L-2021-129, Ssubsequent License Renewal Application: Aging Management Requests for Confirmation Of/Additional Information (Rci/Rai) Set 1 Responses

Ssubsequent License Renewal Application: Aging Management Requests for Confirmation Of/Additional Information (Rci/Rai) Set 1 Responses

ML21189A173
Person / Time
Site:	Point Beach
Issue date:	07/08/2021
From:	Maher W Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML21189A172	List: L-2021-129, Ssubsequent License Renewal Application: Aging Management Requests for Confirmation Of/Additional Information (Rci/Rai) Set 1 Responses
References
L-2021-129
Download: ML21189A173 (63)

Text

ATIACHMENT HP CONTAINS INFORMATION REQUESTED TO BE WITHHELD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390.

NEXTera UPON SEPARATION FROM ATIACHMENT HP, THIS LEITER IS DECONTROLLED.

ENERGY ~

POINT BEACH July 8, 2021 L-2021 -129 10 CFR 54.17 U.S. Nuclear Regulatory Commission Attention: Document Control Desk 11545 Rockville Pike One \Vhite Flint North Rockville, MD 20852-27 46 Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 SUBSEQUENT LICENSE RENEWAL APPLICATION -AGING MANAGEMENT REQUESTS FOR CONFIRMATION OF/ ADDITIONAL INFORMATION (RCI/RAI) SET 1 RESPONSES

References:

1. N extEra 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 Commission (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. Iv1L21006A417)

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. Iv1L21148A116)

5. NRC Email and Attachment dated June 10, 2021, Point Beach SLRA RAI-RCis Set 1 (Final) - Aging Management of Irradiated Concrete and Steel Reactor Vessel Support (ADAMS Accession Nos.

ML21162A003, lvIL21161Al 19)

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 suffic;:ient for docketing (Reference 2), and on January 15, 2021 issued the regulatory 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 RCls and RAls to NEPB (Reference 5). The attachments to this letter provide responses to those information requests.

For ease of reference, the index of attached and enclosed information is provided on page 3 of this letter. The proprietary (non-public) attachment, Attachment 11P ('P' denotes 'proprietary'), is inserted after the last non-Nex!Era Energy Point Beach, LLC 6610 Nuclear Road, Two Rivers, WI 54241

ATIACHMENT 11P CONTAINS INFORMATION REQUESTED TO BE Document Control Desk WITHHELD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2 .390. UPON L-2021-129 Page 2 SEPARATION FROM ATIACHMENT 11P, Tms LETIER IS DECONTROLLED.

propriet:uy (public) attachment, Attachment 13. Certain attachments include associated revisions to the SLRA (Enclosure 3Attachment1 of Reference 1) denoted by strikethrough (deletion) and/or bold red underline (insertion) text. 1P contains information proprietary to Westinghouse Electric Company LLC

(Westinghouse"); it is supported by an Affidavit signed by Westinghouse (Attachments 11 and 11P Enclosure), the owner of the information. The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the Nuclear Regulat01y Conunission ("Conunission") and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations.

Accordingly, it is respectfully requested that the information which is proprietru.y to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Conunission's regulations.

Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse Affidavit should reference CA W-21-5198 and should be addressed to Jill S.

Monahan, Manager, eVinci M,odeling and Analysis, Westinghouse Electric Company, 1000 Westinghouse Drive, Cranbeny Township, Pennsylvania 16066.

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 81" day of July 2021.

Sincerely, William D. Maher Licensing Director - Nuclear Licensing Projects Cc (w / o Attachment 11 P)

Administrator, Region III, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC Public Service Conunission Wisconsin

ATIACHMENT UP CONTAINS INFORMATION REQUESTED TO BE Document Control Desk \VITHHELD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390. UPON L-2021-129 Page 3 SEPARATION FROM ATIACHMENT UP, THIS LEITER IS DECONTROLLED.

Attachments Index A ttachment RAI/ RCI No . Subj ect N o.

1 RAI 3.5.2.2.2.6-1 Uncertainty of Neutron Fluence and Gamma Dose Projections on the Biological Shield Wall 2 RAI 3.5.2.2.2.6-2 Thermal Insulation on the Reactor Coolant Piping Penetrating the Primary Shield Wall 3 RAI 3.5.2.2.2.6-3 Radiation Induced Volumetric Expansion Effects on the Biological Shield Wall Liner 4 RAI 3.5.2.2.2.6-4 Aging Management of Loss of Fracture Toughness of the Biological Shield Wall Liner 5 RAI 3.5.2.2.2.6-5 Temperature Assumptions Associated with Previous Structural Analysis 6 RAI 3.5.2.2.2.7-1 Uncertainty of Neutron Fluence and Gamma Dose Projections on the Reactor Vessel Supports 7 RAI 3.5.2.2.2.7-2 Interaction Ratio Methodology 8 RAI 3.5.2.2.2.7-3 Clarifications on Interaction Ratio Calculations 9 RAI 3.5.2.2.2.7-4 Flaw Assumptions Regarding Interaction Ratio Calculations 10 RAI 3.5.2.2.2.7-5 Use of ASME Section XI, Subsection I\\!F for Aging Management of Irradiation Effects 11, 11P RAI 3.5.2.2.2.7-6 Temperature Assumptions for the Critical Flaw Size Calculations 12 RCI 3.5.2.2.2.7-1 Confirm Conservatisms in the Analysis of Certain Reactor Vessel Support Components 13 RCI 3.5.2.2.2.7-2 Confirm the Applicable Version of the AISC Code Enclosures In dex Attachment E nclosure No. Subject N o.

11, 11P -- Westinghouse Affidavit CA W-21 -5198 Executed July 2, 2021, Application for Withholding Proprietary Information from Public Disclosure

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-1 L-2021-129Attachment 1Page1 of5 Regulatory Basis:

Title 10 of the Code of Federal Regulations (CFR) Section 54.21 (a)(1) requires license renewal applicants to perform an integrated plant assessment (IPA) and their application to identify and list systems, structures, and components (SSCs) that are within the scope of license renewal and subject to aging management review (AMR).

Further, 10 CFR 54.21 (a)(3) requires for the SSCs identified to be subject to AMR, the applicant demonstrate that the effects of aging will be adequately managed such that their intended functions are maintained consistent with the current licensing basis (CLB) for the subsequent period of extended operation. To complete its review and enable the staff to make a reasonable assurance finding on functionality of reviewed SSCs for the subsequent period of extended operation consistent with 10 CFR 54.21, the staff requires under 10 CFR 54 .29(a) additional information be provided regarding the matters described below.

1. SLRA Section 3.5.2.2.2.6, "Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation" RAI 3.5.2.2.2.6-1

Background:

Subsequent License Renewal Application (SLRA) Section 3.5.2.2.2.6 of the Point Beach Nuclear Plant, Units 1 and 2 (PBN), NextEra Energy Point Beach, LLC (NextEra or the applicant), submitted November 16, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20329A264), presents an evaluation of the irradiation effects of the biological shield wall (BSW) for the specified period of extended operation to ensure it will maintain structural integrity and not affect the primary shield wall under design basis loading conditions. However, NRC staff observed that this evaluation is based on neutron fluence and gamma dose results for which the uncertainty has not been assessed. The effect of neutron fluence and gamma dose on the BSW concrete may in fact be greater with consideration of this uncertainty. While no method, generic or specific to PBN, has been approved by the NRC for calculations of exposure for the BSW and primary shield wall (PSW) concrete, the calculations for neutron fluence and gamma dose have generally been found acceptable in prior reviews on the basis that the uncertainty in the calculations necessary for the results to exceed the exposure levels of concern identified in NUREG-2192, "Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants," (SRP-SLR) is substantial (e .g., 200%). In the present evaluation, the neutron fluence and gamma dose for concrete already exceed the NUREG-2192 damage thresholds. The staff is not able to determine whether reasonable assurance exists that the limiting neutron fluence and gamma dose values for concrete were identified with sufficient margin and conservatism to accommodate uncertainties in the fluence analysis methodology associated with calculating exposure at an ex-vessel location.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-1 L-2021-129 Attachment 1 Page 2 of 5 Issue:

In order to ensure the reduction of strength and mechanical properties of concrete due to irradiation are adequately managed, it is necessary to assess the uncertainty associated with the neutron fluence and gamma dose results .

Request:

Provide an estimate of the uncertainty associated with the neutron fluence and gamma dose results at the surface of the BSW.

NEPB Response:

An analytical uncertainty analysis associated with the neutron fluence and gamma dose results at the inner surface of the BSW was not performed for the Point Beach SLRA.

Therefore, a conservative estimate of the uncertainty associated with these results was established using an existing reactor pressure vessel (RPV) extended beltline uncertainty analysis .

The existing RPV extended beltline analysis quantified the analytical uncertainty associated with calculated fast neutron (E > 1.0 MeV) fluence rates at the RPV inner and outer surfaces at various elevations above and below the active fuel. As part of this analysis, numerous parameters that were identified as having a potentially significant contribution to the core neutron source, reactor geometry, coolant temperature, discretization, and modeling approximation uncertainties at the RPV inner and outer surfaces were evaluated. More specifically, each parameter identified was evaluated on an individual basis by determining the maximum relative change in the base-case fluence rate that occurred as the magnitude of that parameter was varied over a bounding range of values . The net analytical uncertainty associated with a given RPV location was then determined by taking the root sum of squares of the individual parameter uncertainty values determined at that location. Given the parameters considered, the magnitudes of the parameter variations evaluated, and the relative proximity of the RPV outer surface to the BSW, the extended beltline uncertainty analysis results for the RPV outer surface were judged to provide a reasonable basis for estimating the analytical uncertainty associated with the BSW neutron and gamma exposures.

The maximum neutron fluence and gamma dose projections at the inner surface of the BSW occur at elevations that are slightly above the core midplane. However, since the extended beltline uncertainty analysis was, by design, focused on the RPV extended beltline region only, it did not consider axial elevations slightly above the core midplane; the elevations nearest the midplane considered were 30 cm above the top and 30 cm below the bottom of the active fuel. Therefore, the extended beltline uncertainty analysis results determined at the RPV outer surface 30 cm above the top of the active fuel were used as the starting point for estimating the uncertainty associated with the BSW neutron and gamma exposures. This is conservative because analytical uncertainties increase with axial distance above the top of the active fuel.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-1 L-2021-129 Attachment 1 Page 3 of 5 In addition to using this bounding RPV location as a starting point, the concrete composition parameter uncertainty value determined at this location was increased by a factor of 2. This value was increased because it was associated with the one parameter evaluated in the RPV extended beltline analysis whose uncertainty was judged to be potentially impacted in a non-negligible manner if a detailed uncertainty analysis for the BSW were performed. Note that the standard concrete composition from the BUGLE-96 documentation (Reference 2) was used for the BSW in both the extended beltline analytical uncertainty analysis base-case calculations and the Point Beach neutron fluence and gamma dose calculations.

Following the process described above, the analytical uncertainty associated with the neutron fluence and gamma dose results at the inner surface of the BSW was conservatively estimated to be 20%. For the following reasons, applying this uncertainty in the limiting direction to the maximum exposures determined at the inner surface of the BSW would not invalidate the results of the BSW structural integrity evaluation summarized in Section 3.5 .2.2.2.6 of Enclosure 3 of the Point Beach SLRA:

• The neutron and gamma exposures used in the BSW structural integrity evaluation were the maximum values determined anywhere on the inner surface of the BSW. Significantly smaller exposures will occur over the majority of the BSW inner surface. For example, the following exposures from LTR-REA-20-28-NP (Reference 3) show the changes in fast neutron (E > 0.1 MeV) fluence and gamma dose exposures that occur as a function of azimuthal angle.

Maximum Neutron and Gamma Exposures at the BSW Inner Surface Time Fast Neutron (E > 0.1 MeV) Fluence (n/cm 2 )

(EFPY) 15° 30° 60° 75° 90° Max 72 3.95E+19 2.81E+19 2.79E+19 3.93E+19 5.23E+19 5.23E+19 Time Gamma Dose (Gy)

(EFPY) 15° 30° 60° 75° 90° Max 72 1.80E+08 1.37E+08 1.43E+08 1.90E+08 2.39E+08 2.39E+08 Reductions in neutron and gamma exposures also occur over axial heights corresponding to those of the active fuel. For example, a sensitivity calculation performed by Westinghouse shows that fast neutron (E > 0.1 MeV) fluence rates at the BSW inner surface would be expected to be approximately 20%

less than their maximum value within 4.5 feet of the core midplane and approximately 50% less than their maximum value within 5.5 feet of the core mid plane.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No . 3.5.2 .2.2.6-1 L-2021-129 Attachment 1 Page 4 of 5 Section 3.5.2.2.2.6 of Enclosure 3 of the Point Beach SLRA documents that the BSW structural integrity evaluation assumed that the maximum irradiation effects applied to the entire vertical surface of the BSW corresponding to the active fuel region.

• Neutron and gamma exposure projections for Point Beach Unit 1 were based on the core power distributions and operating conditions of Unit 1 Cycle 37 (U 1C37) but included a +10% bias on the peripheral and re-entrant corner assembly relative powers; projections for Unit 2 were based on Unit 2 Cycle 38 (U2C38) and also included a +10% bias on the peripheral and re-entrant corner assembly relative powers. A +10% bias applied to the peripheral and re-entrant corner assembly relative powers of a given cycle increases the calculated exposure rates for that cycle by approximately 10%. While the biases used for the Point Beach projection cycles were intended to account for changes in future-cycle exposure rates that are expected to occur as a result of normal variations in future-cycle core designs, they also provide a significant source of margin in the 72-EFPY exposures determined for the SLRA. This is because future cycle designs at Point Beach Unit 1 would not be expected to consistently result in exposure rates 10% greater than the ones determined for U1 C37, and future cycle designs at Point Beach Unit 2 would not be expected to consistently result in exposure rates 10% greater than the ones determined for U2C38 . This conclusion is supported by the RPV neutron exposure rates reported in LTR-REA-20-27-NP, Revision 1 (Reference 4) for Unit 1 Cycles 34-39 and LTR-REA-20-28-NP (Reference 3) for Unit 2 Cycles 32-38 , which show normal cycle-to-cycle variations , but no long-term trend in either the limiting (increasing) or non-limiting (decreasing) direction.

• The depth of concrete needed to reduce the maximum calculated fast neutron (E > 0.1 MeV) fluence at the BSW inner surface of 5.23E+19 n/cm 2 plus 20%

uncertainty to the NUREG-2192 damage threshold would not be significantly different than the 3.35-inch depth reported in Section 3.5 .2.2.2.6 of Enclosure 3 of the Point Beach SLRA. For example, using the approach described in Section 3.5.2.2.2.6 of Enclosure 3 of the Point Beach SLRA along with an attenuation ratio of 1 I (1.2 x 5.23) = 0.16, a fast neutron (E > 0.1 MeV) fluence of 1.2 x 5.23E+19 n/cm 2 would reach the NUREG-2192 damage threshold at a depth of approximately 3.66 inches into the 38-inch thick BSW.

Given the rate at which neutrons are attenuated in concrete, a similarly small increase in the 3.92-inch depth of concrete reported in Section 3.5.2.2.2.6 of Enclosure 3 of the Point Beach SLRA as being subjected to radiation-induced volumetric expansion (RIVE) effects would be expected.

• Using the approach described in Section 3.5.2.2.2.6 of Enclosure 3 of the Point Beach SLRA, the depth of concrete needed to reduce the maximum calculated gamma dose at the BSW inner surface of 2.39E+10 rad plus 20%

analytical uncertainty to the NUREG-2192 damage threshold would be approximately 28.5 inches. This is approximately 4.5 inches greater than the

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-1 L-2021-129Attachment1Page5of5 24-inch depth of concrete evaluated in the BSW structural integrity analysis as undergoing a gamma-induced 20% reduction in strength. In reality, however, the full 20% reduction in strength would only occur near the BSW inner surface; it would decrease with concrete depth in an approximately linear fashion until the gamma dose was attenuated below the NUREG-2192 threshold. The conservative manner in which gamma-induced effects were treated in the BSW structural integrity evaluation is considered to offset the additional depth of concrete that would be needed to attenuate a 20%

increase to the maximum calculated gamma dose at the BSW inner surface.

References:

1. USNRC Request for Additional Information, "Point Beach Nuclear Plant, Units 1 and 2 (PBN) Subsequent License Renewal Application (SLRA) Requests for Additional Information (RAls) and Requests for Confirmation of Information (RCIS) Safety -

Set 1 (Aging Management of Irradiated Concrete and Steel Reactor Vessel Supports)," June 2021. (ADAMS Accession No. ML21161A119).

2. RSICC Data Library Collection DLC-185, "BUGLE-96, Coupled 4 7 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," Radiation Shielding Information Computational Center, Oak Ridge National Laboratory (ORNL), July 1999.

3. LTR-REA-20-28-NP, Revision 0, "Reactor Vessel, Reactor Vessel Supports, and Concrete Bioshield Exposure Data in Support of the Point Beach Unit 2 Subsequent License Renewal (SLR) Time-Limited Aging Analysis (TLAA)," July 2020.

4. LTR-REA-20-27-NP, Revision 1, "Reactor Vessel, Reactor Vessel Supports, and Concrete Bioshield Exposure Data in Support of the Point Beach Unit 1 Subsequent License Renewal (SLR) Time-Limited Aging Analysis (TLAA)," May 2020.

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. 3.5.2.2.2.6-2 L-2021-129 Attachment 2 Page 1 of 13 RAI 3.5.2.2.2.6-2

Background:

SLRA Section 3.5.2.2.2.6 (SLRA Supplement 1, dated April 21, 2021 (ADAMS Accession No. ML21111A155)), states that the "reactor coolant piping which penetrates the PSW is insulated to ensure ambient temperatures remain within design limits."

Information Notice (IN) 2007-21, Supplement 1, "Pipe Wear due to Interaction of Flow-Induced Vibration and Reflective Metal Insulation," states that the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code contains no specific requirements for licensees to remove insulation periodically for visual inspections assessing the integrity of stainless steel reflective metal insulation (RMI) and wear of encapsulated piping. The IN also states that this type of wear, if unchecked , could result in a small break loss of coolant accident and challenge the plant emergency core cooling systems.

To maintain the integrity of insulation, NUREG-2191, "Generic Aging Lessons Learned

- Subsequent License Renewal (GALL-SLR) Report,) Aging Management Program (AMP) Xl.M36, "External Surfaces Monitoring of Mechanical Components," provides measures for monitoring, inspecting, and detecting age-related degradation of stainless steel or aluminum insulation at a frequency not to exceed one refueling cycle. Section B.2.3.23, "External Surfaces Monitoring of Mechanical Components," of the PBN SLRA states that the program monitors the reduction in thermal insulation resistance and evidence of insulation damage.

Issue:

The staff reviewed item 3.5-1, 048 of SLRA Table 3.5-1, "Containment Building Structure and Internal Structural Components - Summary of Aging Management Programs," which states:

There have been no instances of elevated temperatures for PBN plant structures other than containment (which is addressed in item 3.5-1, 003 and Section 3.5.2.2.1.2). In addition, insulation for high-temperature piping (> 200°F) is in scope to assist in maintaining local primary auxiliary building and turbine building concrete temperatures and is managed by the External Surfaces Monitoring of Mechanical Components (B.2.3.23) AMP.

The staff reviewed Item 3.5-1, 003 of SLRA Table 3.5-1, "Containment Building Structure and Internal Structural Components - Summary of Aging Management Programs," and noted that it addresses temperatures of containment penetrations, including the insulated "Main Steam and Feedwater penetrations that experienced elevated temperatures prior to initial license renewal." The staff also reviewed SLRA Table 3.1-1 , "Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System," and noted that "non-metallic thermal insulation associated with reactor coolant piping and piping components does not perform a SLR intended function and is therefore not in scope." The staff further reviewed the PBN "External Surfaces Monitoring of Mechanical Components," program, but it is not clear

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5 .2.2.2.6-2 L-2021-129 Attachment 2 Page 2of13 whether its "scope of program, " program element includes the reactor coolant piping thermal insulation. The aforementioned do not discuss how aging effects of the reactor coolant piping insulation that penetrates the PSW to ensure ambient temperatures remain within design limits are to be managed .

The staff is also not clear what type (e.g ., rigid, flexible, made of stainless steel or other material) of thermal insulation PBN has used where the reactor coolant piping penetrates the primary shield wall (PSW) and whether such insulation aligns with the RMI description provided in Information Notice 2007-21. In addition, the staff is not clear whether the frequency of its inspection is consistent with the guidance provided in the "detection of aging effects" program element of GALL-SLR AMP Xl.M36.

Request:

a. Clarify the type of thermal insulation used on the reactor coolant piping in areas penetrating the PSW and its adequacy of protecting the PSW concrete from potential exposures to radiation and abnormal temperatures.

b. If it is a RMI and IN 2007-21 was applicable to PBN , discuss the IN inspection results, identified problems, and actions taken.

c. Clarify whether the PBN "External Surfaces Monitoring of Mechanical Components,"

program includes in its "scope of program" program element this particular reactor coolant piping insulation.

d. If so, discuss whether accessibility and inspectability of the reactor coolant piping insulation is consistent with gu idance provided in GALL-SLR AMP Xl.M36, "detection of aging effects," other applicable program elements, or other applicable PBN programs, so that it can fulfill its intended function (i.e., protection of the PSW concrete to abnormal temperature exposure) during the period of extended operation.

e. Augment SLRA Section 3.5.2.2.2 .6 to indicate that the effects of aging for the encapsulating insulation to the reactor coolant piping that penetrates the PSW are managed by the "External Surfaces Monitoring of Mechan ical Components" program so that the surrounding PSW concrete ambient temperature remains within design limits.

NEPB Response:

The letters associated with the responses below correspond to the letters in the RAI above .

a. The types of insulation installed on the reactor coolant (RC) piping passing through the PSWs at PBN are a combination of fiberglass blanket insulation with stainless steel jacketing and stainless steel reflective metal insulation (RMI). The design specifications for both types of insulation are that with a RC piping external surface temperature of 650°F that the outside surface temperature of the insulation will be :5 150°F. Additionally, normal containment ambient temperatures are 50 to 105°F with

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 .6-2 L-2021-129 Attachment 2 Page 3 of 13 a non-accident design maximum of 120°F. Therefore, based on the PBN insulation design, the PSW concrete is protected from potential exposures due to abnormal temperatures. Note that the insulation specifications have been uploaded to the ePortal.

With regard to protection from radiation exposures, the PSW concrete surround ing the RC piping passing through the PSW is located away from the source of high radiation exposures. Thus, the projected fluence and gamma dose incident on this concrete at the end of the SPEO will be well below the damage thresholds in NUREG-2192 .

b. Information notice IN-2007-21 issued in 2007 and Supplement 1 to the notice issued in 2020 describe incidents of piping wear due to RMI. The causal factors to this pipe wear were a combination of the use of end caps to clip short segments of RMI insulation together (primarily used on smaller piping) and flow induced piping vibration. When there is relative motion between the piping and the edge of the end cap , pipe wear has occurred . A review was performed to determine the applicability of IN-2007-21 and supplement to the RMI insulation installed on the RC piping passing through the PSWs at PBN . Based on this review , the information notice and supplement were determined to be not applicable to the RMI insulation installed on the RC piping passing through the PSWs at PBN because of the following:

• The size of the piping (hot legs, 34-inch outside diameter and cold legs, 32%-inch outside diameter) and the large insulation segments used where end caps are not required , and

• The minimal flow induced vibration of the reactor coolant loops and the limited relative motion between the insulation and piping due the pipe size (piping and insulation move together).

c. The stainless steel jacketing and RMI insulation installed on the RC piping passing through the PSWs at PBN are included in the scope of the External Surfaces Monitoring of Mechanical Components aging management program (SLRA Section B.2.3.23) . See the additional clarifications included in the Associated SLRA Revisions section below addressing aging management review and aging management program changes.

d. Based on the tight clearances between the insulation and the surface of the PSW inside the penetrations , the guidance provided in element 4, "Detection of Aging Effects", of GALL-SLR AMP Xl.M36 ind icates the following :

"Surfaces that are not readily visible during plant operations and refueling outages are inspected when they are made accessible and at such intervals that would ensure the components' intended functions are maintained. "

This insulation is installed in environment with no source of external moisture based on its location within the PSW. Additionally, the normal operating temperatures of the RC piping are well above the dew point eliminating the need for insulation

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-2 L-2021-129 Attachment 2 Page 4of13 removal. Consistent with the guidance provided in GALL-SLR AMP Xl.M36, visual inspections for loss of material (every refueling outage) and cracking (every ten years) on accessible portions of the RC piping RMI insulation and insulation jacketing will be performed as part of the PBN External Surfaces Monitoring of Mechanical Components aging management program (SLRA Section B.2.3.23).

These inspections will provide an assessment of the condition of the insulation inside the PSW penetrations. See additional enhancement on page 13 below.

e. With regard to augmentation of SLRA Section 3.5.2.2.2.6, see the additional clarifications included in the Associated SLRA Revisions section below including a reference to the PBN External Surfaces Monitoring of Mechanical Components in SLRA Section 3.5.2.2.2 .6.

References:

None.

Associated SLRA Revisions:

SLRA Section 2.1 .5-1, page 2.1-27 is amended as indicated by the following text addition:

With regard to thermal insulation on mechanical components, a screening review was performed as part of the original PBN license renewal project. The review identified only two locations where piping thermal insulation was considered to be in scope of LR. Insulation is installed on the main steam and main feedwater piping at the containment penetrations, and is needed to maintain steady-state concrete temperatures less than 150 degrees F. Therefore, thermal insulation for the main steam and feedwater penetrations is included in the scope of SLR and is addressed in Section 2.4 . As part of the SLR review, additional locations were identified where the installed piping insulation needs to be credited in order to maintain steady-state concrete temperatures less than 150°F. These locations are where the reactor coolant loop piping passes through the primary shield wall.

Accordingly, this insulation is also included in the scope of SLR.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-2 L-2021-129 Attachment 2 Page 5of13 SLRA Table 2.4-1, page 2.4-6, as revised by SLRA Aging Management Supplement 1, is amended as indicated by the following text addition (the text deletion was from SLRA Aging Management Supplement 1):

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 support Penetration sleeves (Mechanical) Pressure boundary Structural support Pressure-retaining bolting Pressure boundary Structural support Primary shield wall (and biological shield wall) Radiation shielding Shelter, protection Structural support

~aEliaAt eAeF§Y sl=lie!Gs i;;:iFe eaFFieF RC Class 1 supports Structural support RC Class 1 support bolting Structural support Reactor cavity seal ring Pressure boundary Refueling components (containment upender, davit arm) Structural support Service Level I coatings Maintain adhesion Sliding surfaces Structural support Tendons (post-tensioning system) Structural support Tendon anchorage and attachments Pressure boundary Structural support Thermal Insulation (high temperature penetrations and Insulate (thermal) reactor coolant giging gassing through the PSW) Insulation jacket integrity

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAJ No. 3.5.2.2 .2.6-2 L-2021-129 Attachment 2 Page 6of13 SLRA Section 3.5.2.1.1, page 3.5-2, is amended as indicated by the following text addition:

3.5.2.1.1 Containment Structure and Containment Internal Structural Components Materials The materials of construction for the Containment structure and internal structural components are:

• Calcium Silicate or Amosite Asbestos with a Silicate Binder (thermal insulation)

• Fiberglass blanket with stainless steel jacketing (thermal insulation)

• Stainless steel reflective metal insulation

• Coatings

• Concrete (reinforced)

• Copper alloy

• Dissimilar metal welds

• Elastomer

• High-strength steel

• Lubrite

• Stainless steel

• Steel (including galvanized steel)

SLRA Section 3.5.2.2.1.2, page 3.5-19, third paragraph, is amended as indicated by the following text addition:

For SLR, the operating experience review did not identify a recurrence of concrete temperatures exceeding UFSAR allowable. The PBN UFSAR states in Section 5.1 .2.4 that temperature has been measured at the main steam penetrations and found to be less than 200°F and no greater than 200°F at small hot piping penetrations. Lastly, the ASME Section XI, Subsection IWE (B.2.3.29)

AMP and ASME Section XI, Subsection IWL (B.2.3.30) AMP provide management of the containment wall and penetrations. Therefore, a plant-specific AMP is not required to manage reduction of strength and modulus due to elevated temperatures for containment penetrations associated with main steam and feedwater lines.

For the reactor coolant piping passing through the primary shield wall (PSW), the insulation on the piping maintains local temperatures to~ 150°F.

The External Surfaces AMP (8.2.3.23) will manage aging of this insulation, so a plant-specific AMP is not required to manage reduction of strength and modulus due to elevated temperatures.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5 .2.2.2.6-2 L-2021-129 Attachment 2 Page 7of13 SLRA Section 3.5.2.2.2 .6, page 3.5-37, as revised by SLRA Aging Management Supplement 1, is amended as indicated by the following text addition :

Other factors regarding temperatures in the reactor cavity are as follows:

• The reactor coolant piping which penetrates passing through the PSW is insulated to ensure ambient temperatures remain within design limits. See SLRA Section 3.5.2.2.1.2 for further discussion of this insulation. The External Surfaces AMP (B.2.3.23) will manage aging of this insulation.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-2 L-2021-129 Attachment 2 Page 8of13 SLRA Table 3.5-1, page 3.5-60 , is amended as indicated by the following text deletion and addition:

Table 3.5-1 Containment Building Structure and Internal Structural Components - Summary of Aging Management Programs Item Component Ag ing Effect Aging Further Evaluation Discussion Number Requiring Management Recommended Management Program 3.1-1 , 134 Non-metallic thermal insulation Reduced thermal AMP Xl.M36 , No t>Jet aF1F1lieaele. Consistent with NUREG-exposed to air, condensation insulation resistance due "External Surfaces 2191.

to moisture intrusion Monitoring of Mechanical Ne-Ft.Non-metallic thermal Components" insulation associated with reactor coolant piping and piping components does not perform a SLR intended function and is therefore not in scope, with exception of the insulation on the reactor coolant piping passing through the primary shield wall.

See discussion in Section 3.5.2.2.1.2.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-2 L-2021-129 Attachment 2 Page 9of13 SLRA Table 3.5-1 , page 3.5-60 , is amended as indicated by the following text addition:

Table 3.5-1 Containment Building Structure and Internal Structural Components - Summary of Aging Management Prog rams Item Component Aging Effect Aging Further Evaluation Discussion Number Requiring Management Recommended Management Program

3. 5-1, 048 Groups 1-5: concrete: all Reduction of strength Plant-specific Yes (SRP-SLR Not required .

and modulus due to ag ing Section 3.5.2.2.1.2) As described in the UFSAR and consiste nt elevated temperature management with the cu rrent renewed licenses,

(>150°F general; program temperatures of containment penetrations

>200°F local) are below the allowable general and local temperature thresholds for reduction of strength and modulus by design . This includes thermal insulation of the Main Steam and Feedwater penetrations that experienced elevated temperatures prior to initial license renewal. The high-temperature Main Steam and Feedwater Penetrations are managed by the ASME Section XI , Subsection IWE (B.2.3.29) AMP (penetration assembly) and ASME Section XI , Subsection IWL (B.2.3.30) AMP (concrete). The insulation on the reactor coolant giging gassing through the grima!Y shield wall ensures grima!Y shield wall concrete temgeratures are maintained below 150°F. The External Surfaces Monitoring of Mechanical Comgonents {B.2.3.23) AMP is credited for aging management of the stainless steel insulation jacketing and reflective metal insulation as described in item 3.5-100 below.

Further evaluation is documented in Section 3.5.2.2.1.2.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-2 L-2021-129 Attachment 2 Page 10of13 SLRA Table 3.5.2-1, page 3.5-95, is amended as indicated by the following text addition:

Table 3.5.2-1: Containment Building Structure and Internal Structural Components - Summary of Aging Management Evaluation Component Type Intended Material Environment Aging Effect Aging Management NUREG-2191 Table 1 Notes Function Requiring Program Item Item Management Tendon anchorage Structural Steel Air- indoor Loss of material ASME Section XI , ll.A1 .C-10 3.5-1, 032 A and attachments support uncontrolled Subsection IWL (B.2.3.30)

Thermal insulation Insulate Calcium Air- indoor None None Vlll.H.S-403 3.4-1, 064 I, 9 (high temperature (thermal) silicate, uncontrolled penetrations) amosite asbestos with a silicate binder Thermal insulation Insulation Stainless Air- indoor Cracking External Surfaces lll.B2.T-37c 3.5-1, 100 ~

(high temQerature Jacket steel uncontrolled Loss of material Monitoring of Qenetrations} lntegritv jacketing Mechanical ComQonents (B.2.3.23}

Thermal insulation Insulate Stainless Air- indoor Cracking External Surfaces 111.82. T-37c 3.5-1, 100 ~

(reactor coolant (thermal} steel uncontrolled Loss of material Monitoring of QiQing Qassing reflective Mechanical through the Qrima!Jl metal ComQonents shield wall} insulation (B.2.3.23}

Thermal insulation Insulate Fiberglass Air- indoor None None IV.C2.R-450 3.1-1, 134 ~

(reactor coolant (thermal} blanket uncontrolled QiQing Qassing through the Qrima!Jl shield wall}

Thermal insulation Insulation Stainless Air- indoor Cracking External Surfaces lll.B2.T-37c 3.5-1, 100 ~

(reactor coolant Jacket steel uncontrolled Loss of material Monitoring of QiQing Qassing lntegrit~ jacketing Mechanical through the Qrima!)l ComQonents shield wall} (B.2.3.23}

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-2 L-2021-129 Attachment 2 Page 11 of 13 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.

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.

I. Aging effect in NUREG-2191 for this component, material and environment combination is not applicable.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-2 L-2021-129 Attachment 2 Page 12of13 SLRA Table 3.5.2-1 , page 3.5-96, is amended as indicated by the following text deletion and addition:

Plant Specific Notes

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 Fa9ade 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 Fa9ade building is non-heated.

3. The tendon gallery adjacent to each PBN Unit's containment, inside the Fa9ade building , is part of the containment basemat 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 Fa9ade 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 AMf .

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, whereas other mechanical penetration sleeves/assemblies are addressed for cumulative fatigue damage.

7. Primary shield wall , and attached biological shield wall , with a 'Y4 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, thermal 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 not subject to wetting so there are no plausible aging effects that could degrade the fiberglass, 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 Ag ing 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.

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 .6-2 L-2021-129 Attachment 2 Page 13of13 SLRA Section B.2.3.23, page B-174, after the fourth bullet, is amended as indicated by the following text addition (new bullet item):

• Revise procedure(s) to specify that visual inspections for loss of material (every refueling outage) and cracking (every ten years) on accessible portions of the RC piping RMI insulation and insulation jacketing will be performed. These inspections will provide an assessment of the condition of the insulation inside the PSW penetrations.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-3 L-2021-129 Attachment 3 Page 1 of 10 RAI 3.5.2.2.2.6-3

Background:

SLRA Section 3.5.2.2.2.6 states that both PBN Units 1 and 2 have a% inch thick steel liner plates installed at the inner face of the biological shield wall (BSW) and that the "liner plates are welded to each other and are anchored to the concrete with steel angle sections, thus enabling composite action with the concrete wall." Attachment 1, "Reactor Vessel Supports, and Concrete Bioshield Exposure Data in Support of the Point Beach Unit 2 Subsequent License Renewal (SLR) Time-Limited Aging Analysis (TLAA)," to Westinghouse LTR-REA-20-28-NP, Revision 0, dated July 31, 2020, and submitted as Attachment 1 to Enclosure 4 of the PBN SLRA. "provides select exposure data applicable to the Point Beach Unit 2 reactor vessel (RV), RV supports, concrete bioshield, and in-vessel and ex-vessel dosimetry." The data indicates that the maximum fluence for E > 0.1 MeV with a 10% bias is 5.23E+19 n/cm 2 for 72 effective full power years (EFPY). The staff notes that the maximum fluence occurs on the BSW at an azimuthal angle of 90°and at an approximate elevation of 43.0 cm above the bottom of active fuel. Table 3.5.2-1, "Containment Building Structure and Internal Structural Components - Summary of Aging Management Evaluation," of SLRA Supplement 1, includes an AMR item for loss of material, distortion of the reactor cavity liner (BSW steel liner) with a radiation shielding intended function . Neither the SLRA or its attachments , however, discuss an evaluation of the liner loss of fracture toughness, so that its radiation shielding and structural support intended functions as noted in SLRA Table 2.4-1 and Table 3.5.2-1, are maintained to the end of the subsequent period of extended operation.

Issue:

The staff, using information contained on the applicant ePortal, calculated the gap between the RV and the% inch thick steel liner to be 6.5". Based on the information provided above, the most severe radiation exposure to the BSW steel liner will occur at or about its mid-height. Given the close proximity of BSW steel liner to the RV, it is not clear whether PBN calculated the harming energy of neutron fluence in terms of dpas (displacement per atom - atoms permanently displaced from their position) at the mid-height of the steel liner and associated weldments, if any, at that location. Although PBN in SLRA Supplement 1 addressed the effects of radiation-induced volumetric expansion (RIVE) of concrete on the BSW steel liner and provided measures to identify its deformation, if any, it is not clear whether such loading coupled to the effects of streaming radiation on the % inch thick steel liner would be factors to its potential cracking. If so, it is not clear what methodology PBN has used to evaluate the liner integrity, its welds, and attachments to concrete for effects of aging due to RIVE of concrete and liner embrittlement due streaming radiation for 72 EFPY of operation.

Request:

a. Discuss whether the RIVE effects on concrete compounded by potential liner embrittlement due to streaming radiation were considered in the evaluation of BSW

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 .6-3 L-2021-129 Attachment 3 Page 2of10 steel liner integrity including its welds and attachments to concrete for projected 72 EFPY of PBN plant operation.

b. If so, discuss, the methodology used to evaluate the structural integrity of the BSW steel liner its weldments, including its anchorage to concrete, with respect to streaming radiation for the subsequent period of extended operation.

c. If not, justify why an evaluation of the BSW steel liner for loss of/reduction to fracture toughness was not necessary for the subsequent period of extended operation .

NEPB Response:

The response below addresses requests a, band c above .

First, an important clarification is required regarding the last sentence in "Background" above . The component line item "Liners (reactor cavity)" in PBN SLRA Tables 2.4-1 and Table 3.5.2-1 would be more clear as "Liner (primary shield wall)" and the SLRA is amended as noted below. As presented in Section 3.5.2.2.2.6 of the PBN SLRA, the liner on the biological shield wall (BSW) does not perform subsequent license renewal intended functions .

The RIVE effects on the BSW concrete are not compounded by liner embrittlement due to streaming radiation . As indicated on page 3.5-36 , second paragraph of the SLRA, the BSW liner plates including those covering the top and bottom of the BSW are welded together, resulting in a continuous plate structure supported on concrete with angle sections typically 2 ft apart from each other. The maximum RIVE effect zone of 3.92 inches is applicable only to a limited number of anchors around the mid-height of the BSW. The rest of the anchors remain effective. Consequently, the overall integrity of the BSW liner is not adversely affected by the RIVE effects . PBN SLRA Aging Management Supplement 1 (Reference 1) provided additional factors that confirm the BSW liner would remain in place considering the RIVE neutron fluence effects and design basis loading conditions. These factors include:

• Actual RIVE distribution: a) use of ACI code concrete strain at ultimate capacity (0.003); b) use of the EPRI normalized neutron flux profile. RIVE is very limited within 2 inches at the center allowing most of the liner anchorage to be effective.

• The structural steel liner and associated welds will continue to maintain its configuration away from the impacted area. Note the steel rebar does not lose its cover at the impacted point based on the bullet above . This rebar is mostly for shrinkage and temperature effects and will continue to perform that function.

The governing failure mode of the PSW/BSW composite system is pure tension failure in the tangential (horizontal) direction under the combination of accident pressure and thermal load . The tangential tensile load capacity of the PSW/BSW is calculated based solely on the horizontal steel reinforcement in the PSW. Any contribution to the tangential tensile capacity from the horizontal reinforcement in the BSW and BSW liner are neglected . Additionally, in addressing the load capacity of the BSW, the BSW will remain structurally sound under dead and seismic loads neglecting any structural

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-3 L-2021-129 Attachment 3 Page 3of10 benefit from the liner and associated welds. Finally, there is sufficient anchorage and load capacity in the liner and associated welds to prevent them from detaching and falling from the BSW under CLB loading conditions and 72 EFPY irradiation effects.

Therefore, irradiation effects on the BSW liner and associated welds are of no consequence. As demonstrated above, the liner and associated welds do not perform subsequent license renewal intended functions and can be neglected when establishing that the BSW will remain structurally sound when considering 72 EFPY irradiation effects. Thus, because the BSW remains structurally sound, the safety related PSW will continue to perform its subsequent license renewal intended functions throughout the SPEO.

References:

1. NEPB Letter L-2021-081 dated April 21, 2021, Point Beach Units 1 and 2 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. 3.5.2.2.2.6-3 L-2021-129 Attachment 3 Page 4of10 Associated SLRA Revisions:

SLRA Table 2.4-1, page 2.4-5 is amended as follows:

Component Type Intended Function(s)

Air lock, equipment hatches and accessories Fire barrier Pressure boundary Concrete Foundation I Basemat Direct flow Pressure boundary Structural support Concrete Walls, Buttresses, Dome and Ring Girder Fire barrier Flood barrier Missile barrier Pressure boundary Shelter, protection Structural suooort 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 Liners (refueling cavity), and covers (sand box, Unit 1 Fire barrier sump A strainer) Pressure boundary Radiation shielding Liners (reactor cavity primary shield wall ) Radiation shielding Structural suooort Liner plate (containment) Direct flow Fire barrier Pressure boundary Structural support Liner plate and keyway channels Direct flow Pressure boundary Structural support Liner plate anchors and attachments Pressure boundary Structural support Liner plate moisture barrier (sealing compound) Shelter, protection Miscellaneous structural components . ' Structural support Penetration assemblies (elastomer) Pressure boundary Structural suooort

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-3 L-2021-129 Attachment 3 Page 5 of 10 SLRA Section 3.5.2.1.1, page 3.5-3, as revised by SLRA Aging Management Supplement 1, is amended as follows:

Aging Effects Requiring Management The following aging effects associated with the Containment structure and internal structural components require management:

• Cracking

• Cumulative fatigue damage

• Distortion

• Increase in porosity and permeability

• Loss of bond

• Loss of coating or lining integrity

• Loss of leak tightness

• Loss of material

• Loss of mechanical function

• Loss of mechanical properties

• Loss of preload

• Loss of prestress

• Loss of sealing

• Loss of strength (also cited as reduction of strength)

SLRA Section 3.5.2.2.2.6, page 3.5-37 as revised by SLRA Aging Management Supplement 1, is amended as follows:

Therefore, the BSW and PSW will continue to satisfy the design criteria considering the long-term radiation effects and a plant specific AMP or enhancements to an existing AMP are not required. The BSW and PSW will continue to be inspected as part of the Structures Monitoring (B.2.3.34) AMP ., Although the BSW liner is not in the scope of SLR, \*.'ith specific attention to the potential fuf--localized distortion of the cavity liner plate as a result will be used as a leading indicator of the RIVE effect on the underlying concrete.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-3 L-2021-129 Attachment 3 Page 6of10 SLRA Table 3.5-1, page 3.5-76, is amended as follows:

Table 3.5-1 Containment Building Structure and Internal Structural Components - Summary of Aging Management Programs Item Component Aging Effect Aging Management Further Evaluation Discussion Number Requiring Program Recommended Management 3.5-1, 096 Groups 6: concrete Cracking due to AMP Xl.S7, No Consistent with NUREG-2191 .

(accessible areas) : all expansion from "Inspection of The Inspection of Water-Control reaction with Water- Control Structures Associated with Nuclear aggregates Structures Associated Power Plants (B.2.3.35) AMP is with Nuclear Power credited with managing cracking for Plants" accessible Circulating Water Pumphouse concrete.

3.5-1 , 097 Group 4: Concrete (reactor Reduction of strength; Plant-specific aging Yes (SRP-SLR Not applicable.

cavity area proximate to loss of mechanical Management program Section 3.5.2.2.2.6) A plant-specific AMP or enhancement the reactor vessel): reactor properties due to or other selected of existing AMPs is not required. The (primary/biological) shield irradiation (i.e. , AMPs , enhanced as impacts of irradiation on the primary wall ; sacrificial shield wall ; radiation interactions necessary shield wall , biological shield wall and reactor vessel support have been reactor vessel with material and evaluated for end of plant life/license support/pedestal structure radiation-induced (the subsequent period of extended heating) operation).

The primary/biological shield wall will continue to satisfy the design criteria considering the long-term radiation effects and loss of RV support fracture toughness does not require management.

The Structures Monitoring (B.2.3.34)

AMP manages primary shield wall, biological shield wall , and reactor cavity associated liner condition.

In addition , the ASME Section XI ,

Subsection IWF (B.2.3.31 ) AMP manages the aging effects for the RV support structure.

Further evaluation is described in Section 3.5.2.2.2.6.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-3 L-2021-129 Attachment 3 Page 7of10 SLRA Table 3.5.2-1 , page 3.5-88 , as revised by SLRA Aging Management Supplement 1, is amended as follows :

Table 3.5.2-1: Containment Bu ilding Structure and Internal Structural Components - Summary of Aging Management Evaluation Com ponent Intended Material Environment Aging Effect Aging NU REG- Tab le 1 Notes Type Function Requiring Management 2191 Item Management Program Item Fuel transfer tube Fire barrier Stainless Air- indoor Loss of material ASME Section lll.B3.T-37b 3.5-1, 100 E, 4 (including Pressure steel uncontrolled XI , Subsection penetration boundary IWE (B.2.3.29) sleeves, Radiation expansion joints, shielding and blind flange)

Liners (refueling Direct flow Fire Stainless Air- indoor Loss of material Structures ll l. B3.T-37b 3.5-1, 100 E, 4 cavity) and barrier steel uncontrolled Monitoring covers (sand box, Pressure (B.2.3.34)

Unit 1 sump A boundary strainer) Radiation shielding Liners (refueling Direct flow Fire Stainless Air- indoor Cracking Structures lll.B3.T-37b 3.5-1 , 100 E,4 cavity) and barrier steel uncontrolled Monitoring covers (sand box, Pressure (B.2.3.34)

Unit 1 sump A boundary strainer) Radiation shielding Liners (reactor Radiation Steel Air- indoor Loss of material Structures Vll.A1 .A-94 3.3-1, 111 c ea¥ity primary shielding uncontrolled Distortion Monitoring shield wall) Structural (B.2.3.34) support Liners (reactor Radiation Steel Air with borated Loss of material Boric Acid lll.B1 .1.T-25 3.5-1, 089 c ea¥ity primary shielding water leakage Corrosion shield wall) Structural (B.2.3.4) support Liner plate Pressure Steel Air- indoor Cumulative TLAA - Section ll.A3.C-13 3.5-1, 009 A boundary uncontrolled fatigue damage 4.6, Containment Structural Liner Plate, Metal support Containments, and Penetrations Fatigue Analysis

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-3 L-2021-129 Attachment 3 Page 8 of 10 SLRA Section 16.2.2.34, page A-36, as revised by SLRA Aging Management Supplement 1, is amended as follows:

16.2.2.34 Structures Monitoring The PBN Structures Monitoring AMP is an existing AMP that consists of periodic visual inspection and monitoring of the condition of concrete and steel structures, structural components, component supports, and structural commodities to ensure that aging degradation (such as those described in ACI 349.3R, ACI 201.1 R, SEl/ASCE 11, and other documents) will be detected, the extent of degradation determined and evaluated, and corrective actions taken prior to loss of intended functions.

Although the biological shield wall liner is not in the scope of SLR, Specific attention is given to the potential ~localized distortion of the reactor cavity liner will be used as a leading indicator plate as a result of the radiation induced volumetric expansion (RIVE) effect on the underlying concrete. Structures are monitored on an interval not to exceed 5 years.

Inspections also include seismic joint fillers, elastomeric materials; steel edge supports and bracings associated with masonry walls, and periodic evaluation of ground water chemistry and opportunistic inspections for the condition of below grade concrete. Quantitative results (measurements) and qualitative information from periodic inspections are trended with sufficient detail, such as photographs and surveys for the type, severity, extent, and progression of degradation, to ensure that corrective actions can be taken prior to a loss of intended function . The acceptance criteria are derived from applicable consensus codes and standards . For concrete structures, the program includes personnel qualifications and quantitative evaluation criteria of ACI 349 .3R.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.6-3 L-2021-129 Attachment 3 Page 9 of 10 SLRA Table 16-3, page A-105 , as revised by SLRA Aging Management Supplement 1, is amended as follows :

No. Aging NUREG-2191 Commitment Implementation Schedule Management Section Program or Activity (Section) 38 Structures Xl.S6 Continue the existing PBN Structures Monitoring AMP , including enhancement to: No later than 6 months prior Monitoring a) Revise inspection procedures to include guidance and acceptance criteria to the SPEO, i.e.:

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

b) Revise implementing procedures to address preventive actions to 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".

c) 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 SCC for stainless steel and aluminum components

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

• Localized distortion of the i:eastei: sa*;ity of the biological shield wall liner~ as a leading indicator of radiation induced volumetric ex1:1ansion of the underl~ing concrete.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No . 3.5.2.2 .2.6-3 L-2021-129 Attachment 3 Page 10of10 SLRA Section B.2.3.34, page B-239, as revised by SLRA Aging Management Supplement 1, is amended as follows:

Element Affected Enhancement Update the governing AMP procedure and other

3. Parameters Monitored applicable procedures to additionally inspect the or Inspected following elements:

G. 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 .

H. Elastomer will also be inspected for loss of material and loss of strength .

I. Pitting and crevice corrosion and evidence of cracking due to sec for stainless steel and aluminum components J. Concrete will be monitored to confirm the absence of water in-leakage K. Localized djstortjon of the bjologjca! shjeld 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 .6-4 L-2021-129 Attachment 4 Page 1 of 2 RAI 3.5.2.2.2.6-4

Background:

The SLRA states that the PBN Structures Monitoring program is consistent with enhancements to the GALL-SLR AMP, Xl.S6. As such, its scope of program includes non-ASME Code related steel structural elements and steel liners. Item 3.5-1, 097 of SLRA Table 3.5-1, "Containment Building Structure and Internal Structural Components

- Summary of Aging Management Programs," states that the program manages the reactor cavity liner condition. SLRA Table 3.5.2-1, "Containment Building Structure and Internal Structural Components - Summary of Aging Management Evaluation, of the SLRA Supplement 1, includes an AMR item for loss of material, the distortion of the reactor cavity liner (BSW steel liner) with the intended function of radiation shielding.

SLRA Section 3.5.2.2.2.6 (Supplement 1), states:

Therefore, the BSW and PSW will continue to satisfy the design criteria considering the long term radiation effects and a plant specific AMP or enhancements to an existing AMP are not required. The BSW and PSW will continue to be inspected as part of the Structures Monitoring (B.2 .3.34) AMP, with specific attention to the potential for localized distortion of the cavity liner plate as a result of the RIVE effect on the underlying concrete.

The SRP-SLR, in Generic Branch Technical Position RLSB-1, identifies the GALL-SLR Report as an approved topical report (TR) for evaluating existing programs generically to document conditions under which they are considered adequate or when they need to be augmented to manage identified effects of aging. It states:

If it is determined that the response to a specific applicant action item will result in the need for augmentation of specific programmatic criteria beyond those activities recommended in the applicable TR, the applicant should define the AMP accordingly to identify the AMP program element or elements that are impacted by the basis for responding to the applicable action item and the adjustments that will need to be made to the TR guidance recommendations, as defined in the impacted program elements for the AMP and applicable to the CLB and design basis for the facility.

Issue:

Although the RV structural steel support assembly, as noted in WCAP-18554-P/NP, was inspected to minimize the possibility of flaws, and the liner is to be inspected for potential distortion for the RIVE effect during the subsequent period of extended operation, there is no discussion whether potential cracking of the BSW (reactor cavity) steel liner was included in such inspections. Neither the Westinghouse attachments to the SLRA or the SLRA discuss the effects of radiation on the liner, liner weldments, or its anchorage to concrete. NUREG/CR-5320, "Impact of Radiation Embrittlement on Integrity of Pressure Vessel Supports for Two PWR Plants," states that "[t]he concern over radiation embrittlement is that it increases the potential for propagation of flaws that might exist."

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 .6-4 L-2021-129 Attachment 4 Page 2 of 2 The staff could not locate an AMR item in Table 3.5.2-1, of the SLRA or in SLRA Supplement 1 addressing loss of/reduction in fracture toughness or cracking due to irradiation embrittlement for managing the effects of aging for the BSW (reactor cavity) steel liner with radiation shielding intended function. Similarly, the staff could not locate an enhancement to the Structures Monitoring program for managing such an aging effect or liner cracking due to radiation embrittlement. It is not clear how PBN will manage the effects of aging for loss/reduction of fracture toughness/cracking of the reactor cavity liner during the subsequent period of extended operation .

Request:

a. Discuss how PBN plans to manage loss of/reduction in fracture toughness/cracking effects of aging of the (BSW) reactor cavity steel liner during the subsequent period of extended operation.

b. Discuss why the PBN SLRA and SLRA Supplement 1 do not provide a Table 2, AMR item and corresponding enhancements to the PBN Structures Monitoring program, applicable program elements for managing loss of/reduction in fracture toughness/cracking aging effect(s) due to radiation embrittlement of the (BSW) reactor cavity steel liner during the subsequent period of extended operation.

c. As an alternative to Request b. above, provide appropriate enhancements to the PBN Structures Monitoring program and include corresponding AMR item(s) and SLRA commitments that demonstrate adequate management of loss of/reduction in fracture toughness or cracking due to irradiation embrittlement of the reactor cavity (BSW) liner. Update the PBN SLRA Basis Document(s) and UFSAR supplement for the Structures Monitoring program as needed.

NEPB Response:

The response below addresses requests a, band c above.

As noted in the response to RAI 3.5.2.2.2 .6-3 (Attachment 3 to this letter), the PSW liner is in the scope of SLR, but the BSW liner is not because it does not perform subsequent license renewal intended functions. Accordingly, PBN does not plan to manage loss of/reduction in fracture toughness/cracking of the BSW liner. However, the response to RAI 3.5.2.2.2.6-3 does include an enhancement to the PBN Structures Monitoring program to monitor for localized distortion of the BSW liner as a leading indicator for RIVE.

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. 3.5.2 .2.2 .6-5 L-2021-129 Attachment 5 Page 1 of 1 RAI 3.5.2.2.2.6-5

Background:

SLRA Supplement 1, Section 3.5 .2.2 .2.2 .6, states that "[t]emperature assumptions were based on the normal operating temperature of the fluid in the RV nozzle of approximately 613°F and cooling of approximately 100°F for each inch away from the heat source . These temperature assumptions are consistent with previous structural analyses ."

Issue:

The SLRA Supplement 1, does not reference the "previous structural analyses ," and the staff is unable to verify the statement.

Request:

State and provide the technical reference(s) to the "previous structural analyses."

NEPB Response:

The structural analys is of record (AOR) was reviewed by the N RC as part of the Extended Power Uprate Licensing Amendment Request (EPU LAR). The AOR provides and justifies the temperatures at which component structural qualification was performed . These temperatures are reduced as the distance from the heat source (RPV nozzle) is increased and are based on assumed temperature values that have been verified to be accurate over the years based on temperature test probe data collected during various system hot functional tests . The fracture mechanics evaluation reduced the temperature a furthe r conservative value (as discussed in RAI 3.5.2.2.2.7-

6) to ensure that the critical flaw sizes were being calculated based on conservative values as documented in WCAP-18554-P/NP.

References:

1. Westinghouse Report, WCAP-18554-P/NP, Revision 1, "Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports for Point Beach Units 1 and 2. "

2. ML091250566, Attachment 5, FPL Energy Point Beach, LLC , Point Beach Nuclear Plant Units 1 and 2, License Amendment Request 261, Extended Power Uprate Licensing Report , Introduction through Chapter 2, Section 2.5.7, April 2009.

3. ML091250569, Attachment 5, FPL Energy Point Beach, LLC, Point Beach Nuclear Plant Units 1 and 2, License Amendment Request 261, Extended Power Uprate, Licensing Report, Chapter 2, Section 2.5 .8 through Appendices, April 2009.

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. 3.5.2.2.2.7-1 L-2021-129 Attachment 6 Page 1 of 5

2. SLRA Section 3.5.2.2.2.7, "Expected Further Evaluation for Loss of Fracture Toughness Due to Irradiation Embrittlement of Reactor Vessel (RV) Supports for NRC Review of the First Three SLRAs" RAI 3.5.2.2.2.7-1

Background:

SLRA Section 3.5.2.2.2.7 presents an evaluation of the irradiation effects on loss of fracture toughness of the reactor vessel (RV) support steel (ring girder and support columns) for the specified period of extended operation to ensure it will maintain its structural integrity. However, NRC staff observed this evaluation is based on neutron fluence and displacements per atom (dpa) results for which the uncertainty has not been assessed. The results of the structural analysis and fracture mechanics evaluation indicate that several RV steel support structure components possess a small amount of margin to interaction ratios and allowable flaw sizes, respectively. The effect of neutron fluence and dpa on the RV support steel with consideration of the uncertainty may cause margins to be exceeded. The staff is not able to determine whether reasonable assurance exists that the limiting neutron fluence and dpa values for the RV support steel were identified with sufficient margin and conservatism to accommodate uncertainties in the fluence analysis methodology associated with calculating exposure at an ex-vessel location.

Issue:

In order to ensure the loss in fracture toughness of steel due to the effects of irradiation are adequately managed, it is necessary to assess the uncertainty associated with neutron fluence and dpa calculation results.

Request:

Provide an estimate of the uncertainty associated with the neutron fluence and dpa results for the RV support steel (ring girder and support columns).

NEPB Response:

An analytical uncertainty analysis associated with the neutron fluence and iron atom displacement (dpa) results for the RPV support columns and ring girder was not performed for the Point Beach SLRA. Therefore, a conservative estimate of the uncertainty associated with these results was established using an existing RPV extended beltline uncertainty analysis.

The existing RPV extended beltline analysis quantified the analytical uncertainty associated with calculated fast neutron (E > 1.0 MeV) fluence rates at the RPV inner and outer surfaces at various elevations above and below the active fuel. As part of this analysis, numerous parameters that were identified as having a potentially significant contribution to the core neutron source, reactor geometry, coolant temperature, discretization, and modeling approximation uncertainties at the RPV inner and outer surfaces were evaluated. More specifically, each parameter identified was evaluated on an individual basis by determining the maximum relative change in the base-case

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.7-1 L-2021-129 Attachment 6 Page 2 of 5 fluence rate that occurred as the magnitude of that parameter was varied over a bounding range of values. The net analytical uncertainty associated with a given RPV location was then determined by taking the root sum of squares of the individual parameter uncertainty values determined at that location. Given the parameters considered, the magnitudes of the parameter variations evaluated, and the relative proximity of the RPV outer surface to the RPV support columns and ring girder, the extended beltline uncertainty analysis results for the RPV outer surface were judged to provide a reasonable basis for estimating the analytical uncertainty associated with the RPV support structure neutron fluence and iron atom displacement exposures.

The maximum neutron fluence and iron atom displacement projections at the RPV support columns occur at elevations that are slightly above the core midplane. However, since the extended beltline uncertainty analysis was, by design, focused on the RPV extended beltline region only, it did not consider axial elevations slightly above the core midplane; the elevations nearest the midplane considered were 30 cm above the top and 30 cm below the bottom of the active fuel. Therefore, the extended beltline uncertainty analysis results determined at the RPV outer surface 30 cm above the top of the active fuel were used as the starting point for estimating the uncertainty associated with the RPV support columns. This is conservative because analytical uncertainties increase with axial distance above the top of the active fuel.

The maximum neutron fluence and iron atom displacement projections at the lower edge of the RPV support ring girder occur at an elevation that is less than 30 cm above the top of the active fuel, while the maximum exposures for the upper edge occur at an elevation that is less than 70 cm above the top of the active fuel. For similar reasons as were given for the RPV support columns, the extended beltline uncertainty analysis results determined at the RPV outer surface 30 cm above the top of the active fuel were used as the starting point for estimating the uncertainty associated with the lower edge of the RPV support ring girder. The results determined at the RPV outer surface 90 cm above the top of the active fuel were used for the upper edge of the ring girder.

In addition to using these bounding RPV locations as starting points, the concrete composition parameter uncertainty values determined at these locations were increased by a factor of 2. These values were increased because they were associated with the one parameter evaluated in the RPV extended beltline analysis whose uncertainty was judged to be potentially impacted in a non-negligible manner if a detailed uncertainty analysis for the RPV support structure were performed. Note that the standard concrete composition from the BUGLE-96 documentation (Reference 2) was used for the BSW in both the extended beltline analytical uncertainty analysis base-case calculations and the Point Beach neutron fluence and iron atom displacement calculations.

Following the process described above, the analytical uncertainty associated with the neutron fluence and iron atom displacement results for RPV support columns and ring girder lower edge was conservatively estimated to be 20%; the analytical uncertainty for the ring girder upper edge was estimated to be 25%. For the following reasons, applying these uncertainties in the limiting direction to the maximum exposures determined for

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5 .2.2.2.7-1 L-2021-129 Attachment 6 Page 3 of 5 the RPV support columns and ring girder would not invalidate the results to the RPV support structure fracture mechanics evaluation summarized in Section 3.5.2.2.2.7 of of the Point Beach SLRA:

• The guidance provided in the NUREG-0933 GSl-15 resolution is to utilize Figure 3-1 of NUREG-1509 to calculate the change in RPV support structure nil-ductility transition temperature (LiNDTT) based on iron atom displacements from neutrons with energies greater than 0.1 MeV. However, the RPV support structure fracture mechanics evaluation performed in support of the SLRA used iron atom displacement exposures that included the contribution from neutrons with energies below 0.1 MeV. A sensitivity calculation performed by Westinghouse shows that excluding the contribution from these lower energy neutrons would:

o reduce the iron atom displacement values used to evaluate the RPV support columns and ring girder lower edge by approximately 11 % and o reduce the iron atom displacement values used to evaluate the RPV support ring girder upper edge by approximately 14%.

• Neutron and gamma exposure projections for Point Beach Unit 1 were based on the core power distributions and operating conditions of U1 C37 but included a

+10% bias on the peripheral and re-entrant corner assembly relative powers; projections for Unit 2 were based on U2C38 and also included a +10% bias on the peripheral and re-entrant corner assembly relative powers . A +10% bias applied to the peripheral and re-entrant corner assembly relative powers of a give cycle increases the calculated exposure rates for that cycle by approximately 10%. While the biases used for the Point Beach projection cycles were intended to account for changes in future-cycle exposure rates that are expected to occur as a result of normal variations in future-cycle core designs, they also provide a significant source of margin in the 72-EFPY exposures determined for the SLRA.

This is because future cycle designs at Point Beach Unit 1 would not be expected to consistently result in exposure rates 10% greater than the ones determined for U1 C37, and future cycle designs at Point Beach Unit 2 would not be expected to consistently result in exposure rates 10% greater than the ones determined for U2C38. This conclusion is supported by the RPV neutron exposure rates reported in LTR-REA-20-27-NP, Revision 1 (Reference 3) for Unit 1 Cycles 34-39 and LTR-REA-20-28-NP (Reference 4) for Unit 2 Cycles 32-38, which show normal cycle-to-cycle variations, but no long-term trend in either the limiting (increasing) or non-limiting (decreasing) direction.

• The iron atom displacement value used to evaluate the RPV support box ring girder bolts was the maximum exposure for any location on the box ring girder.

Since this maximum exposure occurs midway between RPV support columns, it occurs at a location that is significantly different than where the box ring girder bolts are actually located. A sensitivity calculation performed by Westinghouse shows that the iron atom displacement exposure (all neutron energies) at

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5 .2.2.2.7-1 L-2021-129 Attachment 6 Page 4 of 5 72 EFPY determined at a more representative location for the box ring girder bolts would be approximately 50% less than the one used in the RPV support structure fracture mechanics evaluation.

• The iron atom displacement exposure used to evaluate the RPV support shear brace I-beams was the maximum value for any location on the outside surface of the box ring girder. Since this maximum exposure occurs near the RPV support columns , it occurs at a location that is significantly different than where the shear brace I-beams are actually located . A sensitivity calculation performed by Westinghouse shows that the iron atom displacement exposure (all neutron energies) at 72 EFPY determined at a more representative location for the RPV support shear brace I-beams would be approximately 90% less than the one used in the RPV support structure fracture mechanics evaluation.

• Select conservatisms were considered in the box ring girder structural analysis which were not required in the general methodology of NUREG-1509. One such conservatism was the use of welding residual stresses as discussed in Section 4 of WCAP-18554-P, Revision 1 (Reference 5). Moreover, the use of the residual stresses in the fracture mechanics evaluation conside red the conservatisms described in Items 7 and 9 of Section 7 of Reference 5. In addition to the conservatisms in the stress analysis, there was conservatism in the bulk material temperature used in the fracture toughness calculations for the box ring girder, as discussed in the last paragraph of Section 5.1.2 of Reference 5.

References:

1. USNRC Request for Additional Information, "Point Beach Nuclear Plant, Units 1 and 2 (PBN) Subsequent License Renewal Application (SLRA) Requests for Additional Information (RAls) and Requests for Confirmation of Information (RCIS) Safety -

Set 1 (Aging Management of Irradiated Concrete and Steel Reactor Vessel Supports)," June 2021. (ADAMS Accession No. ML21161A119).

2. RSICC Data Library Collection DLC-185 , "BUGLE-96, Coupled 47 Neutron , 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," Radiation Shielding Information Computational Center, Oak Ridge National Laboratory (ORNL) , July 1999.

3. LTR-REA-20-27-NP, Revision 1, "Reactor Vessel, Reactor Vessel Supports, and Concrete Bioshield Exposure Data in Support of the Point Beach Unit 1 Subsequent License Renewal (SLR) Time-Limited Aging Analysis (TLAA) ," May 2020.

4. LTR-REA-20-28-NP, Revision 0, "Reactor Vessel, Reactor Vessel Supports, and Concrete Bioshield Exposure Data in Support of the Point Beach Unit 2 Subsequent License Renewal (SLR) Time-Limited Aging Analysis (TLAA)," July 2020.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2 .2.7-1 L-2021-129 Attachment 6 Page 5 of 5

5. WCAP-18554-P , Revision 1, "Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports for Point Beach Units 1 and 2," September 2020 .

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. 3.5.2.2.2.7-2 L-2021-129 Attachment 7 Page 1 of 3 RAI 3.5.2.2.2.7-2

Background:

to "Point Beach Nuclear Plants Unit 1 and 2 License Amendment Request 261, Extended Power Uprate Licensing Report (EPU LAR)," (ADAMS Accession Nos.

ML091250566, ML091250569) states that the "revised RPV supports loads and load combinations were found to be less than the appropriate allowable load limits with stress interaction ratios (I Rs) indicating that "adequate design margins exist for support loads resulting from EPU conditions." The Staff's safety evaluation (ADAMS Accession Nos. ML110880039, ML110450159), for the EPU LAR, of the RPV supports "concludes that the licensee has demonstrated that the PBNP's RPV and supports will remain structurally adequate to perform their function at EPU conditions and will continue to meet the requirements of PBN [General Design Criteria] GDCs 1, 2, 9 and 40 and the ASME Code Section Ill, Division 1, following implementation of the proposed EPU."

Table 3.5.2.2-4 of SLRA Supplement 1 presents an update to the EPU IRs for RPV support components and states that "[t]hese interaction ratios [IR] have been updated from the ones in [SLRA] References 3.5.4.7 and 3.5.4.8 based on an issue identified and corrected by Westinghouse when performing the critical flaw size analyses for SLR."

Issue:

A comparison of the listed IR values in Table 2.2.2.3-5 of the PBN EPU LAR with those of Table 3.5.2.2-4 of the SLRA Supplement 1, indicates that IRs have increased approximately from 10 to 70 percent. It is not clear what methodology was used in calculating the IRs and where in each of the reported components they occur. It is also not clear whether the newly reported IR values in the Table 3.5.2.2-4 of the SLRA Supplement 1, for each of the reported RV structural steel support components have considered the corresponding Certification of Materials Testing Result (CMTR) reported strength values or the minimum applicable ASTM International (ASTM) material strength values, and the effects of radiation. Given the uncertainty in the fluence and the values provided in the SLRA table, it is not clear what are the actual margins in the critical support components, particularly those components with IRs approximately equal to 1.

Request:

a. Clarify the methodology used in calculating the IRs and the calculated location.

b. Clarify the origin (e.g., CMTRs, ASTM) of the strength values used in calculations of the IRs listed in Table 3.5.2.2.-4 of the SLRA Supplement 1.

c. Clarify whether the values in the SLRA Supplement 1, Table 3.5.2.2-4 consider the effects of radiation. If so, define the projected margins for the RV steel support structure components (i.e., in girders and columns with and without the effects of radiation).

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.7-2 L-2021-129 Attachment 7 Page 2 of 3

d. Given the uncertainty involved in the definition of fluence , discuss how likely is for the margins to increase/decrease.

NEPB Response:

a. The methodology used to calculate the IRs is consistent with the structural analysis of record (AOR) which was reviewed by the NRC as part of the EPU LAR . An FEA model was created using simplified members to distribute the applied load at the nozzle pad location into the other support members. The maximum loads of the members were then used to calculate the maximum stresses of the members using hand calculations . The stresses were conservatively calculated based on the maximum loads regardless of the stress location being evaluated . The interaction ratios (!Rs) are calculated by dividing the maximum calculated member stresses by the allowable stresses for the appropriate loading condition . The allowable stresses are calculated per the licensing basis and the AISC 7th ed . where the published material yield strength (at temperature) is multiplied by the AISC reduction factor. These factors vary based on the loading condition (Normal , Upset, Faulted) .

b. The minimum material strength values used in calculating the !Rs listed in Table 3.5.2.2-4 of SLRA Aging Management Supplement 1 come from the ASME Boiler and Pressure Vessel Code (B&PVC). Use of the ASME code values during design is conservative and appropriate because certified material test reports (CMTR) typically indicate that tested material yield strengths are greater than the ASME code values . The ASME B&PVC values are used because the material properties are provided for various temperatures , while the ASTM specifications typically only provide properties at ambient conditions .

c. The values in SLRA Aging Management Supplement 1, Table 3.5.2.2-4 do not consider the effects of radiation . Radiation effects are not required by any design code to be considered in the calculation of !Rs fo r the RPV supports. The !Rs are based on member stresses calculated due to the applied loadings. These calculated stresses are then divided by allowable stresses that were calculated using published material yield strengths (at temperature) multiplied by reduction factors required by the licensing basis and the AISC 7th ed .

d. The uncertainty involved in the definition of fluence does not influence the margins or !Rs listed in SLRA Ag ing Management Supplement 1, Table 3.5.2.2-4, as radiation effects are not included in the calculations of the margins/IRs. The fracture mechanics evaluations, documented in WCAP-18554-P/NP , consider the influence of radiation effects on the fracture toughness of the RPV supports .

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 .7-2 L-2021-129 Attachment 7 Page 3 of 3

References:

1. "Manual of Steel Construction ," Seventh Edition, American Institute of Steel Construction, New York, N.Y. , including the "Specification for the Design, Fabrication and Erection of Structural Steel for Buildings ," February 12, 1969.

2. ASME Boiler & Pressure Vessel Code, Section Ill and Section II , "Materials ," Part D, "Properties, " 2007 Edition.

3. Westinghouse Report , WCAP-18554-P/NP, Revision 1, "Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports for Point Beach Units 1 and 2."

4. ML091250566, Attachment 5, FPL Energy Point Beach , LLC, Point Beach Nuclear Plant Units 1 and 2, License Amendment Request 261 , Extended Power Uprate Licensing Report, Introduction through Chapter 2, Section 2.5.7, April 2009.

5. ML091250569 , Attachment 5, FPL Energy Point Beach , LLC , Point Beach Nuclear Plant Units 1 and 2, License Amendment Request 261, Extended Power Uprate, Licensing Report, Chapter 2, Section 2.5.8 through Appendices , Ap ril 2009.

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. 3.5 .2.2.2.7-3 L-2021-129 Attachment 8 Page 1of2 RAI 3.5.2.2.2.7-3

Background:

SLRA Section 3.5.2.2 .2.7 , as amended by Supplement 1, defines the loading conditions for the RV structu ral assembly to be as follows :

Normal= Deadweight +Thermal Upset= Normal + QBE Seismic Faulted-1 = Normal+ SSE Seismic Faulted-2 = Normal+ SSE Seismic+ LQCA According to Table 3.5 .2.2-4, "Summary of RPV Support Component Stress Interaction Ratios [!Rs], " of the SLRA Supplement 1, the !Rs for the pipe column supports are the highest, having values 0.9954 and 0.9986, for the Upset and Faulted-2 conditions ,

respectively. The table also lists an IR of0 .7582 forthe Faulted-1 condition . An IR value of 1.0 ind icates that applied load stresses equal those that are allowed by the applicable design codes.

Issue :

The Upset loading condition includes Normal loads plus seismic loads associated with the QBE. The Faulted-1 loading condition includes Normal loads plus the SSE Seismic loads. It is noted that Upset and Faulted-1 loading conditions differ only in seismic forces . For pipe column supports, it is not clear why !Rs for Faulted-1 loading condition are less than those for Upset, when SSE Seismic forces are greater than QBE Seismic forces. It is also not clear why the !Rs for the Faulted-2 loading condition that includes increased seismic and LQCA loads over those of Upset loading condition are incremented only by 0.0032 .

Request:

Discuss the apparent inconsistency in calculations of pipe column !Rs fo r Upset, Faulted 1, and Faulted 2 loading conditions. Clarify why the IRs for the :

i. Faulted-1 loading condition are less than those reported for the Upset loading condition.

ii . Faulted-2 loading condition that includes increased seismic ~nd LQCA loads over those of Upset loading condition are incremented only by 0.0032 .

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.7-3 L-2021 -129 Attachment 8 Page 2 of 2 NEPB Response:

i and ii: The analysis uses the same methodology as the analysis of record (AQR) for the reactor pressure vessel supports structural qualification and was reviewed by the NRC as part of the EPU LAR. The analysis performs evaluations based on the input load combinations Normal , Upset, and Faulted . These are defined as:

Normal = Deadweight + Thermal Upset= Normal+ QBE Seismic Faulted-1 = Normal+ SSE Seismic Faulted-2 = Normal+ SSE Seismic+ LQCA Per the AQR , the QBE loads are to be a minimum of half of the SSE loads . The QBE loads used during the EPU LAR are greater than half of the SSE loads , such that the vertical QBE load is approximately 37 percent higher than half of the vertical SSE load and the horizontal QBE load is approximately 47 percent higher than half of the horizontal SSE load. The stresses calculated based on these load combinations are then compared to allowable stresses defined by the licensing basis and the AISC 7th ed .

based on the loading condition (i.e ., Normal, Upset, Faulted). The allowable stresses for Normal and Upset cases are lower than allowable stresses for the Faulted cases which results in high interaction ratios for the Upset case even though the stresses are lower than those of the Faulted cases.

References:

1. "Manual of Steel Construction ," Seventh Edition, American Institute of Steel Construction, New York , N.Y., including the "Specification for the Design, Fabrication and Erection of Structural Steel for Buildings," February 12, 1969.

2. ML091250566, Attachment 5, FPL Energy Point Beach, LLC, Point Beach Nuclear Plant Units 1 and 2, License Amendment Request 261, Extended Power Uprate Licensing Report , Introduction through Chapter 2, Section 2.5.7 , April 2009.

3. ML091250569 , Attachment 5, FPL Energy Point Beach , LLC , Point Beach Nuclear Plant Units 1 and 2, License Amendment Request 261, Extended Power Uprate, Licensing Report, Chapter 2, Section 2.5.8 through Appendices , April 2009.

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 . 3.5.2.2.2.7-4 L-2021-129 Attachment 9 Page 1 of 3 RAI 3.5.2.2.2.7-4 Background :

Table 3.5.2.2-6, "Summary of Postulated Critical Flaw Sizes for 72 EFPY," of SLRA Section 3.5 .2.2.2.7 , as amended by Supplement 1, itemizes postulated critical flaw sizes for Normal, Upset, Faulted-1, and Faulted-2 loading conditions. A footnote to the aforementioned SLRA table states that the "postulated critical flaw sizes are determined by setting [the] applied stress intensity factor equal to [the] fracture toughness and back-calculating [the] flaw size." The staff also notes that for the definition of fracture toughness, ASME Section Ill and Section XI require consideration of stresses from applicable loadings, including the effects of aging due to irradiation.

Table 3.5.2.2-4, "Summary of RPV Support Component Stress Interaction Ratios [IRs],"

of the SLRA Supplement 1, indicates that the IRs for the pipe column supports for the Upset and Faulted-2 loading conditions are approaching unity, while those of the box ring girder subject to similar loading conditions are much less . Section 3.5 .2.2.2 .7 of SLRA Supplement 1, also states that ASME Section XI , Subsection IWF AMP will perform visual inspections of the RV steel support structure.

Issue:

The staff noted that applicant ePortal documents (e.g. , Standards for Welding T-1 Material Specifications) indicate that a lower yield strength electrode was used for the T-1 weldments , resulting in "undermatched" welds. Although weld fracture toughness is not addressed in the WCAP-18554-P/NP , weld flaws has been addressed through NDTs as part of the RV structural steel assembly fabrication . For those I Rs approaching unity in Table 3.5.2 .2-4 of the SLRA Supplement 1, the inference is that design stresses approach the controlling material yield stress without considering the effects of potential undetected flaws. Conservatively assuming that such material discontinuities exist, it is not clear whether the stress analysis methodology used in the definition of IRs considered potential undetected flaws in the irradiated welded structural steel RV support assembly, potentially resulting in IRs > 1.0. Furthermore, it is not clear: (a) where in the columns the IRs are maximized ; (b) whether the maximized IR locations represent welded joints, and (c) if so whether residual stresses were considered in the calculation of the IRs .

Request:

a. Clarify whether the effects of potential undetected flaws , if any, have been considered in the calculated IR values in Table 3.5 .2.2-4 of the SLRA Supplement 1.

b. If not, discuss the adequacy of the ASME Section XI , Subsection IWF AMP to examine the RV steel support assembly and reasonably assure that potential undetected flaws , including those in welds , would not affect its structural integrity during 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. 3.5 .2.2.2 .7-4 L-2021 -129 Attachment 9 Page 2 of 3

c. Clarify: (a) where in the columns the IRs are maximized ; (b) whether the maximized IR locations represent welded joints , and (c) if so whether residual stresses were considered in the calculation of the IRs .

NEPB Response:

a. The effects of potential undetected flaws in the calculated IR values in Table 3.5.2.2-4 of the SLRA Aging Management Supplement 1 are not considered because they are not required to be per the reactor vessel support design specifications or the AISC 7th ed . code . The methodology used to calculate the IRs is consistent with the structural analysis of record (AOR) which was reviewed by the NRC as part of the PBN extended power uprate (EPU) license amendment (Reference 1). An FEA model was created using simplified members to distribute the applied load at the nozzle pad location into the other support members . The maximum loads of the members were then used to calculate the maximum stresses of the members using hand calculations. The interaction ratios (IRs) are calculated by dividing the maximum calculated member stresses by the allowable stresses. The allowable stresses are calculated per the licensing basis and the AISC 7th ed. where the published material yield strength (at temperature) is multiplied by the AISC reduction factor. Note the AOR mentioned above has been uploaded to the ePortal.

b. Based on the discussion in the Background and Issue of RAI 3.5 .2.2.2.7-4, the question appears to be related to the RV support columns where under Upset and Faulted-2 loading conditions the interaction ratios approach unity. As presented in Table 3.5 .2.2-6 of PBN SLRA Section 3.5 .2.2.2 .7, the fracture mechanics analyses performed by Westinghouse indicates that for the RV support columns (see response to c. below) , the postulated critical flaw sizes for the Upset and Faulted-2 loading conditions were 62 .9% and 44.4%, respectively . When compared to the Section XI allowable flaw size of 3.1 %, the postulated critical flaw sizes are 14 to 20 times larger. Thus , although the IRs are approaching unity, flaws that would impact the structural integrity of the RV support columns would be large enough to be identified through the ASME Section XI , Subsection IWF inspections. Additionally, the applicable aging effects of loss of material and reduction in fracture toughness due to irrad iation embrittlement would typically not generate new flaws or cause existing undetectable flaws to grow appreciably during plant operation . Finally, from SLRA Aging Management Supplement 1 any increase in potential critical flaw size from irradiation embrittlement has been shown to be negligible (<5% over 30 EFPY),

which means that effects from loss of fracture toughness due to irradiation embrittlement are adequately addressed .

Accordingly , there is reasonable assurance that the PBN ASME Section XI, Subsection IWF lnservice Inspection program is adequate to ensure that the RV support assembly will continue to be able to perform its intended functions through the SPEO .

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 .7-4 L-2021-129 Attachment 9 Page 3 of 3

c. The column IRs are calculated using maximum stresses due to maximum column loads regardless of the stress location being evaluated. Hand calculations are used to calculate tension stress, compression stress, shear stress, bending stress and torsional stress for the column. The maximum of these stresses (including their combinations) are used to calculate the IRs listed in Table 3.5.2 .2-4 of the SLRA Aging Management Supplement 1. The column welds are full penetration welds, therefore are treated as part of the pipe cross section. Since the weld metal is an E?O series, the pipe (A53) yield strength is conservatively used to calculate the allowable stresses. Therefore, the IRs are maximized at considered welded joint regions . The calculated allowable stress includes the reduction factors provided by the AISC 7th ed and the licensing basis . Welding residual stresses are not included in the calculations as they are not required to be included by any design code (i.e.,

AISC , ASME Sec. Ill) for the structural qualification.

References:

1. Point Beach, Units 1 and 2 - Safety Evaluation: Extended Power Uprate (TAC Nos.

ME1044 and ME1045) , May 3, 2011 , (ADAMS Accession No. ML110450159).

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. 3.5.2.2 .2.7-5 L-2021-129 Attachment 10 Page 1of3 RAI 3.5.2.2.2. 7-5 Background :

SLRA Section 3.5.2.2.2.7, as amended by SLRA Supplement 1, states in part that "a plant specific AMP or enhancements to an existing AMP are not required to manage loss of fracture toughness due to irradiation embrittlement of the RV supports at PBN."

Revision 0 of the audited PBN ASME Section XI, Subsection IWF AMP Basis Document, FPLCORP00036-REPT-059, reinforces this notion and states that "[f]urther evaluation determined that a plant-specific AMP or enhancements to an existing AMP are not required to manage the aging effect of loss of fracture toughness due to irradiation embrittlement of the RV supports at PBN." Revision 0 of the audited "Primary Shield Wall and Reactor Vessel Support Irradiation Evaluation," FPLCORP00036-REPT-035, basis document also states that:

A review of the aging effects of loss of fracture toughness due to irradiation embrittlement on the PBN supports for SLR was performed ... [and] a plant-specific AMP or enhancements to an existing AMP to manage the effects of concrete and RV support irradiation are not expected to be necessary to ensure the components perform their intended function consistent with the CLB through the SPEO.

The SRP-SLR in its Generic Branch Technical Position RLSB-1, identifies the GALL-SLR Report as an approved topical report (TR) for evaluating existing programs generically to document conditions under which they are considered adequate or when they need to be augmented to manage identified effects of aging. It states:

If it is determined that the response to a specific applicant action item will result in the need for augmentation of specific programmatic criteria beyond those activities recommended in the applicable TR, the applicant should define the AMP accordingly to identify the AMP program element or elements that are impacted by the basis for responding to the applicable action item and the adjustments that will need to be made to the TR guidance recommendations, as defined in the impacted program elements for the AMP and applicable to the CLB and design basis for the facility.

Issue:

SLRA Supplement 1, Table 3.5.2-1: Containment Building Structure and Internal Structural Components - Summary of Aging Management Evaluation," for the RV supports and bolting states that loss of fracture toughness aging effect is managed by the ASME Section XI, Subsection IWF (B.2.3.31) AMP. The staff reviewed the PBN ASME Section XI, Subsection IWF (B.2.3.31) AMP, but could not identify any enhancements associated with loss of fracture toughness aging effect for the RV steel structural supports. It is not clear what program elements of the ASME Section XI, Subsection IWF (B.2.3.31) AMP PBN plans to augment or make adjustments to, so that loss of fracture toughness aging effect due to radiation embrittlement of the RV steel

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No . 3.5.2.2.2.7-5 L-2021-129 Attachment 10 Page 2 of 3 structural support assembly is adequately managed during the subsequent period of extended operation.

Request:

a. Provide measures that would enhance the PBN ASME Section XI, Subsection IWF (B.2.3 .31) AMP and corresponding SLRA commitment(s) and updated UFSAR supplement description, to manage loss of fracture toughness/cracking aging effects due to radiation embrittlement for the RV supports and bolting during the subsequent period of extended operation.

b. If none are intended, state why. Otherwise, update PBN SLRA Basis Document(s) for the ASME Section XI, Subsection IWF program as needed.

NEPB Response:

The letters associated with the responses below correspond to the letters in the RAI above .

a. The SLR evaluation of irradiation effects on the PBN reactor vessel supports was performed consistent with the methodology described in NUREG-1509 and is summarized in PBN SLRA Sections 3.5.2.2 .2.6 and 3.5 .2.2.2.7 as amended by PBN SLRA Aging Management Supplement 1. Further clarification is also contained in the NEPB response to RAI 3.5.2 .2.2 .7-4 (Attachment 9 to this letter) and PBN SLRA Aging Effects Supplement 1. This information demonstrates the following:

• The current IWF AMP (B.2.3.31) manages effects of aging due to loss of fracture toughness. The fracture mechanics analyses performed by Westinghouse confirmed that in a majority of cases, including the columns, the critical flaw sizes for PBN RV support components were much larger than the ASME Section XI allowable flaw sizes . Thus , flaws that would impact the structural integrity of the RV supports would be large enough to be identified through the ASME Section XI, Subsection IWF inspections. Other smaller critical flaw sizes were reviewed on a case-by-case basis as discussed in Section 8.3 of WCAP-18554, conservatisms discussed in Section 7 of WCAP-18554, and resolved through discussions in RAI 3.5.2.2.2 .7-4 and other in-house audit discussions. Based on these discussions it was demonstrated there is sufficient flaw tolerance in the RPV supports has been demonstrated .

• From PBN SLRA Aging Management Supplement 1, any increase in potential critical flaw size from irradiation embrittlement was shown to be negligible (<5%

over 30 EFPY), which when coupled with the initial inspections showing no defects, means that loss of fracture toughness due to irradiation embrittlement is adequately addressed.

Therefore , the evaluations concluded that no additional measures or enhancements were required beyond the inspections and frequencies currently performed as part of the PBN ASME Section XI , Subsection IWF (B.2.3.31) AMP.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.7-5 L-2021-129 Attachment 10 Page 3 of 3 The visual inspections performed and associated acceptance criteria for the PBN ASME Section XI , Subsection IWF AMP, supported by the fracture mechanics analysis of the RV supports which considers current licensing basis loading and irradiation effects, together provide reasonable assurance that loss of fracture toughness due to irradiation embrittlement will not affect the ability of the RV supports to perform their intended functions through the SPEO. Revisions to SLRA Section B.2.3 .31 are provided below to specifically identify that the PBN ASME Section XI, Subsection IWF AMP manages loss of fracture toughness of the reactor vessel support and bolting materials.

b. The basis for no additional measures and/or enhancements to the PBN ASME Section XI, Subsection IWF (B.2.3.31) AMP is provided in response to request a.

above. Additionally, as noted in commitment 51 in Table 16-3 of Appendix A to the PBN SLRA, PBN will continue to follow the ongoing industry efforts that are clarifying the effects of irradiation on concrete and RV support steel and corresponding aging management recommendations.

References:

None.

Associated SLRA Revisions:

SLRA Section B.2.3.31 , page B-225 is amended as follows: The PBN ASME Section XI, Subsection IWF AMP is an existing AMP that consists of periodic visual examination of ASME Code Section XI Class 1, 2, and 3 supports for ASME piping and components for signs of degradation such as corrosion; cracking, deformation; misalignment of supports; missing, detached, or loosened support items; loss of integrity of welds; improper clearances of guides and stops; and improper hot or cold settings of spring supports and constant load supports. Bolting for Class 1, 2, and 3, piping and component supports is also included and inspected for corrosion, loss of integrity of bolted connections due to self-loosening, and material conditions that can affect structural integrity. This program is also credited with managing loss of fracture toughness due to irradiation embrittlement of the reactor vessel support and bolting materials. 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 .

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.7-6 L-2021-129 Attachment 11 Page 1 of 2 RAI 3.5.2.2.2.7-6

Background:

The postulated critical flaw sizes in SLRA Table 3.5.2.2-6 are based on the fracture mechanics analysis in report WCAP-18554-NP, Revision 1, which the applicant included as Attachment 2 of Enclosure 4 to the SLRA (ADAMS Accession No .

ML20329A264). These postulated critical flaw sizes were determined by setting the applied stress intensity factor equal to the fracture toughness and back-calculating flaw size . Fracture toughness depends on the temperature at each of the limiting locations shown in SLRA Table 3.5 .2.2-6. In Section 5.1.2 of WCAP-18554-NP, Revision 1, the applicant stated that the vertical legs of the supports and the corners of the hexagonal ring-beam support (box ring girder) are exposed to considerable movement of ambient temperature air and are , therefore, close to ambient temperatures (-65°F-100°F).

Issue:

The bolts at the ring girder are one of the limiting locations identified in the table of postulated critical flaw sizes in SLRA Table 3.5.2.2-6 . Based on information in WCAP-18554-P, the staff noted that the appropriate temperature of the bolts at the ring girder appeared to be in the range of 65°F to 100°F. However, the temperature for the bolts at the ring girder indicated in Table 5-2 of WCAP-18554-NP, Revision 1 is higher than 65°F to 100°F. A higher temperature means a higher fracture toughness value for the bolts at the ring girder, which results in a larger, less conservative postulated critical flaw size for the bolts at the ring girder.

Request:

Either

a. Recalculate the postulated critical flaw size for the bolts at the ring girder using the ambient temperature at the corners of the hexagonal box ring girder (65°F to 100°F),

which as the staff noted, appeared to be the appropriate temperature for the bolts at the ring girder, or

b. Given that the temperature used for the fracture toughness calculation of the bolts at the ring girder was higher than 65°F to 100°F justify how the I

postulated critical flaw size for the bolts at the ring girder shown in SLRA Table 3.5 .2.2-6 is adequate or explain how the postulated critical flaw size would be affected by using the ambient temperature of 65°F to 100°F at the corners of the hexagonal box ring girder.

Point Beach Nuclear Plant Units 1 and 2 WITHHELD FROM PUBLIC DISCLOSURE Dockets 50-266 and 50-301 UNDER 10 CFR 2.390 NEPB Response to NRC RAI No. 3.5.2 .2.2 .7-6 L-2021-129 Attachment 11 Page 2 of 2 NEPB Response:

During the initial start of the fracture mechanics evaluations in WCAP-18554-P/NP ,

Westinghouse requested the site to provide as-measured temperatures for the Point Beach RPV supports . However, based on discussions with site personnel , the temperatures of the supports are not measured and therefore no as-measured information was available (see Reference 27 of WCAP-18554-P/NP). However, in Reference 27 of WCAP-18554-P/NP, the site stated that a high-level assumption of 65°F to 100°F can be considered as ambient temperatures for the general region of interest at the RPV supports .

In order to use a more appropriate bulk metal temperature for the supports fracture toughness calculations , [

]a,c,e was used in the evaluation of the bolts at the ring girder as reported in Table 5-2 of WCAP-18554-P/NP as an appropriate metal temperature value for the fracture toughness calculations. As a result, the postulated critical flaw size for the bolts at the ring girder shown in SLRA Table 3.5.2.2-6 is adequate based on the temperature considered in Table 5-2 of WCAP-18554-P/NP.

References:

1. Westinghouse Report, WCAP-18554-P/NP, Revision 1, "Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports for Point Beach Units 1 and 2, September 2020 .

Associated SLRA Revisions:

None.

Associated

Enclosures:

Westinghouse Letter CAW-21-5198 Executed July 2, 2021, Application for Withholding Proprietary Information from Public Disclosure

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5 .2.2.2.7-6 L-2021-129 Attachment 11 Enclosure Page 1of4 Enclosure Westinghouse Letter CAW-21-5198 Executed July 2, 2021, Application for Withholding Proprietary Information from Public Disclosure Proprietary Information Notice and Copyright Notice

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.7-6 L-2021-129 Attachment 11 Enclosure Page 2 of 4 COMMONWEALTH OF PENNSYLVANIA:

COUNTY OF BUTLER:

(1) I, Jill S. Monahan, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Elech"ic Company LLC (Westinghouse).

(2) I am requesting NextEra Energy Point Beach (NEPB) Letter L-2021-129 Attachment l lP be withheld from public disclosure under 10 CFR 2.390.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a h"ade secret, privileged, or as confidential commercial or financial information.

(4) Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosme should be withheld.

(i) The information sought to be withheld from public disclosme is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii) The information sought to be withheld is being h"ansmitted to the Commission in confidence and, to Westinghouse's knowledge, is not available in public sources.

(iii) Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosme. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensmate expenses. Also, public disclosure of the information would enable

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2 .2.7-6 L-2021-129 Attachment 11 Enclosure Page 3 of 4 others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

(5) Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types , the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc .) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.7-6 L-2021-129 Attachment 11 Enclosure Page 4 of 4 Westinghouse Non-Proprietary Class 3 CAW-21-5198 Page 3of3 (6) NEPB Letter L-2021-129 Attachments 11 and 11 Pare bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of infomrntion being identified as proprietary or in the margin opposite such infonnation . These lower-case letters refer to the types of infomrntion Westinghouse customarily holds in confidence identified in Sections (5}(a) through (f) of this AffidaYit.

I declare that the aYerments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.

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

Executed on: 7-:;)-;)0d \ q LRP S-rfJ()roJ-._

Jill S. Monahan, Manager eVinci Modeling and Analysis

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RCI No. 3.5.2 .2.2.7-1 L-2021-129 Attachment 12 Page 1of1 RCI 3.5.2.2.2.7-1 Information:

Based on the audit review of report WCAP-18554-P, Revision 1, "Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports for Point Beach Units 1 and 2," Revision 1, the staff noted the conservatisms described in Section 7, Items 7 and 9, which pertain to how stresses were treated in the fracture mechanics evaluation .

Request:

Confirm that the conservatisms described in WCAP-18554-P, Revision 1, Section 7, Items 7 and 9, apply to the steel plates, columns, bolts , pins , and leveling screws of the limiting locations shown in SLRA Table 3.5 .2.2-6 that were analyzed in the fracture mechanics evaluation of the reactor vessel steel support structures in SLRA FE 3.5.2.2.2 .7.

NEPB Response:

Westinghouse confirms that Items 7 and 9 in Section 7 of WCAP-18554-P/NP (Reference 1) apply to all the locations shown in SLRA Table 3.5 .2.2-6.

References:

1. Westinghouse Report, WCAP-18554-P/NP, Revision 1, "Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports for Point Beach Units 1 and 2," September 2020 (SLRA Enclosure 4, Attachment 2 and Enclosure 5, Attachment 2) .

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 RCI No. 3.5.2.2.2.7-2 L-2021-129 Attachment 13 Page 1 of 1 RCI 3.5.2.2.2.7-2 Information:

Based on the audit review of Bechtel Specification No. 6118-C-10, "Specification for Detailing, Fabrication and Delivery of Major Component Support Structures for the Point Beach Nuclear Plant Unit 1. Wisconsin Michigan Power Company. Bechtel Corporation, San Francisco CA for Westinghouse Electric Corporation Atomic Power Divisions,"

Revision 2, the staff noted that Section 9.2 indicates that the design specification shall conform to "Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings," adopted April 17, 1963, and the "Code of Standard Practice for Steel Buildings and Bridges" of the American Institute of Steel Construction (AISC) revised February 20, 1963, and Section 14.5 indicates that welding procedures shall conform to American Welding Society (AWS) D2.0.

Request:

Confirm that the specification 6118-C-10 calls forth the 1963 version of AISC indicated above and AWS D2.0.

NEPB Response:

Westinghouse confirms that Section 9.2 of Bechtel Specification No. 6118-C-10, Revision 2, references the 1963 version of the two AISC codes: "Code of Standard Practice for Steel Buildings and Bridges" and "Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings." Section 14.5 of the Bechtel Specification No. 6118-C-10, Revision 2, references the 1966 version of American Welding Society (AWS) D2.0 (AWS D2.0-66).

References:

None.

Associated SLRA Revisions:

None.

Associated

Enclosures:

None.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 L-2021-129 Proprietary Attachment 11 P End of Non;..Proprietary Attachments Proprietary Attachment 11 P is Inserted Beginning on the Following Page

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RAI No. 3.5.2.2.2.7-6

• L-2021 -129 Attachment 11 P Enclosure Page 1 of 4 Enclosure Westinghouse Affidavit CAW-21-5198 Executed July 2, 2021 Application for Withholding Propriet~ry Information from Public Disclosure Proprietary Information Notice and Copyright Notice

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRC RA! No. 3.5.2.2 .2.7-6 L-2021-129 Attachment 11P Enclosure Page 2 of 4 Westinghouse Non-Proprietary Class 3 CA W-21-5198 Pagelof3 COMMONWEALTH OF PENNSYLVANIA:

COUNTY OF BUTLER:

(1) I, Jill S. Monahan, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2) I am requesting NextEra Energy Point Beach (NEPB) Letter L-2021-129 Attachment l lP be withheld from public disclosure under 10 CFR 2.390.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

(4) Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosme is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii) The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouse's knowledge, is not available in public sources.

(iii) Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a docmnent should be withheld from public disclosme. Nevertheless, public disclosme of this proprietary information is likely to cause substantial hatm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensmate expenses. Also, public disclosme of the information would enable

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 NEPB Response to NRG RAI No. 3.5.2.2.2.7-6 L-2021-129 Attachment 11 P Enclosure Page 3 of 4 Westinghouse Non-Proprietary Class 3 CA W-21-5198 Page2of3 others to use the information to meet NRC requirements for licensing documentation without pmchasing the right to use the information.

(5) Westinghouse has policies in place to identify proprietary infonnation. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, sh*ucture, tool, method, etc .), the application of which data secmes a competitive economic advantage (e.g., by optimization or improved marketability).

(c) Its use by a competitor would reduce his expenditure of resomces or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price infonnation, production capacities, budget levels, or commercial sh*ategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-3 01 NEPB Respon se to NRG RAI No. 3. 5.2. 2.2. 7-6 L-2 021-129 Atta chment 11 P En closure Page 4 of 4 Westinghouse Non-Proprietary Class 3 CAW-21-5198 Page 3of3 (6) NEPB Letter L-2021-129 Attachments 11 and 11 Pare bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of infonnation being identified as proprietary or in the margin opposite such infonnation . These 10\ver-case letters refer to the types of infomrntion Westinghouse customarily holds in confidence identified in Sections (5)(a) through (£) of this Affidavit.

I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.

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

Executed on: 7- :J-;J 0 d \ q LQ_OSf0eJ~

Jill S. Monahan, Manager eVinci Modeling and Analysis