the screening CUFen value (1.06) for the feedwater nozzle stainless steel location to determine the 60-year CUFen value (0.951) in LRA Table 4.3-5. As part of the response, clarify the specific location that the feedwater nozzle stainless steel location refers to (e.g., information on the specific weld location of the feedwater nozzle).

6 Section 4.3.3 Table 4.3-5 4.3-16 LRA Table 4.3-5 identifies the feedwater nozzle stainless steel location as a limiting EAF location for the feedwater nozzle and piping system. The table does not identify a limiting location for the feedwater Class 1 piping. In contrast, NUREG/CR-6260 identifies both the reactor vessel feedwater nozzle and feedwater Class 1 piping as the EAF

1. Clarify the following: (1) whether the feedwater nozzle body is fabricated with carbon steel or low alloy steel; and (2) whether the feedwater nozzle stainless steel location is bounding for the

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions locations that should be included in the EAF analysis as limiting locations for newer vintage boiling water reactors.

In addition, Table 125 of SIA 1300310.303, Rev. 1 (in Calculation B13- 039, Rev. 0) indicates that the feedwater nozzle stainless steel location is bounding for the feedwater piping location (piping point 110, carbon steel). However, these locations are made of different materials and the 60-year screening CUFen values are very close to each other (1.06 for the nozzle stainless steel location versus 0.99 for the carbon steel piping location). The staff needs information supporting that the feedwater nozzle stainless steel location can bound the feedwater piping carbon steel location even though the fabrication materials are different, the screening CUFen values are almost identical (1.06 versus 0.99), the screening CUF for the feedwater piping location (0.409) is greater than that for the feedwater nozzle stainless steel location (0.337), and NUREG/CR-6260 lists both the feedwater nozzle location and feedwater piping location as limiting locations that should be included in the EAF analysis. The staff also needs clarification on whether the feedwater nozzle stainless steel location (listed in LRA Table 4.3-5) is bounding for the feedwater nozzle body steel location in terms of CUF and CUFen feedwater nozzle body (steel) location in terms of the CUF and CUFen.

Explain why the feedwater nozzle stainless steel location is adequate to bound the feedwater piping location (carbon steel) even though the fabrication materials are different, the screening CUFen values are almost identical (1.06 versus 0.99), the screening CUF for the feedwater piping carbon steel location (0.409) is greater than that for the feedwater nozzle stainless steel location (0.337),

and NUREG/CR-6260 lists both the feedwater nozzle location and feedwater piping location as limiting locations that should be included in the EAF analysis.

LRA Section: 4.3.4 - Reactor Vessel Internals Fatigue (TRP 60.4)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.3.4 4.3-18 LRA Section 4.3.4 addresses the fatigue time-limited aging analysis (TLAA) for the reactor vessel internals.

During the audit, the staff noted that the following reference addresses the operating experience related to the jet pumps of the Perry Nuclear Power Plant Unit 1 and the associated fatigue evaluation (

Reference:

CR-2015-07022, 1R15

1. Describe the projected 40- year and 60-year CUF values for the Unit 1 jet pumps 13 and 14. As part of the discussion, describe the transients and transient cycles that are used in the CUF analysis.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Condition Monitoring Exam Results for Jet Pump 13 & 14 Problems Found in 1R12, May 15, 2015). The condition report explains that the vessel side and shroud side set screws prevent the vibration of the jet pump, thereby, avoiding the accumulation of fatigue cumulative usage factor (CUF) for the jet pump.

The condition report also indicates that, based on the screw gaps exceeding a certain criterion, the applicant evaluated the CUF value of Unit 1 jet pumps 13 and 14. Specifically, the condition report indicated that the CUF value for these jet pumps is 0.56 as of the refueling outage 12 of Unit 1 (Spring 2009). However, LRA Section 4.3.4 does not discuss the fatigue analysis for these jet pumps.

2. If jet pumps other than Unit 1 jet pumps 13 and 14 are subject to a fatigue analysis due to the set screw gap exceeding an acceptance criterion, describe the projected 40-year and 60- year CUF values for those jet pumps.

3. Clarify the disposition of the fatigue TLAA for the jet pumps discussed above. As part of the discussion, clarify whether the Fatigue Monitoring AMP monitors the relevant transient cycles to ensure the CUF values for the jet pumps do not exceed the design limit (1.0).

4. Discuss how the applicants inspection activities and the associated aging management programs will monitor the set screw gaps of the jet pumps to minimize the gaps and the effects of fatigue on the structural integrity of the jet pumps.

LRA Section: 4.3.5 - Intermediate High Energy Line Break Location Determination (TRP 60.5)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.3.5 4.3-19 LRA Section 4.3.5 addresses the TLAA on HELB location postulation for ASME Code Section III, Class 1 piping systems.

The section indicates that the TLAA is based on the approach for HELB location postulation in UFSAR Section 3.6.2.1.5.

The staff noted that in addition to the HELB location

1. Describe justification for why LRA Sections 4.3.5 does not identify the HELB analysis as a TLAA for Class 2 and 3 piping systems even though the screening criteria of the HELB

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions postulation for the Class 1 piping systems, UFSAR Section 3.6.2.1.5 addresses the HELB location postulation for Class 2 and 3 piping systems. Specifically, Item a.3 of the UFSAR section explains that the HELB breaks for the Class 2 and 3 piping are assumed in accordance with the criterion based on the allowable stress range for expansion stress (SA),

consistent with the guidance in Branch Technical Position MEB 3-1 (ADAMS Accesso No. ML052340555). SA may need to be adjusted by a stress range reduction factor based on the number of transient cycles. However, LRA Section 4.3.5 does not identify the HELB analysis for Class 2 and 3 piping as a TLAA based on the HELB location postulation that involves SA and the associated cycle-dependent stress range reduction factor.

location postulation involve the dependency on the transient cycles. If justification cannot be provided, identify the HELB analysis as a TLAA for the Class 2 and 3 piping systems (in a consistent manner with the identification of non-Class 1 fatigue TLAA in LRA Section 4.3.2) and discuss the disposition of the TLAA. In addition, clarify whether the UFSAR supplement (LRA Section A.2.3.5) needs to be revised accordingly.

LRA Section: 4.6.6 - Fatigue due to Single Recirculation Loop Operations (TRP 116.6)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.6.6 4.6-6 LRA Section 4.6.6 addresses the fatigue TLAA due to the vibratory loading of the single recirculation loop operation (also called single loop operation or SLO) for the reactor vessel internals.

The LRA section indicates that the applicants evaluation showed that all reactor vessel internals, except the in-core guide tubes, had adequate fatigue life without detailed analysis.

However, the LRA does not clearly discuss how the applicant determined that all reactor vessel internals, except the in-core guide tubes, had adequate fatigue life without detailed analysis for the fatigue due to the vibratory loading of the SLO.

Describe how the applicant determined that all reactor vessel internals, except the in-core guide tubes, have adequate fatigue life.

As part of the discussion, clarify whether the vibratory loading of the SLO does not cause stress exceeding the fatigue endurance limit for the reactor vessel internals other than the in-core guide tubes.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: 4.6.11 - Allowable Stress Analysis of BOP ASME Code Class 1, 2 and 3 Components (TRP 116.11)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.6.11 A.2.6.11 4.6-11 A-59 LRA Section 4.6.6 addresses the allowable stress TLAA for the balance-of-plant (BOP) ASME Code Class 1, 2 and 3 components. However, the LRA does not clearly discuss what aspects of the allowable stresses are time-dependent in the TLAA. The staff also needs clarification on what provisions of ASME Code,Section III are used in the determination of the allowable stresses in the TLAA.

In addition, the title of the TLAA section indicates that the BOP components are evaluated in the TLAA. However, the BOP components are typically Class 2 or 3 components. Therefore, the staff needs clarification on whether ASME Code Class 1 components are included in the TLAA discussed in LRA Section 4.6.6.

1. Describe what aspects of the allowable stresses are time-dependent in the allowable stress TLAA.

2. Describe the specific ASME Code provisions that are used in the determination of the allowable stresses in relation to the allowable stress TLAA. As part of changes to certain time-dependent parameters. If so, clarify the time-dependent parameters and why the changes to the allowable stresses do not affect the existing stress analyses.

3. Clarify whether the allowable stress TLAA is applicable to the ASME Code Class 1 components even though the BOP components are typically ASME Code Class 2 or 3 components and the applicable ASME Code provisions related to the allowable stresses may be different among Class 1, 2 and 3 components (e.g., the effect of fatigue cycles for Class 1 components are mainly addressed by cumulative usage

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions factors rather than adjustments to allowable stresses that are described in ASME Code Section III, NC/ND-3611).

4. Describe which BOP Class 1 components are subject to the discussion and clarify whether the allowable stresses have changed due to the to the allowable stress TLAA.

5. Discuss whether the UFSAR supplement in LRA Section A.2.6.6 needs to be revised based on the discussions above.

LRA Sections: 2.3.1.3 - Reactor Coolant Pressure Boundary (RCPB), 2.3.3.24 - Fire Protection and 2.4.4 - Structural Bulk Commodity (TRP 27)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

Section 2.3.1.3 2.3-7 Table 2.3.1-3 of the LRA does not include the following reactor coolant oil collection system in the scope of license renewal and subject to an aging management review (AMR):

flame arrestor reactor coolant pump oil collection enclosure reactor coolant pump oil collection tank Verify whether the reactor coolant pump oil collection system components listed are within the scope of LRA in accordance with 10 CFR 54.4(a) and whether they are subject to an AMR in accordance with 10 CFR 54.21(a)(1). If any of the listed components are not within the scope of LRA and are not subject to an AMR, the staff requests that the applicant provide justification for the exclusion.

2 Section 2.3.3.24 2.3-101 Table 2.3.3-24 of the LRA does not include the following fire protection components in the scope of license renewal and subject to an aging management review (AMR):

standpipe and interior hose stations Verify whether the fire protection components listed are within the scope of LRA in accordance with 10 CFR 54.4(a) and whether they are

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions hose connections and racks filter housing odorizer hanger and piping support seismic support for standpipes system piping diesel fire pump exhaust silencer traveling screens floor drains and curbs for fire-fighting water dikes for oil spill confinement subject to an AMR in accordance with 10 CFR 54.21(a)(1). If any of the listed components are not within the scope of LRA and are not subject to an AMR, the staff requests that the applicant provide justification for the exclusion.

3 Section 2.4.4 2.4-63 LRA Table 2.4.4-1 of the LRA does not include the following fire barriers:

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

LRA Section: 3.5 - Aging Management of Containments, Structures, and Component Supports (TRP 100)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

3.5 3.5-91 AMR item 3.5.1-95 in Revision 2 of NUREG-1801 indicates there are no aging effects/mechanisms requiring management by an AMP for aluminum, galvanized steel, and stainless steel support members; welds; bolted connections; support anchorage to building structure exposed to indoor uncontrolled air.

LRA Table 3.5.2-2 cites AMR item 3.5.1-95 for managing loss of material of the galvanized steel diesel fuel tank maintenance structures by the Structures Monitoring program. Standard note Please discuss the use of standard note A when the environment, aging effect, and AMP are different.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions A is cited (Consistent with NUREG-1801 item for component, material, environment, and aging effect. AMP is consistent with NUREG-1801 AMP). However, the environment, aging effect, and AMP are not consistent, therefore, the use of standard note A is unclear.

2 AMR item 3.3.1-121 in Revision 2 of NUREG-1801 indicates there are no aging effects/mechanisms requiring management by an AMP for steel piping, piping components, and piping elements exposed to indoor controlled air, dry air, and gas.

LRA Table 3.5.2-1 cites AMR item 3.3.1-121 steel containment vessel electrical penetrations exposed internally to gas. Given that the component is different, the use of standard note A is unclear.

Please discuss the use of standard note A given that the component is different.

LRA Sections: B.2.39 - Reactor Head Closure Stud Bolting Program (TRP 3)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

B.2.39 B-111 LRA Section B.2.39 states, The Refueling Outage 18 (RFO18)

ISI Summary Report identified no unacceptable reactor head closure studs, threads in flange, nuts, or washers. RFO18 Tables 1 and 2 show the components in the scope of the AMP for the current ISI interval were not examined.

Provide a discussion on the applicability of the statement from the LRA as it relates to the RFO18 Report.

2 B.2.39 B-109 The AMP description in LRA Section B.2.39 indicates that the components in scope of the Reactor Head Stud Bolting Program are the reactor head closure studs, nuts, flange threads, and washers.

Confirm that the Perry RPV closure stud holes (i.e., the flange threads) do not have the bushing design.

LRA Section: 4.2.6 - RPV Reflood Thermal Shock (TRP 59.6)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.2.6 4.2-19 The analysis of stresses applied during the analyzed transient are given at 0.052t for the recirculation line break. Typically, this analysis is given for a 0.25t flaw as the more limiting flaw depth, Please provide an explanation of how a 0.052t flaw depth is adequate for the analysis.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions with the application stating that the maximum KIapplied being at the 0.25t location.

2 4.2.6 4.2-19 It is claimed that the LPCI N6 nozzles do not meet the 1.0E+17 n/cm2 threshold for fluence before 54 EFPY. Fluence information is not provided for this nozzle in the application.

Please provide fluence projection data for the LPCI N6 nozzles that shows these nozzles do not exceed the 1.0E+17 n/cm2 threshold before 54 EPFY.

3 4.2.6 4.2-19 It is stated that even though the water level instrument nozzle (WLIN) N12 nozzles exceed the 1.0E+17 n/cm2 fluence threshold, they are not bounding for the present evaluation. The staff reviewed Section 3.2 of SIA calculation 1300341.305, Revision 0, which states that these WLIN N12 nozzles are non-ferritic materials.

What specific materials are the WLIN N12 fabricated from? This information would need to be docketed (via an RCI).

4 4.2.6 4.2-17 The TLAA description cites the following generic references for the subject analyses of the TLAA:

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, Paper G1/5 General Electric Report No. NEDO-10029, An Analytical Study on Brittle Fracture of GE-BWR Vessels Subject to the Design Basis Accident, L.C. Hsu, June 1969.

The staff was unable to find references to these two documents in the UFSAR (Chapters 4 and 5).

Show where these generic references are cited either in a) the UFSAR; or b) other current licensing basis documents (i.e., document(s) that show that these generic analyses are the analyses of record for the 40-year operating period).

LRA Section: 4.2.2 - Upper Shell Energy (TRP 59.2)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

BWRVIP-135, Rev. 4 provides surveillance data for Heat No C2557-1 and 5P6214B.

Explain the methodology used to determine the USE for Heat No C2557-1 and 5P6214B.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: 4.2.3 - Adjusted Reference Temperature (TRP 59.3)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

LRA Section 4.2.3 LRA Table 4.2-3 only addresses ART values for 1/4 T location.

PT limits are currently contained in Tech Spec Section 3.4.11.

Appendix G to 10 CFR Part 50 requires the pressure-temperature limits identified as "ASME Appendix G limits" in table 1 require that the limits must be at least as conservative as limits obtained by following the methods of analysis and the margins of safety of Appendix G of Section XI of the ASME Code.

G-2120 MAXIMUM POSTULATED DEFECT of Section XI provides the requirements for the postulated flaw - this section indicates in part that the postulated defects have a depth of onefourth of the section thickness and a length of 11/2 times the section thickness and defects are postulated at both the inside and outside surfaces.

Discuss whether the ART values for the 3/4T location for this TLAA should be addressed in the LRA?

Discuss whether the assessment of ART for 3/4T would be addressed when P-T limits are updated consistent with Appendix G (i.e.,

discusses 1/4 T flaw from both inside and outside surface).

2 LRA Table 4.2-3 provides material property values/information for PNPP Beltline RPV Materials.

Heat No 5P6214B is listed for the following RPV materials:

Upper-Intermediate Shell Axial Weld (BG, BJ, BK)

Upper-Intermediate Shell Axial Weld (BD and BF)

Lower Intermediate Shell Axial Weld (BE)

Lower Intermediate Shell Axial Weld (BA, BB, BC)

Cu% and Ni% for these RPV materials are consistent with each other except for Lower Intermediate Shell Axial Weld (BE).

Explain the inconsistency in Cu% and Ni% for Heat No 5P6214B for the different RPV materials within LRA Table 4.2-3.

Explain the inconsistency in Cu%

and Ni% for Heat No 5P6214B for the different RPV materials in LRA Table 4.2-3 when compared to the FSAR.

Explain the inconsistency between the Cu% information between LRA Table 4.2-2 and LRA Table 4.2-3? Is

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Cu% and Ni% in LRA Table 4.2-3 is inconsistent with information in FSAR table 5.3-1 and 5.3-3 for Heat No 5P6214B.

Cu% and Ni% for Lower Intermediate Shell Axial Weld (BE) appears to be consistent with information from BWRVIP-86, Rev 1-A. FSAR Table 5.3-2 indicates two values for Initial RTndt for Heat No 5P6214B.

Cu% for these RPV materials in LRA Table 4.2-3 are inconsistent with information in LRA table 4.2-2, except for the Lower Intermediate Shell Axial Weld (BE) with Heat No 5P6214B.

it just because of rounding? if so -

why not round consistently in all the LRA tables?

Confirm whether the basis for Cu%

and Ni% for Lower Intermediate Shell Axial Weld (BE) (Heat No 5P6214B is from BWRVIP-86, Rev 1-A. If not, what is the basis - and provide the associated source/reference documents.

Provide the basis for the Cu% and Ni% for RPV materials with Heat No 5P6214B, except for Lower Intermediate Shell Axial Weld (BE) -

Is it also BWRVIP-86, Rev 1-A with rounding?

Explain the two values for Initial RTndt for Heat No 5P6214B in FSAR Table 5.3-2? Is there any impact for the second initial RTndt in the LRA?

Does the LRA need to be reconciled to have consistent information for the RPV materials with Heat No 5P6214B.

3 CALCULATION B13-030 REV 0.PDF SIA Calc 1300341.301 - page 10 and 11 of 16:

Foot note 3 discusses the basis for the Cu content for C2432-2 and C2453-1. There is reference to data from Shell #1 and Shell#2 Confirm that Shell #1 represents Code No 21-1-1, 21-1-2, and 21-1-3 Confirm that Shell #2 represents Code No 22-1-1, 22-1-2, and 22-1-3

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Footnote 3 appears to apply to C2432 however the Cu content is different than reported for C2432-2 and C2453-1.

Additionally, the Cu and Ni for C2432-1 appears to be from the CMTR - see page 125 of 280 in DI-EA-0272 REV 0.PDF. Thus, its not clear if footnote 3 is relevant for C2432-1 CMTR for C2432 see page 125 of 280 in DI-EA-0272 REV 0.PDF - indicates material is SA-533 Gr.B Class 1 - which is consistent with for C2432-2 and C2453-1 The staff understands that this information is also contained in letter dated June 4, 2002 (PY-CEI/NRR-2627L) (ML021650244)

- see page 151 of 167).

If the above info for Shell #1 and #2 is correct - provide the basis that information for C2432-1 was not used to determine the Cu for C2432-2 and C2453-1 Discuss the relevancy of footnote 3 to C2432-1 in SIA Calc 1300341.301

- page 10 and 11 of 16.

Confirm the Cu and Ni reported for C2432-1 are based on the CMTR (i.e., see page 125 of 280 in DI-EA-0272 REV 0.PDF)

Discuss in more detail the specific material property information that was used to determine the Cu content for Heat No. C2432-2 and Heat No. C2453-1.

4 LRA Table 4.2 PNPP Beltline RPV Material ART Data for 54 EFPY:

Lower Int./Lower Int. Shell Girth Weld - Heat No:

4P7216 (single wire)

Initial RTndt - Cannot read off CMTR on Eportal -

Appears to be on page 149 of 280 in DI-EA-0272 REV 0.PDF. (maybe on the bottom right corner??)

4P7216 (tandem wire)

Initial RTndt - Cannot read off CMTR on Eportal -

Appears to be on page 150 of 280 in DI-EA-0272 REV 0.PDF. (maybe on the bottom right corner??)

If clearer copies of the CMTRs are available - please provide them -

Otherwise -the staff understands that this information is also contained in letter dated June 4, 2002 (PY-CEI/NRR-2627L) (ML021650244) -

see page 49 of 167).

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 5

LRA Table 4.2 PNPP Beltline RPV Material ART Data for 54 EFPY, indicates the following for Lower-Intermediate Shell Plate

- Code No. 22-1-1; Heat No. C2557-1:

Cu% = 0.05 Ni% = 0.63 FSAR Table 5.3 indicates that for Heat No. C2557-1:

Cu% = 0.06 Ni% = 0.61 The basis for the discrepancy in Cu% and Ni% between FSAR and LRA is not clear. It appears that the Cu and Ni for Heat No.

C2557-1 is based on BWRVIP-86, Rev 1-A Confirm that this information in LRA Table 4.2-3 for C2557-1 is from BWRVIP-86 Rev 1-A - Otherwise, provide the basis for the discrepancy?

6 Calc DI-EA-0272 REV 0.PDF on ePortal - The following information was provided from CMTRs in this document C2432-2; Chemistry information - (page 119 of 280 in DI-EA-0272 REV 0.PDF)

C2453-1; Chemistry information - (page 114 of 280 in DI-EA-0272 REV 0.PDF)

C2448-1; Chemistry information - (page 98 of 280 in DI-EA-0272 REV 0.PDF)

C2448-2; Chemistry information - (page 103 of 280 in DI-EA-0272 REV 0.PDF)

A1068-1; Chemistry information - (page 108 of 280 in DI-EA-0272 REV 0.PDF)

On the listed pages of the CMTRs for each of these RPV materials - there are two separate rows that provide the chemical composition (i.e., heat No. and Check)

For these RPV materials - please discuss the meaning for each of these separate rows identified in the CMTRs (i.e., heat No. and Check).

If the lesser value for Cu% and Ni%

values between these two rows was reported in the LRA for these RPV materials - provide a basis that the larger Cu% and Ni% is not appropriate?

Or some means of using all of the available information in determining the copper and nickel content.

Or use of the check measurement if these values represent the sample taken from the finished (solid) product.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Depending on the RPV material, the values for Cu and Ni were the same, higher, or lower between these two rows. The staff noted the chemistry information reported in the LRA was from the first row (i.e., non-check row)

The staff believes that the first row represents the sample taken from a molten sample in the ladle just prior to producing the product, and the second row the sample taken from the finished (solid) product.

Additionally, the staff noted that for the following RPV materials:

Heat No. C2557 Heat No. B6270 Heat No. A1155 The Cu and Ni content values reported in FSAR TABLE 5.3-1 are consistent with the check line from the respective CMTRs loaded on the ePortal.

7 Following approval for License Renewal, if applicable, is there a plan to reconcile the FSAR information for RPV materials?

Example - extended beltline materials New/updated Initial values (USE and RT NDT)

Copper and Nickel content information etcetera 8

Calc DI-EA-0272 REV 0.PDF on ePortal - The following information was provided from CMTRs in this document Heat No.5P621B (single wire); Chemistry information - (page 138 of 280 in DI-EA-0272 REV 0.PDF)

What is the rationale that Heat No.5P621B treated differently than Heat No 4P7216 in the LRA with respect to identifying the specific material properties information for single wire and tandem wire?

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Heat No.5P621B (tandem wire); Chemistry information - (page 141 of 280 in DI-EA-0272 REV 0.PDF)

It appears the limiting values for initial RTndt and USE between the Heat No.5P621B (single wire) and Heat No.5P621B (tandem wire) was used in the development of LRA Tables 4.2-2 and 4.2-3. (See question 2 above related to Cu and Ni content)

LRA Tables 4.2-2 and 4.2-3 addresses a similar situation for Heat No 4P7216 (single wire and tandem wire).

Explain how the material information for Heat No.5P621B (single wire) and Heat No.5P621B (tandem wire) was used to develop/populate the information in LRA Tables 4.2-2 and 4.2-3.

LRA Section: 4.2.5 - RPV Shell Welds Failure Probability Assessment Analysis (TRP 59.5)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.2.5 Calculation B13-035 REV 0 -

On page 25, Table 2: Perry RPV Material Adjusted Reference Temperature for 54 EFPY, indicates the most limiting plate material as A1155-1 and includes additional information in note 1.

On page 18, Table 4 from the Calc and in Table 4.2-5 from the submitted LRA, the most limiting plate material is shown as C2557-1 with an ART value of 56F, whereas, in the same tables A1155-1 has an ART value of 48.8F.

Provide the reasoning for choosing A1155-1 as the limiting plate material when C2557-1 indicates a higher ART, indicating it should be the limiting plate material.

2 4.2.5 RTmax in BWRVIP-329-A is defined as RT(unirradiated) + Shift.

Calculation B13-035 REV 0 - in Table 2: Perry RPV Material Adjusted Reference Temperature for 54 EFPY, EOI RTmax appears to include the margin value from RG 1.99 Rev 2, as well.

Was it intentional to include a margin value when calculating RTmax for 54 EFPY?

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: B.2.7 - Bolting Integrity Program (TRP 19)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

B.2.7 B-30 Existing LRA AMP B.2.7 states that the AMP manages the cracking, loss of material and loss of preload for pressure retaining bolting using preventive measures and inspection activities. It also states that safety-related closure bolting on pressure retaining joints will be managed by LRA AMP B.2.3, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD. It further states that non-safety (states non-ASME) pressure retaining bolted joints will be maintained by AMP XI.M36 External Surfaces Monitoring of Mechanical Components Program, which is a new program at PNPP.

GALL Report, Rev. 2, AMP XI.M18, Bolting Integrity, program to which the enhanced LRA AMP B.2.7 claims consistency, states that the AMP manages pressure retaining closure bolting not within the purview of head closure stud bolting (AMP XI.M3) or structural bolting (AMPs XI.S1, XI.S3, XI.S6, XI.S7, and XI.M23).

However, GALL Report, Rev.2, AMP XI.M 36, External Surfaces Monitoring of Mechanical Components, program states that it manages the effects of aging for metallic, such as piping, piping components, ducting, polymeric components. The AMP does not discuss pressure retaining bolting.

It is not clear how LRA AMP B.2.18, External Surfaces Monitoring of Mechanical Components, program, which is consistent with GALL Report, Rev.2, AMP XI.36, will manage the effects of aging for non-safety (states non-ASME) pressure retaining bolted joints.

Explain how and justify why LRA AMP B.2.18 would be the appropriate aging management program, not the LRA B.2.7 or LRA AMP B.2.3 to manage the effects of aging for non-safety-related structural bolting on pressure retaining joints.

2 Table A.3, A-63, B-31 Existing LRA AMP B.2.7 Bolting Integrity discusses five enhancements will be implemented to the existing Bolting Integrity program. In the 3rd enhancement (LRA Table A.3 Explain how leakage at underwater bolted joints would be monitored, detected, and investigated for the

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Item# 7, and B.2.7 License Renewal Commitments and LRA AMP B.2.7), it states that visual inspection of submerged bolting for the emergency service water pumps, diesel and motor fire pumps, emergency service water screen wash pumps and spent Fuel Rack Grid Structure will be performed using GALL, Rev. 2, AMP XI.M18, to which LRA AMP B.2.7 claims consistency. The program AMP XI.M18 uses inspection leakage, loss of material, cracking, and loss of preload. It is not clear how leakage from pressure retaining components would be monitored and detected under water through program elements Parameters Monitored and Detection of Aging Effects, as enhanced.

enhanced aging management program # 3 (Program Elements Affected: Parameters Monitored and Detection of Aging Effects).

3 Table A.3, Item# 7, and B.2.7 A-63, B-31 Existing LRA AMP B.2.7 Bolting Integrity proposed enhancement to Preventive Actions, program element states that [p]reventive measures will include using bolting material that has an actual measured yield strength limited to less than 1,034 megapascals (MPa) (150 kilo-pounds per square inch

[ksi]). It is not clear what the preventive measures are, the effects of aging they are set to prevent, and whether they are limited to 150 ksi bolts only.

LRA AMP B.2.7 Bolting Integrity is an existing aging management program, selection of bolting material is part of the program, as identified in GALL Report, Revision 2 under Preventive Actions. Does this mean the existing program is deficient in selecting appropriate materials (material with an actual measured yield strength more than 150 ksi) including lubricants/sealants containing sulfur that could cause stress corrosion cracking of bolts? If yes, what remediation measure(s) would be taken for aging management?

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 4

Table A.3, Item# 7, and B.2.7 A-62, B-31 Existing LRA AMP B.2.7 Bolting Integrity proposes to enhance the Parameters Monitored and Detection of Aging Effects program elements for visual inspection of submerged Spent Fuel Rack Grid Structure for loss of material and loss of preload at a 10 year frequency. It is not clear how the proposed enhancements would be implemented to visually inspect the submerged spent fuel pool grid for loss of material and loss of preload. It is also not clear whether the 10 year frequency of inspection is considered adequate.

Explain how loss of material and loss of preload of bolts in the spent fuel rack grid would be inspected (Program Element Affected: Element 4 Detection of Aging Effects).

Justify adequacy of the 10 year periodic visual inspection for cracking, loss of material, and loss of preload for the underwater bolting.

5 B.2.7 B-32 Existing LRA AMP B.2.7 Bolting Integrity states that Element 10 of NUREG-2191 includes the following OE, SCC of A-286 stainless steel closure bolting has occurred when seal cap enclosures have been installed to mitigate gasket leakage at valve body-to-bonnet joints (NRC IN 2012-15) is not applicable because PNPP does not operate with borated water. The staff finds that operating experiences, as documented in CR-2016-08070, CR-2016-11672, CR-2018-08540, and CR-2021-04025, that boron deposits have been observed due to leakage from pipe.

Explain why Element 10 of NUREG-2191 OE SCC of A-286 stainless steel closure bolting would not be applicable.

LRA Sections: B.2.26 - Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program and 4.6.1 - Crane Load Cycles (TRPs 24 & 68)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

LR Application:

1-Section 4.6.1 (TLAA) and Page 4.6-1 The following cranes are clearly listed (in-scope) in Sections 4.6.1 and A.2.6.1 of the LR Application:

(1) Reactor Building Crane, (2) Fuel Handling Building Crane and (3) Emergency Service Water Pump House Crane Further, Reports in the Portal: LRPY-AMP-XI.M23, Revision 3, and LRPY-TLAA-001, Revision 5, also list the same cranes above.

Describe why the cranes in scope were not listed in Sections B.2.26 (AMP) and A.1.26 (UFSAR) in the LR application.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 2-Section B.2.26 (AMP) 3-Section A.1.26 (UFSAR) 4-Section A.2.6.1 (UFSAR)

Portal:

1-LRPY-AMP-XI.M23, Rev. 3, and 2-LRPY-TLAA-001, Rev. 5.

Section 3.7.1 Page B-82 Page A-28 Page A-54 Page 5-14 Page 38 of 75

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Sections: B.2.20 - Fire Protection Program (TRP 26)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.4.2.36 2.4-48 LRA Section 2.4.2.36 states, The concrete curb and fire barriers provide fire protection for the transformers. These are considered as structural bulk commodities, for details, refer to Section 2.4.4.

LRA Table 2.4.4-1 includes component types of Concrete Fire Barrier and Flood Curbs. However, the only Flood Curbs with a fire barrier intended function in LRA Table 3.5.2-4 are steel, not concrete.

Given that LRA Section 2.4.2.36 states that the concrete curb provides fire protection for the transformers, please discuss whether LRA Table 3.5.2-4 should include concrete flood curbs with a fire barrier intended function managed by both the Fire Protection and Structures Monitoring programs.

2 2.4.2.26 2.4-41 LRA Section 2.4.2.26 states the reinforced concrete fuel oil storage tank dike helps prevent fire and is within the scope of license renewal. However, LRA Tables 2.4.2-1 and 3.5.2-2 do not appear to include the component type of fuel oil storage tank dike.

Given that LRA Section 2.4.2.26 states the reinforced concrete fuel oil storage tank dike helps prevent fire and is in scope of license renewal, please discuss whether LRA Tables 2.4.2-1 and 3.5.2-2 should include this component with a fire barrier intended function.

3 2.3, 3.5 2.3-1, 3.5-1 The definition for intended function Pressure Boundary in LRA Table 2.3-1 states, Additionally, for components such as ductwork and fire damper housings, the pressure boundary function includes providing a barrier to the spread of fires.

The following LRA sections indicate the system includes fire dampers with a pressure boundary intended function:

2.3.2.2, Annulus Exhaust Gas Treatment (M15) 2.3.3.1, Auxiliary Building Ventilation (M38) 2.3.3.5, Computer Room HVAC (M27) 2.3.3.13, Control Room HVAC (M25) and Emergency Recirculation (M26) 2.3.3.14, Controlled Access and Miscellaneous Equipment Areas HVAC (M21)

Please discuss how the operating experience in NUREG-2191 associated with fire damper assemblies was considered and discuss why some fire dampers are managed by the Fire Protection program while others are not.

Please discuss why some of the fire dampers in Bulk Commodities were exposed to air either internally or externally while the fire dampers managed by AMPs XI.M36 and XI.M38 appear to be exposed to air internally and externally.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 2.3.3.19, Emergency Closed Cooling Pump Area HVAC (M28) 2.3.3.26, Fuel Handling Area Ventilation (M40) 2.3.3.31, Intermediate Building Ventilation (M33) 2.3.3.36, Miscellaneous Area Ventilation (M46) 2.3.3.37, Miscellaneous Electrical Areas Smoke Ventilation (M49) 2.3.3.41, Offgas Building Ventilation (M36) 2.3.3.49, Radwaste Building Ventilation (M31) 2.3.3.57, Steam Tunnel cooling (M47) 2.3.3.63, Turbine Building Ventilation (M35)

For these systems, the External Surfaces Monitoring of Mechanical Components (XI.M36) and Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (XI.M38) are credited to manage loss of material for steel damper housings exposed externally and internally to indoor uncontrolled air.

For Bulk Commodities (LRA Table 3.5.2-4), the Fire Protection and Structures Monitoring programs are credited to manage the following:

Cracking and loss of material for aluminum, steel, and galvanized steel damper and louver housings and fixed louvers exposed externally to uncontrolled indoor air and outdoor air Loss of material for steel fire damper housings exposed externally to indoor uncontrolled air Los of material for steel louver and damper housings including 2M49 fire dampers exposed internally to indoor uncontrolled air The staff notes that LRA Section 2.1.3 indicates that NUREG-2191 contain potential operating experience applicable to Perry. Volume 1 of NUREG-2191 include AMR item 3.3.1-255 Please discuss whether aging will be managed for galvanized steel Damper and louver housings and fixed louvers4 and Louver and Damper housings including 2M49 Fire Dampers1 exposed externally and internally to indoor uncontrolled air, respectively. In addition, discuss the use of standard note A associated with AMR item 3.5.1-95 for Louver and Damper housings including 2M49 Fire Dampers1 given that the component is different.

Please discuss why RA Sections A.1.20 and B.2.20 and the Fire Protection Basis document (LRPY-AMP-XI.M26) do not discuss fire dampers.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions to manage loss of material and cracking for any material fire damper assembly exposed to air by the Fire Protection program. NUREG-2221 states, The Chapter VII line items that cite ducting and ducting components cite aging management program (AMP) XI.M32, AMP XI.M36, and AMP XI.M38 to manage associated aging effects. Fire dampers are not addressed in these AMPs. A new aging management review (AMR) item (3.3.1-255 citing AMP XI.M26), addresses only fire damper assemblies to eliminate possible confusion about which program manages the associated aging effects for these components.

Therefore, it is unclear whether the operating experience in NUREG-2191 associated with fire damper assemblies was considered, and it is unclear why some fire dampers are managed by the Fire Protection program while others are not.

In addition, it is unclear why some of the fire dampers in Bulk Commodities were exposed to air either internally or externally. The fire dampers managed by AMPs XI.M36 and XI.M38 appear to be exposed to air internally and externally.

It is unclear whether aging will be managed for galvanized steel Damper and louver housings and fixed louvers4 and Louver and Damper housings including 2M49 Fire Dampers1 exposed externally and internally to indoor uncontrolled air, respectively, since the Fire Protection Program and Structures Monitoring program is credited to monitor loss of material for galvanized steel Damper and louver housings and fixed louvers3 exposed externally to outdoor air and steel Louver and Damper housings including 2M49 Fire Dampers exposed internally to indoor uncontrolled air. In addition, it is unclear why standard note A was used for Louver and Damper housings including 2M49 Fire Dampers1 given that AMR item 3.5.1-95 in Revision 2 of NUREG-1801 indicates there are no aging effects/mechanisms requiring management by an AMP

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions for aluminum, galvanized steel, and stainless steel support members; welds; bolted connections; support anchorage to building structure exposed to indoor uncontrolled air.

The staff notes that even though the Fire Protection program is credited to manage fire dampers, LRA Sections A.1.20 and B.2.20 and the Fire Protection Basis document (LRPY-AMP-XI.M26) do not discuss fire dampers.

4 3.3 3.3-297 LRA Table 3.3.2-24 credits the Fire Protection program for managing loss of material for gray cast iron piping exposed internally to raw water with leakage and pressure boundary intended functions (Rows 38 and 56, NUREG-1801 Item VII.G.A-400 (LR-ISG-2012-02) and Table 1 Item 3.3.1-127).

Components managed by the Fire Protection program in the GALL-LR are exposed to air, not raw water.

In addition, LRA Section 3.3.2.2.8 states the Fire Water System program will be enhanced to manage loss of material due to recurring internal corrosion. Therefore, it is unclear why the Fire Protection program is credited to manage loss of material due to recurring internal corrosion.

Please discuss why the Fire Protection program is credited to manage loss of material for gray cast iron piping exposed internally to raw water with leakage and pressure boundary intended functions.

5 2.4 2.4-1 The following LRA sections include roofing that meets fire protection requirements:

2.4.2.1, Turbine Building, Unit 1 2.4.2.2, Turbine power Complex (Condensate Demineralizer Bldg.), Unit 1 2.4.2.13, Service Building However, LRA Tables 2.4.2-1 and 3.5.2-2 do not appear to include any roofing components that have a fire barrier intended function. LRA Table 3.5.2-4 appears to have very specific roofing components, therefore, it is unclear if the roofing that meets fire protection requirements in the previously stated sections are captured.

Please discuss where the roofing that meets fire protection requirements in LRA Sections 2.4.2.1, 2.4.2.2, and 2.4.13 are addressed in the LRA. In addition, please discuss whether the roofing in LRA Sections 2.4.2.16 and 2.4.2.17 meets fire protection requirements.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions It is unclear if the roofing in the following LRA sections meet fire protection requirements:

2.4.2.16, Turbine Building, Unit 2 2.4.2.17, Turbine Power Complex (Condensate Demineralizer Bldg.) Unit 2 6

2.4, 3.5 2.4-1, 3.5-126 The following LRA sections appear to include drywall with a fire resistance rating:

2.4.2.1 2.4.2.2 2.4.2.13 2.4.2.16 2.4.2.17 2.4.2.39, Water Treatment Building The staff also notes that the following documents on the portal refer to gypsum board: SP-2100, PTI-P54-P0054A, PTI-P54-P0054, PAP-1910, and FPI-A-I01.

However, LRA Tables 3.5.2-2 and 3.5.2-4 do not appear to include drywall or gypsum board as a material. The staff notes that LRA Tables 3.5.2-2 includes beams, columns, floor slabs and interior walls with a fire barrier intended function, however, the material is concrete, not drywall.

Drywall only appears to be addressed in LRA Section B.2.20 and the Fire Protection program basis document (LRPY-AMP-XI.M26) as operating experience.

The staff notes that PTI-P54-P0054A and FPI-A-I01 note holes, cracks, and surface removed as applicable aging effects for gypsum board.

SLR-ISG-2021-02-Mechanical (ML20181A434) added AMR items VII.G.A-805, VII.G.A-806, and VII.G.A-807 to Table Please discuss where drywall/gypsum board is addressed in the LRA.

Please discuss why change in material properties, delamination, and separation were not cited in PTI-P54-P0054A and FPI-A-I01 as applicable aging effects for gypsum board.

In addition, discuss whether Fire Protection program instructions/procedures require enhancement, if new aging affects are identified in response to this breakout question.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions VII.G in Volume 1 of NUREG-2191 and Table 3.3-1 in NUREG-2192. The aging effects for cementitious coatings, silicates, and subliming compounds used as fireproofing/fire barriers exposed to air are loss of material, change in material properties, cracking, delamination, and separation. These aging effects are consistent with Section 6, Fire Barriers, of EPRI 3002013084, Long-Term Operations:

Subsequent License Renewal Aging Affects for Structures and Structural Components (Structural Tools), November 2018.

LRA Section 2.1.3 states that SLR-ISG-2021-02 was considered as part of operating experience to optimize aging management effectiveness.

With regards to SLR-ISG-2021-02, the Fire Protection program basis document (LRPY-AMP-XI.M26) states, PNPP structural components contains these commodities, materials and aging effects assigned the Fire Protection and Structures Monitoring Programs. Both program methods are consistent with the Table VII.G items and can detect and manage these aging effects. The Civil AMRs include the aging effect of Cracking/delamination.

Therefore, it is unclear why change in material properties, delamination, and separation were not cited in PTI-P54-P0054A and FPI-A-I01 as applicable aging effects for gypsum board given that it is similar to silicate fireproofing/fire barriers.

7 3.3 3.3-97 Revision 2 of NUREG-1801 includes AMR item 3.3.1-57 for managing increased hardness, shrinkage, and loss of strength due to weathering of elastomer fire barrier penetration seals exposed to indoor uncontrolled air with the Fire Protection program.

The Discussion of AMR item 3.3.1-57 in LRA Table 3.3.1 states, The Fire Protection Program will manage change in material properties and cracking of elastomer fire barrier Please discuss the use of change in material properties for elastomer fire stops, penetration sealant (fire), seismic isolation joint, and shield building electrical penetration seals and sealants.

Please confirm that change in material properties for concrete fire barriers is

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions penetration sealant, seismic isolation Joint, and fire stop commodities subject to aging management and exposed to air

- indoor in the Bulk Civil Commodities.

The staff notes that LRA Section B.2.20 states the Fire Protection program manages change in material properties for elastomer (such as cracking or crazing, swelling, discoloration, and melting).

The LRA is not clear whether change in material properties includes increased hardness, shrinkage, and loss of strength due to weathering where AMR item 3.3.1-57 is cited for elastomer fire stops, penetration sealant (fire), and seismic isolation joint in LRA Table 3.5.2-4.

Revision 2 of NUREG-1801 includes AMR item 3.5.1-33 for managing loss of sealing due to wear, damage, erosion, tear, surface cracks, or other defects for elastomers, rubber and other similar material seals and gaskets exposed to indoor uncontrolled or outdoor air by the 10 CFR Part 50, Appendix J.

Therefore, it is unclear what change in material properties, in instances where AMR item 3.5.1-33 is cited in LRA Table 3.5.2-1 for elastomer shield building electrical penetration seals and sealants, is referring to (i.e., increased hardness, shrinkage, and loss of strength).

Revision 2 of NUREG-1801 includes AMR item 3.3.1-61 for managing cracking and loss of material due to freeze-thaw, aggressive chemical attack, and reaction with aggregates for reinforced concrete structural fire barrier walls, ceilings, and floors exposed to outdoor air by both the Fire Protection and Structures Monitoring programs.

The Discussion for AMR item 3.3.1-61 in LRA Table 3.3.1 states, The Fire Protection program will manage change in referring only to loss of bond and reduction of strength.

In addition, please discuss why loss of material due to freeze-thaw, aggressive chemical attach, and reaction with aggregates was not cited as an aging effect/mechanism for the concrete fire barriers exposed to outdoor air, where AMR item 3.3.1-61 is cited; and what the difference is between Rows 72 and 73 in LRA Table 3.5.2-4.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions material properties and cracking for concrete fire barriers exposed to air outdoor in the Bulk Civil Commodities.

The staff notes that LRA Section B.2.20 states the Fire Protection program manages change in material properties for concrete (loss of bond and reduction of strength).

The staff would like to confirm their understand that change in material properties, in instances where AMR item 3.3.1-61 is cited for concrete fire barriers in LRA Table 3.5.2-4, is only referring to loss of bond and reduction of strength as noted in LRA Section B.2.20.

In addition, it is unclear why loss of material due to freeze-thaw, aggressive chemical attack, and reaction with aggregates was not cited as an aging effect/mechanism for the concrete fire barriers exposed to outdoor air, where AMR item 3.3.1-61 is cited. The staff notes that AMR item 3.3.1-62 (loss of material due to corrosion of embedded steel) was cited to manage loss of material (Row 72) and loss of material due to corrosion of embedded steel reinforcing (Row 73) of concrete fire barriers exposed to outdoor air. It is unclear what is the difference between Rows 72 and 73 in LRA Table 3.5.2-

4.

8 3.5 3.5-139 In LRA Table 3.5.2-4, loss of sealing of elastomer fire stops with FB and SRE intended functions, and elastomer penetration sealant (fire) with EN, FB, FLB, SPB, SNS, and SRE intended functions are managed by the Fire Protection program and Structures Monitoring program. However, only the Fire Protection program is cited for managing change in material properties and cracking of elastomer fire stops with FB and SRE intended functions, and elastomer penetration sealant (fire) with EN, FB, FLB, SPB, SNS, and SRE intended functions.

Please discuss whether the Structures Monitoring program should also be cited for managing change in material properties and cracking of elastomer fire stops and elastomer penetration sealant (fire) associated with the SRE, EN, FLB, SPB, and SNS intended functions.

In addition, discuss why the Fire Protection program instructions/procedures on the portal do

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions It is unclear whether the Structures Monitoring program should also be cited for managing change in material properties and cracking associated with the SRE, EN, FLB, SPB, and SNS intended functions.

The staff notes that LRA Table 3.5.2-1 cites both the Fire Protection and 10 CFR 50, Appendix J programs to manage change in material properties, cracking, and loss of sealing for elastomer shield building electrical penetration seals and sealant with FB, SPB, SSR, FB, and SRE intended functions.

This approach appears to ensure the appropriate programs are used to manage the effects of aging to ensure the fire barrier and structural intended functions are maintained during the period of extended operation.

The staff also notes that the Fire Protection program instructions/procedures on the portal do not appear to address loss of sealing for elastomer fire stops, penetration sealant (fire), and seismic isolation joint.

not appear to address loss of sealing fire stop, penetration sealant (fire), and seismic isolation joint.

9 3.5 3.2-25 The Discussion for AMR item 3.5.1-10 in LRA Table 3.5.1 states, Not applicable - This table row applies to PWRs.

However, LRA Table 3.5.2-1 cites AMR item 3.5.1-10 for several stainless steel components, including the drywell mechanical penetrations also managed by the Fire Protection program.

Section 3.5.2.2.1.6 states, Although an aggressive chemical environment doesnt exist, the potential for SCC is assumed for these components, and the aging effect is managed by the ASME Section XI, Subsection IWE and 10 CFR 50, Appendix J programs for containment penetrations, and by the ASME Section XI, Subsection IWE and Structures Monitoring programs for drywell mechanical penetrations. However, LRA Table 3.5.2-1 does not cite the Structures Monitoring program for managing cracking due to SCC for the stainless steel Please clarify the use of AMR items 3.5.1-10 and 3.5.1-38, and the programs that manage cracking due to SCC of the stainless steel drywell mechanical penetrations.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions drywell mechanical penetrations, it cites AMR item 3.5.1-38 with a note E crediting the Fire Protection program. In addition, the Discussion of AMR item 3.5.1-38 in LRA Table 3.5.1 does not mention the Fire Protection program, in fact, it reads as if it is a not used AMR item. The staff does not disagree that the Fire Protection program should also be credited for managing the effects of aging since the drywell mechanical penetrations have a Fire Barrier intended function.

10 3.5 3.5-79 Revision 2 of NUREG-1801 includes AMR item 3.3-1-58 (VII.G.AP-150) for managing loss of material for steel halon/carbon dioxide fire suppression system piping, piping components, and piping elements exposed to indoor uncontrolled air by the Fire Protection program.

LRA Table 3.5.2-1 cites AMR item 3.3.1-58, with standard note A, for managing loss of material for steel drywell mechanical penetrations and fuel transfer tube penetrations exposed to indoor uncontrolled air by the Fire Protection program.

In addition, Table 3.5.2-4 cites AMR item 3.3.1-58, with standard note A, for managing loss of material for steel containment isolation dampers for U2 containment isolation (2M14), fire damper housing, fire damper housing 2M15, flood curbs, and louver and damper housings including 2M49 fire dampers, exposed to indoor uncontrolled air by the Fire Protection program.

Revision 2 of NUREG-1801 includes AMR item 3.3.1-60 (VII.G-28(A-90)) for managing cracking of reinforced concrete structural fire barrier walls, ceilings, and floors exposed to indoor uncontrolled air by the Fire Protection and Structures Monitoring programs.

LRA Table 3.5.2-1 cites AMR item 3.3.1-60, with standard note A, for managing cracking of concrete drywell, and Given that these components are different from the components associated with AMR items 3.3.1-58 and 3.3.1-60 in Revision 2 of NUREG-1801, please discuss the use of standard note A.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions concrete manway, hatches, manhole covers, and hatch covers exposed to indoor uncontrolled air.

11 As noted in Breakout Question 6, SLR-ISG-2021 Mechanical added AMR items for managing loss of material, change in material properties, cracking, delamination, and separation for cementitious coatings, silicates, and subliming compounds used as fireproofing/fire barriers exposed to air. In addition, the LRA and Fire Protection program basis document (LRPY-AMP-XI.M26) indicate that SLR-ISG-2021-02 was considered as operating experience.

LRA Table 3.5.2-4 includes the following:

Cracking and loss of material of pyrocrete fire proofing managed by the Fire Protection program (Notes F, 505)

Change in material properties, cracking/delamination, and loss of material of 3M Interam fire wrap and radiant energy shield managed by the Fire Protection program (Notes F, 502)

No aging effects or AMP for fiberglass/alumina silicate/calcium silicate/mineral fiber fire wrap and penetration sealant (fire) (Notes F, 503)

Change in material properties and cracking of unimpregnated fiberglass fabric; fiberglass fabric impregnated with elastomer penetration sealant (fire) and SRV tailpipe penetration boot seals managed by the Fire Protection and Structures Monitoring programs (Notes F, 517)

Pyrocrete is explicitly identified as a cementitious coating in AMR item 3.3.1-268 of SLR-ISG-2021-02. 3M Interam is explicitly identified as a subliming compound in AMR item Please address the following:

Why were change in material properties, delamination, and separation not identified as applicable aging effects for Pyrocrete.

Why was separation not identified as an applicable aging effect for 3M Interam.

Why were loss of material, change in material properties, cracking, delamination, and separation not identified as applicable aging effects for fiberglass/alumina silicate/calcium silicate/mineral fiber which appear to be materials similar to the silicates noted in AMR item 3.3.1-269.

Please provide additional details, including specific material names, safety data sheets, and any other documents supporting the cited aging effects, for unimpregnated fiberglass fabric; fiberglass fabric impregnated with elastomer.

Please discuss whether Fire Protection program instructions/procedures require

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 3.3.1-267 of SLR-ISG-2021-02. Marinite, Kaowool',

Cerafiber, Cera blanket, or other similar materials are the material examples for silicates in AMR item 3.3.1-269 of SLR-ISG-2021-02.

It is unclear why change in material properties, delamination, and separation were not identified as applicable aging effects for Pyrocrete.

It is unclear why separation was not identified as an applicable aging effect for 3M Interam.

It is unclear why loss of material, change in material properties, cracking, delamination, and separation were not identified as applicable aging effects for fiberglass/alumina silicate/calcium silicate/mineral fiber which appear to be materials similar to the silicates noted in AMR item 3.3.1-269.

It is unclear what specific materials are being described by unimpregnated fiberglass fabric; fiberglass fabric impregnated with elastomer.

The staff notes that Section 4:00 in SP-2100 lists several approved fireproofing/fire barrier materials (e.g., Thermo-lag).

However, LRA Table 5.3.2-4 only lists elastomer, 3M Interam, Fiberglass/Alumina silicate/Calcium silicate/Mineral fiber, and Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer. Therefore, it is unclear whether all component/material/environment combinations are addressed in the LRA.

The staff also notes that the Fire Protection program instructions/procedures on the portal address holes, and surface removed for pyrocrete; breaks, holes, missing layers, and removed sections for fire wrap; and holes and tears for radiant energy shields.

enhancement if new aging affects identified in response to this breakout question.

Please confirm that all component/material/environment combinations that require aging management are addressed in the LRA.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 12 N/A N/A Section 6.1 of EPRI 3002013084 states, in part, wire and other appurtenances used to secure fire wrap to the item being protected - is considered to be part of the fire wrap itself.

The staff notes that PTI-P54-P0075 refers to mechanical fasteners (bands). The procedure appears to inspect the mechanical fasteners only for being in place/installed and properly spaced.

The staff also noted that SP-2100 refers to stainless steel compression straps, banding, and banding wire.

Please discuss where materials used to secure fire wraps are addressed in the LRA, including AMR items for managing applicable aging effects. In addition, discuss whether Fire Protection program instructions/procedures may need enhancement to manage the effects of aging for materials used to secure fire wraps.

13 Appendix B

B-67 The Detection of Aging Effects program element of the Fire Protection program in Revision 2 of NUREG-1801 states that the visual inspections are performed by fire protection qualified personnel. However, LRA Section B.2.20 and the Fire Protection basis document (LRPY-AMP-XI.M26) do not indicate that the visual inspections are to be performed by qualified personnel.

The staff notes that some Fire Protection program instructions/procedures on the portal do indicate that qualified personnel are required to perform the instruction. However, the staff notes that the following procedures do not state that qualified personnel are required to perform the instruction:

FPI-A-I01, FPI-A-I02, PAP-1910 (discusses fire brigade qualification), PTI-P54-P0056, and PTI-P54-P0044 (unclear if Fire Protection Inspector is defined as qualified personnel).

Therefore, it is unclear why some of the Fire Protection program instructions/procedures do not state qualified personnel are required to perform the instruction.

Please discuss why LRA Section B.2.20, LRPY-AMP-XI.M26, FPI-A-I01, FPI-A-I02, PAP-1910, PTI-P54-P0056, and PTI-P54-P0044 do not clearly state that the inspections are performed by fire protection qualified personnel, consistent with the Fire Protection program as described in Revision 2 of NUREG-1801. In addition, discuss whether Fire Protection program instructions/procedures require enhancement to clearly state this requirement.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 14 N/A N/A Revision 2 of NUREG-1801 includes cracking, spalling, and loss of material as aging effects for fire barrier walls, ceilings, floors.

The staff noted that some Fire Protection program instructions/procedures (e.g., PTI-P54-P0054A, PTI-P54-P0054, FPI-A-I01) on the portal refer to walls and floors, but not ceilings. In addition, LRA Table 3.5.2-4 cites change in material properties, cracking, and loss of material as applicable aging effects for the concrete fire barrier component type. The Fire Protection program instructions/procedures on the portal address holes, scrapes, and surface area removed.

The staff did note that some condition reports (e.g., CR-2019-01234) on the portal addressed cracking of walls and floors.

However, change in material properties and cracking dont appear to be addressed in the Fire Protection program instructions/procedures.

Please discuss the following:

Why some Fire Protection program instructions/procedures on the portal refer only to walls and floors, and not ceilings.

Why change in material properties and cracking dont appear to be addressed in the Fire Protection program instructions/procedures on the portal for fire barrier walls, ceilings, and floors.

15 3.5 3.5-140 LRA Table 3.5.2-4 include steel flood curbs and elastomer seismic isolation joints; however, LRA Sections A.1.20 and B.2.20, Fire Protection program basis document (LRPY-AMP-XI.M26), and Fire Protection program instructions/procedures on the portal do not refer to these components.

Please discuss why flood curbs and elastomer seismic isolation joints are only addressed in LRA Table 3.5.2-4.

16 N/A N/A LRA Table 3.5.2-4 credits the Fire Protection and Structures Monitoring programs for managing the effects of aging for several component types. However, the staff notes that LRA Sections A.1.20 and B.2.20 (Fire Protection program) do not refer to the Structures Monitoring program, and LRA Sections A.1.43 and B.2.43 (Structures Monitoring program) do not refer to the Fire Protection program. However, the staff notes LRA Section B.2.43 does discuss operating experience related to fire barrier penetration seals and appears to include program enhancements related to some of the component types managed by both the Fire Protection and Structures Monitoring programs.

Please discuss why LRA Sections A.1.20 and B.2.20 do not refer to the Structures Monitoring program and LRA Sections A.1.43 and B.2.43 do not refer to the Fire Protection program. Discuss why the Fire Protection and Structures Monitoring program basis documents do not refer to each other. Please discuss how adequate coordination between the two programs may occur if the Fire Protection program instructions/procedures do not provide direction related to this coordination.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions The Fire Protection program basis document (LRPY-AMP-XI.M26) does indicate that the Structures Monitoring program is also credited to manage fire doors and fireproofing materials. However, other component types are listed in LRA Table 3.5.2-4. The Structures Monitoring program basis document (LRPY-AMP-XI.S6) includes only the operating experience in LRA Section B.2.43 stated above.

The staff also noted that the Fire Protection program instructions/procedures available on the portal do not refer to the Structures Monitoring program.

For instance, does a Structures Monitoring and Fire Protection inspector look at the components, or does either a Structures Monitoring or Fire Protection inspector look at the components and communicate deficiencies back to the other program?

17 B.2.43 B-125 LRA Section B.2.43 includes operating experience related to fire protection penetration sealant material. Specifically, an approximate 3/16 gap in sealant material at the floor penetration, which was determined not to challenge the integrity of the fire barrier since it was determined to be insufficiently deep (only an inch or two and not fully penetrating).

The staff also noted the following acceptance criteria in Fire Protection program instructions/procedures on the portal:

FPI-A-I01 and PTI-P54-P0054:

Penetration Seals (less than 3/8 shrinkage, radiant seals (no more than 50 percent of length removed))

Fire Rated Walls/Floors (no more than 100 square inches of a 2-inch-deep scrape per 100 square feet of barrier surface area removed) [Note: Also, in FPI-A-A02]

Gypsum Board (no holes or cracks deeper than one layer of gypsum board and/or no more than 100 square inches per 100 square foot of barrier surface removed)

Pyrocrete (no holes greater than 16 square inches, no more than 100 square inches per 100 square foot of barrier surface removed)

Please discuss whether the acceptance criteria are from the material manufacturer and supported by testing that indicates the fire barrier rating is not impacted, including identifying/providing supporting documents. In addition, please clarify the penetration seal acceptance criteria.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions PTI-P54-P0056:

Penetration Seals (less than 3/8 shrinkage, radiant seals (no more than 50 percent of length removed))

PTI-54-P0054A:

Fire Rated Walls/Floors (no more than 100 square inches of a 2-inch-deep scrape per 100 square feet of barrier surface area removed)

Gypsum Board (no holes or cracks deeper than one layer of gypsum board and/or no more than 100 square inches per 100 square foot of barrier surface removed)

Penetration Seals (less than 3/8 shrinkage, radiant seals (no more than 50 percent of length removed))

[Note: Section 5.1.4 of GMI-0076 states, Radiation barrier seals are required to be the full length of the penetration.]

SP-2100:

Silicon foam and elastomer seals - splits, gaps, or voids less than or equal to 3/8 in width but greater than 1/3 the depth shall be repair to a min depth of 1.

GMI-0077:

Penetration seals with splits, voids, or gaps greater than 3/8 in width and larger/deeper than 9 shall be reworked FPI-A-A02 Gypboard Walls/Ceilings shall not have more than 100 square inches per 100 square feet of one layer missing Penetration seals shall have no missing surface area unless installed 12 inches or more deep, no splits, gaps, or voids more than 3/8 wide and more than 4 deep

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions However, it didnt appear that documents containing the basis for the acceptance criteria are clearly referenced. In addition, there appears to be some minor inconsistencies in the acceptance criteria for penetration seals.

LRA Section B.2.4 - ASME Section XI, Subsection IWE Program (TRP 41)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

Table 3.5.1 Item No.

3.5.1-38 3.5-37 Description of issue and/or background LRA Table 3.5.1, Summary of Aging Management Evaluation for Containment Sand Component Supports, Item 3.5.1-38, consistent with NUREG-1801, Rev. 2 to manage the effects of aging from cracking due to stress corrosion cracking (SCC) lists Item II.B3.2.C-24 which suggest the use of AMP Chapter XI.S1 for IWE and Chapter XI.S4 for Appendix J.

LRA Table 3.5.2-1, Containment Structure, Unit 1, for SCC of Drywell Mechanical penetrations lists item II.B3.2.C-24 with LRA AMP Section B.2.20 (page B-67) Fire Protection Program.

Perry document Significant Deficiency Suppression Pool Stainless Steel Cladding Sensitization, dated March 15, 1982, (ADAM No. ML20042B60 and Perry Stainless steel Cladding Evaluation (ADAM No. ML20042B461) drawing Suppression Pool Cross Section, PNPP, on page 9 (Attachment 2-1) provides details about the suppression pool and containment details.

1. Discuss how the Fire Protection Program is applicable to detect cracking for Structural Steel Elements, particularly for the suppression chamber shell (interior surface) where the GALL Report guidance suggest Item II.B3.2.C-24 and GALL Report AMPs XI.S1 and XI.S4.

2. Discuss whether and how Chapter XI.S1 and Chapter XI.S4 are being used for managing the aging effect of cracking due to SCC aging mechanism.

3. Discuss how LRA B.2.20 is applicable to LRA Table 3.5.1 Item 3.5.1-38.

4. Discuss how the steel containment vessel is protected from corrosion due to inside uncontrolled environment (air-indoor).

2 B.2.4 B-22 LRA Section B.2.4 states that the applicable ASME Code for the current (fourth) 10-year inspection interval for PNPP, which commenced May 18, 2019, and expires on May 17, 2029, is ASME XI, 2013 Edition, as modified by 10 CFR 50.55a or relief granted in accordance with 10 CFR 50.55a.

Discuss if PNPP accounted for, evaluated, and subsequently reconciled differences between ASME Section, Subsection IWE 2004 and 2013 Editions when updating its existing LRA AMP B2.4 to that of GALL Report XI S1.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions The GALL Report NUREG-1801, Revision 2, for AMR is based on ASME Section XI, 2004 Code Edition. It is not clear whether PNPP confirmed continued consistency of its LRA AMP B2.4 to GALL Report XI.S1.

3 B.2.4 B-23 PNPP OE states that, as reported in the inservice inspection summary report for Cycle 14 (June 7, 2011, to May 16, 2013),

general visual examination of the containment and drywell interior and exterior surfaces was performed prior to and during the Cycle 14 refueling outage. Some areas of minor coating degradation were identified and documented on condition reports. None of the reported rusted areas exhibited any significant material loss and evaluations determined that loss of material did not exceed the allowable of 10% of the designed wall thickness.

1. Discuss if all relevant CRs are taken into account for steel containment degradation in LRA AMRs.

2. Discuss if any additional OE information is available related to coating degradation in between 2013 and 2023.

3. If degradation exists, what corrective measures PNPP takes to ensure that compromised coating will not result to loss of material in the steel containment beyond the ASME Code allowable 10%.

4 3.5.1-30 3.5-32 LRA Table 3.5.1, Item 3.5.1-30 lists AMPs XI.S1, ASME Section XI, IWE, and XI.S4, 10CFR Pat 50, Appendix J, for managing the Aging Effect/Mechanism Loss of preload due to self-loosening for Pressure-retaining bolting. LRA Table 3.5.1, Item 3.5.1-30 further states that this is Consistent with NUREG-1801 for component and aging effect.

Additionally, the Condensation environment is considered equivalent to Air - outdoor for evaluation of GALL consistency for Loss of preload for steel bolting.

The staff notes that there is no LRA Table 2, item referencing LRA Table 3.5.1, Item 3.5.1-30. The staff also notes that the NUREG-1801 recommends for AMR item II.A3.CP-150, which lists any-environment.

1. Discuss whether LRA Table 3.5.1, Item 3.5.1-30 is not applicable. If so, state the applicable AMPs and Table 3.5-1 items of NUREG-1801, Rev.2 and NUREG-1800, Rev.2, for this aging effect, component, material, and environment.

Discuss whether there is a concern of using the GALL Report recommendation for any-environment for this aging effect, component, material, and environment.

5 PNPP LRA OE LRPY-OE-001, Rev. 3 (from eDocs eportal)

Page 4 of 6

OE Document LRPY-OE-001, Revision 3, PNPP LRA OE Review Results and Summary lists LRP-AMP-XI-S1, Revision 4 and includes the following CRs:

CR-2013-04102 CR-2013-04105 CR-2013-04102 and CR-2013-04105 discuss Degraded Containment Coatings that exams are being performed in Discuss how the findings of CR-2013-04102 and CR-2013-04105, related to professional engineers (PE) or knowledgeable individuals, are implemented in the PNPP IWE inspection procedures and considered in LRA AMRs.

State whether and how a PE determined that, for indications found, the containment

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions accordance with ASME Section XI, IWE utilizing Visual examination.

Procedure NQI-1042 states that although there is no evidence that structure integrity of the Containment is affected, the acceptance criteria given in NQ-1042 for Containment General Visual exams states, that The Registered Professional Engineer or knowledgeable individual will determine that the recorded area do not affect either the containment structure integrity or leak tightness.

structural integrity and leak tightness were assured.

LRA Section: B.2.6 - ASME Section XI, Subsection IWL Program (TRP 42)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

Table 3.5.1 3.5-12 and 3.5-28 FE 3.5.2.2.1.8, Cracking due to Expansion from Reaction with Aggregate states that there is no OE for cracking due to reaction with aggregates. It also states that [a]ccessible concrete surfaces of the containment fill in annulus are monitored for cracking due to expansion from reaction with aggregates by the ASME Section XI, Subsection IWL program and are addressed under Item Number 3.5.1-19. LRA Table 1 line item 3.5.1-19 assigned to manage cracking due to expansion from reaction with aggregates, states that [a]ging of annulus concrete will be managed by the ASME Section XI, Subsection IWL.

It is clear that PNPP monitors the concrete annulus for cracking due to expansion from reaction with aggregates. It is not clear, however, how PNPP plans to monitor potential cracking of the annulus concrete due to expansion from reaction with aggregates, when there are no Table 2 AMR line items for concrete fill-in-annulus referencing Table 1 line item 3.5.1-19.

1. State why the LRA did not include a Table 2 AMR line item for concrete fill-in-annulus referencing SRP-LR Table 1 line item 3.5.1-19.

2. Provide the necessary Table 2, item to manage effects of aging for annulus concrete including monitoring of its cracking/spalling due to expansion from reaction with aggregates with LRA AMP B.2.6 ASME Section XI, Subsection IWL program.

2 B.2.6 A.1.6 B-27 A-12 LRA AMP B.2.6 ASME Section XI, Subsection IWL as enhanced states that it is an existing condition monitoring program performing periodic visual examinations of accessible areas of the annulus concrete in accordance with ASME Section XI, Subsection IWL, 2013 Code Edition, as modified by 10 CFR 50.55a. The program is enhanced to include ACI 349.3R requirements implicit to Subsection IWL of the 2013 Code Edition.

1. Provide additional OE indicating when and how the noted aging effects of loss of material due to corrosion at the interface of steel-felt material-concrete fill-in-annulus, spalling (with potential cracking) and aggregate exposure of

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions ASME Section XI, Subsections IWE and IWL, 2013 and 2004 Code Editions, the latter which the GALL Report development is based, require general and visual examinations of the steel containment and concrete in annulus. Table IWE-2500-1 Examination Categories, E4.11 requires a 100% detailed visual examination of visible steel containment surfaces. ASME Table IWL-2500-1 Examination Categories, L1.11 requires general visual examination of all accessible concrete surface areas, while L1.12 requires a detail visual examination of suspect to aging areas.

The staff notes that RFO7, ISI Summary Report, dated August 2, 1999 references ASME Section XI, 1989 Code Edition and Drawing SS-305-503-139 and states that IWE 4.11 VT-3 exam found areas of heavy rust at the interface of the annulus concrete pour compressible material and the metal containment. IWL L1.11 VT-3C (general visual) exam found areas of exposed aggregate and spalling, while a L1.12 VT-1C (detail visual) exam found areas of exposed aggregate beneath leaking E32 line.

The staff did not find additional/subsequent OE stating how PNPP addressed these noted aging effects. The staff also notes that compressible material potentially could retain moisture resulting to loss of material due to corrosion on adjoining ferrous based materials (steel). However, the staff notes LRA AMP B.2.6 states that the Structural Maintenance Rule Program performs periodic inspections of the containment annulus and that review of the Maintenance Rule Evaluation Worksheets for more than ten years concluded no adverse findings in the annulus concrete.

Although RFO7 noted that there are no concerns with the annulus concrete, it is not clear whether and how the RFO7 observed deficiencies in spalling annulus concrete (often accompanied by cracking) and heavy rust found at the interface of annulus concrete and containment steel were resolved or are ongoing. It is also not clear what is the condition of the containment coating at the flip side where the corrosion occurred/occurs (blistering?). If the compressible material is damp, has PNPP found the source of the moisture?

concrete fill-in-annulus noted in RFO7 ISI are resolved/managed.

2. Provide Drawing SS-305-503-139 and photos of suspect location.

Discuss current condition of the suspect location. Provide the most recent ASME Section XI, Subsection IWL ISI and examinations results for the concrete fill-in-annulus area.

3. If the compressible material is damp with ongoing containment steel corrosion, state whether loss of material calculations were performed where the heavy corrosion was detected at the interface of steel-felt material-concrete fill-in-annulus so that consistent with 10 CFR 54.21(a)(3) the steel containment will continue to perform its intended function during the period of extended operation.

4. Discuss effectiveness of the Maintenance Rule to meet the regulatory requirements of ASME Section XI, Subsections IWE and IWL, 2013 Code Edition as mandated and modified by 10 CFR 50.55a at the concrete fill-in-annulus area.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: B.2.5 - ASME Section XI, Subsection IWF (TRP 43)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

A.1.5 B.2.5 A-10 B-24 LRA AMP B.2.5, ASME Section XI, Subsection IWF, states that it is an existing program and after enhancements will be consistent with the GALL Report AMP XI.S3. The AMP states that greater than 1 inch nominal diameter high strength (HS) bolts of 150 ksi or greater are used at Perry. GALL Report AMP XI.S3 recommends that molybdenum disulfide (MoS2) not be used as a lubricant to such HS bolts as it is a potential contributor to stress corrosion cracking (SCC).

LRA AMP B.2.5 does not discuss whether molybdenum disulfide lubricants are or have been used at PNPP to such HS bolting. However, the LRA states that PNPP plans to volumetrically examine a sample of its 1 inch nominal diameter high strength (HS) bolts of 150 ksi or greater (see Parameters Monitored\\Inspected and Detection of Aging Effects program element enhancements) to verify their susceptibility to SCC once each 10-year ISI interval.

It is not clear whether the proposed HS bolt sample would be limited to those of ASME Class 1, 2, 3 and MC supports or include such bolting used elsewhere in the plant as well.

1. State whether PNPP used lubricants containing MoS2 on greater than 1 inch nominal diameter HS bolts of greater than or equal to 150 ksi and the including bolting is within the proposed sample.

2. Discuss how LRA AMP B.2.5 plans to refrain from further use of such lubricants particularly in 1 inch nominal diameter HS bolts of 150 ksi or higher during the period of extended operation (PEO).

3. Explain the adequacy of the proposed sample size for volumetric examination of greater than 1 inch nominal diameter HS bolts greater than or equal to 150 ksi to detect SCC particularly if MoS2 was used selectively across the plant.

4. Discuss whether the bolt sample is independent of the component support and hardware sampling of Table IWF-2500-1 (F-A), Examination Category F-A, Supports.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 2

A.1.5 B.2.5 A-10 B-24 LRA section B.2.5 states that the applicant's ASME Section XI, Subsection IWF is an existing program that after enhancements will be consistent with the GALL Report AMP XI.S3. The staff notes that ASME Section XI, Subsection IWF Examination Category F-A, Supports, Table IWF-2500-1 (F-A) states that to the extent practical, the same supports selected for examination during the first inspection interval shall be examined during each successive inspection interval.

It is not clear, whether the same supports selected for examination during the first inspection interval continue to be examined during each successive inspection interval when inspections and examinations identify deficiencies in the component supports and hardware sampled population, but such supports are accepted as is following an engineering evaluation and subsequently reworked to as new condition. The staff notes that samples containing reworked component supports, no longer represent the age-related degradation for the remaining ASME Class 1, 2, 3, and MC component support and hardware population.

1. State whether PNPP IWF of Class 1, 2, 3, and MC components supports samples include reworked supports to as new condition.

2. If so, state if samples are augmented to include additional component supports as representative of the remaining population.

3. If not, explain how the IWF ISI sampled Class 1, 2, 3, and MC component supports and hardware requiring no corrective actions per ASME Code,Section XI, Subsection IWF, acceptance criteria for supports and associated hardware but reworked to as new condition could represent the age-related degradation for the rest of the ASME component support population.

3 A.1.5 B.2.5 A-10 B-24 LRA Section B.2.5 states that the applicable ASME Code for the current (fourth) 10-year inspection interval for PNPP, which commenced May 18, 2019, and expires on May 17, 2029, is ASME XI, 2013 Edition, as modified by 10 CFR 50.55a or relief granted in accordance with 10 CFR 50.55a.

The GALL Report for AMR is based on ASME Section XI, 2004 Code Edition. It is not clear whether PNPP confirmed continued consistency of its LRA AMP B.2.5 to GALL Report XI.S3, even though its ISI and examinations are based on ASME Section XI, Subsection IWF 2013 Code Edition.

Discuss if PNPP accounted, evaluated, and subsequently reconciled differences, if any, between ASME Section XI, Subsection IWF 2004 and 2013 Code Editions when updating its existing LRA AMP B.2.5 to that of GALL Report XI.S3.

4 A.1.5 B.2.5 A-10 B-25 3.5-58 LRA AMP B.2.5 states that its Preventive Action program element is enhanced to specify the following conditions as unacceptable: [d]ebris, Clarify whether PNPP has any sliding surfaces within the purview of LRA AMP B.2.5. If so, provide Table 2, AMR line

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Table 3.5.1 dirt, or excessive wear that could prevent or restrict sliding of the sliding surfaces as intended in the design basis of the support.

LRA Table 3.5.1, Summary of Aging Management Evaluations for Containments, Structures and Component Supports, Item 3.5.1-75, addresses AMR of sliding surfaces. It states that consistent with the GALL Report ASME Section XI, Subsection IWF program will be used to manage loss of distortion, dirt, debris, mechanical function of lubrite sliding surfaces exposed to air-indoor uncontrolled in Structures and Component Supports.

The staff could not locate any Table 2 AMR item using Table 1, item 3.5.1-75. It is not clear whether PNPP has any sliding surfaces that consistent with the GALL Report need to be age managed during the period of extended operation. If so, it is not clear what AMR Table 2 item will support managing effects of aging for sliding surfaces subject to IWF ISIs and examinations.

items lining up with LRA Table 3.5.1, item 3.5.1-75 so that such sliding surfaces effects of aging are managed consistent with 10 CFR54.21(a)(3) during the period of extended operation.

5 A.1.3 B.2.3 A-9 B-20 PNPP 14th ISI Summary Report states that the exterior of the RV skirt support was examined for the first time during a forced outage on 6/15/12. It also states that the accessible portion of the skirt interior was examined during RFO14. However, the total coverage of the skirt VT-3 exam was 55% (100% of the exterior and approximately 10% of the interior). The RFO14 ISI summary references CR-2013-06770, which in its Attachment 1 states that the RV skirt is an ASME Section XI, Subsection IWF, Examination Category F-A, Item F1.40 component support. The CR also states that the skirt support boundary as defined in Figure IWF-1300-1 includes bolting to the RPV pedestal. Although the skirt is identified as a Class 1 support, the LRA states that PNPP plans to manage the effects of aging for accessible and inaccessible portions of the skirt with LRA AMP B.2.3, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program.

The staff reviewed LRA AMP B.2.5 but could not find an exception stating that the effects of aging for the RV skirt Class 1 support will be managed by LRA AMP B.2.3, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program. The staff also reviewed LRA AMP B.2.3, and its basis document LRPY-AMP-XI.M1 for enhancements to programs elements verifying its alignment to GALL XI.S3 guidance to manage the effects of aging for Class 1 RV support skirt but found none.

1. Discuss how PNPP will use LRA AMP B.2.3 to manage the effects of aging for the IWF Class 1 RV support skirt which is clearly within the purview of ASME Section XI, Subsection IWF and GALL AMP XI.S3.

2. Discuss why neither the LRA AMP B.2.3 nor its basis document indicate alignment of its program elements and/or existence of enhancements aligning it to GALL XI.S3 guidance for ISI and examinations.

3. Clarify whether PNPP plans for an exception to LRA AMP B.2.5 scope of program, program element. If so, provide the exception to LRA AMP B.2.5 and necessary enhancements to LRA AMP B.2.3, so that ISIs and examinations for the RV Class 1 support skirt are addressed

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions It is not clear how PNPP will use LRA AMP B.2.3 to manage the effects of aging for the ASME Class 1 RV support skirt. If so, to what extent, if any, PNPP plans to enhance LRA AMP B.2.3 specifically to manage the effects of aging for Class 1 RV support skirt consistent to GALL XI.S3 guidance for the period of extended operation.

appropriately within the regulatory framework during the period of extended operation.

6 A.1.3 B.2.3 A.1.5 B.2.5 A-9 B-20 A-10 B-25 LRA AMP B.2.5 states that [t]he applicable ASME Code for the current (fourth) 10 year inspection interval for PNPP, which commenced May 18, 2019, and expires on May 17, 2029, is ASME XI, 2013 Edition, as modified by 10 CFR 50.55a or relief granted in accordance with 10 CFR 50.55a.

The staff notes that the PNPP requested relief L-14-105 (ML14239A626) for the examination of the ASME Code Section XI, Subsection IWF examination ASME Class 1 support skirt and the NRC SE (ML14069A424) approving the relief, discuss the impracticality of the mandated VT-3 exam to yield the required 100% examination coverage.

to CR-2013-06770 states that in accordance with NRC approved Code Case N-460, essentially full coverage is achieved when a minimum of 90% coverage is achieved, so the Code required coverage is 90%. The staff reviewed ASME

1. Discuss whether and how PNPP plans to use Code Case N-460 applicable to ASME Section XI Class 1 (IWB-2500) and Class 2 (IWC-2500) welds or in conjunction with LRA AMP B.2.3 for ASME Section XI, Subsection IWF ISI and examination of the RV Class 1 support skirt for a potential reduction to its VT-3 examination requirements.

2. Discuss whether PNPP plans to request a relief from ASME Section XI, Subsection IWF, 2013 Code Edition ISI and examination requirements for the RV Class 1

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions BPVC.CC.NC-2023 for applicability of Code Case N-460 Alternative Examination Coverage for Class 1 and Class 2 Welds.

It is not clear whether and how PNPP plans to use Code Case N-460 in conjunction with LRA AMP B.2.3 for ASME Section XI, Subsection IWF inspection and examination of the RV ASME Class1 support skirt during the period of extended operation for potential reduction of the VT-3 examination surface when the Code Case applies only to ASME Section XI Class 1 (IWB-2500) and Class 2 (IWC-2500) welds.

Given the impracticality of the inspection, it is also not clear whether PNPP plans to request a relief from ASME Section XI, Subsection IWF, 2013 Code Edition ISI and examination requirements for the RV Class 1 support skirt when its LRA AMP B.2.3 states that it will manage the effects of aging for the RV skirt.

support skirt from the ASME Section XI, Subsection IWF, 2013 Code Edition, as modified by 10 CFR 50.55a. If so, discuss whether the relief will align with ASME Section XI, Subsection IWF requirements or with those of ASME Section XI, Subsections IWB, IWC, and IWD.

LRA Section: 4.5 - Containment Liner Plate, Metal Containments and Penetrations Fatigue Analysis (TRP 63)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.5.1 A.2.5.1 4.5-1, A-53 LRA Section 4.5.1, Containment Vessel of Perry Nuclear Power Plant (PNPP) describes the fatigue time limited aging (TLAA) analysis for the containment vessel, composed of a free standing pressure retaining steel cylinder with an ellipsoidal dome, secured to a steel-lined, reinforced concrete foundation mat It dispositions the TLAA as 10 CFR 54.21(c)(1)(iii) by stating that the Fatigue Monitoring Program will manage the effects of fatigue based on the highest fatigue usage component of the containment vessel for the period of extended operation. LRA Section 4.5.1, Table 4.5-1 Containment Fatigue Usage - Limiting Locations, identifies P-101 Lower Containment Penetration, as the as the limiting location for containment vessel fatigue TLAA. LRA Section

1. P-101 lower containment penetration was the only location selected as a limiting location for the fatigue TLAA of the containment structure.

Please justify the selection of this component as the bounding component for containment vessel structure fatigue monitoring. Indicate the type of penetration P-101 is.

2. To facilitate the review, please provide the transient events

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 4.5.1, however, does not describe the transients considered for containment vessel components and their design limits, leading to the selection of P-101 Lower Containment Penetration.

Additional information is needed to demonstrate that P-101 Lower Containment Penetration is the limiting location to be included in the Fatigue Monitoring Program tracking of transients consistent with 10 CFR 54.21(c)(1)(iii), so that the effects of cumulative fatigue damage on the intended functions of the containment vessel will be adequately managed for the period of extended operation.

(e.g., OBE, SSE, etc.) along with the number of transient cycles considered in the containment vessel fatigue analysis and the design cycle limits of each transient leading to a projected 60- year CUF of 0.507.

3. Please provide updates to the LRA and UFSAR supplement, as appropriate, to be consistent with the response to the above requests.

2 4.5.2, A.2.5.2 4.5-2 4.5-3, A-53 PNPP LRA Section 4.5.2, discusses the fatigue time limited aging analysis (TLAA) of containment piping penetrations selected for monitoring those of highest fatigue usage. This LRA section disposition the TLAA by stating that consistent with 10 CFR 54.21(c)(1)(iii) the Fatigue Monitoring Program will manage the effects of fatigue on the selected containment penetrations for the period of extended operation. However, LRA Section 4.5.2 does not describe the transients considered for fatigue analysis that led to the selection of these containment piping penetrations and their design limits.

Additional information is needed to demonstrate that the selected piping penetrations listed in LRA Table 4.5-2, Containment Penetration Fatigue Usage - Limiting Locations, bound the containment piping penetration population so that the effects of cumulative fatigue damage, due to cyclic loading, on the intended functions of the containment piping penetration will be adequately managed consistent with 10 CFR 54.21(c)(1)(iii) by the Fatigue Monitoring Program for the period of extended operation

1. Four containment piping penetrations were selected as a limiting location for fatigue monitoring. Please justify the selection of these component as bounding or representing all other containment penetrations.

2. To facilitate the review, please provide the number of transient cycles and associated events considered in the fatigue TLAA and the design cycle limits of each transient.

3. Please provide updates to the LRA and USAR supplement, as appropriate, to be consistent with the response to the above requests.

3 4.5.3, A.2.5.2 SRP-LR acceptance criteria in Section 4.6.2.1.1.1 states that a TLAA is acceptable under 10 CFR 54.21(c)(1)(i) when the existing CUF calculations remains valid because the number of assumed cyclic loads will not be exceeded during the period

1. Please demonstrate that the TLAA analyses for the penetration bellows meets the criteria of 10 CFR 54.21(c)(1)(i),

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions of extended operation.

PNPP LRA Section 4.5.3, states that, the fatigue analyses for these bellows determined they were capable of handling the movement from more normal operation or faulted cycles than were specified. The LRA Section 4.5.3 dispositioned the analysis of the penetration bellows as 10 CFR 54.21(c)(1)(i) by stating that the bellows fatigue analyses will remain valid for the period of extended operation.

PNPP LRA Section 4.5.3 does not include the number of operating transient cycles experienced by the penetration bellows to date and their extrapolation to 60 years of operation, demonstrating that the TLAA remains valid to the end of the period of extended operation and consistent with10 CFR 54.21(c)(1)(i) criteria.

that bellows are qualified for more than the 60-year projected number of startups and shutdowns and provide the following information:

a) Specification of the design cycle limits for all transient cycles considered. end of the period of extended operation.

b) Specification of the number of transient cycles experienced by the penetration bellows to date, and their extrapolation through the extended period of operation c) Demonstration indicating that the existing calculated 500 cycle cumulative usage factor (CUF) remains below the design limit of 1 through the end of the period of extended operation consistent with SRP-LR 4.6.2.1.1.1.

2. Please provide updates to the LRA and USAR supplement, as appropriate, to be consistent with the response to the above requests.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: 3.5.2.2 - Further Evaluation of Aging Management as Recommended by NURGE-1800 (TRP 74)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request

1.

3.5.2.2.1.7 Table 3.5-1 3.5-12 3.5-25 SRP-LR Section 3.5.3.2.1.7 states that a plant-specific program is not required if documented evidence confirms that where the existing concrete had air content of 3% to 8% (including tolerance) and subsequent inspection did not exhibit degradation related to freeze-thaw.

1. The applicant claims AMR item 3.5-1, 011 to be not applicable. PNPP is located in a region where weathering conditions are considered moderate, as shown in ASTM C33-90, Figure 1. Therefore, loss of material (spalling, scaling) and cracking due to freeze-thaw should be an applicable aging effect and subject to AMR.

2. LAR Section 3.5.2.2.1.7 states that inaccessible and accessible areas are designed in accordance with ACI 318 and ASTM standards for selection application and testing of concrete and concrete aggregates. It is unclear to the staff what is the air content of concrete mix used for the containment structure at PNPP.

3. LAR Section 3.5.2.2.1.7 does not appear to include the evaluation on whether plant-specific program is required for managing loss of material (spalling, scaling) and cracking due to freeze-thaw.

4. LAR Section 3.5.2.2.1.7 did not appear to include information about how loss of material (spalling, scaling) and cracking due to freeze-thaw will be managed in inaccessible areas.

1. Please evaluate the claim of no-applicability of AMR item 3.5-1, 011 and provide table 3.5-2 components, if necessary.

2. To facilitate the review, clarify air content of the concrete mix used for the containment structure at PNPP.

3. Evaluate whether a plant-specific program is required.

4. Please explain what AMPs will be used and how these AMPs will manage the aging effects of freeze-thaw in inaccessible areas.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 2

3.5.2.2.1.8 Table 3.5-1 3.5-12 and 3.5-13 3.5-26 SRP Section 3.5.3.2.1.8 indicates that the GALL Report recommends further evaluation of programs to manage cracking due to expansion and reaction with aggregates in inaccessible areas of concrete elements of BWR concrete and steel containments.

Table 3.5-2 did not appear to include components related to AMR item 3.5-1, 012. In accordance with NUREG-1801, Item II.A1.CP-67, cracking due to expansion from reaction with aggregates is applicable to concrete material for any environment, therefore should be applicable to PNPP foundation.

Please provide table 3.5-2 components and update the LAR as necessary.

3.

3.5.2.2.1.9 Table 3.5-1 3.5-13 3.5-26 SRP Section 3.5.3.2.1.9 indicates that the GALL Report recommends further evaluation of programs to manage increase in porosity and permeability due to leaching of calcium hydroxide and carbonation in inaccessible areas of BWR concrete and steel containments.

1. In relation to AMR item 3.5-1, 13. and as part of the OE, in March 1997 the applicant evaluated erosion of cement from porous concrete sub-foundations below the reactor building basemats. The applicant found that porous concrete layer (under the plant buildings) is fairly stagnant and leaching of calcium is localized in the peripheral portions of the porous concrete pad, and not uniformly throughout the pad. The engineering evaluation results that determined that the observed leaching of calcium hydroxide and carbonation in accessible areas has no impact on the intended function of the concrete structure are unclear.

2. 2During the NRC on-site audit on the week of October 23, 2023, the NRC staff observed leaching in accessible areas of the basemat. The staff requested the applicant to upload the operation experience with the key word search:

leaching, leaching AND accessible areas, calcium hydroxide AND carbonation leaching into the e-portal and

1. Provide the engineering evaluation results that determined that the observed leaching of calcium hydroxide and carbonation in accessible areas has no impact on the intended function of the concrete structure.

2. Provide all applicable information regarding your operating experience on the observation of leaching of calcium hydroxide and carbonation in accessible concrete areas, its evaluation for impact on intended functions, and demonstrate how it would be adequately managed in inaccessible concrete areas.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions the applicant indicated that the multiple searches resulted in zero CRs. It is unclear about the applicant operating experience on the observation of leaching of calcium hydroxide and carbonation in accessible concrete areas and its evaluation for impact on intended function.

4.

Table 3.5.1-19 3.5-28

1. The staff found no table 3.5-2 components related to AMR item 3.5-1, 019. In accordance with NUREG-1800, cracking due to expansion from reaction with aggregates is applicable to concrete material for any environment, therefore should be applicable to PNPP top surface of the annulus concrete.

Provide applicable table 3.5-2 components and update the LAR as necessary.

LRA Section: 3.5.2.2.2.4 - Cracking due to Stress Corrosion Cracking and Loss of Material due to Pitting and Crevice Corrosion (TRP 77)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

LRA Table 3.5-1 Page 3.5-48 SPR-SLR Table 3.5-1 Item 4 is applicable to steel elements (inaccessible areas) including drywell shell and drywell head.

1. PNPP Mark III containment design includes a drywell head and drywell wall. The applicant indicated that above the drywell vent region, the inside face of the drywell is formed with a steel plate and the surface of the upper drywell wall and drywell head are considered to be accessible. It is unclear if parts of the drywell all and head are located in inaccessible areas. The applicant did not identify any associated Table 3.5-2 components (e.g., II.B3.1.CP-113) associated with Table 3.5-1 Item 4.

1. Please clarify whether the inside face of the drywell formed with a steel plate and part of the drywell head area are located in inaccessible areas. If applicable, provide Table 3.5-2 components associated with AMR Item 3.5.1-4.

Please update the LRA accordingly.

2 LRA Table 3.5-1 Page 3.5-23 3.5-162 LRA Section 3.5.2.2.1.3, associated with LRA Table 3.5.1 Item 3.5.1-5, addresses steel elements in inaccessible areas for loss of material due to general, pitting, and crevice corrosion.

1. LAR Table 3.5-1, Item 5 indicates that it is consistent with NUREG 1801. For table 3.5.2-4 component 313

1. Please provide the appropriate NUREG-1801 item (i.e., II.B3.2.CP-98 ) applicable to Mark III containments and explain how the Water Chemistry and one time inspections Program will

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Bolting Integrity 3, the applicant cited note E, indicating it proposes to use a program other than the GALL Report-recommended program. It is unclear how the proposed program will manage the loss of material aging effect. In addition, the applicant cited NUREG-1801 Item II.B2.1.CP-63 and clarification is needed since this item appears to be only applicable to Mark II containments.

2. LAR Table 3.5.2 identifies only one component associated with AMR Table 3.5-1, item 5. It also states that the Containment Inservice Inspection IWE and the Containment Leak Rate Programs manage the loss of material of steel elements in this listing. It is unclear whether additional Table 3.5-2 components should be listed as part of AMR Table 3.5-1, Item 5.

3. The SRP-LR Section 3.5.2.2.1.3 states that the GALL Report recommends further evaluation of plant-specific programs to manage this aging effect if corrosion is indicated from the IWE examination. It is unclear whether the IWE program examination has identified instances of significant corrosion and if so, how those were managed.

adequately manage the aging effect of loss of material due to pitting and crevice corrosion for stainless steel components associated with Table 3.5.1, Item 5 (i.e., Table 3.5.2-4) that are exposed to a fluid environment and not accessible for visual inspection.

Indicate how the water chemistry and one time inspection program meets the intent of the GALL report recommended programs (Chapter XI.S1, ASME Section XI, Subsection IWE and Chapter XI.S4, 10 CFR Part 50, Appendix J).

2. If applicable, please provide a complete list for Table 3.5.2 for all components associated with AMR Table 3.5-1 Item 5 and update LAR as applicable.

3. Please provide information on whether the IWE program examinations has identified instances of significant corrosion and if so, provide details on the operating experience.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: 3.5.2.2.2.1 - Aging Management of Inaccessible Areas (TRP 87)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

3.5.2.2.2.1.1 Table 3.5-1 3.5-14 3.5-39 SRP-LR Section 3.5.3.2.2.1.1 states that a plant-specific program is not required if documented evidence confirms that the existing concrete had air content of 3% to 8% and subsequent inspection did not exhibit degradation related to freeze-thaw.

1. LRA Section 3.5.2.2.2.1.1 indicated that for PNPP Groups 1-3, 5, 7-9 structures, water/cement ratios and air entrainment percentages for concrete are within the limits provided in ACI 318. It is unclear whether existing concrete has been documented to have an air content of 3% to 8%.

2. LRA Section 3.5.2.2.2.1.1 and AMR item 3.5.1-42 claimed that the absence of concrete aging effects is confirmed by the ASME Section XI, Subsection IWL and the Structures Monitoring programs. However, it lacks the evaluation on how the ASME Section XI, Subsection IWL and the Structures Monitoring programs are engaged in conforming the absence of concrete aging effects in below grade inaccessible concrete areas, including whether an APM (or AMPs) will be needed to perform opportunistic inspections of normally inaccessible below grade concrete when excavated for any other reason.

3. The applicant claimed that AMR item 3.5.1-42 is consistent with GALL-LR. However, there are no associated Table 2 items (i.e., III.A1.TP-108, III.A2.TP-108, III.A3.TP-108, III.A5.TP-108, III.A7.TP-108, III.A8.TP-108 and III.A9.TP-108) identified and listed in the LRA.

While a plant-specific program may not be required if conditions are met, the loss of material (spalling, scaling)

1. Please clarify the air content of the concrete mix used for Groups 1-3, 5, 7-9 structures at PNPP.

2. To facilitate the review, please provide further evaluation on how the absence of concrete aging effects is confirmed by the ASME Section XI, Subsection IWL and Structures Monitoring programs, including how provisions in AMP program will be used to conduct opportunistic inspections of normally inaccessible below grade concrete when excavated for any other reason. Update the LRA accordingly.

3. Please provide applicable Table 2 items associated with AMR item 3.5.1-42 and update the LRA accordingly.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions and cracking due to freeze-thaw is still an applicable aging effect and is subject to AMR.

Per SLR-ISG-2021-03-STRUCTURES, AMP is revised to Plant-specific aging Management program or AMP XI.S6, Structures Monitoring, enhanced as Necessary.

Note:

Per definition, not applicable means:

The component, material, and environment combination does not exist at the plant as described in the SRP-SLRLR/SLR item.

The component, material, and environment combination does exist at the site; however, the component is not within scope or not subject to aging management review.

The aging effect for the component, material, and environment combination is not applicable (generic note I)

The SRP-SLRLR/SLR associates the item with the other reactor type (e.g., BWR only).

2 3.5.2.2.2.1.2 Table 3.5-1 3.5-14 3.5-15 3.5-39

1. GALL-LR recommends that a plant-specific aging management program is not required if, as described in NUREG-1557, investigations, tests, and petrographic examinations of aggregates performed in accordance with ASTM C295 and other ASTM reactivity tests, as required, can demonstrate that those aggregates do not adversely react within concrete.

LRA Section 3.5.2.2.2.1.2 mentioned that petrographic examination was performed on concrete cores from Unit 2 Auxiliary Building per ASTM C856. It is not clear that ASTM C856 meets the intent of "ASTM C295 and the

1. Please explain and demonstrate that PNPP's petrographic examination in accordance with ASTM C856 meets the intent of "ASTM C295 and the other ASTM reactivity tests". Provide the results of petrographic examination for the staff to review.

2. Please provide applicable Table 2 items associated with AMR item 3.5.1-43 and update the LRA accordingly.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions other ASTM reactivity tests" mentioned by GALL in terms of examining aggregate reactivity.

2. LRA Section 3.5.2.2.2.1.2 and AMR item 3.5.1-43 claimed that the potential aging effect (Cracking due to expansion from reaction with aggregates) is managed. However, there are no associated Table 2 items (i.e., III.A1.TP-204, III.A2.TP-204, III.A3.TP-204, III.A4.TP-204, III.A5.TP-204, III.A7.TP-204, III.A8.TP-204, and III.A9.TP-204) identified and listed in the LRA.

Per SLR-ISG-2021-03-STRUCTURES, AMP is revised to Plant-specific aging Management program or AMP XI.S6, Structures Monitoring, enhanced as Necessary.

3 3.5.2.2.2.1.3 Table 3.5-1 3.5-15 3.5-16 3.5-41 3.5-42 3.5-43

1. Section 3.5.2.2.2.1.3 claimed that more than twenty years of settlement and rebound monitoring has revealed minimal settlement. However, "minimal" is a qualitative term and it is unclear what is the quantitative long-term settlements observed by the PNPPs periodic settlement and rebound monitoring.

2. It is unclear whether any porous concrete sub-foundations are used at PNPP for inaccessible concrete areas of Groups 1-3, and 5-9 structures.

3. The applicant claimed that AMR items 3.5.1-44 through 3.5.1-46 are consistent with GALL-LR, and the Structures Monitoring Program manages the aging effects for these AMR Items. However, there are no associated Table 2 items identified and listed in the LRA. While a plant-specific program is not required because a dewatering system is relied upon for control of settlement at PNPP, the associated aging effects are still applicable and are subject to AMR.

1. To facilitate the review, provide quantitative information on the observed long-term settlement.

Compare the observed long-term settlement to CLB allowable value if applicable. Please provide the results of settlement monitoring for the staff to review.

2. Clarify whether porous sub-foundations are used for inaccessible concrete areas of Groups 1-3, and 5-9 structures.

3. Please provide applicable Table 2 items associated with AMR items 3.5.1-44 thru 3.5.1-46. Update the LRA accordingly.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 4

3.5.2.2.2.1.4 Table 3.5-1 3.5-16 3.5-44

1. SRP-LR Section 3.5.3.2.2.1.4 states that a plant-specific aging management program is not required for the reinforced concrete exposed to flowing water if (1) there is evidence in the accessible areas that the flowing water has not caused leaching of calcium hydroxide and carbonation or (2) evaluation determined that the observed leaching of calcium hydroxide and carbonation in accessible areas has no impact on the intended function of the concrete structure.

LRA Section 3.5.2.2.2.4 claimed that a plant-specific program is not required due to the use of durable concrete, non-aggressive below-grade environment, and the absence of concrete aging effects confirmed by the ASME Section XI, Subsection IWL and the Structures Monitoring programs. However, it lacks the evaluation on how the ASME Section XI, Subsection IWL and the Structures Monitoring programs are engaged in conforming the absence of concrete aging effects.

Note:

Per SLR-ISG-2021-03-STRUCTURES, AMP is revised to Plant-specific aging Management program or AMP XI.S6, Structures Monitoring, enhanced as Necessary.

2. In GALL-LR, for the GALL-LR items associated with Item 47 of SRP-LR Table 3.5-1, the applicable environment is indicated as water-flowing.

Section 3.5.2.2.2.1.4 and AMR Item 3.5.1-47 states that the PNPP below-grade environment is not aggressive.

Therefore, increase in porosity and permeability, and loss of strength due to leaching of calcium hydroxide and carbonation of below grade inaccessible concrete areas are not aging effects requiring management for PNPP Groups 1-5 and 7-9 concrete structures. There are no

1. Provide information on how the absence of concrete aging effects is confirmed by the ASME Section XI, Subsection IWL and the Structures Monitoring programs.

Indicate the test procedures used in the IWL and the Structural Monitoring AMPs to detect the presence of leaching. Verify that the accessible and inaccessible below-grade areas at PNPP are free of OE of calcium hydroxide and carbonation leaching due to flowing water. If leaching is evidenced, provide evaluation whether the observed leaching has impact on the intended function of the concrete structure, and justify why a plant-specific program is not required for below-grade inaccessible concrete areas of Groups 1-5 and 7-9 structures.

Update the LRA as necessary.

2. Clarify the applicability of AMR Item 3.5.1-47. If applicable, provide applicable table 2 items associated with AMR item 3.5.1-

47. Update the LRA accordingly.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions associated Table 2 items identified and listed in the LRA.

It is not clear whether it means AMR Item 3.5.1-47 is not applicable to PNPP.

Note:

Per definition, not applicable means:

The component, material, and environment combination does not exist at the plant as described in the SRP-SLRLR/SLR item.

The component, material, and environment combination does exist at the site; however, the component is not within scope or not subject to aging management review.

The aging effect for the component, material, and environment combination is not applicable (generic note I)

The SRP-SLRLR/SLR associates the item with the other reactor type (e.g., BWR only).

5 3.5.2.2.2.3 Table 3.5-1 3.5-17 3.5-45 SRP-LR Section 3.5.3.2.2.3 states that a plant-specific program is not required if documented evidence confirms that the existing concrete had air content of 3% to 8% and subsequent inspection of accessible areas did not exhibit degradation related to freeze-thaw.

1. AMR item 3.5-1, 49 and LAR Section 3.5.2.2.2.3 indicates that the below grade inaccessible concrete areas of PNPP Group 6 structures were constructed in a manner that minimizes the potential for any freeze-thaw aging effects and that the loss of material (spalling, scaling) and cracking due to freeze-thaw are not aging effects requiring management for PNPP Groups 6 structures. PNPP structures are located in a region where weathering conditions are considered moderate, as shown in ASTM C33-90, Figure 1. Therefore, loss of material (spalling,

1. Clarify the applicability of AMR Item 3.5.1-49. If applicable, provide applicable Table 2 items associated with AMR item 3.5.1-49. Please update the LRA accordingly.

2.Evaluate whether a plant-specific program is required.

3.Explain how and what AMP will manage this aging effect in below grade inaccessible concrete areas.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions scaling) and cracking due to freeze-thaw is an applicable aging effect and subject to AMR.

2. Section 3.5.2.2.2.3.1 lacks the evaluation whether a plant-specific program is required for managing this aging effect.

3. Section 3.2.2.2.2.3.1 mentions the Inspection of Water-Control Structures AMP and the Structures Monitoring AMP for certain inspections. It is not made clear how and what AMP will manage this aging effect in below grade inaccessible concrete areas of Group 6 structures.

6 3.5.2.2.2.3 3.5-18 SRP-LR Section 3.5.3.2.2.1.4 states that a plant-specific aging management program is not required for the reinforced concrete exposed to flowing water if (1) there is evidence in the accessible areas that the flowing water has not caused leaching of calcium hydroxide and carbonation or (2) evaluation determined that the observed leaching of calcium hydroxide and carbonation in accessible areas has no impact on the intended function of the concrete structure.

1. In GALL-LR, for the GALL-LR items associated with Item 51 of SRP-LR Table 3.5-1, the applicable environment is indicated as water-flowing Section 3.5.2.2.2.3 and AMR Item 3.5.1-51 states that the PNPP below-grade exterior reinforced concrete is not exposed to an aggressive environment. Therefore, increase in porosity and permeability, and loss of strength due to leaching of calcium hydroxide and carbonation of below grade inaccessible concrete areas are not aging effects requiring management for PNPP Group 6 concrete structures. It is not clear whether Item 3.5-1, 51 is exposed to the water-flowing environment. No associated Table 2 items were identified and listed in the LRA. It is not clear whether this means that the AMR Item 3.5.1-51 is not applicable to PNPP.

Please clarify the applicability of AMR Item 3.5.1-51. If applicable, provide Table 2 items, provide evaluation in accordance with SRP-LR Section 3.5.3.2.2.1.4 and update the LRA accordingly.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: 3.5.2.2.1.1 - Cracking and Distortion due to Increased Stress Levels from Settlement; Reduction of Foundation Strength, and Cracking due to Differential Settlement and Erosion of Porous Concrete Subfoundations (TRP 96)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

3.5.2.2.1.1 Table 3.5-1 3.5-9 3.5-21

1. LAR Section 3.5.2.2.1.1 states that cracking and distortion due to reduction of foundation strength, and cracking due to differential settlement and erosion of porous concrete subfoundation are not aging effects requiring management for the PNPP concrete containment basemat, and AMR Table 3.5-1, Item 2, its claim to be not applicable. In accordance with Item II.A1.C-07 of the GALL -LR, reduction of foundation strength and cracking due to differential settlement and erosion of porous concrete sub foundation should be an applicable aging effect and subject to AMR.

Note:

Per definition, not applicable means:

The component, material, and environment combination does not exist at the plant as described in the SRP-SLRLR/SLR item.

The component, material, and environment combination does exist at the site; however, the component is not within scope or not subject to aging management review.

The aging effect for the component, material, and environment combination is not applicable (generic note I)

The SRP-SLRLR/SLR associates the item with the other reactor type (e.g., BWR only).

2. In LAR Section 3.5.2.2.1.1, the applicant indicated that the absence of concrete aging effect is confirmed by the Containment Inservice Inspection-IWL program. In addition, the applicant stated that differential settlement can affect the safety related structures, but that measured settlement has been minimal. Although the reported settlements might be

1. Please evaluate the claim of non-applicability of AMR Item 3.5.1-2 and provide Table 2 items.

2. Please provide results for the maximum differential settlement and provide clarification on how the absence of concrete aging effect is confirmed by the Containment Inservice Inspection-IWL program.

3. To facilitate the review, please provide results of PNPP below grade environment chemistry.

4. Please provide updates to the LRA and USAR supplement, as appropriate, to be consistent with the response to the above requests.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions minimal, consistent with NUREG-1801, the Structures Monitoring (B.2.3.34) AMP, should be used to manage potential settlement cracking and distortion for the period of extended operation. The quantitative number result for the differential settlement and the management of concrete aging effect is not made clear.

3. LAR Section 3.5.2.2.1.1 indicates that PNPP below-grade is not aggressive. However, in LAR Section B.2.43 Structure Monitoring program the applicant stated that the program will be enhanced to monitor ground water chemistry including accounting for seasonal variations. The quantitative number results for the below-grade water chemistry of PNPP (e.g., ph, sulfate, chlorides) is unclear.

LRA Section: 3.5 - Aging Management of Containments, Structures, and Component Supports (TRP 102)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

Table 3.5.1 3.5-32 For aging management of BWR containment structures penetration sleeves and the suppression pool liner the GALL Report recommends that either (1) cracking due to cyclic loading be managed by the GALL Report AMPs XI.S1, ASME Section XI, Subsection IWE, and XI.S4, 10 CFR Part 50, Appendix J, if a CLB fatigue analysis does not exist (GALL Report items II.B4.CP-37 and II.B2.1.CP-107); or (2) if a CLB fatigue analysis exist, the cumulative fatigue damage needs to be evaluated as a time limited aging analysis (TLAA) in accordance with 10 CFR 54.21(c) (GALL Report items II.B4.C-13 and II.B2.1.C-45).

1. PNPP LAR indicates that it has a CLB fatigue analysis and that this aging effect is addressed under line Item 3.5.1-9 which correspond to option 2 above were a TLAA is

1. Clarify whether a CLB fatigue analysis exist for all listed components (i.e., anchor bolts, component and piping supports, insulation jacketing, Nuts/washers for in-scope anchorage/embedments2) and state the respective TLAA dispositions for these components in accordance with 10 CFR 54.21(c). If there is no CLB fatigue analysis for any of these components, clarify how the associated aging effects will be managed by the Structure

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions performed. PNPP Table 3.5.2 identified components associated with Item 3.5.1-27 and included Structures Monitoring as the aging management program, corresponding Note E and referencing II.B4.CP-37. The LRA is not clear about the aging management for these components and/or its respective TLAA dispositions.

Monitoring program (as stated in components in Table 3.5.2) such that it meets the intent of the GALL report recommended programs (Chapter XI.S1, ASME Section XI, Subsection IWE and Chapter XI.S4, 10 CFR Part 50, Appendix J) and so that the intended function(s) will be maintained consistent with the CLB for the period of extended operation.

2. Please include appropriate changes into the LAR and USAR.

LRA Section: 3.5 - Aging Management of Containments, Structures, and Component Supports (TRP 103)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.55 A.1.5 B.2.5 A.1.43 B.2.43 2.3-173 A-10 B-24 A-42 B-119 LRA Section 2.3.3.55 includes piping and components within the scope of license renewal. It states that [t]he integrity of non-safety-related, fluid-retaining service water piping in the Unit 1 and 2 auxiliary buildings precludes internal flooding that could affect safety-related components. [10 CFR 54.4(a)(2)]

CR-2020-0933 shows photos of degraded Service Water pipe support 2P41H0007. It states that its lower channel, base plate, associated welds are corroded and delaminated. The CR also states that calculations performed indicate that loss of the support would not adversely affect seismic qualification of the Service Water piping or on any safety related safe shutdown SSCs.

It appears that the aforementioned service water supports are within the scope of license renewal to prevent flooding and subject to an AMR. GALL Report, Revision 2, Quality Assurance for Aging Management Programs, states that SCs

1. State if support 2P41H0007 is exempt from examination and if so, why.

2. If not exempt, discuss effectiveness of the current program managing loss of material and other noted aging effects for this/similar supports.

3. Clarify whether support 2P41H0007 (or similar) is within the scope of license renewal. If so, identify how the noted effects of aging are/will be managed for this or supports in similar condition

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions subject to an AMR consistent with Appendix 10 CFR Part 50, Appendix B, Criterion XVI [m]easures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.

If effects of aging for this or similar supports were managed by the Maintenance Program, it is not clear what LRA AMP will manage the effects of aging for this and similar supports during the period of extended operation. If such supports are ASME Class but exempt from ASME Section XI, Subsection IWF ISI and examination per IWF-1230, how are their aging effects are/will be managed so that consistent with 10 CFR 54.21(a)(3) they continue to perform their intended function during the period of extended operation.

during the period of extended operation.

2 A.1.5 B.2.5 A.1.43 B.2.43 A-10 B-24 A-42 B-119 Multiple Condition Reports (CR-2018-08540, CR-2016-08070, CR-2016-11672) discuss boron built up on ASME components and underlying structural supports. Whether the supports are ASME Class and within the scope of LRA AMP B.2.5 (ASME Section XI, Subsection IWF) or other safety-related and within the scope of LRA AMP B.2.43 (Structures Monitoring) they are affected by boric acid deposits. NUREG/CR-5576 Survey of Boric Acid Corrosion of Carbon Steel Components in Nuclear Plants, provides information on boric acid-induced corrosion of ferritic steel components and their bolted/welded connections in NPP systems important to safety and summarizes several NRC generic communications on the subject detailing incidents of galvanic and acid boric corrosion.

GALL Report, Revision 2, Appendix 10 CFR Part 50, Appendix B, Criterion XVI states that [m]easures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and

1. Discuss current condition of boron affected supports.

2. Describe steps taken to eliminate source of boron/boric acid. If deposits are re-occurring discuss whether they are periodically cleaned.

3. discuss effectiveness of the current program managing loss of material and other noted aging effects for such supports.

4. Discuss how PNPP proposed to manage effects of aging associated for boric acid corrosion affected supports so that consistent with 10 CFR 54.21(a)(3) they continue to

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions equipment, and nonconformances are promptly identified and corrected.

It is clear that boron/boric acid deposit on safety related supports is an unacceptable condition. It is not clear, however, what is the current status of the boron affected supports is and whether the supports are cleaned periodically, and how PNPP proposed to manage effects of aging associated with boric acid corrosion so that consistent with 10 CFR 54.21(a)(3) they continue to perform their intended function during the period of extended operation.

perform their intended function during the period of extended operation.

3 A.1.43 B.2.43 A-42 B-119 Several CRs (e.g., CR-2015-04423, CR-2015-10699, CR-2017-00170, CR-201504423, CR-2021-01630, CR-2013-0738, CR-2017-12449) describe conditions that are unacceptable (e.g., severely degraded anchor bolt concrete, anchors not embedded in concrete, grout pad under a column base plate missing, loose/not bolted supports, bent column base plate).

These CRs though show awareness of degraded structures having widespread deficiencies and unacceptable conditions they also demonstrate potential poor maintenance rule and existing structures monitoring program implementations.

LRA AMP B.2.43, Structures Monitoring Program (SMP) states that when [u]nacceptable conditions found, are evaluated, or corrected in accordance with the corrective action program.

For the described conditions and effects of aging, it is not clear what steps PNPP plans to take to improve awareness of such so that their cause is determined, and corrective action(s) addressed timely to preclude repetition or loss of intended function.

1. Discuss adequacy of the SMP (as proposed to be enhanced) to address the CR identified conditions adverse to quality/effects of aging, timely, so that affected structures and components will continue to perform their intended function consistent with 10 CFR 54.21(a)(3) during the period of extended operation.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: B.2.27 - Internal Coatings/Linings for In Scope Piping, Piping Components, Heat Exchangers, and Tanks Program (TRP 12)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

B.2.27 B-84 to B-86 The Program Description for this AMP requires, in part, that:

a) Inspections of coatings/linings will be performed for signs of coating failures.

b) When acceptance criteria are not met, physical testing will be performed (where possible) in conjunction with repair or replacement of coating/lining.

c) The maximum interval of subsequent coating inspections will be consistent with Table 4a of GALL Report AMP XI.M42 in LR-ISG-2013-01.

A Plant Administrative Procedure uploaded to the Portal identifies the following steel tanks which are internally coated with Rustoleum.

PY-1R45A0002A(B) Division 1 and 2 Fuel Oil Storage Tanks PY-1R45A0003(B) Division 1 and 2 Diesel Generator Fuel Oil Day Tanks PY-1R45A0004 HPCS Fuel Oil Storage Tank PY-1R45A0005 HPCS Fuel Oil Day Tank The LRA references one of the above tanks (the Division 2 fuel oil storage tank) in the OpE section of AMP B.2.27. During the audit, NRC staff reviewed inspection records related to the internal inspection of the above tanks.

During the November 29, 2023 breakout session, the applicant stated that PY-1R45A0005 HPCS Fuel Oil Day Tank was inspected in 2010 but proposed that it be considered inaccessible for inspection because it is fully encased in pyrocrete and none of the piping connections provide access to the internal coatings. Also, Work Order 200358401, which was uploaded to the Portal, describes the 2010 inspection of PY-1R45A0005 as being a new PM internal inspection and The staff requests a discussion of the following:

1. Discuss the reason for performing the 2010 inspection of PY-1R45A0005 and explain the phrase new PM inspection which is found in WO 200358401.

2. Explain the rationale for considering PY-1R45A0005 inaccessible for future inspections, since pyrocrete was removed during the 2010 inspection.

3. Describe the applicants proposed alternative to the internal inspection of PY-1R45A0005.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions documents the removal and reinstallation of Pyrocrete during this inspection.

LRA Sections: B.2.21 - Fire Water System Program (TRP 27)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.25 2.3-107 LRA Section 2.3.3.25 states, Traps are provided on floor drains to prevent the spread of an oil fire through the drain piping in areas where oil is contained in the equipment. In addition, this section states, Diesel generator floor drain traps prevent the spread of an oil fire through the drain piping. However, LRA Tables 2.3.3-25 and 3.3.2-25 do not include component types Drain Trap or Trap.

The staff notes that LRA Tables 2.3.3-54 and 3.3.2-54 include drain traps.

LRA Table 2.4-1 defines the Fire Barrier intended function as Provide rated fire barrier to confine or retard a fire from spreading to or from adjacent areas.

Given that LRA Section 2.3.3.25 states floor drain traps are used to prevent the spread of an oil fire, please discuss whether LRA Tables 2.3.3-25 and 3.3.2-25 should include drain traps or traps. In addition, discuss whether these drain traps or traps have a fire barrier intended function.

2 Appendix A and Appendix B

A-24, B-72 Table 4a in Appendix L of LR-ISG-2012-02 recommends sprinkler testing in accordance with Section 5.3.1 of NFPA 25.

Specifically, Section 5.3.1.1.1 covers standard sprinklers in service for 50 years, Section 5.3.1.1.1.3 covers fast-response sprinklers in service for 20 years, and Section 5.3.1.1.1.6 covers dry sprinklers in service for 10 years.

LRA Sections A.1.21 and B.2.21 include an enhancement to the Detection of Aging Effects program element that states, The program will perform representative sprinkler head sampling (laboratory field service testing) or replacement of sprinkler heads within the scope of license renewal prior to 50 years in-service (installed), and at 10-year intervals thereafter, in accordance with the 2011 Edition of NFPA 25, or until there are no untested sprinkler heads that will see 50 years of service through the end of the period of extended operation.

Please discuss whether the enhancement related to testing or replacing sprinklers should be revised to account for different types of sprinklers to be consistent with NFPA 25. In addition, please confirm that sprinklers that are tested and not replaced at the time indicated in NFPA 25 will continue to be tested at 10 year intervals.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions The staff noted that FSAR Section 9A.4.3.3 discussed fast response type sprinklers. Therefore, it is unclear if the enhancement accounts for the different types of sprinklers present at Perry.

It is unclear what the following statement in the enhancement means, at 10-year intervals thereafter, in accordance with the 2011 Edition of NFPA 25, or until there are no untested sprinkler heads that will see 50 years of service through the end of the period of extended operation.

NFPA 25 requires standard sprinklers in service for 50 years be either tested or replaced. If tested, then the standard sprinklers will continue to be tested at 10-year intervals thereafter.

3 LRA Table 3.3.2-24 credits the Fire Water System program for managing loss of material for the steel heat exchanger (Diesel fire pump HX channel) exposed internally to raw water, and loss of material and reduction of heat transfer for the copper alloy

<15% Zn heat exchanger (Diesel fire pump HX tube) exposed internally to raw water.

LRA Sections A.1.21 and B.2.21 include an enhancement to the Detection of Aging Effects program element that states, The program will require at a minimum of once per 10 year interval visual inspection of the engine heat exchanger for the diesel driven fire water pump to monitor for conditions that could cause a reduction in heat transfer.

LRA Section B.2.21 states the Fire Water System will manage reduction in heat transfer for the heat exchanger through periodic visual inspections for fouling and periodic cleaning, as needed.

Given that LRA Table 3.3.2-24 credits the Fire Water System program for managing reduction of heat transfer for the heat Please discuss the following:

What heat exchanger components will be inspected for conditions that could cause a reduction in heat transfer.

The basis for the 10 year interval visual inspection.

Whether inspections of the heat exchanger tubes are practical, including how the inspections will be performed. In addition, how the visual inspections performed by the Fire Water System program will detect loss of material of the heat exchanger tubes.

Whether the enhancement should also include periodic cleaning, as needed.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions exchanger tubes, and the enhancement states visual inspection of the heat exchanger, it is unclear what heat exchanger components are being visually inspected for conditions that could result in a reduction of heat transfer. In addition, the staff notes that one subsequent license renewal applicant stated that inspection of the heat exchanger tube bundle for degradation is not practical due to the small tube diameter (ML21091A187).

Therefore, it is also unclear how the Fire Water System visual inspections will manage loss of material of the heat exchanger tubes.

The staff notes that CR-2022-03813 discusses shells blocking heat exchanger tubes, therefore, the basis for the 10 year interval visual inspection is unclear given the zebra mussel issue at Perry.

Section 2.3.2 of the Fire Water System program basis document (LRPY-AMP-XI.M27) references PMI-0072 for periodic visual inspection and cleaning, when unacceptable conditions are found. However, it appears to be referring to the closed treated water side of the heat exchanger, and doesnt appear to include inspection and cleaning, if necessary, of the raw water side of the heat exchanger.

Therefore, it is unclear whether the enhancement should also include periodic cleaning. The staff notes that the heat exchanger tubesheet was not explicitly identified in LRA Table 3.3.2-24 as a component requiring aging management.

Whether the heat exchanger tubesheet, or any other components associated with the heat exchanger not currently identified, should be included in LRA Table 3.3.2-24.

4 2.3 2.3-103 LRA Section 2.3.3.24 states that water standpipe and hose system may include a hose cabinet. However, LRA Tables 2.3.3-24 and 3.3.2-24 do not include hose cabinet as a component type. The staff notes that fire hose reels are included in LRA Tables 2.4.4-1 and 3.5.2-4.

Please discuss whether LRA Tables 2.3.3-24 and 3.3.2-24 (or LRA Tables 2.4.4-1 and 3.5.2-4) should include hose cabinets.

5 2.3, 3.3, Appendix A,

2.3-105, 3.3-30, 3.3-292, LRA Section A.1.21 states that the Fire Water System program manages loss of material due to erosion. LRA Section B.2.21 states that the Fire Water System program manages loss of material due to erosion and notes a CR related to a severely Please discuss whether the jockey pump casing is included in LRA Tables 2.3.3-24 and 3.3.2-24. Please discuss whether additional

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Appendix B

A-23, B-69 eroded jockey pump casing. The staff also notes that the Fire Water System program basis document (LRPY-AMP-XI.M27) also states the program manages loss of material due to erosion. However, it does not appear that LRA Table 3.3.2-24 credits the Fire Water System program with managing loss of material due to erosion for any components. In addition, it is unclear whether the jockey pump casing is included in LRA Tables 2.3.3-24 and 3.3.2-24 since it is not explicitly stated.

The Discussion for AMR item 3.3.1-126 states that the Flow-Accelerated Corrosion program manages erosion for components in raw water.

LRA Section 3.3.2.1.24 does not identify the Flow-Accelerated Corrosion program as an AMP that manages the effects of aging on the Fire Protection system. In addition, LRA Table 3.3.2-24 does not credit the Flow-Accelerated Corrosion program for managing loss of material due to erosion for any components in the Fire Protection system.

The staff also notes that CR-2017-08251 and WO 200727052 are related to erosion in two locations on a spool piece (with sight glass connection cast into body) downstream of a relief valve due to potential flow through an orifice upstream of two spool pieces.

components (other than the jockey pump casing) are susceptible to erosion. In addition, for the components susceptible to erosion, please discuss which program will manage it (Fire Water System or Flow-Accelerated corrosion).

6 Appendix A,

Appendix B

A-68, B-72 This question is for staff information only. LRA Sections A.1.21 and B.2.21 include an enhancement to the Parameters Monitored and Detection of Aging Effects program elements related to augmented inspections for portions of the Fire Protection system that are normally dry but periodically wetted and cannot be drained or allow water to collect. Specifically, consistent with LR-ISG-2012-02, the enhancement states, In each 5 year interval, beginning 5 years prior to the period of extended operation, a flow test or flush sufficient to detect potential flow blockage will be conducted, or a visual inspection Given that the applicant is already within the 5-year period prior to the period of extended operation, please discuss if the locations that are normally dry but periodically wetted and cannot be drained or allow water to collect have been identified. In addition, please discuss whether a flow test or flush, or a visual inspection of 100 percent of the internal surface of piping segments

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions of 100 percent of the internal surface of piping segments will be conducted.

have been conducted, and what the results were.

7 3.3.2.2.8 3.3-68 LRA Section 3.3.2.2.8 instances of RIC has been identified in the Fire Water System and states the Fire Water System will manage loss of material due to RIC by augmented inspections. In addition, this section states, The program will be enhanced to require that when visual inspections are used to detect loss of material, the inspection technique is capable of detecting surface irregularities that could indicate wall loss to below nominal pipe wall thickness due to corrosion and corrosion product deposition. Where such irregularities are detected, follow-up volumetric wall thickness examinations will be performed. For the buried piping, visual inspections of the piping interior surfaces will be performed whenever the piping internal surface is made accessible due to maintenance and repair activities. In addition, a portion of the aboveground inspection locations will be selected with process conditions similar to those in the buried piping to use as an indicator of the condition of the buried piping.

The staff notes that LRA Sections A.1.21 and B.2.21 include enhancements to the Parameters Monitored and Detection of Aging Effects program elements as described in LRA Section 3.3.2.2.8. The staff notes that these enhancements are consistent with Appendix L of LR-ISG-2012-02 and are not augmented inspections.

Appendix B of LR-ISG-2012-02 recommends that the applicant states, (a) why the programs examination methods will be sufficient to detect the recurring aging effect before affecting the ability of a component to perform its intended function, (b) the basis for the adequacy of augmented or lack of augmented inspections, (c) what parameters will be trended as well as the decision points where increased inspections would be implemented (e.g., the extent of degradation at individual corrosion sites, the rate of degradation change), (d) how Please discuss the following:

(a) why the programs examination methods will be sufficient to detect the recurring aging effect before affecting the ability of a component to perform its intended function (b) the basis for the adequacy of augmented or lack of augmented inspections (c) what parameters will be trended as well as the decision points where increased inspections would be implemented (e.g., the extent of degradation at individual corrosion sites, the rate of degradation change)

(d) how leaks in any involved buried or underground components will be identified

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions inspections of components that are not easily accessed (i.e.,

buried, underground) will be conducted, and (e) how leaks in any involved buried or underground components will be identified.

Therefore, it appears that LRA Section 3.3.2.2.8 does not clearly address RIC as recommended in Appendix B of LR-ISG-2012-02 (a, b, c, and e stated above).

8 Appendix A,

Appendix B

A-67, B-71 LRA Sections A.1.21 and B.2.21 include an enhancement to the Detection of Aging Effects program element that states, The program will be enhanced to include inspections and testing consistent with Appendix L, Table 4a, Fire Water System Inspection and Testing Recommendations, of License Renewal Interim Staff Guidance LR-ISG-2012-02. The staff notes that this enhancement description lacks specificity which may make it hard to verify implementation of the required enhancements in order to be consistent with Appendix L, Table 4a, LR-ISG-2012-02.

The staff notes that testing or replacing sprinklers prior to 50 years in service is consistent with Section 5.3.1 of NFPA 25 which is included in Table 4a of Appendix L in LR-ISG-2012-02.

However, LRA Sections A.1.21 and B.2.21 include an enhancement to the Detection of Aging Effects program element related to testing or replacing sprinklers prior to 50 years in service. Therefore, it is unclear why more specificity was provided with regards to testing or replacing sprinklers prior to 50 years but not with regards to other required enhancements.

Please provide sufficient detail about the enhancements required to be consistent with Appendix L, Table 4a, LR-ISG-2012-02.

9 3.3 3.3-292 The Discussion for AMR item 3.3.1-64 in LRA Table 3.3.1 states, The Fire Water System program will manage loss of material and flow blockage in steel and copper alloy components exposed to raw water in the Fire Protection system. Steel is used here in the broad category of carbon steel, alloy steel, cast iron, gray cast iron, malleable iron, and high-strength low-alloy, steel.

Please discuss whether the Fire Water System program should be credited for managing flow blockage for additional steel and copper alloy components exposed internally to indoor uncontrolled air, outdoor air,

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions The Discussion of AMR item 3.3.1-66 in LRA Table 3.3.1 does not state that the Fire Water System program will manage flow blockage of stainless steel piping components exposed to raw water.

The Discussion for AMR item 3.3.1-130 in LRA Table 3.3.1 states, The Fire Water System program will manage loss of material and flow blockage for metallic sprinkler heads with internal environments of raw water and air - indoor, uncontrolled.

The Discussion for AMR item 3.3.1-131 in LRA Table 3.3.1 states, The Fire Water System program will manage loss of material and flow blockage for steel and copper alloy fire water piping and piping components with internal environments of air - indoor, uncontrolled and air-outdoor.

LRA Table 3.3.2-24 credits the Fire Water System program for managing flow blockage for the following component types:

Steel piping exposed internally to indoor uncontrolled air Steel piping exposed internally to raw water Copper alloy <15% Zn exposed internally to indoor uncontrolled air and raw water However, the staff notes that the Fire Water System program was not credited to manage flow blockage for any components exposed internally to outdoor air. In addition, the staff notes that there are other steel and copper alloy components exposed internally to indoor uncontrolled air, outdoor air, and raw water that the Fire Water System program was not credited for managing flow blockage. For example, valve bodies, tank (Retarding chamber), strainer bodies, pump casings, piping, orifices, flame arrestor.

and raw water in LRA Table 3.3.2-

24.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 10 3.3 3.3-84 Revision 2 of NUREG-1800 includes AMR item 3.3.1-30x for managing cracking, blistering, and change in color due to water absorption of fiberglass, HDPE piping, piping components, and piping elements exposed internally to raw water by the Open-Cycle Cooling Water System program.

The Discussion for AMR item 3.3.1-30x in LRA Table 3.3.1 states, The Fire Water System program will manage change in material properties and blistering of ASTM D-2996 fiberglass reinforced epoxy piping in the Fire Protection system.

LRA Table 3.3.2-24 credits, with a standard note E, the Fire Water System program for managing change in mechanical properties and blistering for fiberglass piping exposed internally to raw water.

LRA Section B.2.21 states that the Fire Water System program manages change in mechanical properties due to blistering of ASTM D-2996 fiberglass reinforced epoxy piping.

Revision 4 of LRPY-MAMR-P54 includes change in mechanical properties and blistering due to water absorption.

Given the noted inconsistencies, the applicable aging effects/mechanisms are unclear.

The staff notes that Volume 1 of NUREG-2191 includes cracking, blistering, loss of material due to ultraviolet light, ozone, radiation, temperature, or moisture; flow blockage due to fouling for fiberglass piping and piping components exposed to raw water. NUREG-2221 states that loss of material was added based on staff in-field observations, change in color was deleted because it has no impact on the intended function, Please discuss how NUREG-2191 was considered with regards to fiberglass piping exposed to raw water. In addition, please clarify the applicable aging effects/mechanisms, including what is meant by change in material (and/or mechanical) properties.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions and flow blockage was added due to the potential intrusion of fouling products from the raw water source.

Therefore, it is unclear whether NUREG-2191 was considered with regards to fiberglass piping exposed to raw water.

LRA Section: B.2.43 - Structures Monitoring Program (TRP 46)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

B.2.43 B-121 LRA Section B.2.43 provides an enhancement to the scope of program to include a list of Unit 1 and Unit 2 structures and structural bulk commodities within the scope of license renewal.

LRA does not make clear what additional structures and components will be added to the current Structures Monitoring program during PEO.

Scope of the Program:

Clarify what additional structures and components will be added to the current Structures Monitoring program during PEO or clarify a list of Unit 1 and Unit 2 structures and structural bulk commodities.

2 B.2.43 B-121 LRA Section B.2.43 provides an enhancement to the scope of program to list the current procedures/instructions that combine to manage the aging effects of the structural elements.

Procedure TAI-0513 Monitoring the effectiveness of maintenance structure monitoring program, Revision 7, shows a listing of procedures, instructions, and repetitive tasks used for monitoring of structures in the attachment 1.

The NRC staff has the following concerns: 1) This enhancement is very vague; 2) Current procedures/Instructions, structural elements and their aging effects are not clear; and 3) This enhancement does not align with the scope of program element based on GALL-LR Report AMP XI.S6 AMP.

Scope of the Program:

1. Evaluate which Element(s) will be enhanced for this enhancement.

2. Clarify procedures and instructions that will combine.

3. Clarify structural elements and their aging effects.

4. Revise the enhancement accordingly.

3 B.2.43 B-121 LRA Section B.2.43 provides an enhancement to the scope of program to monitor the structural integrity of the plant Scope of the Program:

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions underdrain system including the porous concrete sub-foundation.

GALL-LR Report XI.S6 AMP states that the scope of the program includes all structures, structural components, component supports, and structural commodities in the scope of license renewal that are not covered by other structural AMPs.

This enhancement appears to be inconsistent with GALL-LR Report recommendations.

1. Revise this enhancement to ensure the consistency with the GALL-LR Report recommendations.

2. Explain the procedures how to monitor the unground porous concrete sub-foundation.

4 B.2.43 Table 3.5.2-4 B-121 B-122 3.5-160 3.5-161 LRA Section B.2.43 provides an enhancement to the scope of program to monitor the loss of material and flow blockage in the plant storm drain piping, that provides more information than the scope of program, therefore it is in consistent with GALL-LR Report.

LRA Section B.2.43 also provides an enhancement to the parameters monitored inspected and detection of aging effect to monitor the plant storm drain piping for unacceptable flow blockage and observed structural degradation. This monitoring will consist of direct observation (storm drain grating flow and blockage) and periodic remote visual inspections of sampled portions of storm drain system piping.

LAR Table 3.5.2-4 cites Note H for steel, polymer, and concrete storm drain (1) (3) (Total 6 Table 2 items) exposed to raw water, soil environment to manage aging effects of flow blockage, loss of strength, and loss of material by the Structures Monitoring program.

The NRC staff noted inconsistencies of aging effects between each enhancement and Table 2 items.

The NRC staff further noted there is no enhancement to the acceptance criteria for the plant storm drain piping.

Scope of the Program/ Parameters Monitored or Inspected/Detection of Aging Effect/Acceptance Criteria:

1. Clarify aging effects for the plant storm drain piping.

2. Evaluate and revise these enhancements to ensure consistency with each other.

3. Provide acceptance criteria for the plant storm drain piping.

4. Provide layout of the plant storm drain piping, and construction details.

5. Explain how aging effects of the plant storm drain piping are managed by the Structures Monitoring program.

6. Clarify whether Buried and Underground Piping and Tanks program can manage these aging effects for the underground plant storm draining piping.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions The Structures Monitoring program consists of periodic visual inspections. The Buried and Underground Piping and Tanks program manages the loss of material, cracking, and changes in material properties of external surfaces of piping and tanks exposed to a buried environment. The Buried and Underground Piping and Tanks program also manages the aging of the external surfaces of underground piping.

LRA does not make clear how the Structures Monitoring program manages aging effects of underground plant storm draining piping. LRA does not make clear why Buried and Underground Piping and Tanks program is not used for the underground plant storm draining piping.

7. Clarify periodic remote visual inspections and provide the corresponding procedure.

5 B.2.43 B-121 LRA Section B.2.43 provides an enhancement to the scope of program to inspect the in-scope non-safety related/non-seismic masonry walls for loss of material (spalling, scaling),

change in material properties and cracking due to freeze-thaw.

It appears that this enhancement is related to Elements 1 and

3.

LRA Section B.2.43 states that aging effects associated with the in-scope masonry walls are managed by the structures monitoring program as described in the Masonry Walls Aging Management Program Description.

However, LRA Section B.2.43 does not address following items described in the GALL-LR Report AMP XI.S5 AMP: 1) the steel edge supports and steel bracings are considered component supports and aging effects are managed by the Structures Monitoring program; 2) the primary parameters monitored are potential shrinkage and/or separation and cracking of masonry walls and gaps between the supports and masonry walls that could impact the intended function or potentially invalidate its evaluation basis; and 3) for each masonry wall, the extent of observed shrinkage and/or Various Elements:

1. Evaluate the enhancements to the Structures Monitoring program to address the following items: 1) the steel edge supports and steel bracings, 2) potential shrinkage and/or separation and cracking of masonry walls and gaps between the supports and masonry walls, and

3) that the extent of observed shrinkage and/or separation and cracking of masonry may not invalidate the evaluation basis or impact the walls intended function for each masonry wall.

2. Revise LRA Section B.2.42 accordingly.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions separation and cracking of masonry may not invalidate the evaluation basis or impact the walls intended function.

However, further evaluation is conducted if the extent of cracking and loss of material is sufficient to impact the intended function of the wall or invalidate its evaluation basis.

6 B.2.43 B-121 The GALL-LR Report XI.S6 AMP states that the preventive actions for storage, lubricants, and stress corrosion cracking potential discussed in Section 2 of RCSC publication Specification for Structural Joints Using ASTM A325 or A490 Bolts, need to be used If the structural bolting consists of ASTM A325, ASTM F1852, and/or ASTM A490 bolts.

LRA Section B.2.43 provides an enhancement to the preventive actions for storage of high strength bolting (actual measured yield strength greater than or equal to 150 ksi) from Section 2 of Research Council for Structural Connections publication, Specification for Structural Joints Using ASTM A325 or A490 Bolts, which is not consistent with GALL-LR Report recommendations.

GALL-LR XI.S3 AMP states that operating experience and laboratory examinations show that the use of molybdenum disulfide (MoS2) as a lubricant is a potential contributor to stress corrosion cracking (SCC), especially when applied to high strength bolting. Thus, molybdenum disulfide and other lubricants containing sulfur should not be used. This guidance is applicable to the Structures Monitoring program. LAR does not make clear whether PNPP used lubricants containing MoS2 on high strength (actual measured yield strength 150 ksi or 1,034 MPa) structural bolting greater than 1 inch (25 mm) in diameter.

Preventative Actions:

1. Clarify the bolts used and their preventive actions for storage, lubricants, and stress corrosion cracking potential discussed in Section 2 of RCSC publication.

2. Clarify whether PNPP used lubricants containing MoS2 on high strength (actual measured yield strength 150 ksi or 1,034 MPa) structural bolting greater than 1 inch (25 mm) in diameter.

3. Evaluate and revise the enhancement to ensure consistency with GALL-LR Report recommendations.

7 B.2.43 Figure 3.6-6 B-122 3-115 LRA Section B.2.43 provides enhancements to the parameters monitored or inspected and detection of aging effects to include: 1) plant underdrain system maintenance Parameters monitored or inspected and Detection of aging effects:

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions USAR Figure 2.4-68 and 69 7772 and 7773 of 9336 and inspection; 2) monitoring for building settlement; and 3) monitoring the groundwater elevation.

LRA Fig 3.6-6 shows underdrain system map including pumping manhole and gravity discharge manhole. Figure 2.4-68 and 2.4-69 in USAR shows plot plan - porous concrete underdrain system and plot plan-gravity discharge system.

LRA Section 3.6.4.2 states that the groundwater sampling programs at PNPP include a combination of GPI groundwater monitoring wells, piezometers, and manholes to monitor performance of the underdrain system.

LRA Section B.2.43 Operating Experience states the issue of the loss of calcium from the porous concrete and the precipitation of calcium in the porous concrete, porous concrete pipe, and underdrain manholes had been a subject to investigation since 1983 at PNPP, and engineering evaluations were performed.

LRA Table 2.4.2-1 lists foundations, including caissons and porous concrete subject to aging management.

However, LRA does not include the following: 1) the parameters monitored or inspected in the enhancement; 2) an enhancement to Element 6 acceptance criteria; and 3) Table 2 items.

LRA does not make clear whether 12 thick, porous concrete blanket under safety class structures is subject to aging management.

LRA does not make clear how the Structures Monitoring program manages aging effects of underground plant underdrain system. However, LRA Table 2.3.3-44 lists plant

1. Discuss how the plant underdrain system (including pumping manhole and gravity discharge manhole) works.

2. Clarify parameters monitored or inspected for the plant underdrain system, including steel piping, porous concrete pipe, and porous concrete sub-foundation.3. Provide acceptance criteria for the plant underdrain system.

4. Provide Table 2 items for the plant underdrain system.

5. Clarify which AMP(s) will manage aging effects of the underground plant underdrain system.

6. Discuss 12 thick, porous concrete sub-foundation (blanket) and provide engineering evaluations on porous concrete mentioned in the OE under Section B.2.43.

7. Explain how aging effects of the underground plant underdrain system including porous concrete sub-foundation (blanket) are managed by the Structures Monitoring program or other AMPs, including inspection procedures or testing.

8. Clarify whether Buried and Underground Piping and Tanks program can manage these aging effects for the underground plant underdrain system.

9. Evaluate and revise the enhancements accordingly.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions foundation underdrain component types subject to AMR (bolting and piping), and LRA Table 3.3.2-44 indicates that aging effects of steel piping exposed to raw water and soil environment are managed by Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components and the Buried and Underground Piping and Tanks, respectively.

LRA Table 3.3.2-44 does not address how aging effects of porous concrete pipe are adequately managed during PEO.

Procedure No. TAI-0513 Monitoring the effectiveness of maintenance structure monitoring program states that there will be minimum differential settlement between safety related structures due to the foundations bearing on different materials and differential settlement between safety related and non-safety related structures can also affect safety related SSCs.

LRA does not make clear what acceptance criteria are for the building differential settlement and the groundwater elevation.

LAR Section B.2.43 states that building settlement data are collected on a quarterly basis, but LRA does not make clear of its basis for monitoring building settlement quarterly.

10. Discuss the building settlement monitoring, settlement/differential settlement, the reason for quarterly monitoring, and its acceptance criteria.

11. Discuss the groundwater monitoring and its acceptance criteria.

8 B.2.43 B-122 LRA Section B.2.43 provides enhancements to the parameters monitored or inspected and detection of aging effects to require that structures and structural components are monitored on a frequency not to exceed 5 years, which is only related to the detection of aging effects element based on GALL-LR Report.

Detection of aging effects:

Clarify whether Element 3 can be removed.

9 B.2.43 B-122 LRA Section B.2.43 provides an enhancement to specify that a representative sample of high strength (actual measured yield strength 150 ksi or 1,034 MPa) structural bolts greater than 1 inch (25 mm) in diameter are monitored for stress corrosion cracking (SCC) at least once every 10 years.

Parameters monitored or inspected and Detection of aging effects:

Evaluate and revise the enhancement to be consistent with GALL-LR report recommendations.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions GALL-LR Report XI.S6 AMP requires, in general, all structures to be monitored on a frequency not to exceed 5 years.

GALL-LR Report XI.S6 AMP states that visual inspection of high strength (actual measured yield strength 150 ksi or 1,034 MPa) structural bolting greater than 1 inch (25 mm) in diameter is supplemented with volumetric or surface examinations to detect cracking.

There are inconsistencies between the enhancement and GALL-LR Report recommendations.

10 B.2.43 2.1.2.5 Table 2.4.4-1 B-122 2.1-19 2.4-64 GALL-LR XI.S6 AMP states elastomeric vibration isolators and structural sealants are monitored for cracking, loss of material, and hardening.

LRA Section B.2.43 provides an enhancement to require inspection of elastomeric components for cracking, loss of material and hardening, which is inconsistent with GALL-LR Report recommendations.

LRA Section 2.1.2.5 states that structural sealants were determined to be subject to AMR based on their application and are evaluated as bulk commodities in Section 2.4.4.

LRA Section B.2.43 provides an enhancement to Element 6 (acceptance criteria) for the structural sealants.

The NRC staff noted that LRA Tables 2.4.4-1 and 3.5.2-4 do not include elastomeric components, elastomeric vibration isolators, and structural sealants.

Parameters monitored or inspected and Detection of aging effects:

1. Clarify whether elastomeric components, elastomeric vibration isolators, and structural sealants are used at PNPP.

2. Evaluate and revise the enhancement to ensure consistency with GALL-LR Report recommendations.

3. Include these component types in LRA Tables 2.4.4-1 and 3.5.2-4 based on responses above.

11 B.2.43 B-122 LRA Section B.2.43 provides an enhancement to require (a) evaluation of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas and (b) examination of representative samples of the exposed portions of the below grade concrete, when Parameters monitored or inspected and Detection of aging effects:

Clarify whether Element 3 Parameters monitored or inspected

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions excavated for any reason. If normally inaccessible areas become accessible due to planned activities, an inspection of these areas shall be conducted.

The NRC staff noted this enhancement is only for Element 4 Detection of aging effects based on GALL-LR XI.S6 AMP.

can be removed from this enhancement.

12 B.2.43 B-123 LRA Section B.2.43 provides an enhancement to update plant procedures for prescribing quantitative acceptance criteria based on the acceptance criteria of ACI 349.3R and information provided in industry codes, standards, and guidelines including ACI 318, ANSI/ASCE 11 and relevant AISC specifications. Industry and plant specific operating experience will also be considered in the development of the acceptance criteria.

LRA does not make clear what versions will be used for these codes and standards.

Acceptance criteria:

Specify the versions used for the codes and standards.

13 B.2.43 B-123 LRA Section B.2.43 provides an enhancement that acceptance criteria for high strength structural bolting shall be in accordance with the ASME BPVC,Section V, Article 5, Appendix IV.

The NRC staff noted that ASME BPVC, Section, Article 5, Appendix IV does not include acceptance criteria for high strength structural bolting.

Acceptance criteria:

Evaluate and revise code section used for acceptance criteria for high strength structural bolting.

14 B.2.43 B-123 LRA Section B.2.43 provides an enhancement to require that personnel performing inspections and evaluations meet the qualifications specified within ACI Report 349.3R with respect to knowledge of inservice inspection of concrete and visual acuity requirements.

Acceptance criteria:

Specify the version of ACI 349.3R.

15 A.1.43 A-41 FSAR Supplement in SRP-LR Report states that the program consists of periodic inspection and monitoring the condition of structures and structure component supports to ensure that aging degradation leading to loss of intended functions will

1. Revise FSAR Supplement to include the statements recommended by SRP-LR.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions be detected and that the extent of degradation can be determined. This program is implemented in accordance with NUMARC 93-01, Rev. 2 and RG 1.160, Rev. 2.

The NRC staff noted that UFSAR Supplement in LRA Appendix A, Section A.1.43 does not include above statements based on SRP-LR recommendations.

16 Table 3.5.1 3.5-53 AMR 3.5.1-63 is claimed to be consistent with NUREG-1801, however, TRP tool indicates it to be not applicable.

Leaching was identified at multiple locations (e.g. exterior concrete walls of the buildings, interior walls of the basement) during on-site audits.

CR-2022-00531 and CR-2013-08747 show OE related to leaching. Therefore, leaching is applicable aging effect at site.

However, there are no AMR Table 2 items associated with it.

1. Provide Table two AMR items associated with AMR 3.5.1-63.

17 Table 3.2.1 3.2-44 AMR 3.2.1-71 in Table 3.2.1 states that Structures Monitoring program will manage cracking for aluminum components of metal siding, windows, conduits, damper and louver housings and fixed louvers, and flood curb exposed to air-outdoor in the Turbine Buildings and Associated Structures, Process Facilities, and Yard Structures and Bulk Civil Commodities.

However, the NRC staff finds that conduits, damper and louver housings and fixed louvers, and flood curbs are not listed in Table 2.4.2-1.

Table 3.5.2-2 cites Note E for using Structures Monitoring program to manage aging effect for Table 2 items associated with AMR 3.2.1-71. GALL-LR XI.S6 AMP Element 3 does not include cracking due to stress corrosion cracking for aluminum components.

1. Explain why conduits, damper and louver housings and fixed louvers, and flood curbs subject to aging management review are not listed in Table 2.4.2-1.

2. Explain why Structures Monitoring program can be used to manage aging effect of cracking due to stress corrosion cracking for aluminum components.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 18 Table 3.5.1 Table 3.5.2-1 3.5-46 3.5-49 LRA Table 3.5.1 list all concrete areas for all groups except 6:

concrete (accessible areas) for AMR 3.5.1-54. Table 3.5.2-1 list Table 2 item associated with AMR 3.5.1-54 for Containment sump structures 1 only.

AMR 3.5.1-50 in Table 3.5.1 states that accessible Group 6 concrete is monitored for cracking due to expansion from reaction with aggregates by the Inspection of Water-Control Structures AMP and is addressed under Item Number 3.5.1-

54. However, LAR does not make clear where inaccessible Group 6 concrete will be addressed.

LRA does not provide Table 2 items associated with AMR 3.5.1-54 for the concrete areas of all groups except Containment Unit 1.

The NRC staff reviewed procedure TAI-0513 Monitoring the Effectiveness of Maintenance Structures Monitoring Program and did not find any information how to inspect concrete areas related to cracking due to expansion from reaction with aggregates, such 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.

1. Provide Table 2 items associated with AMR 3.5.1-54 for the concrete areas of all groups, including accessible and inaccessible Group 6 concrete.

2. Clarify what Containment sump structures 1 is.

3. Evaluate whether the Structures Monitoring program needs to be enhanced to manage aging effects of cracking due to expansion from reaction with aggregates.

19 Table 3.5.1 Table 3.5.2-2 Table 3.5.2-3 Table 3.5.3-4 3.5-55 3.5-92 3.5-108 3.5-109 3.5-161 LRA Table 3.5.1 claims AMR 3.5.1-69 to be consistent with NUREG-1801. The Structures Monitoring program will manage cracking of high strength structural bolting.

LRA Table 2 items associated with AMR 3.5.1-69 use GALL-LR item III.A3.TP-300 for low-alloy steel high-strength structural bolting exposed to air-indoor, uncontrolled, or air-outdoor environment.

Table 3.5.2-2 lists Table 2 item associated with AMR 3.5.1-69 (row 26) lists galvanized steel diesel fuel tank maintenance structures exposed to soil environment.

1. Address inconsistencies between Table 2 items associated with AMR 3.5.1-69 and NUREG-1801.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Table 3.5.2-3 lists Table 2 items associated with AMR 3.5.1-69 (rows 9 and 12) lists galvanized steel culvert major steam exposed to raw water and soil environment.

Table 3.5.2-4 lists Table 2 item associated with AMR 3.5.1-69 (row 301) lists high strength steel structural bolting exposed to raw water environment.

It appears that component, material, or environment in above mentioned Table 2 items associated with AMR 3.5.1-69 are inconsistent with NUREG-1801.

20 Table 3.5.1 3.5-57 LRA Table 3.5.1 claims AMR 3.5.1-72 to be consistent with NUREG-1801 with clarification including aging effects for elastomer seals susceptible to change in material properties and cracking which claims to be aligned with aging effect of loss of sealing.

LRA Tables 3.5.2-3 and 3.5.2-4 indicate that some Table 2 items associated with AMR 3.5.1-72 list component type:

exterior walls above grade, fire stops, Insulation, roof membrane, seismic isolation joint, shielding, water stops, which are not listed in GALL-LR item III.A6.TP-7.

LRA Tables 3.5.2-3 and 3.5.2-4 indicate that some Table 2 items associated with AMR 3.5.1-72 also list aging effects of change in material properties and cracking, which are not listed in GALL-LR item III.A6.TP-7.

LAR does not categorize Note correctly based on standard notes listed on Page 3.5-165

1. Evaluate and revise the Note accordingly.

21 Table 3.5.1 Table 3.5.2-4 3.5-66 3.5-132 3.5-135 3.5-136 Table 3.5.2-4 cites note E for Table 2 item associated with AMR 3.5.1-93 for component and piping supports 1, which are managed by IWF program. LRA does not make clear what are component and piping supports 1 and where they are located.

1. Clarify what are component and piping supports 1 and where they are located.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Table 3.5.2-4 cites note E for damper and louver housings and fixed louvers 1 and 2, which are managed by Fire Protection program under AMR 3.3.1-132 (GALL-LR item VII.I.A-405 (LRISG-2012-02)) and 3.3.1-59 (GALL-LR item VII.G.A-22) respectively.

Table 3.5.2-4 also cites note E for damper and louver housings and fixed louvers 3, which are managed by Fire Protection program under AMR 3.5.1-93 (GALL-LR item III.B2.TP-6).

It appears that different GALL-LR items are used for damper and louver housings and fixed louvers 1, 2 and 3.

2. Explain why component and piping supports 1 need to be managed by IWF program.

3. Clarify the applicability of GALL-LR item used for damper and louver housings and fixed louvers 3.

22 Table 3.5.2-4 3.5-133 LRA Table 3.5.2-4 cites Note H for polymer conduit caps to manage aging effect of loss of strength along with plant specific notes 518 Loss of strength does not fail the function.

The loading on the cap to perform the function is minor. As long as the cap is visually leak tight, the function is ensured.

It appears that plant-specific notes 518 is very confusing, especially for wording Loss of strength does not fail the function, and does not make clear what acceptance criteria are used for polymer conduit caps to manage aging effect of loss of strength.

Evaluate and discuss plant-specific notes 518 to ensure that aging effects are adequately managed by the Structures Monitoring program.

23 Table 3.5.2-4 3.5-155 3.5-163 3.5-164 LRA Table 3.5.2-4 cites Note H for roof membrane 1 and waterproofing membranes (1) exposed to concrete, soil, and raw water environments, total 4 Table 2 items.

LRA does not make clear how aging effects are adequately managed by the Structures Monitoring program.

1. Describe the locations of these components and their environment.

2. Clarify how these aging effects for these components are managed by the Structures Monitoring program.

24 Table 3.5.2-2 3.5-98 NUREG-1801 Item III.A5.TP-34 lists aging effect of loss of material (spalling, scaling) and cracking due to freeze-thaw.

LRA Table 3.5.2-2 lists Table 2 items associated with AMR 3.5.1-71, citing Note A, for aging effect of change in material properties, which is not included in the GALL-LR Report.

1. Evaluate and revise the Note to ensure the consistency with GALL-LR Report recommendations.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 25 Table 3.5.2-3 3.5-108 3.5-116 3.5-117 LRA Table 3.5.2-3 cites Note H for galvanized steel culvert, major stream, concrete intake/discharge structures, and concrete intake/discharge tunnel exposed to raw water to manage aging effect of flow blockage by the Structures Monitoring program.

LRA does not make clear how aging effects are managed by the Structures Monitoring program.

Major stream, intake/discharge structures, and intake/discharge tunnel are water-control structures, and they are not within the scope of the Structures Monitoring program, it is not clear why the Structures Monitoring program is used instead of the Inspection of Water-Control Structures program.

1. Provide procedures and explain how aging effects of flow blockage are managed by the Structures Monitoring program.

2. Explain why the Structures Monitoring program is used instead of the Inspection of Water-Control Structures program.

3. Clarify whether divers are involved.

26 N/A N/A GALL-LR Report XI.S6 AMP includes jet impingement shields, sump and pool liners, and trash racks associated with water control structures, electrical duct banks, and tube tracks in the scope of program.

The NRC staff noted there are no Tables 1 and 2 items for the above-mentioned components.

Scope of the Program:

1. Clarify whether these components are included in the scope of LR.

2. Provide Tables 1 and 2 items if necessary.

27 Table 3.5.1 3.5-49 LRA Table 3.5.1 claims AMR items 3.5.1-55 and 3.5.1-74 to be consistent with NUREG-1801, and states that the Structures Monitoring program will be used to manage the aging effects.

TRP tool indicates AMR items 3.5.1-55 and 3.5.1-74 to be not applicable.

The NRC staff noted there are no Table 2 items associated with AMR items 3.5.1-55 and 3.5.1-74.

1. Evaluate the applicability of these AMR items and provide Table 2 items for AMR items 3.5.1-55 and 3.5.1-74 if necessary.

28 B.2.43 N/A During the on-site audit, the NRC staff observed concrete spalling at two roof beam seats in the Fuel Handling Building.

The NRC staff would like to find out how they are addressed.

Provide CRs, engineering evaluations, WOs, and corrective actions for addressing concrete spalling.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 29 B.2.43 B-125 CRs (On-site audits - Inspection Item 476, file: Roof leaks from OE report.pdf) show many roof leaks in multiple buildings.

Some roof leaks may impact operating equipment or damage equipment in the building.

LRA Section B.2.43 states that various condition reports were written to document roof leaks in various plant buildings from 2013 through to date, and those corrective actions included either repair of roofs, roof replacements or in planning for planned future repairs.

GALL-LR Report XI.S6 AMP includes provisions for more frequent inspections of structures and components categorized as (a)(1) in accordance with 10 CFR 50.65.

The NRC staff inquires how the Structures Monitoring program adequately monitors roof leaks.

1. Discuss history of roof leaks, leak frequency, repair/ replacement work, and inspection frequency.

2. Discuss which building roof system had been replaced or will be replaced in the future.

3. Evaluate whether more frequency inspections are needed for roof leaks.

4. Evaluate whether there is a need to perform special inspection of roof leaks immediately following the occurrence of significant natural phenomena, such as large floods, earthquakes, hurricanes, tornadoes, and intense local rainfalls.

30 B.2.43 N/A During the onsite audit, the NRC staff observed standing water on roofs of the Fuel Handling Building and Intermediate Building. The NRC staff also observed the dirt and vegetation on roofs. The NRC staff is not clear whether other building roofs have the similar situations.

The NRC staff also noted that some openings were cut in the roof parapets for roof water overflow in the past.

CR-2019-05086 discusses standing water on the Fuel Handling Roof due to roof drain clog.

On the onsite audit item 432 on the ePortal, the applicant states that the building roofs still receive maintenance rule structural inspections performed at a frequency of 4 years, as a result, the current PM strategy is to perform this work as required.

1. Discuss the history of standing water on all the buildings if applicable, and corrective actions taken.

2. Discuss the cause of standing water on roofs.

3. Evaluate roof overflow system to determine whether standing water weight will exceed the current license basis design loading on the roof.

4. Evaluate whether more frequency inspections are needed for standing water on roofs.

5. Evaluate whether there is a need to perform special inspection of the roof drain system immediately following the occurrence of significant natural phenomena, such

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions GALL-LR Report XI.S6 AMP includes provisions for more frequent inspections of structures and components categorized as (a)(1) in accordance with 10 CFR 50.65.

The standing water may cause roof system to deteriorate more quickly, and it may cause roof leaks to impact operating equipment or damage equipment in the building.

The NRC staff inquires how the Structures Monitoring program adequately monitors roof drains and handle standing water on building roofs.

as large floods, earthquakes, hurricanes, tornadoes, and intense local rainfalls.

6. Provide CRs and corrective actions for the standing water on roofs.

31 B.2.43 B-125 LRA Section B.2.43 states that PNPP had reviewed the Shield Building concrete subsurface laminar cracking and determined that the conditions susceptible to laminar cracking are minimal over the majority of the structure, and no actions are required at PNPP to further prevent laminar cracking due to extreme weather conditions.

1. Discuss laminar cracking and provide reports/documents and drawings to support the PNPPs conclusion.

32 Table 2.4.2-1 Table 3.5.2-2 2.4-51 3.5-96 LAR Table 2.4.2-1 lists foundations, including caissons and porous concrete subject to aging management.

LRA Table 3.5.2-2 lists Table 2 items associated with AMR item 3.5.1-65 for concrete foundations and caissons exposed to soil environment, which is managed by the Structures Monitoring program.

USAR Section 2.4.13.5.1 states that caissons under the north end of the service building and under the Fuel Handling Building have been drilled and or blocked-out through the porous concrete blanket into the shale, and other USAR sections discuss caissons for other buildings as well. Caissons are not discussed in GALL-LR Report.

LRA does not make clear where these caissons are located and how their aging effects can be adequately managed by the Structures Monitoring program.

1. Clarify which building(s) is supported by caissons.

2. Describe the construction of caissons.

3. Explain how aging effects of these caissons can be adequately managed by the Structures Monitoring program.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 33 Table 3.5-1 3.5-46 LRA Table 3.5-1 items 50 and 64 claim to be consistent with NUREG-1801. However, there are no associated Table 3.5.2 items identified and listed in the LRA. The associated aging effects are still applicable and are subject to AMR.

1. Please provide applicable Table 3.5.2 items associated with LRA Table 3.5-1 items 3.5.1-50 and 3.5.1-64.

34 Table 3.5.1 3.5-38 LRA Table 3.5-1 item 41 states: NUREG-1801 items referencing this item are associated with Mark I and II concrete and steel containments. Perry's design is a BWR Mark III Steel containment, such that the PNPP design does not contain components associated with a drywell support skirt.

However, there are several Table 3.5.2 items associated with Item 3.5.1-41.

1. Clarify the need for the Table 3.5.2 items associated with Table 3.5.1 Item 41 when a claim of non-applicability was made.

35 3.5.2.2.1.1 2.3.3.44 3.5-8 2.3-148 LRA Section 3.5.2.2.1.1 states that PNPP does not rely on a dewatering system for control of settlement, but that plant substructures were designed with porous concrete and a permanent underdrain system to reduce hydrostatic pressure.

No associated Table 3.5.2 items were identified and listed in reference to Table 3.5.1 item 1. The associated aging effect are still applicable to the basemat and are subject to AMR.

LRA Section 2.3.3.44 describes the plant foundation underdrain. It indicates that the underdrain system consists of a porous concrete blanket which underlies all of the structures of the nuclear island. In addition, it states that the main objective of the pressure relief underdrain system is to ensure that the groundwater level around the nuclear island does not exceed elevation 590.0 feet.

USAR Section 2.4.13.5.5 Safety Evaluation makes reference to a permanent dewatering system.

For several items in LAR Table 3.5.1 (e.g., Items 1,2,44,45 and

46) the applicant claimed that PNPP does not rely on a dewatering system to control settlement.

1. Please provide applicable Table 3.5.2 items associated with Table 3.5.1 Item 1.

2. Please clarify the purpose of the permanent dewatering system mentioned in USAR Section 2.4.13.5.5 and if it is in fact used to control settlement given that it is used to control the groundwater level (as stated in LAR Section 2.3.3.44) and depending on the properties of the foundation material, moisture can cause swelling and subsequent settlement issues.

3. If dewatering system is relied on to control settlement and settlement caused by erosion of porous concrete blanket under the structures of the nuclear island, please provide further evaluation on how to ensure proper functioning of the dewatering

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions system through the period of extended operation.

LRA Section: B.2.22 - Flow Accelerated Corrosion Program (TRP 18)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

B.2.22 BWR VIP-205 Bottom Head Drain Line (BHDL) Inspection and Evaluation Guidelines shows Perry in the lowest susceptibility category based on never operating on hydrogen water chemistry. However, Perry switched to hydrogen water chemistry in the past. The reactor BHDL is part of the Reactor Water Clean-Up (RWCU) system, and recent inspections of RWCU piping at Perry have shown high wear rates. Although access to the BHDL is limited, other utilities have performed FAC inspections on this piping. Industry discussions about recent RWCU events note that a direct cause was implementation of HWC that resulted in low dissolved oxygen levels.

Discuss implications of changing to hydrogen water chemistry as it relates to the need to consider inspection of the reactor vessel bottom head drain. Based on accelerated wear rates seen in part of the RWCU system, because the bottom head drain line ties into the RWCU system, how is this being considered for the predicted wear rates in the bottom head drain line and the need to do an inspection on the line?

What portions of the RWCU (including the bottom head drain piping) system, have had trace chromium content measured?

LRA Section: B.2.37 - Open-Cycle Cooling Water System Program (TRP 20)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

B.2.37 B-104 The LRA states annual diving inspections have effectively ensured the intake and discharge structures continue to control mussel contamination within acceptable limits. No evidence of Asian clams has been identified. There have been no documented instances of mussel infestation in ESW system piping or safety-related heat exchangers served by the system. The lack of documented instances The listed CR appears to contradict the statement that zebra mussel infestation in the ESW system piping or safety-related heat exchangers has not been documented at Perry.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions ePortal:

indicates that the existing program has been effective in managing aging effects due to biofouling and silting in structures and components serviced by OCCW systems.

CR-2022-03813 (5/3/2022) - Division III Emergency Diesel Generator Jacket Water Heat Exchanger Inspection Unsatisfactory - During the performance of work order 200876542 Operation 0130, the heat exchanger was found to be in unsatisfactory condition. There are 72 tubes that are more than 50% blocked, primarily by zebra mussel shells.

The staff requests discussion on the cited CR as to why it does not provide evidence that zebra mussel infestation has occurred.

2 B.2.37 B-105 ePortal The LRA discusses a CR from February 2022 that documented a degrading trend in the Division III Diesel Generator Jacket Water Heat Exchanger performance. A review of operating experience CRs on the ePortal revealed CR-2022-00950 (Feb-08-2022), which stated that in July 2019, the DGJW HX heat transfer coefficient was 339.18 Btu/hr-ft2-°F and the acceptance criteria was 320.1 Btu/hr-ft2-

°F.

LRPY-AMP-XI.M20 R4 states in Section 2.5.2 that various heat exchanger performance monitoring test results are evaluated to identify components that need cleaning, repair, replacement, or other corrective measure to ensure no loss of intended function. This document then references EMARP-0008 as a guide for processing and trending samples from GMI-0023 or PTI-GEN-0023.

In EMARP-0008, Section 4.7 Surveillance Testing Trending directs the system engineer to receive the monthly surveillance testing pressure/flow data and process it by graphing the data versus time for trending purposes. If the data indicates that Corbicula or Dreissena may be present, request that the system be inspected by Operations, and

1. Discuss the Design Interface Evaluation per NOP-CC-2004.

What does the NOP direct to happen, by whom, when, and how often.

2. CR-2022-00950 shows the trend of the last five performance tests.

According to the time of the last test (Rev 4) and the date the CR was written, it appears that it took 31 months to document the degraded trend in heat exchanger performance. Please clarify if this is accurate and discuss the timeline of events and actions that took place.

3. After the degraded trend was identified in July 2019, how much time passed before the Division III DGJW heat exchanger was cleaned?

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions submit the trend analysis data to the Engineering Records Center for inclusion in Technical Assignment File 80245.

Calcs E22-042, Rev 3 (for Oct-2014 test) and Rev 4 (for Jun-2019 test) both state that the results are sent to the GL 89-13 program owner (via Design Interface Evaluation per NOP-CC-2004) and are used to support trending of heat exchanger performance.

Based on NRC staff evaluation, it appears that the acceptance criteria of 320.1 Btu/hr-ft2-°F would have been exceeded in December 2021, approximately two months before CR-2022-00950 was written. The NRC staff did not see any documentation or discussion in the corrective action documents about the need to perform trending of test data in a timelier manner, to prevent acceptance criteria from being exceeded.

4. Was the heat transfer coefficient recalculated shortly before the heat exchanger was cleaned? If so, what value was obtained? If not, how do you know the heat transfer coefficient of the heat exchanger didnt drop below the minimum acceptance criteria?

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: B.2.18 - External Surfaces Monitoring of Mechanical Components Program (TRP 37)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

B.2.22 OpEx discussion related to failures of stainless steel flexible hoses for Safety Relief Valves.

LRPY-OE-001 includes CR-2021-2296 documenting a leak on an air supply hose for 1B21F051B and cites cracking due to stress corrosion cracking (SCC) The OE document assigns the OpEx to XI.M36 and cites AP-209, which is associated with SRP-LR Further Evaluation 3.3.2.2.3. CR notes that cause of leak is not yet known and failure analysis by vendor should be able to determine cause definitively.

CR-2021-2192 discusses flex hose leak for B21F0041B and states suspected cause would be cyclic fatigue. LRPY-OE-001 says LoM and assigns to Compressed Air AMP.

Both of the above CRs discuss CR-2017-03192 that documents a leak in the flex hose to 0041B in their OpEx section. LRPY-OE-001 does not include the 2017 CR. Both of the 2021 CRs discuss sending hoses out to vendor for destructive analysis.

CR-2023-02318 and CR-2023-02255 say corrective action was written to send out flex hose to a vendor for a destructive analysis.

LRA Table 3.1.2-2 includes stainless steel and nickel alloy flexible hose. Both materials show no aging effects requiring management, and for the internal environment cite Compressed Air Monitoring for managing the aging effects

1. Discuss why CR-2017-03192 was not included in the table of CRs.

2. Discuss why LRPY says the aging effect is SCC, with an associated AMR as AP-209 (which is associated with SRP-LR 3.3.2.2.3 for SCC) yet that section of the LRA says there is no SCC.

3. What did failure analyses determine from the 2021 failures?

4. What did failure analysis determine from the 2023 failures?

5. Based on OpEx, should the flex hoses have an aging effect requiring management?

2 B.2.18 Program description says this is a new program.

The OpEx Section of LRPY-AMP-XI.M36 discusses several condition reports that identified degradation. The section

1. Does Perry have a comparable system walkdown procedure similar to Davis-Besse and are system walkdowns being

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions states: The operating experience shows that routine walkdowns and inspections do identify degradation Davis-Besse (also an Energy Harbor, formerly FENOC plant)

LRA stated that the External Surfaces Monitoring program is an existing program that will be consistent with XI.M36 with enhancements. The Davis-Besse AMP audit (ML11122A014) cites FENOC System Performance Monitoring Program, NOBP-ER-3003 and System Walkdown Checklist NOBP-ER-3009.

routinely performed? If not, discuss why not.

2. The OpEx discussion in B.2.18 appears to say that routine walkdowns are being done. If walkdowns are being performed, then explain why this is considered a new program and not an existing program.

LRA Section: 2.4 - Scoping and Screening Results: Structures (Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.4 2.4-1 LRA Section 2.4 states that this section provides the following information for the Perry structures within the scope of license renewal listed in Table 2.2-2.

However, Table 2.2-2 shows nonradiative Waste Quantities at PNNP.

Provide correct table number.

2 2.4 N/A Perry LR site map indicates that Unit 1 Circulating Water Pump House is in the scope of LR.

LRA Table 2.2-6 states that Circulating Water Pump House is not within the scope of license renewal.

There is discrepancy between Perry LR site map and LRA Table 2.2-6.

Clarify whether Unit 1 Circulating Water Pump House is within the scope of LR and subject to aging management. If yes, provide Tables 1 and 2 items.

3 Figure 3.6-7 Table 2.2-6 3-116 2.2-22 LRA Figure 3.6-7 shows onsite wells.

LRA Section 2.4 does not discuss onsite wells. In addition, LRA Table 2.2-6 does not include onsite wells in the Structures and Structural components not within the scope of License Renewal.

Clarify whether onsite wells are within the scope of LR and subject to aging management. If yes, provide Tables 1 and 2 items.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LAR does not make clear whether onsite wells are within the scope of LR and subject to aging management.

LRA Section: 2.3.3.25 - Floor and Equipment Drains (Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.25 ePortal:

Page 2.3-113 License Renewal Drawing 911-0601 (Coordinates H-7 & H-

10) displays two Drywell sump liners as not subject to AMR.

Table 2.3.3-25 Floor and Equipment Drains Component Types Subject to Aging Management Review does not list a Component Type of Sump Liner. For the Component Type Pump Casing an Intended Function of Pressure Boundary is not listed TS 3.4.7 RCS Leakage Detection System LCO a. Drywell floor drain sump monitoring system indicates that the RCS Leak Detection instrumentation shall be OPERABLE with an APPLICABILITY: MODES 1. 2. and 3.

It is not apparent that the long-term integrity (pressure boundary function) of these liners and the sump pump casings (i.e., pumps not shown on drawing) are not relevant to plant safety with respect to RCS leakage rate measurements within the Drywell during power operations.

Is the drywell floor drain sump monitoring system operability dependent on the integrity of these sump pump components.

Please provide justification of why these components are not color coded as red per 10 CFR 54.4 a(2) functional support requirements and subject to aging management review.

2 2.3.3.25 Page 2.3-112

& 2.3.113 ePortal:

LRA Section 2.3.3.25 reads in part Prevent flooding of the ECCS rooms by backflow through the floor drains. [10 CFR 54.4(a)(2)] and Auxiliary Building floor drains on the 599' and 620' elevation mitigate internal flooding. [10 CFR 54.4(a)(2)].

License Renewal Drawing 911-0617 reads in part Note: 599

& 620 Auxiliary Building floor drains, and flow paths mitigate internal flooding Verify that Valves G51-F576 thru F582 in ECCS Pump Room Sumps should be a(2) functional assuming these valves are N.C. valves controlled by Administrative Procedures.

OR

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions From review of Drawing 911-0617 it is not apparent that common mode failure of the ECCS pumps due to room flooding via backflow of internal flood water from the higher elevations of the Auxiliary Building is not possible.

Provide on the Curtis Wright eportal for staff review an internal flood analysis that addresses this concern.

3 2.3.3.25 ePortal:

Page 2.3-113 License Renewal Drawing 911-0671 (Coordinates E-8 through E-12) the floor drain, piping and four isolation valve bodies in the Control Room Concrete Floor EL. 653-6 may have a Control Room Envelope pressure boundary function (i.e., not just a 10 CFR 54.4a(2) Spatial as displayed on this drawing).

Table 2.3.3-25 Floor and Equipment Drains Component Types Subject to Aging Management Review does not list for the Component Type Valve Body an Intended Function of Pressure Boundary.

Please review for accuracy of function displayed and intended function of valve body on LRA Table 2.3.3-25 Floor and Equipment Drains Component Types Subject to Aging Management Review.

In addition, these four isolation valves appear to have a Fire Protection CLB Function (i.e., may warrant a color code of RED for this regulated event.)

LRA Section: 2.3.3.26 - Fuel Handling Area Ventilation (Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.26 ePortal:

Pages 2.3-116 2.3-153 Renewal Drawing: 912-0617 Coordinates A-9, G-12 et.al.

Flow straighteners are displayed on drawing but are labeled as Flow Elements.

LRA Table 2.3.3-26 Fuel Handling Area Ventilation Component Types Subject to Aging Management Review does not list a Component Type Flow Straightener. Rather it lists the Component Type Flow Element.

In contrast Table 2.3.3-45 Plant Radiation Monitoring and Process Monitoring and Post Accident Radiation Monitoring Component Types Subject to Aging Management Review lists a Component Type of Flow Straightener.

Please discuss with the staff the reason for a perceived lack of consistency.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 2

2.3.3.26 ePortal:

License Renewal Drawing: 912-0617 Coordinate A-9 displays a pointer to another drawing (number is unreadable) with a destination of TO UNIT 1 VENT. What is the drawing number?

Is there radiation monitor on this referred to drawing before the effluent is discharged to the environment? What is the end boundary of the M40 system?

What is the drawing number? Is there radiation monitor on this referred to drawing before the effluent is discharged to the environment? Where is the end boundary of the M40 system?

3 2.3.3.26 Page 2.3-114 ePortal:

LRA Section 2.3.3.26 System Description reads in part: Two barometric pressure relief dampers in the supply duct would relieve any excessive negative building pressure. This is a 10 CFR 54.4(a)(1) CLB function.

Neither review of the Aging Management Review (AMR)

Report for the System M40 on the Curtis Wright MetaStor' eportal nor review of the License Renewal Drawing 912-0617, provided any information as to the location of these dampers.

The staff requests additional information about the location, materials of construction, and the internal and external environments for these two barometric pressure relief dampers.

LRA Section: 2.3.3.29 - Inclined Fuel Transfer System (Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.29 ePortal:

License Renewal Drawings: 302-0972 & 302-0973 Coordinates B-5 through B-12; the three cylinders are NOT identified as a Leakage Boundary Spatial concern.

Staff requests additional information as to the reason for this.

2 2.3.3.29 ePortal:

Page 2.3-123 License Renewal Drawing: 302-0970 Coordinates B-11 through D-13 is color coded green predominantly (if not all) as providing a structural support function.

LRA System Description reads in part The Unit 2 transfer tube, bellows and associated piping components are retired in place, but portions still provide a leakage boundary function.

LRA Table 2.3.3-29 Inclined Fuel Transfer System Please discuss with the staff and/or provide additional information to further staff s understanding.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Component Types Subject to Aging Management Review for the Component Type Bellows (Transfer Tube) does not list either a Leakage Boundary function or Structural Support function.

LRA Section: 2.3.3.32 - Leak Detection (Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.32 ePortal:

Pages 2.3-128 2.3-134 License Renewal Drawing 302-0961, Coordinate H-12:

Drywell Air Cooler Leak Detector Drain Line Flow Element FE-N021 is not identified as a component subject to AMR even though the drain piping to/from is color coded blue as having a leakage boundary spatial function.

This observation is consistent with LAR Table 2.3.3-32 Leak Detection Component Types Subject to Aging Management Review in that the table does not list the Component Type Flow Element.

In contrast the flow elements displayed with drawing symbol of Flow Straightener on License Renewal Drawing for the 912-0609 are color coded red as being subject to AMR are displayed on LRA Table 2.3.3-35 MCC Switchgear and Miscellaneous Electrical Area HVAC, and Battery Room Exhaust Component Types Subject to Aging Management Review as Component Type Flow element.

Please discuss with the staff and/or provide additional information to further staff s understanding of this apparent inconsistency.

LRA Section: 2.3.3.33 - Liquid Radwaste Disposal (Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.33 Page 2.3-130 LRA Section 2.3.3.33 subsection System Functions (and scoping criteria, if intended function) reads in part: Nonsafety-Please discuss with the staff and/or provide additional information to

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions ePortal:

Pages 2.3-128 2.3-134 related piping up to and including the first equivalent anchor beyond safety/nonsafety interface(s) provides mechanical support for safety-related SSCs.[10 CFR 54.4(a)(2)]

License Renewal Drawings: 302-0731, 302-0733, 302-0734, 302-0736, 302-0737, 302-0738:

The staff notes that none of identified G50 components as subject to AMR are color coded light green to indicate a structural integrity intended function - there were several build interfaces hard stopped dark green to identify the transition to a non-safety building.

Piping External Environment of Concrete is identified in the AMR report for System G50 as being displayed on LR Drawing 302-0737 and appears to be associated with Containment penetration 42B.The piping on either side of this penetration is not color coded as light green to indicate a structural integrity intended function.

further staff s understanding of this inconsistency between LRA Section 2.3.3.33 and the License Renewal Drawings.

LRA Section: 2.3.3.34 - Liquid Radwaste Sumps (Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.34 ePortal:

License Renewal Drawing 301-0739, Coordinate C-12 Drywell Equipment Drain Sump liner. Same issue as Issue #1 for the Floor and Equipment Drains (P68) above.

Note that UFSAR Section 11.2.10.k Sumps reads in part:

With the exception of those sumps that are normally nonradioactive, all sumps are lined with stainless steel for leakage control and to facilitate decontamination.

Is the drywell floor drain sump monitoring system operability dependent on the integrity of these sump pump components.

Please provide justification of why drain pump liner is not color coded as red per 10 CFR 54.4 a(2) functional support requirements and subject to aging management review.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 2

2.3.3.34 Page 2.3-132 ePortal:

LRA Section 2.3.3.34 System Functions (and scoping criteria, if intended function) reads in part Nonsafety-related piping up to and including the first equivalent anchor beyond safety/nonsafety interface(s) provides mechanical support for safety-related SSCs. [10 CFR 54.4(a)(2)].

License Renewal Drawings 302-0739; 302-0740; and 302-0741 do not color code (light green) any sections of piping subject to AMR that the LRA section criteria applies to. These drawings do contain building interfaces hard stopped dark green to identify the transition to a non-safety building per drawing 302-001 Legend. But this does not satisfy the need to identify the relevant section of Nonsafety-related piping up to and including the first equivalent anchor per the criteria above.

Moreover, the inboard and outboard piping sections connected to the Containment Isolation valves MCV-F075 and MCV-F080 at Containment Penetration #417 on drawing 302-0739 should be color coded light green out to first equivalent anchor to ensure the structural supports for these piping sections are subject to aging management regardless of the fluid contained (e.g., gas, air, or leakage boundary spatial for fluid).

Please discuss with staff the apparent inconsistency between the License Renewal Drawings versus the cited criteria of LRA Section 2.3.3.34.

LRA Section: 2.3.3.36 - Miscellaneous Area Ventilation (Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.36 ePortal:

License Renewal Drawing: 912-0629 Coordinate G-5; Component Fixed Louver (no component number assigned) is displayed as not being subject to AMR. LRPY-CAMR-CMD, Revision 13 does not reference License Renewal Drawing:

912-0629.

Please confirm that this louver is contained within the aging management program for Structural Bulk Commodities (LRA Section 2.4.4).

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions This fixed louver appears to provide an Intended Function (SRE) of a(3) similar to the exhaust piping contained therein (i.e., Fire Pump room exhaust piping).

FSAR 9.4.12.2.6 refers to this louver as a ducted relief louver.

LRA Section: 2.3.3.37 - Miscellaneous Electrical Areas Smoke Ventilation (Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.37 ePortal:

Page 2.3-136 License Renewal Drawing: 912-0629 Coordinate C-2; Component F0059 has an associated TD label.

From review of the Aging Management Review Report for System 49 Miscellaneous Electrical Areas Smoke Ventilation, Revision 2, it does not appear that Component PY-1M49F0059 is addressed by the AMR report.

This appears to be a tornado damper and relevant to the first sentence of the System Description of LRA Section 2.3.3.37 which reads in part With the exception of the tornado dampers and associated ducting, the miscellaneous electrical areas smoke venting system is nonsafety and functions only when a high smoke condition occurs in the electrical areas Please clarify where the aging management of the subject safety related tornado damper(s) is addressed in the LRA.

LRA Section: 2.3.3.40 - Nuclear Closed Cooling (Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.40 ePortal:

License Renewal Drawings 302-0612 and 352-0612; Instrument Air Dryers for Unit 1 vs Unit 2 are not color coded consistently as a(2) leakage boundary spatial. Unit 1 dryers are not color coded at all.

Please discuss with the staff as to why this difference exists.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 2

2.3.3.40 ePortal:

Page 2.3-143 License Renewal Drawings 302-0611; 302-0612; 302-0613; and 352-0612: Flow elements (orifice type) throughout these drawings are color coded as a(2) leakage boundary spatial.

However, LRA Table 2.3.3-40 Nuclear Closed Cooling Component Types Subject to Aging Management Review does not contain a line item for the Component Type Flow Element.

Please discuss with staff as to why the Component Type Flow Element is not listed in LRA Table 2.3.3-40.

3 2.3.3.40 ePortal:

License Renewal Drawing 302-0611, Coordinate 5-D displays a Level Gauge (LG0298) for the Nuclear Closed Cooling Surge Tank as being subject to AMR for its a(2) Leakage Boundary Spatial Intended Function. What type of level gauge is this (i.e., sight glass)? If it is a sight glass, material Glass does not appear in the AMR Report for System P43 Nuclear Closed Cooling Revision 3.

The staff requests confirmation the LF0298 is accurately represented in the aging management programs for the plant.

4 2.3.3.40 Page 2.3-142 ePortal:

LRA Section 2.3.3.40 System Functions (and scoping criteria, if intended function) reads in part Nonsafety-related piping up to and including the first equivalent anchor beyond safety/nonsafety interface(s) provides mechanical support for safety-related SSCs. [10 CFR 54.4(a)(2)].

License Renewal Drawings 302-0611; 302-0612; 302-0613; and 352-0612 do not color code (light green) any sections of piping subject to AMR that the above criteria could apply to.

Drawing 352-0612 does contain a building interface hard stopped dark green to identify the transition to a non-safety building per drawing 302-001 Legend. However, this does not satisfy the need to identify the relevant sections of Nonsafety-related piping up to and including the first equivalent anchor per the criteria above.

The staff request clarification of this apparent discrepancy between the LRA and the License Renewal Drawings.

5 2.3.3.40 ePortal:

License Renewal Drawing Renewal Drawing 302-0613: the inboard and outboard piping sections connected to the Containment Isolation valves and Drywell Isolation valves are The staff request confirmation that the subject structural supports are subjected to aging management.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions not color-coded light green out to first equivalent anchor to ensure the structural supports for these piping sections are subject to aging management.

The staff notes that regardless of the fluid contained (e.g.,

gas, air, or leakage boundary spatial for fluid) these structural supports are germane as being scoped within LR and subject to aging management. More specifically, if the fluid were air or gas, a leakage boundary spatial intended function would not be assigned but the structural supports for such lines would still need to be maintained per the aging management program.

LRA Section: 2.3.3.41 - Offgas Building Ventilation (Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.41 ePortal:

Page 2.3-145 License Renewal Drawing 912-0622, Coordinates C-4 thru D-4 (Filter Train 1M36D001A) and C-6 thru D-6 (Filter Train 1M36D001B) displays instrumentation piping and valves attached to the filter train housing pressure boundary and to the flow elements from each filter train.

This instrumentation piping is a continuation of the pressure boundary. However, this instrumentation piping and valves are not displayed as being subject to AMR.

The staff acknowledges that Note 2 on the drawing reads Nonsafety-related piping does not provide support to base mounted ventilation units or ductwork components.

LRA Table 2.3.3-41 Offgas Building Ventilation Component Types Subject to Aging Management Review does not list an Intended function of Pressure Boundary for neither Component Type Piping nor Valve Body.

Please discuss with the staff as to why this piping and these valves need not be subjected to aging management during the period of extended plant operations.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 2

2.3.3.41 ePortal:

Page 2.3-143 License Renewal Drawing 912-0622, Coordinate D-4 (Filter Train 1M36D001A) and Coordinate D-6 (Filter Train 1M36D001B) displays drain piping and isolation valve as subject to AMR due to providing a Leakage Boundary spatial Intended Function.

However, during operation of the filter trains it appears that the piping to the solenoid-controlled isolation valves along with the valve body would provide an intended function of Pressure Boundary.

Please confirm to the staff that this piping and these valves perform no Pressure Boundary Intended Function during operation of the filter trains.

3 2.3.3.41 ePortal:

Page 2.3-145 License Renewal Drawing 912-0622, Coordinate D-7:

Instrumentation piping to System D17 radiation detector is a continuation of the pressure boundary of the M36 Duct to which it is attached.

However, this piping is not displayed as being subject to AMR for neither its Pressure Boundary Intended Function nor its required Structural Support Intended Function.

LRA Table 2.3.3-41 Offgas Building Ventilation Component Types Subject to Aging Management Review does not list an Intended function of Pressure Boundary for Component Type Piping.

Please confirm to the staff that this instrumentation piping performs no Pressure Boundary Intended Function during plant operation.

4 2.3.3.41 ePortal:

Page 2.3-145 License Renewal Drawing 912-0622 displays 15 to 20 pitot tube test connection symbols.

The term pitot tube or pitot does not appear in the LRA.

Table 2.3.3-41 Offgas Building Ventilation Component Types Subject to Aging Management Review does not contain a line item for Component Type Pitot tube or Pitot tube test connection.

LRA section 2.3.3.41 reads in part: Components located below elevation 660 of the offgas building are designed to satisfy system space requirements and to satisfy the requirements for Safety Class 3 and Seismic Category I Please confirm that the Component Type Duct in LRA Table 2.3.3-41 envelopes the term Pitot tube test connection and that these ports are subject to AMR.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions items. Please clarify whether the Component Type Duct in LRA Table 2.3.3-41 envelopes the term Pitot tube test connection and that these ports are subject to AMR.

5 2.3.3.41 ePortal:

License Renewal Drawing 912-0622, Coordinate D-4 (Filter Train 1M36D001A) and Coordinate D-6 (Filter Train 1M36D001B) displays piping from the P54 Fire Protection system providing fire suppression to the charcoal beds of each Filter Train. However, this piping is not displayed as being subject to AMR.

It appears that the integrity of this piping would provide a Pressure Boundary function and/or could provide Leakage Boundary Spatial function to Safety Related components within the Off-Gas Building. The integrity of this piping may also perform a 10 CFR 50.48 Fire protection regulated event function.

Please discuss this observation with the staff.

6 2.3.3.41 Page 2.3-144 UFSAR App 9A.4.13.1 reads in part:

The ventilation system for the offgas building consists of supply plenums and supply fans blowing cooled outdoor air to various areas.

Revision 9A.4-213 December, this supply air is discharged to the atmosphere by the exhaust fans.

UFSAR 9A.4.13.12 Analysis reads in part:

Combustibles in the offgas building include charcoal and hydrogen gas. Special consideration was given to the charcoal filters and to a possible explosive hydrogen mixture, as hazards in this building. The charcoal filters are provided with heat sensors that initiate signals in the control room so that the deluge system can be manually actuated. The components and piping for the offgas system up to the Please provide additional information as to why these functions are not relevant to the LRAs M36 Intended Functions and the aging management of the plant during the period of extended plant operation.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions recombiners are designed to withstand a hydrogen explosion. The ventilation system supplies sufficient circulation of room air so that any hydrogen leakage will be limited to levels below 4 percent by volume hydrogen concentration.

The staff notes that the Off-Gas Building is a safety related structure. LRA Section 2.3.3.41 reads in part:

Components located below elevation 660 of the offgas building are designed to satisfy system space requirements and to satisfy the requirements for Safety Class 3 and Seismic Category I items.

Neither LRA Section 2.3.3.41 nor License Renewal Boundary drawings 302-0751, 302-0752, 912-0622 provide information about the outside air supply side of the Off-Gas Ventilation System. The system intended function of limiting hydrogen leakage to levels below 4 percent by volume hydrogen concentration nor the systems components and piping design for withstanding a hydrogen explosion.

LRA Section: 2.3.3.44 - Plant Foundation Underdrain (Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.41 ePortal:

Page 2.3-148 License Renewal Drawing 302-0871: From review of LRA Section 2.3.3.44 and its UFSAR citations it is not clear which piping on drawing 302-0871 is porous concrete pipe vs steel pipe.

LRA Table 3.3.2-44 Auxiliary Systems - Plant Foundation Underdrain Summary of Aging Management Evaluation does Staff request clarification of P72s design and how it is reflected on drawing 302-0871

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions not list a material type of Porous Concrete for the Component Type Piping. Staff request clarification of P72s design and how it is reflected on drawing 302-0871.

LRA Section: 2.3.3.47 - Potable Water Supply (Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.41 ePortal:

Page 2.3-160 License Renewal Drawing 919-0022, Coordinates B-4 & B-6 displays drain traps F-3 and F-5 as being part of the Control Room Envelope (CRE) Pressure Boundary [a(1) function].

License Renewal Note 2 on this drawing also refers to other components F-1 and F-2 that are part of the CRE.

LRA Table 2.3.3-47 Potable Water Supply Component Types Subject to Aging Management Review does not list a Pressure Boundary Intended function for Component Types urinal, water closet, kit. single sink, lavatory.

Staff request clarification of where in the LRA is the aging management of the Pressure Boundary function of these Component Types addressed?

LRA Section: B.2.1 - 10 CFR Part 50, Appendix J Program (TRP 44)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

CR 2020-00889.pdf 25, 27 NEI 94-01 Rev 3A, Section 10.2.3.4 states in part: If Type C test results are not acceptable,... a cause determination should be performed, and corrective actions identified that focus on those activities that can eliminate the identified cause of a failure with appropriate steps to eliminate recurrence. Cause determination and corrective action should reinforce achieving acceptable performance.

NEI 94-01 Rev 3A, Section 10.2.3.3 states in part: An as-left Type C test shall be performed following maintenance, repair, modification or adjustment activity unless an a)

Provide reports which identifies the cause determination of the valve failure and corrective actions completed.

b)

Provide the As Left Type C test results demonstrating valve F0040 is performing within the acceptance criteria.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions alternate testing method or analysis is used to provide reasonable assurance that such work does not affect a valves leak tightness and a valve will still perform its intended function.

On the CR-2020-00889 report, it was observed that valve F0040 failed LLRT multiple times. However, there are no work orders or engineering evaluations discussing cause of failures and corrective actions for this valve.

2 200710364.pdf Page 165 NEI 94-01 Rev 3A, Section 10.2.3.3 states in part: An as-left Type C test shall be performed following maintenance, repair, modification or adjustment activity unless an alternate testing method or analysis is used to provide reasonable assurance that such work does not affect a valves leak tightness and a valve will still perform its intended function.

Section 4.5.3 of PNPP Appendix J Testing Program {NOP-ER-2008} states: As Left Testing is to be performed whenever repairs or adjustments are performed that affect the leakage rate of a component subject to Type B or Type C testing prior to it being relied upon for establishing or maintaining primary containment integrity.

It is unclear whether an As Left Test of MOV isolation valve 1G33F0039 after it was replaced as part of the WO200710364.

a) Provide the As Left Type C test results indicating the replaced valve 1G33F0039 is performing within the acceptance criteria.

LRA Section: 4.6.5 - Fatigue due to Partial Feedwater Heating (TRP 116.5)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.6.5 4.6-5 The referenced page of the application states:

The applicants TLAA description and evaluation did not further discuss the re-investigation of the effect of

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions The effect of PFH on the acoustic and flow induced loads on the reactor shroud, shroud support and jet pumps were re-investigated to ensure that design limits are not exceeded.

partial feedwater heating (PFH) on the acoustic and flow induced loads on the reactor shroud, shroud support and jet pumps.

a) Discuss the time-limited aspects (if any) of the effect of PFH on the acoustic and flow induced loads on the reactor shroud, shroud support and jet pumps.

b) Discuss changes/updates (if any) to the current licensing basis acoustic and flow induced loads on the reactor shroud, shroud support and jet pumps.

LRA Section: 2.3.3.35 - MCC Switchgear and Miscellaneous Electrical Area HVAC, and Battery Room Exhaust (Scoping &

Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.3.3.35 ePortal License Renewal Drawings: 912-0609 Coordinates B-3 & B-

14. Wall dampers (component numbers unreadable) appear to be protected by missile shields.

Please discuss with the staff why these dampers are not subject to AMR.

2 2.3.3.35 ePortal:

Pages 2.3-134 2.3-153 License Renewal Drawing: 912-0609 Coordinates B-6, B-11 et.al. Flow straighteners displayed on drawing.

LRA Table 2.3.3-35 MCC Switchgear and Miscellaneous Electrical Area HVAC, and Battery Room Exhaust Component Types Subject to Aging Management Review does not list a Component Type Flow straightener.

In contrast, LRA Table 2.3.3-45 Plant Radiation Monitoring and Process Monitoring and Post Accident Radiation Please discuss with staff the inconsistency between the labeling of the Component Type Flow Straightener versus Flow Element throughout the LRA.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Monitoring Component Types Subject to Aging Management Review lists a Component Type of Flow straightener.

LRA Section: 4.6.5 - TLAA Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.6.5 4.6-5 The referenced page of the application states:

The effect of PFH on the acoustic and flow induced loads on the reactor shroud, shroud support and jet pumps were re-investigated to ensure that design limits are not exceeded.

The applicants TLAA description and evaluation did not further discuss the re-investigation of the effect of partial feedwater heating (PFH) on the acoustic and flow induced loads on the reactor shroud, shroud support and jet pumps.

a) Discuss the time-limited aspects (if any) of the effect of PFH on the acoustic and flow induced loads on the reactor shroud, shroud support and jet pumps.

b) Discuss changes/updates (if any) to the current licensing basis acoustic and flow induced loads on the reactor shroud, shroud support and jet pumps.

LRA Sections: B.2.10 - BWR Feedwater Nozzle Program (TRP 5)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

B.2.10 B-39 Via ML22006A167, the NRC authorized alternative IR-063 which specifies feedwater nozzle-to-vessel welds and feedwater nozzle inner radii examinations to be done on a sampling basis of 25 percent of the population each inservice inspection interval in accordance with Table 6-1 of FE NE-Request the licensee specify whether the sampling methodology is to select nozzles and inner radii randomly each interval or whether the inspections are done on a rotating basis across the population.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 523-A71-0594 Rev. 1 instead of 100 percent of the population in accordance with Section XI of the ASME B&PV Code.

LRA Sections: AMP XI M9 (TRP 9)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

Appendix C.1 C-10, C-12 Action Items BWRVIP-76, Rev.1 (4) and BWRVIP-76 (8) state that LRA applicants shall reference staff-approved topical reports BWRVIP-99-A and BWRVIP-100-A. The PNPP Response column states that BWRVIP-99-A and BWRVIP-100-A are discussed in LRA Appendix B Section B.2.14.

Neither BWRVIP-99-A nor BWRVIP-100-A are discussed in LRA Appendix B Section B.2.14.

BWRVIP-99-A is mentioned in LRPY-AMP-XI.M9 Section 2.5.1 where it is quoted from NUREG-1801 Rev. 2 but not discussed directly by the applicant and BWRVIP-100-A isnt cited. Are they discussed/cited elsewhere as per the BWRVIP-76 Rev. 1 action items?

Has the applicant addressed the EPRI notification (ML21804A164) regarding the non-conservativism of BWRVIP-100-1A?

2 Appendix C.1 C-14 Action Item BWRVIP-139-R1-A (2) states that applicants applying the BWRVIP-139-R1-A and associated Appendix B shall describe in the USAR supplement summary description for the AMP how the BWRVIP-139-R1-A report and Appendix B of the report will be used to manage aging in the plants steam dryer assembly components. The PNPP Response column states the USAR supplement discussing the PNPP BWR Vessel Internals program is provided in LRA Appendix A, Section A.1.14 LRA Appendix A Section A.1.14 does not discuss the AMP specifically or provide a reference to another document which discuss the issue. Is the result of the action item fully captured as necessary?

3 Appendix B.2.14 B-50, B-51 NUREG-1801 Rev. 2 AMP XI.M9 Section 7 Corrective Actions notes that when cracking is observed in the top guides, sample size and inspection frequencies are increased.

Please provide where the information of this corrective action is captured in documentation? Although Perry has not observed indications in the top

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRPY-AMP-XI.M9 AMP Eval Pg.21 guides, the changes to sample size and frequency should be noted in documentation so theyre not missed if indications are found in the future.

4 Appendix B.2.14 B-53 Operating Experience for the steam dryer notes various IGSCC in the upper support ring and that subsequent inspections have shown increased number and size of indications, although the longest linear indication did not grow during inspections in 2017 or 2019.

Is Perry reviewing the indications against BWRVIP-139 R-1-A or BWRVIP-181 R-2 to determine when repair of the upper support ring becomes necessary?

LRA Section: 2.3.1.3 - Reactor Coolant Pressure Boundary (TRP 27)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1A Section 2.3.1.3 2.3-7 Section 2.3.1.3 of the LRA states that the reactor coolant pressure boundary to include all pressure-containing components such as pressure vessels, piping, pumps, and valves. Further, Section 2.3.1.3 states that the system is within the scope of license renewal because it contains safety-related components that are relied upon to remain functional during or following a design basis event, and because it is relied on to perform a function that demonstrates compliance with the regulations for Fire Protection (FP) and Equipment Qualification (EQ). The reactor coolant pressure boundary contains the coolant under operating temperature and pressure conditions and limits leakage and the release of fission products. [10 CFR 54.4(a)(1), (a)(3) - (FP)].

Table 2.3.1-3 of the LRA does not include the reactor coolant pressure boundary FP systems and components within the scope of license renewal and subject to an aging management review (AMR).

Verify whether the reactor coolant pressure boundary fire protection systems and components are within the scope of license renewal in accordance with 10 CFR 54.4(a) and whether they are subject to an AMR in accordance with 10 CFR 54.21(a)(1). If any, provide the list of FP systems and components.

If there are no FP systems and components within the scope of license renewal and not subject to an AMR, the staff requests that the applicant provide justification for the exclusion.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Sections: TLAA/AMP/Scoping and Screening (TRP 55 & Scoping & Screening)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.1.2.3 2.1-18 NUREG 1800 (SRP-LR) Table 2.1-2, Specific Staff Guidance on Scoping, recommends that the applicant describe the basis for the commodity groupings.

PNPP LRA Section 2.1.2.3, Screening of Electrical Systems, states:

All electrical components within in-scope systems were included within the scope of license renewal. In-scope electrical components were placed into commodity groups and were evaluated as commodities during the screening process.

a-Provide the basis for placing the in-scope electrical components into commodity groups.

b-Clarify if I&C components are also included in scope of LR and in the commodities groups.

2 2.5.2 2.5-2 PNPP LRA Section 2.5.2, Application of Screening Criteria 10 CFR 54.21(a)(1)(i) to electrical and I&C component commodity groups, states:

The screening determination with respect to the passive criterion is taken directly from NEI 95-10. Appendix B of NEI 95-10 delineates which commodity groups are active and which are passive. The active components are excluded from further review, by the direction of 10 CFR 54.21(a)(1)(i).

Specific Perry documents were reviewed to determine the applicability of the industry standard commodity groups (i.e.,

single-line drawings, maintenance rule functions, UFSAR

[Reference 1.3-6] Chapter 7, Chapter 8, Appendix 9A.4, and electrical layout drawings, etc.). The screening review also evaluated the environmental qualification status of the electrical and I&C components. The screening review did not identify any additional commodity groups for evaluation.

Clarify if all electrical and I&C passive components and commodity groups identified for PNPP LR are consistent with (i.e., same as) the electrical and I&C passive components and commodity groups identified in Appendix B of NEI 95-10, which is incorporated into NUREG 1800 (SRP-LR) Table 2.1-5.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 3

Table 2.2-3 2.2-9 PNPP LRA Table 2.2-3, Electrical and I&C Systems, includes 1) thermocouples/electrodes intended for monitoring of corrosion of the drywell wall suppression pool liner and 2) temperature and vibration sensors used for plant start-up testing.

Appendix B of NEI 95-10 (NUREG 1800 (SRP-LR) Table 2.1-

5) indicated that elements, resistance temperature detectors, sensors, thermocouples, transducers, and electric heaters, are active components or commodity groups that meet 10 CFR 54.21(a)(1)(i) if they have pressure boundary functions.

Clarify if the components or commodity groups of elements, resistance temperature detectors, sensors, thermocouples, transducers, and electric heaters have passive boundary functions for LR at PNPP. If they do, discuss how they are screened in PNPP LR.

4 2.1.1.3.5 2.1-12 PNPP LRA Section 2.1.1.3.5, Station Blackout (10 CFR 50.63), states that the boundary for SBO recovery is extended to the first interconnective devices (i.e., 345kV switchyard circuit breakers S-612, S-620, S-621, S-650, S-652, S-660 and S-661) that would restore offsite power to the main switchyard busses and then to the step-up station transformer (refer to scoping boundary drawing 206-0010 LR)

The staff notes that there is no step-up transformer connected to the 345kV circuit breakers and breaker S-612 is not highlighted on drawing 206-0010 LR.

a-Clarify if the startup transformer is the step-up station transformer for Unit 1 and Unit 2 b-Clarify why breaker S-612 is not highlighted in the boundary drawing 206-0010 LR that is provided in the portal.

5 2.1.1.3.5 2.1-12 PNPP FSAR 8.2.1.2.1, Transmission Station, indicates that there are 2 direct current systems that provide power for the 345kV switchyard circuit breakers and are in the transmission station control house.

During the onsite audit, the applicant noted that control power/circuit for the 345 kV switchyard breakers associated with the SBO recovery path is not within scope of since they are crediting manipulation of a manual disconnect instead to restore power following an SBO event.

1-Describe the process for recovering power during an SBO with PNPP 4-hour coping duration.

2-Provide justification for not including the control circuits and structures associated with the 345kV switchyard breakers in the scope of LR.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions The staff challenged the applicants approach to credit manipulation of a manual disconnect since:

1) 10 CFR 50.63) requires and related guidance (NUMARC 87-00 and RG 1.155) notes that the plant must be able to withstand and recover from an SBO in a specified duration (i.e., coping duration); and 2) NUREG 1800 (SRP-LR) section 2.5.2.1.1, Components within the Scope of SBO (10 CFR 50.63), notes that the control circuits and structures associated with breakers that are within the SBO recovery path are in scope of LR.

6 B.2.24 TRP 55 B-79 5-PNPP LRA B.2.24, Fuse Holders Program, states that the Perry Aging Management Program for Fuse Holders is a new program that is consistent with the 10 elements of an effective aging management program as described in NUREG-1801,Section XI.E5, Fuse Holders.

The staff reviewed Perry AMP for fuse holders in the portal and noted that the elements of parameters monitored/inspected, detection of aging effects, acceptance criteria, and operating experience of PNPP fuse holders AMP are not consistent with those of NUREG-1801 AMP XI.E5.

Perry AMP Parameters monitored/inspected - This statement:

Contacts are tested by thermography, contact resistance, or other appropriate method at least once every 10 years for indication of the metallic clamps of the fuse holders, is not in the NUREG 1801 AMP XI.E5 element 3 Perry AMP Detection of aging effects - No contact resistance testing or other appropriate testing methods are provided in the detection of aging effects. The only testing provided is thermography.

Explain how Perry AMP for fuse holders is consistent with the elements of NUREG-1801 AMP XI.E5. If there is no consistency, provide justification for the differences between the Perry AMP elements for fuse holders and the NUREG-1801 AMP XI.E5 elements and revise PNPP LRA B.2.24 accordingly.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Perry AMP Acceptance criteria - Criteria for temperature and resistance of the metallic clamps of the fuses holders are not included in the acceptance criteria.

Perry AMP Operating experience - the documents considered for the AMP are not provided in the operating experience.

LRA Sections: B.2.31, and B.2.33 (TRPs 49, 50, 51 & 54)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

B.2.31 B-91 In the Program Description of the Non-EQ Electrical Cable Connections Program, the applicant notes that this one-time test program provides reasonable assurance that the intended functions of the metallic parts of electrical cable connections and susceptible to age-related degradation resulting in increased resistance due to corrosion are maintained.

Explain the sampling basis methodology that adequately addresses corrosion on metallic parts of electrical cable connections through the period of extended operation and how will the cables be managed in the XI.E6 AMP.

2 B.2.31 B-92 The existing program provides a statement that theres a preventative maintenance program that is in place and is effective such that a periodic inspection program is not required. However, theres visual corrosion on the cable connections program. How does the preventative maintenance program cover visual corrosion conditions?

Are maintenance procedures being used to age manage electrical cable connections scoped in XI.E6? if so, are there any actions within the procedures that address visual corrosion on equipment? If available, please upload the procedures.

3 B.2.32 B-93 The program description notes the manholes and underground vaults are inspected for water accumulation and water removed. However there appears to be a pattern of cable wetting in the manholes.

Explain how the design change implementation will comply with cable condition monitoring requirements and adequately manage the condition of cables (specifically XI.E3) during the period of extended operation?

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 4

OpE OpE Explain the cause of the deficiencies for the manholes and underground vaults (CR-2017-02787, CR-2013-11870, CR2015-07181, CR-2016-09778, CR-2022-02632, CR-2019-07599, CR-2016-14536, CR-2016-00770, CR-2022-06118, CR-2020-05243, CR-2020-05686, CR-2014-12127. Requesting more images of in scope manholes to confirm the actual condition of cables are being adequately age managed?

5 OpE OpE What is the process to address the identification of reduced insulation resistance due to moisture, oil and exposed to heat-related conditions for the XI.E1 AMP? Are there procedures in place?

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

7 OpE OpE It appears cable submergence has occurred for long periods of time.

Requesting more images of in-scope manholes to determine the actual conditions of the manholes.

8 B.2.33 B-95 The program description assures the intended functions of sensitive, high-voltage, low-signal cables exposed to adverse localized equipment environments caused by heat, radiation and moisture will be maintained.

Based on the following CRs CR-2017-02456, CR2013-16629, CR-2022-08732, CR2015-05688, CR-2015-03042, CR-2013-08641.

Are there any procedures to age

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions manage the instrumentation circuit cables for XI.E2 AMP?

LRA Sections: 2.5.5.4 and B.2.31 (TRPs 54 & 80)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

2.5.5.4 3.6.2.1.6 2.5-6 3.6-5 The staff was able to perform a walkdown from the vicinity of the switchyard to the startup transformers of the plant. While visually assessing the conditions of the high-voltage insulator some of the transmission line insulators showed corrosion (rust) at the connection portion between the transmission conductor and the insulator. GALL includes an item for managing the aging effect for high voltage insulators, such as loss of material.

Based on the observations during the walkdowns, there were signs of corrosion in the connection portion of the High-voltage insulators. Since Perrys LRA notes that no significant aging mechanism affects high-voltage insulators, therefore, no aging management program will be created, discuss how that plant will manage the observed aging effects (i.e., corrosion/rust) for high-voltage insulators.

2 B.2.31 While performing the walkdowns, the NRC staff noted corrosion (rust) on the outside of the LH-1-A and its connections. PNPP staff to provide photos of the portions of the transformers that were not visible. The photos of the cables didnt show any visual degradations. However, the staff observed visual signs of corrosion (i.e., rust) of the metallic parts of the LH-1-A (e.g., cable conduits, and junction boxes) and its connections. Currently the Non-EQ Electrical Cable Connections Program is a one-time test program.

Discuss how PNPP plans to monitor aging degradation of non-eq cable connections thru the period of extended operating time once the initial test of the equipment is completed?

LRA Section: 3.6 - Aging Management of Electrical Commodities (TRPs 57 & 61)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

3.6 (Table 3.6.2 -

3.6-22 The staff notes that the applicant did not address EQ in table 3.6.2 which corresponds to Table 3.6.1 Item 3.6.1-1.

While EQ is a TLAA, the applicant provided (as is typical The following elements should be included and addressed in Table 3.6.2 - Electrical Commodities of

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Electrical Commodities) and expected) an aging management review of their EQ program in Sections A.1.17, A.2.4, and B.2.17.

the Perry LRA to be consistent with GALL and prior reviews:

Electrical Equipment subject to 10 CFR 50.49 EQ requirements for Electrical Continuity, Insulate (Electrical),

Various Organic Polymers in Adverse Localized Environment (Ext),

Various Degradations (in Accordance with 10 CFR 50.49),

TLAA - Environmental Qualification (Section 4.4),

VI.B.L-05 3.6-1, 001, Note A LRA Section: 3.5 - Water Chemistry (TRP 2)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

3.5 3.5-87 and 3.5-131 In the initial set of breakout questions, questions 7 and 9 refer to aluminum component and piping supports subject to cracking and aluminum upper containment pool gates subject to cracking. It is unclear to the staff why GALL consistency was determined to be different for two components with material, environment, and aging effects that appear to be the same. Row 56 of Table 3.5.2-4 addresses cracking of aluminum supports in treated water. Row 141 of Table 3.5.2-1 addresses cracking of aluminum upper containment pool gates in treated water. In the first case the response to question 7 states that a comparable aging management item (3.3.1-136 from LR ISG 2012-02 and GALL item VII.G.A-412) is used with Note E, and the component type is not considered relevant. In the second case, Table 3.5.2-1 uses Note H, indicating the aging effect is not in the GALL report for this component, material, and environment combination.

Please clarify how the consistency with the GALL report was determined to be different (Note E vs. Note H) for managing cracking of these two components, which appear to be exposed to the same environment and made of the same material.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: B.2.41 - Water Control Structures (TRP 47)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

Table 3.5.1 3.5-49 LRA states AMR 3.5.1-56 Loss of material due to abrasion, cavitation is N/A Please explain why its not applicable (N/A) to Perry.

2 Table 3.5.1 3.5-51 LRA states AMR 3.5.1-60 Loss of material (spalling, scaling) and cracking due to freeze thaw.is N/A Please explain why its not applicable (N/A) to Perry. (see ASTM C33-90, Figure 1) 3 AMP B.2.41 B-115 Last sentence states: No signs of MAJOR crack, spalls or Please define the term MAJOR and explain how you intend to measure it.

LRA Section: 4.2.2 - Neutron Fluence Projections (TRP 59.1)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

4.2.2 4.2-3 Discuss the use of RAMA code and BUGLE-96 cross-section library and explain why they adhere to RG 1.190 for the duration of the PNPP requested license renewal application. In addition, discuss any RAMA methodology benchmarking that is applicable to PNPP LRA related analyses.

2 4.2.2 4.2-3 Discuss fluence (number of neutrons per square centimeter that contacts the reactor vessel wall/shell and its internal components) values with respect to the limit for:

a-Jet Pump Assemblies b-Recirculation inlet nozzle thermal sleeves c-Steam Dryer

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions d-Core Spray Lines1 and Liquid Control Line e-Vessel Head Spray Nozzle f-Differential Pressure Sensing Lines g-In-Core Flux Monitor Guide Tubes h-Surveillance Sample Holders i-Low Pressure Coolant Injection (LPCI) Lines LRA Section: AMP XI.M4 - BWR Vessel ID Attachment Welds (TRP 4)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 1

A.1.13 A-17 A.1.13 BWR Vessel ID Attachment Welds Program states, in part:

The BWR Vessel ID Attachment Welds aging management program is an existing condition monitoring program that includes the inspection and evaluation recommendations within BWRVIP-48 and the requirements of ASME Code,Section XI, Subsection IWB.

NUREG-1801, Rev. 2 states:

The program includes inspection and flaw evaluation in accordance with the guidelines of a staff-approved boiling water reactor vessel and internals project (BWRVIP-48-A) to provide reasonable assurance of the long-term integrity and safe operation of boiling water reactor (BWR) vessel inside diameter (ID) attachment welds.

NUREG-1801 addresses the BWR vessel ID attachment welds in AMP XI.M4. The AMP specifically identifies BWRVIP-48-A for meeting program recommendation of GALL Report.

The UFSAR supplement cites BWRVIP-48. Clarify which version of the report is denoted by BWRVIP-48, as applicable to the PEO after the 4th ISI interval concludes (May 2029).

If version other than BWRVIP-48-A will be used following the 4th 10-year ISI interval - provide basis/justification.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions Discuss if UFSAR supplement needs to modify for clarity.

2 B.2.13 B-46 The AMP identified an exception to NUREG-1801: The BWR Vessel ID Attachment Welds Program is based on the requirements of BWRVIP-48, R1 in lieu of BWRVIP-48-A.

In the Safety Evaluation dated Jan. 29, 2021 (ML20363A006),

the NRC staff accepted the use of identified BWRVIP guidelines, including BWRVIP-48, Revision 1, in lieu of the ASME Code requirements for the fourth 10-year ISI interval (scheduled to conclude on May 17, 2029). The current operating license for PNPP expires November 7, 2026.

The LRA states, It is understood that prior to entering the subsequent 10-year ISI intervals, PNPP would have to either comply with the ASME Code requirements or request relief from the requirements consistent with what PNPP has done during the initial operating period.

NUREG-1801 specifically identifies BWRVIP-48-A for meeting AMP XI.M4 recommendations. Previous NRC approval of BWRVIP-48, Revision 1, is limited to and expires at the end of the fourth 10-year ISI interval.

Identify the version of BWRVIP-48 that will be applicable to the AMP upon completion of the fourth 10-year ISI interval. If an approach other than following ASME Code Section XI requirements and BWRVIP-48-A will be implemented upon completion of the fourth 10-year ISI interval (during the PEO), clarify, and justify.

LRA Section: 3.5 - Steel-other (TRP 100)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 3

3.5 3.5-91 In the response to Breakout Question 1, the applicant proposed to revise the LRA to cite AMR item 3.5.1-93 for the galvanized steel diesel fuel tank maintenance structures exposed externally to outdoor air instead of AMR item 3.5.1-

95.

AMR item 3.5.1-93 (III.B2.TP-6 and III.B4.TP-6) in Revision 2 of NUREG-1801 is for managing loss of material of galvanized steel, aluminum, and stainless steel support members, welds, bolted connections, and support anchorage Please discuss citing AMR item 3.5.1-93 for managing cracking for galvanized steel conduits2 exposed externally to soil, and loss of material for stainless steel conduits4 exposed externally to soil. Please also discuss the use of standard note A?

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions to building structures exposed to outdoor air by the Structures Monitoring program.

LRA Table 3.5.2-4 cites AMR item 3.5.1-93 for managing cracking and loss of material for galvanized steel conduits2 exposed externally to soil by the structures monitoring program, and loss of material for stainless steel conduits4 exposed externally to soil by the structures monitoring program. Standard note A used in each instance.

The Discussion of AMR item 3.5.1-93 in LRA Table 3.5-1, as proposed to be revised, does not cover its use for managing cracking or an external soil environment.

LRA Section: B.2.43: Structures Monitoring (TRP 46)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 4A (Follow-up to question

4)

B.2.43 Table 3.5.2-4 B-121 B-122 3.5-160 3.5-161 Aging management of the plant storm drain system is plant-specific, which cannot not be managed by current Structures Monitoring program. The applicant needs to follow Branch Technical Positions described in Appendix A of SRP-LR to provide plant-specific aging management program or enhancements to the Structures Monitoring program.

LRA Section B.2.43 discuss the scope of the plant storm drain system, however, wording loss of material and flow blockage does not belong to the scope.

LRA Section B.2.43 does not make clear whether roof drain system is a part of the plant drain system.

LRA Section B.2.43 does not make clear whether polymer storm drain piping is monitored for cracking, loss of material, and hardening.

Discuss any necessary enhancements to the Structures Monitoring program and provide an explanation for how the enhanced structures monitoring program will meet the 10 Aging Management Program Elements of Appendix A of the SRP-LR for the plant storm drain system.

Specifically clarify:

The applicability of loss of material and flow blockage in the enhancement to the scope of program element.

Whether or not the roof drain system should be included in an enhancement to the scope of program element.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Table 3.5.2-4, rows 292 and 293 does not include flow blockage, therefore it does not make clear whether aging effect of flow blockage is needed for polymer storm Drain1.

The parameters monitored or inspected for the plant drain system listed in LRA do not match what are described on the ePortal.

LRA B.2.43 does not make clear of inspection frequency for the plant drain system.

The statement This monitoring will consist of direct observation. in an enhancement to Element 3 belongs to Element 4.

LRA Section B.2.43 states periodic remote visual inspections of sampled portions of storm drain system piping, it does not make clear of periodic remote visual inspections, sampled portions and their selections.

The applicant provides additional enhancement to measure wall thickness of corrugated metal pipe shall be included during opportunistic excavations of storm drain, However, the applicant does not include wall thickness measurement of polymer and concrete storm drain piping in the enhancement on the ePortal.

LRA Section B.2.43 does not make clear of acceptance criteria for storm drain piping.

Whether or not the plant procedure will be revised to include inspection guidelines to maintain the intended function of the storm drain system.

What aging effects are applicable to polymeric portions of the storm drain system and whether they are consistent with Table 2 items of polymer storm drain piping.

Which parameters are monitored/inspected for the storm drain system in the structures monitoring program.

The proposed inspection frequency associated with monitoring the storm drain system/basis for that frequency.

How sample locations/sizes are determined to monitor the plant storm drain system.

What types of opportunistic inspections are performed on which materials when portions of the storm drain system are excavated.

What are examples of acceptable inspection techniques (i.e., what are examples of acceptable remote visual inspection or

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions wall thickness examination techniques).

Whether wall thickness measurement of polymer and concrete storm drain piping should be included in the enhancement to the detection of aging effect element.

What acceptance criteria will be used for inspections of the storm drain system.

7A (Follow-up to question

7)

B.2.43 Figure 3.6-6 USAR Figure 2.4-68 and 69 B-122 3-115 7772 and 7773 of 9336 Aging management of the plant underdrain system is plant-specific, which cannot not be managed by current Structures Monitoring program. The applicant needs to follow Branch Technical Positions described in Appendix A of SRP-LR to provide plant-specific aging management program or enhancements to the Structures Monitoring program.

LRA Section B.2.43 provides an enhancement to scope of program for the structural integrity of the plant drain system.

The staff noted wording structural integrity to be not applicable to the scope of program.

LRA states that the plant underdrain system includes porous concrete sub-foundation and 12 porous concrete piping. It also includes steel gravity discharge piping, which their aging effects are managed by the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting. Therefore, LRA Section B.2.43 does not make clear of the scope of program, which is managed by the Structures Monitoring program.

LRA Section B.2.43 states The program will be enhanced to take credit for the existing monitoring program elements in place to maintain the structural integrity of the plant underdrain system Discuss any necessary enhancements to the Structures Monitoring program and provide an explanation/clarification for how the enhanced structures monitoring program will meet the 10 Aging Management Program Elements of Appendix A of the SRP-LR for the plant underdrain system.

Specifically clarify:

The applicability of structural integrity in any enhancement to the scope of program element.

The portions of the plant underdrain system that will be managed by the Structures Monitoring program.

Which parameters will be monitored/inspected, including for the porous concrete pipe and concrete sub-foundation

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions including the porous concrete sub-foundations in the enhancement to the parameters monitored or inspected.

However, this statement is note related to this Element.

LRA Section B.2.43 does not make clear of parameter monitored or inspected for the plant underdrain system.

LRA Section B.2.43 does not make clear of inspection frequencies for the plant underdrain system, settlement monitoring and groundwater elevation monitoring.

Similarly, LRA Section B.2.43 does not make clear whether direct observation (blockage), periodic remote visual inspections of sampled portions of porous concrete pipe, and wall thickness measurement of porous concrete pipe are needed (see follow-up questions on the plant storm drain pipe) during opportunistic excavations of porous concrete pipe.

LRA Section B.2.43 does not make clear whether non-destructive examination technique such as Ultrasonic Testing will be used to determine wall thickness of the porous concrete pipe.

LRA Section B.2.43 does not make clear of acceptance criteria for the plat underdrain system, building settlement, and groundwater elevation.

The proposed inspection frequencies associated with monitoring the plant underdrain system, settlement monitoring, and groundwater elevation monitoring/the bases for those frequencies.

How the porous concrete pipe (i.e. flow blockage) will be monitored/detected/cleaned to ensure continuity of intended function through the PEO What acceptance criteria will be used for inspections of the plant underdrain system, the monitoring of building settlement, and the monitoring of groundwater elevation.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions 36 (New Question) 3.5.2.2.1.6 AMR 3.5.1-10 3.5-11 3.5-25 SRP-LR Section 3.5.3.2.1.6 states that Containment ISI IWE and leak rate testing may not be sufficient to detect cracks, especially for dissimilar metal welds, and additional appropriate examinations to detect SCC in bellows assemblies and dissimilar metal welds are recommended to address this issue.

LRA Table 3.5.1 claims AMR 3.5.1-10 to be not applicable, however, LRA provides Table 2 items associated with AMR 3.5.1-10.

LRA Section 3.5.2.2.1.6 states these elements are not subject to an aggressive chemical environment. LRA Table 3.5.1 on Page 3.5-37 describes this environment as high temperature, moist or wetted environment or the environment contaminated with chlorides, fluorides, or sulfates.

LRA Section 3.5.2.2.1.6 states aging effects of drywell mechanical penetrations are managed by the IWE program and the Structures Monitoring program. LRA does not make clear whether drywell mechanical penetrations are subject to Appendix J testing.

1. Evaluate the non-applicable claim of AMR 3.5.1-10 and address the discrepancy.

2. Clarify whether the penetration sleeves and bellows and associated welds including drywell mechanical penetrations are subject to high temperature, moist or wetted environment or the environment contaminated with chlorides, fluorides, or sulfates. If yes, clarify what examination methods are capable of detecting the cracking due to stress corrosion cracking.

3. Clarify whether drywell mechanical penetrations are subject to Appendix J testing and explain why the IWE program and Structures Monitoring program can be used to detect this aging effect of drywell mechanical penetrations.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions LRA Section: B.2.43: Structures Monitoring (TRP 46)

Question Number LRA Section LRA Page Background / Issue Discussion Question / Request 37 (Follow-up on item 22)

Table 3.5.2-4 3.5-133 LRA Table 3.5.2-4 cites Note H for polymer conduit caps to manage aging effect of loss of strength along with plant specific notes 518 Loss of strength does not fail the function.

The loading on the cap to perform the function is minor. As long as the cap is visually leak tight, the function is ensured.

It appears that plant-specific notes 518 is very confusing, especially for wording Loss of strength does not fail the function, and does not make clear what acceptance criteria are used for polymer conduit caps to manage aging effect of loss of strength.

GALL-LR XI.S6 does not include polymer material for structural components. However, GALL-LR XI.M36 includes aging effects for the flexible polymeric materials and states that hardening and loss of strength and loss of material due to wear for flexible polymeric materials are expected to be detectable prior to any loss of intended function. It appears that LRA does not address all the aging effects for the polymer conduit caps.

1. Evaluate whether hardening and loss of material are applicable aging effects for the polymer conduit caps.

If not, provide technical justification for why they are not applicable.

2. Evaluate and discuss plant-specific notes 518 to ensure that aging effects are adequately managed by the Structures Monitoring program.

3. Evaluate acceptance criteria for the polymer conduit caps.

38 (Follow up on item 23)

Table 3.5.2-4 3.5-155 3.5-163 3.5-164 LRA Table 3.5.2-4 cites Note H for roof membrane 1 and waterproofing membranes (1) exposed to concrete, soil, and raw water environments, total 4 Table 2 items (rows 259, 321, 322, and 323).

LRA does not make clear how aging effects are adequately managed by the Structures Monitoring program.

GALL-LR XI.S6 does not include aging effects for the elastomer roof membrane, but it states that elastomeric vibration isolators and structural sealants are monitored for

1. Evaluate whether loss of material and hardening are applicable aging effects for the elastomer roof membrane. If not, provide the justification for why they are not applicable.

2. Clarify how these aging effects for these components are managed by the Structures Monitoring program.

3. Evaluate acceptance criteria for the elastomer roof membrane.

Perry Nuclear Power Plant, Unit 1 License Renewal Application (LRA) Breakout Audit Questions cracking, loss of material, and hardening, which maybe applicable to the elastomer roof membrane.

LRA Table 3.5.2-4, row 258 cites Note A for roof membrane associated with AMR 3.5.1-72, which has different aging effect from the corresponding GALL item III.A6.TP-7.

4. Evaluate AMR item on row 258 in LRA Table 3.5.2-4 and address the inconsistency with the GALL item.