L-2021-144, Subsequent License Renewal Application-Aging Management Requests for Additional Information (RAI) Set 2 Responses

Subsequent License Renewal Application-Aging Management Requests for Additional Information (RAI) Set 2 Responses

ML21223A308
Person / Time
Site:	Point Beach
Issue date:	08/11/2021
From:	Maher W Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2021-144
Download: ML21223A308 (177)

Text

August 11, 2021 U.S. Nuclear Regulatory Corrunission Attention: Document Control Desk 11545 Rockville Pike One \\,\\!hite Flint North Rockville, JvID 20852-2746 Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 NEXTera ENERGY.

POINT BEACH L-2021-144 10 CFR 54.17 SUBSEQUENT LICENSE RENEWAL APPLICATION -AGING MANAGEMENT REQUESTS FOR ADDITIONAL INFORMATION (RAI) SET 2 RESPONSES

References:

1.

NextEra Energy Point Beach, LLC (NEPB) Letter NRC 2020-0032 dated November 16, 2020, Application for Subsequent Renewed Facility Operating Licenses (ADAMS Package Accession No. ML20329A292)

2.

U.S. Nuclear Regulatory Corrunission (NRC) Letter dated January 15, 2021, Point Beach Nuclear Plant, Units 1 and 2 - Determination of Acceptability and Sufficiency for Docketing, Proposed Review Schedule, and Notice of Opportunity to Request a Hearing Regarding the NextEra Energy Point Beach, LLC Application for Subsequent License Renewal (EPID No. L-2020-SLR-0002)

(ADAMS Accession No. ML21006A417)

3.

NRC Letter dated January 15, 2021, Point Beach Nuclear Plant, Units 1 and 2 - Aging Management Audit Plan Regarding the Subsequent License Renewal Application Review (ADAMS Accession No. ML21007A260)

4.

US Nuclear Regulatory Commission Meeting with NextEra Energy Concerning the Point Beach Subsequent License Renewal Application Review - June 3, 2021 Public Meeting (ADAMS Accession No. ML21148A116)

5.

NRC Email and Attachment dated July 13, 2021, Point Beach SLRA RAis Set 2 Final (ADAMS Accession Nos. ML21208A191, ML21208A189)

NEPB, owner and licensee for Point Beach Nuclear Plant (PBN) Units 1 and 2, has submitted a subsequent license renewal application (SLRA) for the Facility Operating Licenses for PBN Units 1 and 2 (Reference 1).

On January 15, 2021, the NRC determined that NEPB's SLRA was acceptable and sufficient for docketing (Reference 2), and on January 15, 2021 issued the regulato1y audit plan for the aging management portion of the SLRA review (Reference 3). Based on the information exchanged and discussions held during the public meeting held on June 3, 2021 (Reference 4), the NRC issued its Set 2 RAis to NEPB (Reference 5). The attachments to this letter provide responses to those information requests.

For ease of reference, the index of attached information is provided on page 3 of this letter. Certain attachments include associated revisions to the SLRA (Enclosure 3 Attachment 1 of Reference 1) denoted by NextEra Energy Point Beach, LLC 6610 Nuclear Road, Two Rivers, WI 54241

Document Control Desk L-2021-144 Page 2 strikethrough (deletion) and/or bold red underline (insertion) text. Previous SLRA revisions are denoted by bold black text. SLRA table revisions are included as excerpts from each affected table.

Should you have any questions regarding this submittal, please contact me at (561) 304-6256 or William.Maher@fpl.com.

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

Executed on the 11th day of August 2021.

Sincerely, William Maher Digitally signed byWilllam Maher ON: cn=W1lllam Maher,o=Nuclear, ou=Nudear licensing Projects, emalt=wllliam..mahere rpl.com, c=US Oate:2021.08.1114:16:17-04'00' William D. Maher Licensing Director - Nuclear Licensing Projects Cc:

Administrator, Region III, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC Public Service Commission \\'{fisconsin

Document Control Desk L-2021-144 Page 3 Attaclunent RAI No.

No.

1 2.3.3.6-1 2

3.5.2.2.2.4-1 3

3.5.2.2.2.4-2 4

B.2.2.1 -1 5

B.2.3.2-1 6

B.2.3.8-1 7

B.2.3.8-2 8

B.2.3.8-3 9

B.2.3.10-1 10 B.2.3.10-2 11 B.2.3.10-3 12 B.2.3.16-1 13 B.2.3.27-1 14 B.2.3.27-2 15 B.2.3.27-3 16 B.2.3.27-4 17 B.2.3.29-1 18 B.2.3.29-2 19 B.2.3.29-3 20 B.2.3.29-4 21 B.2.3.31-1 22 B.2.3.31-2 23 3.5-1, 068-1 24 3.5-1, 085-1 25 B.2.3.34-1 26 3.5.2.11-1 27 3.3.1, 263-1 28 3.3.1, 111-1 29 4.3.1-1 30 4.3.1-2 Attachments Index Subject Fire Protection - Dry Chemical Extinguishing System Cracking Due to Stress Corrosion Cracking and Loss of Material Due to Pitting and Crevice Corrosion (AMR Item 3.5.1-099)

Cracking Due to Stress Corrosion Cracking and Loss of Material Due to Pitting and Crevice Corrosion (AMR Item 3.5.1-100)

Fatigue Monitoring - Reactor Vessel Internals (R\\TJ) Baffle Bolts

\\X!ater Chemistry-Aging Management Evaluation for Stainless Steel Valve Body in Treated Water Flow Accelerated Corrosion - Revision Level of NSAC-202L Flow Accelerated Corrosion - Software Quality Assurance Flow Accelerated Corrosion - Applicability of Erosion/Flow-Accelerated Corrosion (FAC) to Main and Auxiliary Steam System and Feedwater and Condensate System Components Exposed to Treated \\'V'ater Steam Generators - Carbon Steel Blowdown Piping Nozzles and Secondary Side Shell Penetrations Exposed to Treated Water Steam Generators - Tube Plugging Materials Steam Generators - Fatigue Analysis Fire \\'V'ater System - Components Subject to \\'V'et-Dry Cycle Buried and Underground Piping and Tanks - Cathodic Protection Buried and Underground Piping and Tanks - Uncoated Fire Protection Piping Buried and Underground Piping and Tanks - Interface Corrosion for Tanks Buried and Underground Piping and Tanks - Preventive Action Category C ASME Section XI, Subsection IWE - NUREG-2191 Consistency Statement ASME Section XI, Subsection IWE - Volumetric Examination and Schedule ASME Section XI, Subsection IWE - Detection of Cracking due to Cyclic Loading or Stress Corrosion Cracking ASME Section XI, Subsection I\\'V'E - Containment Components that Serve a Moisture Barrier Function ASME Section XI, Subsection IWF - Preventive Actions ASME Section XI, Subsection IWF - Detection of Stress Corrosion Cracking and Cracking Assessment of High-Strength Bolting Samples ASME Section XI, Subsection I\\'V'F - Detection of Stress Corrosion Cracking in High-Strength Steel Bolting ASME Section XI, Subsection I\\'V'F - Stainless Steel Structural Bolting Exposed to Treated \\'V'ater Structures Monitoring - Epoxy Grouted Anchors and Bolts Structures Monitoring - Earthen Berm Structures Structures Monitoring - Polystyrene (Polymer) Inserts Used in Manhole Covers Structures Monitoring - Inconsistency in Applicability of Item 3.3-1, 111 Metal Fatigue Class 1 Components year Allowable Transient Cycles Metal Fatigue Class 1 Components - Pressurizer Spray Piping and Cycle Monitoring

Document Control Desk L-2021-144 Page 4 Attachment RAI No.

No.

31 4.3.3-1 32 4.3.4-1 33 4.3.4-2 34 16.3.6-1 35

4. 7.1-1 36 4.7.6-1 Attachments Index Subject Metal Fatigue of Non-Class 1 Components - Transients for (1) Feedwater and Condensate; (2) Main and Auxiliary Steam; (3) Reactor Coolant, non-Class 1; and (4) Safety Injection Environmentally Assisted Fatigue - Pressurizer Surge Line, RHR Tee and Accumulator Safety Injection Nozzle Environmentally Assisted Fatigue - Results and Leading Locations UFSAR Supplement for Containment Liner and Steel Piping Penetrations TLAA Leak-Before-Break of Reactor Coolant System Loop Piping-Weld Inlay Material Applied to Unit 1 Steam Generator Inlet and Outlet Nozzles Crane Cycle Load Limit Apparent Inconsistency

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 2.3.3.6-1 L-2021-144 Attachment 1 Page 1 of 3

1. SLRA Section 2.3.3.6, "Fire Protection" Regulatory Basis:

10 CFR 54.4, "Scope," defines the scope of license renewal as those plant SSCs, as well as the process used to identify the SSCs that are subject to an aging management review, as required by 10 CFR 54.21 (a)(1 ); (a) that are safety-related; (b) whose failure could affect safety-related functions; and (c) that are relied on to demonstrate compliance with the NRC's regulations for fire protection, environmental qualification, pressurized thermal shock, anticipated transients without scram, and station blackout.

In particular, Section 54.4(a)(3) of 10 CFR includes within the scope of license renewal all SSCs relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with Commission's regulations for fire protection, 10 CFR 50.48.

RAI 2.3.3.6-1

Background:

[No background was stated in this RAI.]

Issue:

In License Amendment Request 271, Transition to 10 CFR 50.48(c) - National Fire Protection Association (NFPA) 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition (ADAMS Accession No. ML13182A351) Table 4-3, "Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features," indicates that a dry chemical extinguishing system protects fire area A01-E, Fire Zones 322 and 547. It appears to the NRC staff that the dry chemical extinguishing system is necessary to meet the requirement of 10 CFR 50.48. However, the staff is unable to find information in the subsequent license renewal application indicating that this system is within the scope of the license renewal.

Request:

Verify whether the dry chemical extinguishing system is within the scope of license renewal in accordance with 10 CFR 54.4(a) and whether it is subject to an aging management review (AMR) in accordance with 10 CFR 54.21 (a)(1 ), because it appears to be necessary to meet the requirements of 10 CFR 50.48. If it is not within the scope of license renewal and is not subject to an AMR, provide justification for the exclusion.

NEPB Response:

Per RAI response 2.3.3.6-10 issued during PB N's first license renewal the dry chemical suppression system for the turbine-generator bearings and the gas turbine exhaust bearing are in scope in accordance with 10 CFR 54.4(a). The dry chemical containers are managed similar to a fire extinguisher, where they are routinely monitored, and replaced as needed (reference SLRA Section 2.1.5.6, Consumables). The fixed components subject to aging management are represented by component types "piping," "nozzle," and "valve body". "Nozzle" and "valve body" are already included in

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 2.3.3.6-1 L-2021-144 Attachment 1 Page 2 of 3 SLRA Tables 3.3.2-6. SLRA Table 3.3.2-6 is revised to include galvanized steel piping used in internal and external environments and SLRA Table 3.3-1 is revised to reflect that item 3.3-1, 116 is now applicable.

References:

1. FPL Response to Request for Additional Information Regarding the Point Beach Nuclear Plant License Renewal Application, TAC Nos. MC2099 and MC 2100, October 8, 2004, Accession No. ML042940439

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 2.3.3.6-1 L-2021 -144 Attachment 1 Page 3 of 3 Associated SLRA Revisions:

SLRA Table 3.3-1, item 116 (page 3.3-52) is revised as follows:

Table 3.3-1: Summary of Aging Management Evaluations for the Auxiliary Systems Item Component Aging Aging Management Further Number Effect/Meehan ism Program (AMP)/TLAA Evaluation Recommended 3.3-1,

Galvanized steel None None No Discussion

~Jet a13131ieaele.

116 piping, piping

+l=leFe aFe Ae §alvaAieed steel 13i13iA§ components exposed aAEI 13i13iA§ eeffi13eAeAts e:x:13esed te to air - indoor iAdeeF b!A69Atmlled aiF iA tAe uncontrolled A1:1:x:iliai=y Systeffis.

Consistent with NUREG-2191.

There are no aging effects for galvanized steel 12i12ing ex12osed to air - indoor uncontrolled.

SLRA Table 3.3.2-6 (page 3.3-182) is revised as follows:

Table 3.3.2-6: Fire Protection System - Summary of A ::iing Management Evaluation Component Intended Material Environment Aging Effect Aging NU REG-Table 1 Notes Type Function Requiring Management 2191 Item Management Program Item Pi12ing Pressure Galvanized steel Air-indoor None None Vll.J.AP-3.3-1, 116 A

bounda!Jl uncontrolled 13 (ext)

Pi12ing Pressure Galvanized steel Air-indoor None None Vll.J.AP-3.3-1, 116 A

bounda!Jl uncontrolled 13 (int)

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.4-1 L-2021-144 Attachment 2 Page 1 of 3

2. SLRA Section 3.5.2.2.2.4, "Cracking Due to Stress Corrosion Cracking and Loss of Material Due to Pitting and Crevice Corrosion" (Further Evaluation)

Regulatory Basis:

1 O CFR 54.21 (a)(3) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

RAI 3.5.2.2.2.4-1

Background:

AMR item 3.5.1-099 is applied to manage aging effects of loss of material due to pitting and crevice corrosion, cracking due to stress corrosion cracking (SCC) for aluminum and stainless steel (SS) support members, welds, bolted connections or anchorage to structure for ASME Class 1, 2, 3 or MC (metal containments) components. SLRA Table 3.5-1, as amended by Attachment 29 of SLRA Supplement 1 dated April 21, 2021 (ADAMS Accession No. ML21111A155), states that AMR item 3.5.1-099 is not used, and the aging effects of loss of material and cracking of aluminum and SS supports and other structural components are addressed in AMR item 3.5.1 -100.

Issue:

Based on the staff's review of the information provided in the SLRA, it is not clear how components associated with item 3.5.1 -099 were addressed under AMR item 3.5.1, 100 or if the AMR item is not applicable for the site. The staff noted that table 2 AMR items associated with item 3.5.1 -100 do not include any aluminum and SS support members, welds, bolted connections or anchorage to structure for ASME Class 1, 2, 3 or MC components as suggested in the discussion section of Table 1 item 3.5.1-099.

Request:

Clarify whether item 3.5.1 -099 is not applicable (i.e. this component/environment/aging effects does not exist at the site) or not used (i.e. another line item is used to manage the same aging effects for these components). If not applicable, provide a technical justification. Provide revised SLRA Tables 3.5-1 and 3.5-2 accordingly, if necessary.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.4-1 L-2021-144 Attachment 2 Page 2 of 3 NEPB Response:

SLRA Table 3.5-1 item 3.5-1, 099 is not used. Another item, 3.5-1, 100, is used to manage stainless steel supports as indicated. Furthermore, SLRA Table 3.5.2-1, for RC Class 1 supports including RV supports, and SLRA Table 3.5.2-13, for ASME Class 2 and 3 supports identify only steel supports for ASME Class 1, 2 or 3 components and are addressed by other Table 3.5-1 line items (068, 087, 089, and 091 ). In addition, SLRA Supplement 1 clarified these tables and included galvanized steel supports but did not otherwise alter the determination regarding materials of ASME Class 1, 2 or 3 supports.

As such, supports for ASME Class 1, 2 or 3 components at PBN are not stainless steel.

However, there are stainless steel supports for non-ASME components, as well as other non-ASME structural components that are stainless steel or aluminum, which are addressed by Table 3.5-1, item 3.5-1, 100 and 3.5.2.2.2.4.

SLRA Table 3.5-1, item 3.5-1, 099, as amended by Supplement 1, is further clarified in this response.

References:

1. "Point Beach Nuclear Plant Units 1 and 2 Subsequent License Renewal Application (Public Version)," Enclosure 3, Attachment 1, dated November 2020 (ADAMS Accession# ML20329A247)

2. NextEra Energy Point Beach, LLC (NEPB) Letter to NRC L-2021-081 dated April 21, 2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML21111A155)

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.4-1 L-2021-144 Attachment 2 Page 3 of 3 Associated SLRA Revisions:

SLRA Table 3.5-1 (page 3.5-77), as amended by SLRA Aging Management Supplement 1 (Attachment 22), is revised as follows:

rrable 3.5-1 Containment, Structures and Structural Components/Commodities Summary of Aciinc Manaciement Procirams Item Component Aging Effect Aging Further Evaluation Discussion Number Requiring Management Recommended Management Program 3.5-1, 099 Aluminum, stainless steel Loss of material due to ~MP Xl.M32, rYes (SRPSLR Not used.

support members; welds; pitting and crevice "One Time Section 3.5.2.2.2.4) bolted connections; corrosion, cracking Inspection," AMP Loss of material and cracking of non-support anchorage to due to sec

~l.S3,

l\\SME stainless steel supports and building structure "ASME non-ASME stainless steel or Section XI, aluminum supports and other Subsection IWF,"

structural components isare addressed or AMP Xl.M36, in item 3.5-1, 100.

"External Surfaces Further evaluation is documented in Monitoring of Section 3.5.2.2.2.4.

Mechanical Components Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.4-2 L-2021-144 Attachment 3 Page 1 of 9 RAI 3.5.2.2.2.4-2

Background:

Table 3.5.2-1 AMR line item, "Liners (refueling cavity) and covers (sand box, Unit 1 sump A strainer)" components "Stainless steel" material associate with Table 1 item 3.5.1-100, describes several intended functions including "pressure boundary." The AMR line item identifies the Structures Monitoring program to manage aging effect.

As discussed in NUREG-2221, the staff finds the use of visual inspection for the AMR line items acceptable when performing periodic inspections to manage the aging effects for structural supports, and when it can easily be demonstrated that, for these type of structural supports, minor loss of material or cracking that might not be visually detectable during a walkdown inspection will likely not impact the intended function of the support.

Issue:

It is not clear whether regular visual inspections through the Structural Monitoring program are sufficient to detect small cracks from SCC (usually too small to be seen) to retain the pressure boundary intended function associated with the components described above.

Request:

Clarify if "pressure boundary" is an intended function for each of the components associated with the line item. If so, justify how regular visual inspections are adequate to detect small cracks from sec so that the pressure boundary intended function will be maintained for the subsequent period of extended operation.

NEPB Response:

Pressure boundary is not an intended function for the identified stainless steel covers (for the sandbox, and Unit 1 sump A strainer) associated with Table 3.5-1, line item 3.5-1, 100. The function of these covers is shelter, protection. A 2010 modification installed a cover above the A strainer in Unit 1 recirculation sump to protect the A strainer from debris that could be washed down with containment spray activation. The configurations of the B strainer in Unit 1 and both strainers in Unit 2 do not require such protection. The sandbox covers around the RPV flange allow for inspection of the top of the RPV support and reactor cavity.

As listed in SLRA Tables 3.5.2-1, 3.5.2-8, 3.5.2-11, 3.5.2-13, and 3.5.2-14, clarified by Supplement 1, other components associated with Table 3.5-1, item 3.5-1, 100 also do not have pressure boundary functions. Supplement 1 Attachment 29 also clarified the aging effects and aging management programs for the external surface of the fuel transfer tube, which does have a pressure boundary function, removing the association with Table 3.5-1, item, 3.5-1, 100.

The refueling cavity liner has a pressure boundary function along with direct flow, and radiation shielding. The fire barrier function assigned to this component was mistakenly

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.4-2 L-2021-144 Attachment 3 Page 2 of 9 carried over from the interior walls to which it is attached. In addition, the refueling cavity liner should have been assigned a treated borated water environment similar to the reactor cavity seal ring and other refueling components.

To clarify this oversight, the following updates to the SLRA, as amended by SLRA Aging Management Supplement 1, are provided:

Table 2.4 clarify component types and associated functions Section 3.5.2.2.2.4 - remove (refueling cavity) liner Table 3.5-1, item 100 discussion - remove (refueling cavity) liner and Table 3.5.2 replace the existing 'Liners (refueling cavity) and covers (sand box, Unit 1 sump A strainer)' item and clarify pertinent plant specific notes.

Additionally, the generic notes for Water Chemistry and One Time Inspection AMPs are updated for consistency with other (mechanical) items that credit these AMPs. In the process of these clarifications, an additional minor clarification was identified in Table 3.5.2-1 (as updated in Supplement 1, Attachments 21 and 29) to clarify the generic note used for sliding surfaces as the IWF AMP has an exception.

References:

1. "Point Beach Nuclear Plant Units 1 and 2 Subsequent License Renewal Application (Public Version)," Enclosure 3, Attachment 1, dated November 2020 (ADAMS Accession# ML20329A247)

2. NextEra Energy Point Beach, LLC (NEPB) Letter to NRC L-2021-081 dated April 21, 2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML21111A155)

Associated SLRA Revisions:

SLRA Table 2.4-1, Section 3.5.2.2.2.4, Table 3.5-1, item 100, and Table 3.5.2-1, as amended by SLRA Aging Management Supplement 1 (Attachment 22 (Section 3.5.2.2.2.4, and Table 3.5-1 ), and Attachment 29 (Table 2.4-1 and Table 3.5.2-1 )), are revised.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.4-2 L-2021-144 Attachment 3 Page 3 of 9 SLRA Table 2.4-1 (pages 2.4-5, 2.4-6), as amended by SLRA Aging Management Supplement 1 (Attachment 29), is revised as follows:

Table 2.4-1 Containment Structure and Internal Structural Components Subject to Aging M

t R ana~emen ev1ew Component Type Intended Function(s)

IAir lock, equipment hatches and accessories Fire barrier Pressure boundary Concrete Foundation I Basemat Direct flow Pressure boundary Structural suooort Concrete Walls, Buttresses, Dome and Ring Girder Fire barrier Flood barrier Missile barrier Pressure boundary Shelter, protection Structural support Concrete Internal Columns, Beams, Slabs and Walls Fire barrier Flood barrier Missile barrier Shelter, protection Structural support Concrete Tendon Gallery Walls Shelter, protection Construction truss Structural support H-Piles Structural support Fuel transfer tube (including penetration sleeves, Fire barrier expansion joints and blind flange)

Pressure boundary Radiation shielding Structural support II '--*-

.:... \\

,J I 1-:< ~ -

, __ I\\ Fire barrier

,_ - - \\' -

'~J JI -*,_

l".f........... ~-

. -\\

Pressure boundary I

n-..J:-<=--

~h '- 1..J : --

Liners (reactor cavity)

Radiation shielding Structural suooort Liner plate Direct flow Fire barrier Pressure boundary Structural suooort Liner plate and keyway channels Direct flow Pressure boundary Structural support Liner plate anchors and attachments Pressure boundary Structural suooort Liner plate moisture barrier (sealing compound)

Shelter, protection Miscellaneous structural components1 Structural support Penetration assemblies (elastomer)

Pressure boundary Structural support

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.4-2 L-2021 -144 Attachment 3 Page 4 of 9 1 Miscellaneous structural components inside Containment include ladders, stairs, handrails, gratings such as Unit 1 flow diverters, platforms, and the RV core barrel support stand.

Table 2.4-1 Containment Structure and Internal Structural Components Subject to Aging M

R anaaement ev1ew Component Type Intended Function(s)

Penetration assemblies (Electrical)

Fire barrier Pressure boundary Structural suooort Penetration assemblies (Mechanical)

Pressure boundary Structural support Penetration sleeves (Electrical)

Pressure boundary Structural suooort Penetration sleeves (Mechanical)

Pressure boundary Structural sunnort Pressure retaining bolting Pressure boundary Structural suooort Primary shield wall (and biological shield wall)

Radiation shielding Shelter, protection Structural sunnort D--1:........ ~ -*- ---*

,..L'

• -1.....

~iFe l:laFFieF

........... '- - ':::J}

'IJ

'~

RC Class 1 supports Structural support RC Class 1 support bolting Structural support Reactor cavity seal ring Pressure boundary Refuelina cavitv liner Direct flow Pressure boundaty Radiation shieldina Refueling components (containment upender, davit arm)

Structural support Sand box sumo strainer A IUnit 1 l cover Shelter1 12rotection Service Level I coatings Maintain adhesion Sliding surfaces Structural support

[Tendons (posttensioning system)

Structural support tTendon anchorage and attachments Pressure boundary Structural support tThermal Insulation (high temperature penetrations)

Insulate (thermal)

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.4-2 L-2021-144 Attachment 3 Page 5 of 9 SLRA Section 3.5.2.2.2.4 (page 3.5-32), as amended by SLRA Aging Management Supplement 1 (Attachment 22), is revised as follows:

used, the SLRA states the specific alloy or temper used for the applicable inscope components.

Cracking due to sec and loss of material due to pitting and crevice corrosion is possible in stainless steel and aluminum structural components exposed to any air, condensation, or underground environment where sufficient halides (e.g., chlorides) and moisture are present, and for tank foundation anchor bolts where water may collect. The air environment for stainless steel new fuel storage racks, (refueling cavity) liner, sandbox and ECCS strainer covers, supports or anchorage or aluminum manhole covers, fire barrier penetration seals and fire wraps, or insulation jacketing is not expected to be aggressive enough, in rural Wisconsin, to cause cracking or localized loss of material for stainless steel or aluminum exposed to indoor or outdoor air in the presence of wetting.

In addition, stainless steel structural components are limited in number in comparison to the amount of stainless steel mechanical components and there are no aluminum supports. Furthermore, there has been no site operating experience of cracking or localized corrosion of stainless steel or aluminum SSCs. As such, cracking due to sec and loss of material due to pitting and crevice corrosion is conservatively an applicable aging effect at PBN for stainless steel and aluminum and is managed with the External Surfaces Monitoring of Mechanical Components (B.2.3.23) AMP, which will interface with the Structures Monitoring (B.2.3.34) AMP, the Fire Protection (B.2.3.15) AMP and the ASME Section XI, Subsection IWE (B.2.3.29) AMP if degradation is detected in the mechanical components.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.4-2 L-2021-144 Attachment 3 Page 6 of 9 SLRA Table 3.5-1, item 100 (page 3.5-78), as amended by SLRA Aging Management Supplement 1 (Attachment 22), is revised as follows:

Table 3.5-1 Containment, Structures and Structural Components/Commodities Summary of Aain!= Manaaement Proarams Item Component Aging Effect Aging Further Evaluation Discussion Number Requiring Management Recommended Management Program 3.5-1, 100 Aluminum, stainless steel Loss of material due to ~MP Xl.M32,

!Yes (SRPSLR Consistent with NUREG2191, as support members; welds; pitting and crevice "One Time Section 3.5.2.2.2.4) clarified.

bolted connections; corrosion, cracking Inspection," AMP The External Surfaces Monitoring of support anchorage to due to sec IX1.S6, building structure "Structures Mechanical Components (B.2.3.23) AMP Monitoring," or is credited with managing loss of material

~MP Xl.M36, and cracking of stainless steel and "External aluminum insulation jacketing. The Surfaces Structures Monitoring (B.2.3.34) AMP is Monitoring of credited with managing loss of material Mechanical and cracking of stainless steel new fuel Components" storage racks (refueling cavity) liners, sandbox, anchorages and ECCS strainer covers, as well as aluminum manhole covers exposed to air. The ASME Section XI, Subsection IWE (B.2.3.29)

~MP is credited with managing loss of material for the stainless steel

~ransfer tube. The Fire Protection (B.2.3.15) AMP is credited with managing loss of material in stainless steel fire barrier penetrations, and rollup door as well as aluminum (foil) and stainless steel {bands) used in fire wraps.

Further evaluation is documented in Section 3.5.2.2.2.4.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.4-2 L-2021-144 Attachment 3 Page 7 of 9 SLRA Table 3.5.2-1 (page 3.5-88), as amended by SLRA Aging Management Supplement 1 (Attachment 29) is revised as follows:

able 3.5.2-1: Containment Buildina Structure and Internal Structural Comoonents Summarv of Aaina Manaaement Evaluation Component I Intended I Material I Environment I Aging Effect IAging Management' NUREG2191 j Table 1 I Notes Type Function Requiring Program Item Item Manaaement E, 4...11 E, 4...11

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.4-2 L-2021-144 Attachment 3 Page 8 of 9 SLRA Table 3.5.2-1 (pages 3.5-93 to 3.5-94), as amended by SLRA Aging Management Supplement 1 (Attachment 29),

is revised as follows:

Table 3.5.2-1: Containment Building Structure and Internal Structural Comoonents Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-2191 Table 1 Notes Type Function Requiring Program Item Item Manaaement RC Class 1 Structural High IAir - indoor Loss of preload IASME Section XI, lll.B1.1.TP229 3.5-1, 087 B, 8 support bolting support strength uncontrolled Subsection IWF steel B.2.3.31 )

RC Class 1 Structural High Air-indoor Cracking IASME Section XI, lll.B1.1.TP41 3.5-1, 068 B, 8 support bolting support strength uncontrolled Subsection IWF steel B.2.3.31 )

Reactor cavity Pressure Stainless lrreated Loss of material OneTime Inspection Vll.A2.A99 3.3-1, 125 c

seal ring Boundary steel borated (B.2.3.20) water Water Chemistry D

'B.2.3.2)

Refuelina Direct flow Stainless Treated Loss of material OneTime lnsnection Vll.A2.A99 3.3-1 125 f

cavitv liner Pressure

~

borated

'B.2.3.20\\

boundarv

~

Water Chemistrv Q

Radiation 8.2.3.2\\

shieldina Refueling Structural Stainless Treated Loss of material OneTime Inspection Vll.A2.A99 3.3-1, 125 f

components support steel borated (B.2.3.20)

(containment water Water Chemistry D

upender, davit (B.2.3.2) arm)

Sandbox sumo Shelter Stainless Air-indoor Loss of material Structures Monitorina lll.83.T37b 3.5-1 100 E 4 11 strainer A lunit orotection steel uncontrolled B.2.3.34\\

1) cover Sandbox sumo Shelter Stainless Air-indoor Crackina Structures Monitorina lll.83.T37b 3.5-1 100 E 4 11 strainer A lunit orotection steel uncontrolled B.2.3.34\\

1) cover Service Level I Maintain Coatings IAir - indoor Loss of coating or Protective Coating ll.A3.CP152 3.5-1, 034 A

coatings adhesion uncontrolled lining integrity Monitoring and Maintenance (B.2.3.36)

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.4-2 L-2021-144 Attachment 3 Page 9 of 9 Sliding Surfaces Structural Lubrite Air-indoor support uncontrolled Associated

Enclosures:

None.

Loss of mechanical

~unction

~

lll.B21.1.TP4}5 3.5-1,

A§ 10.2.:.:"; ASME 0745 Section XI, Subsection IWF (B.2.3.31 )

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.2.1-1 L-2021-144 Attachment 4 Page 1 of 4

3. SLRA Section 8.2.2.1, "Fatigue Monitoring" RAI B.2.2.1 -1 Regulatory Basis:

10 CFR 54.21 (a)(3) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

Background:

The "parameters monitored or inspected" program element of GALL-SLR AMP X.M1, "Fatigue Monitoring," states that the program monitors all applicable plant transients that cause cyclic strains and contribute to fatigue, as specified in the fatigue analyses, and monitors or validates appropriate environmental parameters that contribute to Fen values. SLRA Section B.2.2.1 addresses the Fatigue Monitoring program as a consistent program with GALL-SLR AMP X.M1.

SLRA Section B.2.2.1 states that SLRA Table 4.3.1-1 identifies the design cycles utilized in the component fatigue analyses and concludes that the projected cycles through the subsequent period of extended operation (SPEO) will not exceed the design cycles assumed in the analyses. In a similar manner, the fatigue transients and their allowable cycle numbers are provided in SLRA Appendix A, UFSAR supplement, Table 4.1 -8.

Issue:

Generally speaking, the UFSAR supplement table in SLRA Appendix A is less comprehensive than SLRA Table 4.3.1 -1 in terms of the design transients identified in the tables. For example, SLRA Appendix A, UFSAR supplement, Table 4.1-8 does not identify the accumulator safety injection, loss of charging flow, loss of letdown flow, or pressurizer heatup transient as a design transient, while these transients are included in SLRA Table 4.3.1-1.

In addition, SLRA Appendix A, UFSAR supplement, Table 4.1-8 specifically identifies the more limiting allowable cycle numbers for RVI baffle bolts (also called baffle former bolts) as well as the general limits to the design transient cycle numbers that are applied to the other reactor vessel internal (RVI) and piping components. In contrast, the

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.2.1-1 L-2021 -144 Attachment 4 Page 2 of 4 specific allowable cycle numbers for the baffle former bolts are not described in SLRA Table 4.3.1-1.

Request:

1. Reconcile the difference between SLRA Table 4.3-1 and SLRA Appendix A, Table 4.1 -8 regarding the allowable transient cycles for the RVI baffle former bolts.

2. Explain why the design transients listed in SLRA Table 4.1.3-1 and SLRA Appendix A, Table 4.1-8 are different. If the difference cannot be justified, identify a consistent design transient table for both SLRA Section 4.3.1 and SLRA Appendix A.

NEPB Response:

1. SLRA Table 4.3.1-1 Note 1 inadvertently omitted the baffle bolt limitations contained in SLRA Appendix A, Table 4.1-8. Note 2 is added to reflect the UFSAR table baffle bolt design cycles. This supersedes the information in SIA Report Number 2000088.401, Revision 2 Table 2.1-3.

2. The design transients listed in SLRA Appendix A, Table 4.1-8 titled "THERMAL AND LOADING CYCLES" only apply to RCS design transients used for equipment design purposes and are not intended to reflect operating experience. SLRA Table 4.3.1-1 titled "80-Year Projected Cycles - PBN Units 1 and 2" includes design cycles from Table 4.1-8, additional system transients used in the fatigue analysis of components that are subject to these transients (e.g., Accumulator, Aux Spray and HPSI Injections, loss of charging, loss of letdown, etc.) and includes transient cycle projections and allowable cycles for 80 years of operation based on plant operating experience. Accordingly, there is no need to make a consistent design transient table.

References:

None.

Associated SLRA Revisions:

SLRA Table 4.3.3-1, page 4.3-7, is revised as follows:

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.2.1-1 L-2021-144 Attachment 4 Page 3 of 4 Table 4.3.1-1 80-Year Projected Cycles - PBN Units 1 and 2 Current Cycles 80-year Design Year-end 2019 Allowable 80-year Allowable Item Transie Unit 1 Unit 2 Projection Cycles Cycles 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 nt 10% Step Load Decrease 25 30 44 2000 100 10% Step Load Increase 0

1 2

2000 20 50% Step Load Decrease 46 20 66 200 100 Accumulator Safety Injection 4

1 7

89 8

Auxiliary Spray Actuation 0

0 0

10 2

HPSI Injection 2

0 4

89 4

Inadvertent Accumulator Slowdown 0

0 0

4 2

Inadvertent RCS Depressurization 0

0 0

20 2

Loss of Charging Flow 16 17 34 60 50 Loss of Letdown Flow 20 17 46 200 75 Pressurizer Cooldown 78 61 115 200 120 Pressurizer Heatup 79 62 116 200 120 Primary Side Hydrostatic Test 1

2 3

5 3

Primary Side Leak Test 36 38 57 94 60 Primary to Secondary Leak Test 2

9 14 27 15 RCS Cooldown 79 62 118 200 120 RCS Heatup 80 63 119 200 120 RPV Safety Injection 0

0 0

89 2

Reactor Trip 68 52 107 300 120 Refueling 49 46 77 80 80 Relief Valve Actuation 1

3 7

100 8

Secondary to Primary Leak Test 38 33 57 128 60 Trip Due to Loss of RCP 1

2 4

100 4

Unit Loading 5%/min 1691 1806 2478 11600 (Note 2) 8000 (Note 1 and 2)

Unit Unloading 5%/min 1544 1670 2295 11600 {Note 2} 8000 (Note 1 and 2)

FW Cycling at Hot Standby NC NC N/A 2000 Not provided Boron Concentration Eq.

NC NC N/A 23360 Not provided Loss of Load (Trip)

NC NC N/A 80 Loss of Power (Trip)

NC NC N/A 40 Loss of Flow (Trip)

NC NC N/A 80 Turbine Roll Test NC NC N/A 10 Control Rod Drop NC NC N/A N/A Excessive FW Flow NC NC N/A N/A OBE NC NC N/A N/A Notes 1 - Specific 80-year allowable cycles are limited to the following values for specific RCS components due to environmentally-assisted fatigue (EAF) values presented in Table 4.3.4-1 :

(a) CROM upper latch housings are limited to 2700 loading and unloading cycles at 5%/min 80 40 80 10 80 30 10 (b) Vessel flanges are limited to 5000 loading and unloading cycles at 5%/min (c) The fatigue crack growth (FCG) analysis of longitudinal flaws in reactor

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.2.1-1 L-2021-144 Attachment 4 Page 4 of 4 coolant loop cast austenitic stainless steel piping components (Reference 4.8.15) utilizes a limit of 3,000 loading and unloading cycles.

2 - Cycle limit for reactor vessel internal baffle bolts is 2485 Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.2-1 L-2021-144 Attachment 5 Page 1 of 2

4. SLRA Section 8.2.3.2, "Water Chemistry" Regulatory Basis:

10 CFR 54.21 (a)(3) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

RAI B.2.3.2-1

Background:

The last row of SLRA Table 3.3.2-1, "Chemical Volume and Control - Summary of Aging Management Evaluation," is for loss of material for a stainless steel valve body in treated water. The NUREG-2191 Item listed is Vlll.01.SP-87, and the SLRA Table 1 Item listed is 3.3-1, 085. The use of Item 3.3-1, 085 appears to be an error because it corresponds to hardening or loss of strength for elastomer materials.

Issue:

The reference to Table 1 Item 3.3-1, 085 in the last row of SLRA Table 3.3.2-1 appears to be an error because it applies to elastomer materials and would not effectively manage loss of material for stainless steel valve bodies.

Request:

Please provide the correct information for the entries in the last row of Table 3.3.2-1, Page 3.3-103 of the SLRA.

NEPB Response:

The reference to Table 1 Item 3.3-1, 085 in the last row of SLRA Table 3.3.2-1 is corrected to Item 3.4-1, 085.

References:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.2-1 L-2021-144 Attachment 5 Page 2 of 2 Associated SLRA Revisions:

SLRA Table 3.3.2-1 (page 3.3-103) is revised as follows:

Table 3.3.2-1: Chemical and Volume Control - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Type Function Requiring Management Management Program Valve body Pressure Stainless steel Treated water Loss of material Water Chemistry boundary (int)

(B.2.3.2)

One-Time Inspection (B.2.3.20)

Associated

Enclosures:

None.

NU REG-Table 1 Notes 2191 Item Item Vlll.01.SP-

~ A 87 we 3.4-1, 085

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.8-1 L-2021-144 Attachment 6 Page 1 of 2

5. SLRA Section 8.2.3.8, "Flow Accelerated Corrosion" Regulatory Basis Section 54.21 (a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the U.S. Nuclear Regulatory Commission (NRC) staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

RAI 8.2.3.8-1 (Revision Level of NSAC-202L)

Background:

During the audit of the Flow-Accelerated Corrosion (FAC) program, the NRC staff sought clarification on which revision of EPRI 3002000563, "Recommendations for an Effective Flow-Accelerated Corrosion Program" (NSAC-202L), the Point Beach Nuclear Plant (PBN) FAC program incorporates.

Commitments a) and d) for No. 12, Flow-Accelerated Corrosion (16.2.2.8), in Subsequent License Renewal Application (SLRA) Table 16-3, "List of SLR Commitments and Implementation Schedule;" and the Enhancements to Element 1, "Scope of Program," and Element 4, "Detection of Aging Effects," as described in SLRA Section B.2.3.8, "Flow-Accelerated Corrosion," were revised in SLRA Supplement 1, dated April 21, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. 21111A155), to reference Revision 4 of NSAC-202L.

Issue:

Section 16.2.2.8, "Flow-Accelerated Corrosion," of the Updated Final Safety Analysis Report (UFSAR) Supplement in Appendix A of the SLRA states that the PBN FAC program is based on Revision 4 of NSAC-202L. However, Reference 61 in Section 16.5, "References," of the UFSAR Supplement is for Revision 3 of NSAC-202L.

Request:

Please provide a basis to justify the discrepancy between Sections 16.2.2.8 and Reference 61 in Section 16.5 of the UFSAR with regards to the revision of NSAC-202L.

Alternatively, revise Section 16.5 of the UFSAR Supplement to be consistent with the changes previously made for the revision level of NSAC-202L.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.8-1 L-2021-144 Attachment 6 Page 2 of 2 NEPB Response:

The PBN Flow Accelerated Corrosion AMP is based on Revision 4 of NSAC-202L.

Reference 61 in Section 16.5, "References," of the UFSAR Supplement is revised to Revision 4 of NSAC-202L for alignment with program documents and Section 16.2.2.8, "Flow-Accelerated Corrosion," of the Updated Final Safety Analysis Report.

References:

None.

Associated SLRA Revisions:

SLRA Appendix A, Section 16.5 is amended as follows:

61.

EPRI, NSAC-202L, Revision 1~. "Recommendations for an Effective Flow-Accelerated Corrosion Program (30020005631015425)," Electric Power Research Institute, Palo Alto, California, Nuclear Safety Analysis Center (NSAC),

November 26, 2013August 10, 2007.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.8-2 L-2021-144 Attachment 7 Page 1 of 3 RAI B.2.3.8-2 (Software Quality Assurance)

Background:

The "scope of program" program element for Aging Management Program (AMP)

Xl.M17, "Flow-Accelerated Corrosion," in Volume 2 of NUREG-2191, "Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report - Final Report" (ADAMS Accession No. ML17187A204), states that the program, described by NSAC-202L, includes procedures and administrative controls to assure that structural integrity is maintained for piping components. Section 3.1, "Governing Document," of Revision 4 of NSAC-202L, recommends the inclusion of quality assurance requirements.

During the audit of the PBN FAC program, the NRC staff noted that Revision 13 of IM-AA-101, "Software Quality Assurance Program," provides the essential elements to meet the quality assurance standards established in the Quality Assurance Topical Report. Revision 13 of IM-AA-101 also defines four levels (A through D) of software classification based on the task for which the output is to be used. In addition, during the audit, the staff noted that Revision 4 of ER-AA-111-1000, "Flow-Accelerated Corrosion (FAC) Activities," mentions other software that may be used for analysis, such as FAC Manager and IDDEAL. For post-inspection activities, this procedure also states that ultrasonic test analyses may be performed using trending software analysis functions other than those built into CHECWORKS'.

The NRC staff notes ER-AA-111, "Flow-Accelerated Corrosion (FAC) Program,"

Revision 3, states that ultrasonic inspection data should be evaluated using an approved (i.e., validated and verified) software program. The applicant stated during the audit of the PBN FAC program that the software quality assurance classification for CHECWORKS' was Level C (which corresponds to Business Critical in procedure IM-AA-101 ). The staff also notes that Level C software does not require "Software Verification and Validation" in all instances and does not require "Error Reporting and Corrective Action" as provided in ER-AA-101, Sections 5.12 and 5.14, respectively. The staff further notes that during a recent, previous license renewal review, a number of inspection deferrals were required because the FAC Manager software inaccurately designated components as requiring inspections in the next outage.

Issue:

The SLRA, Revision 3 of ER-AA-111, and Revision 4 of ER-AA-111-1000 do not state whether software other than CHECWORKS' is used in the PBN FAC program. In addition, none of these documents identify the software quality assurance classification for CHECWORKS' or any other software used in the PBN FAC program. Also, although ER-AA-111 states that the software evaluating the ultrasonic inspection data should be validated and verified, it does not include IM-AA-101 as a developmental reference for accomplishing this, and because CHECWORKS' is classified as Level C software, it does not require validation and verification. Further, error notification for

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.8-2 L-2021-144 Attachment 7 Page 2 of 3 other software used in the PBN FAC program (e.g., FAC Manager) would not be required if comparably classified as Level C software.

Request:

1. Please identify all software products used in the PBN FAC program.

2. Please provide the software quality assurance classification and the bases for the classification for each software product used in the PBN FAC program. If verification and validation or error notification activities are currently being performed but are not required by the software classification level, provide information regarding assurances that these activities will be continued during the subsequent period of extended operation.

NEPB Response:

The following numbered responses correspond to the numbered requests in the RAI.

1.

The software programs utilized in the PBN FAC AMP are CHECWORKS' Steam/Feedwater Application (SFA) and FAC Manager Web Edition (FMWE).

IDDEAL has been retired and no longer used.

2.

CHECWORKS' SFA and FMWE are classified as SQA Level C software per IM-AA-101 for the current licensing basis. This same classification is also applicable throughout the subsequent period of extended operations (SPEO) and the basis is documented on the two respective software classification determination (SCD) forms. Appropriate classification of software is the responsibility of the applicable Nuclear Fleet department head or designee. Per IM-AA-101, SQA Level C, the software is considered the following:

"Regulatory (Not Meeting Level B) I Business Critical I Plant Reliability I Other Non Safety-Related Process Control (Not Meeting Level B) and I or monitoring not associated with operational decisions."

Although not specifically required for Category C software, error notification activities are captured as responsibilities of the FAC Program Fleet Engineer as noted in ER-AA-111-1000:

Section 3.4.J, "Maintain and update FAC Program elements, including the Program Basis Document, CHECWORKS' Models, Susceptible Non-Modeled (SNM) Program, and the FAC Program database in use of inspected components."

Section 3.4.M, "Maintain awareness of industry experiences and practices regarding FAC, and shares experience with the industry."

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.8-2 L-2021-144 Attachment 7 Page 3 of 3 Section 4.2.A.(1)d, "Industry experiences should be considered when selecting components for FAC inspections, including the following :

o Reported failures or other experience. (e.g., CHECWORKS' Users Group (CHUG) information including Plant Experience Reports, Industry Reporting and Information System (IRIS), NRC communications, etc."

The above activities will be continued during the SPEO. Although not specifically required, verification and validation {V&V) is performed on a 5-7 year frequency, or as needed after any major modifications to the plant, to ensure the Checworks ' SFA input and functionality are correct. Some of the inputs include global data, component level data, and replacement and installation data. A V&V was recently performed in January 2019.

References:

None.

Associated SLRA Revisions:

None.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.8-3 L-2021-144 Attachment 8 Page 1 of 5 RAI B.2.3.8-3 (AMR Items)

Background:

NUREG-2191, Volume 1 identifies wall thinning due to FAC for steel piping, piping components exposed to treated water (e.g., GALL-SLR item Vlll.F.S-16, SRP item 3.4.1-005). In addition, NUREG-2191, Volume 1 identifies wall thinning due to erosion for metallic piping, piping components exposed to treated water (e.g., GALL-SLR item Vlll.B1.S-408, SRP item 3.4.1-060).

Issue:

SLRA Table 3.4.2-1, "Main and Auxiliary Steam - Summary of Aging Management Evaluation," does not cite wall thinning - FAC or wall thinning - erosion for the following carbon steel components exposed to a treated water (int [internal]) environment:

Drain trap Flow element Piping Piping and piping components Valve body In addition, SLRA Table 3.4.2-1 does not cite wall thinning - erosion for the following components exposed to a treated water (int) environment:

Stainless steel drain trap Stainless steel piping Stainless steel piping and piping components Copper alloy valve body SLRA Table 3.4.2-2, "Feedwater and Condensate - Summary of Aging Management Evaluation," does not cite wall thinning - FAC and/or wall thinning - erosion for the following carbon steel components (intended function in parenthetical) exposed to a treated water (int) environment:

Piping (Leakage boundary (spatial))

Piping and piping components (Pressure boundary)

Piping and piping components (leakage boundary (spatial))

In addition, SLRA Table 3.4.2-2 does not cite wall thinning - erosion for stainless steel piping and piping components (all intended functions) exposed to a treated water (int) environment.

During the audit of the PBN FAC program, the applicant stated that many of these components were aligned with a April 29, 2005, response to a request for additional information related to PBN's original license renewal application (ADAMS Accession

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.8-3 L-2021-144 Attachment 8 Page 2 of 5 No. ML051300355). However, the NRC staff did not find these specific components, materials, and environments addressed as part of the Main and Auxiliary Steam System or the Feedwater and Condensate System in the April 29, 2005, response.

Request:

Please discuss, with justification as appropriate, whether wall thinning due to FAC and/or erosion is an applicable aging effect requiring management for the components identified above in the Main and Auxiliary Steam System and Feedwater and Condensate System exposed to a treated water (internal) environment.

NEPB Response:

Carbon steel piping and piping components, and valve bodies in the Main and Auxiliary Steam systems are subject to high flow, and therefore, susceptible to wall thinning due to erosion and FAC. Copper alloy valve bodies and stainless steel piping in the Main and Auxiliary Steam systems are also subject to high flow, and therefore, susceptible to wall thinning due to erosion. SLRA Table 3.4.2-1 is revised to include the necessary table items corresponding to the above aging effects.

In the SLRA, the "piping and piping components" commodity group is used with a "pressure boundary" or "leakage boundary (spatial)" intended function as a catchall for components with a certain material that are subject to fatigue. The individual component types (e.g. "piping" or "valves") are listed for non-fatigue related aging effects. For the "structural integrity (attached)" intended function, the "piping and piping components" commodity group is used more broadly to represent all component types that are part of the non-safety related portion of the system that is attached to and provides structural integrity for the safety-related portion of the system. As such, the "piping and piping components" component type with a "structural integrity (attached)"

intended function needs to list all associated aging effects in the SLRA tables. Based on this approach, SLRA Table 3.4.2-1 is missing a wall thinning due to FAC and a wall thinning due to erosion table items for the carbon steel "piping and piping components" component type with a "structural integrity (attached)" intended function so these table items are added to the table.

The stainless steel "piping and piping components" table item with a "pressure boundary" intended function does not need to address wall thinning due to erosion because the individual component types that are stainless steel have associated wall thinning due to erosion table items in SLRA Table 3.4.2-1. In addition, carbon steel drain traps and flow elements along with stainless steel drain traps are not subject to high flow rates and are not susceptible to wall thinning due to erosion and FAC.

Carbon steel piping exposed to treated water in the Feedwater and Condensate systems are subject to high flows and are therefore, susceptible to wall thinning due to erosion and FAC and is managed by the FAC program. A carbon steel piping wall thinning due to erosion table item is already included in SLRA Table 3.4.2-2 but a table item for wall thinning due to FAC has been added. Additionally, stainless steel piping and piping components exposed to treated water in the Feedwater and Condensate

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.8-3 L-2021-144 Attachment 8 Page 3 of 5 systems are subject to high flows and are therefore, susceptible to wall thinning due to erosion only for the "structural integrity (attached)" intended function. SLRA Table 3.4.2-2 is revised to include the above changes.

Consistent with the above discussion stainless steel piping and piping components providing a "pressure boundary" or "leakage boundary (spatial)" intended function are included in SLRA Table 3.4.2-2 only for cumulative fatigue damage so no table items need to be added.

In addition, loss of material and cracking table items for stainless steel "piping and piping components" with a "structural integrity (attached)" intended function are also added to SLRA Table 3.4.2-2 for completeness.

References:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.8-3 L-2021-144 Attachment 8 Page 4 of 5 Associated SLRA Revisions:

SLRA Section 3, Table 3.4.2-1 is revised as follows:

Table 3.4.2-1: Main and Auxiliary Steam - Summary of Agin i Management Evaluation Component Intended Aging Effect Aging Management Type Function Material Environment Requiring Program Management Piping Leakage Carbon Treated Wall Flow-Accelerated bounda!Jl steel water (int}

thinning -

Corrosion (B.2.3.8}

(spatial}

erosion Piping Leakage Carbon Treated Wall Flow-Accelerated bounda!Jl steel water {int}

thinning -

Corrosion {B.2.3.8}

{spatial}

FAC Piping Leakage Stainless Treated Wall Flow-Accelerated bounda!Jl steel water {int}

thinning -

Corrosion {B.2.3.8}

{spatial}

erosion Piping and Structural Carbon Treated Wall Flow-Accelerated piping integrity steel water {int}

thinning -

Corrosion {B.2.3.8}

comoonents (attached) erosion Piping and Structural Carbon Treated Wall Flow-Accelerated piping integrity steel water (int}

thinning -

Corrosion {B.2.3.8}

comoonents (attached)

FAC Valve body Leakage Carbon Treated Wall Flow-Accelerated bounda!Jl Steel water (int}

thinning -

Corrosion f B.2.3.8}

{spatial}

erosion Valve body Leakage Carbon Treated Wall Flow-Accelerated boundaty Steel water {int}

thinning -

Corrosion {B.2.3.8}

{spatial}

FAC Valve body Leakage Copper Treated Wall Flow-Accelerated bounda!Jl alloy water {int}

thinning -

Corrosion {B.2.3.8}

{spatial}

erosion NU REG-Table 2191 Item 1 Item Notes Vlll.01.S-408 3.4-1, A

060 Vlll.01.S-16 3.4-1, A

005 Vlll.01.S-408 3.4-1, A

060 Vlll.01.S-408 3.4-1, A

060 Vlll.01.S-16 3.4-1, A

005 Vlll.01.S-408 3.4-1 t A

060 Vlll.01.S-16 3.4-1, A

005 Vlll.G.S-408 3.4-1, A

060

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.8-3 L-2021-144 Attachment 8 Page 5 of 5 SLRA Section 3, Table 3.4.2-2 is revised as follows:

Table 3.4.2-2: Feedwater and Condensate - Summary of Aging Management Evaluation Intended Aging Effect Aging Component Type Function Material Environment Requiring Management Management Program Piping Leakage Carbon Treated Wall thinning -

Flow-bounda!)l steel water {int}

FAC Accelerated

{spatial}

Corrosion (8.2.3.8)

Piping and piping Structural Stainless Treated Cracking Water Chemist!)l components integritv steel water >140°F

{8.2.3.2}

{attached}

{int}

One-Time Inspection (8.2.3.20)

Piping and piping Structural Stainless Treated Loss of Water Chemist!)l components integri~

steel water {int}

material

{8.2.3.2}

{attached}

One-Time Inspection (8.2.3.20)

Piping and piping Structural Stainless Treated Wall thinning -

Flow-components integritv steel water {int}

erosion Accelerated

{attached}

Corrosion C8.2.3.8)

Associated

Enclosures:

None.

NU REG-Table 2191 Item 1

Notes Item Vlll.E.S-16 3.4-1, A

005 Vlll.01.SP-88 3.4-1,

§_

011 A

Vlll.01.SP-87 3.4-1,

§_

085 A

Vlll.01.S-408 3.4-1, A

060

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.10-1 L-2021-144 Attachment 9 Page 1 of 3

6. SLRA Section 8.2.3.10, "Steam Generators" Regulatory Basis:

Section 54.21 (a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the U.S. Nuclear Regulatory Commission (NRC) staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

RAI 8.2.3.10-1

Background:

Subsequent License Renewal Application (SLRA) Table 3.1.2-5, "Steam Generators -

Summary of Aging Management Evaluation," November 2020 (Agencywide Documents Access Management System (ADAMS) Accession No. ML20329A247), includes Generic Aging Lessons Learned (GALL) - Subsequent License Renewal (SLR) Item IV.D1.RP-368, 3.1-1, 012 for managing loss of material for the carbon steel blowdown piping nozzles and secondary side shell penetrations exposed to treated water by the ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD and Water Chemistry aging management programs (AMPs).

SLRA Supplement 1, dated April 21, 2021 (ADAMS Accession No. 21111A155), revised SLRA Table 3.1.2-5 by adding GALL-SLR Item IV.D1.RP-161, 3.1-1, 072 for managing loss of material for the carbon steel blowdown piping nozzles and secondary side shell penetrations exposed to water by the Steam Generators and Water Chemistry AMPs.

Issue: 0 of SLRA Supplement 1 states, in part, "... SLRA Table 3.1.2-5, Table 3.1-1 Item 072... are revised to clarify that the Steam Generators and Water Chemistry AMPs will manage... loss of material for blowdown piping nozzles and secondary side shell penetrations." No GALL-SLR items related to the carbon steel blowdown piping nozzles and secondary side shell penetrations were deleted in SLRA Supplement 1.

Therefore, it is unclear to the NRC staff whether GALL-SLR Item IV.01.RP-161, 3.1-1, 072 was in addition to or replaced GALL-SLR Item IV.D1.RP-368, 3.1-1, 012.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.10-1 L-2021-144 Attachment 9 Page 2 of 3 Request:

Please clarify if loss of material for the carbon steel blowdown piping nozzles and secondary side shell penetrations exposed to treated water will be managed by the ASME Section XI lnservice Inspection; Subsections IWB, IWC, and IWD; Steam Generators; and Water Chemistry AMPs.

NEPB Response:

Loss of material in the PBN steam generators blowdown piping nozzles and secondary side shell penetrations will be managed using the ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD and the Water Chemistry AMPs using NUREG-2191 line item IV.01.RP-368. Table 3.1.2-5 is revised as shown below to remove NUREG-2191 line item IV.01.RP-161 which was crediting the Steam Generators and Water Chemistry AMPs.

References:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.10-1 L-2021-144 Attachment 9 Page 3 of 3 Associated SLRA Revisions:

SLRA Table 3.1.2-5 page 3.1-100 is revised as follows:

Table 3.1.2-5: Steam Generators - Summan of Ac:iinc:i Mana:tement Evaluation Component Intended Type Function Blowdown piping Pressure nozzles and boundary secondary side shell penetrations BlawaawR Prnss1ue 13i13iR9 R92:2:les bauRaary aRa sesaRaary siae shell

~---

~

Associated Enclosures None.

Material Carbon steel Garbe A steel Aging Effect Aging Environment Requiring Management ManaC1ement ProC1ram Treated water Loss of ASME Section XI material lnservice Inspection, Subsections IWB,

IWC, and IWD (B.2.3.1)

Water Chemistry (B.2.3.2)

Treatea water bass af Steam GeReratars material

{B.2.3.10)

Water Chemistry

{B.2.3.2)

NUREG-2191 Table 1 Item Item Notes IV.01.RP-368 3.1-1, 012 c D

IV.01.RP 1 e1 3.1 1, 072 g

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.10-2 L-2021-144 Attachment 10 Page 1 of 2 RAI 8.2.3.10-2

Background:

SLRA Table 3.1.2-5 includes GALL-SLR Items IV.D1.R-40, 3.1-1, 070 and IV.C2.RP-23, 3.1-1, 088 for managing cracking and loss of material, respectively, for the nickel alloy tube plugs exposed to reactor coolant by the Steam Generators and Water Chemistry AMPs.

During the audit of the Point Beach Nuclear Plant (PBN) Steam Generators program, the NRC staff noted that Revision 15 of Site Engineering Manual 7.11.20, "Eddy Current Testing of the Unit 1 Steam Generators," states that there are three welded alloy 600 plugs in Steam Generator A of PBN Unit 1 that were installed during manufacture. The staff also noted that Revision 21 of plant procedure NP 7.7.17, "Requirements for Steam Generator Primary Side Activities," states that all Alloy 600 plugs were replaced with Alloy 690 plugs during refueling outage 26 (U1 R26) for PBN Unit 1.

The applicant clarified during the audit of the PBN Steam Generators program that only the Alloy 600 plugs in Steam Generator B of PBN Unit 1 were replaced with Alloy 690 plugs during U1 R26, and that there remains three Alloy 600 welded plugs in Steam Generator A of PBN Unit 1.

Issue:

SLRA Section B.2.3.10 states, "Tube plugs installed are fabricated from heat treated lnconel Alloy 690 material." In addition, B.2.3.10 does not state whether there are any Alloy 600 tube plugs installed in the PBN steam generators. Therefore, it is unclear to the NRC staff whether this statement in SLRA Section B.2.3.10 is referring to current and/or future installed tube plugs.

Request:

1. Please confirm that the only Alloy 600 plugs in the PBN steam generators that will be managed by the Steam Generators and Water Chemistry AMPs are the three Alloy 600 welded plugs in Steam Generator A in PBN Unit 1.

2. Please make any necessary changes to SLRA Section B.2.3.10 to clearly describe the tube plugging materials in the PBN steam generators.

NEPB Response:

The only Alloy 600 plugs installed in the PBN steam generators are in the three tubes with Alloy 600 welded plugs installed in PBN Unit 1 steam generator A (i.e., a total of 6 Alloy 600 welded plugs). Note that the original Alloy 600 plugs in steam generator B were replaced in 1990 with a Westinghouse-designed "plug in a plug" (PIP), then subsequently replaced in U1 R26 with Alloy 690 mechanical plugs. Currently, all Alloy 600 plugs in PBN Unit 1 steam generator B have been replaced with Alloy 690 mechanical plugs. SLRA Section B.2.3.10 is revised as shown below to clearly describe the tube plugging materials in the PBN steam generators.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.10-2 L-2021-144 Attachment 10 Page 2 of 2

References:

None.

Associated SLRA Revisions:

SLRA Section B.2.3.10 pages B-91 are revised as follows:

SG tubes not meeting the TS limits for continued operation are removed from service by installation of tube plugs. This plug installation redefines the reactor coolant pressure boundary and loss of steam generator tube plug integrity can impact the ability of the steam generators to perform its intended function if permitted to continue without corrective action. Tube plugs installed are fabricated from heat treated lnconel Alloy 690 material, with the exception of six plugs (in three tubes) in Unit 1 steam generator A that had Alloy 600 welded plugs installed during manufacturing. Regardless of plug material or installation method, all plugs are inspected during periodic steam generator tube inspection to ensure the integrity of the reactor coolant pressure boundary. /\\!though these plugs have a high resistance to primary v1ater stress corrosion cracking (PVVSGC), they are routinely inspected.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.10-3 L-2021-144 Attachment 11 Page 1 of 3 RAI 8.2.3.10-3 Regulatory Basis:

Section 54.21 (a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the U.S. Nuclear Regulatory Commission (NRC) staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

Background :

In the SLRA (Agencywide Documents Access Management System (ADAMS)

Accession No. ML20329A247), Table 3.1.2-5, "Steam Generators - Summary of Aging Management Evaluation," Items IV.D1.R-437 and 3.1-1, 125 are listed for cracking management of nickel alloy U-tubes exposed to treated water >140 degrees Fahrenheit and steam, by the Steam Generators aging management program (AMP).

In Table 3.1-1 of the Standard Review Plan (SRP)- SLR, Item 125 is a new item that addresses cracking due to flow induced vibration and high cycle fatigue of nickel alloy steam generator tubes at support plate locations exposed to secondary feedwater or steam. During initial license renewal applications, high cycle fatigue of steam generator tubes at tube support plate locations was addressed by a time-limited aging analysis and was covered by item 2 in Table 3.1-1 of the SRP-LR.

As a result of both U.S. and international operating experience with high-cycle fatigue failure of steam generator tubes at tube support plates, the NRC issued Bulletin No.

1988-002, Rapidly Propagating Fatigue Cracks in Steam Generator Tubes," ADAMS Accession No. ML031220043, Information Notice 2005-29, "Steam Generator Tube and Support Configuration," ADAMS Accession No. ML052280011, and Information Notice 2007-37, "Buildup of Deposits in Steam Generators," ADAMS Accession No. ML072910750. Westinghouse Electric Company issued a Nuclear Safety Advisory Letter (NSAL 12-7) on September 24, 2012, in response to NRC staff communications for licensees to identify the as-built anti vibration bar insertion depths in applicable steam generators, so as to identify the potential for additional tube fatigue failures.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.10-3 L-2021-144 Attachment 11 Page 2 of 3 A generic analysis of Westinghouse Model 44F steam generators was produced by Westinghouse Electric Company for the Electric Power Research Institute in April 2016 1. The purpose of this report was to define, on a SG model-specific basis, the information required to complete a plant-specific U-bend fatigue analysis consistent with the requirements of NRC Bulletin 88-02."

1. Steam Generator Management Program: Generic Elements of U-Bend Tube Vibration Induced Fatigue Analysis for Westinghouse Model 44F Steam Generators.

EPRI, Palo Alto, CA: 2016. 3002007562.

Issue:

While the generic analysis of the Westinghouse Model 44F steam generators was performed based on the Point Beach Unit 1 steam generators, the report was intended as a reference for utilities to use in completing a plant specific fatigue analysis of potentially susceptible tubes. The results of the plant specific analysis were not discussed in the SLRA or provided as a reference on the portal for NRC staff review.

Request:

Provide the results of the plant-specific fatigue analysis for NRC staff review.

NEPB Response:

In response to NRC requests for plants to identify their as-built anti-vibration bar (AVB) insertion depths in the U-bend area of operating steam generators (SGs) considering domestic and international experiences with high-cycle fatigue cracking described in References 1 through 3, Westinghouse issued NSAL 12-7 (Reference 4) to alert plants of the potential for high-cycle fatigue cracking based on as-built AVB insertion depth and local flow conditions. Reference 5 concluded that plants affected by the high-cycle fatigue issue are limited to Westinghouse-designed SGs built prior to 1987. The Point Beach Unit 1 Westinghouse Model 44 SGs were listed as being potentially affected, while the Point Beach Unit 2 replacement SGs were not identified as a potentially affected SG design. Reference 5 recommended that affected plants (i.e., Point Beach Unit 1) perform as-built U-bend AVB insertion depth mapping to identify locations where the AVB supports may not be inserted to the design specifications and to identify AVB insertion conditions that could lead to locally increased flows conditions (i.e., flow peaking). Should the AVB insertion depth mapping identify either condition of AVB supports not inserted to depth as designed or conditions of flow peaking, Reference 5 recommended to perform an analytical evaluation to determine the potential for high-cycle fatigue failure.

An AVB insertion depth mapping effort was completed for the Point Beach Unit 1 Model 44F SGs (Reference 6). The insertion depth mapping confirmed that all AVB supports satisfied the design specification required depth of insertion and all tubes are supported in the U-bend region as designed. All but two AVB supports penetrated to at least two or more tubes deeper than the design specification criteria thus providing margin to the design criteria for AVB tube support. The lowest tube row of unsupported tubes was two rows below the design criteria and was for only two tubes. The potential for flow-

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.10-3 L-2021-144 Attachment 11 Page 3 of 3 induced vibration and high-cycle fatigue is greatly reduced the deeper into the bundle the AVB supports are inserted, especially to tube rows lower than the design criteria row. The as-built AVB insertion configurations were generally uniform without significant local insertion depth variations that could lead to flow peaking or flow channeling to exacerbate flow-induced vibration conditions. Since all as-built AVB insertion depths satisfied the design specification criteria and no local flow peaking conditions are present, a site-specific high-cycle fatigue analysis was not required for the Point Beach Unit 1 Model 44F SGs.

The Point Beach Unit 2 SGs were replaced in 1997 with Westinghouse Model Delta 47 replacement SGs. The replacement SGs contain advanced design features, thermally treated Alloy 690 tubing and improved materials of construction. The Point Beach Unit 2 replacement SGs are not subject to the high cycle fatigue experiences described in References 1 through 3 and are not listed as a potentially affected plant by Westinghouse NSAL 12-7 (Reference 4).

References:

1. U.S. NRC Bulletin 88-02, "Rapidly Propagating Fatigue Cracks in Steam Generator Tubes," February 5, 1988. (ADAMS Accession No. ML031220043).

2. U.S. NRC Information Notice 2005-29, "Steam Generator Tube and Support Configuration," October 27, 2005. (ADAMS Accession No. ML052280011).

3. U.S. NRC Information Notice 2007-37, "Buildup of Deposits in Steam Generators,"

November 23, 2007. (ADAMS Accession No. ML072910750).

4. Westinghouse Nuclear Safety Advisory Letter NSAL-12-7, Revision 0, "Insufficient Insertion of Anti-Vibration Bars in Alloy 600TT Steam Generators with Quatrefoil Tube Support Plates," September 2012.

5. Electric Power Research Institute (EPRI) Technical Report 3003007562, "Steam Generator Management Program: Generic Elements of U-bend Tube Vibration Induced Fatigue Analysis for Westinghouse Model 44F Steam Generators," April 2016. EPRI Palo Alto, CA. (EPRI Proprietary).

6. Westinghouse Document SG-SGMP-13-23, Revision 0, "Determination of the As-Built AVB Depth Insertion in the Point Beach Unit 1 Steam Generators," November 2013. (Westinghouse Proprietary).

Associated SLRA Revisions:

None.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.16-1 L-2021-144 Attachment 12 Page 1 of 4

7. SLRA Section B.2.3.16, "Fire Water System" Regulatory Basis:

Section 54.21 (a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the U.S. Nuclear Regulatory Commission (NRC) staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

RAI 8.2.3.16-1 (Fire Protection System Components Subject to Wet-Dry Cycle)

Background:

NUREG-2191, "Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report," Volume 2 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17187A204), states portions of water-based fire protection system components that have been wetted but are normally dry are subject to augmented testing and inspections beyond those in Table Xl.M27-1, "Fire Water System Inspection and Testing Recommendations." The augmented tests and inspections are conducted on piping segments that cannot be drained or piping segments that allow water to collect. The augmented tests and inspections are:

Every 5 years, beginning 5 years prior to the subsequent period of extended operation (SPEO), conduct either a flow test or flush sufficient to detect potential flow blockage, or a 100 percent visual inspection of the internal surface of piping segments that cannot be drained or piping segments that allow water to collect. If the results of the 100 percent internal visual inspection are acceptable and the segment is not subsequently wetted, then further augmented tests or inspections are not necessary.

Every 5 years during the SPEO, perform volumetric wall thickness examinations on 20 percent of the length of piping segments that cannot be drained or piping segments that allow water to collect.

Subsequent License Renewal Application (SLRA) Supplement 1, dated April 21, 2021 (ADAMS Accession No. ML21111A155), revised SLRA Section B.2.3.16, "Fire Water System," by clarifying the operating experience related to the fire water piping systems subject to the wet-dry cycle. The fire water piping systems susceptible to the wet-dry

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.16-1 L-2021-144 Attachment 12 Page 2 of 4 cycle are the gas turbine building and low voltage auxiliary transformers suppression system and the warehouse #2 suppression system.

SLRA Supplement 1 stated that the internal piping of the warehouse #2 suppression system was inspected in April 2014 and there was little to no wear in the piping and no head blockage. It also stated, "An additional action request was issued to perform future inspections of warehouse #2 dry system branches as applicable [emphasis added]."

SLRA Supplement 1 stated that a low point in the gas turbine building and low voltage auxiliary transformers suppression system was drained and internally inspected and there was a normal amount of internal wear/corrosion. It also stated, "A recent walkdown determined there were several additional low points, therefore, a new action request was issued to perform future inspections of the gas turbine building and low voltage auxiliary transformers' dry system branches as applicable [emphasis added]."

Issue:

SLRA Supplement 1 did not state whether the inspection of the warehouse #2 suppression system was 100 percent of the internal surface of piping segments that cannot be drained or piping segments that allow water to collect. Due to the use of the phrase "as applicable," it is unclear whether the future inspections will be consistent with augmented tests and inspections in NUREG-2191, Volume 2. In addition, SLRA Supplement 1 did not discuss when the future inspections will be performed.

Request:

1. Please clarify whether 100 percent of the internal surface of piping segments that cannot be drained or piping segments that allow water to collect in the warehouse #2 suppression system were inspected.

2. If the inspections of the gas turbine building and low voltage auxiliary transformers suppression system and the warehouse #2 suppression system have been completed, please describe the results of the inspections.

3. Discuss what is meant by the phrase "as applicable" as it relates to the future inspections of the gas turbine building and low voltage auxiliary transformers suppression system and the warehouse #2 suppression system. Alternatively, confirm that future inspections of the suppression systems subject to the wet-dry cycle will be consistent with the augmented tests and inspections in NUREG-2191,

Volume 2.

NEPB Response:

The numbered responses below correspond to the numbered requests in RAI B.2.3.16-1.

1. The internal visual inspection of a representative branch of the warehouse #2 normally dry suppression system was last performed by a state-licensed sprinkler vendor in May 2014. The branch line was inspected to check for internal corrosion

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.16-1 L-2021-144 Attachment 12 Page 3 of 4 and/or head blockage in accordance with NFPA 25, 2011 Edition, Section 14.2. Per NFPA 25, this included opening a flushing connection at the end of one main and removing a sprinkler toward the end of a branch line and internally inspecting between those locations. The inspection did not inspect 100 percent of the internal nondraining surfaces, since only one representative branch line was inspected. The inspection was not originally a license renewal (LR) commitment. The need to periodically inspect these nondraining sections was first identified after NRC Information Notice (IN) 2013-06 was published in accordance with the PBN corrective action program.

2. The 2014 inspection of a warehouse #2 dry branch have been completed and the results stated that there was little to no internal piping wear and no sprinkler head blockage. The next inspection for a warehouse #2 dry suppression system branch is scheduled for June 2022. The results of the 2014 inspection of the nondraining normally dry suppression system for the G-05 gas turbine building and 1/2X-04 transformers stated that the inspected pipe had 2 to 3 gallons of water that was drained as well as some minor corrosion and normal pipe wear. The next inspection of the nondraining normally dry suppression system for the G-05 gas turbine building and 1/2X-04 transformers is scheduled for August 2021.

3. The phrase "as applicable" was intended to mean other branches within the warehouse #2 system that meet the wet-dry criteria. The IN 2013-06 related inspections to be performed during the PEO are independent from the SLR Commitment inspections that start 5 years prior to the SPEO. New periodic preventive maintenance activities will be created to internally inspect fire suppression system components exposed to the wet-dry cycle prior to the SPEO and will be consistent with NUREG-2191, Volume 2. The proposed enhancements were clarified in Supplement 1, Attachment 14, as follows:

Additionally, the new procedure will specify that portions of water-based fire protection system components that have been wetted but are normally dry, such as dry-pipe or preaction sprinkler system piping and valves, are subjected to augmented testing and inspections beyond those of NUREG-2191 Table Xl.M27-1.

The augmented tests and inspections are conducted on piping segments that cannot be drained or piping segments that allow water to collect:

In each 5-year interval, beginning 5 years prior to the SPEO, either conduct a flow test or flush sufficient to detect potential flow blockage, or conduct a visual inspection of 100 percent of the internal surface of piping segments that cannot be drained or piping segments that allow water to collect.

In each 5-year interval of the SPEO, 20 percent of the length of piping segments that cannot be drained or piping segments that allow water to collect is subject to volumetric wall thickness inspections. Measurement points are obtained to the extent that each potential degraded condition can be identified (e.g., general corrosion, MIC). The 20 percent of piping that is inspected in each 5-year interval is in different locations than previously inspected piping.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.16-1 L-2021-144 Attachment 12 Page 4 of 4 If the results of a 1 GO-percent internal visual inspection are acceptable, and the segment is not subsequently wetted, no further augmented tests or inspections are necessary.

References:

1. NextEra Energy Point Beach, LLC (NEPB) Letter to NRC L-2021 -081 dated April 21, 2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML21111A155)

Associated SLRA Revisions:

None.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-1 L-2021-144 Attachment 13 Page 1 of 4

8. SLRA Section B.2.3.27, "Buried and Underground Piping and Tanks" Regulatory Basis:

10 CFR 54.21 (a)(3) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 1 O CFR 54.29(a), the staff requires additional information in regard to the matters described below.

RAI B.2.3.27-1 (Breakout Topic No. 4: Cathodic Protection)

Background:

SLRA Section B.2.3.27, "Buried and Underground Piping and Tanks," states "[t]he PBN Buried and Underground Piping and Tanks AMP, with enhancements, will be consistent with exception [not related to the subject RAls] with the 10 elements of NUREG-2191, Section Xl.M41, "Buried and Underground Piping and Tanks.""

GALL-SLR Report AMP Xl.M41, "Buried and Underground Piping and Tanks,"

Table Xl.M41-1, "Preventive Actions for Buried and Underground Piping and Tanks,"

recommends that cathodic protection is provided for buried steel piping and tanks at least 5 years prior to the subsequent period of extended operation (SPEO). In addition, the "preventive actions" program element of GALL-SLR Report AMP Xl.M41 states the following:

Failure to provide cathodic protection in accordance with Table Xl.M41-1 may be acceptable if justified in the SLRA. The justification addresses soil sample locations, soil sample results, the methodology and results of how the overall soil corrosivity was determined, pipe to soil potential measurements and other relevant parameters. If cathodic protection is not provided for any reason, the applicant reviews the most recent 10 years of plant specific operating experience (OE) to determine if degraded conditions that would not have met the acceptance criteria of this AMP have occurred. This search includes components that are not in scope for license renewal if, when compared to in scope piping, they are similar materials and coating systems and are buried in a similar soil

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-1 L-2021 -144 Attachment 13 Page 2 of 4 environment. The results of this expanded plant specific OE search are included in the SLRA.

During its review of the SLRA, the staff noted the following steel components will be managed for loss of material using the Buried and Underground Piping and Tanks program: (a) steel piping exposed to soil in the service water, emergency power, and fire protection systems; and (b) the half buried emergency fuel oil storage tank.

During its audit, the staff noted the following : (a) Point Beach does not have any buried tanks that are cathodically protected; (b) the cathodic protection systems at Point Beach were originally installed to provide corrosion control for the containment structures and buried circulating water piping.

Issue:

Based on its audit and review of the SLRA, the staff could not determine if cathodic protection will be provided for steel piping and tanks exposed to soil (within the scope of subsequent license renewal (SLR)). The SLRA does not specify which systems and/or components will receive cathodic protection.

Request:

Provide clarification regarding if cathodic protection will be provided for steel piping and tanks exposed to soil (within the scope of SLR) at least 5 years prior to the SPEO. If all or portions of the subject piping and tanks will not be cathodically protected, state the basis, according to the "preventive actions" element of GALL-SLR Report AMP Xl.M41,

for why cathodic protection will not be provided.

NEPB Response:

The only buried tank exposed to soil at PBN is the emergency fuel oil storage tank, T-072, which is partially buried. This tank is not cathodically protected and no activities are planned to install cathodic protection for this tank. The basis for excluding cathodic protection on this tank are as follows:

The buried portion of tank T-072 has other preventive measures applied, which include the following :

o A bitumastic coating.

o The respective backfill includes clean granular backfill soil that excludes frozen material, ice lenses, top soil, humus, brush, roots, peat, sod, cinders, rubbish, other perishable materials, clay, and sufficiently-sized stones that could interfere with compaction. The backfill soil was compacted in 9-inch layers to 95% of maximum density. The applicable specification implies that the backfill soil could have been imported at the contractor's discretion, but even if local soil was used, soil analyses determined that the soil was not aggressive.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-1 L-2021-144 Attachment 13 Page 3 of 4 The buried portion of tank T-072 has no operating experience (OE) which has resulted in loss of intended function of the tank. The results of the most recent visual and volumetric examinations, performed in August 2020, show that for the locations with the greatest wall loss rate, the projected time until minimum wall thickness may be reached was greater than 20 years. The tank will be reinspected and reevaluated on a 10-year interval. If an evaluation projects the minimum wall thickness being reached prior to the end of the SPEO, then per Commitment 31, Items "m" and "n", a corrective action must be performed. An acceptable corrective action can include performing the tank inspections at a higher frequency (e.g., 5 years) to provide reasonable assurance that minimum wall thickness remains maintained through the SPEO. Other appropriate corrective actions, such as repairs or cathodic protection, can also be evaluated and implemented as necessary.

The 2015 cathodic protection survey for PBN states that the following buried piping systems have cathodic protection: circulating water supply and return, service water supply and return, fire protection water, fuel oil, and propane system lines. However, not all of these buried piping systems are completely cathodically protected. No activities are currently planned to increase the coverage of cathodic protection system to 100 percent of the buried SLR-scope piping. The basis for excluding cathodic protection on portions of the buried piping is provided as follows in accordance with NUREG-2191, Section Xl.M41, Subsection 2.g.iv:

The soil was determined to be non-aggressive as proven by the following soil analyses:

o A 1992 analysis of soil samples was obtained during the installation of 4 groundwater monitoring wells in the Unit 1 and Unit 2 facades, near the containment structures. The soil samples were extracted every 5 feet during the well borings and the samples were analyzed for pH, resistivity, and chlorides. The samples had an average resistivity of 16,740 ohm-cm, which was considered "mildly corrosive" per the sampling manual used, "Corrosion Control," Air Force Manual (AFM), No. 88-9. The average pH was 9.52, which was considered to be within the optimum range of 8.5 to 11.0. The average amount of chlorides was approximately 59 ppm, which was well below the 500 ppm minimum for an aggressive chloride environment. In general, the soil analyses showed no signs of aggressive chemical exposure to subsurface systems around the containment structures at PBN.

o A 2009 analysis of soil samples in the immediate vicinity of the buried fire protection system piping was performed after the piping had been excavated for 10-year inspections. The moisture content was analyzed in accordance with ASTM 02974-87; resistivity was analyzed in accordance with EPA 120.1; pH was analyzed in accordance with EPA 9045; oxidation and reduction potential were analyzed in accordance with SM 25808; and

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-1 L-2021-144 Attachment 13 Page 4 of 4 IC anions were analyzed in accordance with EPA 300.0. The sample results indicated that resistivity was within the 13,800-16,600 Ohm-cm range, redox potential had a range of 81.9-172 mV, the soil pH was 7.9, chlorides were measured at 31.8 mg/kg, sulfides were within a range of 11.6-13.4 mg/kg, and moisture content was at 19.1 percent. The associated action request (AR) stated that the soil sample results proved that the soil was nonaggressive.

A review of PBN plant-specific operating experience (OE) spanning the 10-year operating period prior to January 1, 2020 was performed and was documented in SLRA Section B.2.3.27. The OE review was inclusive of all buried components, including components outside the scope of SLR. No aging-related failures were identified for buried piping or tank components. The only failure identified for a pressure-retaining component was related to freeze-induced cracking of a fire hydrant header in 2015. The OE review indicated that when excavations were performed in 2009 and 2016, no evidence of wall loss was identified.

The response to RAI B.2.3.27-4 states that PBN will commit to performing SLR inspections of buried piping in accordance with NUREG-2191, Table Xl.M41-2, Preventive Action Category E, rather than Category C. That RAI response also provides further clarification on why Preventive Action Category E is the appropriate category. The additional inspections associated with Preventive Action Category E will provide reasonable assurance that aging is appropriately managed for the buried piping sections that lack cathodic protection.

References:

1. NextEra Energy Point Beach, LLC (NEPB) Letter to NRC L-2021-081 dated April 21, 2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML21111A155)

Associated SLRA Revisions:

None.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-2 L-2021-144 Attachment 14 Page 1 of 4 RAI B.2.3.27-2 (Breakout Topic Nos. 5 and 6: Uncoated Buried Fire Protection Piping)

Background:

GALL-SLR Report Table Xl.M41-1 recommends that external coatings are provided for buried metallic piping. In addition, GALL-SLR Report AMP Xl.M41 states additional inspections, beyond those in GALL-SLR Report Table Xl.M41-2, "Inspection of Buried and Underground Piping and Tanks," may be appropriate if exceptions are taken to the "preventive actions" program element or in response to plant specific OE. Based on its audit and review of the SLRA, the staff noted there is uncoated buried piping in the fire protection system. The staff could not determine the extent of uncoated buried piping in the fire protection system or which material types (i.e., steel, gray cast iron, ductile iron) are uncoated. Therefore, additional inspections may be appropriate, according to GALL-SLR Report AMP Xl.M41.

Issue:

The recommended inspections in GALL-SLR Report AMP Xl.M41 are based on external coatings being provided for buried metallic piping. Based on plant specific OE indicating that there is uncoated buried piping in the fire protection system, the staff seeks additional clarification regarding why additional inspections of buried fire protection system piping are not appropriate.

Request:

Provide additional OE (or other technical justification) to demonstrate that additional inspections, beyond those recommended in GALL SLR Report AMP Xl.M41, are not appropriate for buried fire protection system piping.

NEPB Response:

The general buried piping inspections recommended by NUREG-2191, Table Xl.M41-2, Preventive Action Category E (see response to RAI B.2.3.27-4, Attachment 16 to this letter) will be supplemented by additional inspections for the uncoated buried fire protection piping. Since the uncoated buried fire protection piping was installed in accordance with NFPA 24 and the piping will be subject to periodic flow testing in accordance with NFPA 25, section 7.3.1, the piping is exempt from the preventive actions listed in NUREG-2191, Table Xl.M41-1. As a supplement to the flow testing preventive measure, continuous pressure monitoring of the fire water system is currently performed by the PBN Fire Water System AMP and will continue to be performed through the SPEO, so that loss of system pressure is immediately detected and corrected when acceptance criteria are exceeded.

PBN has a current license renewal (LR) commitment to perform 10-year inspections on a "susceptible" location in the fire protection system (i.e., uncoated or unwrapped piping). NEPB proposes that this current inspection commitment be extended to be performed through the SPEO. This commitment will be in addition to the inspections on the other buried piping described in the response to RAI B.2.3.27-4. This extended

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-2 L-2021-144 Attachment 14 Page 2 of 4 inspection Commitment will focus on uncoated/unwrapped fire protection piping for future inspections. Accordingly, an SLR commitment is added.

References:

1. NextEra Energy Point Beach, LLC (NEPB) Letter to NRC L-2021-081 dated April 21, 2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML21111A155)

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-2 L-2021-144 Attachment 14 Page 3 of 4 Associated SLRA Revisions:

SLRA Appendix A, Section 16.4, Table 16-3 (Item 31), page A-100, as revised by SLRA Aging Management Supplement 1, is amended as follows:

f No.

Aging NUREG-2191 Management Section Program or Activity (Section) m

~

Table 16-3 di Commitment Schedul

[Refer to the RAI B.2.3.27-4 response for related changes.]

Perform insgections on the uncoated/unwragged gortions of the buried fire grotection S)lstem giging no earlier than 10 )!'.ears grior to the SPEO and at least eve!)l 10 )!'.ears during the SPEO. The insgections include at least two 10-ft segments of uncoated/unwragged fire grotection QiQing.

Implementation Schedule

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-2 L-2021-144 Attachment 14 Page 4 of 4 SLRA Appendix B, Section B.2.3.27, page B-197, as revised by SLRA Aging Management Supplement 1, is amended as follows:

Element Affected Enhancement

4. Detection of Aging PBN manuals, procedures, etc. will be enhanced to:

Effects

[Refer to the RAI 8.2.3.27-4 response for related changes.]

[Refer to the RAI 8.2.3.27-4 response for related changes.]

Clarif)l that insQections will be Qerformed on the uncoated/unwrai;med QOrtions of the buried fire Qrotection S)lstem QiQing no earlier than 10 )£ears Qrior to the SPEO and at least evert 10 )£ears during the SPEO.

The insQections include at least two 10-ft segments of uncoated/unwrapped fire protection PiPinQ.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-3 L-2021-144 Attachment 15 Page 1of5 RAI 8.2.3.27-3 (Breakout Topic Nos. 3 and 7: Interface Corrosion for Tanks)

Background:

SLRA Table 3.3.2-8, "Emergency Power - Summary of Aging Management Evaluation,"

states the emergency diesel generator (EOG) fuel oil storage tanks are exposed to an external environment of concrete. As amended by letter dated April 21, 2021 (ADAMS Accession No. ML21111A155), SLRA Table 3.3.2-8 was revised to state that the subject tanks are also exposed to an underground external environment.

SLRA Table 3.3.2-8 states the emergency fuel oil storage tank is exposed to an external environment of soil. As amended by letter dated April 21, 2021, SLRA Table 3.3.2-8 was revised to state that the subject tank is also exposed to a concrete and underground external environment.

As amended by letter dated April 21, 2021, the last enhancement associated with the "detection of aging effects" program element of SLRA Section B.2.3.27, "Buried and Underground Piping and Tanks," states the following:

Issue:

"[c]larify that examinations of the buried portion of tank T-072 are conducted from the external surface of the tank using visual techniques or from the internal surface of the tank using volumetric techniques. A minimum of 25% of the buried surface is examined. The inspected area includes at least some of both the top and bottom of the tank [emphasis added]. If the tank is inspected internally by volumetric methods, the method must be capable of determining tank wall thickness and general and pitting corrosion and qualified at PBN to identify loss of material that does not meet acceptance criteria. The double wall tanks, T-175A and T-175B shall be examined by monitoring.the annular space for leakage."

The inspection recommendations for buried and underground tanks provided in the enhancement above are consistent with GALL-SLR Report AMP Xl.M41 recommendations (e.g., as italicized above, the inspected area includes at least some of both the top and bottom of the tank). However, these recommendations are based on tanks being exposed to a buried or underground environment (i.e., not tanks with an air-to-soil, air-to-concrete, or soil-to-concrete external interface, where there is an increased potential for degradation at interface locations). The staff seeks additional clarification regarding how inspections of the EOG fuel oil storage tanks (exposed externally to a concrete and underground (i.e., air) environment) and the emergency fuel oil storage tank (exposed externally to soil, concrete, and underground environments) will account for the potential for corrosion at air-to-soil, air-to-concrete, or soil-to-concrete interfaces.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-3 L-2021 -144 Attachment 15 Page 2 of 5 Request:

Provide additional information regarding how inspections of the EOG fuel oil storage tanks and emergency fuel oil storage tank will account for the potential for corrosion at air-to-soil, air-to-concrete, or soil-to-concrete interfaces. Alternatively, revise the subject enhancement to clarify that inspected areas will also include interface regions (where there is a transition from one external environment to another).

NEPB Response:

The EOG fuel oil storage tanks, T-175A and T-175B, are located in the underground vault portion of the diesel generator building. Both are horizontally-placed cylindrical tanks resting on metal tank supports. Each tank is encased within a liner. Bottom supports as well as an annular air space separate the exterior tank surface from the liner. The EOG tank liner is surrounded by the underground/uncontrolled vault air.

These tanks do not have any interfaces where an external environment transitions to another environment. The EOG tanks do not have manways installed, so it is not practical to do internal inspections. Rather, the annular air space has a drain hole which allows for the removal of moisture. The drain also allows for the monitoring of tank leakage which will be performed in lieu of tank inspections. Although the EOG tank supports are in contact with concrete, the actual tank performing the pressure boundary function is not. Therefore, the SLRA is amended to remove the respective rows from the aging management evaluation summary table.

The emergency fuel oil storage tank, T-072, is a horizontally-placed cylindrical tank resting on metal tank supports. A portion of tank T-072 is located within the fuel oil pumphouse, where it is in contact with the underground uncontrolled pumphouse air, as well as the concrete wall. The other portion of tank T-072 is located on the other side of the concrete wall, where it is buried in soil that meets certain backfill specifications.

Thus, there is no air-to-soil interface. This tank has two environment transition interfaces: a) backfill soil-to-concrete, and b) underground uncontrolled air-to-concrete.

The inspection enhancement for this tank is amended to clarify that, in addition to the external visual inspection of the underground uncontrolled air-to-concrete interface, internal volumetric examinations will be performed and the internal volumetric examinations will record wall thickness measurements at both interfaces.

References:

1. NextEra Energy Point Beach, LLC (NEPB) Letter to NRC L-2021-081 dated April 21, 2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML21111A155)

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-3 L-2021-144 Attachment 15 Page 3 of 5 Associated SLRA Revisions:

SLRA Table 3.3.2-8, page 3.3-221, as revised by SLRA Aging Management Supplement 1, is amended to remove the rows as follows:

Table 3.3.2-8: Emeraencv Power Svstem - Summarv of Aaina Manaaement Evaluation Component Intended Material Environment Aging Effect Aging NUREG-2191 Table 1 Item Notes Type Function Requiring Management Item Manaaement Proa ram Tank (EOG PFess1:1Fe Cmbon Concmte Cmcking 81:1FieEl anEl VII.I.I\\ 425 3.3 1, 144 g

fuel bo1:1nElary steel fe*1t 6lnEleFgFO!:lnEl stomge)

Piping anEl Tanks I D"> 'l ">"'7\\

Tank (EOG PFess1:1Fe Cmbon Concmte Loss of matmial 81:1FieEl anEl VII.I.AP 198 3.3 1, 109 g

fuel bo1:1nElary steel fe*1t 6lnEleFgFO!:lnEl stomge)

Piping anEl Tanks r n,.,..,,.,..,,

"LJ, "";E.... ;,--*;*.L..*

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-3 L-2021-144 Attachment 15 Page 4 of 5 SLRA Appendix A, Section 16.4, Table 16-3 (Item 31 ), page A-98, as revised by SLRA Aging Management Supplement 1, is amended as follows:

Table 16-3 List of SL No.

Aging NUREG-2191 Commitment Implementation Management Section Schedule Program or Activity (Section) k)

Perform the examinations of the buried portion of tank T-072 from eitAef the external surface of the tank using visual techniques, which include ins12ection of the air-to-concrete wall interface1 ei: and from the internal surface of the tank using volumetric techniques.:...-A minimum of 25% of the ffi:H:ieG-tank surface is examined. The inspected area includes at least some of both the top and bottom of the tank. If the tank is inspected internally by volumetric methods, t_!he method must be capable of determining tank wall thickness and general and pitting corrosion and qualified at PBN to identify loss of material that does not meet acceptance criteria. Volumetric wall thickness measurements must also be recorded for the tank interfaces with the air-to-concrete and concrete-to-soil exterior environment transitions. The double wall tanks, T-175A and T-1758 shall be examined by monitoring the annular space for leakage.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-3 L-2021 -144 Attachment 15 Page 5 of 5 SLRA Section B.2.3.27, page B-198, as revised by SLRA Aging Management Supplement 1, is amended as follows:

Element Affected Enhancement than 10 feet, then 10 feet of piping is inspected. If the entire run of piping of that material type is less than 10 feet in total length, then the entire run of piping is inspected.

Clarify that the visual inspections will be supplemented with surface and/or volumetric nondestructive testing if evidence of wall loss beyond minor surface scale is observed.

State that PBN site-specific conditions can result in transitioning to a higher number of inspections than originally planned at the beginning of a 10-year interval as specified in NUREG-2192, Section 4.a of Xl.M41.

Clarify the guidance for piping inspection location selection as follows: (a) a risk ranking system software incorporates inputs that include coating type, coating condition, cathodic protection efficacy, backfill characteristics, soil resistivity, pipe contents, and pipe function; (b) opportunistic examinations of nonleaking pipes may be credited toward examinations if the location selection criteria are met; and (c) the use of guided wave ultrasonic examinations may not be substituted for the required inspections.

Select one of the alternatives to visual examination of piping from NUREG-2191 pages Xl.M41-9 and Xl.M41-10.

Clarify that examinations of the buried portion of tank T-072 are conducted from the external surface of the tank using visual techniques1 which include insl;!ection of the air-to-concrete wall interface1--GF and from the internal surface of the tank using volumetric techniques. A minimum of 25% of the btlfietl tank surface is examined. The inspected area includes at least some of both the top and bottom of the tank.-

If the tank is inspected internally by volumetric methods, t_

! he method must be capable of determining tank wall thickness and general and pitting corrosion and qualified at PBN to identify loss of material that does not meet acceptance criteria. Volumetric wall thickness measurements must also be recorded for the tank interfaces with the air-to-concrete and concrete-to-soil exterior environment transitions. The double wall tanks, T-175A and T-1758 shall be examined by monitoring the annular space for leakage.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-4 L-2021-144 Attachment 16 Page 1 of 9 RAI 8.2.3.27-4 (Breakout Topic No. 8: Preventive Action Category C)

Background:

As amended by letter dated April 21, 2021, the SLRA states the following with respect to meeting the criteria for and transitioning from Preventive Action Category C:

SLRA Appendix A, "Updated Final Safety Analysis Report Supplement," Section 16.2.2.27, "Buried and Underground Piping and Tanks," states:

[b]ased on excellent plant OE and the combination of good soil conditions, preventive design features in place, and inspections, the buried steel piping at PBN would meet the criteria for Preventive Action Category C. However, if these conditions were to change, the Preventive Action Category would require reevaluation and could potentially change. Thus, the number of inspections for each 10-year inspection period, commencing 10 years prior to the SPEO, based on the inspection quantities identified in GALL-SLR Table Xl.M41-2 (adjusted for a 2-unit plant site) is two, so long as the Preventive Action Category C remains applicable.

SLRA Appendix A, Table 16-3, "List of SLR Commitments and Implementation Schedule," commitment No. 31, subpart (g) states in part:

[p]erform inspections of buried and underground piping and tanks in accordance with NUREG-2191 Table Xl.M41-2 Category C steel, unless a reevaluation of future OE and soil conditions determines that another Preventive Action Category is more applicable.

The first enhancement associated with the "detecting of aging effects" program element in SLRA Section B.2.3.27, "Buried and Underground Piping and Tanks,"

states the in part:

[c]larify that inspections of buried and underground piping and tanks will be conducted in accordance with NUREG-2191 Table Xl.M41-2 Category C steel, unless a reevaluation of future OE and soil conditions determines that another Preventive Action Category is more applicable.

GALL-SLR Report Table Xl.M41-2, "Inspection of Buried and Underground Piping and Tanks," states:

Preventive Action Category C applies when (a) cathodic protection was installed or refurbished 5 years prior to the end of the inspection period of interest; (b) cathodic protection has operated at least 85 percent of the time either since 10 years prior to the subsequent period of extended operation or since installation/refurbishment, whichever is shorter; and (c) cathodic protection has provided effective protection for buried piping as evidenced by meeting the acceptance criteria of Table Xl.M41-3, "Cathodic Protection Acceptance Criteria," of this AMP at least 80 percent of the time either since 10 years prior to the subsequent period of extended operation or since installation/refurbishment, whichever is shorter.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-4 L-2021-144 Attachment 16 Page 2 of 9 Preventive Action Category E applies when a cathodic protection system has been installed but all or portions of the piping covered by that system fail to meet any of the criteria of Preventive Action Category C piping, provided (a) coatings and backfill are provided in accordance with the "preventive actions" program element of this AMP; (b) plant-specific OE is acceptable (i.e., no leaks in buried piping due to external corrosion, no significant coating degradation or metal loss in more than 10 percent of inspections conducted); and (c) soil has been determined to not be corrosive.

Issue:

The conditions to meet Preventive Action Category C, as delineated in GALL-SLR Report Table Xl.M41-2, are based solely on cathodic protection efficacy. Factors such as soil corrosivity, plant-specific OE, results of inspections, and preventive actions other than cathodic protection (i.e., coatings and backfill quality) more closely align with the criteria for meeting Preventive Action Category E. The staff seeks clarification regarding why factors other than cathodic protection efficacy are used in relation to meeting the criteria for and transitioning from Preventive Action Category C.

Request:

State the basis for why factors other than cathodic protection efficacy are used in relation to meeting the criteria for and transitioning from Preventive Action Category C.

Alternatively, revise the SLRA, as appropriate, to reflect that cathodic protection efficacy will be used in relation to meeting the criteria for and transitioning from Preventive Action Category C.

NEPB Response:

Since not all of the buried piping within the scope of SLR is cathodically protected and no activities are planned to increase the cathodic protection coverage to 100 percent, a clarification is added to state that buried SLR-scope piping will be classified as Category E, unless a reevaluation based on future OE and soil conditions, as defined in NUREG-2191, Table XI.M41-2, determines that another Preventive Action Category, such as Category F, is more applicable. Preventive Action Category C will not be used.

The response to RAI B.2.3.27-1 provides justification in accordance with NUREG-2191, Section Xl.M41, Item 2.g.iv for why cathodic protection does not need to be installed in locations where it is impractical. Preventive Action Category E is appropriate for the buried piping for the following reasons:

All buried piping, except for a portion of the fire protection piping, is wrapped and/or coated and lined per the original construction requirements. For the uncoated fire protection piping, additional inspections are performed, as described in the RAI B.2.3.27-2 response, to compensate and provide reasonable assurance that aging of the piping exterior is managed through the SPEO.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-4 L-2021-144 Attachment 16 Page 3 of 9 All original construction backfill for buried piping systems was specified to meet certain requirements including clean granular backfill soil that excluded frozen material, ice lenses, top soil, humus, brush, roots, peat, sod, cinders, rubbish, other perishable materials, clay, and sufficiently-sized stones that could interfere with compaction. The backfill was also required to meet compaction requirements, which vary depending on the application. The fire protection ring header installed after original construction meets the backfill requirements of NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appurtenances, which states that a) backfill shall be tamped in layers or puddled under and around pipes to prevent settlement or lateral movement and shall contain no ashes, cinders, refuse, organic matter, or other corrosive materials; b) rocks shall not be placed in trenches; Frozen earth shall not be used for backfilling; and c) in trenches cut through rock, tamped backfill shall be used for at least 6 inches under and around the pipe and for at least 2 feet above the pipe.

A review of PBN plant-specific operating experience (OE) spanning the 10-year operating period prior to January 1, 2020 was performed and documented in SLRA Section B.2.3.27. The OE review was inclusive of all buried components, including components outside the scope of SLR. No aging-related failures were identified for buried piping or tank components. The only failure identified for a pressure-retaining component was related to freeze-induced cracking of a fire hydrant header in 2015. The OE review indicated that when excavations were performed in 2009 and 2016, no evidence of wall loss was identified.

Soil analyses in accordance with NUREG-2191, Table Xl.M41-2 will be performed. To ensure that Preventive Action Category E remains valid prior to and throughout the subsequent period of extended operation (SPEO), a Commitment is added to perform soil sample analysis in accordance with NUREG-2191 Table Xl.M41-2, Item E.b.iii, starting no earlier than 10 years prior to the SPEO, but no later than 6 months or last refueling outage prior to the SPEO, and recurring every 10 years thereafter through the SPEO.

The relevant PBN SLRA, as modified by SLRA Aging Management Supplement 1, commitment and enhancement wording is revised to reflect the NUREG-2191 guidance.

References:

1. NextEra Energy Point Beach, LLC (NEPB) Letter to NRC L-2021-081 dated April 21, 2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML21111A155)

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-4 L-2021-144 Attachment 16 Page 4 of 9 Associated SLRA Revisions:

SLRA Appendix A, Section 16.2.2.27, page A-32, as revised by SLRA Aging Management Supplement 1, is amended as follows:

Based on excellent plant OE and the combination of good soil conditions, preventive design features in place, and inspections, and cathodic protection coverage on portions but not all of the buried piping, the buried steel piping at PBN wG-HW meet~ the criteria for Preventive Action Category Gg. However, if these conditions were to change, the Preventive Action Category would require reevaluation and could potentially change to Preventive Action Category F. Thus, the number of inspections for each 10-year inspection period, commencing 10 years prior to the SPEO, based on the inspection quantities identified in GALL-SLR Table XI.M41-2 (adjusted for a 2-unit plant site) is the smaller of 5% of the total piping length or five 10-foot segments. This is in addition to the 10-year inspection of the uncoated buried fire water system piping. Soil testing is conducted starting no earlier than 10 years prior to the SPEO, but no later than 6 months or last refueling outage prior to the SPEO, and soil testing recurs every 10 years thereafter through the SPEO. tvvo, so long as the Preventive Action Category C remains applicable.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-4 L-2021-144Attachment16 Page 5of9 SLRA Appendix A, Section 16.4, Table 16-3 (Item 31 ), page A-97, as revised by SLRA Aging Management Supplement 1, is amended as follows:

No.

Aging Management Program or Activity (Section)

List of SLR C Table 16-3 "tments and I tation Schedul NUREG-2191 Commitment Section g)

PeFfem1 iAspeGtieAs ef 91:JFiee aAe 1:JAEieF§Fe1:JAe pipiA§ aAe taAks iA aGG9FSaAGe with NUREG 2191 Taele Xl.M41 2 Gate§eFY G steel, 1:1Rless a Feeval1:1atieR ef f1:1t1:1Fe GE aRel seil GeRelitieRs eleteFmiRes that aRetheF PFeveRtive AstieR Categeiy is mern applisable. Perform inspections of buried piping in accordance with NUREG-2191 Table Xl.M41-2 Catego!)l E steel, unless a reevaluation based on future OE and soil conditions, as defined in NUREG-2191, Table Xl.M41-2, determines that another Preventive Action Catego!)l is more applicable. The inspections will be distributed evenly among the units. Since PBN is a two-unit site, the inspection quantities are 50% greater than NUREG-2191 Table Xl.M41-2 and are rounded up to the nearest whole inspection. Thus, the number of inspections for each 10-year inspection period, commencing 10 years prior to the SPEO and continuing during the SPEO, iR aGGeFelaRGe 1.vith PrnveRtive AGtieR Categeiy C, is as follows:

Buried Piping: The smaller of G:-5% of the piping length or twe five 10-foot segments.

Buried Tank: One inspection for tank T-072.

Underground Tanks: In lieu of inspections, Mmonitor annular space of double walled tanks T-175A and T-1 75B for leakage.

Implementation Schedule

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-4 L-2021-144 Attachment 16 Page 6 of 9 No.

Aging NUREG-2191 Management Section Program or Activity (Section)

Commitment Implementation Schedule When the inspections for a given material type is are based on percentage of length and results in an inspection quantity of less than 10 feet, then 10 feet of piping is inspected. If the entire run of piping of that material type is less than 10 feet in total length, then the entire run of piping is inspected.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-4 L-2021-144Attachment 16 Page 7 of9 SLRA Appendix A, Section 16.4, Table 16-3 (Item 31 ), page A-100, as revised by SLRA Aging Management Supplement 1, is amended as follows:

No.

Aging Management Program or Activity (Section)

List of SLR C Table 16-3 "tments and I tation Schedul NUREG-2191 Commitment Section m Perform soil samQle anall£ses in accordance with NUREG-2191, Table Xl.M41-2, as follows to confirm that soil is not corrosive for the resQective QiQing material tl£Qe:

Obtain a minimum of three sets of soil samQles in each soil environment {e.g., moisture content, soil comQosition} in the vicinitv in which in-scoQe comQonents are buried.

Test the soil for soil resistivitv, corrosion accelerating bacteria, QH, moisture, chlorides, sulfates, and redox QOtential.

Determine the QOtential soil corrosivitv for each material tvQe of buried in-scoQe QiQing. In addition to evaluating each individual Qarameter, the overall soil corrosivitl! shall be determined.

Conduct soil testing no earlier than 10 l£ears Qrior to the SPEO, but no later than 6 months or last refueling outage Qrior to the SPEO, and conduct soil testing evert 10 l£ears thereafter through the SPEO.

Implementation Schedule

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-4 L-2021-144 Attachment 16 Page 8 of 9 SLRA Appendix B, Section B.2.3.27, page B-197, as revised by SLRA Aging Management Supplement 1, is amended as follows:

Element Affected Enhancement

4. Detection of Aging PBN manuals, procedures, etc. will be enhanced to:

Effects GlaFify tl=lat iRs13estieRs ef 81:1FieEI aREI l:IREleF§JF91:1REI 13i13iR§J aREI taRks will se 69RSl:IGteEI iR asseFElaRGe llJitl=l t>J6J~eG 2~ 9~

+asle Xl.M4 ~ 2 Gate§l9Py' G steel, !:JR less a rne>.ial1:JatieR ef f1:Jt1:JFe Gi aRd seil seRditieRs detel'miRes that aRethel' PrnveRtive AstieR Categel'y is mme applisable. Clarify that ins12ections of buried 12i12ing will be conducted in accordance with NUREG-2191 Table Xl.M41-2 Categorv E steel 1 unless a reevaluation based on future OE and soil conditions 1 as defined in NUREG-2191 1 Table Xl.M41-2 1 determines that another Preventive Action Categort is more a1212licable. The inspections will be distributed evenly among the units. Since PBN is a two-unit site, the inspection quantities are 50% greater than NUREG-2191 Table Xl.M41-2 and are rounded up to the nearest whole inspection. Thus, the number of inspections for each 10-year inspection period, commencing 10 years prior to the SPEO and continuing during the SPEO, iR asse!'GaRse with PrnveRtive AstieR Categel)' C, is as follows:

0 -

Buried Piping: The smaller of G-:-5% of the piping length or twe five 10-foot segments.

0 Buried Tank: One inspection for tank T-072.

0 Underground Tanks: In lieu of ins12ections, Mmonitor the annular space of double walled tanks T-175A and T-175B for leakage.

Clarify that soil sam12le analyses will be 12erformed in accordance with NUREG-2191 1 Table Xl.M41-2 1 as follows to confirm that soil is not corrosive for the res12ective 12i12ing material ty12e:

0 Obtain a minimum of three sets of soil sam12les in each soil environment {e.g. 1 moisture content1 soil com12osition} in the vicinity in which in-sco12e com12onents are buried.

0 Test the soil for soil resistivity 1 corrosion accelerating bacteria 1 12H 1 moisture1 chlorides 1 sulfates 1 and redox 12otential.

0 Determine the 12otential soil corrosivity for each material ty12e of buried in-sco12e 12i12ing. In addition to evaluating each individual 12arameter1 the overall soil corrosivity shall be determined.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.27-4 L-2021 -144 Attachment 16 Page 9 of 9 Element Affected Enhancement 0

Conduct soil testing no earlier than 10 l£ears 12eriod 12rior to the SPE0 1 but no later than 6 months or last refueling outage 12rior to the SPE0 1 and conduct soil testing evert 10 l£ears thereafter through the SPEO.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-1 L-2021 -144 Attachment 17 Page 1 of 3

9. SLRA Section B.2.3.29, "ASME Section XI, Subsection IWE" Regulatory Basis:

Paragraph 54.21 (a)(3) of 10 CFR requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis for the subsequent period of extended operation. As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL-SLR Report when evaluation of the matter in the GALL-SLR Report applies to the plant.

RAI B.2.3.29-1

Background:

SLRA Section B.2.3.38 under the heading "NUREG-2191 Consistency" states: "The PBN ASME Section XI, Subsection IWE AMP, with enhancements, will be consistent with the 10 elements of NUREG-2191, Section Xl.S1, "ASME Section XI, Subsection IWE" as modified by SLR-/SG-Structures-2020-XX, Updated Aging Management Criteria for Structures Portions of the Subsequent RA/ B. 2. 3. 29-1 License Renewal."

(emphasis added)

Issue:

The GALL-SLR AMP Xl.S1, "ASME Section XI, Subsection IWE," was not modified by the cited Draft Interim Staff Guidance (ISG) SLR-ISG-Structures-2020-XX (issued for comment in June 2020), as claimed in the consistency statement SLRA Section B.2.3.29. The staff further noted that the option to perform a fatigue waiver analysis, for containment pressure-retaining boundary components subject to cyclic loading that do not have a CLB fatigue analysis, provided in Appendix A of the ISG related to the SR P-SLR further evaluation 3.5.2.2.1.5 was not used in the PBN SLRA. The staff is unable to make a finding of the NUREG-2191 consistency for SLRA B.2.3.29 AMP.

Request:

Provide an appropriately revised NUREG-2191 consistency statement for SLRA B.2.3.29, "ASME Section XI, Subsection IWE," AMP against which staff can make its finding.

NEPB Response:

SLRA Section B.2.3.29, as amended by SLRA Aging Management Supplement 1, is revised as described below to provide a more accurate NUREG-2191 consistency statement for the PBN ASME Section XI, Subsection IWE, AMP.

References:

1. NextEra Energy Point Beach, LLC (NEPB) Letter to NRC L-2021-081 dated April 21, 2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML21111A155)

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-1 L-2021-144 Attachment 17 Page 2 of 3 Associated SLRA Revisions:

SLRA Section B.2.3.29 (page B-214), as amended by SLRA Aging Management Supplement 1, is revised as follows:

peeling, discoloration, wear, pitting, excessive corrosion, arc strikes, gouges, surface discontinuities, dents, or other signs of surface irregularities. Pressure-retaining bolting is examined for loosening and material conditions that cause the bolted connection to affect either containment leak-tightness or structural integrity. Moisture barriers are visually inspected for degradation per Category E-A.

Cumulative fatigue damage for the PBN liner and steel piping (and ventilation) penetrations for the containment structures is addressed in the Containment Liner Plate, Metal Containments, and Penetrations Fatigue Analysis TLAA for SLR (Section 4.6). Cracking due to cyclic loading of all non-piping penetrations (hatches, electrical penetrations, etc.) that are subject to cyclic loading but have no current licensing bases fatigue analysis will be managed by the 10 CFR Part 50, Appendix J AMP (Section B.2.3.32) or supplemental surface examinations or enhanced visual examinations_using the ASME Section XI, Subsection IWE AMP.

This AMP will also include supplemental one-time inspections within 5 years prior to the SPEO for a representative sample of stainless steel penetrations and dissimilar metal welds, including the fuel transfer tubes, that may be susceptible to cracking due to sec and are leading indicators relative to cyclic loading.

Examinations and evaluations are performed in accordance with the requirements of ASME Section XI, Subsection IWE, which provides acceptance standards for the containment pressure boundary components. Areas identified with damage or degradation that exceed acceptance standards require an engineering evaluation or require correction by repair or replacement. Such areas are corrected by repair or replacement in accordance with IWE-3122 or accepted by engineering evaluation.

NUREG-2191 Consistency The PBN ASME Section XI, Subsection IWE AMP, with enhancements, will be consistent with the 10 elements of NUREG-2191, Section Xl.S1, "ASME Section XI, Subsection IWE~" as modified by SLR ISG Structures 2020 XX, Updated Aging Management Criteria for Structures Portions of the Subsequent License Renewal Guidance.

Exceptions to NUREG-2191 None.

Enhancements The PBN ASME Section XI, Subsection IWE AMP will be enhanced as follows for alignment with NUREG-2191. The one-time inspections for sec will be started no earlier than five years prior to the SPEO. The enhancements will be implemented

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-1 L-2021-144 Attachment 17 Page 3 of 3 and one-time inspections completed no later than six months prior to entering the SPEO.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-2 L-2021-144 Attachment 18 Page 1of7 RAI B.2.3.29-2

Background:

SRP-SLR Section 1.2.1 states, in part: "If a GALL-SLR Report AMP is selected to manage aging, the applicant may take one or more exceptions to specific GALL-SLR Report AMP program elements. Exceptions are portions of the GALL-SLR Report AMP that the applicant does not intend to implement, which the staff will review on a case-by-case basis. Any deviation or exception to the GALL-SLR Report AMP should be described and justified."

SLRA Section B.2.3.29, "ASME Section XI, Subsection IWE," program, as amended by Supplement 1 dated April 21, 2021, with enhancements (and no exceptions) claims consistency with GALL-SLR AMP Xl.S1. To make the program element consistent with that in GALL-SLR AMP Xl.S1, the SLRA included an enhancement to the "detection of aging effects" program element of the AMP, related to supplemental one-time supplemental volumetric examination of the containment liner (SLRA Table 16-3,

• License Renewal (LR) Commitment 33(f)), that states:

Issue:

Augment existing procedures to implement a one-time supplemental volumetric inspection of metal liner surfaces that samples randomly selected as well as focused locations susceptible to loss of thickness due to corrosion from the concrete side if triggered by plant-specific OE identified through code inspections after the date of issuance of the first renewed license for each unit. This sampling is conducted to demonstrate with 95% confidence, that 95% of accessible portion of the liner is not experiencing greater than 10% wall loss.

[emphasis added]

The staff is unable to determine that the "detection of aging effects" program element, with the above stated enhancement, will be consistent with that in GALL-SLR AMP Xl.S1 because of the following issues identified with regard to the enhancement (LR Commitment 33(f)):

1) The trigger specified in the GALL-SLR is the site-specific occurrence or recurrence of the stated plant-specific OE without regard to the method, program, or process by which (how) it is identified. Contrary to this, the SLRA enhancement states that the triggering OE would be specific to that identified through code inspections (emphasis added), which would be an unjustified exception to the GALL-SLR AMP Xl.S1. Past nuclear power industry operating experience has identified corrosion of the containment liner that originated from the inaccessible (concrete) side during repair/replacement activities, maintenance walkdowns, Appendix J activities, and can be identified by several other means in an operating plant. Contrary to the intent of the GALL-SLR AMP Xl.S1, tying the OE identification specifically to "code inspections" introduces technicalities for potentially not performing one-time supplemental volumetric examination if identified by other plant processes or programs.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAJ No. B.2.3.29-2 L-2021-144 Attachment 18 Page 2 of 7

2) There is no implementation schedule provided for LR Commitment 33(f) for the case where the triggering OE occurs during the subsequent period of extended operation.

Request:

1) Provide a revised enhancement to the "detection of aging effects" program element in SLRA Section B.2.3.29, related to one-time supplemental volumetric examination, and the associated LR Commitment 33(f) that would make the PBN AMP program element consistent with that in GALL-SLR AMP Xl.S1, that addresses Issue 1 above with regard to how (method, program or process) the triggering operating experience is identified.

2) Additionally, provide the implementation schedule for LR Commitment 33(f) for the case where the triggering OE occurs during the subsequent period of extended operation.

NEPB Response:

1) SLRA Supplement 1 Attachment 29 (SLRA Table 16-3 Commitment 33(f) and Section B.2.3.29) is revised as described below to provide further clarification that volumetric examinations will be triggered by plant specific OE after the date of issuance of the first renewed license for each unit.

2) SLRA Supplement 1 Attachment 29 (SLRA Table 16-3 Commitment 33(f)) is revised as described below to clarify that the one-time volumetric examination will be completed on a schedule established by the PBN Corrective Action Process. The inspection will be scheduled to provide reasonable assurance that the metal liner intended function is maintained consistent with the CLB through the SPEO.

References:

1. NextEra Energy Point Beach, LLC (NEPB) Letter to NRC L-2021-081 dated April 21, 2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML21111A155)

Associated SLRA Revisions:

SLRA Supplement 1 Attachment 29 (SLRA Table 16-3 Commitment 33(f) and Section B.2.3.29) is revised.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-2 L-2021-144 Attachment 18 Page 3 of 7 SLRA Table 16-3 (pages A-100 through A-103, item 33), as amended by SLRA Aging Management Supplement 1, is revised as follows:

Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging Management NUREG-2191 Commitment Implementation Schedule Program or Activity Section (Section) 33 ASME Section XI, Xl.S1 Continue the existing PBN ASME Section XI, No later than 6 months prior Subsection IWE Subsection IWE AMP, including enhancement to:

to the SPEO, or no later (16.2.2.29) a) Augment existing procedures to specify that than the last refueling whenever replacement of bolting is required, outage prior to the SPEO, bolting material, installation torque or tension, i.e.:

and use of lubricants and sealants are in PBN1 : 04/05/30 accordance with the guidelines of EPRI PBN2: 09/08/32 NP-5769, "Degradation and Failure of Bolting in Start the one-time Nuclear Power Plants," EPRI TR-104213, inspections for cracking due "Bolted Joint Maintenance & Application Guide,"

to sec no earlier than five and the additional recommendations of years prior to the SPEO.

NUREG-1339, "Resolution of Generic Safety Complete one-time Issue 29: Bolting Degradation or Failure in Nuclear Power Plants.";

inspection of containment liner b) Augment existing procedures to specify that for locations in both units if structural bolting consisting of ASTM A325, degradation from ASTM F1852, and/or ASTM A490 bolts, the inaccessible {concrete}

preventive actions for storage, lubricants, and side is identified, in either stress corrosion cracking potential discussed in unit, on a schedule Section 2 of RCSC (Research Council for established b~ the PBN Structural Connections) publication

• Corrective Action "Specification for Structural Joints Using ASTM Program. Inspection will A325 or A490 Bolts," will be used.

be scheduled to provide c) Augment existing procedures to specify that reasonable assurance pressure retaining bolting is inspected for that the metal liner loosening and material condition affecting leak intended function is tightness or structural integrity.

maintained consistent

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-2 L-2021-144 Attachment 18 Page 4 of 7 Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging Management NUREG-2191 Commitment Program or Activity Section (Section) d)

• e)

• f)

Augment existing procedures to implement a one-time supplemental volumetric inspection of metal liner surfaces that samples randomly selected as well as focused locations susceptible to loss of thickness due to corrosion from the concrete side if triggered by plant-specific OE i11SAtifi151 Um1ei~l9 st5115 iAsi;issti@As after the date of issuance of the first renewed license for each unit. This sampling is conducted to demonstrate with 95%

confidence, that 95% of accessible portion of the liner is not experiencing greater than 10% wall loss.

• SLRA Revision associated with RAI B.2.3.29-3.

Implementation Schedule with the CLB through the SPEO.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-2 L-2021-144 Attachment 18 Page 5 of 7 SLRA Section B.2.3.29 (pages B-213 through B-215), as amended by SLRA Aging Management Supplement 1, is revised as follows:

B.2.3.29 ASME Section XI, Subsection IWE Program Description The PBN ASME Section XI, Subsection IWE AMP is an existing AMP that was formerly part of the ASME Section XI, Subsections IWE and IWL lnservice Inspection AMP. This AMP is performed in accordance with ASME Code Section XI, Subsection IWE, and consistent with 10 CFR 50.55a "Codes and Standards," with supplemental recommendations. This program will use the edition and addenda of ASME Section XI required by 10 CFR 50.55a, as reviewed and approved by the NRC staff for aging management under 10 CFR 54. Alternatives to these requirements that are aging management related will be submitted to the NRC in accordance with 10 CFR 50.55a prior to implementation.

This AMP includes periodic visual, surface, and volumetric examinations, where applicable, of the steel liner of each concrete containment and their integral attachments for signs of degradation, damage, irregularities including discernable liner plate bulges, and for coated areas distress of the underlying metal shell or liner, and corrective actions. Acceptability of inaccessible areas of steel containment shell or concrete containment steel liner is evaluated when conditions found in accessible areas indicate the presence of, or could result in, flaws or degradation in inaccessible areas.

If site-specific OE identified after the date of issuance of the first renewed license for each unit triggers the requirement to implement a one-time supplemental volumetric examination, then this inspection is performed by sampling randomly selected, as well as focused, liner locations susceptible to corrosion that are inaccessible from one side. Guidance provided in EPRI TR-107514 will be considered for sampling determinations. The trigger for this one-time examination is site-specific occurrence or recurrence of liner corrosion (base metal material loss exceeding 10 percent of nominal plate thickness) that is determined to originate from the inaccessible (concrete) side. /\\ny such instance would be identified through code inspections performed since 10/05/10 for Unit 1 or 03/08/13 for Unit 2.Based on a review of current PBN operating experience, no such triggers have occurred.

Coated surfaces are visually inspected for evidence of conditions that indicate degradation of the underlying base metal. Coatings are a design feature of the base material and are not credited with managing loss of material. The PBN Protective Coating Monitoring and Maintenance AMP (Section B.2.3.36) is used for the monitoring and maintenance of protective containment coatings in relation to reasonable assurance of emergency core cooling system operability. Concrete portions of containments are inspected by the separate PBN ASME Section XI, Subsection IWL AMP (Section B.2.3.30).

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-2 L-2021-144 Attachment 18 Page 6 of 7 SLRA Section B.2.3.29 (pages B-213 through B-215), as amended by SLRA Aging Management Supplement 1, revision continued:

Surface conditions are monitored through visual examinations to determine the existence of corrosion. Surfaces are examined for evidence of flaking, blistering, peeling, discoloration, wear, pitting, excessive corrosion, arc strikes, gouges, surface discontinuities, dents, or other signs of surface irregularities. Pressure-retaining bolting is examined for loosening and material conditions that cause the bolted connection to affect either containment leak-tightness or structural integrity. Moisture barriers are visually inspected for degradation per Category E-A.

Cumulative fatigue damage for the PBN liner and steel piping (and ventilation) penetrations for the containment structures is addressed in the Containment Liner Plate, Metal Containments, and Penetrations Fatigue Analysis TLAA for SLR (Section 4.6). Cracking due to cyclic loading of all non-piping penetrations (hatches, electrical penetrations, etc.) that are subject to cyclic loading but have no current licensing bases fatigue analysis will be managed by the 10 CFR Part 50, Appendix J AMP (Section B.2.3.32) aAtfor supplemental surface examinations \\Gf-other appropriate examination/evaluation methods) or enhanced visual examinations using the ASME Section XI, Subsection IWE AMP. This AMP will also include supplemental one-time inspections within 5 years prior to the SPEO for a representative sample of stainless steel penetrations and dissimilar metal welds, including the fuel transfer tubes, that may be susceptible to cracking due to sec and are leading indicators relative to cyclic loading.

Examinations and evaluations are performed in accordance with the requirements of ASME Section XI, Subsection IWE, which provides acceptance standards for the containment pressure boundary components. Areas identified with damage or degradation that exceed acceptance standards require an engineering evaluation or require correction by repair or replacement. Such areas are corrected by repair or replacement in accordance with IWE-3122 or accepted by engineering evaluation.

NUREG-2191 Consistency Exceptions to NUREG-2191 None.

• SLRA Revision associated with RAI B.2.3.29-1.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-2 L-2021-144 Attachment 18 Page 7 of 7 SLRA Section B.2.3.29 (pages B-213 through B-215), as amended by SLRA Aging Management Supplement 1, revision continued :

2.

Enhancements The PBN ASME Section XI, Subsection IWE AMP will be enhanced as follows for alignment with NUREG-2191. The one-time inspections for SCC will be started no earlier than five years prior to the SPEO. The enhancements will be implemented and one-time inspections completed no later than six months prior to entering the SPEO.

Element Affected Enhancement Preventive Actions Augment existing procedures to specify that whenever replacement of bolting is required, bolting material, installation torque or tension, and use of lubricants and sealants are in accordance with the guidelines of EPRI NP-5769, "Degradation and Failure of Bolting in Nuclear Power Plants," EPRI TR-104213, "Bolted Joint Maintenance & Application Guide," and the additional recommendations of NUREG-1339, "Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants."

2. Preventive Actions Augment existing procedures to specify that for structural bolting consisting of ASTM A325, ASTM F1852, and/or ASTM A490 bolts, the preventive actions for storage, lubricants, and stress corrosion cracking potential discussed in Section 2 of RCSC (Research Council for Structural Connections) publication "Specification for Structural Joints Using ASTM A325 or A490 Bolts," will be used.

3. Parameters Monitored Augment existing procedures to specify that pressure retaining bolting or Inspected is inspected for loosening and material condition affecting leak tightness or structural inteority.

4. Detection of Aging Effects

4. Detection of Aging Effects

4. Detection of Aging Augment existing procedures to implement a one-time supplemental Effects volumetric inspection of metal liner surfaces that samples randomly selected as well as focused locations susceptible to loss of thickness due to corrosion from the concrete side if triggered by plant-specific OE ileRtifiel tt;irn1s1~R ssle iRs13estisRs after the date of issuance of the first renewed license for each unit. This sampling is conducted to demonstrate with 95% confidence, that 95% of accessible portion of the liner is not experiencing greater than 10% wall loss.

7. Corrective Actions If sec is detected as a result of the supplemental one-time inspections, additional inspections will be conducted in accordance with the site's corrective action process.

• SLRA Revision associated with RAI B.2.3.29-3.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-3 L-2021-144 Attachment 19 Page 1of8 RAI B.2.3.29-3

Background:

SLRA Section B.2.3.29, "ASME Section XI, Subsection IWE, program, as amended by Supplement 1 dated April 21, 2021, includes enhancements to the "detection of aging effects" program element and corresponds to LR Commitments 33(d) and 33(e) that identify "surface examinations or enhanced visual examinations" methods to detect cracking due to cyclic loading or due to stress corrosion cracking (SCC). (emphasis added)

With regard to the enhanced visual examination method, GALL-SLR AMP Xl.M32, "One-Time Inspection, identifies EVT-1 as an acceptable inspection method to detect cracking due to sec or cyclic loading for pressure-retaining components.

Issue:

With regard to the enhanced visual alternative, the staff is not clear what specific enhanced visual inspection method is intended to be used with regard to LR Commitments 33(d) and 33(e) and corresponding program enhancements to make its determination of the capability of the technique to detect cracking due to cyclic loading or sec.

Request:

1) Explicitly state the specific enhanced visual inspection method (e.g. EVT-1) in LR Commitments 33(d) and 33(e) and the corresponding B.2.3.29 program enhancements (and FSAR Supplement 16.2.29) that will be used to detect cracking due to cyclic loading or sec.

2) If it is other than the EVT-1 method, describe the specific enhanced visual method and justify its adequacy to detect cracking due to cyclic loading or sec.

NEPB Response:

SLRA Commitments 33(d) and 33(e), along with the enhancements in Section B.2.3.29, as revised by SLRA Aging Management Supplement 1, is amended as described below to clarify that the specific enhanced visual inspection method used to detect cracking due to cyclic loading or SCC will be EVT-1.

References:

1. NextEra Energy Point Beach, LLC (NEPB) Letter to NRC L-2021-081 dated April 21, 2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML21111A155)

Associated SLRA Revisions:

SLRA Section 16.2.2.29, as amended by SLRA Aging Management Supplement 1, is revised as follows:

The PBN ASME Section XI, Subsection IWE AMP is an existing AMP that was formerly part of the ASME Section XI, Subsections IWE and IWL lnservice Inspection

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-3 L-2021-144 Attachment 19 Page 2 of 8 AMP. This condition monitoring AMP is in accordance with ASME Code Section XI, Subsection IWE, and consistent with 10 CFR 50.55a, "Codes and Standards," with supplemental recommendations. This program will use the edition and addenda of ASME Section XI required by 10 CFR 50.55a, as reviewed and approved by the NRC staff for aging management under 10 CFR 54. Alternatives to these requirements that are aging management related will be submitted to the NRC in accordance with 10 CFR 50.55a prior to implementation.

The AMP includes periodic visual, surface, and volumetric examinations, where applicable, of the steel liner of each concrete containment and their integral attachments for signs of degradation, damage, irregularities including discernable liner plate bulges, and for coated areas distress of the underlying metal shell or liner, and corrective actions. Acceptability of inaccessible areas of steel containment shell or concrete containment steel liner is evaluated when conditions found in accessible areas indicate the presence of, or could result in, flaws or degradation in inaccessible areas.

In addition, the program includes supplemental surface examination QL enhanced visual examination (EVT-1) to detect cracking for specific pressure-retaining components including all non-piping penetrations (hatches, electrical penetrations, etc.) that are subject to cyclic loading but have no CLB fatigue analysis and are not subject to local leak rate testing.

If triggered by plant-specific OE, a one-time supplemental volumetric examination will be performed by sampling randomly selected as well as focused locations susceptible to loss of thickness due to corrosion of containment shell or liner that is inaccessible from one side. Inspection results are compared with prior recorded results in acceptance of components for continued service.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-3 L-2021-144Attachment19 Page 3 of 8 SLRA Table 16-3 (pages A-100 through A-103, item 33), as amended by SLRA Aging Management Supplement 1 is revised as follows:

Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging Management NUREG-2191 Commitment Implementation Schedule Program or Activity Section (Section) 33 ASME Section XI, Xl.S1 Continue the existing PBN ASME Section XI, Subsection No later than 6 months prior to Subsection IWE IWE AMP, including enhancement to:

the SPEO, or no later than the (16.2.2.29) a) Augment existing procedures to specify that last refueling outage prior to the whenever replacement of bolting is required, SPEO, i.e.:

bolting material, installation torque or tension, and PBN1 : 04/05/30 use of lubricants and sealants are in accordance PBN2: 09/08/32 with the guidelines of EPRI NP-5769, Start the one-time inspections for "Degradation and Failure of Bolting in Nuclear cracking due to sec no earlier Power Plants," EPRI TR-104213, "Bolted Joint than five years prior to the SPEO.

Maintenance & Application Guide," and the additional recommendations of NUREG-1339, "Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants.";

b) Augment existing procedures to specify that for structural bolting consisting of ASTM A325, ASTM F1852, and/or ASTM A490 bolts, the preventive actions for storage, lubricants, and stress corrosion cracking potential discussed in Section 2 of RCSC (Research Council for Structural Connections) publication "Specification for Structural Joints Using ASTM A325 or A490 Bolts," will be used.

c)

Augment existing procedures to specify that pressure retaining bolting is inspected for loosening and material condition affecting leak tightness or structural integrity.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-3 L-2021-144 Attachment 19 Page 4 of 8 Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging Management NUREG-2191 Commitment Program or Activity Section (Section) d) Augment existing procedures to implement periodic supplemental surface examinations or enhanced visual examination (EVT-1) at intervals no greater than 10-years to detect cracking due to cyclic loading of all non-piping penetrations (hatches, electrical penetrations, etc.) that are subject to cyclic loading but have no current licensing bases fatigue analysis and are not subject to local leak rate testing.

e) Augment existing procedures to implement supplemental one-time surface examinations or enhanced visual examinations (EVT-1),

performed by qualified personnel using methods capable of detecting cracking due to sec, comprising (a) a representative sample (two) of the stainless steel penetrations or dissimilar metal welds associated with high-temperature (temperatures above 140°F) stainless steel piping systems in frequent use on each unit; and (b) the stainless steel fuel transfer tube on each unit. If cracking is detected as a result of the supplemental one-time inspections, additional inspections will be conducted in accordance with the site's corrective action process. This will include 1 additional penetration with dissimilar metal welds associated with greater than 140

°F stainless steel piping systems for each unit until SCC is no longer detected. Periodic inspection of subject penetrations with dissimilar metal welds for cracking will be added to the PBN ASME Section XI, Implementation Schedule

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-3 L-2021-144 Attachment 19 Page 5 of 8 Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging Management NUREG-2191 Commitment Program or Activity Section (Section)

Subsection IWE AMP if necessary, depending on the inspection results.

f)

• SLRA Revision associated with RAI B.2.3.29-2.

Implementation Schedule

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 8.2.3.29-3 L-2021-144 Attachment 19 Page 6 of 8 SLRA Section B.2.3.29 (pages 8-213 through 8-215), as amended by SLRA Aging Management Supplement 1, is revised as follows:

B.2.3.29 ASME Section XI, Subsection IWE Program Description The PBN ASME Section XI, Subsection IWE AMP is an existing AMP that was formerly part of the ASME Section XI, Subsections IWE and IWL lnservice Inspection AMP. This AMP is performed in accordance with ASME Code Section XI, Subsection IWE, and consistent with 10 CFR 50.55a "Codes and Standards," with supplemental recommendations. This program will use the edition and addenda of ASME Section XI required by 10 CFR 50.55a, as reviewed and approved by the NRC staff for aging management under 10 CFR 54. Alternatives to these requirements that are aging management related will be submitted to the NRC in accordance with 10 CFR 50.55a prior to implementation.

This AMP includes periodic visual, surface, and volumetric examinations, where applicable, of the steel liner of each concrete containment and their integral attachments for signs of degradation, damage, irregularities including discernable liner plate bulges, and for coated areas distress of the underlying metal shell or liner, and corrective actions. Acceptability of inaccessible areas of steel containment shell or concrete containment steel liner is evaluated when conditions found in accessible areas indicate the presence of, or could result in, flaws or degradation in inaccessible areas.

Coated surfaces are visually inspected for evidence of conditions that indicate degradation of the underlying base metal. Coatings are a design feature of the base material and are not credited with managing loss of material. The PBN Protective Coating Monitoring and Maintenance AMP (Section B.2.3.36) is used for the monitoring and maintenance of protective containment coatings in relation to reasonable assurance of emergency core cooling system operability. Concrete portions of containments are inspected by the separate PBN ASME Section XI, Subsection IWL AMP (Section B.2.3.30).

Surface conditions are monitored through visual examinations to determine the existence of corrosion. Surfaces are examined for evidence of flaking, blistering,

• SLRA Revision associated with RAI 8.2.3.29-2.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-3 L-2021-144 Attachment 19 Page 7 of 8 SLRA Section B.2.3.29 (pages B-213 through B-215), as amended by SLRA Aging Management Supplement 1, revision continued:

peeling, discoloration, wear, pitting, excessive corrosion, arc strikes, gouges, surface discontinuities, dents, or other signs of surface irregularities. Pressure-retaining bolting is examined for loosening and material conditions that cause the bolted connection to affect either containment leak-tightness or structural integrity. Moisture barriers are visually inspected for degradation per Category E-A.

Cumulative fatigue damage for the PBN liner and steel piping (and ventilation) penetrations for the containment structures is addressed in the Containment Liner Plate, Metal Containments, and Penetrations Fatigue Analysis TLAA for SLR (Section 4.6). Cracking due to cyclic loading of all non-piping penetrations (hatches, electrical penetrations, etc.) that are subject to cyclic loading but have no current licensing bases fatigue analysis will be managed by the 10 CFR Part 50, Appendix J AMP (Section B.2.3.32) or supplemental surface examinations or enhanced visual examinations (EVT-1) using the ASME Section XI, Subsection IWE AMP. This AMP will also include supplemental one-time inspections within 5 years prior to the SPEO for a representative sample of stainless steel penetrations and dissimilar metal welds, including the fuel transfer tubes, that may be susceptible to cracking due to SCC and are leading indicators relative to cyclic loading.

Examinations and evaluations are performed in accordance with the requirements of ASME Section XI, Subsection IWE, which provides acceptance standards for the containment pressure boundary components. Areas identified with damage or degradation that exceed acceptance standards require an engineering evaluation or require correction by repair or replacement. Such areas are corrected by repair or replacement in accordance with IWE-3122 or accepted by engineering evaluation.

NUREG-2191 Consistency Exceptions to NUREG-2191 None.

Enhancements The PBN ASME Section XI, Subsection IWE AMP will be enhanced as follows for alignment with NUREG-2191. The one-time inspections for sec will be started no earlier than five years prior to the SPEO. The enhancements will be implemented and one-time inspections completed no later than six months prior to entering the SPEO.

• SLRA Revision associated with RAI B.2.3.29-1.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-3 L-2021-144 Attachment 19 Page 8 of 8 SLRA Section B.2.3.29 (pages B-213 through B-215), as amended by SLRA Aging Management Supplement 1, revision continued:

Element Affected Enhancement

2. Preventive Actions Augment existing procedures to specify that whenever replacement of bolting is required, bolting material, installation torque or tension, and use of lubricants and sealants are in accordance with the guidelines of EPRI NP-5769, "Degradation and Failure of Bolting in Nuclear Power Plants," EPRI TR-104213, "Bolted Joint Maintenance & Application Guide," and the additional recommendations of NUREG-1339, "Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants."

2. Preventive Actions Augment existing procedures to specify that for structural bolting consisting of ASTM A325, ASTM F1852, and/or ASTM A490 bolts, the preventive actions for storage, lubricants, and stress corrosion cracking potential discussed in Section 2 of RCSC (Research Council for Structural Connections) publication "Specification for Structural Joints Using ASTM A325 or A490 Bolts," will be used.

3. Parameters Monitored Augment existing procedures to specify that pressure retaining bolting or Inspected is inspected for loosening and material condition affecting leak tightness or structural integrity.

4. Detection of Aging Augment existing procedures to implement periodic supplemental Effects surface examinations or enhanced visual examinations (EVT-1) at intervals no greater than 10-years to detect cracking due to cyclic loading of all non-piping penetrations (hatches, electrical penetrations, etc.) that are subject to cyclic loading but have no current licensing bases fatigue analysis and are not subject to local leak rate testing.

4. Detection of Aging Augment existing procedures to implement supplemental one-time Effects surface examinations or enhanced visual examinations (EVT-1),

performed by qualified personnel, comprising (a) a representative sample (two) of the stainless steel penetrations or dissimilar metal welds associated with high-temperature (temperatures above 140°F) stainless steel piping systems in frequent use on each unit; and (b) the stainless steel fuel transfer tube on each unit. This will include 1 additional penetration with dissimilar metal welds associated with greater than 140 °F stainless steel piping systems for each unit until cracking is no longer detected. Periodic inspection of subject penetrations with dissimilar metal welds will be added to the PBN ASME Section XI, Subsection IWE AMP if necessary, depending on the inspection results.

4. Detection of Aging Effects

7. Corrective Actions If SCC is detected as a result of the supplemental one-time inspections, additional inspections will be conducted in accordance with the site's corrective action process.

• SLRA Revision associated with RAI B.2.3.29-2.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-4 L-2021-144 Attachment 20 Page 1 of 5 RAI 8.2.3.29-4

Background:

SLRA Section B.2.3.29, under the "Operating Experience" subsection, addresses PBN's evaluation of NRC Information Notice (IN) 2014-07 "Degradation of Leak Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner," and Regulatory Issue Summary (RIS) 2016-07 "Containment Shell or Liner Moisture Barrier Inspection" regarding leak-chase interface components that serve a moisture barrier function, and it states in part:

..., the plant does have accessible capped lines for the leak chase. The locations of the leak chase channel vents were documented on applicable PBN drawings and included in IWE visual examinations.

It further states, with regard to RIS 2016-07, Issue:

PBN has been proactively performing exams on moisture barriers once each Period during the 2nd Interval, and this RIS has been incorporated into the 3rd IWE Interval program plan.

The AMR results in SLRA Table 3.5.2-1 do not include line item(s) that correspond to the above components that serve a moisture barrier function for inaccessible areas of containment floor liner and that correspond to the above statements made in the SLRA related to OE described in IN 2014-07 and RIS 2016-07. The staff is unable to verify the above referenced statements in the SLRA nor that the AMR results in SLRA Table 3.5.2-1 reflect aging management for containment components that serve a moisture barrier function per recommendations for industry OE described in IN 2014-07 and RIS 2016-07.

Request:

1) Clarify and state if accessible leak-chase channel interface components at or near the containment floor that serve a moisture barrier function (i.e., prevent intrusion of moisture into inaccessible portions of liner) are monitored by the ASME Section XI, Subsection IWE program, consistent with the recommendations in NRC IN 2014-07 and RIS 2016-07 and as alluded to in the SLRA.

2) Provide appropriate AMR line items in the SLRA Table 3.5.2-1 for above containment components that serve a moisture barrier function per NRC IN 2014-07 and RIS 2016-07. If not, justify the SLRA statements referenced in the Background section above and justify the adequacy of the PBN ASME Section XI, Subsection IWE program to manage liner degradation in inaccessible areas related to operating experience described in NRC IN 2014-07 and RIS 2016-07.

NEPB Response:

1) The seal-welded leak chase channel caps are level with the floor rather than in a depression in the floor such that moisture is unlikely to collect on or around the

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-4 L-2021-144 Attachment 20 Page 2 of 5 caps. PBN is currently committed to the 2007 Edition with 2008 Addenda of ASME Section XI. Table IWE-2500-1 of that edition indicates that containment moisture barrier materials include "caulking, flashing, and other sealants."

Currently, the leak chase channel caps are accessible surfaces (Table IWE-2500-1 Category E1.11) and the moisture barriers provide the moisture barriers function (Table IWE-2500-1 Category E1.30). The frequencies of examinations are specified in the 2007 edition through 2008 addenda of ASME Section XI, Subsection IWE, as modified by 10 CFR 50.55a and approved alternatives. Both categories are inspected 100% during each inspection period, both in the first inspection interval and in successive inspection intervals.

2) The seal-welded leak chase channel caps are part of the "Liner plate anchors and attachments (accessible)" component type in SLRA Table 3.5.2-1, which corresponds with NUREG-2191 Item ll.A1-CP-35 and Table 1 Item 3.5-1, 035. A new plant-specific Note 14 is added to SLRA Table 3.5.2-1 as described below to clarify that the seal welded leak chase channel caps are included.

3) There is also a minor editorial correction to Plant-Specific Note 10 reflected below.

References:

None.

Associated SLRA Revisions:

SLRA Table 3.5.2-1 (Page 3.5-89) and Plant-Specific Notes (Page 3.5-96), as updated by Supplement 1 (Attachments 21 and 29) are further revised.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-4 L-2021-144 Attachment 20 Page 3 of 5 SLRA Table 3.5.2-1 (page 3.5-89), including plant-specific notes (page 3.5-96), as amended by SLRA Aging Management Supplement 1, is revised as follows:

Table 3.5.2-1: Containment Building Structure and Internal Structural Components - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-2191 Table 1 Notes Type Function Requiring Program Item Item Management Liner plate Pressure Steel Air-indoor Loss of 10 CFR Part 50, ll.A1.CP-35 3.5-1, 035 A

(accessible) boundary uncontrolled material Appendix J Structural (B.2.3.32) support ASME Section XI, Subsection IWE (B.2.3.29)

Liner plate Pressure Steel Air-indoor Loss of 10 CFR Part 50, ll.A1.CP-98 3.5-1, 005 A

(inaccessible) boundary uncontrolled material Appendix J Structural (B.2.3.32) support ASME Section XI, Subsection IWE (B.2.3.29)

Liner plate Direct flow Steel Air-indoor Loss of 10 CFR Part 50, ll.A1.CP-35 3.5-1, 035 A

and keyway Pressure uncontrolled material Appendix J channel boundary (B.2.3.32)

(accessible)

Structural ASME Section XI, support Subsection IWE (B.2.3.29)

Liner plate Direct flow Steel Air-indoor Loss of 10 CFR Part 50, 11.A1.CP-98 3.5-1, 005 A

and keyway Pressure uncontrolled material Appendix J channel boundary (B.2.3.32)

(inaccessible)

Structural ASME Section XI, support Subsection IWE (B.2.3.29)

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-4 L-2021-144 Attachment 20 Page 4 of 5 Table 3.5.2-1: Containment Building Structure and Internal Structural Components - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-2191 Table 1 Type Function Requiring Program Item Item Management Liner plate Pressure Steel Air-indoor Loss of 10 CFR Part 50, ll.A1.CP-35 3.5-1, 035 anchors and boundary uncontrolled material Appendix J attachments Structural (B.2.3.32)

(accessible) support ASME Section XI, Subsection IWE (B.2.3.29)

Liner plate Pressure Steel Air-indoor Loss of 10 CFR Part 50, ll.A1.CP-98 3.5-1, 005 anchors and boundary uncontrolled material Appendix J attachments Structural (B.2.3.32)

(inaccessible) support ASME Section XI, Subsection IWE (B.2.3.29)

Plant Specific Notes Notes A.._H A

1. Copper alloy is not addressed as a structural component in NUREG-2191. However, the environment, aging effects (cracking and loss of material) and aging management programs for steel air lock, hatch components are conservatively also applicable to the copper alloy airlock bushings.

2. PBN containments are located entirely inside the Fac;ade building and are not associated with an air - outdoor environment.

However, freeze-thaw conditions are still possible during winter months where water or groundwater collects as the Fac;ade building is non-heated.

3. The tendon gallery adjacent to each PBN Unit's containment, inside the Fac;ade building, is part of the containment base mat in the top few feet. The tendon galleries are not associated with an air - outdoor environment. However, freeze thaw conditions are still possible during winter months where water or groundwater collects as the Fac;ade building is non-heated.

4. Structural stainless steel that is exposed to air - indoor uncontrolled during normal plant operation is inspected under the Structures Monitoring (B.2.3.34) AMP, or in the case of the transfer canal the ASME Section XI, Subsection IWE (B.2.3.29)

AMP, the structural equivalent of the NUREG-2191 Xl.M36, Externals Surfaces Monitoring of Mechanical Components AMP.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.29-4 L-2021-144 Attachment 20 Page 5 of 5

5. Liner moisture barriers are at the junction where the liner is embedded in the concrete slab and for the core holes in the concrete slab that allow inspection of the liner.

6. Penetration assemblies for high temperature stainless steel piping systems only for sec, or as a leading indicator for cyclic loading of other mechanical penetration sleeves/assemblies with dissimilar metal welds.

7. Primary shield wall, and attached biological shield wall, with a % inch steel liner surrounds the reactor cavity and the reactor vessel support structure passes through and is attached to it at certain points. Existing inspections, through the Structures Monitoring (B.2.3.34) AMP, manage the condition of the shield wall.

8. As described in the RAI responses/supplements for the first 2 PWRs with renewed licenses for 80 years, irradiation embrittlement of the steel reactor vessel support structure columns and beams requires analysis. Existing inspections, through the ASME Section XI, Subsection IWF (B.2.3.31 ) AMP, manages the condition of the reactor vessel support.

9. Insulation for main steam and feedwater penetrations are encased in steel penetration covers in the annulus and there are no plausible aging effects that could degrade the calcium silicate or amosite asbestos (with a silicate binder) insulation.

Furthermore, temperature measurements for the penetrations are within UFSAR allowable.

10. Based on SLR-ISG-Structures-2020-XX, "Updated Aging Management Criteria for Structures Portions of Subsequent License Renewal Guidance.:.", the existing Structures Monitoring (B.2.3.34) AMP is credited rather than a plant-specific AMP and is supplemented by the ASME Section XI, Subsection IWL (B.2.3.32) AMP as appropriate.

11. Component also provides a fire barrier function as evaluated in the Fire Protection Program Design Document that is physically equivalent to the structural functions managed under the associated Containment structural programs.

12. Steel mechanical (and ventilation) penetration assemblies are covered by a fatigue waiver, as described in Section 4.6.

13. The loss of fracture toughness aging effect is managed by the ASME Section XI, Subsection IWF (B.2.3.31) AMP.

14. Includes seal-welded leak chase channel caps.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.31 -1 L-2021 -144 Attachment 21 Page 1 of 4 10.SLRA Section B.2.3.31, "ASME Section XI, Subsection IWF" Regulatory Basis:

Paragraph 54.21 (a)(3) of 10 CFR requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis for the subsequent period of extended operation. As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL-SLR Report when evaluation of the matter in the GALL-SLR Report applies to the plant.

RAI B.2.3.31-1

Background:

SLRA Section B.2.3.31 states that the PBN ASME Section XI, Subsection IWF aging management program (AMP), with enhancements, will be consistent with one exception

[to scope of program] to NUREG-2191 [GALL-SLR] Section Xl.S3, "ASME Section XI,

Subsection IWF." The SLRA AMP does not take any exception to the "preventive actions" program element.

SLRA Section B.2.3.31, as amended by Supplement 1 dated April 21, 2021, includes an amended enhancement (SLR Commitment 35(d)) to the "preventive actions" program element of the SLRA "... Additionally, Molybdenum disulfide [MoS2] thread lubricants are not used."

The corresponding program element in GALL-SLR AMP Xl.S3 states, in part, that "molybdenum disulfide and other lubricants containing sulfur should not be used."

(emphasis added)

During the audit, the staff reviewed Maintenance Instruction (Ml) 29.1, Rev. 14, "Use of Thread Lubricants and Sealants," and noted that it does not prohibit other lubricants containing sulfur.

SRP-SLR Section 1.2.1 states, in part: "If a GALL-SLR Report AMP is selected to manage aging, the applicant may take one or more exceptions to specific GALL-SLR Report AMP program elements. Exceptions are portions of the GALL-SLR Report AMP that the applicant does not intend to implement, which the staff will review on a case-by-case basis. Any deviation or exception to the GALL-SLR Report AMP should be described and justified."

Issue:

Contrary to the applicant's claim of consistency with the GALL-SLR AMP, the above referenced enhancement to the applicant's B.2.3.31 AMP does not appear to be consistent with the corresponding program element in the GALL-SLR Xl.S3 AMP with regard to the GALL-SLR recommendation that other lubricants containing sulfur should also not be used, in addition to molybdenum disulfide. It is not clear whether the applicant plans to continue using (other than MoS2) lubricants containing Sulphur on high-strength bolting. The SLRA does not take an exception to the preventive actions

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.31-1 L-2021-144 Attachment 21 Page 2 of 4 program element for other than MoS2 lubricants containing Sulphur. Also, the language of the SLRA B.2.3.31 AMP enhancement is not structured the same as corresponding SLR Commitment 35(d).

Request:

1) Provide a revised enhancement and a revised SLR Commitment 35(d) that would make the "preventive actions" program elem*ent consistent, as claimed in the SLRA, with the GALL-SLR AMP Xl.S3 regarding the recommendation that other lubricants containing sulfur should also not be used in addition to molybdenum disulfide.

2) Alternatively, provide the technical justification for the exception to the "preventive actions" program element GALL-SLR AMP Xl.S3 specifically with regard to the recommendation that, in addition to MoS2, other lubricants containing Sulphur should also be not used on high-strength bolting.

NEPB Response:

NEPB does not intend to continue to use lubricants containing sulfur and has initiated a procedure change to prohibit the use of molybdenum disulfide and other lubricants containing sulfur.

1) The enhancement to the preventive actions element of the PBN ASME Section XI, Subsection IWF and SLR Commitment 35(d) are revised to clarify that in addition to molybdenum disulfide, lubricants containing sulfur will not be used.

2) No exception to the preventive actions program element of GALL-SLR AMP Xl.S3 is being proposed.

References:

1. NextEra Energy Point Beach, LLC (NEPB) Letter to NRC L-2021-081 dated April 21, 2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML21111A155).

Associated SLRA Revisions:

SLRA Table 16-3 (Item No. 35(d))-Appendix A, Section 16.4 (Page A-103, as amended by SLRA Aging Management Supplement 1) is revised as follows:

d) Augment existing procedures to specify that for structural bolting consisting of ASTM A325, ASTM F1852, ASTM F2280 and/or ASTM A490 bolts, the preventive actions for storage, lubricants, and stress corrosion cracking potential discussed in Section 2 of RCSC (Research Council for Structural Connections) publication "Specification for Structural Joints Using ASTM A325 or A490 Bolts," will be used. Additionally, molybdenum disulfide and other thread lubricants containing sulfur will not be used.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.31-1 L-2021-144 Attachment 21 Page 3 of 4 SLRA Section B.2.3.31 (page B-225, as amended by SLRA Aging Management Supplement 1) is revised as follows:

This AMP emphasizes proper selection of bolting material, lubricants, and installation torque or tension to prevent or minimize loss of bolting preload of structural bolting and cracking of high: strength bolting. As noted below in the enhancement discussion, the AMP also includes the preventive actions for storage requirements of high-strength bolts and ensuring that molybdenum disulfide and other thread lubricants containing sulfur are not used for structural bolting. The requirements of ASME Code Section XI, Subsection IWF are supplemented to include volumetric examination of high: strength bolting for cracking. This AMP will also include a one-time inspection within 5 years prior to the SPEO of an additional 5 percent of piping supports from the remaining IWF population that are considered most susceptible to age-related degradation. Inspections of elastomeric vibration isolation elements to detect hardening are also included if the vibration isolation function is suspect.

SLRA Section B.2.3.31 (page B-227, as amended by SLRA Aging Management Supplement 1) is revised as follows:

Element Affected Enhancement

2. Preventive Actions Augment existing procedures to specify that whenever replacement of bolting is required, bolting material, installation torque or tension, and use of lubricants and sealants are in accordance with the guidelines of EPRI NP-5769, "Degradation and Failure of Bolting in Nuclear Power Plants," EPRI TR-104213, "Bolted Joint Maintenance &

Application Guide," and the additional recommendations of NUREG-1339, "Resolution of Generic Safety Issue 29:

Bolting Degradation or Failure in Nuclear Power Plants."

2. Preventive Actions Augment existing procedures to specify that for structural bolting molybdenum disulfide thread lubricants are not used and for bolting consisting of ASTM A325, ASTM F1852, ASTM F2280 and/or ASTM A490 bolts, the preventive actions for storage, lubricants, and stress corrosion cracking potential discussed in Section 2 of RCSC (Research Council for Structural Connections) publication "Specification for Structural Joints Using ASTM A325 or A490 Bolts," will be used. Additionally, molybdenum disulfide and other lubricants containing sulfur will not be used.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.31-1 L-2021-144 Attachment 21 Page 4 of 4 Element Affected Enhancement

3. Parameter Monitored or Augment existing procedures to specify that bolting within Inspected the scope of this program is inspected for loss of integrity of bolted connections due to self-loosening.

3. Parameter Monitored or Augment existing procedures to specify that elastomeric or Inspected polymeric vibration isolation elements are monitored for cracking, loss of material, and hardening.

4. Detection of Aging Effects Perform and document a one-time inspection of an additional 5% of the sample populations for Class 1, 2, and 3 piping supports. The additional supports will be selected from the remaining population of IWF piping supports and will include components that are most susceptible to age-related degradation.

4. Detection of Aging Effects Augment existing procedures to include tactile inspection (feeling, prodding) of elastomeric vibration isolation elements to detect hardening if the vibration isolation function is suspect.

4. Detection of Aging Effects Augment existing procedures to specify that, for ASME Class 1, 2 or 3 component supports, high: strength bolting greater than one inch nominal diameter, volumetric examination comparable to that of ASME Code,Section XI, Table IWB-2500-1, Examination Category B-G-1 will be performed to detect cracking in addition to the VT-3 examination. A re12resentative sam12le of bolts will be ins12ected during the ins12ection interval 12rior to the start of the SPEO and Hn each 10-year period during the subsequent period of extended operation, a FepFeseAtative sample of bolts 'Nill be iAspected. The sample will be 20% of the population (for a material I environment combination) up to a maximum of 25 bolts.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.31-2 L-2021-144 Attachment 22 Page 1 of 3 RAI 8.2.3.31-2

Background:

The "preventive actions" program element of GALL-SLR AMP Xl.S3 states: "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. The "parameters monitored or inspected" program element of GALL-SLR AMP Xl.S3 recommends that high strength bolting (actual measured yield strength greater than or equal to 150 ksi) in sizes greater than 1 inch nominal diameter should be monitored for sec.

SLRA Section B.2.3.31, as amended by Supplement 1, includes an amended enhancement (SLR Commitment 35(d)), which states, in part, that Molybdenum disulfide thread lubricants will not be used during the subsequent period of extended operation (SPEO).

During the audit, the staff reviewed Maintenance Instruction (Ml) 29.1, Rev. 14, "Use of Thread Lubricants and Sealants." Section 3.1.3 of this Ml states the solid components

[of Thread Lubricants] may include, among others, Molybdenum disulfide. Attachment A of the Ml includes Molykote as a proprietary brand of thread lubricants used at Point Beach, which is of molybdenum disulfide material. Also, the Ml does not prohibit other lubricants containing sulfur. It appears that Molybdenum disulfide lubricant may have been used at Point Beach in the past.

Since high-strength bolting is used and will continue to be used, and molybdenum disulfide or other lubricants containing Sulphur may have been used at PBN, SLRA Section B.2.3.31, as amended by Supplement 1 dated April 21, 2021, includes an enhancement (SLR Commitment 35(i) in Table 16-3 as amended by Supplement 1) to the "detection of aging effects" program element to include volumetric examinations of a representative sample of high-strength bolting greater than 1 inch diameter, to detect cracking due to SCC, in each 10-year interval during the SPEO.

Issue:

1) The staff noted that visual inspections of the currently implemented ASME Section XI, Subsection IWF program cannot detect SCC aging effect in high-strength bolting from the use of lubricants containing sulfur in high-strength bolting. The staff also noted that preventive measures are not currently in place to inhibit such an aging effect. Hence, it is possible that this aging effect may be present now or become present due to continued use of high-strength bolts as replacements coated with lubricants containing Sulphur, including MoS2, prior to SPEO. Hence, it is possible that such an aging effect may remain undetected until SLRA AMP B.2.3.31 volumetric examinations of a sample containing high-strength bolts coated with lubricants containing Sulphur are performed, which may be as much as ten years into the SPEO. The SRP-SLR Branch Technical Position RLSB-1 Section A.1.2.3.4, however, states that "detection of aging effects should occur before there is a loss of the SC-intended function(s). Therefore, for the period of time between the start of

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.31-2 L-2021-144 Attachment 22 Page 2 of 3 the SPEO and when the volumetric examinations are performed, it is not clear how the aging effect of cracking due to sec will be detected prior to a loss of intended function.

2) If PBN continues the use of MoS2 (or other lubricants containing Sulphur)-coated high-strength bolts susceptible to sec, there is the potential to increase the population of installed high-strength bolts (i.e., install additional high-strength bolts as replacement bolting) susceptible to SCC. It is not clear how a sample for volumetric examination representing the entire population of high-strength bolts will be established. It is also not clear how the program will assess the sample size and scope to ensure that it continues to monitor suspect high-strength bolts coated with Sulphur based lubricants, especially those that have used/are using MoSz.

Request:

1) Since volumetric examinations per SLRA Commitment 35(i) are planned for some time into the SPEO (could be as much towards the end of the first 10-year interval in the SPEO), provide information on whether and how the aging effect of cracking due to sec will be detected for the population of existing high strength bolts such that this aging effect can be managed consistent with SRP-SLR Branch technical Position RLSB-1 Section A.1.2.3.4 prior to entering the SPEO.

2) Discuss how the "parameters monitored or inspected" program element will identify and assess the adequacy of the representative high-strength bolting sample inspected for cracking due to SCC for existing and/or when additional susceptible high-strength bolts are installed.

3) Update applicable portions of the SLRA as necessary.

NEPB Response:

To ensure that the aging effect due to sec is managed prior to entering the SPEO:

1) Prior to the SPEO, Point Beach Nuclear will inspect a representative sample of high-strength bolting greater than one-inch nominal diameter for ASME Class 1, 2, or 3 component supports.

2) The sample will be comprised of 20 percent of the high-strength bolts (for a material/environment combination) up to a maximum of 25 bolts. The inspections will be performed using techniques capable of detecting cracking. If additional high-strength bolting is installed, this sample will continue to represent the most susceptible locations since molybdenum disulfide and other thread lubricants containing sulfur will be prohibited from use at Point Beach (see NEPB Response to RAI B.2.3.31 -1 in Attachment 21 to this letter).

3) Commitment 35(i) and SLRA Section B.2.3.31 are revised to reflect this change.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.31 -2 L-2021-144 Attachment 22 Page 3 of 3

References:

1. NextEra Energy Point Beach, LLC (NEPB) Letter to NRC L-2021-081 dated April 21,

2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML21111A155).

Associated SLRA Revisions:

SLRA Table 16-3 (Item No. 35(i))-Appendix A Section 16.4 (Page A-103, as amended by SLRA Aging Management Supplement 1) is revised as follows.

i)

Augment existing procedures to specify that, for ASME Class 1, 2, or 3 component supports, high.:strength bolting greater than one.:inch nominal diameter, volumetric examination comparable to that of ASME Code,Section XI, Table IWB.:2500-1, Examination Category B-G-1 will be performed to detect cracking in addition to the VT.:3 examination. A representative sample of bolts will be inspected during the inspection interval prior to the start of the SPEO and iin each 10-year period during the SPEO, a representative sample of bolts 'Nill be inspected. The sample will be 20% of the population (for a material/environment combination) up to a maximum of 25 bolts.

SLRA Section B.2.3.31 (page B-227, as amended by SLRA Aging Management Supplement 1) is revised as shown in Attachment 21 to this letter.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 068-1 L-2021-144 Attachment 23 Page 1 of 9 RAI 3.5-1, 068-1

Background:

SLRA Table 3.5-1, AMR item 3.5-1, 068 (cracking due to stress corrosion cracking (SCC) of high strength steel bolting) in the "Discussion" column states, in part: "....

Cracking of any high-strength [bolting] used in non-ASME component supports is managed by the Structures Monitoring (B.2.3.34) AMP."

SLRA Table 3.5.2-13 "Component Supports Commodity Group - Summary of Aging Management Evaluation," includes on SLRA page 3.5-138 one GALL-SLR Report line item lll.B1.1.TP-41 that corresponds to Table 3.5-1, item 3.5-1, 068 with generic Note E.

This SLRA AMR item credits the Structures Monitoring Program (B.2.3.34) (in lieu of the Xl.S3 "ASME Section XI, Subsection IWF" AMP credited in the corresponding GALL-SLR and SRP-SLR Table 3.5-1 items) to manage the effects of aging for cracking of high-strength steel structural bolting for non-ASME structural supports.

Generic Note E (Refer SLRA page 3.5-138) states: "Consistent with material, environment, and aging effect but a different aging management program is credited or NUREG-2191 credits a plant-specific aging management program."

The GALL-SLR Report line item lll.B1.1.TP-41 associated with SRP-SLR item 3.5-1, 068 recommends the Xl.S3 "ASME Section XI, Subsection IWF" AMP. The SLRA B.2.3.31 "ASME Section XI, Subsection IWF" AMP performs periodic volumetric examinations on a representative sample of high-strength bolting to detect cracking due to sec, as recommended in Xl.S3 AMP.

Issue:

The SLRA B.2.3.34 "Structures Monitoring" AMP included an enhancement (SLRA Table 16-3, Commitments 38(a) and 38(d)) to include guidance and acceptance criteria for inspections of stainless steel and aluminum components for pitting and crevice corrosion, and evidence of cracking due to SCC. However, there is no enhancement provided in the SLRA Structures Monitoring AMP that relates to detecting cracking due to SCC for high-strength steel bolts. It is not clear how the general visual examinations of the SLRA Structures Monitoring AMP is adequate to detect cracking due to SCC in high-strength bolting. Further, for high-strength steel bolting, it is also not clear which specific non-ASME components are proposed to be managed by the referenced AMR line item.

The staff does not have sufficient information to make its finding regarding the adequacy of the inspection method of SLRA Section B.2.3.34 "Structures Monitoring" to detect cracking due to sec in high-strength steel bolting.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 068-1 L-2021-144 Attachment 23 Page 2 of 9 Request:

1) Clarify what specific bolting/system will be examined under the scope of the general non-ASME "high strength components supports" group, including safety significance of their function, for which the specific GALL-SLR line item with a generic note E in SLRA Table 3.5.2-13 that corresponds to SLRA Table 3.5-1, item 3.5-1, 068 will be used.

2) Discuss the examination method and frequency that is proposed to be used in the SLRA Section B.2.3.34 including its adequacy to detect cracking due to SCC in non-ASME high-strength steel bolting considering the significance of the function of the specific components examined.

3) Update the applicable SLRA Sections for conforming changes, as necessary.

NEPB Response:

1) SLRA Table 3.5.2-13 includes a line item for cracking of structural bolting made of high-strength steel, managed by the PBN Structures Monitoring AMP, aligned with Table 1 Item 3.5-1, 068. The PBN Structures Monitoring AMP is credited for managing structural bolting, including any high-strength structural bolting, that is within the scope of subsequent license renewal, but is not covered by other structural AMPs; i.e., "ASME Section XI, Subsection IWE" (GALL-SLR Report AMP Xl.S1); "ASME Section XI, Subsection IWL" (GALL-SLR Report AMP Xl.S2); "ASME Section XI, Subsection IWF" (GALL-SLR Report AMP Xl.S3);

"Masonry Walls" (GALL-SLR Report AMP Xl.S5); and NRC RG 1.127, "Inspection of Water-Control Structures Associated with Nuclear Power Plants" (GALL-SLR Report AMP Xl.S?).

2) The PBN Structures Monitoring AMP will be enhanced prior to the SPEO to specify that, for non-ASME high-strength bolting in scope for SLR and greater than one inch nominal diameter, volumetric examination capable of detecting cracking will be performed in addition to the VT-3 examination. Within 10 years prior to entering the SPEO, and in each 10-year period during the SPEO, a representative sample of bolts will be inspected. The sample will be 20% of the population (for a material I environment combination) up to a maximum of 25 bolts.

3) Revisions to the SLRA are described below to address this new enhancement.

4) Additionally, Commitment 38(c) and the enhancement to Element 2 (Preventive Actions) of the Structures Monitoring AMP are updated for consistency to include a similar enhancement as the PBN ASME Section XI, Subsection IWF AMP regarding lubricants containing sulfur.

References:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 068-1 L-2021-144 Attachment 23 Page 3 of 9 Associated SLRA Revisions:

SLRA Table 16-3 and Section B.2.3.34, as amended by SLRA Aging Management Supplement 1, are revised.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 068-1 L-2021-144 Attachment 23 Page 4 of 9 SLRA Table 16-3 Item No. 38 (pages A-105 and A-106), as amended by SLRA Aging Management Supplement 1, is revised as follows:

Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging NUREG-2191 Commitment Implementation Management Section Schedule Program or Activity (Section) 38 Structures Xl.S6 Continue the existing PBN Structures Monitoring AMP, including No later than 6 Monitoring enhancement to:

months prior to the (16.2.2.34) a)

Revise inspection procedures to include guidance and acceptance SPEO, i.e.:

criteria on inspections of stainless steel and aluminum components PBN1 : 04/05/30 for pitting and crevice corrosion, and evidence of cracking due to PBN2: 09/08/32 SCC. Perform an evaluation if stainless steel or aluminum surfaces exhibit evidence of sec, pitting, or crevice corrosion.

b)

• c) Revise implementing procedures to include preventive actions to ensure bolting integrity for replacement and maintenance activities by specifying proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high strength bolting. Also, ensure proper selection and storage of high strength bolting in accordance with Section 2 of the Research Council for Structural Connections publication, "Specification for Structural Joints Using High-Strength Bolts".

Additionalllf'. 1 mollf'.bdenum disulfide and other lubricants containing sulfur will not be used.

d) Revise inspection procedures to additionally inspect for the following items:

Increase in porosity and permeability, loss of strength, and reduction in concrete anchor capacity due to local concrete degradation in concrete structures.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 068-1 L-2021-144 Attachment 23 Page 5 of 9 Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging NUREG-2191 Commitment Management Section Program or Activity (Section)

Pitting and crevice corrosion, and evidence of cracking due to sec for stainless steel and aluminum components.

Confirmation of the absence of water in-leakage through concrete.

e)

Revise inspection procedures to include guidance on MEB inspection for loss of material (external bus duct enclosure surfaces and structural supports) and elastomer degradation (exterior housing gaskets, boots, and sealants).

f)

Clarify that if ground water leakage is identified then engineering evaluation, more frequent inspections, or destructive testing of affected concrete (to validate properties and determine pH) are required. When leakage volumes allow, assessments may include analysis of the leakage pH, along with mineral, chloride, sulfate, and iron content in the water.

g) Revise inspection procedure to specify that the responsible engineer (RE) shall be a registered professional engineer with knowledge in the design, evaluation, and in-service inspection of concrete structures and performance requirements of nuclear safety-related structures; or a degreed civil or structural engineer with at least ten years' experience in the design, construction, and inspection of concrete structures, with knowledge of the performance requirements of nuclear safety-related structures and potential degradation processes.

h) Revise inspection procedure to specify that accessible areas subject to similar conditions (material, environment, etc.) may be inspected in lieu of inaccessible areas, and include guidance for evaluating the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate Implementation Schedule

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 068-1 L-2021-144 Attachment 23 Page 6 of 9 Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging NUREG-2191 Commitment Management Section Program or Activity (Section) the presence of, or result in, degradation to the inaccessible areas.

i)

UQdate the governing Qrocedure to SQecifv that1 for non-ASME high-strength bolting in scoQe for SLR and greater thah one inch nominal diameter1 volumetric examination caQable of detecting cracking will be Qerformed in addition to the VT-3 examination. Within 10 y:ears Qrior to entering the SPE0 1 and in each 10-Jlear Qeriod during the SPE0 1 a reQresentative samQle of bolts will be insQected. The samQle will be 20% of the QOQulation {for a material I environment combination} UQ to a maximum of 25 bolts.

ilit Ensure quantitative baselines have been established for all structures within the scope of LR prior to entering the SPEO.

.!slli Revise inspection procedure to include the following acceptance criteria:

For Bolting and Fasteners: Loose bolts and nuts are not acceptable unless accepted by engineering evaluation.

For Structural Sealants: Observed loss of material, cracking, and hardening will not result in loss of sealing.

llkt Localized distortion of the reactor cavity liner due to radiation induced volumetric expansion of the underlyin~ concrete.

• SLRA Revision associated with RAI 3.3.1, 263-1.

Implementation Schedule

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 068-1 L-2021-144 Attachment 23 Page 7 of 9 SLRA Section B.2.3.34 table, pages B-239 and B-240, as amended by SLRA Aging Manaqement Supplement 1, are revised as follows:

Element Affected

1. Scope

2. Preventive Actions

3. Parameters Monitored or Inspected

4. Detection of Aging Effects Enhancement Update the governing AMP procedure and other applicable procedures to add stainless steel and aluminum as a material that is inspected for pitting and crevice corrosion, and evidence of cracking due to sec.

Update the governing AMP procedure and other applicable procedures to include preventive actions to ensure bolting integrity for replacement and maintenance activities by specifying proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high strength bolting. Also, ensure proper selection and storage of high strength bolting in accordance with Section 2 of the Research Council for Structural Connections publication, "Specification for Structural Joints Using High Strength Bolts". Additionally, molybdenum disulfide and other lubricants containina sulfur will not be used.

Update the governing AMP procedure and other applicable procedures to additionally inspect the following elements:

Concrete Structures will be inspected for increase in porosity and permeability, loss of strength, and reduction in concrete anchor capacity due to local concrete degradation.

Pitting and crevice corrosion and evidence of cracking due to sec for stainless steel and aluminum components Concrete will be monitored to confirm the absence of water in-leakage Localized distortion of the reactor cavity liner due to radiation induced volumetric expansion of the underlying concrete.

Update the governing AMP procedure and other applicable procedures to include guidance on inspections for pitting and crevice corrosion, and evidence of cracking due to SCC for stainless steel and aluminum components.

Update the governing AMP procedure and other applicable procedures to include guidance on MEB inspection for loss of material (external bus duct enclosure surfaces and structural supports) and elastomer degradation (exterior housing gaskets, boots, and sealants).

• SLRA Revision associated with RAI 3.3.1, 263-1.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 068-1 L-2021-144 Attachment 23 Page 8 of 9 SLRA Section B.2.3.34, pages B-239 and B-240, as amended by SLRA Aging Manaqement Supplement 1, rev1s1on continued:

Element Affected Enhancement Update the governing AMP procedure and other applicable procedures to clarify that if ground water leakage is identified then engineering evaluation, more frequent inspections, or destructive testing of affected concrete (to validate properties and determine pH) are required. When leakage volumes allow, assessments may include analysis of the leakage pH, along with mineral, chloride, sulfate and iron content in the water.

Update the governing procedure to specify that the responsible engineer (RE) shall be a registered professional engineer with knowledge in the design, evaluation, and in-service inspection of concrete structures and performance requirements of nuclear safety-related structures; or a degreed civil or structural engineer with at least ten years' experience in the design, construction, and inspection of concrete structures, with knowledge of the performance requirements of nuclear safety-related structures and potential degradation processes.

Revise inspection procedure to specify that accessible areas subject to similar conditions (material, environment, etc.) may be inspected in lieu of inaccessible areas, and include guidance for evaluating the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to the inaccessible areas.

Update the governing procedure to specify that, for non-ASME high-strength bolting in scope for SLR and greater than one inch nominal diameter, volumetric examination capable of detecting cracking will be performed in addition to the VT-3 examination. In each 10-year period during the SPEO. a representative sample of bolts will be inspected. The sample will be 20% of the population (for a material I environment combination) uo to a maximum of 25 bolts.

5. Monitoring and Trending Ensure quantitative baselines have been established for all structures within the scope of LR prior to entering the SPEO.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 068-1 L-2021-144 Attachment 23 Page 9 of 9 Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 085-1 L-2021-144 Attachment 24 Page 1 of 6 RAI 3.5-1, 085-1

Background:

SLRA Table 3.5-1, AMR item 3.5.1, 085 (loss of material due to pitting, crevice corrosion for stainless steel bolting exposed to treated water), as amended by Supplement 1 dated April 21, 2021 (ADAMS Accession No. ML21111A155, Attachment 22 page 22 of 48) in the "Discussion" column states:

Consistent with NUREG-2191 as clarified and with exception for Water Chemistry. The Water Chemistry (B.2.3.2) AMP and One-Time Inspection (B.2.3.304- [B.2.3.20]) AMP are credited with managing loss of material for stainless steel bolting exposed to treated borated water in the spent fuel pool.

The ASME Section XI, Subsection IWF (B.2.3.31) AMP addresses bolting specific considerations regarding lubricants and storage.

Further, SLRA Table 3.5-1, item 3.5-1, 090 in "Discussion" column states: "Not Used.

Stainless steel bolting exposed to treated borated water in the spent fuel pool is addressed in item 3.5-1, 085."

Generic Note A (Refer SLRA page 3.5-125) states: "Consistent with component, material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP is consistent with NUREG-2191 AMP description."

The GALL-SLR Report credits AMR items lll.B1.1.TP-232, 111.81.2.TP-232, or 111.81.3.TP-232 for loss of material due to pitting and crevice corrosion of stainless steel structural bolts exposed to treated water as summarized in SRP-SLR Table 3.5-1, item 085.

Issue:

SLRA Table 3.5.2-9 "Spent Fuel Pool Structure - Summary of Aging Management Evaluation" on page 3.5-124 includes an AMR line item corresponding to item 3.5.1, 085 with generic Note A. The SLRA AMR line item assigns NUREG-2191 item lll.A6.TP-221 and Water Chemistry (SLRA Section 8.2.3.2) and One-Time Inspection (SLRA Section 8.2.3.20) AMPs to manage the effects of aging for loss of material due to general, pitting, and crevice corrosion for structural bolting in the spent fuel pool.

This designation is not in agreement with the material, environment and AMPs of GALL-SLR Report for disposition of SLRA item 3.5-1, 085 that credits GALL-SLR AMR items 111.81.1.TP-232, 111.81.2.TP-232, or 111.81.3.TP-232 and Water Chemistry and ASME Section XI, Subsection IWF AMPs for consistency and conformance to the assigned generic Note A. Furthermore, the GALL-SLR Xl.S3, "ASME Section XI, Subsection IWF," AMP addresses ISi of Class 1, 2, 3 for MC piping and components and their supports.

It is not clear whether the designation of generic Note A indicating complete consistency of for this SLRA AMR line item with the GALL-SLR Report guidance is appropriate. It is also not clear how the One-Time Inspection (B.2.3.20) program currently credited (in lieu of the ASME Section XI, Subsection IWF (B.2.3.31) AMP) will be used to manage

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 085-1 L-2021-144 Attachment 24 Page 2 of 6 loss of material due to pitting and crevice corrosion in stainless steel bolting in treated/borated water environment of the spent fuel pool.

The staff needs additional information to make its finding that the aging management of components for which item SLRA 3.5-1, 085 is credited is consistent with the corresponding GALL-SLR Report line item as claimed in SLRA Table 3.5-1 and SLRA Table 3.5.2-9, and its adequacy.

Request:

1) Identify the component(s)/component supports for which stainless steel structural bolting are proposed to be managed by the referenced AMR item 3.5-1, 085 in SLRA Table 3.5-1 and Table 3.5.2-9.

2) Provide revised SLRA Table 3.5-1 and Table 3.5.2 AMR line items for item 3.5.1, 085 (or any other applicable SRP-SLR Table 3.5-1 line item) justifying GALL-SLR Report consistency for an assigned generic Note A and a plant-specific note if applicable.

3) Alternatively, explain how the combination of One-Time Inspection (SLRA Section B.2.3.20) and Water Chemistry (SLRA Section B.2.3.2) AMPs will be used to adequately manage stainless steel structural bolting exposed to treated water for pitting and crevice corrosion aging effects (i.e., material, environment and aging effect combination). Accordingly, provide revised Table 3.5-1 and applicable Table 3.5.2s AMR line items with justifiable generic and plant-specific notes.

NEPB Response:

1) The line item in SLRA Table 3.5.2-9 that aligns with Table 1 Item 3.5-1, 085 is for stainless steel structural bolting exposed to treated borated water in the spent fuel pool.

2) SLRA Table 3.5-1 Item Number 3.5-1, 085 will be revised as described below to reflect that the Water Chemistry (B.2.3.2) and ASME Section XI, Subsection IWF AMPs will manage loss of material for stainless steel structural bolting exposed to treated borated water in the spent fuel pool.

SLRA Table 3.5.2-9 will be revised as described below to credit NUREG-2191 Item 111.81.2.TP-232 instead of lll.A6.TP-221, which was previously credited.

SLRA Table 3.5.2-9 will also be revised as described below to credit the PBN ASME Section XI, Subsection IWF AMP in addition to the Water Chemistry AMP and in place of the One-Time Inspection AMP for managing loss of material for the structural bolting located in the spent fuel pool.

The generic notes in SLRA Table 3.5.2-9, as related to the Water Chemistry AMP, are also updated below to reflect that the PBN Water Chemistry AMP contains exceptions.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 085-1 L-2021-144 Attachment 24 Page 3 of 6

References:

None.

Associated SLRA Revisions:

SLRA Table 3.5-1, Item Number 3.5-1, 085, as amended by SLRA Aging Management Supplement 1, and SLRA Table 3.5.2-9 are revised.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 085-1 L-2021-144 Attachment 24 Page 4 of 6 SLRA Table 3.5-1, Item Number 3.5-1, 085 (page 3.5-72), as amended by SLRA Aging Management Supplement 1, is revised as follows:

Table 3.5-1: Containment Building Structure and Internal Structural Components - Summary of Aging Management Programs Item Component Aging Aging Management Further Evaluation Discussion Number Effect/Mechanism Proc:iram (AMP)/TLAA Recommended 3.5-1,

Structural Loss of material due AMP Xl.M2, "Water No Consistent with NUREG-2191 as-085 bolting to pitting, crevice Chemistry," and AMP clarified and with exception for the corrosion Xl.S3, "ASME Section XI, Water Chemistry (B.2.3.2) and Subsection IWF" ASME Section XI, Subsection IWF (B.2.3.31) AMPs.

The Water Chemistry (B.2.3.2) AMP and One Time Inspection (B.2.a.ao)

ASME Section XI, Subsection IWF (B.2.3.31) AMP are credited with managing loss of material for stainless steel structural bolting exposed to t-!=treated borated water in the spent fuel pool.

Tl=le ASMe Section XI, Sl,lbsection IWF {B.2.3.3~} AMP addresses boltin§ specific considerations re§ardin§ ll,lbricants and stoi=a§e.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 085-1 L-2021-144 Attachment 24 Page 5 of 6 SLRA Table 3.5.2-9 (page 3.5-124) is revised as follows:

Table 3.5.2-9: Spent Fuel Pool Structure - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management Type Function Requiring Program Management Liner Pressure Stainless steel Treated borated Cracking Water Chemistry boundary water Loss of material (B.2.3.2) and monitoring of the spent fuel pool water level and leakage from the leak chase channels Miscellaneou Structural Steel Air with borated Loss of material Boric Acid Corrosion s structural support water leakage (B.2.3.4) components Seals Pressure Elastomer Treated borated Loss of sealing Structures Monitoring boundary water (B.2.3.34)

Spent fuel Structural Stainless steel Treated borated Loss of material Water Chemistry pool upender support water (B.2.3.2)

One-Time Inspection (B.2.3.20)

Spent fuel Structural Stainless steel Treated borated Loss of material Water Chemistry storage racks support water (B.2.3.2)

One-Time Inspection (B.2.3.20)

Structural Structural Steel Air-indoor Loss of preload Structures Monitoring bolting support uncontrolled (B.2.3.34)

Structural Structural Steel Air-indoor Loss of material Structures Monitoring bolting support uncontrolled (B.2.3.34)

Structural Structural Stainless steel Treated borated Loss of material Water Chemistry bolting support water (B.2.3.2)

ASME Section XI, Subsection IWF

{B.2.3.31 }GRe +ime 1~~- -

-*=-~ ro..., '> "ll"I \\

NU REG-Table Notes 2191 Item 1 Item lll.A5.T-14 3.5-1,

~A 078 111.85.T-25 3.5-1,

A 089 111.A6.TP-3.5-1,

A 7

072 Vll.A2.A-3.3-1,

Q C 99 125 Vll.A2.A-3.3-1,

~ A 99 125 lll.A6.TP-3.5-1,

A 261 088 lll.A6.TP-3.5-1,

A 248 080 111.81.2.T 3.5-1,

~

P-085 232~

+~ 22~

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5-1, 085-1 L-2021-144 Attachment 24 Page 6 of 6 Table 3.5.2-9: Soent Fuel Pool Structure - Summarv of Aaina Manaaement Evaluation Component Intended Material Environment Aging Effect Aging Management Type Function Requiring Program Manaaement NU REG-Table Notes 2191 Item 1 Item Transfer Pressure Stainless steel Treated borated Loss of material Water Chemistry Vll.A2.A-3.3-1,

Q C canal gates boundary water (B.2.3.2) 99 125 One-Time Inspection (B.2.3.20)

Generic Notes A. Consistent with component, material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP is consistent with NUREG-2191 AMP description.

B. Consistent with component. material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP has exceptions to NUREG-2191 AMP description.

C. Component is different, but consistent with material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP is consistent with NUREG-2191 AMP description.

D. Component is different. but consistent with material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP has exceptions to NUREG-2191 AMP description.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 1 of 18 11. SLRA Section B.2.3.34, "Structures Monitoring" Regulatory Basis:

Paragraph 54.21 (a)(3) of 10 CFR requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis for the subsequent period of extended operation. As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL-SLR Report when evaluation of the matter in the GALL-SLR Report applies to the plant.

RAI 8.2.3.34-1

Background:

AMR items in SLRA Table 3.5.2-13, "Component Supports Commodity Group," as revised by the supplemental letter dated April 21, 2021, states that building concrete at locations of expansion and grouted anchors, and grout pads for support base plates will be managed for the aging effects of reduction in concrete anchor capacity by the Structures Monitoring program. The AMR item, associated with Table 1 item 3.5-1, 055, cites generic note A to indicate that this AMR item is consistent with the GALL-SLR Report. The AMR item also cites plant-specific Note 2 to indicate this type of component includes epoxy grout/anchors as subject to the same aging effects.

(emphasis added)

The GALL-SLR Report does not generically address epoxy grouted anchors as a component subject to an AMR or provide a comprehensive list of all potential aging effects that may be applicable to the epoxy grouted anchors. As described in the SRP-SLR, the applicant may use a plant-specific AMP or plant-specific aging management activities (within an existing AMP) as the basis for aging management of a specific structure or component. For those components, materials and aging effects combinations that are not generically addressed by the GALL-SLR Report, the NRC staff reviews the proposed AMPs or activities in accordance with the program element criteria that are defined in the SRP-SLR Appendix A.1, Subsection A. 1.2.3, to ensure that the effects of aging for those structures or components will be adequately managed during the period of extended operation.

NRC Information Notice No. (IN) 83-40, discusses industry operating experiences regarding the use of epoxy grouts for anchor bolts installations, the potential degradations of epoxy formulations due to heat and radiation, and potential degradations due to the relatively low creep strength of epoxies. Furthermore, the industry and manufacturers have recognized the importance of proper installation (e.g.

proper mixing, hole cleaning, application, etc.) of these types of anchors to ensure the desired performance. NRC IN 2010-01 discusses industry operating experiences regarding improper installation of post-installed anchors in pipe supports.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 2 of 18 Issue:

Based on the staff review of the information provided in the SLRA and the audited documents, additional information is necessary to demonstrate that the aging effects for epoxy grouted anchors/bolts will be adequately managed by the Structures Monitoring Program. Specifically, the additional justification is needed for the following issues:

Plant-specific Note 2 states that epoxy grouted anchors are subject to the same aging effects as typical grouted anchors, however no technical justification has been provided to support this claim.

The proposed aging effects under AMR item 3.5-1, 055 is limited to the aging effects of reduction in concrete anchor capacity due to concrete degradations, however the SLRA did not consider or address other potential aging effects identified in relevant industry operating experience for epoxy grout anchors (see NRC IN 83-40).

During the audit, the staff reviewed plant procedure NP 7.7.9, "Facilities Monitoring Program," Revision 21, and noted that epoxy grouted anchors/bolts components are not specifically discussed or addressed within the procedure to ensure that they are adequately managed for the subsequent period of extended operation.

The scope within the Structures Monitoring program for this type of anchors is not clear. Additional information is needed regarding the type of systems or components (e.g., safety-related systems, ASME Class 1, 2, 3 components, non-safety related components) that the epoxy grouted anchors are associated with and their specific environments (e.g., areas with potential for high temperatures or radiation), if there are any limitations in place to prevent the use of epoxy grouted anchors for certain applications or locations, and the type of epoxy grout used at PBN and its material qualification based on its intended function(s) and application (e.g., safety-related, non safety-related).

It is not clear whether installation procedures followed ensure that manufactures recommendations for proper installation and personnel qualification are met to prevent any premature deterioration/aging effects (see e.g., IN 2010-01 for installation related issues).

It is not clear whether the epoxy grouted anchors have been qualified to sustain design basis loads to the end of the subsequent period of extended operation.

Request:

Provide clarification or additional justification to the issues identified above to demonstrate that the aging effects for epoxy grouted anchors/bolts will be adequately managed by the Structures Monitoring Program so that the relevant CLB and their intended function(s) will be maintained to the end of the subsequent period of extended operation.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 3 of 18 NEPB Response:

The conclusion stated by Plant Specific Note 2 in SLRA Table 3.5.2-13 is based on a review of publicly available records for information related to potential aging mechanisms associated with epoxy (adhesive) anchors and epoxy resin-based grout that was performed as part of the aging management review (AMR) for Plant Structures at Point Beach Nuclear Plant (PBN). The review determined the following:

The primary function of a Component Support incorporating an epoxy (adhesive) anchor or epoxy resin-based grout is to provide anchorage so that the supported element can perform its intended function. When within the scope of license renewal, such a support is a passive, long-lived structure/component that requires AMR.

The Component Support tabulations in Chapter 111 of Reference 1 contain no line item listing "polymer" or "polymeric" (the generalized term used in the AMR tables that encompasses "epoxy" or "epoxy resins") as a material of construction.

Similarly, there is no line item listing "polymer or "polymeric" as an environment to which bolting material (providing support anchorage to building structures) would be exposed.

Aside from information on the use of adhesive anchoring systems in structural support applications documented in Reference 2, NEPB has identified no mention of "adhesive" or "epoxy" anchor systems in any recent license renewal application correspondence with the NRC.

Epoxy (adhesive) anchor and epoxy resin-based grout vendors provide information demonstrating chemical resistance for adhesives but no indication that the base resins, curatives, and fillers comprising their products degrade bolting materials used in the anchoring applications.

o One vendor provides a summary of testing to demonstrate resistance of the adhesives to degradation following immersion in a variety of chemicals. The tabulation notes that in most actual service conditions, the majority of the anchoring adhesive is not exposed to the chemical and thus some period of time is required for the chemical to saturate the entire adhesive. An adhesive anchor would be expected to maintain bond strength and creep resistance until a significant portion of the adhesive is saturated.

o In response to an in,quiry from NextEra Energy procurement engineering regarding aging concerns or testing for a particular material, another vendor stated that there is "no specific test for the accelerated aging of construction adhesives that are used for post installed anchors." However, this vendor also indicated that "millions" of their anchors are installed yearly "without a single negative feedback regarding age-related degradation" and that "there have been no reported problems regarding

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 4of18 the performance of the product due to aging" during approximately 35 years of use in the construction industry.

Available technical literature identifies reduction in load capacity (in both wet and dry environments) as an aging effect for epoxy adhesive materials.

At PBN, epoxy (adhesive) anchors and epoxy resin-based grout have been evaluated or included in support plate modifications for the Service Water and Component Cooling

• Water pumps (safety related) and in locations outside the scope of license renewal (e.g., mounting anti-vortex cages for the Circulating Water pumps). The following attributes were among those considered as part of the standard design process for these installations:

Similar to other "polymer" or "polymeric" materials used in various mechanical systems, protection for epoxy (adhesive) anchors and epoxy resin-based grout depends on chemical resistance to the environment to which the materials are exposed. The epoxy resin-based grout installations at PBN are not submerged-but rather are exposed only to air (indoor - uncontrolled; outdoor).

Acceptability for the use of such materials in each location was a design-driven criterion that considered relevant industry operating experience. None is located in an environment associated with the potential aging effects identified by NRC IN 83-40 (i.e., radiation fields inside containment; elevated temperatures beyond those ordinarily caused by changes in weather; appreciable preload tension).

Design documentation specifies conformance to the plant maintenance instructions for grouting and the need to follow the manufacturer's printed installation instructions for anchor systems-paying particular attention to drilling and cleaning the holes, confirming that the adhesive is thoroughly mixed, avoiding air voids when injecting the adhesive, properly inserting the anchor, and ensuring no disturbance until the cure time has elapsed. Manufacturer's requirements are provided at the vendor websites along with product technical guides (providing design calculation templates and tabulations for proper sizing of anchoring systems) and evaluation reports (issued by the ICC Evaluation Service-a subsidiary of the International Code Council-and documenting code compliance for anchoring systems).

With the implementation of standard engineering practices and in the absence of meaningful evidence to the contrary, the AMR concluded that installations employing epoxy (adhesive) anchors and epoxy resin-based grout at PBN are subject to the same aging effects for steel and grout identified by Chapter Ill in Reference 1 and are managed by the PBN Structures Monitoring AMP, consistent with Section Xl.S6 in Reference 3.

The implementing procedure for the PBN Structures Monitoring AMP currently does not specifically discuss or address epoxy (adhesive) anchors and epoxy resin-based grout.

To provide reasonable assurance that components comprising such installations are adequately managed for the subsequent period of extended operation, the PBN

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 5of18 Structures Monitoring AMP is updated to include a new enhancement specifying their inspection for degradation that could cause a loss of anchor capacity.

New rows are added for the applicable component types in SLRA Table 3.5.2-13 (i.e.,

anchorage I embedment; building concrete at locations of expansion and grout anchors, grout pads for support base plates) citing Generic Note F ("Material not in NUREG-2191 for this component.") and Plant Specific Note 2. The text of Plant Specific Note 2 is revised for clarification.

References:

1. NUREG-2191, Volume 1, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, July 2017 (ADAMS Accession No. ML17187A031 ).

2. FPL Letter L-2019-103 to NRC dated May 9, 2019, Turkey Point, Units 3 and 4 -

Use of Adhesive Anchoring Systems in Structural Supports (ADAMS Accession No. ML19133A061 ).

3. NUREG-2191, Volume 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, July 2017 (ADAMS Accession No. ML17187A204).

4. NextEra Energy Point Beach, LLC (NEPB) Letter to NRC L-2021-081 dated April 21, 2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML21111A155).

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 6of18 Associated SLRA Revisions:

SLRA Table 3.5.2-13 (pages 3.5-136 to 3.5-138, as amended by SLRA Aging Management Supplement 1) is revised as follows:

Table 3.5.2-13: Component Supports Commodity Group - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-2191 Table 1 Notes Type Function Requiring Program Item Item Management Anchorage I Structural Steel Air-indoor Loss of preload Structures lll.A3.TP-261 3.5-1, 088 A,:;,

embedment support uncontrolled Monitoring (B.2.3.34)

Anchorage I Structural Steel Air-indoor Loss of preload Structures F,2 embedment support uncontrolled Monitoring (B.2.3.34}

Anchorage I Structural Stainless Air-indoor Loss of material Structures lll.B3.T-37b 3.5-1, 100 A

embedment Support steel uncontrolled Monitoring (B.2.3.34)

Anchorage I Structural Steel Air - outdoor Loss of material Structures 111.83.TP-248 3.5-1, 080 A

embedment support Monitoring (B.2.3.34)

Anchorage I Structural Steel Air with Loss of material Boric Acid Corrosion 111.81.1.T-25 3.5-1, 089 A

embedment support borated water (B.2.3.4) leakage ASME Class Structural High-strength Air-indoor Cracking ASME Section XI, 111.81.1.TP-41 3.5-1, 068 B, 1 2 and 3 support steel uncontrolled Subsection IWF structural (B.2.3.31 )

bolting

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 7of18 Table 3.5.2-13: Component Supports Commodity Group-Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-2191 Type Function Requiring Program Item Management ASME Class Structural Steel Air-indoor Loss of preload ASME Section XI, lll.B1.2.TP-229 2 and 3 support uncontrolled Subsection IWF structural (B.2.3.31 )

bolting ASME Class Pipe whip Steel Air with Loss of material Boric Acid Corrosion lll.B1.1.T-25 2 and 3 restraint borated water (8.2.3.4) supports Structural leakage support ASME Class Pipe whip Steel Air-indoor Loss of material ASME Section XI, lll.B1.1.T-24 2 and 3 restraint uncontrolled Subsection IWF supports Structural (B.2.3.31 )

support ASME Class Structural Steel Air-indoor Loss of ASME Section XI, 111.81.2.T-28 2 and 3 support uncontrolled mechanical Subsection IWF supports function (B.2.3.31 )

(hangers,

guides, stops)

Building Structural Concrete Air-indoor Reduction in Structures lll.B2.TP-42 concrete at support (reinforced) uncontrolled concrete anchor Monitoring locations of capacity (B.2.3.34) expansion and grouted anchors; grout pads for support base plates Table 1 Notes Item 3.5-1, 087 8, 1 3.5-1, 089 B, 1 3.5-1, 091 8, 1, 3 3.5-1, 057 A,4 3.5-1, 055 A

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 8of18 Table 3.5.2-13: Component Supports Commodity Group-Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-2191 Type Function Requiring Program Item Management Building Structural Concrete Air - outdoor Reduction in Structures lll.B2.TP-42 concrete at support (reinforced) concrete anchor Monitoring locations of capacity (B.2.3.34) expansion and grouted anchors; grout pads for support base plates Building Structural Grout Air - outdoor Reduction in Structures lll.B2.TP-42 concrete at support concrete anchor Monitoring locations of capacity (B.2.3.34) expansion and grouted anchors; grout pads for support base plates Building Structural Grout Air-indoor Reduction in Structures lll.B2.TP-42 concrete at support uncontrolled concrete anchor Monitoring locations of capacity (B.2.3.34) expansion and grouted anchors; grout pads for support base plates Table 1 Notes Item 3.5-1, 055 A

3.5-1, 055 A,-2 3.5-1, 055 A,-2

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 9of18 Table 3.5.2-13: Component Supports Commodity Group-Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-2191 Type Function Requiring Program Item Management Building Structural Grout Air-indoor Reduction in Structures concrete at su1rnort uncontrolled concrete anchor Monitoring locations of capacity (8.2.3.34) expansion and grouted anchors; grout pads for support base plates Component Structural Stainless Air-indoor Cracking Structures lll.82.T-37b supports support steel uncontrolled Loss of material Monitoring Air - outdoor (B.2.3.34)

Component Structural Steel Air-indoor Loss of material Structures lll.B4.T-43 supports support uncontrolled Monitoring Air - outdoor (B.2.3.34)

Component Structural Steel Air with Loss of material Boric Acid Corrosion lll.B4.T-25 supports support borated water (B.2.3.4) leakage Electrical

Shelter, Steel Air with Loss of material Boric Acid Corrosion lll.B3.T-25 Enclosures -

protection borated water (B.2.3.4)

Panels, Structural leakage

boxes, support

cabinets, consoles, raceways Table 1 Notes Item F,2 3.5-1, 100 A

3.5-1, 092 A

3.5-1, 089 A

3.5-1, 089 A

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 10of18 Table 3.5.2-13: Component Supports Commodity Group-Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-2191 Type Function Requiring Program Item Management Electrical

Shelter, Steel Air-indoor Loss of material Structures 111.83.T-43 Enclosures -

protection uncontrolled Monitoring

Panels, Structural (B.2.3.34)

boxes, support

cabinets, consoles, raceways Insulation Insulation Stainless Air - outdoor Cracking External Surfaces lll.B2.T-37c Jacket Steel Loss of material Monitoring of integrity Mechanical Components (B.2.3.23)

Insulation Insulation Aluminum Air - outdoor Cracking External Surfaces lll.B2.T-37c Jacket Loss of material Monitoring of integrity Mechanical Components (B.2.3.23)

Insulation Insulation Stainless Air-indoor Cracking External Surfaces lll.B2.T-37c Jacket Steel uncontrolled Loss of material Monitoring of integrity Mechanical Components (B.2.3.23)

Insulation Insulation Aluminum Air-indoor Cracking External Surfaces lll.B2.T-37c Jacket uncontrolled Loss of material Monitoring of integrity Mechanical Components (B.2.3.23)

Table 1 Notes Item 3.5-1, 092 A

3.5-1, 100 c 3.5-1, 100 c 3.5-1, 100 c 3.5-1, 100 c

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 11 of 18 Table 3.5.2-13: Component Supports Commodity Group - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-2191 Type Function Requiring Program Item Management Pipe Pipe whip Steel Air with Loss of material Boric Acid Corrosion 111.82.T-25 restraints and restraint borated water (B.2.3.4)

HVAC duct Structural leakage supports support Pipe Pipe whip Steel Air-indoor Loss of material Structures 111.82.TP-43 restraints and restraint uncontrolled Monitoring HVAC duct Structural (B.2.3.34) supports support Structural Structural Steel Air-indoor Loss of preload Structures lll.A3.TP-261 bolting support uncontrolled Monitoring (B.2.3.34)

Structural Structural Steel Air-indoor Loss of material Structures 111.83.TP-248 bolting support uncontrolled Monitoring (B.2.3.34)

Structural Structural Steel Air - outdoor Loss of material Structures lll.A3.TP-274 bolting support Monitoring (B.2.3.34)

Structural Structural High-strength Air-indoor Cracking Structures 111.81.1.TP-41 bolting support steel uncontrolled Monitoring (B.2.3.34)

Vibration Structural Non-metallic; Air-indoor Reduction or loss ASME Section XI, 111.81.1.T-33 isolation support Elastomer uncontrolled of isolation Subsection IWF elements function (B.2.3.31 )

Vibration Structural Non-metallic; Air-indoor Reduction or loss Structures 111.84.TP-44 isolation support Elastomer uncontrolled of isolation Monitoring elements function (B.2.3.34)

Table 1 Notes Item 3.5-1, 089 A

3.5-1, 092 A

3.5-1, 088 A

3.5-1, 080 A

3.5-1, 082 A

3.5-1, 068 E

3.5-1, 094 B

3.5-1, 094 A

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 12 of 18 Generic Notes A.

Consistent with component, material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP is consistent with NUREG-2191 AMP description.

B. Consistent with component, material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP has exceptions to NUREG-2191 AMP description C. Component is different, but consistent with material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP is consistent with NUREG-2191 AMP description.

E. Consistent with NUREG-2191 material, environment, and aging effect but a different aging management program is credited or NUREG-2191 identifies a plant-specific aging management program.

F.

Material not in NUREG-2191 for this component.

Plant Specific Notes

1. RCS Class 1 major equipment supports are addressed in Table 3.5.2-1.

2. Component type ilncludes epoxy ~(adhesive) anchors or epoxy resin-based grout, which are not generically addressed by NUREG2191. The use of such materials is a design-driven criterion. AMR concluded that installations employing these materials at PBN are subject to the same aging effects as those identified by other items for anchors and grout.

3.

Galvanized steel ASME Class 2 and 3 supports are considered same as carbon steel supports.

4.

Passive portions of constant and variable spring hangers (e.g., attachment to structure).

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 13of18 SLRA Table 16-3 (Item No. 38)-Appendix A, Section 16.4 (pages A-105 and A-106, as amended by SLRA Aging Management Supplement 1) is revised as follows:

Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging Management NU REG-Commitment Implementation Program or Activity 2191 Schedule (Section)

Section 38 Structures Monitoring Xl.S6 Continue the existing PBN Structures Monitoring AMP, including No later than 6 (16.2.2.34) enhancement to:

months prior to the a) Revise inspection procedures to include guidance and acceptance SPEO, i.e.:

criteria on inspections of stainless steel and aluminum components PBN1 : 04/05/30 for pitting and crevice corrosion, and evidence of cracking due to PBN2: 09/08/32 SCC. Perform an evaluation if stainless steel or aluminum surfaces exhibit evidence of sec, pitting, or crevice corrosion.

b) Revise inspection procedure scope to include polystyrene foam that is mounted to the underside of manhole covers as an elastomer material.

c)

Revise implementing procedures to include preventive actions to ensure bolting integrity for replacement and maintenance activities by specifying proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high strength bolting. Also, ensure proper selection and storage of high strength bolting in accordance with Section 2 of the Research Council for Structural Connections publication, "Specification for Structural Joints Using High-Strength Bolts".

d) Revise inspection procedures to additionally inspect for the following items:

Increase in porosity and permeability, loss of strength, and reduction in concrete anchor capacity due to local concrete degradation in concrete structures.

Loss of material and loss of strength for elastomers.

Pitting and crevice corrosion, and evidence of cracking due to sec for stainless steel and aluminum components.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 14of18 Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging Management NU REG-Commitment Program or Activity 2191 (Section)

Section Confirmation of the absence of water in-leakage through concrete.

e)

Revise inspection procedures to include guidance on MEB inspection for loss of material (external bus duct enclosure surfaces and structural supports) and elastomer degradation (exterior housing gaskets, boots, and sealants).

f)

Clarify that if ground water leakage is identified then engineering evaluation, more frequent inspections, or destructive testing of affected concrete (to validate properties and determine pH) are required. When leakage volumes allow, assessments may include analysis of the leakage pH, along with mineral, chloride, sulfate, and iron content in the water.

g} Update the governing AMP procedure and other applicable procedures to specifv inspection of structural support applications emplo~ing epo~ {adhesive} anchors and epo~

resin-based grout for degradation that could cause a loss of anchor capacity.

h!t) Revise inspection procedure to specify that the responsible engineer (RE) shall be a registered professional engineer with knowledge in the design, evaluation, and in-service inspection of concrete structures and performance requirements of nuclear safety-related structures; or a degreed civil or structural engineer with at least ten years' experience in the design, construction, and inspection of concrete structures, with knowledge of the performance requirements of nuclear safety-related structures and potential degradation processes.

ill) Revise inspection procedure to specify that accessible areas subject to similar conditions (material, environment, etc.) may be inspected in lieu of inaccessible areas, and include guidance for evaluating the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to the inaccessible areas.

Implementation Schedule

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 15of18 Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging Management NU REG-Commitment Program or Activity 2191 (Section)

Section m Ensure quantitative baselines have been established for all structures within the scope of LR prior to entering the SPEO.

.!sf) Revise inspection procedure to include the following acceptance criteria:

For Elastomers: No loss of material and no indications of loss of strength such as unacceptable surface cracking, crazing, scuffing, dimensional change (e.g., "ballooning" and "necking"),

shrinkage, discoloration, or hardening.

For Bolting and Fasteners: Loose bolts and nuts are not acceptable unless accepted by engineering evaluation.

For Structural Sealants: Observed loss of material, cracking, and hardening will not result in loss of sealing.

Implementation Schedule

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 16of18 SLRA Section B.2.3.34 (pages B-239 and B-240, as amended by SLRA Aging Management Supplement 1) is revised as follows:

Element Affected

1. Scope

2. Preventive Actions

3. Parameters Monitored or Inspected

4. Detection of Aging Effects Enhancement Update the governing AMP procedure and other applicable procedures to add stainless steel and aluminum as a material that is inspected for pitting and crevice corrosion, and evidence of cracking due to sec.

Update the governing AMP procedure scope to include polystyrene foam that is mounted to the underside of manhole covers as an elastomer material.

Update the governing AMP procedure and other applicable procedures to include preventive actions to ensure bolting integrity for replacement and maintenance activities by specifying proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high strength bolting. Also, ensure proper selection and storage of high strength bolting in accordance with Section 2 of the Research Council for Structural Connections publication, "Specification for Structural Joints Using High Strength Bolts".

Update the governing AMP procedure and other applicable procedures to additionally inspect the following elements:

Concrete Structures will be inspected for increase in porosity and permeability, loss of strength, and reduction in concrete anchor capacity due to local concrete degradation.

Elastomer will also be inspected for loss of material and loss of strength.

Pitting and crevice corrosion and evidence of cracking due to sec for stainless steel and aluminum components Concrete will be monitored to confirm the absence of water in-leakaqe Update the governing AMP procedure and other applicable procedures to include guidance on inspections for pitting and crevice corrosion, and evidence of cracking due to SCC for stainless steel and aluminum components.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 17of18 SLRA Section B.2.3.34 (pages B-239 and B-240) revision continued :

Element Affected Enhancement Update the governing AMP procedure and other applicable procedures to include guidance on MEB inspection for loss of material (external bus duct enclosure surfaces and structural supports) and elastomer degradation (exterior housing gaskets, boots, and sealants).

Update the governing AMP procedure and other applicable procedures to clarify that if ground water leakage is identified then engineering evaluation, more frequent inspections, or destructive testing of affected concrete (to validate properties and determine pH) are required. When leakage volumes allow, assessments may include analysis of the leakage pH, along with mineral, chloride, sulfate and iron content in the water.

Uudate the governing AMP urocedure and other auulicable urocedures to suecif~ insuection of structural SUQQOrt aQQlications emQIO~ing eQOX~

{adhesive} anchors and euox~ resin-based grout for degradation that could cause a loss of anchor cauacit~.

Update the governing procedure to specify that the responsible engineer (RE) shall be a registered professional engineer with knowledge in the design, evaluation, and in-service inspection of concrete structures and performance requirements of nuclear safety-related structures; or a degreed civil or structural engineer with at least ten years' experience in the design, construction, and inspection of concrete structures, with knowledge of the performance requirements of nuclear safety-related structures and potential degradation processes.

Revise inspection procedure to specify that accessible areas subject to similar conditions (material, environment, etc.) may be inspected in lieu of inaccessible areas, and include guidance for evaluating the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to the inaccessible areas.

. I

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. B.2.3.34-1 L-2021-144 Attachment 25 Page 18 of 18 SLRA Section B.2.3.34 (pages B-239 and B-240) revision continued :

Element Affected Enhancement

5. Monitoring and Trending Ensure quantitative baselines have been established for all structures within the scope of LR prior to entering the SPEO.

6. Acceptance Criteria Update the governing AMP procedure and other applicable procedures to include acceptance criteria on inspection of stainless steel and aluminum components for pitting and crevice corrosion, and evidence of cracking due to sec. In addition, require performance of an evaluation if stainless steel or aluminum surfaces exhibit evidence of sec, pitting, or crevice corrosion.

Update the governing AMP procedure and other applicable procedures to include the following acceptance criteria:

Elastomers: No loss of material and no indications of loss of strength such as unacceptable surface cracking, crazing, scuffing, dimensional change (e.g.,

"ballooning" and "necking"), shrinkage, discoloration, or hardening.

Bolting and Fasteners: Loose bolts and nuts are not acceptable unless accepted by engineering evaluation.

Structural Sealants: Acceptable if the observed loss of material, cracking, and hardening will not result in loss of sealing.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.11 -1 L-2021-144 Attachment 26 Page 1 of 9 RAI 3.5.2.11-1 (earthen berm structures)

Background :

An AMR item in SLRA Table 3.5.2-11, states that loss of form and loss of material for earthen berm structures exposed to air - outdoor environment will be managed by the Structures Monitoring program. The AMR item cites generic note J to indicate that neither the component nor the material and environment combination are evaluated in the GALL-SLR Report.

Issue:

The GALL-SLR Report includes item lll.A6.T-22, for earthen water-control structures, which includes loss of material and loss of form as the applicable aging effects for earthen dams or embankments. During the audit, the staff reviewed the proposed enhancements in SLRA Section B.2.3.34, "Structures Monitoring," and plant procedure NP 7.7.9, "Facilities Monitoring Program," Revision 21, and noted that Section 2 of Attachment D of the procedure addresses the aging effects and acceptance criteria of loss of material for similar structural components (i.e., earthen embarkments).

However, none of the reviewed documents address monitoring for the loss of form aging effect and the associated acceptance criteria for the earthen berm structures.

Therefore, it is not clear how the Structures Monitoring program will adequately manage the loss of form aging effect for earthen berm structures if the associated program elements are not clearly defined within the Structures Monitoring program.

Request:

Clarify how the Structures Monitoring program will adequately manage loss of form and loss of material for earthen berm structures, including those parameters to be monitored or inspected and acceptance criteria, for the subsequent period of extended operation.

NEPB Response:

The berm in question is an earthen berm surrounding the fuel oil storage tanks, and its intended function is to provide a fire barrier in the case of a fuel oil spill and any resulting fire. It does not function as a water-control structure and is not within the scope of the PBN Water-Control Structures AMP, and therefore was not aligned with GALL-SLR Item lll.A6.T-22. The aging effects listed in SLRA Table 3.5.2-11 do, however, include loss of form and loss of material, which are the same aging effects listed in GALL-SLR Item lll.A6.T-22 for earthen water-control structures.

The implementing procedure for the PBN Structures Monitoring AMP currently includes loss of material as an aging effect requiring management for the earthen berm around the fuel oil tanks.

The PBN Structures Monitoring AMP will be enhanced to include loss of form as an aging effect for the earthen berm surrounding the fuel oil storage ta.nks. Acceptance criteria for both loss of material and loss of form will include no evidence of: settlement (unusual localized or overall settlement, depressions, sinkholes); slope instability (variance from originally constructed slopes, unusual changes from original crest

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.11-1 L-2021-144 Attachment 26 Page 2 of 9 alignment and elevation, evidence of movement); and, erosion (gullies or notches in slope).

SLRA Table 16-3 (Appendix A, Section 16.4) and Section B.2.3.34, as revised by SLRA Aging Management Supplement 1, are amended as described below to include enhancements to the Parameters Monitored or Inspected (Element 3) and Acceptance Criteria (Element 6) elements to address the loss of form aging effect for the earthen berm surrounding the fuel oil storage tanks.

References:

None.

Associated SLRA Revisions:

SLRA Table 16-3 and Section B.2.3.34, as amended by SLRA Aging Management Supplement 1, are revised.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.11-1 L-2021-144 Attachment 26 Page 3 of 9 SLRA Table 16-3 (Appendix A, Section 16.4), pages A-105 and A-106, as amended by SLRA Aging Management Supplement 1, are revised as follows:

Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging NUREG-2191 Commitment Implementation Management Section Schedule Program or Activity (Section) 38 Structures Xl.S6 Continue the existing PBN Structures Monitoring AMP, including No later than 6 Monitoring enhancement to:

months prior to the (16.2.2.34) a) Revise inspection procedures to include guidance and acceptance SPEO, i.e.:

criteria on inspections of stainless steel and aluminum components PBN1: 04/05/30 for pitting and crevice corrosion, and evidence of cracking due to PBN2: 09/08/32 SCC. Perform an evaluation if stainless steel or aluminum surfaces exhibit evidence of sec, pitting, or crevice corrosion.

b) Revise inspection procedure scope to include polystyrene foam that is mounted to the underside of manhole covers as an elastomer material.

c)b1Revise implementing procedures to include preventive actions to ensure bolting integrity for replacement and maintenance activities by specifying proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high strength bolting. Also, ensure proper selection and storage of high strength bolting in accordance with Section 2 of the Research Council for Structural Connections publication, "Specification for Structural Joints Using High-Strength Bolts".

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.11-1 L-2021-144 Attachment 26 Page 4 of 9 SLRA Table 16-3 (Appendix A, Section 16.4), pages A-105 and A-106, as amended by SLRA Aging Management Supplement 1, revision continued:

Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging NU REG-Commitment Implementation Management 2191 Schedule Program or Section Activity (Section) d)6fRevise inspection procedures to additionally inspect for the following items:

Increase in porosity and permeability, loss of strength, and reduction in concrete anchor capacity due to local concrete degradation in concrete structures.

Loss of material and loss of strength for elastomers.

Pitting and crevice corrosion, and evidence of cracking due to SCC for stainless steel and aluminum components.

Confirmation of the absence of water in-leakage through concrete.

Loss of form of the earthen berm surrounding the fuel oil storage tanks.

e)G1Revise inspection procedures to include guidance on MEB inspection for loss of material (external bus duct enclosure surfaces and structural supports) and elastomer degradation (exterior housing gaskets, boots, and sealants).

f)et Clarify that if ground water leakage is identified then engineering evaluation, more frequent inspections, or destructive testing of affected concrete (to validate properties and determine pH) are required. When leakage volumes allow, assessments may include analysis of the leakage pH, along with mineral, chloride, sulfate, and iron content in the water.

g) Revise inspection procedure to specify that the responsible engineer (RE) shall be a registered professional engineer with knowledge in the design, evaluation, and in-service inspection of concrete structures and performance requirements of nuclear safety-related structures; or a degreed civil or structural engineer

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.11-1 L-2021 -144 Attachment 26 Page 5 of 9 SLRA Table 16-3 (Appendix A, Section 16.4), pages A-105 and A-106, as amended by SLRA Aging Management Supplement 1, revision continued:

Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging NUREG-2191 Commitment Implementation Management Section Schedule Program or Activity (Section) with at least ten years' experience in the design, construction, and inspection of concrete structures, with knowledge of the performance requirements of nuclear safety-related structures and potential degradation processes.

h) Revise inspection procedure to specify that accessible areas subject to similar conditions (material, environment, etc.) may be inspected in lieu of inaccessible areas, and include guidance for evaluating the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to the inaccessible areas.

i)

Ensure quantitative baselines have been established for all structures within the scope of LR prior to entering the SPEO.

j)ft Revise inspection procedure to include the following acceptance criteria:

For Elastomers: No loss of material and no indications of loss of strength such as unacceptable surface cracking, crazing, scuffing, dimensional change (e.g., "ballooning" and "necking"),

shrinkage, discoloration, or hardening.

For Bolting and Fasteners: Loose bolts and nuts are not acceptable unless accepted by engineering evaluation.

For Structural Sealants: Observed loss of material, cracking, and hardening will not result in loss of sealing.

For earthen berm: No evidence of:

0 Settlement - unusual localized or overall settlement, deoressions. sinkholes

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.11-1 L-2021-144 Attachment 26 Page 6 of 9 Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging NUREG-2191 Commitment Management Section Program or Activity (Section) 0 Slo~e instabilitv - variance from originalll£ constructed slo~es 1 unusual changes from original crest alignment and elevation 1 evidence of movement 0

Erosion - gullies or notches in slo~e Implementation Schedule

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.11-1 L-2021-144 Attachment 26 Page 7 of 9 SLRA Section B.2.3.34, pages B-239 and B-240, as amended by SLRA Aging Management Supplement 1, are revised as follows:

Element Affected

1. Scope

2. Preventive Actions

3. Parameters Monitored or Inspected

4. Detection of Aging Effects Enhancement Update the governing AMP procedure and other applicable procedures to add stainless steel and aluminum as a material that is inspected for pitting and crevice corrosion, and evidence of cracking due to sec.

Update the governing AMP procedure scope to include polystyrene foam that is mounted to the underside of manhole covers as an elastomer material.

Update the governing AMP procedure and other applicable procedures to include preventive actions to ensure bolting integrity for replacement and maintenance activities by specifying proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high strength bolting. Also, ensure proper selection and storage of high strength bolting in accordance with Section 2 of the Research Council for Structural Connections publication, "Specification for Structural Joints Using High Strength Bolts".

Update the governing AMP procedure and other applicable procedures to additionally inspect the following elements:

Concrete Structures will be inspected for increase in porosity and permeability, loss of strength, and reduction in concrete anchor capacity due to local concrete degradation.

Elastomer will also be inspected for loss of material and loss of strength.

Pitting and crevice corrosion and evidence of cracking due to sec for stainless steel and aluminum components Concrete will be monitored to confirm the absence of water in-leakage Earthen berm will be monitored for loss of form Update the governing AMP procedure and other applicable procedures to include guidance on inspections for pitting and crevice corrosion, and evidence of cracking due to sec for stainless steel and aluminum components.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.11-1 L-2021-144 Attachment 26 Page 8 of 9 SLRA Section B.2.3.34, pages B-239 and B-240, as amended by SLRA Aging ManaQement Supplement 1, rev1s1on continued:

Element Affected Enhancement Update the governing AMP procedure and other applicable procedures to include guidance on MEB inspection for loss of material (external bus duct enclosure surfaces and structural supports) and elastomer degradation (exterior housing gaskets, boots, and sealants).

Update the governing AMP procedure and other applicable procedures to clarify that if ground water leakage is identified then engineering evaluation, more frequent inspections, or destructive testing of affected concrete (to validate properties and determine pH) are required. When leakage volumes allow, assessments may include analysis of the leakage pH, along with mineral, chloride, sulfate and iron content in the water.

Update the governing procedure to specify that the responsible engineer (RE) shall be a registered professional engineer with knowledge in the design, evaluation, and in-service inspection of concrete structures and performance requirements of nuclear safety-related structures; or a degreed civil or structural engineer with at least ten years' experience in the design, construction, and inspection of concrete structures, with knowledge of the performance requirements of nuclear safety-related structures and potential degradation processes.

Revise inspection procedure to specify that accessible areas subject to similar conditions (material, environment, etc.) may be inspected in lieu of inaccessible areas, and include guidance for evaluating the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to the inaccessible areas.

5. Monitoring and Trending Ensure quantitative baselines have been established for all structures within the scope of LR prior to entering the SPEO.

6. Acceptance Criteria Update the governing AMP procedure and other applicable procedures to include acceptance criteria on inspection of stainless steel and aluminum components for pitting and crevice corrosion, and evidence of cracking due to sec. In addition, require performance of an evaluation if stainless steel or aluminum surfaces exhibit evidence of sec, pitting, or crevice corrosion.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.11-1 L-2021-144 Attachment 26 Page 9 of 9 SLRA Section B.2.3.34, pages B-239 and B-240, as amended by SLRA Aging M

tS I

t1 f

d anagemen upp emen

, rev1s1on con 1nue :

Element Affected Enhancement Update the governing AMP procedure and other applicable procedures to include the following acceptance criteria:

Elastomers: No loss of material and no indications of loss of strength such as unacceptable surface cracking, crazing, scuffing, dimensional change (e.g.,

"ballooning" and "necking"), shrinkage, discoloration, or hardening.

Bolting and Fasteners: Loose bolts and nuts are not acceptable unless accepted by engineering evaluation.

Structural Sealants: Acceptable if the observed loss of material, cracking, and ha.rdening will not result in loss of sealing.

Earthen berm: No evidence of:

0 Settlement - unusual localized or overall settlement1 de!;!ressions 1 sinkholes 1 0

Slo!;!e instabilit~ - variance from original!~

constructed slo!;!es 1 unusual changes from original crest alignment and elevation 1 evidence of movement 0

Erosion - gullies or notches in slo!;!e Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.3.1, 263-1 L-2021-144 Attachment 27 Page 1 of 8 RAI 3.3.1, 263-1 (polystyrene (polymer) inserts used in manhole covers)

Background:

An AMR item in SLRA Table 3.5.2-11, "Yard Structures," as revised by the supplemental letter dated April 21, 2021, states that polystyrene (polymer) inserts used in manhole covers will be managed for the aging effects of blistering, cracking, hardening, loss of material, and loss of strength by the Structures Monitoring program.

The AMR item, associated with Table 1 item 3.3-1, 263, cites generic note E to indicate that this AMR item is consistent with the GALL-SLR Report for material, environment, and aging effects, but a different aging management program is credited. SLRA Section B.2.3.34, as revised by the supplemental letter dated April 21, 2021, provides an enhancement to the Structures Monitoring program to ensure that the polystyrene (polymer) foam components that are mounted on the underside of the manhole covers are added to the scope of the program as an elastomer material.

As described in the SRP-SLR, the GALL-SLR Report contains the NRC staff's generic evaluation of plant AMPs and establishes the technical bases for their adequacy. The applicant may use a plant-specific AMP or plant-specific aging management activities (within an existing AMP) as the basis for aging management of a specific structure or component. However, the GALL-SLR Report AMP Xl.S6, "Structures Monitoring," does not generically address the aging effects for polymeric materials. For those components, materials and aging effects combinations that are not generically addressed by the GALL-SLR Report, the NRC staff reviews the proposed AMPs or activities in accordance with the program element criteria that are defined in the SRP-SLR Appendix A.1, Subsection A.1.2.3, to ensure that the effects of aging for those structures or components will be adequately managed during the period of extended operation.

Issue:

During the audit, the staff reviewed PBN's procedure NP 7.7.9, Revision 21, and noted, in Attachment D, that the Structures Monitoring program addresses the acceptance criteria and aging effects of cracking and changes in material properties (i.e. hardening) for elastomers. However, the existing procedure does not clearly define or address the acceptance criteria and aging effects of blistering, loss of material, and loss of strength for elastomer components. Therefore, it is not clear how the Structures Monitoring program will adequately manage the aging effects for polystyrene (polymer) inserts used in manhole covers if the associated program elements are not clearly defined within the Structures Monitoring program. Additionally, due to the difference in the terminology used to described this component material, it is not clear if the insert components behaves more as an elastic material, as described GALL-SLR Report in Table IX.C for elastomers, or as a more solid/inelastic material (e.g. Polystyrene/foam).

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.3.1, 263-1 L-2021-144 Attachment 27 Page 2 of 8 Request:

Clarify the type of the polymer material used (e.g., elastic material similar to a flexible seal, inelastic material similar to solid or foam) to better characterize the inserts used for the manhole covers.

Clarify how the Structures Monitoring program will adequately manage the aging effects of polystyrene (polymer) inserts used with the manhole covers, including what parameters will be monitored or inspected, how frequently the components will be inspected, and the acceptance criteria that will be used, so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation.

NEPB Response:

The polymer material used in the manhole cover inserts is inelastic polystyrene foam material that shares a similar chemical makeup to other polymers, including plastics and elastomers. Polystyrene is made up of polymer chains with large side groups, which causes the material to be less flexible at room temperature and more flexible at higher temperatures [Ref. 1]. Lightweight polystyrene foams are thermoplastics that are advertised as good thermal insulators, inert, durable and resistant to water damage.

Therefore, for the purposes of aging management, the polystyrene foam is considered as a polymer. The manholes in the yard are inspected every five years by the Structures Monitoring AMP. In the outdoor environment, the manhole covers and inserts are exposed to cold temperatures during the winter, ultraviolet light and ozone, and precipitation.

SLRA Table 16-3 (Appendix A, Section 16.4) and Section B.2.3.34 include enhancements to inspect elastomer components for loss of material and loss of strength, with acceptance criteria including no loss of material and no indications of loss of strength such as unacceptable surface cracking, crazing, scuffing, dimensional change (e.g., "ballooning" and "necking"), shrinkage, discoloration, or hardening.

SLRA Table 16-3 (Appendix A, Section 16.4) and Section B.2.3.34, as amended by SLRA Aging Management Supplement 1, are revised as described below to include the blistering aging effect and polymers in the above enhancements.

References:

1. EPRI 1014800, Plant Support Engineering : Elastomer Handbook for Nuclear Power Plants, August 2007 Associated SLRA Revisions:

SLRA Table 16-3 (Appendix A, Section 16.4) and Section B.2.3.34, as amended by SLRA Aging Management Supplement 1, are revised.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.3.1, 263-1 L-2021-144 Attachment 27 Page 3 of 8 SLRA Table 16-3 (Appendix A, Section 16.4), pages A-105 and A-106, as amended by SLR Aging Management Supplement 1, are revised as follows:

Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging NUREG-2191 Commitment Implementation Management Section Schedule Program or Activity (Section) 38 Structures Xl.S6 Continue the existing PBN Structures Monitoring AMP, including enhancement No later than 6 Monitoring to:

months prior to the (16.2.2.34) a)

Revise inspection procedures to include guidance and acceptance SPEO, i.e.:

criteria on inspections of stainless steel and aluminum components for PBN1: 04/05/30 pitting and crevice corrosion, and evidence of cracking due to sec.

PBN2: 09/08/32 Perform an evaluation if stainless steel or aluminum surfaces exhibit evidence of sec, pitting, or crevice corrosion.

b) Revise inspection procedure scope to include polystyrene foam that is mounted to the underside of manhole covers as an elastomer/polymer material.

c)

• d) Revise inspection procedures to additionally inspect for the following items:

Increase in porosity and permeability, loss of strength, and reduction in concrete anchor capacity due to local concrete degradation in concrete structures.

Loss of material, blistering, and loss of strength for elastomers/~oly:mers (including ~oly:stvrene inserts for manhole covers).

Pitting and crevice corrosion, and evidence of cracking due to SCC for stainless steel and aluminum components.

Confirmation of the absence of water in-leakage through concrete.

e)Revise inspection procedures to include guidance on MEB inspection for loss of material (external bus duct enclosure surfaces and structural

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.3.1, 263-1 L-2021-144 Attachment 27 Page 4 of 8 Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging NUREG-2191 Commitment Management Section Program or Activity (Section) supports) and elastomer degradation (exterior housing gaskets, boots, and sealants).

f)

Clarify that if ground water leakage is identified then engineering evaluation, more frequent inspections, or destructive testing of affected concrete (to validate properties and determine pH) are required. When leakage volumes allow, assessments may include analysis of the leakage pH, along with mineral, chloride, sulfate, and iron content in the water.

g) Revise inspection procedure to specify that the responsible engineer (RE) shall be a registered professional engineer with knowledge in the design, evaluation, and in-service inspection of concrete structures and performance requirements of nuclear safety-related structures; or a degreed civil or structural engineer with at least ten years' experience in the design, construction, and inspection of concrete structures, with knowledge of the performance requirements of nuclear safety-related structures and potential degradation processes.

h) Revise inspection procedure to specify that accessible areas subject to similar conditions (material, environment, etc.) may be inspected in lieu of inaccessible areas, and include guidance for evaluating the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to the inaccessible areas.

ll

• f+il Ensure quantitative baselines have been established for all structures within the scope of LR prior to entering the SPEO.

H!sl Revise inspection procedure to include the following acceptance criteria:

Implementation Schedule

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.3.1, 263-1 L-2021-144 Attachment 27 Page 5 of 8 Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging NUREG-2191 Commitment Management Section Program or Activity (Section)

For Elastomers/(!Ollf'.mers {including (!Ollf'.Stlf'.rene inserts for manhole covers): No loss of material, no blistering, and no indications of loss of strength such as unacceptable surface cracking, crazing, scuffing, dimensional change (e.g., "ballooning" and "necking"), shrinkage, discoloration, or hardening.

For Bolting and Fasteners: Loose bolts and nuts are not acceptable unless accepted by engineering evaluation.

For Structural Sealants: Observed loss of material, cracking, and hardening will not result in loss of sealing.

ktll Localized distortion of the reactor cavity liner due to radiation induced volumetric expansion of the underlying concrete.

• SLRA Revision associated with RAI 3.5-1, 068-1.

Implementation Schedule

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.3.1, 263-1 L-2021-144 Attachment 27 Page 6 of 8 SLRA Section B.2.3.34, pages B-239 and B-240, as amended by SLR Aging Management Supplement 1, are revised as follows:

Element Affected

1. Scope

2. Preventive Actions

3. Parameters Monitored or Inspected

4. Detection of Aging Effects Enhancement Update the governing AMP procedure and other applicable procedures to add stainless steel and aluminum as a material that is inspected for pitting and crevice corrosion, and evidence of cracking due to sec.

Update the governing AMP procedure scope to include polystyrene foam that is mounted to the underside of manhole covers as an elastomer material.

Update the governing AMP procedure and other applicable procedures to additionally inspect the following elements:

Concrete Structures will be inspected for increase in porosity and permeability, loss of strength, and reduction in concrete anchor capacity due to local concrete degradation.

Elastomer/polymer (including polystyrene inserts for manhole covers) will also be inspected for loss of material, blistering, and loss of strength.

Pitting and crevice corrosion and evidence of cracking due to sec for stainless steel and aluminum components Concrete will be monitored to confirm the absence of water in-leakage Localized distortion of the reactor cavity liner due to radiation induced volumetric expansion of the underlying concrete.

Update the governing AMP procedure and other applicable procedures to include guidance on inspections for pitting and crevice corrosion, and evidence of cracking due to SCC for stainless steel and aluminum components.

Update the governing AMP procedure and other applicable procedures to include guidance on MEB inspection for loss of material (external bus duct enclosure surfaces and structural supports) and elastomer degradation (exterior housing gaskets, boots, and sealants).

Update the governing AMP procedure and other applicable procedures to clarify that if ground water leakage is identified then engineering evaluation, more frequent inspections, or destructive testing of affected concrete (to validate properties and determine pH) are required. When

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.3.1, 263-1 L-2021-144 Attachment 27 Page 7 of 8 Element Affected Enhancement leakage volumes allow, assessments may include analysis of the leakage pH, along with mineral, chloride, sulfate and iron content in the water.

Update the governing procedure to specify that the responsible engineer (RE) shall be a registered professional engineer with knowledge in the design, evaluation, and in-service inspection of concrete structures and performance requirements of nuclear safety-related structures; or a degreed civil or structural engineer with at least ten years' experience in the design, construction, and inspection of concrete structures, with knowledge of the performance requirements of nuclear safety-related structures and potential degradation processes.

Revise inspection procedure to specify that accessible areas subject to similar conditions (material, environment, etc.) may be inspected in lieu of inaccessible areas, and include guidance for evaluating the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to the inaccessible areas.

5. Monitoring and Trending Ensure quantitative baselines have been established for all structures within the scope of LR prior to entering the SPEO.

6. Acceptance Criteria Update the governing AMP procedure and other applicable procedures to include acceptance criteria on inspection of stainless steel and aluminum components for pitting and crevice corrosion, and evidence of cracking due to SCC. In addition, require performance of an evaluation if stainless steel or aluminum surfaces exhibit evidence of SCC, pitting, or crevice corrosion.

Update the governing AMP procedure and other applicable procedures to include the following acceptance criteria:

Elastomers/polymers (including polystyrene inserts for manhole covers): No loss of material.1...

no blistering, and no indications of loss of strength such as unacceptable surface cracking, crazing, scuffing, dimensional change (e.g., "ballooning" and "necking"), shrinkage, discoloration, or hardening.

Bolting and Fasteners: Loose bolts and nuts are not acceptable unless accepted by engineering evaluation.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.3.1, 263-1 L-2021-144 Attachment 27 Page 8 of 8 Element Affected Enhancement Structural Sealants: Acceptable if the observed loss of material, cracking, and hardening will not result in loss of sealing.

• SLRA Revision associated with RAI 3.5-1, 068-1.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.3.1, 111-1 L-2021-144 Attachment 28 Page 1 of 2 RAI 3.3.1, 111-1 (inconsistency in applicability of item 3.3-1, 111)

Background:

SLRA Table 3.3-1, as revised by the supplemental letter dated April 21, 2021, states that item 3.3-1, 111 is not applicable and that item 3.5.1, 100 is used instead to adequately manage the stainless steel new fuel storage racks at PBN. However, two line items in SLRA Table 3.5.2-1, "Containment Building Structure and Internal Structural Components," associated with Table 1 item 3.3-1, 111, are used by the SLRA to demonstrate that other components (i.e. steel liner in the reactor cavity and miscellaneous steel structural components) exposed to an air - indoor environment are being managed for loss of material and/or distortion by the Structures Monitoring program.

The GALL-SLR Report item Vll.A1.A-94 (page VII A1-2, Table A1), associated with SRP-SLR Table 1 item 3.3-1, 111, recommends the Structures Monitoring program to manage the aging effects of loss of material due to general, pitting, or crevice corrosion for new fuel storage racks made out of steel material that are exposed to an air - indoor uncontrolled environment.

Issue:

It is not clear how the SLRA Table 2 line items are consistent with the SLRA disposition of Table 1 item 3.3-1, 111, which states that the Table 1 item is not applicable for PBN.

Request:

Clarify the inconsistency identified between the SLRA Table 1 item and SLRA Table 2 line items associated with AMR item 3.3-1, 111.

NEPB Response:

Item 3.3-1, 111 is not applicable to the new fuel storage racks at PBN, which are stainless steel and are aligned with Item 3.5-1, 100. However, Item 3.3-1, 111 is applicable to other components in SLRA Table 3.5.2-1 and for these components, the evaluation of the original SLRA should have been retained. SLRA Table 3.3-1 is revised as described below for consistency.

References:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.3.1, 111-1 L-2021-144 Attachment 28 Page 2 of 2 Associated SLRA Revisions:

SLRA Table 3.3-1 item 111, as amended by SLRA Aging Management Supplement 1, is revised as follows:

Table 3.3-1: Summary of Aging Management Evaluations for the Auxiliary Systems Item Component Aging Aging Management Further Discussion Number Effect/Mechanism Program (AMP)/TLAA Evaluation Recommended 3.3-1,

Steel structural steel Loss of material due AMP Xl.S6, No Not ApplicableConsistent with 111 exposed to air -

to general, pitting, "Structures Monitoring" NUREG-2191.

indoor uncontrolled crevice corrosion The Structure Monitoring AMP is used to manage loss of material in structural steel ex~osed to uncontrolled indoor air. This line item is used to evaluate structural items in Section 3.5.

New Fuel storage racks at PBN are stainless steel and addressed in item 3.5.1, 100.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.1-1 L-2021-144 Attachment 29 Page 1 of 3

12. SLRA Section 4.3.1, "Metal Fatigue Class 1 Components" Regulatory Basis:

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

RAI 4.3.1-1

Background:

SLRA Table 4.3.1-1 describes the allowable 80-year transient cycle numbers in comparison with the original 40-year design cycle numbers for the applicant's fatigue analyses.

Issue:

For the "control rod drop," "excessive FW [feedwater] flow," and QBE [operating basis earthquake] transients, SLRA Table 4.3.1-1 does not identify existing design allowable cycle numbers. However, the table lists specific 80-year allowable cycle numbers for these transients (table items 32, 33 and 34). The basis for the 80-year allowable cycles is not clear.

For the "feedwater cycling at hot standby" and "boron concentration equilibrium" transients, the existing design allowable cycle numbers are described as 2000 and 23360 respectively in SLRA Table 4.3-1. However, the table does not provide the 80-year allowable design cycles for these transients (table items 26 and 27). It is not clear why 80-year allowable cycles are not specified.

Request:

1. Clarify the basis for the 80-year allowable cycles for the "control rod drop,"

"excessive FW flow," and QBE transients.

2. Explain why SLRA Table 4.3.1-1 does not identify the 80-year allowable cycle numbers for the "feedwater cycling at hot standby" and "boron concentration equilibrium" transients as part of fatigue monitoring activities.

NEPB Response:

1. These values were selected as being typical design values for Westinghouse plants, since UFSAR Table 4.1-8 does not include these but they were used in some fatigue analyses. Since the design cycles in UFSAR Table 4.1-8 will not be exceeded in 80-years of operation, it was judged to be reasonable to use typical design values for the control rod drop, excessive FW flow, and QBE transients.

2. An engineering evaluation was performed to establish a conservative basis for estimating the acceptable number of "feedwater cycling at hot standby" and "boron concentration equilibrium" transients.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.1-1 L-2021-144 Attachment 29 Page 2 of 3 For the "feedwater cycling at hot standby" transient, a review of periods of hot standby operation showed the largest temperature difference was 11 °F and the average temperature difference was 4.5°F as compared to the design temperature excursion of -27.5/+32.5°F (60°F l'i T). Based on a conservatively estimated number of cycles and magnitude of temperature changes, the estimated current CUF is less than 0.1. Assuming 12 cycles per each startup and shut down cycle, the plant can experience 18,000 additional startup and shutdown cycles as compared to 2,000 design transient cycles.

For the "boron concentration equilibrium" transient, the engineering evaluation determined that an average of one equilibrium cycle per month will be bounding for the current and subsequent period of extended operation as compared to a conservative estimate of two per day for the first 20 years of operation when the plant performed load following operation. The engineering evaluation determined that subtracting the estimated number of cycles to date from the number of design cycles leaves 8, 760 cycles for the balance of plant life. At this rate, the plant would not exceed the number of allowable cycles for more than 700 years.

SLRA Table 16-3 is updated to add commitment item 1 (d) to include monitoring of "feedwater cycling at hot standby" and "boron concentration equilibrium" transients to ensure they remain within limits.

References:

None

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.1-1 L-2021-144 Attachment 29 Page 3 of 3 Associated SLRA Revisions:

SLRA Table 16-3 item 1, page A-64, is revised as follows:

16.4.

Sybseguent License Renewal lSLRl Commitments List Table 16-3 List of SLR Commitments and Implementation Schedule No.

Aging NU REG-Commitment Management 2191 Program or Section Activity (Section) 1 Fatigue X.M1 Continue the existing PBN Fatigue Monitoring AMP, including enhancement to:

Monitoring a)

Update the plant procedure to monitor chemistry parameters that provide (16.2.1.1) inputs to Fen factors used in CU Fen calculations.

b)

Update the plant procedure to identify and require monitoring of the 80-year projected plant transients that are utilized as inputs to CU Fen calculations.

c)

Update the plant procedure to identify the corrective action options to take if component specific fatigue limits are approached.

d} U~date the ~lant ~rocedure to include monitoring of "feedwater C:lf'.cling at hot standb:ll and "boron concentration eguilibrium" transients to ensure thev remain within limits.

Associated

Enclosures:

None.

Implementation Schedule No later than 6 months prior to the SPEO, i.e.:

PBN1 :

04/05/2030 PBN2:

09/08/2032

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.1-2 L-2021-144 Attachment 30 Page 1of1 RAI 4.3.1-2

Background:

SLRA Section 4.3.1 states that the cumulative usage factor (CUF) for the pressurizer spray piping has been projected to the end of the subsequent period of extended operation. The SLRA section also indicates that stratification cycles are conservatively projected based on thermocouple data with a leaking spray control valve that is assumed to leak throughout 80 years of plant operation.

In addition, SLRA Section 4.3.1 explains that the fatigue analysis for the piping results in a CUF value of 0.369 for 80 years of operation. The section states that, due to the conservatism applied to this analysis, cycle monitoring is not required for the piping.

Issue:

Additional information is needed to confirm that the previous inspection results support the applicant's approach that cycle monitoring is not used for the pressurizer spray piping.

Request:

Clarify whether the previous inspection results support the applicant's approach that cycle monitoring is not used for the pressurizer spray piping. If the inspections identified a relevant indication of fatigue in the piping, provide justification for (1) why cycle monitoring is not needed for the piping and (2) why inspection activities are not proposed to manage fatigue for the piping.

NEPB Response:

The pressurizer spray piping location with the maximum calculated fatigue usage is associated with Unit 1 ISi weld RC-03-PS-1002-24. This location is included in the ISi program to be performed every 10 years. The ultrasonic NOE inspections performed at this location have not identified any recordable indications. The most recent ultrasonic examination was performed in 2017. Therefore, considering the conservatively calculated CUF at 80 years of operation has significant margin to the allowable usage and inspections performed at 10-year intervals would identify relevant indications, cycle monitoring is not needed for this piping. Inspection activities are performed in accordance with the ASME Section XI ISi program.

References:

None.

Associated SLRA Revisions:

None.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.3-1 L-2021-144 Attachment 31 Page 1 of 7

13. SLRA Section 4.3.3, "Metal Fatigue of Non-Class 1 Components" Regulatory Basis:

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

RAI 4.3.3-1

Background:

SLRA Section 4.3.3 addresses the fatigue analysis waiver for non-Class 1 components in accordance with the ANSI B31.1 code. In the SLRA section, Table 4.3.3-2 discusses the projected numbers of equivalent full temperature cycles to demonstrate that the total transient cycle for each evaluated piping system does not exceed the 7000 cycle threshold that allows no reduction to the allowable stress range.

Specifically, Table 4.3.3-2 indicates that, in accordance with the design transient cycles specified in UFSAR supplement (SLRA Appendix A) Table 4.1 -8, the 7000 cycle threshold is not exceeded by the following systems: (1) feedwater and condensate system; (2) main and auxiliary steam system; (3) reactor coolant system (Non-Class 1 );

and (4) safety injection system.

Issue:

In comparison, ANSI B31.1 indicates that, if the range of temperature change varies for transient cycles, the equivalent full temperature cycles may be computed by considering the lesser temperature change ratios based on the partial temperature changes of the transient cycles. However, SLRA Section 4.3.3 does not clearly discuss which design transients can contribute to the equivalent full temperature cycles for the evaluated systems in addition to the plant heatup and cooldown transients.

Request:

1. Clarify which design transients, other than plant heatup and cooldown transients, in USFAR supplement Table 4.1-8 and SLRA Table 4.3.1-1 can contribute to the equivalent full temperature cycles for each of the following systems: (1) feedwater and condensate system; (2) main and auxiliary steam system; (3) reactor coolant system, non-Class 1; and (4) safety injection system.

2. Considering the design transients identified in item 1 and their cycles, confirm that the equivalent full temperature cycles do not exceed the 7000 cycle threshold in each system.

NEPB Response:

1. The following design transients, in addition to 200 plant heatup and cooldown transients, have been conservatively considered to contribute to the equivalent full

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.3-1 L-2021-144 Attachment 31 Page 2 of 7 temperature cycles for each of the following PBN Units 1 and 2 systems: (1) feedwater and condensate system; (2) main and auxiliary steam system; (3) reactor coolant system, non-Class 1; and (4) safety injection system:

Reactor trip - 400 cycles Trip due to loss of RCP - 4 cycles Loss of load (trip) - 80 cycles Loss of power (trip) - 40 cycles Loss of flow (trip) - 80 cycles 10% step load increase/decrease - 2000 cycles 50% step load decrease - 200 cycles Unit loading/unloading 5% - 18,300 cycles (assume 915 equivalent full temperature cycles)

Note that even though the 10% step load increase/decrease and 50% step load decrease design transients are not equivalent full temperature cycles, the total number of their design cycles have been conservatively included in the total cycle count for the subject systems. However, the 18,300 design cycles for the unit loading/unloading 5% transient from UFSAR Table 4.1-8 have been reduced by a factor of twenty (20) to approximate the number of equivalent full temperature cycles. The total of these transients plus the 200 heatup and cooldown transients equals 3919 equivalent full temperature cycles, which is significantly less than the cycle limit of 7000.

2. Confirmation that the equivalent full temperature cycles for the 1) feedwater and condensate system; 2) main and auxiliary steam system; 3) reactor coolant system, non-Class 1; and 4) safety injection system do not exceed the 7000 cycle threshold is provided in the update to PBN SLRA Table 4.3.3.2.

References:

None

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.3-1 L-2021-144 Attachment 31 Page 3 of 7 Associated SLRA Revisions:

SLRA Section 4.3.3, pages 4.3-10 through 4.3-11, and Table 4.3.3-2 are revised as follows:

Table 4.3.3-1 Stress Range Reduction Factors for ANSI 831.1 Piping Number of Equivalent Full Stress Range Temperature Cycles Reduction Factor 7,000 and less 1.0 7,000 to 14,000 0.9 14,000 to 22,000 0.8 22,000 to 45,000 0.7 45,000 to 100,000 0.6 100,000 and over 0.5 The reduction factor is 1.0 provided the number of anticipated cycles is limited to 7000 equivalent full temperature cycles for piping and tubing. A review of the ANSI 831.1 piping within the scope of SLR was performed in order to identify those systems that operate at elevated temperature and to establish a conservative number of projected cycles based on 80 years of operation.

Typically, these piping and tubing systems are subject to continuous steady-state operation and experience temperature cycling only during plant heatup and cooldown, during plant transients, or during periodic testing.

From the EPRI Report TR-104534, "Fatigue Material Handbook" Volume 2, Section 4 (Reference 4.8.16) and the EPRI Non-Class 1 Mechanical Implementation Guideline and Mechanical Tools (Reference ML12335A508), piping and tubing systems subject to thermal fatigue due to temperature cycling are described as, "For initial screening, systems in which the fluid temperature can vary more than 200°F in austenitic steel components and more than 150°F in carbon and low alloy steel components are potentially of concern for fatigue due to thermal transients. Thus, carbon steel systems or portions of systems with operating temperatures less than 220°F and stainless steel systems or portions of systems with operating temperatures less than 270°F may generally be excluded from such concerns, since room temperature represents a practical minimum exposure temperature for most plant systems."

All non-Class 1 mechanical systems within the scope of the PBN SLRA were initially screened for the TLAA associated with metal fatigue. PBN SLR mechanical systems with maximum fluid temperatures below the limits specified above are not considered to be susceptible to metal fatigue and a TLAA is not applicable. Therefore, any PBN non-Class 1 mechanical system or portions of systems with operating temperatures above 220°F are conservatively evaluated for metal fatigue. The non-Class 1 piping and tubing systems requiring evaluation for the metal fatigue SLR are listed in Table 4.3.3-2 below.

Once a system is established to operate at a temperature above 220°F, system operating characteristics are established, and a determination is made as to

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.3-1 L-2021-144 Attachment 31 Page 4 of 7 whether the system is expected to exceed 7000 full temperature cycles in 80 years of operation. In order to exceed 7000 full temperature cycles during the 80-year SPEO, a system would be required to need to be subjected to a full temperature cycle heatup and cooldowfl approximately once every four days.

Thermal cycling of base-loaded nuclear power plant systems at this rate is not practical for PBN Units 1 and 2. For the systems that are subjected to elevated temperatures above the fatigue threshold, an evaluation was performed to determine a conservative number of projected full temperature cycles for 80 years of plant operation. These projections, which are presented in Table 4.3.3-2, indicate that 7000 thermal cycles will not be exceeded for 80 years of operation.

TLAA Disposition: 10 CFR 54.21(c)(1)(i)

The ANSI 831.1 allowable stress calculations remain valid for the SPEO. The results demonstrate that the number of assumed thermal cycles will not be exceeded in 80 years of plant operation.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.3-1 L-2021-144 Attachment 31 Page 5 of 7 Table 4.3.3-2 P. tdN b

fFllT ro1ec e um er o LI t

empera ure c :ye es Description Conservative Basis for Cycle Projection Auxiliary Feedwater The turbine driven AFW pumps are normally in standby service and are started quarterly by in-service testing procedures. This would equate o 4 times per year x 80 years = 320 cycles leaving significant margin to 7000 cycles to account for any additional pump starts in support of plant startup and shutdown.

Chemical and Normal charging and letdown during the reactor Volume Control coolant system power operation is at steady state temperature. Conservatively assume 4 hermal and loading cycles for the chemical and volume control system per

~ear.

Emergency Power PBN surveillance requirements require the emergency diesel generators to be started once per 31 days (12 times per year). SR 3.8.1.5, 3.8.1.6 and 3.8.1.7 require the emergency diesel generators to be started once per 18 months (assume once per year). Total cycles = 13 cycles per year x 80 years =

1040 cycles.

Feedwater and Feedwater and condensate equivalent full Condensate emoerature cvcles transients relative to pewef-

~~eration (plant heatup and 1 -

-r-

• ---~=- ";) are consistent with reactor coolant system transients from !JFSAR Table 4.1-8.

~s demonstrated below for the reactor coolant svstem the total for these events eauals 3919 eauivalent full temoerature cvcles which is sianificantlv less than the cvcle limit of 7000.

Fire Protection Fire pump diesel engine cycles only during pump testing, which occurs weekly. 52 cycles/year x 80 years = 4, 160 cycles.

Projected Cycles for 80 years Less than 7,000 cycles Less than 7,000 cycles Less than 7,000 cycles Less than 7,000 cycles Less than 7,000 cycles

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.3-1 L-2021-144 Attachment 31 Page 6 of 7 Table 4.3.3-2 P. tdN b

fFllT ro1ec e um er o u

t empera ure c vc es Description Conservative Basis for Cycle Projection Heating Steam Cycles based on seasonal heating.

Conservatively assume 85 cycles per year (80 jYears x 85/year = 6,800 cycles)

Main and Auxiliary Main and auxiliary steam equivalent full Steam

~emoerature cvcles transients relative to power 1..... ~...... 1-

• ~~ 1~1 h~-

,~ ~~,.J ~~~.. _,

"~' are

\\I"' *- *

--*- r- -* -

' I consistent with reactor coolant system transients

~ram !,!FSAR Table 4.1-8. As demonstrated below for the reactor coolant svstem the otal for these events eauals 3919 eauivalent

~ull temoerature cvcles which is sianificantlv less than the cvcle limit of 7000.

Plant Sampling A specific hot leg sample evaluation was performed for PBN license renewal and it resulted in a total of 6702 cycles for 60 years of operation. The critical assumption in the evaluation that would be used for the additional 20 year operating period is thermal cycling of the hot leg sample line occurs only 10 times per jYear. This would result in an additional 200 cycles for the hot line sample line, resulting in a

~otal number of 6902 cycles for the 80-year SPEO.

Reactor Coolant Reactor coolant system thermal and loading System cycle limits are provided in UFSAR Table 4.1-8.

A conservative oroiection of eauivalent full emoerature cvcles for the reactor coolant 1Svstem for the SPEO would include the

~ollowina transients:

• Plant heatup and cooldown = 200 limits are 200 cycles each.

• Reactor tri12 = 400

• Trig due to loss of RCP = 4

• Loss of load (tri12l = 80

• Loss of 12ower (tri12l = 40

• Loss of flow (tri12l = 80

• 10% ste12 load increase/decrease= 2000

• 50% ste12 load decrease = 200

• Unit loading/unloading 5% = 915 The total for these desian transients eauals 3919 equivalent full tem12erature c~cles 1 which is significant!~ less than the c~cle limit of 7000.

Projected Cycles for 80 years Less than 7,000 cycles Less than 7,000 cycles Less than 7,000 cycles Less than 7,000 cycles

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.3-1 L-2021-144 Attachment 31 Page 7 of 7 Table 4.3.3-2 p

rorecte dN b

fFllT um er o u

emperature c :ye es Description Conservative Basis for Cycle Projection Residual Heat Residual heat removal system piping is Removal heated during shutdowns and startups.

Assume four thermal cycles per heatup and cooldown each plant shutdown.

Safety Injection Portions of safety injection system~ +s are 12art of G8RReGtetl te the reactor coolant system 12ressure bounda~ 12i12ing and +s fil!LSUbjected to reactor coolant system thermal and loading cycle limits provided in UFSAR Table 4.1-8. As demonstrated above for the reactor coolant system 1 the total for these events eguals 3919 eguivalent full tem12erature cycles 1 which is significantly less than the cycle limit of 7000.

Safety injection pump testing occurs at temperature less than 200°F.

Waste Disposal Conservatively assume high temperature portions of the liquid waste management system are cycles once per week. Total cycles = 52 cycles per year x 80 years =

4, 160 cycles.

Associated

Enclosures:

None.

Projected Cycles for 80 years Less than 7,000 cycles Less than 7,000 cycles Less than 7,000 cycles

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.4-1 L-2021-144 Attachment 32 Page 1 of 3

14. SLRA Section 4.3.4, "Environmentally Assisted Fatigue" Regulatory Basis:

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

RAI 4.3.4-1

Background:

SLRA Section 4.3.4 addresses the environmentally assisted fatigue (EAF) time-limited aging analysis (TLAA). GALL-SLR AMP X.M1, "Fatigue Monitoring," and SRP-SLR Section 4.3.2.1.2 recommend that an EAF TLAA evaluate the component and piping locations identified in NUREG/CR-6260.

SLRA Section 4.3.4 includes EAF evaluations for most of the sample locations addressed in NUREG/CR-6260 for the Westinghouse-designed reactor. However, the SLRA does not clearly address the following locations identified in NUREG/CR-6260:

(1) pressurizer surge line piping locations other than the reactor coolant loop surge line nozzle and pressurizer vessel surge line nozzle; and (2) RHR system Class 1 piping locations other than the RHR tee and accumulator safe injection nozzle locations.

Issue:

The existing fatigue analysis for the pressurizer surge line described in WCAP-13509 indicates that the maximum CUF value is estimated at a surge line piping location rather than the reactor coolant loop surge nozzle or the pressurizer vessel surge nozzle.

However, the SLRA does not clearly address this location identified in WCAP-13509 in the evaluation of the NUREG/CR-6260 locations. In addition, SLRA Section 4.3.4 does not clearly discuss the basis for selecting the RHR tee and accumulator safety injection nozzle locations as the leading EAF locations for the RHR piping system.

Request:

1. Reconcile the potential inconsistence of pressurizer surge line leading locations between SRLA Section 4.3.4 (hot leg surge and pressurizer surge nozzles) and WCAP-13509 (surge line piping location).

2. Clarify the basis for selecting the RHR tee and accumulator safety injection nozzle locations as the leading EAF locations for the RHR system Class 1 piping.

NEPB Response:

1. The CLB ASME Code Section Ill analysis of record (AOR) remains WCAP-13509, "Structural Evaluation of the Point Beach Units 1 & 2 Pressurizer Surge Lines, Considering the Effects of Thermal Stratification." This analysis used design cycles and design severity transients. In this analysis the surge piping near the terminal end

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.4-1 L-2021-144 Attachment 32 Page 2 of 3 of the pressurizer surge nozzle was determined to be the highest fatigue location using idealized transient definitions and conservative analysis methodology.

In contrast, Point Beach's fatigue monitoring system uses measured plant data, reflective of actual operating practices over time. The Point Beach fatigue monitoring calculates fatigue at both of these terminal ends of the surge piping, and thus addresses both the NUREG/CR-6260 location and the location with maximum calculated fatigue in the CLB analysis. The fatigue monitoring system has demonstrated that the hot leg surge nozzle (the NUREG/CR-6260 location) is limiting with respect to fatigue, in comparison to the other terminal end near the pressurizer surge nozzle. This analysis (NMC-03Q-307 Rev 0, Surge line and Pressurizer Lower Head Fatigue Analysis) was performed using design cycles and actual transient severity.

2. The basis for the leading SI and RHR system EAF locations is consistent with and based on work performed for initial LR for the SI (accumulator) nozzle and the RHR system.

Since PBNP is a B31.1 piping plant, there were no original fatigue analyses for the Class 1 piping locations. Therefore, a plant-specific piping model was created that addressed the SI and RHR piping (see figure below). A fatigue analysis of this piping was then performed in accordance with ASME Code Section Ill, NB-3600 rules using design cycles. The locations identified as leading EAF locations in that analysis are the SI (accumulator) nozzle and the RHR tee. The SI (accumulator) nozzle is the same location as that identified in NUREG/CR-6260. The RHR piping NUREG/CR-6260 location is an 8"x8"x8" tee in the alternate RHR line connecting to the normal RHR line adjacent to the RHR connection to the accumulator tank C line.

Since PBNP is not configured with the alternate RHR line, the modeled piping included the 1 O"x1 O"x10" normal RHR connection to the accumulator tank, which is the leading EAF location in the RHR piping.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.4-1 L-2021-144 Attachment 32 Page 3 of 3 I"..

flgv r e A: S1h!IOA1l!> v1u:

References:

None.

Associated SLRA Revisions:

None.

Associated Encfosures:

None.

PipinQ Reaion 1

2 3A 3B 4

5A 5B 6

7 8

Node Points Start End 105 110 36 105 21 36 36 48 48 61 1

17 17 21 17 236 236 248 248 250

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.4-2 L-2021-144 Attachment 33 Page 1 of 2 RAI 4.3.4-2

Background:

GALL-SLR AMP X.M1, "Fatigue Monitoring," and SRP-SLR Section 4.3.2.1.2 recommend that an EAF analysis include the evaluation of the sample locations identified in NUREG/CR-6260 and that, if plant-specific locations are more limiting than those sample locations, the plant-specific location should be also evaluated in the EAF TLAA.

SLRA Section 4.3.4 explains that the piping systems of PBN were originally designed and constructed in accordance with the ANSI B31.1 code. After the commence of the operation, some piping systems were evaluated in accordance with the fatigue analysis provisions of ASME Code, Section Ill.

The SLRA also indicates that the sample locations specified in NUREG/CR-6260 and the additionally identified pressurizer spray piping are evaluated in the EAF TLAA to address the piping systems that were designed per ANSI B31.1 but not analyzed per ASME Code, Section Ill.

The applicant did not perform a detailed screening analysis for the ANSI B31.1 piping systems. Instead, the applicant used the previous EAF screening results of Westinghouse-designed reactors (i.e., Surry plant) as the basis of selecting the NUREG/CR-6260 locations and pressurizer spray piping for further EAF evaluation.

Issue:

It is not clear to the staff why the EAF analysis results of the Surry plant can represent the applicant's reactors in terms of identifying plant-specific locations that may be more limiting than the NUREG/CR-6260 locations. In addition, additional information is needed to clarify why the leading locations identified in the applicant's EAF evaluation are bounding for the piping systems designed per ANSI 831.1 but not analyzed per ASME Code, Section Ill or the guidance on leading EAF locations of NUREG/CR-6260.

Request:

1. Provide justification for why the EAF analysis results of the Surry plant can represent the applicant's reactors in terms of identifying plant-specific locations that may be more limiting than the NUREG/CR-6260 locations.

2. Justify why the leading locations identified in the applicant's EAF evaluation are bounding for the piping systems designed per ANSI B31.1 but not analyzed per ASME Code, Section Ill or the guidance on leading EAF locations of NUREG/CR-6260.

NEPB Response:

1. The Surry plant was not used as a direct comparison to Point Beach because none of the Westinghouse 2-Loop plants have had a similar review as that performed for Surry. Surry was used as an example to show that a comprehensive review of the Class 1 B31.1 piping demonstrated that the plant design and vintage locations

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.3.4-2 L-2021-144 Attachment 33 Page 2 of 2 selected in NUREG/CR-6260 are reasonably representative of the most limiting locations. Since there is no requirement to perform Class 1 fatigue analysis of additional locations designed to the 831.1 Power Piping Code, it's reasonable to conclude that the locations other than the NUREG/CR-6260 and pressurizer spray piping are not more limiting.

2. The current licensing basis analyses for the Class 1 piping is 831.1 other than locations where fatigue analyses were performed to evaluate NUREG/CR-6260 locations and the pressurizer spray piping. There is no requirement to perform fatigue analysis in addition to the current licensing basis analysis. Furthermore, where new plant-specific piping analyses were performed to ASME Section Ill that encompassed much of the system piping, the NUREG/CR-6260 locations and pressurizer spray line location were the controlling locations.

References:

None.

Associated SLRA Revisions:

None.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAJ No. 16.3.6-1 L-2021-144 Attachment 34 Page 1 of 3

15. SLRA Section 4.6, "Containment Liner Plate, Metal Containments, and Penetrations" Regulatory Basis:

10 CFR § 54.21 (c)(1) requires the applicant to evaluate time limited aging analyses (TLAA). 10 CFR § 54.21 (d) requires that the FSAR supplement for the facility must contain a summary description of the evaluation of the TLAA for the period of extended operation determined by 54.21 (c)(1 ).

RAI 16.3.6-1

Background:

Section 4.6, as amended by Attachment 29 of SLRA Supplement 1 dated April 21, 2021 (ADAMS Accession No. ML21111A155), dispositioned the TLAA for fatigue of the containment liner plate and carbon steel piping penetrations in accordance with 10 CFR 54.21 (c)(1 )(i) for the subsequent period of extended operation. SLRA Appendix A, Section 16.3.6 "Containment Liner Plate and Penetrations Fatigue Analysis" which provides the UFSAR supplement description for this TLAA was not amended by Supplement 1 for conforming changes.

Issue:

SLRA Appendix A, Section 16.3.6 continues to include (specifically in the last two paragraphs on SLRA pages A-58 and A-59) description and partial disposition in accordance with 10 CFR 54.21 (c)(1 )(iii) that is inconsistent with the evaluation description and disposition in SLRA Section 4.6, as amended by Attachment 29 of Supplement 1 dated April 21, 2021. The staff is unable to make its determination that the UFSAR supplement provides an adequate summary description for the TLAA.

Request:

Provide a revised SLRA Appendix A, Section 16.3.6 with an UFSAR supplement summary description for the containment liner and steel piping penetrations TLAA that is consistent with the TLAA description and disposition in SLRA Section 4.6, as amended by Supplement 1 dated April 21, 2021.

NEPB Response:

A revised SLRA Appendix A, Section 16.3.6 with an UFSAR summary description for the containment liner and steel piping penetrations TLAA that is consistent with the TLAA description and disposition in SLRA Section 4.6, as amended by Supplement 1 dated April 21, 2021, is included in the associated SLRA Appendix A, Section 16.3.6 revision below.

References:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 16.3.6-1 L-2021-144 Attachment 34 Page 2 of 3 Associated SLRA Revisions:

SLRA Appendix A, Section 16.3.6, pages A-58 and A-59, are revised as follows:

plate. The attachment sleeves are designed in accordance with ASME Boiler and Pressure Vessel Code, Section Ill, fatigue considerations.

For item one, the number of thermal cycles due to annual outdoor temperature variations was increased from 60 to 80 for the SPEO. The effect of this increase is insignificant in comparison to the assumed 500 thermal cycles due to containment interior temperature varying during heatup and cooldown of the reactor coolant system (RCS). The 500 thermal cycles includes a margin of 300 thermal cycles above the 200 allowable RCS design heatup and cooldown cycles, which is sufficient margin to accommodate the additional 20 cycles of annual outdoor temperature variation.

For item two, the assumed 500 thermal cycles was evaluated based on the more limiting heatup and cooldown design cycles (transients) for the major components of the RCS. As indicated in Table 4.1-8 of the PBN UFSAR, the major components of the RCS were designed to withstand 200 heatup and cooldown cycles and is conservative enough to envelop the projected cycles for the SPEO.

For item three, the assumed value for thermal cycling due to the design basis accident remains valid. No maximum design basis accident has occurred to date and none is expected, therefore, this assumption is considered valid for the SPEO.

For item four, the design of the containment piping penetrations was previously reviewed as part of the original PBN license renewal application and the EPU project. The containment penetrations were designed, fabricated, inspected, and tested in accordance with ASME Boiler and Pressure Vessel Code, Section Ill, Class B, 1968 edition and all addenda. The main steam, feedwater, blowdown, and letdown systems are the only piping systems penetrating the containment that contribute significant thermal loading on the liner plate. Due to the higher operating temperature, the main steam piping penetration was considered bounding and was therefore analyzed for fatigue through the SPEO. The analysis of the main steam piping penetration sleeve and the sleeve end fitting connecting the pressure piping to the sleeve verified that the six conditions of ASME Code, Section Ill, Subsection A, N 415.1, 1965, are satisfied for the PEO and a fatigue analysis of the piping penetrations is not required.

PBN has been unable to locate the original fatigue analysis or confirm if a fatigue waiver exists for the PBN containment penetrations other than steel piping penetrations, piping penetrations with dissimilar metal welds, and for the expansion joints of the containment structure reactor fuel transfer tube.

Therefore, consistent 1Nith ~JUREG 2192, cracking due to cyclic loading of non piping containment penetrations will be managed by the PBN /\\SME Section XI, Subsection IWE /\\MP (Section 16.2.2.29) and periodic supplemental surface examinations incorporated into and consistent with the frequency of the PBN 10 GFR Part 50, Appendix J /\\MP (Section 16.2.2.32).

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 16.3.6-1 L-2021-144 Attachment 34 Page 3 of 3 Therefore, the fatigue analyses associated with the containment liner plate and steel piping penetrations have been evaluated and determined to remain valid for the SPEO in accordance with 10 CFR 54.21(c)(1)(i). The fatigue analyses associated with the containment personnel airlocks, equipment hatch, personnel hatch, electrical penetrations, piping penetrations with dissimilar metal welds, and expansion joints of the containment structure fuel transfer tube on each unit will be managed by the PBN ASME Section XI, Subsection IWE AMP (Section 16.2.2.29) and the PB~J10 CFR Part 50, Appendix J AMP (Section 16.2.2.32) in accordance with 10 CFR 54.21(c)(1)(iii).

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.7.1-1 L-2021-144 Attachment 35 Page 1 of 2

16. SLRA Section 4. 7.1, "Leak-Before-Break of Reactor Coolant System Loop Piping" Regulatory Basis:

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

RAI 4.7.1-1

Background:

[No background was stated in this RAI.]

Issue:

During the staff review of Section 4.7.1 (Leak Before Break of Reactor Coolant System Loop Piping) of the SLRA, the applicant stated that the PBN Unit 2 steam generator (SG) inlet and outlet nozzles contain Alloy 82/182 dissimilar metal welds which are susceptible to primary water stress corrosion cracking (PWSCC). To mitigate PWSCC due to the existence of Alloy 82/182, Alloy 52/152 weld inlays were applied to the SG primary nozzle safe end welds that are exposed to primary coolant. In Section 4.7.1, there is no mention of the material or whether weld inlays were applied to the SG inlet and outlet nozzles in Unit 1.

Request:

Please provide this information and if this material is resistant to PWSCC.

NEPB Response:

As noted in Section 2.1 of WCAP-14439 (Reference 1 ), materials susceptible to primary water stress corrosion cracking (PWSCC) are not found in the primary loop piping and nozzles of Point Beach Unit 1. Reference 2 shows the Steam Generator inlet and outlet nozzles of Point Beach Unit 1 are a carbon steel base metal with austenitic stainless steel buttering (Type 309L and 308L weld filler metal, per Reference 3).

A stainless steel field weld joins the stainless steel buttered Steam Generator nozzles (inlet and outlet) to the respective RCL piping elbows, which are cast stainless steel (A351 CF8M). These welds do not utilize Alloy 82/182 for joining dissimilar materials and, therefore, are not susceptible to PWSCC. As such, an inlay of Alloy 52/152, for the mitigation of PWSCC, is not applied to the Point Beach Unit 1 Steam Generator nozzle-to-pipe welds.

References:

1. Westinghouse Report, WCAP-14439-P/NP, Revision 4, "Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.7.1-1 L-2021 -144 Attachment 35 Page 2 of 2 Point Beach Nuclear Plant Units 1 and 2 for the Subsequent License Renewal Program (80 Years)," June 2020.

2. Westinghouse Drawing, 1184J63, Revision 4, "Steam Generator Model '44 F' Channel Head Clad & Machine," (three sheets), July 1983. (Westinghouse Proprietary)

3. Westinghouse Process Specifications; 82121 PL, "Stainless Steel (P-8) Submerged Arc (SAW) Buttering on Carbon (P-1) and Low Alloy (P-3) Steels," March 1981. (Westinghouse Proprietary) 82121 JH, "Welding Nuclear Power Components to ASME Section Ill Stainless Steel Buttering on Carbon Steel or Low Alloy Steel, Manual and Automatic Methods,"

October 1977. (Westinghouse Proprietary)

Associated SLRA Revisions:

None.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAJ No. 4.7.6-1 L-2021 -144 Attachment 36 Page 1 of 5

17. SLRA Section 4. 7.6, "Crane Cycle Load Limit" Regulatory Basis:

10 CFR 54.21 (c)(1 )(i) requires an applicant to demonstrate time-limited aging analyses remain valid for the period of extended operation.

RAI 4.7.6-1

Background:

In Section 15.4.2, "Fatigue, under title "Crane Load Cycle Limit, of the UFSAR, the applicant states "Cranes designed in accordance with CMAA-70 Class "A" service are designed from 20,000 to 200,000 load cycles." In SLRA Section 4.7.6 and Appendix A, Section 16.3.7.6, the applicant states, "Table 2.8-1 of CMAA Specification 70 states that a range of load cycles from 20,000 to 100,000 was considered for cranes in Service Class A service... " And, in Section 4.7.6, "Crane Load Cycle Limit, of FPLCORP00036-REPT-038, Revision 0, the applicant determined the service life of the cranes as 100,000 load cycles based on CMAA Specification# 70-1975.

Issue:

Based on the apparent discrepancy between the load cycle limit identified in the SLRA and in Section 15.4.2 of the UFSAR, it is unclear to the staff which load cycle limit is correct.

Request:

Clarify whether this discrepancy needs to be reconciled, and if so, identify the correct crane load cycle limit for the TLAA and update the impacted documents accordingly.

NEPB Response:

The correct load cycle limit for PBN cranes that are in the scope of subsequent license renewal (SLR) is 20,000 to 200,000 load cycles. This load cycle limit is consistent with the limit stated in PBN UFSAR Section 15.4.2 and Section 4.3.13 of NUREG-1839 "Safety Evaluation Report Related to the License Renewal of the Point Beach Nuclear Plant, Units 1 and 2." PBN SLRA 4.7.6 and Appendix A, Section 16.3.7.6 are revised accordingly.

References:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.7.6-1 L-2021-144 Attachment 36 Page 2 of 5 Associated SLRA Revisions:

SLRA Section 4.7.6, pages 4.7-10 through 4.7-12, are revised as follows:

Method of Evaluation - Acceptance Criteria All PBN cranes were originally designed and constructed to meet the requirements of Specification 61 of the Electric Overhead Crane Institute (Reference 4.8.44).

EOCl-61 did not require a specific fatigue analysis for cranes. NUREG-0612 requires that the design of heavy load overhead handling systems meet the intent of CMAA-70. As stated in Section 4.3.13 of NUREG-1839, CMAA Specification 70 and ANSI Standard 830.2-1976 (Reference 1.6.44) apply to the PBN containment polar cranes, auxiliary building cranes, and turbine building crane. In accordance with PBN UFSAR Section 15.4.2 and Section 4.3.13 of NUREG-1839, the current licensing basis (CLB) load cycle limit for PBN cranes that are in the scope of subsequent license renewal (SLR) is 20,000 to 200,000 load cycles.

CMAA SpeGifiGation 70 presents the bounding Gombinations of the number of load GyGles and mean effeGtive load faGtors for eaGh serviGe Glass. These define the aGGeptable serviGe limits for the TLI\\/\\. The following paragraph desGribes the method of seleGting the serviGe Glass from CM/\\A SpeGifiGation 70 that Gorresponds to the serviGe Glass originally speGified from EOCI SpeGifiGation 61 This serviGe Glass is used with CM,t\\,t\\ SpeGifiGation 70, Table 2.8 1 to identify the appliGable A-1::1ffiBer of load GyGles for that speGifiG serviGe Glass.

A.wendix A of EOCI SpeGifiGation 61 defines Class/\\, as:

"Standby serviGe: For suGh use as powerhouse, pump rooms, motor rooms, transformer repair, etG. 'Nhere the Grane is used very infrequently. These Granes must be substantially designed to handle expensive loads."

The Gorresponding serviGe Glass stated in SeGtion 70 2 of CMAA SpeGifiGation 70 is Class/\\ serviGe, whiGh is defined as the folio~

"Standby or Infrequent serviGe: This serviGe Glass Govers Granes whiGh may be used in installations suGh as powerhouses, publiG utilities, turbine rooms, motor rooms and transformer stations where preGise handling of equipment at slow speeds with long, idle periods between lifts required. CapaGity loads may be handled for initial installation of equipment and for infrequent maintenanGe."

Based on the Gomparison of serviGe Glasses desGribed in the original design SfleGifiGation (EOCI SpeGifiGation 61) to CM/I.A SpeGifiGation 70, the appliGable sePJiGe Glass for the PBN Gontainment polar Granes, auxiliary building Granes, and turbine building Grane is Class A.

Table 2.8 1 of CM/\\/\\ SpeGifiGation 70 states that a range of load GyGles from 20,000 to 100,000 vvas Gonsidered for Granes in ServiGe Class A serviGe thus establishing the envelope for the aGGeptable number of load GyGles for this TLAA.

The 80-year total projected load cycles for the PBN containment polar cranes,

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.7.6-1 L-2021-144 Attachment 36 Page 3 of 5 auxiliary building cranes, and turbine building crane based on past and future use are summarized in Table 4.7.6-1.

Crane Table 4.7.6-1 Evaluation Summary of Crane Operation CMAA Maximum Projected Number Service Number of Load of Load Cycles for Class Cycles 80 years Valid for 80 years Containment Polar Class A 4-£00,000 96,000 Yes Cranes Auxiliary Building Crane Class A 4-£00,000 8,384 Yes Turbine Building Crane Class A 4-£00,000 Note 1 Yes Note 1: as stated in Section 4.3.13 of NUREG-1839, the primary auxiliary building (PAB) crane is the most limiting for rated load lifts, while the containment crane is most limiting for partial load lifts. Therefore, both bound the load lifts for the turbine building crane.

Containment Polar Crane Evaluation Section 4.3.13 of NUREG-1839 provides the basis for the number of lifts the containment polar cranes would be subjected to during the 60-year period of extended operation (PEO). This total number of lifts assumed 60 outages, with 20 days of lifting per outage, and a total of 40 lifts per day. This equates to a total number of 48,000 lifts. Conservatively doubling this lift total for the 80-year SPEO equals 96,000 lifts which is less than the 4-£00,000 lift limit for PBN cranes Service Class A in CM/\\!\\ Specification 70; therefore, the TLAA for the containment polar crane remains valid.

Auxiliary Building Crane Evaluation Section 4.3.13 of NUREG-1839 provides the basis for the number of lifts the auxiliary building crane would be subjected to during the 60-year period of extended operation (PEO). This total number of lifts assumed 2700 fuel cask lifts (NUHOMS),

600 maintenance load lifts, and 892 original fuel cask lifts (VSC-24). This equates to a total number of 4, 192 lifts. Conservatively doubling this lift total for the 80-year SPEO equals 8,384 which is significantly less than the 4-£00,000 lift limit for PBN cranes Service Class /I, in CMA/I, Specification 70; therefore, the TLAA for the containment polar crane remains valid.

TLAA Disposition: 10 CFR 54.21 (c)(1 )(i)

The containment polar cranes, auxiliary building crane, and turbine building crane load cycle evaluation has been demonstrated to remain valid through the SPEO.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.7.6-1 L-2021-144 Attachment 36 Page 4 of 5 SLRA Appendix A, Section 16.3.7.6, page A-63, is revised as follows:

16.3.7.5 Crane Load Cycle Limits A review of design specifications for cranes within the scope of SLR was performed to identify those cranes that were designed or otherwise required to meet the intent of the Crane Manufacturers Association of America (CMAA) Specification 70-4-9+-a and, therefore, have defined service life as measured in load cycles.

The cranes potentially subject to TLAA are those in compliance with NUREG-0612. As documented in UFSAR Appendix A.3, the following cranes comply with NUREG-0612 and are included in SLR scope:

Containment polar cranes Auxiliary building crane Turbine building crane All PBN cranes were originally designed and constructed to meet the requirements of Specification 61 of the Electric Overhead Crane Institute.

EOCl-61 did not require a specific fatigue analysis for cranes. NUREG-0612 requires that the design of heavy load overhead handling systems meet the intent of CMAA-70. As stated in Section 4.3.13 of NUREG-1839, CMAA Specification 70 and ANSI Standard 830.2-1976 apply to the PBN containment polar cranes, auxiliary building cranes, and turbine building crane. In accordance with PBN UFSAR Section 15.4.2 and Section 4.3.13 of NUREG-1839, the current licensing basis (CLB) load cycle limit for PBN cranes that are in the scope of subsequent license renewal (SLR) is 20,000 to 200,000 load cycles.

CM/\\/\\ Specification 70 presents the bounding combinations of the number of load cycles and mean effective load factors for each service class. These define the acceptable service limits for the TLA/\\. Based on the comparison of service classes described in the original PBN crane design specification EOCI Specification 61 to CM/\\/\\ Specification 70, the applicable service class for the PBN containment polar cranes, auxiliary building cranes, and turbine building crane is Class A. Table 2.8 1 of CM/\\A Specification 70 states that a range of load cycles from 20,000 to 100,000 was considered for cranes in Service Class/\\ service thus establishing the

~e for the acceptable number of load cycles for this TU\\/\\.

The total projected 80-year load cycles for the PBN cranes within the scope of this TLAA are as follows:

Containment polar cranes - 96,000 load cycles Auxiliary building crane - 8,384 load cycles Both of these 80-year load cycles are less than the 4-£00,000 load cycle limit specified for PBN cranes Class A service. Note that as stated in Section 4.3.13 of NUREG-1839, the primary auxiliary building cranes and the containment cranes are limiting for load lifts. Therefore, they bound the load lifts for the turbine building crane.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 4.7.6-1 L-2021-144 Attachment 36 Page 5 of 5 The containment polar cranes, auxiliary building cranes, and turbine building crane load cycles remain valid through the SPEO in accordance with the requirements of 10 CFR 54.21 ( c)(1 )(i).

Associated

Enclosures:

None.