ML25028A425

Breakout Questions - Aging Management Audit - Browns Ferry Units 1, 2, and 3 - Subsequent License Renewal Application

ML25028A425
Person / Time
Site:	Browns Ferry
Issue date:	01/29/2025
From:	Office of Nuclear Reactor Regulation
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Download: ML25028A425 (1)

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Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Proprietary Breakout Questions Aging Management Audit Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application March 18, 2024 - January 29, 2024

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions aging effects. However, there are AMR Table 2 items in Table 3.1.2-1 that cite this item with generic note E and plant-specific note 1 which states, The BWR Vessel Internals program is substituted to manage the aging effect(s) applicable to this component type, material, and environment combination. In Table 3.1.1 there is no discussion of the BWR Vessel Internal Program being used for Table 2 items associated with item 3.1-1, 085.

Inspection programs will be used to manage the aging effects when there are components associated with this item that use the BWR Vessel Internals Program to manage the aging effects.

5 SLRA Table 3.3.1, SLRA Table 3.4.2-2 and SLRA Table 3.4.2-3 3-335, 3-601 and 3-610 The discussion for item 3.3-1, 020 in SLRA Table 3.3.1 states that this item is consistent with NUREG-2191 and that the Water Chemistry and One-Time Inspection programs will be used to manage the aging effects. However, there are AMR Table 2 items in Table 3.4.2-2 that cite this item with generic note E and plant-specific note 3 which states, The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program (B.2.1.24) is substituted to manage the aging effect(s) applicable to this component type, material, and environment combination. There are also AMR Table 2 items in Table 3.4.2-3 that cite this item with generic note E and plant-specific note 2 which states, The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program (B.2.1.24) is substituted to manage the aging effect(s) applicable to this component type, material, and environment combination. In Table 3.3.1 there is no discussion of the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program being used for Table 2 items associated with item 3.3-1, 020.

• Please describe the basis for saying that item 3.3-1, 020 is consistent with NUREG-2191 and stating that the Water Chemistry and One-Time Inspection programs will be used to manage the aging effects when there are components associated with this item that use the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program to manage the aging effects.

6 SLRA Table 3.3.2-19 3-468 SLRA Table 3.3.2-19 states that steel, piping, piping components and tanks exposed to sodium pentaborate will use the Water Chemistry and One-Time Inspection programs to manage loss of material.

Please provide the following information about the steel components subject to loss of material and in a sodium pentaborate internal environment.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions These items cite generic note F and plant-specific note 1 which states, The Water Chemistry program (B.2.1.2) (relating to Standby Liquid Control) is used to manage the aging effect(s) applicable to this component type, material and environment combination. Because SLC systems are constructed primarily from stainless steels (e.g., NUREG/CR-6001, ML040340671), the staff requests additional information about the steel components and conditions of exposure.

a) The specific components requiring aging management.

b) Location of the components in the SLC system.

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

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

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

f) The grades of steel used for these components.

7 SLRA Table 3.1.2-3 3-147 SLRA Table 3.1.2-3 states that steel and stainless steel piping, piping components (class 1) and piping, components ((class 1) Valve body)) exposed to a treated water internal environment will use the Water Chemistry and One-Time Inspection programs to manage loss of material. These items cite generic note G and plant-specific note 1 that states, The aging effects identified for this material/environment combination are consistent with industry experience.

Because the application states that this is a different environment than what is in GALL-SLR the staff Please provide the following information about the steel and stainless steel components subject to loss of material in a treated water internal environment:

a) The specific components requiring aging management.

b) The basis for aligning these items to item 3.1-1, 079.

c) The basis for using generic note G for these items.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions requests additional information about these components.

d) The nature of these components exposure to treated water.

e) The nature of the treated water environment that these components are exposed to.

f) How the Water Chemistry and One-Time Inspection Programs will be used to manage loss of material for these components.

g) Please clarify what industry experience is being referred to in plant-specific note 1 and identify any references used to document this experience.

8 SLRA Table 3.1.2-3 and SLRA Table 3.1.2-4 3-147 SLRA Table 3.1.2-3 states that steel and stainless steel piping, piping components (class 1) and piping, components ((class 1) Valve body)) exposed to a treated water internal environment will use the Water Chemistry and One-Time Inspection programs to manage loss of material. These items cite generic note G which indicates that the environment is different from GALL-SLR. These items also cite Table 1 item 3.1-1, 079. However, in SLRA table 3.1.2-4 there are stainless steel components which also cite table 1 item 3.1-1, 079 such as flow device (class 1) exposed to Treated Water Treated water >60°C

[>140°F] (Internal) which cites generic note A. Also, in SLRA Table 3.1.2-4 stainless Piping, piping components: Class 1 greater than or equal to 4" NPS cites generic note A for Treated water >93°C [>200°F]

(Internal). Also, stainless steel Piping, piping components (Valve Body (Class 1)) in table 3.1.2-4 cites Table 1 item 3.1-1, 079 and cites generic note A for loss of material in an internal treated water environment. In these cases, treated water is not considered in the SLRA as an environment that is Please clarify the following:

a) The basis for designating note G for treated water internal items that cite Table 1 item 3.1-1, 079 in Table 3.1.2-3 when in Table 3.1.2-4 treated water classified with Note A when citing item 3.1-1, 079.

b) The basis for using Table 1 item 3.1-1, 016 for steel components subject to loss of material in treated water internal environments or item 3.1-1, 022 for stainless steel components subject to loss of material in treated water internal environment when similar items reference different table 1 items.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions different from GALL-SLR. Also, in Table 3.1.2-4 there are instances where Table 1 item 3.2.-1, 016 is being used for steel components subject to loss of material in internal treated water environments. Steel Piping, piping components: Class 1 piping, fittings and branch connections less than 4" NPS and greater than or equal to 1" NPS is one example of this. Also, item 3.2-1, 022 is being used for stainless steel components subject to loss of material in treated water internal environment in SLRA Table 3.1.2-4.

One such example is, stainless steel Piping, piping components (Valve Body). The staff requests more clarification because there seems to be multiple ways of classifying the same combination of material, environment, and aging effect.

SLRA Section3.5.2.2.2.6: Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question / Request F1 3.5.2.2.2.6 3-700 The nature of the analysis performed is unclear.

Pg 3-700 states Fluence calculations were performed to confirm attenuation effects through RV internals, the RV and outwards to the CBS. The result of this analysis is a peak fluence of 6.15E-04dpa at the inner surface of the biological shield wall. However, it is unclear from the description what the nature is of the analysis performed.

Please clarify the nature of this analysis and if necessary, provide a reference for the calculations. For example, were the fluence calculations performed by the RAMA methodology or were the calculations similar to the determination of the 0.1MeV fluence exposure and gamma dose for the concrete?

F2 3.5.2.2.2.6 3-700 Clarification regarding support structure fluence analysis with respect to concrete.

Pg 3-700 states Fluence calculations were performed to confirm attenuation effects through RV internals, the RV and outwards to the CBS. The As a follow up to the previous question, if the RAMA methodology was applied to determine the peak fluence in dpa at the biological shield wall inner surface, why was this approach not applied for determination

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions considered in the TRANSFX analyses compare to the secondary gamma production from the hematite grout concrete of the Browns Ferry CBSW).

C1 3.5.2.2.2.6 3-694 through 3-701 This breakout question is to help support the NRC staff evaluation with a general understanding of structures and components (SCs) that fall under the scope of SLRA Section 3.5.2.2.2.6.

a) Clarify in sufficient detail using illustrative drawings or diagrams, the general arrangement and configuration of the structures and components (SCs) that fall under the scope of this further evaluation for cumulative effects of aging due to combined aging mechanisms including those of irradiation.

b) Emphasize in your effort descriptions, illustrations of the as-built concrete biological shield wall (CBSW), CBSW star truss, RV stabilizers/brackets, RV annulus and cavity, RV anchorage, CBSW WF beams and columns, CBSW WF columns anchorage to the structure of the concrete pedestal, extent of the CBSW inner and outer liners, ring girder and sole plate, and their anchorage.

c) Indicate using illustrative drawings or diagrams the extent of the bottom, top, and central portions of the CBSW infill concretes and their proximity/location with regard to the fuel core mid plane.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions C2 3.5.2.2.2.6 B.2.1.30 B.2.1.33 3-694 through 3-701 B-154 through B-159 B-166 through B-173 SLRA Section 3.5.2.2.2.6, Subsection SLR Reactor Vessel Support Steel Evaluations states:

In addition to the potential aging effects due to irradiation of reinforced concrete, a loss (or reduction) in fracture toughness due to irradiation embrittlement of the reactor vessel (RV) support steel is a potential aging effect considered in this subsection. NUREG-1509, May 1996, Radiation Effects on Pressure Vessel Supports, is a resource for addressing the issue for subsequent license renewal []

NUREG-1509 is a resource for addressing the potential effects of irradiation on the steel elements of the CBS for SLR.

As noted in NUREG-1509 the prerequisite for its use to evaluate the effects of radiation is a physical examination of the supports to detect visible signs of degradation, including, but not limited to rust, corrosion, cracks or permanent deformation of the members followed-up with a comparison to the initial construction condition, and the degree of degradation predicted by the end of plant life.

It is not clear what is the current general condition and aging effects, if any, of reactor cavity area materials, structures and components (SCs) and that predicted to the end of plant life to qualify the use of NUREG-1509 in this FE.

The staff needs to identify the current state of materials, structures and components associated with the RV structural support and steel elements of the CBSW for SLR and a prediction of their a) Discuss current general material and structural condition of the CBSW annulus and reactor cavity, CBSW star truss, RV seismic restraint (stabilizers/brackets), RV support skirt, RV support cavity, RV pedestal and anchorages based on past performed inspections.

b) For the SCs discussed above, provide representative photos indicating their current/most recent condition.

c) Discuss periodicity and adequacy of visual inspections and other examinations associated with BFN SLRA AMPs B.2.1.30 and B.2.1.33 particularly in high ALARA areas for the aforementioned SCs (e.g.,

RV support and CBSW steel elements) demonstrating, in accordance with 10 CFR 54.21(a)(3), that the effects of aging will be adequately managed so that the[eir] intended function(s) will be maintained consistent with the CLB for the period of extended operation.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions anticipated degradation to the end of plant life, in this case to the end of the SPEO.

The staff notes, however, that BFN will periodically evaluate the state of metallic SCs associated with this FE through inspections and examinations mandated by ASME Code Section XI, Subsection IWF and the guidance provided by GALL-SLR AMPs XI.S3 and XI.S6 implement by BFN in AMPs B.2.1.30 and B.2.1.33.

C3 (can be discussed with question C4, below) 3.5.2.2.2.6 3.1.2.2.1 Table 3.1.1.

3-698 through 3-699 3-16 SLRA Section 3.5.2.2.2.6 states that the integrity of the reactor vessel supports is assured, and no additional aging management of reactor vessel supports beyond the current ASME Section XI, Subsection IWF program (B.2.1.30) is necessary for aging effects due to irradiation during the subsequent period of extended operation of BFNP.

The SLRA assigns Table 3.1.1 Item Numbers 3.1-1, 004 and 3.1-1, 113 and corresponding Table 3.1.2-1, Reactor Vessel - Summary of Aging Management Evaluation AMR items which consider SLRA AMPs B.3.1.1 Fatigue Monitoring and B.2.1.1 ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD, to manage the aging effects of RV skirt fatigue and for loss of material.

The RV skirt is an ASME Class 1 support and consistent with ASME Section XI, its Inservice Inspection (ISI) is under the jurisdiction of Subsection IWF. The staff reviewed SLRA AMP B.2.1.30 but could not find an exception stating that the effects of aging with regard to acceptance criteria program element of GALL-SLR AMP XI.S3 that include loss of material due to corrosion, cracks or permanent a) Discuss how BFN will use SLRA AMP B.2.1.1 to manage the effects of aging for the ASME Class 1 RV support skirt which is clearly within the 10 CFR 50.55a regulatory review of ASME Section XI, Subsection IWF and GALL-SLR AMP XI.S3 guidance.

b) Discuss why neither the SLRA AMP B.2.1.1 nor its basis document indicate alignment of program elements and/or existence of enhancements to program elements consistent with the GALL-SLR XI.S3 guidance for the RV skirt ISIs and examinations.

c) Clarify apparent discrepancies in SLRA AMPs B.2.1.30 and B.2.1.1 jurisdictions with regard to ISIs and examinations for the RV ASME Class 1 support skirt.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions lower. There is also a lack of clarity whether ventilation has been considered for reduction of temperature, Ionization at the RV annulus and air cavity, and to that within the CBSW.

C9 3.5.2.2.2.6, Table 3.5-1 Table 3.5.2-1 3.5-38, 3.5-74, 3.5-76 SLRA Section 3.5.2.2.2.6 states that the Structures Monitoring program (B.2.1.33) will be used to manage the potential for reduction of strength, loss of mechanical properties, and cracking due to irradiation of concrete near reactor vessel (CBS) during the SPEO.

SLRA Table 3.5-1 item 3.5.1-097 states that [t]he Structures Monitoring program (B.2.1.33) will be used to manage reduction of strength and loss of mechanical properties of the reinforced concrete elements exposed to the specified environment (air indoor uncontrolled) in the Group 4 structure. and includes in Table 3.5.2-2, Primary Containment Structures - Summary of Aging Management Evaluation, two (2) corresponding AMR items crediting the Structures Monitoring Program to manage aging effects of irradiation of concrete, for Group 4 structures and for high density shielding concrete in an air indoor uncontrolled environment.

However, the credited:

a) SLRA AMP B.2.1.33 program description, its program elements and those in the ePortal Program Basis Document (PBD) do not appear to include reduction of strength; loss of mechanical properties due to irradiation (i.e., radiation interactions with material and radiation-induced heating) as aging effects/mechanism that will be managed by the program. Similarly, the AMP and its PBD do not appear to discuss its use to address the effects of aging for high density shielding against high energy line break (HELB).

a) Revise the SLRA AMP B.2.1.33 and its PBD as appropriate to include managing the aging effects/mechanism corresponding to SLRA item 3.5-1, 097 for which the AMP is credited in the SLRA Table 3.5.2-2 AMR items.

b) Consistent with Table 3.5.2-2, revise the Table 1 item 3.5-1, 097 discussion to indicate applicability of the Structures Monitoring program to manage the effects of aging for reduction of strength; loss of mechanical properties due to irradiation (i.e., radiation interactions with material and radiation induced heating) of concrete associated with high energy line break(s) (HELB).

c)

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions neutrons/cm2s. The staff noted that the units of neutrons/cm2s is flux not fluence.

S3 (may be discussed with C8) 3.5.2.2.2.6 3-698 On this page of the SLRA, the applicant states lowest service temperature (LST) of 100°F.

a) It is not clear why this value is different than the normal operating temperature (average, Mode 1) in the drywell of 136°F mentioned on page 3-700 of the SLRA.

b) There is no discussion of the assumption(s) in the normal operating temperature (average, Mode 1) in the drywell of 136°F.

a) Clarify why the LST value of 100°F is different than the normal operating temperature (average, Mode 1) in the drywell of 136°F mentioned on page 3-700 of the SLRA.

b) Discuss the assumption(s) in the normal operating temperature (average, Mode 1) in the drywell of 136°F.

S4 3.5.2.2.2.6 3-699 On this page of the SLRA, the applicant stated that in the material receipt records, the steel elements of the concrete biological shield, consisting of the columns, 1/4-inch-thick steel liner plates, ring girders, and transfer beams, are fabricated from steel conforming to ASTM A36 low carbon steel. The applicant assumed at initial nil-ductility transition temperature (NDTT) plus 1.3 of 39°F from Table 4-1 of NUREG-1509.

Confirm that there were no plant-specific certified material testing reports that contain the plant-specific initial NDTT of the ASTM 36 steel, and that therefore, the value in Table 4-1 of NUREG-1509 was used.

S5 3.5.2.2.2.6 3-700 On this page of the SLRA, the applicant stated that the peak fluence at the shield wall inner surface for Unit 3 64 EFPY (80 years) equates to a displacement per atom = 6.15E-04 dpa. It is noted that Unit 3 provides the most bounding DPA value for use in the sacrificial steel evaluation.

The staff noted that on SLRA page 4-11, it is stated that Unit 1 is projected to reach 50 EFPY in 80 years, Unit 2 is projected to reach 64 EFPY in 80 years, and Unit 3 is projected to reach 62 EFPY in 80 years.

Correct/clarify that the peak fluence at the shield wall inner surface should be for Unit 2 64 EFPY or Unit 3 62 EFPY. If the former, then Unit 2 provides the most bounding dpa value.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions S6 (may be discussed with C4)

FSAR 12.2 Anchorage detail in Table 12.2-10 Need discussion of impact of combined effects of aging associated with irradiation during the SPEO for the steel ring girder assembly at the bottom of the RV skirt.

Discuss with C4 S7 (may be discussed with C7) 3.5.2.2.2.6 3-695 This page of the SLRA states that the central portion of the CBSW is comprised of high-density concrete with hematite aggregate. Need discussion of effect on CBSW steel liner of potential thermal expansion coefficient difference between hematite/concrete fill and steel liners.

Discuss with C7 SLRA Section B.2.1.1 ASME Section XI, Inservice Inspections, IWB, IWC, and IWD Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.1.1 B-12 The SLRA Section related to Operating Experience states, A weld indication was identified during Unit 2 Refueling Outage 21 in 2021 and entered into the Corrective Action Program and evaluated. The results of the evaluation determined that the observed indication is acceptable for continued operation per the requirements of ASME Code,Section XI, IWB-3640 and Appendix A. Limits were established bounding the evaluation and follow up inspections will be performed.

ASME Code,Section XI, Subsection IWB-3640 is related to flaw evaluations of austenitic or ferritic piping, Appendix A is intended for flaw evaluations of ferritic components greater than 4 inches in thickness and is usually not intended for austenitic stainless steel.

Please discuss the component, and materials of construction where the weld indication was discovered in 2021.

Discuss the established limits and results of the follow up inspections.

2 B.2.1.1 B-12 The SLRA Section related to Operating Experience states, A weld flaw was identified during the Unit 1 Refueling Outage 13 in 2020 and entered into the Corrective Action Program and evaluated. An evaluation of the flaw indicated that a weld overlay repair was required. This repair was completed in 2022. This weld is inspected every refueling outage, and therefore will continue to be evaluated for any Please discus the component and materials of construction for the flaw identified in 2020. Why was the weld overlay repair completed in 2022, and how is it inspected every refueling outage?

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions additional degradation.

3 NA NA NRCs Power Reactor Status indicated that Unit 1 was in a recent maintenance outage from 1/25/2024 to 1/30/2024.

Discuss the nature of the outage, and provide details of the outage if it was due age related degradation of any ASME Code, Class 1, 2, or 3 components.

SLRA Section B.2.1.4, BWR Vessel ID Attachment Welds Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.1.4 and A.2.1.4 Use of alternative guidance was identified as an exception to the GALL-SLR report guidance for aging management.

In AMP XI.M4, the GALL-SLR report states, The program includes inspection and flaw evaluation in accordance with the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, and the guidance in BWR Vessel and Internals Project, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines [Boiling Water Reactor Vessel and Internals Project (BWRVIP)-48-A] to provide reasonable assurance of the long-term integrity and safe operation of BWR vessel ID attachment welds.

In its SLRA, TVA proposes to use BWRVIP-48, Revision 2, as the basis for inspecting reactor pressure vessel (RPV) interior attachment welds in lieu of applicable ASME Section XI, Examination Category B-N-2 requirements. In Section 4.4.2 of the basis document for aging management program, BWR Vessel ID Attachment Welds, the applicant states that BWRVIP-48, Rev. 2, is used in lieu of ASME Section XI. The FSAR Supplement included in Appendix A of the SLRA states that the guidance in

a. Clarify that separate NRC approval will be sought to use BWRVIP-48, Rev. 2, as a basis for inspecting RPV interior attachment welds during the SPEO as an alternative to ASME Section XI requirements and the guidance in BWRVIP-48-A, through the 10 CFR 50.55a alternative request process.

b. In the SLRA, the applicant proposes to use the guidance in BWRVIP-48, Rev. 2, as a substitute for the requirements of ASME Sec. XI.

Discuss whether the FSAR supplement needs to be modified.

c. As an exception to the GALL-SLR, the applicant proposes to use the guidance in BWRVIP-48, Rev. 2, as a substitute for the guidance in BWRVIP-48-A. Explain the changes between BWRVIP-48-A and Revision 2 of the document and the impact of those changes on the effectiveness of associated in-

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions BWRVIP-48, Rev. 2, is to be, substituted for the requirements of ASME Code,Section XI, Table IWB-2500-1, Examination Category B-N-2.

service examinations. Explain how the revised guidelines would continue to ensure that any degradation in vessel ID attachment welds would be detected and properly characterized, consistent with effective aging management of these components.

2 B.2.1.4 The applicant completed a focused self-assessment of the reactor pressure vessel internals inspection (RPVII) program in 2015, with results described in TVA CR Vault Summary Report, 1065478. The self-assessment identified a concern that some industry operating experience (OE) that may be relevant to the RPVII program may not have been reviewed or evaluated for applicability. The report states that actions would be entered into a corrective action program to evaluate recent industry OE and to revise procedure documentation to prevent omissions in the future.

Provide the results of any reevaluation completed on recent industry OE to determine its applicability to the RPVII program. Explain what procedure modifications have been documented and implemented to prevent future omissions of relevant OE from the review and evaluation process.

SLRA Section B.2.1.6, BWR Penetrations Question Number LRA/SLRA Section LRA/SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

SLRA Section B.2.1.6 B-27 The applicant took an exception and proposed an enhancement to GALL-SLR Report AMP XI.M8. Specifically, the exception and enhancement are to use Boiling Water Reactor Vessel and Internals Project (BWRVIP)-57, Revision 1, Instrument Penetration Repair Design Criteria, not approved by NRC in place of BWRVIP-57-A. The applicant provided justification as follows:

Section 1 and Appendix A of BWRVIP-57-A were revised in Revision 1 to add recent operational experience and descriptions of additional instrument penetration nozzle repairs that have been or could be implemented in the BWR fleet. The aging management

a. In a supplement to the submittal, identify specific contents, statements, paragraphs, and pages in BWRVIP-57, Revision 1 that are different than the ones in BWRVIP A.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions requirements in BWRVIP-57, Revision 1 were screened using NEI-03-08, Revision 4, Appendix C, Document Screening, resulting in a determination that it could be generically released for implementation by U.S. BWRVIP members without prior U.S. NRC review and approval.

The staff notes that the applicant provided BWRVIP-57, Revision 1 on the portal rather than as supplement to SLRA for docketing as part of staffs review of the applicants exception and enhancement to GALL-SLR Report AMP XI.M8.

The staff also notes that the applicants justifications for exception and enhancement lack specifics. For example, the applicant did not identify specific contents, statements, paragraphs, and pages in BWRVIP-57, Revision 1 that are different from BWRVIP-57-A. For each identified difference, the applicant should provide discussions on whether the difference is a technical (e.g., inspection, evaluation, frequency, and/or design related) or as an administrative (e.g., OE related) in nature. If the identified difference is a technical, the applicant needs to provide justification to demonstrate that the effects of aging on reactor vessel penetrations at Browns Ferry units will be adequately managed so that the intended function(s) will be maintained consistent with the CLB for the subsequent period of extended operation, as required by 10 CFR 54.21(a)(3).

According to page xxxix of GALL-SLR Report titled Guidance on Use of Later Editions/Revisions of Various Industry Documents, the applicant may use later editions/revisions of an industry generated documents subject to the following provisions:

i.

If the later edition/revision has been explicitly reviewed and approved/endorsed by the NRC staff for license renewal via a NRC Regulatory Guide endorsement, a safety evaluation for generic use such as BWRVIP, incorporation into 10 CFR, or license renewal interim staff guidance.

ii.

If the later edition/revision has been explicitly reviewed and approved on a plant-specific basis by the NRC staff in its Safety

b. For identified differences, provide discussions on whether the difference is a technical (e.g.,

inspection, evaluation, frequency, and/or design related) or administrative (e.g.,

OE related) in nature.

c. If the identified difference is a technical, the applicant should demonstrate that the effects of aging on reactor vessel penetrations at the Browns Ferry units will be adequately managed so that the intended function(s) will be maintained consistent with the CLB for the subsequent period of extended operation, as required by 10 CFR 54.21(a)(3).

d. Provide BWRVIP-57, Revision 1 as a supplement to SLRA for docketing and

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Evaluation Report for another applicants SLRA (a precedent exists). Applicants may reference this and justify applicability to their facility via the exception process in Nuclear Energy Institute 95-10.

If either of these methods is used as justification for adopting a later edition/revision than specified in the GALL-SLR Report, the applicant shall reference the information pertaining to the NRC endorsement/approval of the later edition/revision.

According to Section 3.1.2.4, Aging Management Programs, of GALL-SRP Report, if the applicant identifies an exception to any of the program elements of the cited GALL-SLR Report AMP, the SLRA AMP should include a basis demonstrating how the criteria of 10 CFR 54.21(a)(3) would still be met. The reviewer should then confirm that the SLRA AMP with all exceptions would satisfy the criteria of 10 CFR 54.21(a)(3). The reviewer should document the disposition of all SLRA-defined exceptions. Furthermore, the SLRA should identify any enhancements that are needed to permit an existing licensee AMP to be declared consistent with the GALL-SLR Report AMP to which the licensee AMP is compared. The reviewer is to confirm both that the enhancement, when implemented, would allow the existing licensee AMP to be consistent with the GALL-SLR Report AMP and that the applicant has a commitment in the FSAR Supplement to implement the enhancement prior to the subsequent period of extended operation.

The reviewer should document the disposition of all enhancements.

staffs review of the applicants exception and enhancement to GALL-SLR Report AMP XI.M8.

2 SLRA Section B.2.1.6 B-27 The applicant indicated that the Browns Ferry BWR Penetrations program augments the requirements of ASME Code,Section XI by incorporating the inspection and flaw evaluation recommendations of NRC-approved BWRVIP guidelines.

According to NRC safety evaluations of BWRVIP-27-A, BWRVIP-47-A, BWRVIP-49-A, BWRVIP-53-A, BWRVIP-55-A, BWRVIP-57-A, and BWRVIP-58-A, the applicant can use NRC-approved BWRVIP guidelines as an alternative to ASME Code,Section XI provided that a request for alternative pursuant to 10 CFR 50.55a(z) to be submitted and approved for each 10-year ISI interval.

a. Supplement SLRA Section B.2.1.6 to state that a proposed alternative request for use of the BWRVIP guidelines will be submitted to NRC approval during the Browns Ferry subsequent period of extended operation

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The staff notes that the Browns Ferry BWR Penetrations program falls short stating that: (a) an alternative request has been submitted and approved by the NRC for use of the above BWRVIP guidelines for each 10-year ISI interval of Browns Ferry units during the period of extended operation, and (b) a proposed alternative request for use of each of the BWRVIP guidelines will be submitted to NRC approval for each 10-year ISI interval during the subsequent period of extended operation.

b. In the portal, provide the NRC safety evaluations of a request for alternative to use of the BWRVIP guidelines during the Browns Ferry period of extended operation.

SLRA Section AMP: B.2.1.7 BWR Vessel Internals Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

A.2.1.7 A-10 The FSAR supplement includes enhancement 2 to implement BWRVIP-315-A and subsequent revisions approved by the NRC but does not discuss the current state of the guidance, namely the draft NRC BWRVIP-315 safety evaluation and the BWRVIP comments to the draft submitted by letter dated January 20, 2022 or that the NRC has issued a final safety evaluation which includes 5 conditions to be incorporated into BWRVIP-315-A when issued.

Can the FSAR supplement be updated to include a discussion of the ongoing development of guidance for the subsequent period of extended operations, similar to MNGP SLRA Supplement 2 (ML23177A218)?

2 A.2.1.7 B.2.1.7 A-11 B-40 Enhancement 6 doesnt contain any of the discussion in Exception 6 about advanced manufacturing processes.

Clarify whether the intent of Enhancement 6 is to speak to BWRVIP-84 generally or advanced manufacturing specifically.

3 B.2.1.7 B-25 Exception 2 of the BWR Vessel Internals AMP proposes to use BWRVIP-62 Revision 2 in lieu of BWRVIP-62 Revision A.

Discuss which primary and secondary parameters are used by Browns Ferry to monitor

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions water chemistry and whether any of parameters required new analyses due to the extended power uprate.

4 B.2.1.7 Appendix C

B-31 B-40 C-1 C C-32 The application discusses the applicants response to the draft NRC safety evaluation of BWRVIP-315 of January 20, 2022. The NRC issued a final safety evaluation October 31, 2023 (ML23251A056).

Discuss how Browns Ferry meets the five conditions specified in the NRC final safety evaluation of BWRVIP-315 in lieu of the two license conditions and four applicant actions items discussed in Appendix C.

5 B.2.1.7 B-38 The applicant intends to utilize BWRVIP-100 Revision 2 which is still undergoing NRC review. The BWRVIP issued a Part 21 notice (ML21084A164).

How did Browns Ferry respond to the Part 21 notice?

6 B.2.1.7 B-39 The applicant proposes to use BWRVIP-84 Revision 3 in lieu of Revision 2-A. The major differences discussed are the addition of significant technical detail related to advanced manufacturing processes that arent captured by current ASME B&PV Code or Code Cases.

Does the applicant intend to use the allowances in BWRVIP-84 Revision 3 to manufacture components using processes that are not already controlled by ASME Code?

7 B.2.1.7 B-40 Enhancement 3 discusses revising implementing procedures to incorporate a requirement but doesnt cite what requirement is being referred to.

Explain the source of requirements being discussed in Enhancement 3.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 8

Appendix C Action Items BWRVIP-76-R1 (8)

C-17 BWRVIP-76R1-A action item (8) requires the applicant to reference NRC approved topical reports BWRVIP-99-A and BWRVIP-100-A. The Browns Ferry response states that the BWR Vessel Internals AMP is being enhanced to explicitly reference BWRVIP-14-A, BWRIVP-99-A, and BWRVIP-100-A.

BFN BWR Vessel Internals AMP exception 4 and enhancement 4 both refer to BWRVIP-100 Revision 2. Other action items (e.g. BWRVIP-76R1-A (4)) reference BWRVIP-100-A but also discuss program guidance to incorporate new or revised BWRVIP reports. BWRVIP R1-A action item 8 does not discuss this guidance or BWRVIP-100 Revision 2.

Should the discussion in BWRVIP-76R1-A action item (8) refer to BWRVIP-100 Revision 2?

9 Appendix C

Action Item BWRVIP-139-R1-A (2)

C-20 The applicant notes that citing report BWRVIP-139-R1-A (and Appendix B of the report) in the FSAR would be done to satisfy 10 CFR 54.21(d) but BWRVIP-139-R1-A does not apply to Browns Ferry so no discussion of the steam dryer aging management program is included in Section A.2.1.7.

Can a short summary of the aging management plan described in the Browns Ferry Response to the action item, or as discussed in Exception 1 to B.2.1.7, be included in the FSAR to meet the intent of 10 CFR 54.21(d)?

10 Appendix C

Action Item BWRVIP-41-R4-A BWRVIP-315 Applicant Action Item 4

C-25 C-31 The discussion in Action Item BWRVIP-41-R4-A and BWRVIP-315 Applicant Action Item 4 discuss the guidance related to degradation mechanisms at high fluences and determine that the enhancements and revisions to BWRVIP-41-R4-A identified in BWRVIP-315 are not needed for the subsequent period of extended operation.

There is no discussion of BWRVIP-315 Section 4.5.1 Limitation 4 which states that the scope expansion exemption for large diameter jet pump diffuser, adapter, and lower ring welds inspected by UT is based on an assumption of a 60-year service life and that the exemption should be limited to plants not intending to operation beyond 60 years.

Discuss how Browns Ferry captures the BWRVIP-315 Section 4.5.1 limitation or whether a commitment to enhance the BWR Vessel Internals AMP prior to entering the subsequent period of extended operations by following the latest revision of BWRVIP-41 is necessary.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 11 Appendix C

Action Item BWRVIP-47-A BWRVIP-315 Applicant Action Item 3

C-27 C-30 Action Item BWRVIP-47-A states that guidance provided in BWRVIP-47-A is incorporated into the BWR vessel internals program and recommended changes are clarifying in nature, so no action is required.

BWRVIP-315 Applicant Action Item 3 further notes that Section 4.3.5.3 of BWRVIP-315 states that BWRVIP will issue a set of revised guidelines to address the subsequent period of extended operations published two years prior to the first BWR plant operating beyond 60 years.

BWRVIP-315 Section 4.5.1 Limitation 2 states in part that:

owners submitting an application for operation beyond 60 years should commit to implementing a future version of BWRVIP-47 that addresses extended operations or propose a set of plant-specific activities to manage age-related degradation of CRGTs.

The NRC issued a final safety evaluation October 31, 2023 (ML23251A056) which includes condition 3 which states:

Applicants for renewed operating licenses extending beyond 60 years must describe and justify in the application plant-specific re-inspection plans for the lower plenum in the application.

E.g. see MNGP SLRA Supplement 4 Enclosure 4 Pg 6 of 9 (ML23199A154)

Describe and justify the plant-specific re-inspection plans for the lower plenum and discuss whether a commitment to enhance the BWR Vessels Internals AMP by following the latest revision of BWRVIP-47 is necessary.

SLRA Section B.2.1.8, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

SLRA Table 3.2.1 3.199 SLRA Table 3.2.1, AMR Item 3.2-1, 010 shows consistency with NUREG-2191, however there are no Table 2 AMR items associated with this Table 1 AMR item.

It is not clear to the staff whether the above Table 1 AMR item is applicable, not applicable, or not used. If the above Table 1 item is Provide Table 2 AMR items for the Table 1 AMR item.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions used, what the associated Table 2 AMR items are for the Table 1 AMR item.

SLRA Section B.2.1.27: Buried and Underground Piping and Tanks Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

B.2.1.27 B-123 GALL-SLR Report AMP XI.M41, Buried and Underground Piping and Tanks, states [i]f cathodic protection is not provided for any reason, the applicant reviews the most recent 10 years of plant-specific operating experience (OE) to determine if degraded conditions that would not have met the acceptance criteria of this AMP have occurred.

This search includes components that are not in-scope for license renewal if, when compared to in-scope piping, they are similar materials and coating systems and are buried in a similar soil environment [emphasis added by staff]. The results of this expanded plant-specific OE search are included in the SLRA.

SLRA Section B.2.1.27, Buried and Underground Piping and Tanks, states based on review of BFN-specific operating experience, no leaks in the subject buried piping due to external corrosion have been observed and no significant buried piping coating degradation has been observed.

During its audit, the staff noted several instances of buried piping leaks and an instance of buried piping not being coated in accordance with design specifications. The first three bullets are related to buried service air system piping (not in-scope based on the systems listed in SLRA Section B.2.1.27 (page B-119)) and the last two bullets are related to buried fire protection system piping (in-scope based on the systems listed in SLRA Section B.2.1.27 (page B-119)).

The staff reviewed CR 793555 and noted the suspected cause of the leak was corrosion similar to that seen in two service air piping leaks earlier in 2013.

The staff requests a discussion with respect to the following:

1. Whether the OE noted by the staff during its audit (i.e., instances of through-wall leakage due to external corrosion) is representative of the condition of in-scope piping.

2. If out-of-scope service air piping is constructed from similar materials and coating systems when compared to in-scope piping.

3. If out-of-scope service air piping is exposed to a similar soil environment

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The staff reviewed CR 1120048 and noted the source of the leak was determined to be service air and not fire protection.

The staff reviewed CR 898407 and noted (a) a section of service air piping had no visible exterior tape coating or polyethylene coating as required by design drawings; and (b) the uncoated section of piping had multiple thru wall holes.

The staff reviewed CR 828934 an noted it was indeterminate whether the leak was coming from a slipjoint between sections of pipe or a hole in the pipe wall.

The staff reviewed CR 1102016 and noted that a contributing cause of a break in buried fire protection piping was outer diameter graphitic corrosion.

when compared to in-scope piping.

4. Why out-of-scope buried piping not being coated in accordance with design specifications would also not be applicable to in-scope piping.

2 B.2.1.27 B-121 B-122 B-123 B-124 SL-016653, Evaluation of Cathodic Protection for Buried Piping, notes that a site-wide impressed current cathodic protection (ICCP) system is not likely to be successful; however, it also notes that protecting a small scope of buried piping that is high-risk due to the pipe construction, process fluid, or the soil conditions in the area is likely achievable. A similar discussion is provided in BP-2023-0027-03-TR, Browns Ferry License Renewal Buried Piping Cathodic Protection Review.

The discussion provided in SLRA Section B.2.1.27 focuses on why installation of a site-wide ICCP system is impractical; however, it appears that protecting a limited scope of in-scope piping may be practical.

The staff requests a discussion on this topic.

3 B.2.1.27 B-131 SLRA Section B.2.1.27 states [f]rom the 2009 soil sample APEC survey there were 6 recommended inspection locations that could possibly reflect localized areas of potential coating damage and active corrosion. It was recommended that these locations be excavated and directly inspected. There were also 6 locations recommended for excavation and direct examination based on the 8 soil samples from 2023 combined with the 48 total native potential measurements recorded as CIS.

The 2009 and 2023 surveys both recommended six inspection locations that could possibly reflect localized areas of potential coating damage and active corrosion. The staff notes that this also corresponds to the recommended number of

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions inspections for Preventive Action Category E in GALL-SLR Report Table XI.M41-2, Inspection of Buried and Underground Piping and Tanks (adjusted for a three-unit site). Since coatings and backfill are the only barriers to aging, and coatings will continue to degrade over time, the staff requests a discussion with respect to why increased inspections quantities in each successive 10-year interval (i.e., more inspections in the 70-80 year interval when compared to the 60-70 year interval, more inspections in the 60-70 year interval when compared to the 50-60 year interval) are not necessary to provide reasonable assurance.

4 B.2.1.27 B-137 Enhancement No. 2 shows two inspections for steel piping exposed to concrete, and two inspections for stainless steel piping exposed to concrete.

It is unclear to the staff where two inspections come from. The staff requests a discussion on this topic.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 5

B.2.1.27 B-141 Enhancement No. 30 states [r]evise implementing procedures to state that opportunistic examinations of non-leaking pipes may be credited toward examinations if the location selection criteria are met Does this imply that opportunistic inspections could replace the targeted inspections in SLRA Table B.2.1.27-1 and SLRA Figure B.2.1.27-1 (i.e., digs #1 through #12)? The staff requests a discussion on this topic.

6 B.2.1.27 B-137 B-140 Enhancement No. 3 states [r]evise implementing procedures to require that soil testing using the guidance in the EPRI Report 3002018353, Revision 2, Buried and Underground Piping and Tank Reference Guide, be conducted in conjunction with the periodic direct inspections of buried piping.

Enhancement No. 24 states [r]evise implementing procedures to require that BFN will take soil samples prior to planned excavations to confirm that the chemistry of the backfill is nonaggressive.

GALL-SLR Report AMP XI.M41 defines non-corrosive soil as nine points or less using AWWA C105, Polyethylene Encasement for Ductile-Iron Pipe Systems, Table A.1, Soil Test Evaluation. Draft Revision 1 to the GALL-SLR Report added 10 points or less using EPRI Report 3002005294, Soil Sampling and Testing Methods to Evaluate the Corrosivity of the Environment for Buried Piping and Tanks at Nuclear Power Plants, Table 9-4, Soil Corrosivity Index from BPWORKS, as an additional definition of non-corrosive soil. The staff requests a

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions discussion with respect to how non-corrosive soil will be determined.

7 B.2.1.27 B-124 SLRA Section B.2.1.27 states [t]he steel pipe is coated in accordance with American Water Works Association standard, AWWA C203-66, Section A1.5 - Coal-Tar Enamel, Fibrous-Glass Mat, and Bonded Asbestos-Felt Wrap.

The staff requests a discussion with respect to if buried copper alloy piping is externally coated in accordance with AWWA C203, if another coating system applies to this material, or if this piping is not coated.

8 N/A N/A It is the staffs understanding that buried stainless steel piping is not externally coated. GALL-SLR Report AMP XI.M41 recommends external coatings for buried stainless steel piping in chloride containing environments; however, soil corrosivity data from 2009 (SLRA Page B-127) and 2023 (SLRA Page B-128) shows low levels of chlorides at BFN.

The staff requests a discussion with respect to the following: (a) whether the soil corrosivity data collected in 2009 and 2023 was in the vicinity of in-scope uncoated buried stainless steel piping; and (b) whether deicing salts have been used in the vicinity of in-scope uncoated buried stainless steel piping.

9 B.2.1.27 B-119 B-136 SLRA Section B.2.1.27 includes the following systems in the scope of the program: condensate demineralized water, residual heat removal, raw service water, high pressure fire protection, condenser circulating water, containment, standby gas treatment, emergency equipment cooling water, radwaste, containment atmosphere dilution, and hardened containment venting.

The staff notes that the systems identified in SLRA Section B.2.1.27 are either included in the initial LRA or are included in Enhancement No. 1 to the SLRA, except for the hardened

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Enhancement No. 1 states [r]evise implementing procedures to explicitly include the Condensate Demineralized Water System and the Radwaste System within the scope of the SLR Buried and Underground Piping and Tanks Aging Management Program.

containment venting system. The staff requests a discussion with respect to why this system is not included in Enhancement No. 1.

10 B.2.1.27 B-137 Enhancement No. 2 includes a table of inspections for steel and stainless steel piping exposed to soil and concrete.

The staff requests a discussion with respect to why underground steel piping and buried copper alloy piping are also not included in this table.

11 B.2.1.27 B-142 Enhancement No. 41 states [r]equire that the section of underground carbon steel piping (in an isolation valve pit) in the Hardened Containment Venting System which is not coated consistent with GALL SLR Element 2 for underground steel piping, will be coated in accordance with Table 1 of NACE SP0169-2007 prior to the subsequent period of extended operation [emphasis added by staff].

The inspection quantities listed in GALL-SLR Report Table XI.M41-2 are based on meeting the preventive actions identified in GALL-SLR Report Table XI.M41-1, Preventive Actions for Buried and Underground Piping and Tanks.

Based on this, it is unclear to the staff how the inspection quantities listed in GALL-SLR Report Table XI.M41-2 are appropriate for underground steel piping in the 10-year interval prior to the SPEO, since this underground piping will not be coated during this 10-year interval. The

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions staff requests a discussion on this topic.

12 B.2.1.27 B-120 SLRA Section B.2.1.27 states [t]he GALL-SLR defines nonaggressive groundwater and soil as having pH > 5.5, chlorides < 500 ppm, and sulfates <1,500 ppm. Based on the GALL-SLR definition, the soil at BFN is nonaggressive.

The following criteria come from GALL-SLR Report AMP XI.S6, Structures Monitoring.

It is not clear to the staff why either AWWA C105 (Table A.1) or EPRI Report 3002005294 (Table 9-4) are not referenced in lieu of these criteria (see breakout question No. 6 above). The staff requests a discussion on this topic.

13 Table 3.2.2-7 3-269 SLRA Table 3.2.2-7, Containment Atmosphere Dilution System -

Summary of Aging Management Evaluation, includes buried stainless steel piping which will be managed for loss of material using the Buried and Underground Piping and Tanks program.

The staff requests a discussion with respect to why cracking due to SCC is also not cited.

SLRA Section AMP: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

B.2.1.28 B-144 B-146 SLRA Section B.2.1.28, Program Description (page B-144),

says the in-scope systems managed by this aging management program (AMP) are the High Pressure Fire Protection (Diesel Driven Pump) System, Control Air System, Service Air System, Condensate/Demineralized Water System, and Residual Heat Removal Service Water (RHRSW) System. SLRA Section B.2.1.28, Operating Experience (page B-146), describes four examples from the RHRSW System and states that these Discuss operating experience, from the following in-scope systems if available, that provides objective evidence that SLRA Section B.2.1.28 AMP inspections will be effective in identifying and managing aging effects and loss

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions examples provide objective evidence that the program inspections will be effective in identifying and managing aging effects.

of coating integrity for the following in-scope systems:

High Pressure Fire Protection (Diesel Driven Pump) System Control Air System, Service Air System, Condensate/Demineralized Water System 2

B.2.1.28 B-146 Condition Reports (CRs) 1272939, 1272516, and 1272948 on the Portal further describe loss of coating integrity in the 2A RHRSW Heat Exchanger which was observed during a 2017 inspection.

Discuss operating experience learned from the inspections described in CRs 1272939, 1272516, and 127948. Discuss how similar, potentially flow-blockage inducing loss of coating integrity of RHRSW components will be managed during the period of extended operation.

3 Table 3.4.1 3-573 Table 3.4.1, Summary of Aging Management Evaluations for the Steam and Power Conversion Systems, of the Browns Ferry SLRA specifies that for Item 3.4-1, 066, the applicants Outdoor and Large Atmospheric Metallic Storage Tanks program (stated to be consistent with GALL-SLR AMP XI.M29 with one exception) has been substituted for the applicants Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program (stated to be consistent with GALL-SLR AMP XI.M42) to manage loss of coating integrity of carbon steel tanks with internal coating exposed to treated water in the condensate/demineralizer water system. NRC staff note the following issues with the applicants proposed AMP substitution:

Identify any internally coated carbon steel tanks for which loss of coating integrity is to be managed by an SLRA AMP.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions a) Table 3.4.2-2, Condensate/Demineralized Water System - Summary of Aging Management Evaluation, of the Browns Ferry SLRA does not list any internally coated carbon steel tanks for which the loss of coating integrity is to be managed by any SLRA AMP.

b) Contrary to the scope of GALL-SLR Report AMP XI.M29, the SLRA AMP for Outdoor and Large Atmospheric Metallic Storage Tanks does not contain the recommendations of GALL-SLR Report AMP XI.M42 (including exceptions and enhancements if applicable).

c) Contrary to the scope of GALL-SLR Report AMP XI.M29, the FSAR supplement for the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program does not include a reference to the Outdoor and Large Atmospheric Metallic Storage Tanks program.

4 Table 3.4.1 3-573 See question 3.

For any internally coated carbon steel tanks identified in response

1. 3 above, specify which SLRA AMP will be used to manage for loss of coating integrity.

5 Table 3.4.1 3-573 See question 3.

For any internally coated carbon steel tanks using the SLRA AMP for Outdoor and Large Atmospheric Metallic Storage Tanks to manage for loss of coating integrity, explain how the recommendations of GALL-SLR Report AMP XI.M42 (including

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions exceptions and enhancements if applicable) are incorporated into the SLRA AMP.

6 Table 3.4.1 3-573 See question 3.

If internally coated carbon steel tanks are being managed for loss of coating integrity using the Outdoor and Large Atmospheric Metallic Storage Tanks program, amend the FSAR supplement for the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program to reference the Outdoor and Large Atmospheric Metallic Storage Tanks program.

SLRA Section B.2.1.9 Flow-Accelerated Corrosion Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

N/A N/A In the Flow Accelerated Corrosion Program - System Susceptibility Evaluation (SSE) for each unit identifies the systems that are susceptible to FAC. The following documents include System Code UN and System Name Unspecified that is identified as susceptible to FAC.

Documents:

17-0291-TR-006, Rev. 0 o PDF Page 36 17-0291-TR-011, Rev. 0 o PDF Page 36 17-0291-TR-001, Rev. 0 PDF Page 39 Please define System Code UN and System Name Unspecified that is susceptible to FAC.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions probe. No inspection for Unit 2 and 3 were noted. However, there was no discussion of the results of the Unit 1 stick and probe inspection.

Additionally, Task Report SLR-BFN-0091 states, Based on a review of industry actions, BFN will review Radiation Protection (RP) requirements and dose levels for inspecting the bottom head drain elbow and associated piping downstream of the bottom head drain in lower dose areas during the 2020 - U3R19 outage. Additionally, review of remote inspection methods will be pursued to include crawlers, ultrasonic thickness - UT on a stick, and guided wave ultrasonic technology.

associated with the inspection.

Have any inspections on the Unit 2 and 3 bottom head drain been identified since Revision 0 of Task Report SLR-BFN-0091. If so, please provide the results of the inspections.

Have the reviews noted in Task Report SLR-BFN-0091 been completed and, if so, what were the outcomes?

Please provide the isometric drawings of the bottom head drain piping for each Unit.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA Section /AMP: XI.M20 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

Table 3.2.1 3-202 SLRA Table 3.2.1, item 3.2.1, 023 addresses steel [emphasis added]

components exposed to raw water and notes this item is Not Applicable and states, There are no stainless steel [emphasis added]

heat exchanger components, piping, or piping components exposed to raw water in the ESF and RCIC Systems. It is unclear whether item 3.2-1-023 is not applicable, because the discussion addresses stainless steel and does not address steel components.

Clarify the proper material classification for item 3.2-1, 023.

2 Table 3.2.1 3-203 SLRA Table 3.2.1, item 3.2-1, 025 addresses stainless steel heat exchanger components exposed to raw water and notes this item is Not Applicable and states, There are no stainless steel heat exchanger components exposed to raw water in the ESF and RCIC Systems. However, SLRA Table 3.1.2-4, Reactor Recirculation System, includes this item for the stainless steel variable frequency drive heat exchangers and cites generic note B (indicating there are exceptions to the aging management program) for the Open-Cycle Cooling Water program. It is unclear whether item 3.2-1, 025 is not applicable, because even though there are no associated components in the ESF and RCIC systems, this item is being used in the reactor recirculation system. Also, it is not clear what exception is associated with the Open-Cycle Cooling Water System program associated with the generic note B Clarify the proper applicability for item 3.2-1, 025 in Table 3.2.1.

Also, clarify the exception that makes item 3.2-1, 025 in Table 3.1.2-4 a Note B instead of a Note A, since there are no stated exceptions in the Open-Cycle Cooling Water AM 3

Table 3.2.1 3-203 SLRA Table 3.2.1, item 3.2-1, 027 addresses steel and stainless steel heat exchanger tubes exposed to raw water and notes this item is Not Applicable and states, There are no stainless steel heat exchanger tubes exposed to raw water in the ESF and RCIC systems. However, SLRA Table 3.3.2-5, Raw Water Cooling System, includes this item for steel heat exchanger tubes. It is unclear whether item 3.2-1, 027 is not applicable, because even though there are no associated components in the ESF and RCIC systems, this item is being used in an auxiliary system and there is no discussion about steel heat exchanger tubes Clarify the proper applicability for item 3.2-1, 027 in Table 3.2.1 making sure to address both stainless steel and steel materials.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA Section: B.2.1.15 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

2.0, 3.0, B.2.1.15 2-31, 2-90, 3-7, 3-343, 3-417, 3-445, B-77 SLRA Table 2.2-1 states that the halon fire protection system is not in scope of SLR.

SLRA Table 2.3.3-7 includes component type Halon/carbon Dioxide Fire Suppression System Piping, Piping Components.

SLRA Table 3.0-1 states the following in the description of gas: The GALL-SLR Report AMP XI.M26, Fire Protection, is used for the periodic inspection and testing of the halon/carbon dioxide fire suppression system.

The discussion of AMR item 3.3-1, 058 in Table 3.3.1 does not indicate that the halon fire suppression system is not in scope.

SLRA Tables 3.3.2-7 and 3.3.2-13 include component type Halon/carbon dioxide fire suppression system piping, piping components.

Under the Operating Experience section in SLRA Section B.2.1.15, it states, The program also includes periodic inspection and test of halon/carbon dioxide fire suppression systems, and This operating experience provides objective evidence that the Fire Barrier, Diesel-Driven Fire Pumps and Halon / Carbon Dioxide Fire Suppressions systems are effectively being monitored and tested to assure that these components will continue to perform their intended Functions.

The staff recognizes that the use of halon/carbon dioxide may have been because it is used in GALL-SLR, however, where it is used, as described above, could be considered inconsistent with SLRA Table 2.2-1, without additional clarifying statements.

Please confirm that the halon fire suppression system is not in scope of SLR. Please discuss SLRA changes to clarify whether the halon fire suppression system is in scope or not.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The staff notes that SLR-BFN-0060, Rev. 2, Aging Management Program Basis Document - Fire Protection, also similarly refers to halon/carbon dioxide fire suppression systems (4.1.2, 4.5.2, 4.10.2, Table 2).

2 2.0, 3.0 2-94, 2-97, 2-102, 3-428, 3-431, 3-445 SLRA Tables 2.3.3-9 and 2.3.3-10 include component type ducting, ducting components with intended functions of pressure boundary, fire barrier, and structural support. However, SLRA Tables 3.3.2-9 and 3.3.2-10 do not include component type ducting, ducting components, with a fire barrier intended function.

SLRA Table 3.3.2-13 cites AMR item 3.3-1, 255 for managing the affects of aging for the steel fire damper assemblies by the Fire Protection program. However, SLRA Table 2.3.3-13 does not include component type fire damper assemblies.

The fire damper assemblies in SLRA Tables 3.3.2-9 and 3.3.2-10 have a fire barrier and structural support intended functions, and pressure boundary, fire barrier, and structural support intended functions, respectively. Only the Fire Protection program is cited to manage the effects of aging associated with these intended functions.

The discussion of AMR item 3.3-1, 255 in SLRA Table 3.3.1 does not state it is used for Structural Commodities (Hazard Barriers and Elastomers).

Please discuss the following:

1. Whether the component type ducting, ducting components for the normal ventilation and air conditioning systems have a fire barrier intended function and should also be managed by the Fire Protection program.

2. Why SLRA Table 2.3.3-13 does not include component type fire damper assemblies.

3. Discuss how the Fire Protection program will manage the effects of aging for the intended functions, other than the fire barrier intended function, or whether another AMP (e.g., Structures Monitoring program) should be credited to

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions manage the intended functions, other than the fire barrier intended function.

4. Why the discussion of AMR item 3.3-1, 255 in SLRA Table 3.3.1 does not state it is also used for Structural Commodities (Hazard Barriers and Elastomers).

3 2.0, 3.0 2-94, 297, 3-429 SLRA Table 2.3.3-9 includes component type piping, piping components with the following intended functions: pressure boundary, fire barrier, and structural support. However, SLRA Tables 3.3.2-9 does not include component type piping, piping components, with a fire barrier intended function.

SLRA Table 2.3.3-10 includes component type external surfaces with a fire barrier intended function. However, SLRA Table 3.3.2-10 does not include component type external surfaces with a fire barrier intended function.

Please discuss whether the component types piping, piping components and external surfaces in SLRA Tables 3.3.2-9 and 3.3.2-10, respectively, have a fire barrier intended function and should also be managed by the Fire Protection program.

4 2.0, 3.0 2-179, 2-220, 2-214, 3-919 SLRA Section 2.4.10 states, Fire barriers (doors, dampers, fire rated enclosures, fire proofing material, penetration seals, fire barrier function of walls and slabs) are evaluated with the Fire Protection System. In addition, SLRA Section 2.4.34 states, Fire barriers (doors, dampers, fire rated enclosures, fire proofing material, penetration seals, fire barrier function of walls and slabs - including oil retaining dikes and walls that separate transformers) are evaluated with the Fire Protection System. With the exception, of fire rated enclosures, these components appear to be evaluated with the Structural Commodities (Hazard Barriers and Elastomers).

Please discuss whether the statements in SLRA Sections 2.4.10 and 2.4.34 should have stated the fire barriers are evaluated with the Structural Commodities (Hazard Barriers and Elastomers) instead of evaluated with the Fire

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions With regards to fire rated enclosures, SLRA Tables 2.4-26 and 3.5.2-36 do not include component type fire rated enclosures.

Protection System. In addition, discuss why component type fire rated enclosure is not in SLRA Tables 2.4-26 and 3.5.2-36.

5 2.0, 3.0 2-102, 3-286, 3-311, 3-445 SLRA Table 3.3.2-13 includes component types not included in SLRA Table 2.3.3-13. For example, fire damper assemblies, halon/carbon dioxide fire suppression piping, piping components, and piping elements. The staff recognizes that some of the component types used in SLRA Table 3.3.2-13 are from the GALL-SLR. However, it may appear as if component types were omitted from scoping and screening of the CO2 Storage, Fire Protection/Purge System when these tables are not consistent.

SLRA Section 3.3.2.1.13 states that the CO2 Storage, Fire Protection/Purge System components are exposed to an environment of any. However, the environment of any is not used in SLRA Table 3.3.2-13. The staff recognizes that this could have been associated with AMR Items 3.3-1, 002 and 3.3-1, 015 which have an any environment in GALL-SLR.

SLRA Section 3.3.2.2.1 states that AMR Item 3.3-1, 002 evaluates piping, piping components exposed to diesel exhaust for cumulative fatigue damage due to fatigue, and states it was evaluated for the CO2 Storage, Fire Protection/Purge System. SLRA Table 3.3.2-13 cites cumulative fatigue damage as an applicable aging effect for steel piping, piping components exposed only to air in the CO2 Storage, Fire Protection/Purge System. SLRA Section 2.3.3.13 states that the CO2 Storage, Fire Protection/Purge System mitigate consequences of a fire in the Diesel Generator Building, therefore, diesel exhaust may be an applicable environment for components of this system. However, SLRA Table 3.3.2-13 does not cite diesel exhaust as an applicable environment for any components. In addition, Section 3.3.2.1.13 does not identify diesel exhaust as an applicable environment.

Please discuss the following:

1. Why the component types are different in SLRA Tables 2.3.3-13 and 3.3.2-13.

2. How the environment of any is used for the CO2 Storage, Fire Protection/Purge System components.

3. Whether diesel exhaust is an applicable environment for CO2 Storage, Fire Protection/Purge System components and should be added to SLRA Section 3.3.2.1.13 and SLRA Table 3.3.2-13.

4. Why SLRA Tables 2.3.3-13 and 3.3.2-13 do not include lubricating oil or fuel oil as applicable environments for CO2 Storage, Fire

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA Section 2.3.3.13 states that the CO2 Storage, Fire Protection/Purge System contain lubricating oil or fuel oil components.

However, SLRA Tables 2.3.3-13 and 3.3.2-13 does not include lubricating oil or fuel oil as applicable environments for CO2 Storage, Fire Protection/Purge System components.

The steel fire damper assemblies do not have a fire barrier intended function in SLRA Table 3.3.2-13. However, the effects of aging are being managed by the Fire Protection program.

Protection/Purge System components.

5. Whether the fire damper assemblies in SLRA Table 3.3.2-13 should have a fire barrier intended function. If the fire barrier intended function is in addition to the pressure boundary intended function, discuss how the Fire Protection program will manage the effects of aging for the pressure boundary intended function, or whether another AMP should be credited to manage the pressure boundary intended function.

6 3.0 3-344, 3-921 AMR item 3.3-1, 057 in GALL-SLR manages hardening, loss of strength, shrinkage due to elastomer degradation for elastomer fire barrier penetration seals.

SLRA Table 3.5.2-36 cites AMR item 3.3-1, 057, with notes F, 15, for grout fire barrier penetration seals. Generic note F is Material not in NUREG-2191 for this component. Plant-specific note 15 states, The Fire Protection program (B.2.1.15) will manage the hardening, loss of strength for fire barrier penetration seals made of elastomers or grout.

Please discuss citing AMR item 3.3-1, 057, with notes F, 15, for grout fire barrier penetration seals instead of AMR item 3.3-1, 060 as indicated in SLRA Table 3.3.1.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The discussion for AMR item 3.3-1, 057 of SLRA Table 3.3.1 does not state that it is used for grout fire barrier penetration seals.

However, the staff notes that the discussion of AMR item 3.3-1, 060 in SLRA Table 3.3.1 states it is used to manage cracking and loss of material of grout fire barriers in SLRA Table 3.5.2-36. Which seems more appropriate than citing AMR item 3.3-1, 057, since grout is typically a cementitious material.

7 Appendices A and B A-22, A-98, B-76 SLRA Section A.2.1.15 identifies seven enhancements to the Fire Protection program. However, SLRA Table A.5 and SLRA Section B.2.1.15 identify six enhancements to the Fire Protection program.

The staff notes that enhancement 5 in SLRA Section A.2.1.15 is a sub-bullet to enhancement 4 in SLRA Table A.5 and SLRA Section B.2.1.15.

Please discuss revising the SLRA to consistently number the enhancements to the Fire Protection program in SLRA Table A.5 and SLRA Sections A.2.1.15 and B.2.1.15.

8 Appendices A and B Enhancement 4 in SLRA Table A.5 and SLRA Section B.2.1.15 states, Revise implementing procedures to require that results of inspections of the aging effects of cracking and loss of material on fire barrier penetration seals, fire barriers, fire damper assemblies, and fire doors be trended to provide for timely detection of aging effects so that appropriate corrective actions be taken and:

The staff recognizes that the language in the enhancement is consistent with the GALL-SLR, however, the GALL-SLR and the BFN SLRA identifies aging effects in addition to cracking and loss of material for fire barrier penetration seals, fire barriers, fire damper assemblies, and fire doors. For example, hardening, loss of strength, shrinkage due to elastomer degradation, delamination, change in material properties, and separation. The staff notes that Rev. 1 of the GALL-SLR proposed to clarify the Monitoring and Trending program element in AMP XI.M26 to indicate inspection results of all aging effects be trended.

Please discuss whether all aging effects on fire barrier penetration seals, fire barriers, fire damper assemblies, and fire doors will be trended to ensure timely detection and correction actions. In addition, please confirm that the results of the periodic CO2 fire suppression system testing are being trended consistent with GALL-SLR. If not, please discuss whether it should be included in the enhancement related to

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The Monitoring and Trending program element of AMP XI.M26 in GALL-SLR states, The performance of the halon/CO2 fire suppression system is monitored during the periodic test to detect any degradation in the system. These periodic tests provide data necessary for trending. Section 4.5.2 of the Fire Protection program basis document states the above information from GALL-SLR related to monitoring and trending the periodic test of the CO2 fire suppression system. The staff did not see reference to trending in the CO2 procedures on the portal.

trending inspection results.

9 3.0 3-919 SLRA table 3.5.2-36 includes controlled leakage doors with multiple intended functions, including a fire barrier intended function, however, the Fire Protection program is not credited to manage the effects of aging associated with the Fire Barrier intended function.

Please discuss whether the controlled leakage doors in SLRA Table 3.5.2-36 have a fire barrier intended function and should also be managed by the Fire Protection program. If the Fire Protection program will not be credited, discuss how the programs credited (One-Time Inspection, Structures Monitoring, Inspection of Water-Control Structures Associated with Nuclear Power Plants) will manage the effects of aging for the fire barrier intended function.

10 2.0, 3.0 3-390, 3-659, 3-923 SLR-ISG-2021-02-Mechanical, Updated Aging Management Criteria for Mechanical Portions of Subsequent License Renewal Guidance (ML20181A434) added GALL-SLR Items 3.3-1, 267, 3.3-1, 268, and 3.3-1, 269. The aging effects for subliming compounds, cementitious coatings, and silicates used as fireproofing/fire barriers exposed to air are loss of material, cracking/delamination, change in material Please discuss the following:

Why SLRA Table 3.5.2-36 does not include all the aging

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions properties, and separation. These aging effects are consistent with Section 6, Fire Barriers, of EPRI 3002013084, Long-Term Operations: Subsequent License Renewal Aging Affects for Structures and Structural Components (Structural Tools), November 2018.

For subliming fire barriers, SLRA Table 3.3.1 is consistent with the aging effects/mechanisms for subliming compounds in SLR-ISG-2021-02-Mechanical. However, SLRA Table 3.5.2-36 includes only cracking due to vibration and loss of material due to abrasion and flaking for subliming compounds.

The staff notes that SLRA Table 3.5.2-36 uses Subliming compounds (Thermolag, and other similar materials), however, SLRA Section 3.5.2.1.36 uses Subliming compounds (Thermolag). This discrepancy makes it unclear whether there are subliming compounds, other than Thermolag.

For cementitious coating fireproofing, fire barriers, SLRA Table 3.3.1 is consistent with the aging effects/mechanisms for cementitious coatings in SLR-ISG-2021-02-Mechanical. However, SLRA Table 3.5.2-36 includes only loss of material due to abrasion, flaking; cracking/delamination due to settlement; change in material properties due to gamma irradiation exposure; separation for gypsum fireproofing, fire barriers.

effects/mechanisms identified in SLR-ISG-2021 Mechanical for AMR item 3.3-1, 267.

Are there subliming compounds, other than Thermolag, used at BFN?

Why SLRA Table 3.5.2-36 does not include all the aging effects mechanisms identified in SLR-ISG-2021 Mechanical for AMR item 3.3-1, 268.

11 3.0 3-370, 3-923 The discussion of AMR item 3.3-1, 179 in SLRA Table 3.3.1 states, The Fire Protection program (B.2.1.15) and Masonry Walls program (B.2.1.32) will be used to manage cracking due to restraint shrinkage, creep, aggressive environment; loss of material (spalling, scaling) and cracking due to freeze thaw of masonry walls: structural fire barriers exposed to air in Table 3.5.2-36: Structural Commodities (Hazard Barriers and Elastomers).

However, SLRA Table 3.5.2-36 does not include loss of material (spalling, scaling) and cracking due to freeze thaw.

Please discuss whether loss of material and cracking due to freeze thaw are applicable aging effects for the masonry walls structural fire barrier walls.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 12 Appendix B B-77 SLRA Section B.2.1.15 states that the Fire Protection program includes a diesel-driven fire pump inspection program. Specifically, the pump is periodically tested to ensure that the fuel supply line can perform the intended function.

Please identify the Fire Protection program procedure for periodically testing the diesel-driven fire pump fuel supply line. In addition, please discuss whether the SLRA includes an AMR item for the fuel supply line.

13 N/A N/A The Note in Section 5.1.1 of 0-SI-4.11.G.1.c(2) states that Kaowool blankets shall be strapped down using metal straps.

Section 6.1 of EPRI 3002013084, Long-Term Operations: Subsequent License Renewal Aging Effects for Structures and Structural Components (Structural Tools), November 2018, states, in part, wire and other appurtenances used to secure fire wrap to the item being protected - is considered to be part of the fire wrap itself.

Please identify the materials used for securing fire wraps and where they are addressed, including AMR items for managing applicable aging effects.

14 N/A N/A The frequency for 0-SI-4.11.G.1.a(1) and 0-SI-4.11.G.1.A(b) is at least once per 36 months. Based on the NFPA 805 Fire Protection Requirements Manual, the frequencies for 9.4.11.G.1a and 9.4.11.G.1b are 18 months and 12 months.

Please discuss the basis for the 36-month frequency, including identification of supporting reference documents.

15 N/A N/A Section 6.2.1 of 0-SI-4.11.G.1.A(b) states, A loss of protection of >1 sq. ft. cumulative per beam is unacceptable. This applies to Albi Clad on structural steel beams.

Please identify supporting reference documents for the square footage limit. In addition, is there a document that includes all acceptance criteria (length, width, depth limits) for fireproofing and fire barrier materials?

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA Section: B.2.1.16 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

3.3 3-279, 3-416 SLRA Section 3.3.2.1.7 identifies gas as an applicable environment for the High Pressure Fire Protection (Diesel Driven Pump) System.

However, SLRA Table 3.3.2-7 does not include any components exposed to gas.

Please discuss whether components in the High Pressure Fire Protection (Diesel Driven Pump)

System are exposed to gas.

2 3.3, Appendix B 3-279, 3-319, 3-361, 3-416, B-78 SLRA Section 3.3.2.2.7 states, in part, The review of plant specific OE has identified no recurring internal corrosion in metallic piping, piping components, tanks exposed to raw water, raw water (potable), treated water, or waste water in the screened in portions ofHigh Pressure Fire Protection This section states, However, since internal Operating Experience has shown that there have been significant numbers of piping through wall leaks and/or minimum wall thickness readings within the BFN raw water systems, recurring internal corrosion is assumed. In addition, this section states, In addition to being covered by the Open Cycle Cooling Water System Program, as described below, BFN will implement the Fire Water System program (B.2.1.16) to manage the potential for recurring internal corrosion in the Fire Water System.

SLRA Section B.2.1.16 states, A review of BFN operating experience has revealed instances of recurring internal corrosion in the fire water system piping that is within the scope of the Fire Water System program. Inspections are performed on the fire water piping by non-intrusive volumetric examinations, to ensure that aging effects are managed, and that wall thickness is within acceptable limits.

The Discussion of AMR item 3.3-1, 127 in SLRA Table 3.3.1 states it is not applicable and that There are no metallic piping, piping components, tanks exposed to raw water, raw water (potable), treated Please discuss the following:

1. Is loss of material due to recurring internal corrosion an applicable aging effect for the fire water system?

2. If so, please discuss SLRA changes to clarify that loss of material due to recurring internal corrosion is an applicable aging effect for the fire water system.

3. Please discuss which AMR items are used for loss of material due to recurring internal corrosion if AMR item 3.3-1, 127 is not. The staff notes

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions water, waste water that have not been addressed by other more specific item numbers in screened in portions of Auxiliary Systems.

The staff notes that SLRA Section 3.3.2.2.7 and SLRA Table 3.3.2-7 do not include loss of material due to recurring internal corrosion as an applicable aging effect.

that the GALL-SLR includes specific AMR items for loss of material due to recurring internal corrosion.

The staff notes that there is a Breakout Question under TRP 82 regarding the level of information in SLRA Section 3.3.2.2.7.

3 2.3, 3.3 2-90, 3-363, 3-417 SLRA Table 2.3.3-7 includes component type Fire Water Storage Tanks.

The discussion of AMR item 3.3-1, 136 in SLRA Table 3.3.1 states, The Fire Water System program (B.2.1.16) will be used to manage loss of material of the steel fire water storage tanks exposed to air, condensation, soil, concrete, raw water, raw water (potable), treated water in the High Pressure Fire Protection (Diesel Driven Pump)

System.

SLRA Table 3.3.2-7 cites AMR item 3.3-1, 136 for managing loss of material for steel fire water storage tanks exposed to treated water.

The staff notes that Section 3.5 in SLR-BFN-0061, Revision 2, Aging Management Program Basis Document - Fire Water System, states the following:

BFN does not have carbon steel fire water tanks in the screened in portions for the fire water system.

BFN source of water to the fire main is Wheeler Reservoir and considered unlimited. There are not Fire Water storage tanks managed by this AMP.

Please discuss the following:

1. Why does Revision 2 of SLR-BFN-0061 state there is no Fire Water Storage Tank?

2. Are the Fire Water Storage Tank bottom surfaces exposed to soil or concrete?

3. If the Fire Water Storage Tank bottom surfaces are exposed to soil or concrete, why does SLRA Table 3.3.2-7 not include AMR items for managing the effects of aging for the Fire Water Storage Tank by AMP XI.M29?

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The Detection of Aging Effects program element for AMP XI.M27 in NUREG-2191, Volume 2, states, Fire water storage tank bottom surfaces exposed to soil or concrete are inspected in accordance with GALL-SLR Report AMP XI.M29, Outdoor and Large Atmospheric Metallic Storage Tanks, Table XI.M29-1. For indoor fire water storage tanks exposed to concrete, this only applies if the tank bottom to concrete interface surface is periodically exposed to moisture.

The Scope of the Program program element for AMP XI.M42 in NUREG-2191, Volume 2, states, The aging effects associated with fire water tank internal coatings/linings are managed by Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report aging management program (AMP) XI.M27, Fire Water System, instead of this AMP. However, where the fire water storage tank internals are coated, the Fire Water System Program and Final Safety Analysis Report (FSAR) Summary Description of the Program should be enhanced to include the recommendations associated with training and qualification of personnel and the corrective actions program element. The Fire Water System Program should also be enhanced to include the recommendations from the acceptance criteria program element.

4. Is the Fire Water Storage Tank located outdoors or indoors?

5. If the Fire Water Storage Tank is located indoors and the bottom surfaces are exposed to concrete, is the interface between the tank bottom and concrete periodically exposed to moisture?

6. Are the Fire Water Storage Tank internals coated?

7. If the Fire Water Storage Tank internals are coated, please discuss changes to SLRA Sections A.2.1.16 and B.2.1.16 to include the recommendations from AMP XI.M42.

4 3.3 3-417 SLRA Table 3.3.2-7 includes component type Heat exchanger components, and cites AMR item 3.3-1, 050 for managing reduction of heat transfer due to fouling of the copper alloy heat exchanger tubes by the Closed Treated Water System program.

The staff notes that one subsequent license renewal applicant stated that inspection of the heat exchanger tube bundle for degradation is not practical due to the small tube diameter (ML21091A187).

Please identify the specific heat exchanger components included in component type Heat exchanger components (e.g., shell, channel, tubesheet). In addition, please discuss whether inspections of the heat

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions exchanger tubes are practical.

5 Appendix A and Appendix B A-23 SLRA Sections A.2.1.16 and B.2.1.16 state, Abnormal results are entered into the Corrective Action Program for review and resolution.

In addition, SLRA Section B.2.1.16 states, Abnormal or unacceptable results are entered into the Corrective Action Program for review and resolution.

Similar statements are made in SLR-BFN-0061, Revision 2.

However, the NRC staff did not find what is meant by abnormal degradation in the procedures associated with the Fire Water System program.

Please discuss what is meant by abnormal results.

6 3.3, Appendix A and Appendix B 3-279, 3-421, A-24, A-100, B-80 SLRA Sections A.2.1.16 and B.2.1.16 and SLRA Table A.5 include an enhancement to the Fire Water System program to Revise implementing procedures to ensure that visual examinations of cementitious materials will be conducted to detect indications of loss of material and cracking that could affect the system's ability to maintain pressure.

SLRA Table 3.3.2-7 cites AMR item 3.3-1, 138 for managing loss of coating or lining integrity and loss of material or cracking for cementitious coatings/linings by the Internal Coatings/ Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program. This appears to be consistent with the statements in SLRA Sections A.2.1.16 and B.2.1.16 that state, The fire main buried cement lined pipe aging effects such as cracking are managed by the Internal Coating/Linings for In Scope Piping, Piping Components, Heat Exchangers, and Tanks program (B.2.1.28).

There are no other cementitious materials identified in SLRA Table 3.3.2-7. The staff notes that the material for the component where AMR item 3.3-1, 138 is cited in SLRA Table 3.3.2-7 is steel and not steel Please discuss the following:

1. The basis for the Fire Water System program enhancement related to cementitious materials when it appears the only cementitious material is that used to line the fire main buried piping which is managed by the Internal Coatings/

Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions with internal coatings/linings. In addition, SLRA Section 3.3.2.1.7 does not include cementitious or steel with internal coatings/linings.

2. Why SLRA Table 3.3.2-7 and SLRA Section 3.3.2.1.7 does not include steel with internal coatings/linings as a material for the High Pressure Fire Protection (Diesel Driven Pump System.

7 Appendix A and Appendix B A-25, A-100, B-80 SLRA Sections A.2.1.16 and B.2.1.16 and SLRA Table A.5 include an enhancement to the Fire Water System program to Revise implementing procedures to meet guidelines of Revision 0 of NUREG-2191, Table XI.M27-1, Fire Water System Inspection and Testing Recommendations (NFPA 25, 2011 Edition, Guidelines).

The staff notes that this enhancement lacks specificity which may make it hard to verify implementation of the required enhancements in order to be consistent with Table XI.M27-1 in Revision 0 of NUREG-2191.

The enhancement, as written, is difficult to determine what changes are required. For example, Table XI.M27-1 in NUREG-2191, Volume 2, recommends inspection and testing of hydrants in accordance with Section 7.3.2 of the 2011 Edition of NFPA 25. Specifically, Section 7.3.2.1 of NFPA 25 states, Each hydrant shall be opened fully and water flowed until all material has cleared. In addition, Sections 7.3.2.4 and 7.3.2.5 of NFPA 25 are related to the hydrant barrel taking no longer than 60 minutes to drain.

The staff notes that Surveillance Instructions 0-SI-4.11.F.1.a, Revision 0014, and FP-0-026-INS003, Revision 0028, include acceptance criteria related to flushing the hydrant thoroughly, hydrant flow is unobstructed, and drains are functional and drain the hydrant barrel.

However, they also make statements related to flushing until sediment is minimized, which is different from the NFPA 25 requirement that all Please provide sufficient detail about the enhancements required to be consistent with Table XI.M27-1 in Revision 0 of NUREG-2191.

In addition, please discuss whether the hydrant Surveillance Instructions will be revised to be consistent with the recommendations in NFPA 25.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions material has cleared. In addition, these procedures do not appear to include a timeframe for when the hydrant barrel should drain.

8 Appendix A and Appendix B A-25, A-102, B-81 SLRA Sections A.2.1.16 and B.2.1.16 and SLRA Table A.5 include an enhancement to the Fire Water System program to Revise implementing procedures to ensure that inspections with quantitative results, degradation identified will be projected until the next scheduled inspection. Similar statements are made in SLR-BFN-0061, Revision

2.

The terminology inspections with quantitative results is not used in the Monitoring and Trending program element in AMP XI.M27 in Revision 0 of NUREG-2191, Volume 2, which states, Where practical, degradation identified is projected until the next scheduled inspection.

The staff notes that Section A.1.2.3.5 in Revision 0 of NUREG-2192 states, Although aging indicators may be quantitative or qualitative, aging indicators should be quantified, to the extent possible, to allow trending. The parameter or indicator trended should be described.

While it states aging indicators should be quantified to the extent possible, it does not indicate that qualitative aging indicators should not be considered for trending.

Please discuss whether the use of inspections with quantitative results is too limiting.

9 Appendix A and Appendix B A-26, A-102, B-81 SLRA Sections A.2.1.16 and B.2.1.16 and SLRA Table A.5 include an enhancement to the Fire Water System program to Revise implementing procedures to specify acceptance criteria for minimum design wall thickness, and for loose fouling products in systems that could cause flow blockage in the sprinklers or deluge nozzles.

The Acceptance Criteria program element in AMP XI.M27 in Revision 0 of NUREG-2191, Volume 2, states in part, The acceptance criteria are:(b) minimum design wall thickness is maintained, and (c) lo loose fouling products exists in systems that could cause flow blockage in the sprinklers or deluge nozzles.

Given that the enhancement does not state what the acceptance criteria is, please discuss whether the acceptance criteria for minimum design wall thickness and loose fouling products will be consistent with the Acceptance Criteria program element in AMP

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions XI.M27 in Revision 0 of NUREG-2191, Volume 2.

10 Appendix B B-83 The Operating Experience discussion in SLRA Section B.2.1.16 states that a condition report was generated to have a specialty contractor review fire protection system piping to identify and correct pipe slope issues. SLR-BFN-0061, Revision 2, identified the condition report as CR-1847803, dated April 2023.

Please discuss the status of this review.

11 3.3 3-363, 3-417 The Discussion of AMR item 3.3-1, 136 in SLRA Table 3.3.1 states, Consistent with NUREG-2191. The Fire Water System program (B.2.1.16) will be used to manage loss of material of the steel fire water storage tanks exposed to air, condensation, soil, concrete, raw water, raw water (potable), treated water in the High Pressure Fire Protection (Diesel Driven Pump) System. However, the steel fire water storage tank in SLRA Table 3.3.2-7, where AMR item 3.3-1, 136 is cited, is only exposed to a treated water environment.

Please discuss why the applicable environments for the steel fire water storage tank differ in the Discussion of AMR item 3.3-1, 136 in SLRA Table 3.3.1 and SLRA Table 3.3.2-7.

12 3.3 3-345, 3-419 The Discussion of AMR item 3.3-1, 065 in SLRA Table 3.3.1 states, Consistent with NUREG-2191. The Fire Water System program (B.2.1.16) will be used to manage loss of material due to pitting, crevice corrosion; flow blockage due to fouling while exposed to raw water for aluminum piping, piping components exposed to raw water, treated water, raw water (potable) in screened in portions of the High Pressure Fire Protection (Diesel Driven Pump) System. However, the aluminum piping and piping components in SLRA Table 3.3.2-7, where AMR item 3.3-1, 065 is cited, is only exposed to a treated water environment.

Please discuss why the applicable environments for the aluminum piping and piping components differ in the Discussion of AMR item 3.3-1, 065 in SLRA Table 3.3.1 and SLRA Table 3.3.2-7. In addition, what is meant by while exposed to raw water?

13 3.3 3-279, 3-286, 3-350, 3-362, 3-421, 3-447 The Discussion of AMR item 3.3-1, 089 in SLRA Table 3.3.1 states, Not Applicable. There is no steel piping, piping components exposed to condensation (internal) in screened in portions of Auxiliary Systems.

However, AMR item 3.3-1, 131 is cited to manage flow blockage due to fouling of the steel piping and piping components exposed to air, condensation.

Please discuss whether loss of material should be managed for piping, piping components, and spray nozzles exposed to air, condensation in the High Pressure Fire Protection (Diesel Driven

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The Discussion of AMR item 3.3-1, 131 in SLRA Table 3.3.1 states, Consistent with NUREG-2191. The Fire Water System program (B.2.1.16) will be used to manage flow blockage of the steel, stainless steel, copper alloy, aluminum piping, piping components, and spray nozzles exposed to air, condensation in the High Pressure Fire Protection (Diesel Driven Pump) and CO2 Storage Fire Protection/Purge Systems. Therefore, it is unclear why flow blockage due to fouling would be managed for steel piping, piping components, and spray nozzles exposed to air, condensation, but not loss of material.

The staff notes that AMR item 3.3-1, 131 is cited for piping and piping components in SLRA Tables 3.3.2-7 and 3.3.2-13, but not cited for spray nozzles. In addition, spray nozzles are not included in SLRA Tables 2.3.3-1 and 2.3.3-13.

Pump) and CO2 Storage Fire Protection/Purge Systems. In addition, please discuss why spray nozzles are not in SLRA Tables 2.3.3-1, 2.3.3-13, 3.3.2-7, and 3.3.2-13.

14 3.3 3-344, 3-418 The Discussion for AMR item 3.3-1, 064 in SLRA Table 3.3.1 states, in part, The Fire Water System program (B.2.1.16) will be used to manage loss of materialflow blockage due to foulingin steel, copper alloy piping, piping components exposed to raw water, treated water, and raw water (potable) However, no steel or copper alloy piping, piping components in SLRA Table 3.3.2-7 is exposed to raw water (potable).

Please discuss whether any steel or copper alloy piping, piping components in the High Pressure Fire Protection (Diesel Driven Pump) system are exposed to raw water (potable).

15 3.3 3-362 The Discussion for AMR item 3.3-1, 130 in SLRA Table 3.3.1 states, in part, The Fire Water System program will be used to manage Loss of materialflow blockage due to fouling on metallic sprinklers exposed to air, condensation, raw water, raw water (potable), treated water in the High Pressure Fire Protection System. However, air and condensation are the only environments cited for sprinklers in SLRA Table 3.3.2-7.

Please discuss whether the sprinklers in the High Pressure Fire Protection (Diesel Driven Pump) system are exposed to environments other than air and condensation.

16 2.3, 3.3, Appendix B 2-88, B-78, 3-416 SLRA Section 2.3.3.7 includes the following components: aqueous foam systems, hose racks, and hose connections.

Please discuss how components of the aqueous foam system, hose racks, hose connections, valve

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA Section B.2.1.16 includes the following components: valve bodies, fire pump casings, hose stations, standpipes, strainers, and flow devices.

SLR-BFN-0061, Revision 2 includes the following components: valve bodies, fire pump casings, hose stations, standpipes, strainers, and flow devices.

However, these components do not appear to be included in SLRA Tables 2.3.3-7 and 3.3.2-7.

The staff notes the following:

SLR-BFN-0061, Revision 2, states that hose stations and standpipes are considered piping in the SLRA. However, it did not appear to make similar statements regarding the other components noted above.

The SLRA includes component type piping, piping components (strainers) for other systems. In addition, strainers typically also have a filtration function, however, no components in SLRA Tables 2.3.3-7 and 3.3.2-7 have a filtration function.

The SLRA includes component type flow devices for other systems.

bodies, fire pump casings, strainers, and flow devices are addressed in the SLRA.

17 Appendix A,

Appendix B A-24, B-79 SLRA Sections A.2.1.16 and B.2.1.16 state, The water-based fire protection system is normally maintained at required operating pressure and is monitored such that loss of system pressure is immediately detected and corrective actions initiated.

SLR-BFN-0061, Revision 2, does not appear to provide additional information regarding how the operating pressure is monitored.

Please discuss how the operating pressure is monitored.

18 Appendix A,

Appendix B A-24, B-79 Table XI.M27-1 in NUREG-2191, Volume 2, recommends testing of sprinklers in accordance with Section 5.3.1 of the 2011 Edition of NFPA

25.

SLRA Sections A.2.1.16 and B.2.1.16 state, Sprinkler heads for each unit will be replaced or inspected. Either sprinklers are replaced before Please discuss whether there are dry sprinklers at BFN that would need to be replaced/tested in accordance with Section

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions reaching 50 years in-service or a representative sample of sprinklers from one or more sample areas is tested by using the guidance of NFPA 25, 2011 Edition. Additionally, if sprinklers are not replaced, they shall be inspected prior to the end of the specified service life and at specified intervals thereafter in accordance with NFPA 25. [Note, SLRA Sectio B.2.1.16 also includes the title of NFPA 25, 2011 Edition]

The Operating Experience section of SLRA Section B.2.1.16 states, One deficiency was related to testing / replacement of fast acting sprinklers, which NFPA 25 requires replacement or testing after 20 years service. This issue was initially addressed by stating that the original licensing commitment included all sprinkler types and that they would be replaced / tested at the 50 year interval. After further review, it was determined that BFN should follow the EPRI guidance for the testing / replacement and possible de-scoping of the sprinklers, and develop a project to complete this activity. This EPRI report may be used to support the service life assessment of the 50 year sprinklers as well as the fast acting sprinkler heads and development of a technical justification regarding the testing or replacement requirement. This project is scheduled to complete in February 2024.

SLR-BFN-0061, Revision 2, states similar information but includes references to CRs 1776704 and 1835584, and EPRI Reports 3002016097 and 3002018285.

The staff notes that if sprinkler replacement/testing for standard, fast-response, and dry sprinklers will not be in accordance with Section 5.3.1 of the 2011 Edition of NFPA 25, then an exception would be needed for the SLRA with an adequate technical justification, including any supporting technical references.

5.3.1.1.1.6 of NFPA 25, 2011 Edition.

In addition, provide an overview and status of the sprinkler service life assessment which was to be completed in February 2024. Include whether you plan to include an exception to the recommendation in Table XI.M27-1 in NUREG-2191, Volume 2, which recommends testing of sprinklers in accordance with Section 5.3.1 of the 2011 Edition of NFPA 25.

19 N/A N/A Surveillance Instruction 0-SI-4.11.E.1.A, Revision 0021, is titled Outside Fire Hose Replacement. However, it appears to be focused on inspecting fire hoses rather than replacing them.

Please discuss why Surveillance Instruction 0-SI-4.11.E.1.A, Revision 0021, is considered a fire hose replacement procedure.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The staff notes that Surveillance Instructions 0-SI-4.11.E.1.a(2) and 0-SI-4.11.E.1.B(2) appear to be related to replacement of outside and safety related fire hoses, respectively.

20 N/A N/A The staff notes that the frequency for FP-0-026-INS002, Revision 0013, is As required.

Please discuss the frequency for the fire hose hydrostatic test.

SLRA Section: B.2.1.17 Outdoor and Large Atmospheric Metallic Storage Tank Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

SLRA -

Appendix B Section B.2.1.17 XI. M29 Outdoor and Large Atmospheric Metallic Storage Tank SLRA -

Page B-86 SLRA Exception #1:

The guidance of the GALL-SLR require to apply sealant or caulking at the interface between the tank external surface and concrete or earthen surface for the condensate storage tanks.

The use of caulking between the tank external surface and the concrete and earthen surface is not used as specified in Elements 2, 4, 6 and 7 of the NUREG-2191, XI.M29. As an alternate to the GALL-SLR Elements 2. 4, 6 and 7 requirements, the BFN application takes an exception to applying sealant or caulking at the interface between the tank external surface and concrete earthen surface and instead uses the fiber board in the top of the concrete ring of the foundation of the tanks to prevent interference between the tank external surface and the concrete and earthen surface at BFN for condensate storage tank.

What is the basis to determine that the fiber board in the top of the concrete ring of the foundation of the tanks (between the tank external surface and the concrete and earthen surface) can substitute for the use of sealant or caulking at the interface between the tank external surface and concrete or earthen surface for the condensate storage tanks?

What type of fiber board was used? What are the manufacturers recommendations for the fiber board routine maintenance and/or inspection?

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions What environment will the board be exposed to, and is there any aging effects that need to be managed because of this environment?

Is the fiber board regularly inspected for age related degradation?

Is there any OE on the fiber board or OE of moisture intrusion between the interfaces?

SLRA Section: 3.1.2-3 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

SLRA Section 3.1.2-3 Page:

3-148 3-150 The applicant has identified, loss of material due to pitting, crevice corrosion an aging effect that is not in the GALL-SLR Report, which sites Note H.

It is not clear to the staff why 3.1-1, 136 is a note H. It appears that the aging affect is in the GALL-Report.

2 SLRA Section 3.2.2.2.3 Page:

3-186 As stated in SLRA Section 3.2.2.2.3, Plant-specific OE has revealed corrosion on the carbon steel nipple between the header and a BFN Unit 2 torus spray copper alloy nozzle. The corrosion was attributed to leakage in the upstream valves that created a prolonged wetted condition.

Will this component be included in the sample population for the One-Time inspection to verify that loss of material is not an issue?

3 SLRA Section 3.1.2.2.16 Page 3-42 As stated in SLRA Section 3.2.2.16, this item evaluates the loss of material due to pitting and crevice corrosion in stainless steel and nickel alloy piping, piping components exposed to air and condensation. BFN-specific OE review revealed only one previously written Condition Report (CR) in the Corrective Action Program with the potential to Will this component be included in the sample population for the One-Time inspection to verify

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions represent an aging effect requiring management. This CR documented several areas on the Unit 1 reactor vessel head and flange in 2014, where pitting was found that exceeded General Electric (GE) criteria, and repairs were performed which included welding and honing.

that loss of material is not an issue?

SLRA Section B.2.1.21: Selective Leaching Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

B.2.1.21 B-104 SLRA Section B.2.1.21, Selective Leaching, states [d]uring an inspection in August 2011, the Emergency Diesel Generator (EDG) cooling water temperature control valve for the D Diesel Generator had indications of selective leaching in the valve internals.

SLR-BFN-0066, Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Subsequent License Renewal Aging Management Basis Document - Selective Leaching, notes that BFN has operating experience (OE) of selective leaching associated with the Diesel Generator Temperature Control Valves (subjected to closed-cycle cooling water (CCCW)), therefore the Diesel Generator closed cooling water system will not be included in the One-Time Inspection grouping.

The statement that the diesel generator closed cooling water system will not be included in the one-time inspection grouping does not appear to be reflected in the SLRA. The staff requests a discussion with respect to this topic.

Components susceptible to selective leaching and a CCCW environment are also within the scope of the air conditioning and reactor building closed cooling water systems. The staff requests a discussion with respect to why one-time inspection are appropriate for these components, given the OE in the standby diesel generators system.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 2

B.2.1.21 B-102 SLRA Section B.2.1.21 states [a] representative sample consists of three percent of each material and environment population per unit or a maximum of 10 components per population per unit.

NUREG-2222, Disposition of Public Comments on the Draft Subsequent License Renewal Guidance Documents NUREG-2191 and NUREG-2192, provides the basis for reducing the extent of inspections for selective leaching during the subsequent period of extended operation (i.e., 3 percent with a maximum of 10 components per GALL-SLR guidance) when compared to the extent of inspections for selective leaching during the initial period of extended operation (i.e., 20 percent with a maximum of 25 components per GALL Report, Revision 2 guidance). Part of the basis for reducing the extent of inspections is that industry OE had not identified instances of loss of material due to selective leaching which had resulted in a loss of intended function for the component.

SLR-BFN-0066 notes that plant-specific OE has identified some thru-wall failures have occurred in fire protection piping in environments that are susceptible to selective leaching where the corrosion mechanism that resulted in these thru-wall failures was not determined.

CR 1102016 notes that a contributing cause of a break in buried fire protection piping was outer diameter graphitic corrosion (i.e., selective leaching).

Based on OE at BFN, the staff requests a discussion with respect to using the reduced sample size (i.e., 3 percent with a maximum of 10 components) for gray cast iron piping exposed to soil.

3 B.2.1.21 B-102 SLRA Section B.2.1.21 states [m]aterials susceptible to selective leaching which are in the scope of this program are gray cast iron, ductile iron, and copper alloys containing greater than 15 percent zinc.

Copper alloys containing greater than 8 percent aluminum (aluminum bronze) are also susceptible to selective leaching.

The staff requests a discussion with respect to if aluminum bronze components are within the scope of the Selective Leaching program. Several aging management review (AMR) line items cite

[c]opper alloy (>15% Zn

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions or >8% Al) as the material.

4 N/A N/A Draft NUREG-2221, Revision 0, Supplement 1, Technical Bases for Changes in the Subsequent License Renewal Guidance Documents, Draft NUREG-2191, Revision 1, and Draft NUREG-2192, Revision 1, Draft Report for Comment, documents recent SLRA plant-specific operating experience as a basis for the inclusion of malleable iron as a material susceptible to selective leaching.

The staff requests a discussion with respect to if there are mechanical malleable iron components in-scope for subsequent license renewal (the SLRA shows malleable iron for electrical insulators only).

Section B.2.1.22, ASME Code Class 1 Small-Bore Piping Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

B.2.1.22 B-106 to B-108 The SLRA section identified applicable examples of cracking in ASME Code Class 1 small-bore piping, as follows.

1) A condition report, dated May 13, 2009, documented a planar flaw that exceeded the acceptance criteria of ASME Section XI, Table IWB-3514-2 in the Unit 2 Reactor Vessel Instrumentation Nozzle N12A safe end to pipe weld.

2) On May 5, 2012, a cracked weld at a 1-inch Unit 3 Residual Heat Removal test valve was identified during the ASME Section XI System Pressure Test.

3) On August 3, 2014, an entry was made due to rising unidentified leakage rate in Unit 2 drywell with the reactor at approximately 13% power. During this entry, a leak was identified downstream from a 20-inch Unit 2 RHR flow control valve.

4) On March 26, 2017, a small through wall leak in the 1-inch piping between a primary penetration location and a flow element was identified during the reactor vessel leakage test.

5) On October 15, 2020, a circumferential indication was identified during Ultrasonic Test (UT) examination of the Unit 1 safe end to pipe weld on an instrument line connected to pressure vessel Please provide on the portal the CRs which document the subject OE along with the root cause analysis reports or metallurgical analysis reports, if available.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions nozzle and connecting upstream of a primary containment penetration.

6) A Condition Report, dated October 20, 2020, documented an aging related crack that was found on the safe end to pipe weld is part of the Standby Liquid Control system in Unit 1.

B.2.1.22 B-105 The SLRA section states that, The operating experience relative to the ASME Code Class 1 Small-Bore Piping program did not identify an adverse trend in performance.

Discuss the technical bases to substantiate the statement.

3 B.2.1.22 B-106 to B-108 The SLRA section states that, The new ASME Code Class 1 Small-Bore Piping program will provide reasonable assurance that the identified aging effects will be adequately managed In light of the recent OE cases, discuss detailed processes and plans to manage aging during the subsequent period of extended operation.

AMP: XI.M38 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

GALL-SLR B.2.1.24 3.3.2.2.7 BFN SLR Aging Related OE.pdf XI.M38-1 B-111 3-189 The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components AMP, XI.M38, in the GALL-SLR inspects the internal surfaces of piping, piping components, ducting, and heat exchanger components exposed to aggressive environments. This AMP also manages aging effects associated with elastomers and flexible polymeric components in the open-cycle, closed-cycle cooling water, ultimate heat sink, and fire water systems. As noted in the GALL-SLR, this AMP is not intended for use on components in which recurring internal corrosion (RIC) is evident based on plant specific operating experience (OE).

Based on the Program Description in the SLRA (page B-111), this AMP at BFN is focused on elastomeric and polymeric materials but is also used to manage stress corrosion cracking in stainless steel components.

Based on review of the site boundary drawings and the specific wording used in the CRs/SRs (referenced in the adjacent column), the staff was not able to determine whether the portions of the RCW system referenced in the CRs/SRs were in scope for SLR. However, since Section 3.3.2.2.7 of the SLRA has assumed RIC, it is unclear why internal

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions However, page B-112 describes internal visual inspections used to assess loss of material due to corrosion and corrosion product deposition in steel components.

The last paragraph on page B-112 states that a review of plant-specific OE for RIC did not identify RIC.

However, Section 3.3.2.2.7 of the SLRA states that since internal OE has shown that there have been significant numbers of piping through wall leaks and/or minimum wall thickness readings within the BFN raw water systems, RIC is assumed. Therefore, BFN will implement the Open-Cycle Cooling Water System program (B.2.1.11) to manage RIC in the piping, piping components, tanks exposed to raw water, raw water (potable), treated water, or waste water systems.

A keyword search for RCW in BFN SLR Aging Related OE.pdf on the CERTREC portal resulted in 146 matches, which equated to about 35 CRs/SRs that documented through-wall leaks or minimum wall thickness violations in raw cooling water piping or piping components.

surface inspections of piping are referenced in AMP XI.M38. It appears that BFN is planning to use the Open-Cycle Cooling Water AMP, XI.M20, to manage RIC of piping, piping components, and tanks exposed to raw water, raw water (potable),

treated water, or waste water.

The staff requests a discussion to clarify:

the scope of inspections covered by AMP XI.M38 whether the RCW system has met the definition in the SRP-SLR for RIC whether the scope of AMP XI.M20 is just safety-related metallic piping, piping components, and tanks or all metallic piping, piping components, and tanks exposed to raw water, raw water (potable), treated water, or waste water.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA Section: B.2.1.26 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

N/A N/A In Browns Ferrys Response to Generic Letter 2016-01 it states that, TVA anticipates that there are enough coupons for the surveillance program for the life of the SFP. (BFN Unit 3 is currently licensed to operate until 2036). It is unclear to the staff how TVA will ensure that an adequate number of coupons remain for testing during the subsequent period of operation.

Please discuss what provisions are in place to ensure that there are an adequate number of coupons remaining for testing during the subsequent period of operation.

2 B.2.1.26 B-117 In Browns Ferrys response to GL 2016-01 it states that Unit 3 is representative of Units 1 and 2 and in 0-TI-116 High Density Fuel Storage and Surveillance Program, it states that the coupon testing for the coupons in Unit 3 is representative of all 3 units. However, section B.2.1.26 of the SLRA states that the new condition monitoring program will be consistent with NUREG-2191 and it does not mention using the coupons for Unit 3 to cover Units 1 and 2. It is unclear to the staff how the program will be consistent with the GALL-SLR report for Units 1 and 2 if testing of the unit specific absorber material is not being done for Units 1 and 2.

Please discuss how the Monitoring of Neutron Absorbing Materials Other Than Boraflex program will be consistent with the GALL-SLR report for units 1 and 2.

SLRA Section B.2.1.30 ASME Section XI, Subsection IWF Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

B.2.1.30 Pg. B-154 ASME Section XI, Subsection IWF has been used throughout this subsection except in one instance where BFN states Indications are evaluated against the acceptance standards of ASME Code Section XI. Although it can be inferred that the acceptance standards are those of Subsection IWF, there is a lack of clarity whether other acceptance criteria than those of ASME Code Section XI, Subsection IWF may be followed at specific areas of programmatic inspections.

a) Clarify whether the referenced ASME Section XI acceptance standards are those of Subsection IWF.

b) Discuss whether BFN plans to revise ASME Section XI to read

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions ASME Section XI, Subsection IWF for consistency.

2 B.2.1.30

'Pg. B-154/ 155/

156 The last sentence of the fourth paragraph within Program Description Section, of SLRA AMP B.2.1.30 includes only the inspection of elastomeric vibration isolation elements. BFN then provides Enhancement 7 and a corresponding commitment (in commitment number 30) to its B.2.1.30 AMP revising its procedures to ensure that, in addition to elastomeric, polymeric vibration isolation elements will be monitored in accordance with parameters monitored or inspected program element of GALL-SLR AMP XI.S3. It is not clear whether BFN has also polymeric vibration isolation elements. If so, whether it monitors their effects of aging under this AMP, or in tandem with B.2.1.23 AMP, External Surfaces Monitoring of Mechanical Components, where explicitly polymeric components are discussed; or simply does not have polymeric vibration isolation elements at this time.

a) Clarify why polymeric vibration isolation elements are not included in B.2.1.30 AMP program description.

b) Clarify whether the effects of aging of such polymeric components, if in existence at BFN, are to be managed by BFN AMP B.2.1.30 or AMP B.2.1.23, or by both.

c) Revise AMP B.2.1.30 program description if necessary.

3 B.2.1.30; A.2.1.30; Table A.5 Item 30 Pg. B-157; A-45; A-123 Enhancement 14 and associated Commitment 30 to Acceptance Criteria program element of SLRA AMP B.2.1.30 references ASME Code Section XI, IWF-3410(a) for enhancement of implementing procedures for [l]oss of material, cracking, and hardening of elastomeric or polymeric vibration isolation elements that could reduce the vibration isolation function.

The staff notes that the corresponding GALL-SLR XI.S3 program element amplifies the ASME Code Section XI, IWF-3410(a) by adding the aforementioned statement in other unacceptable conditions to the statement of the program element.

Clarify the statement made in Enhancement 14 that the revised procedures for the aforementioned unacceptable conditions for elastomeric and polymeric vibration isolators will conform to topical report GALL-SLR XI.S3 guidance instead of refenced IWF-3410(a).

a) Discuss why the other unacceptable conditions addressed under the acceptance criteria program element of GALL-SLR XI.S3 AMP are not associated with Enhancement 14 and Commitment 30 (#14).

b) Revise Enhancement

(#14) to state, instead with compliance to IWF-3410(a), its consistency with GALL-SLR AMP XI.S3, Acceptance Criteria, program element.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 4

B.2.1.30; TVA SLR-BFN-0039 R2 (PBD)

Section 4.1.2 B-155; PBD Pg.

24 The SLRA Section states, The program will also perform inspections of the seismic restraints in the RHRSW pump pit.

The AMP Program Basis Document (PBD) SLR-BFN-0039 R2 Section 4.1.2 states, BFN will continue to perform periodic inspections of the seismic restraints in the RHRSW pump pit (previously performed by Open-Cycle Cooling Water [OCCW] ILR AMP B.2.1.17). These inspections will be performed no sooner than 24 months prior to the subsequent period of extended operation and within 10 years after entering the SPEO. These inspections will be performed using underwater camera or other methods or techniques available at the time of the inspection. The acceptance criteria for these inspections will be no significant material loss due to corrosion or wear. The frequency of these inspections will be evaluated and adjusted based on system performance and identified deficiencies not to exceed 10 years. All inspection deficiencies will be entered into the corrective action program.

The staff notes that the inspection of the seismic restraints in the RHRSW pump pits is one of the initial license renewal (ILR) commitments and it can be inferred that BFN plans to continue the LRA commitment made for the PEO into the SPEO. However, both SLRA AMP B.2.1.30 and its PBD do not discuss such commitment and the reason of changing the LRA AMP B.2.1.17 Open-Cycle Cooling Water System Program during the PEO to that of SLRA AMP B.2.1.30 ASME Code Section XI, Subsection IWF during the SPEO to manage the aging effect of the seismic restraints in the RHRSW pump pits.

The staff notes that the seismic restraints in the RHRSW pump pits are ASME Code Class 3 equivalent per reference BFN-1-47E858-1-ISI.

It is not clear whether the omission and such change would not require enhancement(s), revision(s) or exception(s) to the SLRA AMP B.2.1.30 and implementing program procedures to meet the of ASME Code Section XI, Subsection IWF requirements and consistency with the GALL-SLR AMP XI.S3.

a) Discuss the continuation of the ILR commitment into the SPEO. Discuss how the performance of inspection(s) of the seismic restraints in the RHRSW pump pits would differ from that of PEO during the SPEO?

b) Confirm that switch of AMPs from OCCW to IWF for the seismic restraints in the RHRSW pump pit for SPEO was due to classification of the restraints as ASME Code Class 3 supports.

Discuss inclusion of these supports in ISI sample in accordance with ASME Code Section XI, Subsection IWF Table IWF-2500-and implementing ISI program 0-TPP-ENG-467 c) Identify which Table 2 AMR line item(s), lining up with a corresponding SLRA Table 3.5.1 item is/are used so that the aging effects are managed consistent with GALL-SLR AMP XI.S3 during the SPEO.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 5

Table 3.5.2-2 Pg. 3-759 For the third AMR line item on this Table and in the referenced page, Molykote@321 or equal is identified as a material for sliding surfaces providing structural support and subject to loss of mechanical function aging effect. It references NUREG-2191 item III.B1.3.TP-45, which is assigned to Lubrite; graphitic tool steel; Fluorogold; Lubrofluor.

The staff has found that Molykote D-321 is not a tool steel or other solid material but a lubricant. Molykote D-321 Safety Data Sheet (SDS) indicates that it may contain up to 10% of molybdenum disulfide (MoS2), which is a potential contributor to stress corrosion cracking (SCC) when there is an environment conducive of such. The staff also notes that Molykote D-321 is not a long-lived material having a usable life of 24 months from the date of production.

It is not clear whether Molykote 321 has been used, currently is being used and also it is planned to be used to lubricate load bearing component surfaces to the end of the SPEO, and if so where.

It is also not clear whether there is a plan to update the relevant implementing procedures to identify Molykote@321 so that it will not be accidently or mistakenly used to lubricate high-strength bolts with nominal diameter greater than 1 inch or on sliding surfaces that are anchored with high-strength bolts with nominal diameter greater than 1 inch or greater.

Further, it is not clear whether 10 CFR Part 54 provides the regulatory framework for managing effects of aging on the lubricant having a usable life of 24 months.

a) Discuss how BFN plans to prevent Molykote@321 from being used for high strength bolts with nominal diameter greater than 1 inch.

b) Provide details of this component type to which the Molykote@321 is planned to be applied.

c) Justify the use of 10 CFR Part 54 to manage effects of aging and loss of function for the limited life Molykote 321 material.

6 Table 3.5.2-36 Pg. 3-965/

3-966 Generic note A is assigned and NUREG 2191 items III.B1.1.T-28, III.B1.2.T-28, and III.B1.3.T-28 are referenced for the 3rd and 5th line items on Pg. 3-965, and the 2nd line item on Pg. 3-966, respectively.

The staff found that the material, i.e., stainless steel, is not in GALL-SLR for this component, i.e., Constant and variable load spring hangers; guides; stops. According to NEI 17-01, the generic note F should be chosen.

a) Discuss the use of generic note A for NUREG 2191 items III.B1.1.T-28, III.B1.2.T-28, and III.B1.3.T-28 associated Table 3.5.2-36 items.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions It is not clear why BFN references generic note A in the aforementioned Table 2 items.

b) Revise the generic note if necessary.

7 B.2.1.30 Pg. B-154 to B-159 The SLRA ASME Section XI, Subsection IWF program does not appear to discuss the reactor vessel (RV) supports, stabilizer, and star truss.

Specifically, there is no discussion on their condition and Industry Operating Experience as it is applied to BFN.

a) Discuss whether and how industry Operating experience regarding inspection of RV supports (i.e., steel skirt and weldment, anchorage, pedestal, RPV stabilizer brackets, stabilizer, rods and trusses etc.) has been considered at BFN.

b) Discuss/explain in general the history of RV support (including steel skirt and weld, anchorage, pedestal, RV stabilizer brackets, stabilizer, rods and trusses, etc.) inspections at BFN, how and when these inspections are/were performed, and any findings of degradation from the inspections. Illustration of inspected parts on an existing drawing(s) will be helpful if possible.

8 A.2.1.30 B.2.1.30 TVA SLR-Pg. A-43

/44 Pg. B-154

/156 From PBD Section 4.1.2, the staff notes that the applicable ASME Code for current 10-year inspection interval for BFN, which began on February 2016, is the ASME XI, 2007 Edition with the 2008 Addenda.

The SLRA states that BFN Unit 1 license (DPR-33) expires at midnight a) Clarify the Inservice Inspection interval for implementing provisions of the ASME XI, 2007

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions BFN -0039 R2 (PBD)

PBD Pg. 9

/12 /30 /47

/63 /66 /68

/70 /72 on December 20, 2033. BFN Unit 2 license (DPR-52) expires at midnight on June 28,2034. BFN Unit 3 license (DPR-68) expires at midnight on July 2, 2036.

The SLRA provides Enhancement No. 9 to Detection of Aging Effects program element (Element 4 in IWF AMP) that discusses the provisions of ASME Code Section XI, 2007 Edition, 2008 Addenda. The staff notes that at time of SLR consistent with the requirements of 10 CFR 50.55a(g)(4)(ii) during the SPEO will be a different code.

10 CFR 50.55a(g)(4)(ii) requires that Inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the ASME Code incorporated by reference in paragraph (a) of 10 CFR 50.55a 18 months before the start of the 120-month inspection interval.

SLRA does not make clear the Inservice Inspection interval for implementing provisions of the ASME XI, 2007 Edition with the 2008 Addenda. It also does not make clear how 10 CFR 50.55a(g)(4)(ii) is followed.

Edition with the 2008 Addenda.

b) Clarify in the SLRA how to follow the requirements of 10 CFR 50.55a (g)(4)(ii).

c) Clarify whether the provisions of ASME Code Section XI, 2007 Edition, 2008 Addenda described in Enhancement 9 will still be applicable during the SPEO.

9 B.2.1.30 N/A BFN-LRA Section B.2.1.33 IWF AMP makes exceptions to NUREG-1801 as follows:

The aging effects for supports of MC components will be managed by the Structures Monitoring Program (B.2.1.36), or Water Chemistry Program (B.2.1.5) with associated One-Time Inspection Program (B.2.1.29) for submerged supports during the extended period of operation.

The SLRA states that The ASME Section XI, Subsection IWF program is an existing condition monitoring program that consists of periodic visual examination of supports for ASME Class 1, 2, 3, and MC piping and Components.

Clarify how these exceptions in the initial license renewal application are addressed in the enhanced IWF AMP during the SPEO.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The SLRA does not make clear how these exceptions in the initial license renewal application are addressed in the enhanced IWF AMP during the SPEO.

10 A.2.1.30 B.2.1.30 Pg. A-44 Pg. B-156 Both SLRA UFSAR A.2.1.30 and AMP B.2.1.30 Detection of Aging Effects program element (AMP Enhancement #10) state, [t]he extent, frequency, and examination methods are designed to detect, evaluate, or repair age-related degradation before there is a loss of component support intended function.

The SLRA does not make clear what specific enhancements to the IWF AMP will be made to the extent, frequency, and examination methods during the SPEO.

Clarify the specifics of the extent, frequency, and examination methods discussed in the enhancement to Detection of Aging Effects program element of SLRA AMP B.2.1.30.

11 B.2.1.30 Pg. B-157 to B-159 SLRA B.2.1.30 OE #2 and #4 discuss the past inspection examples regarding the discovery of the loose bolts on pipe supports. It states that following the identification of the loose bolts the issues were entered into the corrective action program and engineering evaluations were performed. The evaluations determined that sufficient margin was available to document acceptability of the condition.

It is not clear whether the safety factor consistent with BFN CLB was used in the evaluation to determine the safety margin.

Confirm the anchor bolt safety factors that are consistent with BFN CLB as identified in BFN USFAR Appendix C.3.6.1 were used in the OE engineering evaluation to determine the acceptability of condition.

12 B.2.1.30 Pg. B-154 and B-157 SLRA Section B.2.1.30 states in program description that examinations that reveal flaws or relevant conditions that exceed the referenced acceptance standard will be expanded to include additional examinations during the current outage.

The SLRA provide Enhancement 15 to Corrective Actions program element (AMP Element 7) for ensuring that identification of unacceptable conditions triggers an expansion of the inspection scope, in accordance with IWF-2430.

The applicant followed GALL-SLR but the two statements mentioned above conflict with each other, and SLRA does not make clear of a) Clarify specific conditions that trigger an expansion of the inspection during the SPEO.

b) Discuss the OEs (provide CRs if any) with identified flaws that exceed the referenced acceptance criteria and that require corrective

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions specific conditions when additional examinations will be triggered during the current outage of the SPEO.

The staff reviewed OEs identified in NRC Item #128 on the ePortal. The staff finds that the applicant identified the flaws that do not meet the acceptance standards of IWF-3410(a), evaluated the condition for the intended function per IWF-3122.3, and voluntarily took corrective actions (measures) without performing additional examinations as required by IWF-2430.

In addition, the staff noted that the provisions of ASME Code Section XI, 2007 Edition, 2008 Addenda lacks acceptance criteria to be used to verify acceptability of a component support or a portion of a component support for service by evaluation or test in accordance with IWF-3122.3.

The IWF scope of inspection for supports is based on sampling depending on the ASME Class of support on a percentage of the total support population as required by Table IWF-2500-1 of the Code. The intent of additional examinations is to ensure the full extent of deficiencies is identified for additional component supports during the current outage.

measures, which trigger additional examinations in accordance with IWF-2430.

c) Discuss the OEs and clarify acceptance criteria that were used to accept the component support for service by IWF-3122.3.

13 B.2.1.30 Pg. B-154 In the second line of the fourth paragraph within Program Description Section, the. preceding every 10 years appears to be a typographical error.

Suggest correction if this is a typographical error.

14 TVA SLR-BFN-0039 R2 (PBD)

Section 3.1.2 PBD Pg.

9/10 There appears to be a typographical error, i.e., instead of using ASME, AMSE is used in three places.

Suggest correction of the typographical error.

15 B.2.1.30 Pg. B-158 First sentence of OE #2 A review of implementing activities and Corrective Action Program condition reports generated as a result of those activities shows was conducted examples reviewed include does not appear to be phrased properly.

Suggest revising as appropriate.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 16 Table 3.5.2-36 Pg. 3-971 and 3-982 ASME Section XI, Subsection IWF AMP (B.2.1.30) is assigned to the last line item on Pg. 3-971 with generic note F and plant-specific note

2.

On Pg. 3-982, plant-specific note 2 for Table 3.5.2-36 states Consistent with NUREG-2191 Volume 2, steel, stainless steel, and cast iron materials are susceptible to corrosion. The Structures Monitoring program (B.2.1.33) will manage the loss of material due to corrosion for the steel fire barriers, stainless steel controlled leakage doors, miscellaneous structural steel, penetrations, pipe whip restraints and jet impingement shields, non-ASME stainless steel piping support members; welds; bolted connections; support anchorage to building structure; whereas the ASME Section XI, Subsection IWE program (B.2.1.29) will manage the ASME Class 2 cast iron, steel, and stainless steel piping support members; welds; bolted connects; support anchorage to building structure; and the One-Time Inspection program (B.2.1.20) will manage the loss of material and cracking due to SCC for the stainless steel and aluminum components in the Miscellaneous Steel commodity Group as well as for the stainless steel components in the Pipe Whip Restraints and Jet Impingement Shields Commodity Group.

The staff does not find ASME Section XI, Subsection IWF AMP (B.2.1.30) be included in the plant-specific note 2.

It is not clear why plant-specific note 2 is applicable to this line item or whether ASME Section XI, Subsection IWF AMP is the intended AMP for this line item.

a) Clarify in plant-specific note 2 whether or not the mention of ASME Section XI-IWE (B.2.1.29) AMP is intended to mean ASME Section XI-IWF (B.2.1.30) and correct accordingly.

17 B.2.1.30 B-154 Regulatory Basis:

Title 10 of the Code of Federal Regulations Section 54.21(a)(3) requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis (CLB) for the period of extended operation. As described in the SRP-SLR, an applicant may demonstrate compliance

1. Provide an evaluation concerning the effectiveness of the SLRA IWF AMP and its implementation to ensure that deficiencies in the remaining piping supports are identified

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions with 10 CFR 54.21(a)(3) by referencing the GALL-SLR Report when evaluation of the matter in the GALL-SLR Report applies to the plant.

Background:

The detection of aging effects program element in GALL-SLR AMP XI.S3 includes a supplemental provision to include a one-time inspection of an additional 5 percent of the sample size specified in Table IWF-2500-1 for Class 1, 2, and 3 piping supports. The one-time inspection is conducted within 5 years prior to entering the subsequent period of extended operation. The additional supports are selected from the remaining population of IWF piping supports. However, the responsible engineer should ensure that the sample includes components that are most susceptible to age-related degradation (i.e.,

based on time in service, aggressive environment, etc.).

The technical bases for including this provision in GALL-SLR AMP provided in Table 2-30 on page 2-384 of NUREG-2221 (ML17362A126) include: (1) verifying the same IWF representative sample of supports that are periodically inspected is indeed representative of the entire population, including the supports that have never been inspected; (2) identifying age-related degradation occurring in supports that have never been inspected; (3) use of plant operating experience as an important consideration in determining need for additional activities for 60-80 year period of operation; and (4) addition of select number of random inspections and focused inspections inclusive of aging effects or environments most susceptible to degradation to allow for reasonable assurance that the IWF AMP sample will be representative of the aging of the entire component support population during the subsequent period of extended operation (SPEO).

SLRA Section B.2.1.30 provides the related enhancement (Commitment 30.9 in SLRA Table A.5 on page A-121) to the detection of aging effects program element in accordance with GALL-SLR XI.S3 AMP for including one-time inspection of an additional 5 percent of the sample size specified in Table IWF-2500-1 for Class 1, 2, and 3 piping and corrected prior to loss of intended function and the inspected sample remains representative of the entire population of piping supports during the SPEO.

2. Clarify, for the one-time inspection of additional 5-percent of the sample size commitment, whether the applicant would include additional examination (scope expansion) if relevant conditions are identified in one or more piping supports that do not meet Design Specification and Owners Requirements in the CLB.

2.a. If yes, clarify how additional examination (scope expansion) will be implemented such that the described intent/technical bases for the supplemental one-time inspection are met.

2.b. If not, provide technical justification

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions supports. The one-time inspection will be conducted within 5 years prior to entering the subsequent period of extended operation.

The NRC staff reviewed OEs identified in NRC Item #128 and breakout session information on the ePortal and notes that: (1) the applicant identified relevant conditions (such as loose bolts for piping supports and improper cold settings of spring supports) that are unacceptable for continued service per the acceptance standards of IWF-3410(a); (2) the applicant subjectively evaluated the relevant conditions that do not meet the acceptance standards of IWF-3410 per IWF-3122.3; and (3) the applicant voluntarily took corrective measures to restore the component support to its original design condition. However, the NRC staff noted that corrective measures were not taken in accordance with IWF-3122.2 even when the CLB design specifications and owners requirements were not met. In addition, the NRC staff noted that the provisions of ASME Code Section XI, Subsection IWF, 2007 Edition, 2008 Addenda lack acceptance criteria to be used to verify acceptability of a component support or a portion of a component support for service by evaluation or test in accordance with IWF-3122.3.

During the breakout session on 7/9/2024, the applicant explained that they used ASME Section XI interpretation XI-1-86-30R that defines this evaluation as analyzed and/or tested to substantiate its integrity for its intended service, and asserted that their process is aligned with the industry, and no changes are necessary to their IWF evaluation process, methods, or acceptance standards. In addition, the applicant used the same sample of piping supports for examinations in accordance with ASME Section XI, Subsection IWF.

The NRC staff further reviewed the evaluations provided by the applicant on the ePortal and noted the following different approaches used (1): the applicant used design acceptance criteria for accepting piping support for continued service in the as-found configuration (e.g.

NOI U1R10-001); (2): the applicant concluded that pipe support anchors would have been overstressed in the tension capacity by approximated 18% and 21% based on design acceptance criteria but how the one-time inspection of an additional 5 percent sample of piping supports and the IWF AMP would be effective in adequately managing aging effects consistent with 10 CFR 54.21(a)(3).

3. Revise the SLRA accordingly consistent with the responses above.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions the applicant accepted pipe support for continued service based on being able to perform its design function (e.g. UNI U3R20-001); and (3) the applicant applied design acceptance criteria but used engineering judgment to state that the riser clamp would have been expected to resist this load and be fully functional based on the configuration of the riser clamp, bolts and nuts that would limit the scenario of the nut fully falling off and the bolt completely falling out of the clamp (e.g. UOI U3R19-001). These inconsistent and subjective evaluation approaches used to demonstrate that the intended function of the degraded piping supports are not necessarily consistent with the CLB prior to their corrective measures, and resulted in no corrective measures or repair/replacement activity in accordance with IWF-3122.2, and therefore would not trigger additional examinations (scope expansion) in accordance with IWF-2430. Also, during the breakout session, the applicant was not able to point to any case where additional examinations were performed in accordance with IWF-2430 when unacceptable relevant conditions were found. Therefore, the applicants evaluation approaches do not appear to be adequate to meet requirements set in 10 CFR 54.21(a)(3).

The IWF scope of inspection for piping supports is based on sampling of the total support population. The sample size varies depending on the ASME Class. Discovery of support deficiencies during regularly scheduled inspections should trigger an increase of the inspection scope in order to ensure that the full extent of deficiencies is identified and corrected, as necessary, and the inspected sample remains representative of the entire population of piping supports. Supports requiring corrective actions are re-examined during the next inspection period.

Issue:

Based on the applicants implementation approach described above, the applicant is able to circumvent the requirement in IWF-2430 for additional examinations (scope expansion) of relevant conditions exceeding the acceptance standard of IWF-3400 that do not meet the CLB design specifications and owners requirements, and thereby

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions requiring corrective measures or repair/replacement in accordance with IWF-3122.2. This implementation appears contrary to the acceptance criteria for detection of aging effects AMP element in SRP-SLR Section A.1.2.3.4 and appears to render the sample-based part of the AMP ineffective for aging management. The NRC staff is seeking to understand how aging effects of the remaining piping supports (75% to 90%) that have not been examined will be adequately managed during the SPEO. The SLRA is unclear how one-time inspection of additional 5-percent of the sample size commitment will be implemented, with regard to inspection scope expansion and additional actions if unacceptable relevant conditions are found, to assure that the intent of the supplemental one-time inspection described in the Background section will be met to assure adequate aging management during the SPEO.

SLRA Section B.2.1.31 10 CFR Part 50, Appendix J Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

B.2.1.31; TVA SLR-BFN-0042 (Program Basis

Document, PBD)

Section 3.1.2 Pg. B-159/160; PBD Pg. 9 Consistency of Scope of Program Section 3.1.2 of PBD states, in part: Components required to be leak rate tested or excluded from testing are identified in UFSAR Table 5.2-

2. The one-time inspection (SLR-BFN-0065) program, water chemistry (SLR-BFN-0040) program, inspection of internal surfaces in miscellaneous piping and ducting components (SLR-BFN-0045) program, external surfaces monitoring of mechanical components SLR-BFN-044) program, BWR stress corrosion cracking (SLR-BFN-0041) program, ASME Section XI inservice inspection, Subsection IWB, IWC, and IWD (SLR-BFN-0032) program and the Flow-Accelerated Corrosion (SLR-BFN-0063) program are included among the aging management programs that manage the aging effects associated with components excluded from leak rate testing..

While SLRA Section B.2.1.31 states, in part: Components required to be leak rate tested or excluded from testing are identified in UFSAR

1. Discuss whether BFN plans to revise SLRA Section 2.1.31 Program Description section to make clear that the aging effects of the components that are within the scope of SLR but are excluded from Type B or C Appendix J testing will be

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Table 5.2-2., it does not include or state the AMPs, as identified in the PBD, that will be used to manage the aging effects of components excluded from testing.

The scope of program element of GALL-SLR (NUREG-2191) AMP XI.S4 states, The aging effects associated with containment pressure-retaining boundary components within the scope of subsequent license renewal and excluded from Type B or C Appendix J testing must still be managed. Other programs may be credited for managing the aging effects associated with these components; however, the component and the proposed AMP should be clearly identified..

It is not clear for those components (valves, penetrations, and others) that have been excluded from 10 CFR Part 50 Appendix J program whether they are included in the scope and which AMPs will be used to manage the aging effects for each of the excluded components. The staff is unable to verify consistency with the scope of program element of the SLRA AMP with the corresponding element of GALL-SLR AMP XI.S4.

managed by other AMPs identified in the PBD to be in consistent with the scope of program element of GALL-SLR (NUREG-2191)

AMP XI.S4.

2. Identify the SLRA Tables (i.e.,

Table 2s) that contain the AMR line items for these excluded components.

SLRA Section B2.1.32 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

B.3.1.32 Masonry

Walls, SLR-BFN-0028 Masonry Walls, Rev 1,

LCEI-CI-C9 SLRA B-163 SLRA Section B.3.1.32, Masonry Walls states, The enhanced Masonry Walls aging management program will be consistent with the 10 elements of NUREG-2191 Subsection XI.S5.

There are 10 enhancements spread amongst Elements 3, 5, 6, and 7, and all start with, Revise Implementation procedures... Furthermore, the AMP Basis document on the portal appears to reference LCEI-CI-C9, Procedure for Walkdown of Structures for Maintenance Rule and Licensing.

Section 7.2.5, Masonry Walls, of LCEI-CI-C9 provides a brief description of what Masonry Walls shall be inspected for. It is unclear

1. Please provide a markup of the implementation procedure for the 10 proposed enhancements that clearly demonstrates how each enhancement relates to the respective program element.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions how this procedure will be enhanced for each of the element categories, 3) Parameters Monitored or Inspected, 5) Monitoring and Trending, 6) Acceptance Criteria, and 7) Corrective actions, because the procedure does not have separate subsections related to the enhanced elements.

2 B.3.1.32 Masonry

Walls, BFN-50-c-7100 B-164 Enhancements 7 states in part, require each masonry wall that has observed degradation to be assessed against the evaluation basis.

Section 2.0, Evaluation Basis, of BFN-50-C-7100, is strictly related to design of the masonry walls. It is unclear which evaluation basis document applies to degradation assessment for the masonry walls.

Enhancements 8, 9, and 10 make references to acceptance criteria, however, there is no acceptance criteria in procedure LCEI-CI-C9.

1. Please provide the evaluation basis document for masonry walls degradation assessment on the portal for review.

2. Please provide the implementation procedure that provides the acceptance criteria for the masonry walls inspection program.

3 SLR-BFN-0028 Masonry Walls, Rev 1

Section 4.1.5 Section 4.1.5 states in part, BFN Masonry Walls AMP with enhancements is consistent with NUREG-2191 XI.S5 element 1.

However, Section 4.1.4 indicates there are no enhancements for element 1.

3. Clarify whether an enhancement applies to Element 1.

SLRA Section B.2.1.33: Structures Monitoring Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

B.2.1.33 B-166 GALL-SLR XI.S6 AMP includes the following SCs and structural commodities in the scope of program, such as transmission towers, panels and other enclosures, racks, electrical cable trays and conduits, Scope of the Program:

1. Clarify whether these SCs and structural

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions trash racks and traveling screens associated with water-control structures, electrical duct banks, manholes, penetration seals, and structural sealants.

SLRA does not make clear whether these SCs and structural commodities are already included in the existing Structures Monitoring program.

It appears that SLRA does not include Tables 1 and 2 AMR items for the SCs and structural commodities mentioned above.

commodities are within the scope of existing Structures Monitoring program. If not, provide an enhancement to including these SCs and structural commodities.

2. Provide Tables 1 and 2 items for those missing SCs and structural commodities.

2 B.2.1.33 B-167 SLRA XI.S6 AMP provides an enhancement No. 3 to the parameters monitored or inspected program element for revising implementing procedures to be consistent with ACI 349.3R-02 and SEI/ASCE 11-99.

During onsite audit discussions, the applicant will voluntarily delete word coatings and clarify roof waterproofing membranes.

SLRA Table 3.5.2-36 lists membrane with aging effect of loss of weatherproofing integrity due to cracking, dying, organic decomposition, separation, shrinkage, wear, weathering.

ACI 349.3R-02, Section 5.1.4 states that absence of degradation in any waterproofing membrane protecting below-grade concrete surfaces within the inspected area, which is different application of waterproofing membrane from roof waterproofing membrane.

Parameters monitored or inspected:

Address the conflict and provide parameters monitored or inspected for roof waterproofing membrane.

3 B.2.1.33 Table 3.5.2-1 B-168 3-752 3-754 SLRA XI.S6 AMP provides an enhancement No. 5 to parameters monitored or inspected program element for revising implementing procedures to ensure steel, aluminum, and non-ferrous material components are monitored for cracking, loss of material due to corrosion or mechanical wear, and general degradation.

Parameters monitored or inspected and acceptance criteria:

1. Address the inconsistency of aging effects for the aluminum

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA XI.S6 AMP also provides an enhancement No. 20 to the acceptance criteria program element for revising implementing procedures to ensure no significant cracking, no significant loss of material due to corrosion or mechanical wear, and no significant signs of general degradation for steel, aluminum, and non-ferrous material components.

Wording no significant and general degradation are very vague.

SLRA Table 3.5.2-1 lists aluminum spent fuel pool gates bolting and non-ferrous structural bolting with aging effect of loss of preload due to self-loosening, which is managed by the Structures Monitoring program.

It appears that there is inconsistency of aging effects between enhancements and SLRA Table 2 AMR items.

In addition, SLRA Table 2 AMR items do not include aging effect of cracking (due to SCC) for steel, aluminum, and non-ferrous material components managed by the Structures Monitoring program.

and non-ferrous material components between the enhancements No. 5 and 20 to the Structures Monitoring program and SLRA Table 2 AMR items.

2. Provide Table 2 items for steel, aluminum, and non-ferrous material components with aging effect of cracking (due to SCC).

3. Discuss general degradation listed in the enhancements.

4. Evaluate enhancement No. 20 to determine whether the acceptance criteria can be derived from applicable codes and standards.

4 B.2.1.33 Table 3.5.2-36 B-168 3-925 GALL-SLR XI.S6 AMP states that elastomeric vibration isolators, structural sealants, and seismic joint fillers are monitored for cracking, loss of material, and hardening.

SLRA XI.S6 AMP provides an enhancement No. 7 to the parameters monitored or inspected program element Parameters monitored or inspected:

1. Clarify whether membrane is a roof waterproofing membrane.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions for revising implementing procedures to include membranes in addition to elastomeric vibration isolators, structural sealants, and seismic joint fillers.

During the on-site audit, the applicant clarified that membrane is a roof waterproofing membrane. Combining membranes with elastomeric vibration isolators, structural sealants, and seismic joint fillers are confusing since they have different aging effects.

SLRA Table 3.5.2-36 lists membrane with aging effect of loss of weatherproofing integrity due to cracking, dying, organic decomposition, separation, shrinkage, wear, weathering, which is different from aging effect listed in the enhancement No. 7.

2. Address the inconsistency of aging effects for membranes between the enhancement No. 7 and SLRA Table 2 items.

2. Revise enhancement No. 7 by separating membranes from other components.

5 B.2.1.33 B-168 SLRA XI.S6 AMP provides an enhancement No. 8 to the parameters monitored or inspected program element for revising implementing procedures to include monitoring for earth berms for loss of material and loss of form due to erosion, settlement, sedimentation, frost action, waves, currents, surface runoff, and seepage.

SLRA XI.S6 AMP also provides an enhancement No. 26 to the acceptance criteria program element for revising implementing procedures to add acceptance criteria for earth berms. Wording no significant loss of material and no significant loss of form is very vague.

For similar structures managed by the Inspection of Water-Control Structures program, the instruments are usually installed to measure the behavior of water-control structures (e.g. Dam, Embankment).

However, the Structures Monitoring program consists primarily of periodic visual inspections, and it may not manage these aging effects adequately.

GALL-SLR XI.S7 AMP states that acceptance criteria for earthen structures, such as canals and embankments, are consistent with programs falling within the regulatory jurisdiction of the FERC or the USACE Parameters monitored or inspected and acceptance criteria:

1. Explain how the Structures Monitoring program can detect aging effects of loss of material and loss of form for rock and earth fill embankment by visual inspections.

2. Evaluate and revise acceptance criteria for earth berms.

3. Discuss the OE (CR 1682962 and WO 12203978) for how loss of material and loss of form are determined.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 6

B.2.1.33 B-168 SLRA XI.S6 AMP provides an enhancement No. 9 to the parameters monitored or inspected program element for revising implementing procedures to monitor groundwater chemistry on a frequency not to exceed five years, which includes an enhancement to the detection of aging effect program element.

Parameters monitored or inspected and Detection of aging effects:

Clarify whether enhancement No. 9 includes both parameters monitored or inspected and detection of aging effect program elements.

7 B.2.1.33 B-168 GALL-SLR XI.S6 AMP states that leakage volumes and chemistry are monitored and trended for signs of concrete or steel reinforcement degradation.

SLRA XI.S6 AMP provides an enhancement No. 10 to the parameters monitored or inspected program element for revising implementing procedure to monitor and trend for signs of concrete or steel reinforcement degradation if through-wall leakage or groundwater infiltration is identified.

The enhancement No. 10 is inconsistent with GALL-SLR report.

Parameters monitored or inspected:

Revise enhancement No.

10 to be consistent with GALL-SLR report.

8 B.2.1.33 B-169 SLRA XI.S6 AMP provides an enhancement No. 16 to the monitoring and trending program element for revising implementing procedures to establish quantitative baseline inspection data for structures and structural commodities for which baseline inspection data has not been established or for which the baseline acceptance criteria used are not comparable to the GALL-SLR acceptance criteria, prior to the subsequent period of extended operation.

SLRA on page 3-682 states that Test results for groundwater and raw water samples showed that at a limited number of locations (Monitoring Wells (MW): MW-10, MW-11, MW-13 and MW-17), the pH limits are Monitoring and trending:

1. Clarify whether new quantitative baseline inspection data for structures and structural commodities exposed to aggressive groundwater/soil environment needs to be established prior to the SPEO.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions below the lower threshold limit for pH of 5.5 per NUREG-1557, Table B3, and therefore qualify as aggressive.

SLRA does not make clear whether new quantitative baseline inspection data for structures and structural commodities exposed to aggressive groundwater/soil environment needs to be established.

SLRA does not make clear whats the baseline acceptance criteria used that are not comparable to the GALL-SLR acceptance criteria.

2. Clarify what structures and structural commodities have the baseline acceptance criteria used that are not comparable to the GALL-SLR acceptance criteria.

9 B.2.1.33 B-169 SLRA XI.S6 AMP provides an enhancement No. 22 to the acceptance criteria program element for revising implementing procedures to add acceptance criteria for structural steel bracing and edge supports associated with masonry block walls.

The enhancement No. 22 does not address aging effect of loss of material for structural steel bracing and edge supports associated with masonry block walls.

Acceptance criteria:

Clarify the acceptance criteria for loss of material for structural steel bracing and edge supports associated with masonry block walls.

10 B.2.1.33 B-170 GALL-SLR XI.S6 AMP states that structural sealants are acceptable if the observed loss of material, cracking, and hardening will not result in loss of sealing, and elastomeric vibration isolation elements are acceptable if there is no loss of material, cracking, or hardening that could lead to the reduction or loss of isolation function.

SLRA XI.S6 AMP provides an enhancement No. 25 to the acceptance criteria program element for revising implementing procedures to add acceptance criteria for elastomeric vibration isolators, membranes, structural sealants, and seismic joint fillers.

SLRA Table 3.5.2-36 lists membrane with aging effect of loss of weatherproofing integrity due to cracking, dying, organic decomposition, separation, shrinkage, wear, weathering, which is different from aging effect of membranes listed in this enhancement.

Acceptance criteria:

1. Clarify whether elastomeric membrane is a roof waterproofing membrane.

2. Address the inconsistency of aging effects for membranes.

3. Revise enhancement No. 26 by separating membranes form structural sealants and seismic joint filler.

4. Address the inconsistency of the

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA does not make clear whether membranes are roof waterproofing membranes. In addition, this enhancement No. 25 is inconsistent with GALL-SLR report.

enhancement No. 25 with the GALL-SLR report.

11 B.2.1.33 B-170 SLRA on page 3-682 states that test results for groundwater and raw water samples showed that at a limited number of locations (Monitoring Wells (MW): MW-10, MW-11, MW-13 and MW-17), the pH limits are below the lower threshold limit for pH of 5.5 per NUREG-1557, Table B3, and therefore qualify as aggressive.

SLRA on page 3-683 also states that soil was found not to be aggressive.

Attachment F Groundwater well low pH evaluation of CDQ0003032023000000, Revision 2, shows that some structures are located below the water table level. SLRA does not make clear of the relationship between water table levels and underground concrete elevations of the structures exposed to the aggressive groundwater/soil environment.

XI.S6 AMP basis document, Section 4.4.2 on page 27, states that BFN has been determined to have non-aggressive groundwater/soil (pH is less than 5.5, chlorides is greater than 500 ppm, or sulfates is greater than 1500ppm). However, GALL-SLR XI.S6 AMP defines non-aggressive groundwater/soil as pH is greater than 5.5, chlorides is less than 500 ppm, and sulfates is less than 1500ppm.

UFSAR O.1.34 Inspection of Water-Control Structures Program on page O.1-14 states that raw water in close proximity to the Intake Pumping Station is required to be periodically monitored for the requirements of an aggressive environment as described in NUREG-1557.

SLRA XI.S6 AMP provides an enhancement No. 27 to the acceptance criteria program element for revising implementing procedures to Detection of aging effects:

1. Clarify whether the enhancement to the acceptance criteria program element can be changed into the enhancement to the detection of aging effects program element for consistency with GALL-SLR report.

2. Discuss the precedent and provide specific and detailed information for actions that will be taken during the SPEO to adequately address aging management of both accessible and inaccessible areas exposed to potentially aggressive groundwater/soil environment.

3. Discuss whether there is a need to add additional monitoring wells for testing from

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions ensure the groundwater chemistry within the specified ranges and include evaluations, destructive testing and focused inspections and so on if groundwater chemistry values indicate aggressive groundwater/soil environment, which is related to the detection of aging effects program element in the GALL-SLR XI.S6 AMP. In addition, this enhancement lacks detailed information how to address aging management of inaccessible areas exposed to aggressive groundwater/soil environment.

The NRC has approved the Oconee SLRA for addressing this issue (ML21349A005), which is one of the methods that is acceptable to the NRC.

locations that are representative of the potential aggressive groundwater in contact with structures within the scope of SLR and define the areas of the structures subject to aggressive groundwater/soil environment for more focused inspections.

4. Explain how aging effects of inaccessible concrete areas are adequately managed and clarify whether the observed degradations (all aging effects) in accessible areas have impact on the intended function of the concrete structures.

5. Discuss the test data and criteria used for the soil to be non-aggressive.

6. Address the inconsistency of groundwater chemistry values between AMP basis document and GALL-SLR report.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions

7. Revise relevant SLRA accordingly based on the responses above.

12 B.2.1.33 B-169 Section 4.6.2 in XI.S6 AMP basis document describes acceptance criteria for concrete surfaces based on the second-tier evaluation criteria provided in Chapter 5 of ACI 349.3R-02.

SLRA or plant procedures do not make clear whether this acceptance criteria is used in the existing Structures Monitoring program.

Acceptance criteria:

Clarify whether an enhancement to the Structures Monitoring program is needed to include this acceptance criteria for concrete surfaces based on the second-tier evaluation criteria.

13 B.2.1.33 B-171 SLRA XI.S6 AMP states in OE item No. 2, Although the conditions were evaluated at the time of identification of the degradation, no evaluation of the effect on aging management as a result of canceling or delaying these work orders was performed. This condition was entered into the Corrective Action Program with the actions taken which include an evaluation of each delayed or canceled WO for the effect on the component or structure.

SLAR does not make clear why these work orders were cancelled or delayed and whether the SCs with degradations in these delayed work orders are repaired or replaced.

Operating experience:

1. Provide the justification for the work orders that were cancelled.

2. Explain why work ordered are delayed and clarify whether the SCs with degradations in these delayed work orders are repaired or replaced so that their intended functions are maintained.

14 Table 3.5.2-1 3-747 GALL-SLR Table IX.D states that water in the definition of water-flowing can include rainwater, raw water, groundwater, or water flowing under a foundation.

SLRA Table 3.5.2-1 list concrete components in reactor building exposed to condensation, groundwater, and treated water environment

1. Clarify whether groundwater environment can be eliminated.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions (citing notes G and 4) for the Structures Monitoring program to manage increase in porosity and permeability; loss of strength due to leaching of calcium hydroxide and carbonation.

Note 4 states that the reactor building is exposed to condensation, groundwater, treated water. The listed combination was screened in to conservatively capture all potential combinations that stem from these environments.

GALL-SLR Table IX.D states on page IX D-6 that there are two categories of treated water, the first category generally contains minimal amounts of any additions. The second category contains corrosion inhibitors and also may contain biocides or other additives.

SLRA does not make clear whether concrete components will have other applicable aging effects when they are exposed to treated water environment.

2. Describe where the treated water is generally used.

3. Evaluate whether treated water environment will trigger other aging effects for the concrete components.

15 Table 3.5.2-1 3-750 GALL-SLR item III.A2.TP-29 lists concrete components exposed to groundwater/soil environment.

SLRA Table 3.5.2-1 list concrete components in reactor building exposed to treated water environment (citing notes G and 4) for the Structures Monitoring program to manage increase in porosity and permeability; loss of strength due to leaching of calcium hydroxide and carbonation.

Note 4 states that the reactor building is exposed to condensation, groundwater, treated water, where groundwater does not match groundwater/soil in the GALL-SLR report.

SLRA does not include Table 2 items associated with Table 1 AMR item 3.5-1, 067 for concrete components exposed to condensation environment.

GALL-SLR Table IX.D states on page IX D-6 that there are two categories of treated water, the first category generally contains

1. Provide Table 2 items associated with Table 1 AMR item 3.5-1, 067 for concrete components exposed to condensation environment.

2. Discuss terms used in Note 4 for the consistency with GALL-SLR report.

3. Describe where the treated water is generally used.

4. Evaluate whether treated water environment will trigger

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions minimal amounts of any additions. The second category contains corrosion inhibitors and also may contain biocides or other additives.

SLRA does not make clear whether concrete components will have other applicable aging effects when they are exposed to treated water environment.

other aging effects for the concrete components.

16 Table 3.5.2-36 3-979 Table 3.5.2-36 lists steel sliding support bearings and sliding support surfaces exposed to air-indoor and air-outdoor environment (citing notes F and 8) for managing loss of mechanical function due to corrosion, distortion, dirt or debris accumulation, overload, wear.

Note 8 states, Though noted for Class 1, NUREG 2191 Volume 2 notes that steel sliding surfaces can be susceptible to a loss of mechanical function. Conservatively, this aging effect will be monitored at BFN using the Structures Monitoring program consistent with steel sliding surfaces being scoped into the Structures Monitoring Program by NUREG 2191 Volume 2.

GALL-SLR Table IX.E states on page IX E-4, Loss of mechanical function in Class 1 piping and components (such as constant and variable load spring hangers, guides, stops, sliding surfaces, and vibration isolators) fabricated from steel or other materials, such as Lubrite, can occur through the combined influence of a number of aging mechanisms. Such aging mechanisms can include corrosion, distortion, dirt accumulation, overload, fatigue due to vibratory and cyclic thermal loads, or elastomer or polymer hardening.

Clearances being less than the design requirements can also contribute to loss of mechanical function.

SLRA does not make clear where these components are located and whether other aging mechanisms or aging effects will be applicable to these components fabricated from steel.

1. Provide drawing or photos for these components for discussions.

2. Clarify whether there are materials listed in GALL-SLR item III.B2.TP-46/47 fabricated for these components.

3. Evaluate whether sliding support bearings and sliding support surfaces fabricated from steel will trigger other aging mechanisms or aging effects.

17 Table 3.5.2-31 3-890 3-940 SLRA Table 3.5.2-31 lists steel components exposed to groundwater/soil environment (citing notes G and 2) for the Structures Monitoring program to manage loss of material due to corrosion.

1. Describe steel components in a buried environment.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Table 3.5.2-36 Note 2 on page 3-892 states, Steel with a buried environment that includes groundwater/soil is conducive to the potential for loss of material due to corrosion. The Structures Monitoring program will be used to manage this aging effect.

SLRA does not make clear how the Structures Monitoring program can detect aging effects for the buried steel components.

SLRA Table 3.5.2-36 lists steel components exposed to treated water environment (citing notes G and 2) for the Structures Monitoring program to manage loss of material due to corrosion.

SLRA Table 3.5.2-36 lists steel support members and support anchorage to building structure exposed to treated water environment (citing notes G and 2) for the Structures Monitoring to manage loss of material due to general, pitting corrosion.

Note 2 on pages 3-982 and 1250 states, Consistent with NUREG-2191 Volume 2, steel, stainless steel, and cast iron materials are susceptible to corrosion. The Structures Monitoring program will manage the loss of material due to corrosion for the steel fire barriers, stainless steel controlled leakage doors, miscellaneous structural steel, penetrations, pipe whip restraints and jet impingement shields, non-ASME stainless steel piping support members; welds; bolted connections; support anchorage to building structure; whereas the ASME Section XI, Subsection IWE program will manage the ASME Class 2 cast iron, steel, and stainless steel piping support members; welds; bolted connects; support anchorage to building structure; and the One-Time Inspection program will manage the loss of material and cracking due to SCC for the stainless steel and aluminum components in the Miscellaneous Steel commodity Group as well as for the stainless steel components in the Pipe Whip Restraints and Jet Impingement Shields Commodity Group.

2. Explain how the visual examinations of the Structures Monitoring program can detect aging effects for the buried steel components.

3. Describe where the treated water is generally used.

4. Evaluate whether treated water environment will trigger other aging effects for the steel components.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions GALL-SLR Table IX.D states on page IX D-6 that there are two categories of treated water, the first category generally contains minimal amounts of any additions. The second category contains corrosion inhibitors and also may contain biocides or other additives.

SLRA does not make clear whether concrete components will have other applicable aging effects when they are exposed to treated water environment.

18 Table 3.5.2-36 3-920 3-921 SLRA Table 3.5.2-36 lists non-ferrous copper alloys controlled leakage door exposed to air-indoor and air-outdoor environment for the Structures Monitoring program to manage loss of material due to mechanical wear.

SLRA Table 3.5.2-36 also lists stainless steel controlled leakage door exposed to air-outdoor environment for the Structures Monitoring program to manage loss of material due to mechanical wear.

The staff noted loss of material due to corrosion and loss of material due to general, pitting, crevice corrosion for steel controlled leakage door.

SLRA does not make clear whether loss of material due to corrosion and loss of material due to general, pitting, crevice corrosion are applicable to non-ferrous copper alloys controlled leakage door and stainless steel controlled leakage door.

SLRA also does not make clear whether stainless steel controlled leakage door is exposed to air-indoor environment.

1. Evaluate whether loss of material due to corrosion and loss of material due to general, pitting, crevice corrosion are applicable to non-ferrous copper alloys controlled leakage door and stainless steel controlled leakage door.

2. Clarify whether stainless steel controlled leakage door is exposed to air-indoor environment. If yes, provide Table 2 AMR items.

19 Table 3.5.2-36 3-925 SLRA Table 3.5.2-36 lists membrane component exposed to air-outdoor environment for the Structures Monitoring program to manage loss of weatherproofing integrity due to cracking, drying, organic decomposition, separation, shrinkage, wear, weathering.

SLRA does nor make clear where the membrane is used and how the Structures Monitoring program.

1. Clarify whether this membrane is roof waterproofing membrane.

2. Explain how the Structures Monitoring

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions program can adequately manage the aging effects of the membrane.

20 Table 3.5.2-36 3-924 GALL-SRP states on page 2.4-1 that these passive, long-lived SCs are subject to an AMR.

SLRA Table 3.5.2-36 lists radiation protection blankets exposed to air-indoor environment (citing notes J and 7) for the Structures Monitoring program to manage SS mesh cracking, decomposition due to general degradation.

SLRA does not make clear where the radiation protection blankets are used and why they are subject to aging management.

1. Clarify where the radiation protection blankets are used

2. Explain why radiation protection blankets are subject to aging management.

SLRA Section B.2.1.36: XI.E1 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

B.2.1.36; B-183 NUREG 2191 Vol 2. under the detection of aging effects element notes the cable and connection electrical insulation condition monitoring program utilizes a sampling criteria. In addition, the BFN basis document SLR-BFN-0047 includes a technical basis for sample selection. However, the Accessible Non-Environmental Qualification Cables and Connections Inspection Program 0-TI-566 Rev. 0003 is silent on the sample size criteria.

Explain the sample size criteria in the implementing inspection program and why the delta between the GALL/SLR, BFN basis document and implementing procedure?

SLRA Section B.2.1.38/39/40: XI.E3A, B, C Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

A.2.1.38 /

B.2.1.38; A.2.1.39 /

B.2.1.39; A-58 / B-191; A-59 / B-195; Appendix A (FSAR) last paragraph in section A.2.1.38 (and 39 and

40) state that the inspections and tests that are required to be completed prior to entering Subsequent PEO will be completed no When is the latest time prior to subsequent PEO that initial inspections and cable testing may be

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions A.2.1.40 /

B.2.1.40 A-59 / B-199 later than six months prior to subsequent PEO, or no later than the last refueling outage prior to the subsequent PEO.

Section B.2.1.38 (and 39 and 40) (AMPs) state the first inspections and tests will be conducted prior to subsequent PEO.

performed? The FSAR has tighter restrictions than what is listed in the AMP.

Will the last refueling outage prior to SPEO be earlier than 6 months prior to SPEO?

2 SLR-BFN-

0051, Section 4.6.2 23 This document is the Basis document for Inaccessible Low-Voltage power cables, and the last paragraph in this section states Visual inspection results show that instrumentation and control cable jacket material are free from unacceptable..

Verify whether this paragraph should reference Low-Voltage power cable jacket material, vs I&C jacket material.

3 B.2.1.38/

39/40 B-192; B-196; B-199 NUREG-2191 Consistency paragraph for each AMP does not address whether the AMP will be consistent with the Guidance contained in SLR-ISG-2021-04-ELECTRICAL.

Explain whether the ISG-2021-04 guidance was incorporated into these Inaccessible Cable AMPs.

4 G-38, Section 19.4.7.F.4 Pg 174 The repeat test frequency for EPMI type cables states EPMI cables are moisture impervious. No periodic testing is required.

AMP B.2.1.38 states Test frequencies are adjusted based on test results and plant-specific OE. This indicates that EPMI cables may require periodic testing based on the plant-specific OE or test results.

Clarify whether the current wording in G-38 could cause discrepancies with the AMP for EPMI cables with the current wording, if plant or industry-OE were to indicate the need for periodic testing be performed.

5 RIMS No:

R27 090812 306 pg 37-40 of 67 The water penetration resistance test report documents that the tests failed for both Part A (pg 39) and Part B (pg 40).

Explain how these cable tests support industry requirements of IEEE 383 for being qualified for long-term cable submergence.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 6

G-38, Section 19.2 &

19.3 Pg 165 &

166 of 302 This section discusses Instrument/Thermocouple and Control and LV Power cables. Section 19.4 for Medium Voltage cables references the BFN License Renewal commitment as defined in FSAR App. O, Section O.1.3.

Explain whether this document (G-38), along with FSAR App O, Sections O.1.2 (for I&C cables) and other sections for Low Voltage cables are going to be updated with license renewal commitments, similar to the Medium Voltage section O.1.3.

SLRA Section/TLAA/AMP/Scoping and Screening: Fuse Holders Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

B.2.1.42 B-205 Browns Ferry SLRA AMP B.2.1.42, Fuse Holders, states: The Fuse Holders aging management program will be consistent with the 10 elements of aging management program XI.E5, Fuse Holders, specified in NUREG-2191.

The staff noted that the program description of SLRA AMP B.2.1.42 and the acceptance criteria element in the Browns Ferry fuse holders AMP basis document in the portal appear inconsistent with those of NUREG-2191 (GALL-SLR) AMP XI.E5.

The program description of SLRA AMP B.2.1.42 does not include the following statement in GALL-SLR: Fuse holders are visually inspected for electrical insulation surface anomalies indicating signs of reduced insulation resistance due to thermal/thermoxidative degradation of organics, radiolysis and photolysis [ultraviolet (UV) sensitive materials only] of organics, radiation-induced oxidation, and moisture intrusion as indicated by signs of embrittlement, discoloration, cracking, melting, swelling, or surface contamination.

a-Explain how the acceptance criteria element for Browns Ferry AMP for fuse holders is consistent with the corresponding element of GALL-SLR AMP XI.E5 considering the identified missing statements. If there is no consistency, provide justification for the difference between the Browns Ferry AMP for fuse holders and the NUREG-2191 AMP XI.E5 and revise the SLRA AMP B. 2.1.42 accordingly.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The acceptance criteria element of Browns Ferry AMP for fuse holders in the basis document does not include the following statements in GALL-SLR:

When thermography is used, the metallic clamp of the fuse holder needs to be below the maximum allowed temperature for the application Test acceptance criteria show that fuse holders are free from the unacceptable aging effects of increased resistance of connection due to chemical contamination, corrosion, and oxidation or fatigue caused by ohmic heating, thermal cycling, electrical transients, frequent removal and replacement, or vibration. Visual inspection acceptance criteria show that fuse holders are free from unacceptable electrical insulation surface anomalies indicating signs of reduced insulation resistance due to thermal/thermoxidative degradation of organics, radiolysis and photolysis (UV sensitive materials only) of organics; radiation-induced oxidation, and moisture intrusion as indicated by signs of embrittlement, discoloration, cracking, melting, swelling, or surface contamination.

Browns Ferry Thermography Program document 0-TI-230T Appendix C states that a satisfactory/advisory result is a temperature rise of less than 15º F above established reference temperatures.

b-Clarify if the satisfactory/ advisory temperature results in 0-TI-230T Appendix C are less than the maximum allowed temperature for the applications.

SLRA SectionB.3.1.1, Fatigue Monitoring Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

B.3.1.1 B-210 SLRA Section B.3.1.1 addresses Enhancement 1 regarding the scope of program program element of the Fatigue Monitoring Aging Management Program (AMP).

Enhancement 1 is to revise implementing procedures to require monitoring of the refueling containment skirt within the scope of the program.

1. Discuss the background and basis for the inclusion of the refueling containment skirt in the scope of the program for the subsequent period of

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions However, it is not clear to the staff why the applicant needs to include the refueling containment skirt in the scope of the program for the subsequent period of extended operation (i.e., 60 to 80 years of operation).

extended operation (e.g., a new analysis of the component or consideration of an increased effect of fatigue).

2 B.3.1.1 B-211 SLRA Section B.3.1.1 addresses Enhancement 2 regarding the scope of program program element of the Fatigue Monitoring AMP.

Enhancement 2 is to revise implementing procedures to require component locations that are in the scope of the program to be revised based on operating experience, plant modifications, and inspection findings.

However, the SLRA does not clearly discuss whether this enhancement is related to a generic enhancement to include the evaluation of operating experience, plant modifications and inspection findings for identifying potentially new fatigue locations for aging management. If this enhancement is intended to address specific previous operating experience, plant modifications or inspection findings, the staff needs to confirm that the planned revision to implementing procedures are adequate to manage the aging effects of fatigue based on the specific operating experience, degradation or components.

1. Clarify whether Enhancement 2 is a generic enhancement to include the evaluation of operating experience, plant modifications and inspection findings to identify potentially new fatigue locations for aging management. If this enhancement is intended to address specific previous operating experience, plant modifications or inspection findings, describe the planned revision to implementing procedures and its technical basis based on the specific operating experience, degradation or components.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 3

B.3.1.1 B-211 SLRA Section B.3.1.1 addresses Enhancement 3 regarding the parameters monitored or inspected program element of the Fatigue Monitoring AMP.

Enhancement 3 is to revise implementing procedures to ensure periodic review of chemistry parameters that give inputs to Fen (environmental fatigue corrections factor) used in CUFen (environmentally adjusted cumulative usage factor) calculations for environmentally-assisted fatigue calculations.

However, the enhancement does not clearly discuss whether the chemistry parameters used in the Fen calculations rely on the water chemistry monitoring of the Water Chemistry AMP (SLRA Section B.2.1.2) and the associated data.

1. Clarify whether the chemistry parameters used in the Fen calculations rely on the water chemistry monitoring of the Water Chemistry AMP (SLRA Section B.2.1.2) and the associated data. If not, provide justification for why the Water Chemistry AMP is not used to provide the water chemistry data for the calculations of Fen values.

4 B.3.1.1 B-211 SLRA Section B.3.1.1 addresses Enhancement 4 regarding the parameters Monitored or inspected program element of the Fatigue Monitoring AMP Enhancement 4 states that analysis has been completed to re-evaluate the cumulative fatigue limit for the recirculation inlet nozzle safe ends, and the limits will be revised in FatiguePro [fatigue monitoring software]

prior to entry into the subsequent period of extended operation.

However, the following items are not clear to the staff: (1) the meaning of the completed cumulative usage limit for the recirculation inlet nozzle safe ends in their fatigue reevaluation; and (2) the meaning of the limits that will be revised in the fatigue monitoring software (FatiguePro software).

1. Clarify the following items described in Enhancement 4: (1) the meaning of the completed cumulative usage limit for the recirculation inlet nozzle safe ends; and (2) the meaning of the limits that will be revised in the fatigue monitoring software (e.g., e.g.,

limits on relevant transient cycles used in the CUFen calculations).

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions

2. Clarify whether the 80-year projected CUFen values for the recirculation inlet nozzle safe ends have been already calculated and described in SLRA Table 4.3.1-4. If not, (1) revise the enhancement accordingly to clarify that the enhancement will perform a revaluation of CUFen values for the recirculation inlet nozzle safe ends and (2) provide the technical basis for why the future reevaluation will have reasonable assurance that the CUFen values, as will be revised, do not exceed the design limit (1.0).

5 B.3.1.1 B-211 SLRA Section B.3.1.1 addresses Enhancement 5 regarding the parameters monitored or inspected program element of the Fatigue Monitoring AMP Enhancement 5 states that FatiguePro [fatigue monitoring software],

Version 4, will be implemented prior to entry into the subsequent period of extended operation.

1. Clarify whether the applicant is already using FatiguePro Revision 4. If so, discuss why this enhancement is needed even though

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions In comparison, the following reference indicated that the applicants fatigue monitoring is already using FatiguePro, Revision 4 (

Reference:

Structural Integrity Associates (SIA) Report 1801147.301, Revision 2, BFN 80 Year Cycle Projection Update for SLR, Section 4.0 and Figure 1).

The staff needs clarification on whether the applicant is already using FatiguePro Revision 4. If so, the staff needs clarification on why Enhancement 5 is needed. If not, the staff clarification on the differences between the currently used version of FatiguePro and Revision 4.

The staff also needs clarification as to whether Revision 4 of FatiguePro relies on the guidance in NUREG/CR-6909, Revision 1 in the calculation of CUFen values.

FatiguePro Revision 4 is already used.

2. If the applicant is current using FatiguePro, Revision 3 or earlier version, clarify the following:

(1) the currently used version of FatiguePro; and (2) the difference between the currently used version of FatiguePro and Revision 4 that will be used for the subsequent period of extended operation. As part of the discussion, describe the improved areas of FatiguePro Revision 4 compared to the currently used version.

3. Clarify whether Revision 4 of FatiguePro relies on the guidance in NUREG/CR-6909, Revision 1 in the calculation of CUFen values. If not,

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions resolve the potential inconsistency between Enhancement 5 and SLRA Section 4.3.5 that indicates the use of guidance in NUREG/CR-6909, Revision 1 for environmentally assisted fatigue (EAF) analysis.

6 B.3.1.1 B-211 SLRA Section B.3.1.1 addresses Enhancement 7 regarding the acceptance criteria program element of the Fatigue Monitoring AMP Enhancement 7 is to revise implementing procedures to reflect the re-evaluated cumulative fatigue values for the Units 1, 2, and 3 recirculation inlet nozzle safe ends.

However, it is not clear to the staff which aspect of the acceptance criteria program element will be enhanced in related to the reevaluated cumulative fatigue values for the recirculation inlet nozzle safe ends in Enhancement 7.

The staff also clarification on whether this enhancement is applied to both cumulative usage factor (CUF) and environmentally adjusted CUF (CUFen) for the recirculation inlet nozzle safe ends.

1. Clarify which aspect of the acceptance criteria program element will be enhanced in related to the reevaluated cumulative fatigue values for the recirculation inlet nozzle safe ends in Enhancement 7 (e.g.,

limits on relevant transient cycles used in the CUFen calculations).

2. Clarify whether this enhancement is applied to both cumulative usage factor (CUF) and environmentally adjusted CUF (CUFen)

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions for the recirculation inlet nozzle safe ends. If not, provide justification for why the enhancement is not applied to both CUF and CUFen for the components.

SLRA Section3.5.2.2, AMR Results for Which Further Evaluation is Recommended by the GALL-SLR Report Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

FE 3.5.2.2.1.2 FSAR 5.2.3.2 3-664 5.2-3

1. SLRA FE Section 3.5.2.2.1.2 states that each drywell is cooled during normal operation of the unit by a closed loop ventilation system.

However, FSAR Section 5.2.3.2 states, in part, that this temperature is maintained by recirculating the drywell atmosphere across forced draft air cooling units which, in turn, are cooled by the Reactor Building Closed Cooling Water System.

SLRA Section 3.3.2.1.23 describes the reactor building closed cooling water system.

SLRA FE Section 3.5.2.2.1.2 does not make clear what system is used to cool drywell during normal operation.

2. SLRA FE Section 3.5.2.2.1.2 also states that insulation and air gaps are provided to reduce thermal stress regarding maintenance of local temperature limit (200°F).

SLRA does not make clear which line items are used for the insulation mentioned above and what AMP manages its aging effect.

3. SLRA FE Section 3.5.2.2.1.2 states that the evaluation determined that the localized temperature rise in the shield wall resulted in a localized concrete temperature of less than 180oF.

1. Describe the closed loop ventilation system and clarify what AMP manages aging effects of this system.

2. Clarify Tables 1 and 2 AMR items used for the insulation, and what AMP manages its aging effect.

3. Discuss the evaluation mentioned in SLRA FE Section 3.5.2.2.1.2.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 3

FE 3.5.2.2.1.3 Table 3.5.1 3-679 3-716 AMR 3.5-1, 044 in SLRA Table 3.5.1 claims to be consistent with NUREG-2191, but it states that that cracks, distortion, increase in component stress level due to settlement are not aging effects requiring management for structures founded on rock or bearing piles.

SLRA Table 3.5.2-1 shows Table 2 item associated with AMR item 3.5-1, 044, citing Note I with plant-specific notes 1.

If the combination of component, material and environment exist, this AMR item is applicable and the component requires aging management.

AMR 3.5-1, 046 in SLRA Table 3.5.1 claims to be not applicable, but it also states that the aging effect pertaining to settlement and foundations/subfoundations was captured visa item number 3.5-1.

044.

It appears that AMR 3.5-1, 046 is not used, and their aging management is addressed under AMR 3.5-1, 044.

SLRA does not make clear what is not applicable or not used.

1. Discuss whether the cracks, distortion, increase in component stress level due to settlement are applicable aging effects requiring management for structures founded on rock or bearing piles.

2. Discuss the applicability of AMR 3.5-1, 046.

4 FE 3.5.2.2.2.1.4 FE 3.5.2.2.2.3.3 3-680 3-689 GALL-SRP states that a plant-specific AMP is not required for the reinforced concrete exposed to flowing water if (1) there is evidence in the accessible areas that the flowing water has not caused leaching of calcium hydroxide and carbonation or (2) evaluation determined that the observed leaching of calcium hydroxide and carbonation in accessible areas has no impact on the intended function of the concrete structure.

FE Section 3.5.2.2.2.1, item 4 states that it was determined that this aging mechanism is not applicable to BFN Group 4 structures (which include the biological shield wall, RPV Pedestal, and RPV Support Skirt), as BFN has a Mark 1 steel containment (not a concrete containment). It appears that BFN containment internal structure includes concrete components.

1. Clarify what constitutes the Group 4 structures and discuss the applicability of aging mechanism for BFN Group 4 structures.

2. Discuss OEs associated with Increase in porosity and permeability; loss of strength due to leaching of calcium

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions It states, The effects of carbonation have not been observed at BFN, and Operating experience at BFN, which looks for concrete deterioration due to any mechanism, has not identified porosity and permeability and loss of strength due to these mechanisms, and Concrete degradation due to chemistry attack or leaching has not been observed at BFN. These statements may not match OE at BFN site.

SLRA does not make clear whether the observed leaching of calcium hydroxide and carbonation in accessible areas has impact on the intended function of the concrete structure.

SLRA Table 3.5.2-1 lists Table 2 items associated with AMR item 3.5-1, 047, cite Note G and 4. Note 4 states, The reactor building is exposed to condensation, groundwater, treated water. However, air-indoor uncontrolled and outdoor environments are also listed in the SLRA Table 3.5.2-1.

SLRA does not make clear whether there are other applicable aging effects when concrete components are exposed to condensation and treated water.

hydroxide and carbonation and revise their conclusions if necessary.

3. Provide an evaluation whether the observed leaching of calcium hydroxide and carbonation in accessible areas has impact on the intended function of the concrete structure.

4. Explain how aging effects of inaccessible concrete areas are adequately managed by the Structures Monitoring program.

5. Address the inconsistency of environments described in the plant-specific note 4 for Table 2 items associated with AMR item 3.5-1, 047.

6. Evaluate whether there are other applicable aging

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions effects for concrete components exposed to condensation and treated water environment.

5 FE 3.5.2.2.2.3 3-684 FE Section 3.5.2.2.2.2.3, items 1, 2, and 3, state that the CCW conduits are completely inaccessible during normal operations, the conduits are inspected from the inside during outages.

SLRA does not make clear of interval between plant outages and whether the CCW conduits are monitored on an interval not to exceed 5 years.

1. Clarify the interval between plant outages.

2. Clarify the inspection interval for monitoring the CCW conduits.

6 FE 3.5.2.2.2.3 3-687 FE Section 3.5.2.2.2.3, item 2 states that cracking associated with expansion due to reaction with aggregates has not been observed on BFN Group 6 concrete structures, as described in operating experience. It also states that significant concrete deformation/loss of material due to concrete expansion has not been detected in Group 6 structures, which have comparatively high exposure to water compared to other reinforced concrete structures at BFN, which conflict with the statement above.

FE Section 3.5.2.2.2.3, item 2 does not make clear how the cracking due to expansion and reaction w/aggregates in inaccessible areas are adequately managed by the Structures Monitoring program.

1. Clarify whether cracking associated with expansion due to reaction with aggregates are present at BFN site.

2. Explain how aging effects in inaccessible areas are adequately managed by the Structures Monitoring program. This question also applies to other further evaluation sections.

SLRA SectionTRP: 82 Recurring Internal Corrosion Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

3.3.2.2.7 3-318 The GALL 2192 under the further evaluation section 3.3.2.2.7 states that there are five questions that should be addressed. The applicant Please define the basis for the adequacy of

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions states: (a) why the program's examination methods will be sufficient to detect the recurring aging effect before affecting the ability of a component to perform its intended function, (b) the basis for the adequacy of augmented or lack of augmented inspections, (c) what parameters will be trended as well as the decision points where increased inspections would be implemented (e.g., the extent of degradation at individual corrosion sites, the rate of degradation change), (d) how inspections of components that are not easily accessed (i.e., buried, underground) will be conducted, and (e) how leaks in any involved buried or underground components will be identified.

This section does not currently identify all aspects which are needed for a complete review.

augmented or lack of augmented inspections.

Please define what parameters will be trended.

Please expand on how leaks in any involved buried or underground components will be identified. The application section 3.3.2.2.7 discusses opportunistic inspection of buried or underground piping but does not identify leak process.

Further Evaluations for Pitting Corrosion, Cre SLRA Sections 3.2, 3.3, 3.4 Stress Corrosion Cracking and Loss of Material (pitting, crevice) for Stainless Steel, Nickel Alloys, and Aluminum Alloys Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

3.3.2.2.3 3-312 SLRA Section 3.3.2.2.3, Item 3.3-1, 004, states that cracking has been identified as an aging effect, possibly caused by stress corrosion cracking (SCC) in the presence of moisture and halides in the Diesel Generator Starting Air System. It also states that based on the conservative assumption of SCC due to halides, BFN will implement the External Surfaces Monitoring of Mechanical Components program to demonstrate cracking is not occurring in piping, piping components, and tanks exposed to air-indoor uncontrolled, air-outdoor, and condensation for the Diesel Generator Starting Air System. It does not appear that periodic inspection was specified for other systems with Please discuss how the potential SCC of stainless steel in the Diesel Generator Starting Air System was considered in determining the need for aging management for SCC and LOM due to pitting and crevice corrosion for stainless

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions these environments based on this operating experience or for loss of material (LOM) due to pitting or crevice corrosion for any system.

In the SLR-SRP (and reproduced in the SLRA), the further evaluation sections for loss of material and SCC of stainless steel, nickel alloys, and aluminum alloys (e.g., Section 3.3.2.2.3) state that One-Time Inspection is an acceptable program to demonstrate that the aging effect is not occurring. Several acceptable programs with periodic inspections are listed as acceptable options if the aging effect has occurred.

Although it was assumed that contaminants can be introduced into the emergency diesel generator building and cause SCC of stainless steel, it is not clear how this operating experience was applied to other systems for SCC, or to LOM due to pitting or crevice corrosion for the Diesel Generator Starting Air System or other systems.

steel, nickel alloys, and aluminum alloys in uncontrolled indoor air, outdoor

air, and condensation.

The corresponding further evaluation sections are 3.2.2.2.2, 3.2.2.2.4, 3.2.2.2.8, 3.2.2.2.10, 3.3.2.2.3, 3.3.2.2.4, 3.3.2.2.8, 3.3.2.2.10, 3.4.2.2.2, 3.4.2.2.3, 3.4.2.2.7, and 3.4.2.2.9.

2 3.3.2.2.3 3.3.2.2.4 Table 3.3.1 Table 3.3.2-19 3-317 3-381 3-467 Based on the typical design of Standby Liquid Control (SLC) systems, the Discussion in Table 3.3.1 for Item 3.3-1, 232, the discussion of item 3.3-1, 232 in further evaluation (FE) Section 3.3.2.2.4, and the use of item 3.3-1, 232 in Table 3.3.2-19, the staff infers that the insulated SLC tank is a stainless steel tank managed for loss of material (LOM) in air and condensation due to pitting and crevice corrosion using the One-Time Inspection program. This aging management approach is consistent with the guidance in FE Section 3.3.2.2.4 given that this aging effect has not occurred.

One-Time inspection would also be consistent with the guidance in FE Section 3.3.2.2.3 for cracking due to SCC of the SLC tank in air and condensation; however, there does not appear to be an AMR item in Table 3.3.2-19 for managing cracking due to SCC for the SLC tank (e.g., item 3.3-1, Please address the following:

a.

Provide a description of the SLC tank and how aging management plans meet the guidance in FE Sections 3.3.2.2.3 and 3.3.2.2.4 for SCC and LOM due to pitting and crevice corrosion in air and condensation environments.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions If the intent is to manage aging of the SLC system tank for LOM but not SCC, please discuss the extent to which this approach is being taken for other stainless steel, nickel

alloy, and aluminum alloy components subject to further evaluation for LOM and SCC in SLRA Sections 3.2, 3.3, and 3.4.

3 3.3.2.2.4 Table 3.3.1 3-315 3-316 3-330 For item 3.3-1, 006, FE Section 3.3.2.2.4 lists several systems for which operating experience identified LOM in stainless steel components. This item pertains to environments of air or condensation.

In Section 3.3.2.2.4 and in the Discussion for item 3.3-1, 006 in Table 3.3.1, periodic inspections are proposed for components in these systems, but One-Time Inspection is proposed for stainless steel components exposed to these environments in the other auxiliary systems.

It is not clear to the staff why the use of periodic inspections, in response to the operating experience, appears to be system-based rather than materials/environment-based. The guidance in FE Section 3.3.2.2.4 (and other similar FE sections) is based on material and environment combinations that could result in the aging effect.

Please describe how the system-based approach to specifying periodic or one-time inspections considers all of the stainless steel components potentially exposed to the environmental conditions that resulted in the LOM observed.

4 3.3.2.2.8 3.3.2.2.10 Table 3.3.1 3-322 3-326 3-387 AMR item 3.3-1, 254 addresses SCC of aluminum heat exchanger components exposed to air or condensation. Table 3.3.1 says this item is not applicable, but FE Section 3.3.2.2.8 states that Core Spray and RHR Room Cooler fins are aligned to this item. The staff did not find this item assigned to these components or any others.

Please address the following:

a.

Clarify how item 3.3-1, 254 is being applied for aging

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions In addition, it is unclear why there are aluminum heat exchanger components managed for pitting and crevice corrosion in condensation using the One-Time Inspection program (i.e., FE Section 3.3.2.2.10, Item 3.3-1, 242) but not for SCC.

management and where that appears in the SLRA.

b.

Discuss why aluminum heat exchanger components are assumed to be susceptible to pitting and crevice corrosion in condensation (item 3.3-1, 242) but not susceptible to SCC (item 3.3-1, 254).

5 3.4.2.2.2 3-543 3-557 3-597 3-603 3-615 3-623 Section 3.4.2.2.2 contains the following statements about cracking due to SCC that appear to be contradictory:

Cracking has not been identified as an aging effect at BFN for stainless steel screened-in components in these environments, and Cracking has been identified as an aging effect at BFN for stainless steel as a result of transportable halogens, indicating that the environments do contain sufficient halides (e.g., chlorides) in the present of moisture to result in SCC.

In addition, the Condensate Head Drain Tank is identified under item 3.4-1, 002 in this section as a stainless steel tank, but it is not clear if it is included in the aging management tables using item 3.4-1, 002 or another AMR item.

Please address the following:

a. Clarify the apparent contradiction in the statements in SLRA Section 3.4.2.2.2 that, (a) cracking has not been identified as an aging effect at BFN for stainless steel screened-in components in these environments, and, (b) cracking has been identified as an aging effect at BFN for stainless steel as a

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions There is also a discrepancy associated with item 3.4-1, 002 because the Discussion in Table 3.4.1 states that the External Surfaces Monitoring of Mechanical Components program is used for this item, but the Table 3.4.2-X entries for 3.4-1, 002 list One-Time Inspection as the AMP.

result of transportable halogens.

b. Discuss whether the condensate head drain tank is included in the one-time inspections for piping and piping components managed by item 3.4.1-002 as described in SLRA Section 3.4.2.2.2. If it is not, please clarify the discussion in Section 3.2.2.2.2 and aging management for the tank.

c. Clarify which AMPs are associated with item 3.4-1, 002, given the discrepancy between the Table 3.4.1 discussion (External Surfaces Monitoring program) and the Table 2 AMR items (One-Time Inspection program).

Identify and justify any inconsistencies with the SLR guidance.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 6

3.4.2.2.2 3-543 3-544 The discussion of item 3.4-1, 002 in FE Section 3.4.2.2.2, summarizes a sample probe failure associated with CR1858867 dated May 2023.

The probes are identified as stainless steel that failed due to SCC in a condensate environment upstream of the condensate booster pump.

Because this FE section and AMR item are associated with air or condensation environments, but the probe failure appears to have occurred in condensate, the appropriate aging management response for this operating experience is unclear to the staff.

Please address the following:

a. Clarify the environment in which the sample probe cracking occurred.

b. If the cracking occurred in a treated water environment, explain why it is discussed in FE Section 3.4.2.2.2 and associated with item 3.4-1, 002, and whether there are other AMR items for managing SCC of these probes in treated water.

c. It appears that the decision not to require periodic inspection of the new probes is based in part on observing no failures with the new probe design. Discuss the length of operating time supporting this decision.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions

d. Describe any evaluations and corrective actions related to the Water Chemistry program that were performed in response to the failure.

7 3.4.2.2.7 Table 3.4.1 Table 3.5.2-36 3-550, 3-583, 3-949 The discussions in SLRA Table 3.4.1 and Subsection 3.4.2.2.7 state that AMR item 3.4-1, 102 is not applicable because there are no aluminum tanks (within the scope of AMP XI.M29, Outdoor and Large Atmospheric Metallic Storage Tanks) exposed to air, condensation, soil, concrete raw water, waste water within the screened-in portions of the Steam and Power Conversion Systems.

Although there are no aluminum tanks in these exposure conditions in the steam and power conversion systems, AMR item 3.4-1, 102 is cited in SLRA Table 3.5.2-36, Structural Commodities (Penetrations and Sleeves) for aluminum electrical penetrations and instrumentation and controls penetrations exposed to concrete.

Because item 3.4-1, 102 is cited for components, the statement in SLRA Table 3.4.1 that this item is not applicable appears to be incorrect.

In addition, it is not clear to the staff whether the One-Time Inspection program is appropriate for the aluminum penetrations exposed to concrete. This uncertainty is based on two cases of through-wall leaks in aluminum penetrations exposed to concrete (CR1695960 and CR1717380).

Please address the following:

a. Clarify the applicability and use of item 3.4-1, 102.

b. Discuss any changes needed in Table 3.4.1, Section 3.4.2.2.7, or other parts of the SLRA to correctly describe the applicability of item 3.4-1, 102.

c. Discuss the relevance of the operating experience with leaking aluminum penetrations (CR1695960 and CR1717380) to the penetrations that would be managed by item 3.4-1, 102.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA Section: N/A TRP: 085 No Aging Effects Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

3.2 3-193, 3-216, 3-750, 3-950 The Discussion in SLRA Table 3.2.1 for AMR item 3.2-1, 091 states, Not applicable. See Subsection 3.2.2.2.9.

SLRA Section 3.2.2.2.9 states AMR item 3.2-1, 091 is not applicable because there are no stainless steel piping or piping components exposed to concrete in the ESF and RCIC systems.

While there are no stainless steel piping or piping components exposed to concrete in the ESF and RCIC systems, AMR item 3.2-1, 091 is cited in SLRA Table 3.5.2-1 for the stainless steel fuel pool liner and SLRA Table 3.5.2-36 for stainless steel mechanical penetrations. Therefore, the statement in SLRA Table 3.2.1 indicating it is not applicable appears to be incorrect since it is cited for components.

Please discuss why SLRA Table 3.2.1 states AMR item 3.2-1, 091 is not applicable when it is cited in SLRA Table 3.5.2-1 for the stainless steel fuel pool liner and in SLRA Table 3.5.2-36 for stainless steel mechanical penetrations.

In addition, please discuss why the stainless steel fuel pool liner and mechanical penetrations exposed to concrete are not addressed in SLRA Section 3.2.2.2.9.

2 3.2 3-193, 3-208, 3-209, 3-229, 3-234, 3-833, 3-898, 3-909 SLRA Section 3.2.2.2.9 states that the carbon steel piping and piping components exposed to concrete in the Standby Gas Treatment System and Containment System are potentially exposed to groundwater and that AMR item 3.2-1, 052 is used for managing loss of material of these components by the Buried and Underground Piping and Tanks program.

However, the Discussion of AMR item 3.2-1, 055 in SLRA Table 3.2.1 reiterates the use of AMR item 3.2-1, 052, but also states, Consistent with NUREG-2191, there is no Aging Management Program required for the steel piping components exposed to concrete in the Containment System. Therefore, SLRA Section 3.2.2.2.9 and SLRA Table 3.2.1 appear to be conflicting.

Please discuss the following:

• Conflicting statements regarding steel piping, piping components in the Containment System.

• Where AMR item 3.2-1, 055 is cited in SLRA Tables 3.2.2-1, 3.5.2-15, 3.5.2-33, and 3.5.2-36, confirm that the exclusion criteria in Further

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA Tables 3.2.2-1, 3.5.2-15, 3.5.2-33, and 3.5.2-36 cite AMR item 3.2-1, 055 for steel components exposed to concrete.

The Discussion of AMR item 3.2-1, 052 in SLRA Table 3.2.1 states, The Buried and Underground Piping and Tanks program (B.2.1.27) will be used to manage loss of material of the carbon steel piping, piping components exposed to concrete and soil in the Standby Gas Treatment System and Containment System.

SLRA Table 3.2.2-1 cites both AMR items 3.2-1, 052 (soil, concrete) and 3.2-1, 055 (concrete) for steel piping and piping components. It is unclear why both AMR items are cited.

In SLRA Table 3.3.2-2, AMR item 3.2-1, 052 is cited for managing loss of material of steel piping and piping components exposed to soil in the Standby Gas Treatment System. Concrete is not identified as an applicable Environment for the Standby Gas Treatment System in either SLRA Section 3.2.2.1.2 or SLRA Table 3.3.2-2. This appears to be inconsistent with statements in SLRA Section 3.2.2.2.9 and SLRA Table 3.2.1.

In addition, AMR item 3.2-1, 055 is cited for steel components exposed to concrete in SLRA Tables 3.5.2-15, 3.5.2-33, and 3.5.2-36. However, the Discussion for AMR item 3.2-1, 055 in SLRA Table 3.2.1 does not state the item is used for components in 3.5.

Evaluation Section 3.3.2.2.9 in NUREG-2192 have been met and discuss why the steel components exposed to concrete in these tables were not addressed in SLRA Section 3.3.2.2.9.

• Why both AMR items 3.2-1, 055 and 3.2-1, 052 are cited for steel piping, piping components exposed to concrete in SLRA Table 3.2.2-1.

• Whether steel piping, piping components in the Standby Gas Treatment System are exposed to concrete.

• Why SLRA Table 3.2.1 does not explain that AMR item 3.2-1, 055 is used for components in 3.5.

3 3.3 3-324, 3-355, 3-365, 3-375, SLRA Section 3.3.2.2.9 states that AMR items 3.3-1, 107 and 3.3-1, 144 will address loss of material and cracking, respectively, of stainless steel piping, piping components exposed to concrete in Auxiliary Systems.

However, SLRA Table 3.3.1 states both AMR items 3.3-1, 107 and 3.3-1, 144 are not applicable and no Auxiliary System aging management evaluation table cites these AMR items.

Please clarify how stainless steel piping, piping components exposed to concrete in the Auxiliary Systems are addressed in the SLRA.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 4

3.3 3-323, 3-357 SLRA Section 3.3.2.2.9 states, A review of OE for BFN indicates there are occurrences of concrete degradation, in some systems, that could lead to the penetration of water to the metal surface; therefore, a loss of material due to general, pitting, and crevice corrosion of steel piping and tanks exposed to concrete is an aging effect that requires management. However, the discussion of AMR item 3.3-1, 112 in SLRA Table 3.2.1 states no condition reports were identified. These statements appear to be conflicting.

AMR item 3.3-1, 112 is cited in SLRA Tables 3.3.2-4, 3.3.2-7, 3.3-1, 112, 3.3.2-26, 3.3.2-27, for steel piping, piping components, and tanks exposed to concrete.

SLRA Section 3.3.2.2.9 states loss of material is addressed by AMR item 3.3-1, 109 for steel piping, piping components exposed to concrete and potentially exposed to groundwater.

The Discussion of AMR item 3.3-1, 112 in SLRA Table 3.2.1 appears to state that loss of material of steel piping, piping components exposed to concrete in the Residual Heat Removal Service Water will be managed by the Buried and Underground Piping and Tanks. However, AMR item 3.3-1, 109 is only cited for steel piping and piping components exposed to soil.

Please address the following:

• Clarify whether there has been operating experience at BFN where steel piping, piping components exposed to concrete have been exposed to groundwater.

• Where AMR item 3.3-1, 112 is cited in SLRA Tables 3.3.2-4, 3.3.2-7, 3.3-1, 112, 3.3.2-26, 3.3.2-27, confirm that the exclusion criteria in Further Evaluation Section 3.3.2.2.9 in NUREG-2192 have been met and discuss why the steel components exposed to concrete in these tables were not addressed in SLRA Section 3.3.2.2.9.

• Are there steel piping, piping components exposed to concrete in the Residual Heat Removal Service Water that is potentially exposed to groundwater?

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 5

3.2 3-269 AMR item 3.2-1, 063 in Vol. 1 of NUREG-2191 cites no aging effects or aging management program for stainless steel piping, piping components exposed to gas.

SLRA Table 3.2.2-7 cites AMR item 3.2-1, 063 for stainless steel tanks exposed to gas with generic note 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.

Given that the component in SLRA Table 3.2.2-7 is different than the component in AMR item 3.2-1, 063, discuss the use of generic note A.

SLRA Table 3.5.2-23, Condensate Water Storage Tanks and Foundations, Trenches, and Tunnels TRP: 091, Non-GALL AMR - Structural Components Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

Table 3.5.2-23 2.4.23 3-867 2-198 SLRA Table 3.5.2-33 lists rock and sand earthfill component exposed to air-outdoor, groundwater/soil, water-flowing, and water-standing environment, citing notes J and 1.

Note J states that neither the component nor the material and environment combination is evaluated in NUREG-2191.

Plant-specific note 1 states that the earthen materials of the Condensate Water Storage Tanks foundation interior base are protected from aging effects by the concrete perimeter ring and Condensate Water Storage Tank bottom.

SLRA Section 2.4.23 states that each of the three in-scope Condensate Water Storage Tanks is supported on a foundation consisting of a concrete ring, under the perimeter of the tank bottom, which surrounds a bed of compacted sand, which does not include rock material as described in SLRA Table 3.5.2-23.

If the combination of component, material and environment exist, this AMR item is applicable and the component requires aging management.

1. Discuss the construction of the earthfill and concrete ring.

2. Clarify whether the earthfill component consists of rock.

3. Evaluate and clarify the aging effects for the earthfill.

4. Clarify what AMP will be used and explain how the aging effects can be adequately managed by this AMP.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions A bed of compacted sand underneath the in-scope Condensate Water Storage Tanks is inaccessible, but it requires aging management.

SLRA Section TLAA: 4.2.10 - Jet Pump Slip Joint Repair Clamp TRP: 142.10 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.2.10 SLRA Section 4.2.10 states The structural evaluation states that the cold bolt preload is 550 pounds at 100ºF, the initial hot preload at 550ºF is 500 pounds, and the minimum, end-of life preload is 350 pounds at 550ºF. Therefore, 150 pounds is attributed for loss of preload during the life of the component, associated with a neutron fluence exposure of 3.02E+18 n/cm2 It appears the original analysis determined the minimum preload needed to maintain the intended function of the component; thus, its not clear how the neutron fluence exposure was incorporated into the original analysis.

Did the original analysis determine the minimum preload needed at 550F is 350-lbs to maintain the component intended function?

If so, how was neutron fluence incorporated/assessed in the original analysis?

How much loss of preload was determined in the original analysis?

Confirm that the SLRA and SIA CALC -

2200107.406P - the analysis is demonstrating at the end of 80 years -

the remaining preload is greater than the minimum preload of 350-lbs determined in the original analysis.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Is the %loss preload for 80 years determined in SIA CALC -

2200107.406P for the cold bolt preload, the initial hot preload or some other configuration.

SLRA Section TLAA: 4.2.11 - Jet Pump Aux Spring Wedge Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.2.11 SIA CALC - 2200107.406 provides the applicants assessment of the Jet Pump Aux Spring Wedge.

Tables 2, 3 and 4 indicate a neutron fluence exposure for the Jet Pump Aux Spring Wedge at Units 1, 2 and 3, respectively.

Table 6 in Appendix B provides a Max BFN fluence for the Jet Pump Aux Spring Wedge that does not correspond to the neutron fluences in Tables 2, 3 or 4.

Tables 3 and 4 - For the Jet Pump Aux Spring Wedge, the neutron fluence is the same - but %relaxation amount is different between Units 2 and 3 Clarify the discrepancy in neutron fluences between Tables 2, 3, and 4, and the neutron fluence in Table 6 of Appendix B.

Clarify the discrepancy in neutron fluences and

%relaxation in between Table 3 and Table 4 in SIA CALC -

2200107.406P.

2 4.2.11 Equation 2 in SIA CALC - 2200107.406P - refers to Figure 3 in ML12321A319.

It appears that Figure 3-2 in ML12321A319 is the relevant figure for Irradiation Creep of Alloy X-750.

Figure 3-2 in ML12321A319 - provides several equations based on the data set being fitted that have a similar functional form as Equation 2 in SIA CALC - 2200107.406P.

Discuss the development/basis for Equation 2 in SIA CALC

- 2200107.406P.

Discuss if/how Figure 3-2 in ML12321A319 was used in the development

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions of Equation 2 in SIA CALC - 2200107.406P.

SLRA Section TLAA: 4.2.12 - Jet Pump Riser Repair Clamp Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.2.12 SIA CALC - 2200107.406P provides the %relaxation for the Jet Pump Riser Repair Clamp The calculated %relaxation in Table 4 of SIA CALC - 2200107.406P is notable Based on SLRA Section 4.2.12 and results in SIA CALC -

2200107.406P - the components appear that they can tolerate additional loss of preload than that determined SIA CALC -

2200107.406P - especially the "support bolt" given the neutron fluence at the end of 80 years is less than that assumed in the original analysis.

Comparison of stress values, and loss of preload from CLB analyses compared with the results in SIA CALC -

2200107.406P for %

relaxation for the jet pump riser repair clamp and support bolts.

2 4.2.12 SLRA Section 4.2.12 Jet Pump Riser Repair Clamp Loss of Preload Analysis states The stress analysis of the clamp analyzed preload losses based on neutron fluence of 1.89E+20 n/cm2 for the clamp bolt and 3.15E+20 n/cm2 for the support bolt.

SLRA Section 4.2.12 - indicates that this TLAA was dispositioned in accordance with 10 CFR 54.21(c)(1)(i) and that the analysis for jet pump riser repair clamp loss of preload remains valid through the subsequent period of extended operation.

During its audit, the staff noted that the neutron fluence values referenced in SLRA 4.2.12 was provided to the applicant via communication with GE. It appears that this communication was via e-mail.

Since this TLAA is being disposition in accordance with 10 CFR 54.21(c)(1)(i) - the staff is seeking applicant documentation that is auditable and retrievable that indicates that the applicants TLAA for Jet Provide applicant documentation that is auditable and retrievable that indicates that the applicants TLAA for Jet Pump Riser Repair Clamp Loss of Preload was based on an assumed values cited in the SLRA.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Pump Riser Repair Clamp Loss of Preload was based on an assumed values cited in the SLRA.

SLRA Section TLAA: 4.2.13 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.2.13 4-56 LRA Section 4.2.13 states BFN Unit 1 has had only a small population of core support plate plugs replaced. The original core support plate plugs had a service life of approximately 14 years. BFN will replace the original core support plate plugs during the outage which corresponds to the end of their service life.

The TLAA analysis only discusses the extended life core support plate plugs, which appears to be applicable only to BFN Units 2, and 3, which have had all the original core support plate plugs replaced, and for the small population of core support plate plugs replaced at BFN Unit 1.

Discuss what programmatic mechanism enforces the replacement from the original core support plate plugs to the extended life core support plate plugs prior to the subsequent period of extended operation, if the SLRA is approved.

Discuss whether the disposition of the TLAA in accordance with 10 CFR 54.21(c)(1)(ii) is applicable/appropriate for the core support plate plugs that have NOT been replaced at BFN, Unit 1.

SLRA Section TLAA: 4.2.14 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.2.14 4-57 The TLAA Description notes that:

Clarify whether the exclusion of the guide tubes in the TLAA

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Section 4.7.6 of the BFN LRA concludes that the expected fluence on the core shroud, top guide, core plate, and in-core instrumentation dry tubes and guide tubes exceed All further discussion in the TLAA refers only to the core shroud, top guide, core plate, and in-core instrumentation dry tubes.

Evaluation and TLAA Disposition was an administrative oversight.

If the omission was intentional, clarify whether the TLAA is applicable to the guide tubes and/or how the analysis of the guide tubes will be dispositioned.

2 A.4.2.14 A-75 Similar to 1 but for the FSAR supplement.

Similar to 1 but for the FSAR supplement.

SLRA TLAA Section 4.2.15 - Core Spray Replacement Piping Bolting Loss of Preload Evaluation Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request

1.

4.2.15 4-58 In Section 4.2.1.2 it states that the 80-year fluence projections will be validated by the BFN Neutron Fluence Monitoring program.

Verify that the BFN Neutron Monitoring Program will be utilized for the Core Spray Replacement Piping Bolting.

2.

4.2.15 4-58 Section 4.2.15 discusses using the RAMA fluence methodology to project the neutron fluence to 62 EFPY as described in Section 4.2.1.2.

Reviewing Table 4.2.1.2-3 for the Unit 3 Reactor Vessel Internal Component Fluence Projections, there is no reference or Fluence identified for the Core Spray Replacement Bolting.

Verify that the Core Spray Replacement Bolting should be included in Table 4.2.1.2-3 since it states that the vessel internal component fluence analyses will be used for the TLAA evaluations in Sections 4.2.8 through 4.2.19.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA Section 4.2.16 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.2.16 4-58 to 4-59 Topic: SLRA Section 4.2.16, Core Spray Sparger Repair Clamp Loss of Preload Evaluation SLRA Section 4.2.16, Core Spray Sparger Repair Clamp Loss of Preload Evaluation, states that the 50 EFPY fluence value for the Unit 1 core spray sparger repair clamp location was determined to be 2.94E+19 n/cm2.

By letter dated February 23, 2023, Structural Integrity Associates (SIA),

Inc forwarded TVA a letter report, titled Stress Relaxation of Core Spray and Jet Pump Component Hardware for Subsequent License Renewal, Report No. 2200107.406P. Table 2 of the SIA letter report shows that at 50 EFPY, the core spray repair clamp at Browns Ferry Unit 1 will receive a neutron fluence of 2.43E+19 n/cm2. The staff notes the discrepancy between the two neutron fluence values.

The staff notes that Table 4 of the SIA letter report shows that the core spray sparger clamp will receive a neutron fluence of 2.94E+19 n/cm2 at Unit 3.

The staff notes that Section 4.2.16 only discusses the core spray sparger repair clamp at Unit 1. It appears that the Unit 3 core spray sparger also has a clamp installed; however, the Unit 3 clamp is not discussed in Section 4.2.16.

a.Discuss why the two neutron fluences are different for the same core spray sparger repair clamp at Unit 1.

b. Discuss why Section 4.2.16 does not discuss the loss of preload evaluation for the core spray sparger repair clamp at Unit 3.

2 4.2.16 4-58 to 4-59 Section 4.2.16 states that In order to determine if this fluence assumption will remain valid through 80 years of operation, the neutron fluence was projected to 50 EFPY using the RAMA fluence methodology previously described in Section 4.2.1.2. The 50 EFPY fluence value for the Unit 1 core spray sparger repair clamp location was determined to be 2.94E+19 n/cm2 The staff notes that SLRA Tables 4.2.1.2-1, 4.2.1.2-2, and 4.2.1.2-3 provide neutron fluence projections for various reactor vessel internal a.Discuss how 2.94E+19 n/cm2 was derived. e.g.,

at what location was this fluence projected.

b. Discuss why Tables 4.2.1.2-1, 4.2.1.2-2, and 4.2.1.2-3 do not show a

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions components, but not the core spray sparger. Also, the staff could not find 2.94E+19 n/cm2 in Tables 4.2.1.2-1, 4.2.1.2-2, and 4.2.1.2-3.

neutron fluence of 2.94E+19 n/cm2.

c. Discuss why these tables do not project neutron fluence for the core spray sparger.

3 4.2.16 4-58 to 4-59 Topic: SLRA Section 4.2.16, Core Spray Sparger Repair Clamp Loss of Preload Evaluation Table 2 of the Structural Integrity Associates, Incs letter report reports the core spray sparger repair clamp will have stress relaxation at 50 EFPY for Unit 1.

Table 4 of the SIAs letter report indicates that the core spray sparger repair clamp will have stress relaxation at 62 EFPY for Unit 3.

SIA Report No. 2200107.406P provides a method for determining the loss of preload of the Core Spray Sparger Repair Clamp as a function of neutron fluence.

The staff noted that if the stress relaxation screening criteria cited in SLRA Section 4.2.16 is applied to the method in SIA Report No.

2200107.406P, the core spray sparger repair clamp would be expected to have greater than 50% loss of preload for both Units 1 and 3.

a.Discuss how the stress relaxation was derived and reconcile the amount of stress relaxation that is predicted in SIA Report No.

2200107.406P with the amount that is allowed by the design of the component.

b.Provide the allowable percentage of stress relaxation for the repair clamp.

4 4.2.16 4-58 to 4-59 Section 4.2.16 states that a repair clamp device was installed during the Unit 1 restart effort.

Discuss what year and month was the clamp installed. If this information is available in a previous submittal, please provide the ADAMS Accession number.

5 4.2.16 4-58 to 4-59 Section 4.2.16 does not provide information regarding the clamp.

a.Discuss the clamp size (length, width and thickness).

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions

b. Discuss the material of the repair clamp. If this information has been submitted to the NRC previously, provide the ADAMS Accession Number.

SLRA Section TLAA: 4.2.17 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.2.17 4-59 The access hole cover repair is scheduled to be installed in Unit 3 in year 2024.

Will the applicant supplement the SLRA once the access hole covers in Unit 3 are replaced? This would allow the staff to issue the SE without equivocation.

SLRA Section TLAA: 4.2.18 - Jet Pump Hold-Down Beam Assembly Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.2.18 SLRA Section 4.2.18 states An evaluation was prepared to determine if the calculated preload loss in the beam bolt was acceptable for the subsequent period of extended operation. The analysis found that the jet pump beam assembly can maintain functionality up to a fluence level of approximately 7.17E+20 n/cm2 (E > 1 MeV) for Group 2 beam bolt assemblies and 6.12E+20 n/cm2 (E > 1 MeV) for Group 3 beam bolt assemblies. The peak fast neutron fluence for the BFN jet pump beam assemblies is projected to be 1.57E+20 n/cm2 (E > 1 MeV).

Therefore, the BFN jet pump hold-down beam assembly loss of preload analysis is acceptable and has been projected to the end of the subsequent period of extended operation.

Beyond the use of plant-specific design operating temperature at BFN Units 1, 2, and 3 -

Discuss any other differences in assumptions, methodology, inputs, etc between plant-specific evaluation in SIA Calc No. 2200107.308P and the generic assessment

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Appendix G of BWRVIP-315 provides a generic assessment that presents stress relaxation calculations that determine the total loss of preload in the jet pump beam and compares those values to the jet pump beam required preload.

By email dated October 31, 2023, the NRC transmitted the Final Safety Evaluation for BWRVIP-315, Reactor Internals Aging Management Evaluation for Extended Operations (ML23251A056).

SIA Calc No. 2200107.308P provides the applicants assessment for the Jet Pump Hold-Down Beam Assembly for BFN, Units 1, 2 and 3 for 80-years of operation. The staff noted that the methodology in SIA Calc No. 2200107.308P and Appendix G of BWRVIP-315 are consistent and/or similar. It was noted that the plant-specific design operating temperature at BFN Units 1, 2, and 3 was used in lieu of the operating temperature assumed in the generic analysis of BWRVIP-315 in Appendix G of BWRVIP-315. For any difference provide an explanation/basis.

Based on the peak 80-year projected neutron fluences for Units 1, 2 and 3 jet pump beam assemblies - it appears the neutron fluence criteria in the generic assessment in Appendix G of BWRVIP-315 was met - Why did the applicant decide to perform a plant-specific evaluation that appears to have followed the same methodology of the generic assessment in Appendix G of BWRVIP-315.

SLRA Section TLAA: 4.2.19 - Jet Pump Sensing Line Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.2.19 SLRA Section 4.2.19 states "The original design of the sensing line clamp assumed a fluence value of 1.0E+19 n/cm2 corresponding to a preload relaxation of 5% for the 40-year design life."

Tables 2, 3 and 4 of SIA CALC - 2200107.406P indicate that

%relaxation in these components would be greater than the original design analysis at a much lower neutron fluence.

Discuss the discrepancy in information in SLRA Section 4.2.19 and SIA CALC - 2200107.406P related to the preload relaxation of the jet pump sensing line.

2 4.2.19 SLRA Section 4.2.19 Jet Pump Sensing Line Clamps Loss of Preload Analysis states The original design of the sensing line clamp assumed Provide applicant documentation that is

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions a fluence value of 1.0E+19 n/cm2 corresponding to a preload relaxation of 5% for the 40-year design life.

SLRA Section 4.2.19 - indicates that this TLAA was dispositioned in accordance with 10 CFR 54.21(c)(1)(i) and that the jet pump sensing line clamp loss of preload analysis remains valid through the subsequent period of extended operation.

During its audit, the staff noted that the neutron fluence value referenced in SLRA 4.2.19 was provided to the applicant via communication with GE. It appears that this communication was via e-mail.

Since this TLAA is being disposition in accordance with 10 CFR 54.21(c)(1)(i) - the staff is seeking applicant documentation that is auditable and retrievable that indicates that the applicants TLAA for Jet Pump Sensing Line Clamps Loss of Preload was based on an assumed value cited in the SLRA.

auditable and retrievable that indicates that the applicants TLAA for Jet Pump Sensing Line Clamps Loss of Preload was based on an assumed value cited in the SLRA.

SLRA Section /TLAA/SLRA Section 4.2.5 and 4.2.6: Reactor Vessel Circumferential Weld Failure Probability Analyses & Reactor Vessel Axial Weld Failure Probability Analyses Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.2.5 &

4.2.6 4-43 & 4-47 In the TLAA Calculations and Evaluations section of the portal there is a set of analyses titled Reactor Pressure Vessel Axial and Circumferential Weld Analyses with BWRVIP-329-A" used to justify the disposition to 10 CFR 54.21(c)(1)(ii) and 10 CFR 54.21(c)(1)(iii).

The staff noted that SLRA Tables 4.2.5 -1, 3 provides the relevant information for BFN Units 1, 2 and 3, respectively for the limiting plate circumferential weld for the Reactor Vessel Circumferential Weld Failure Probability Analyses TLAA.

The staff noted that SLRA Tables 4.2.6 -1, 3 provides the relevant information for BFN Units 1, 2 and 3, respectively for the limiting plate Confirm if EOI RTMAX (°F) was calculated consistent with RG 1.99, Rev. 2 for calculating ART at the 0T location.

Was it intentional to include a margin value when calculating EOI RTMAX (°F) for 50, 64, and 62 EFPY for Units 1, 2, and 3, respectively.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions axial weld for the Reactor Vessel Circumferential Weld Failure Probability Analyses TLAA.

These analyses utilize guidance of BWRVIP-329-A to justify elimination of ASME Code,Section XI examination of RPV circumferential welds and reduction of examination of RPV axial welds.

Contained within the analyses are tables 2, 3, and 4, (pages 5-7) and contain material adjusted reference temperature data for unit 1, 2, and 3 respectively.

The staff reviewed the data in the each of the tables and observed the calculation of EOI RTMAX (°F), appears to include the margin value contained in Regulatory Guide (RG) 1.99, Revision 2.

2 4.2.5 &

4.2.6 4-43 & 4-47 In the TLAA Calculations and Evaluations section of the portal there is a set of analyses titled Reactor Pressure Vessel Axial and Circumferential Weld Analyses with BWRVIP-329-A" used to justify the disposition to 10 CFR 54.21(c)(1)(ii) and 10 CFR 54.21(c)(1)(iii).

CALCULATION 2200107.401P.R0 - BFN BWRVIP-329-A.PDF on ePortal - plant-specific analyses of BWRVIP-329-A, The staff noted that SLRA Tables 4.2.5 -1, 3 provides the relevant information for BFN Units 1, 2 and 3, respectively for the limiting plate circumferential weld for the Reactor Vessel Circumferential Weld Failure Probability Analyses TLAA.

The staff noted that SLRA Tables 4.2.6 -1, 3 provides the relevant information for BFN Units 1, 2 and 3, respectively for the limiting plate axial weld for the Reactor Vessel Circumferential Weld Failure Probability Analyses TLAA.

These analyses utilize guidance of BWRVIP-329-A to justify elimination of ASME Code,Section XI examination of RPV circumferential welds and reduction of examination of RPV axial welds.

Contained within the analyses are tables 2, 3, and 4, (pages 5-7) and contain material adjusted reference temperature data for unit 1, 2, and 3 respectively.

Clarify whether the licensee is crediting the plant-specific calculated EOI RTMAX (°F) values for the limiting plate and welds or the EOI RTMAX

(°F) values generated from the surveillance materials from the BWRVIP ISP (i.e., the ones contained in tables 2, 3, and 4, or the ones contained in the footnotes).

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions At the footnotes of these tables the staff observed the inclusion of the licensees EOI RTMAX (°F) from the Integrated Surveillance Program, as values different from values that corresponds to each of the limiting plate and weld values included in the tables.

SLRA Item 3.1.1-003 and Sections 3.1.2.2.1, Cumulative Fatigue Damage and 4.3, Metal Fatigue Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

Table 3.1.1, Item 3.1.1-003; Sections 3.1.2.2.1 and 4.3; 3-42 SLRA Table 3.1.1, Item 003 (also called Item 3.1-1, 003) describes the aging management review (AMR) results for the stainless steel and nickel alloy reactor vessel internal (RVI) components that are subject to cumulative fatigue damage. This AMR item is also related to SLRA Section 3.1.2.2.1 for the further evaluation of the aging effect of cumulative fatigue damage and Section 4.3 for metal fatigue time-limited aging analysis (TLAA).

In comparison, SLRA Table 3.1.2-2 indicates that SLRA Table 3.1.1, Item 003 is applied to the following RVI components that are also associated with GALL-SLR Item IV.B1.R-53: (1) jet pump assemblies:

jet pump sensing line; (2) RVI components: fuel supports and control rod drive assemblies; (3) jet pump assemblies: thermal sleeve inlet header, riser brace arm, hold-down beams, and wedges; (4) RVI components: top guide; (5) core shroud and core plate: access hole cover (mechanical - Units 1 and 2); (6) core shroud and core plate:

core shroud (upper, central, lower); (7) core shroud and core plate:

access hole cover (welded - Unit 3); (8) core shroud and core plate:

core shroud support structure (shroud support cylinder, shroud support plate, shroud support legs); (9) core spray lines and spargers: core spray lines (headers), spray rings, spray nozzles, thermal sleeves; (10) core spray sparger nozzle elbows; and (11) jet pump assemblies:

castings (inlet elbow, mixing assembly, diffuser casting, restrainer bracket).

1. Clarify which subsections of SLRA Section 4.3 describe the fatigue TLAA for the components that are associated with SLRA Table 3.1.1, Item 3 (also called Item 3.1-1, 003) and GALL-SLR Item IV.B1.R-53, which are listed in the issue and background section:

2. If the RVI components are bounded by certain limiting components, identify the limiting components for the RVI components associated with SLRA Table 3.1.1, Item 003.

3. If a new fatigue TLAA section or a revision to the existing fatigue

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions However, SLRA Section 4.3 does not clear describe which subsection of SLRA Section 4.3 address the fatigue TLAA for the RVI components discussed above.

TLAA section is needed to address the components associated with SLRA Table 3.1.1, Item 003, discuss a plan to revise SLRA Section 4.3.

SLRA Section 4.3.1, Transient Cycle and Cumulative Usage Projections for 80 Years Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.3.1 4-62 SLRA Section 4.3.1 addresses the transient cycle and cumulative usage projections for 80 years of operation.

The applicant did not identify the transient cycle projections and cumulative usage (CUF and CUFen) projections as a time-limited aging analysis even though the projections are time-dependent.

The staff needs clarification on why the projections of transient cycles and cumulative usage parameters are not identified as a TLAA.

1. Discuss why the projections of transient cycles and cumulative usage parameters are not identified as a TLAA.

2 4.3.1 4-62 SLRA Tables 4.3.1-1, 4.3.1-2, and 4.3.1-3 describe the cumulative transient cycles (up to December 31, 2022) and 80-year design transient cycles for Browns Ferry Nuclear Plant (BFN) Units 1, 2 and 3, respectively.

In addition, the following reference indicates that the applicant used the most recent 10-year transient cycles up to December 31, 2022 to project the 80-year cycles (

Reference:

SIA Report 1801147.301, Revision 2, BFN 80 Year Cycle Projection Update for SLR ).

However, it is not clear to the staff why only the most recent 10-year cycle data (up to December 31, 2022) were used in the 80-year cycle projections.

2. Clarify the technical basis for why only the most recent 10-year cycle data (up to December 31, 2022) were used in the 80-year cycle projections. As part of the discussion, describe why the most recent 10-year data can represent the operation and its characteristics for the

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions subsequent period of extended operation.

3 4.3.1 4-65 to 4-73 SLRA Tables 4.3.1-1, 4.3.1-2, and 4.3.1-3 describe the cumulative transient cycles (up to December 31, 2022) and 80-year design transient cycles for Browns Ferry Nuclear Plant (BFN) Units 1, 2 and 3, respectively.

In the cycle projections for some of the transients, the applicant used an accumulation rate of zero in the cycle projections for the operation after December 31, 2022. The transients, which are projected to have no additional cycle accumulation after December 31, 2022, are listed below.

• Unit 1: Transients 5, 6, 7, 8, 14, 15, 17, 18, 29, and 33

• Unit 2: Transients 5, 6, 7, 8, 10, 12, 14, 17, 18, 19, 23, 25, and 29

• Unit 3: Transients 5, 8, 12, 15, 17, 22, 23, 25, 26, 27, 28, 29, and 33 For Transient 8 (hydrostatic test to 1563 psig transient) of each unit, SLRA Section 4.3.1 explains that the test has been performed during plant construction and is not expected to occur again.

However, for the other transients that are projected to have no additional cycles after December 31, 2022, the SLRA does not clearly describe why these transients are projected to have no additional cycles.

1. For the transients that are projected to have no additional cycles after December 31, 2022, provide justification for why each of these transients is projected to have no additional cycles after December 31, 2022. If justification cannot be provided, describe the cycle projections for these transients using non-zero cycle accumulation rates.

2. If the projection no additional cycle is based on the actual cycle data of the most recent 10-year operation (up to December 31, 2022),

discuss why the most recent 10 year operation is different from the previous operating period in terms of cycle accumulation and related operating characteristics.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 4

4.3.1 4-65 to 4-73 SLRA Tables 4.3.1-1, 4.3.1-2, and 4.3.1-3 describe the cumulative transient cycles (up to December 31, 2022) and 80-year design transient cycles for Browns Ferry Nuclear Plant (BFN) Units 1, 2 and 3, respectively.

In the cycles projections for some of the transients,, the applicant added very small numbers of additional cycles for the operation after December 31, 2022. For example, Unit 1 Transient 22 (scram - other scrams transient) has cumulative cycles of 52 up to December 31, 2022, but the transient is projected to have additional cycles of only 7 after December 31, 2022.

Since the Unit 1 operating period between the start of operation (December 20, 1973) and the ending reference point for cumulative cycles (up to December 31, 2022) is approximately 39 years, the additional operating period after December 31, 2022 for 80 years operation is approximately 41 years (i.e., operating period for projections of additional cycles), which is comparable to the operating period for the up-to-date cycle calculation (39 years). However, the additional projected cycle number of 7 for the rest of the operation (41 years) after December 31, 2022 is very small compared to the cumulative cycles up to December 22, 2022 (i.e., 52 cycles).

In a similar manner to the discussion above, the following transiens are projected to have very small number of additional cycles after December 31, 2022 even though the additional projected cycles are not zero.

Unit 1: Transients 19, 22, 23, 25, 26, 27, and 28 Unit 2: Transients 15, 22, 26, 28, and 36 Unit 3: Transients 6, 7, 10, 14, and 18 For these transients, the ratio of the additional cycles after December 31, 2022 to the cycles up to December 31, 2022 is in the range from 0.02 to 0.2 approximately. The staff needs justification for why the number of the additional cycles projected after December 31, 2022 is

1. For the transients discussed in the issue section, provide justification for why the number of the additional cycles projected after December 31, 2022 is very small compared to number of the accumulated cycles up to December 31, 2022 in the 80-year cycle projections for the transients discussed above.

2. If the projections of the additional cycles after December 31, 2022 are based on the actual cycle data of the most recent 10-year operation (up to December 31, 2022), discuss why the most recent 10 year operation is different from the previous operating period in terms of cycle accumulation and related operating characteristics.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions so small compared to the number of the accumulated cycles up to December 31, 2022 in the 80-year cycle projections for the transients discussed above.

5 4.3.1 4-65 SLRA Tables 4.3.1-1, 4.3.1-2, and 4.3.1-3 describe the cumulative transient cycles (up to December 31, 2022) and 80-year design transient cycles for Browns Ferry Nuclear Plant (BFN) Units 1, 2 and 3, respectively.

Transient 2 in these tables is design basis accident transient.

However, these table does not clearly describe what the design basis accident transient is.

1. Describe the definition of the design basis accident transient (e.g., specific accident involved in the transient and the associated temperature and pressure changes during the transient).

6 4.3 4-62 SLRA Section 4.3 indicates that ASME Section VIII, Division 2 may require a fatigue analysis or assume a stated number of full-range thermal and displacement transient cycles.

However, SLRA Section 4.3 does not clearly discuss whether the applicants fatigue TLAA includes any component subject to the fatigue analysis per ASME Section VIII, Division 2.

1. Clarify whether the applicants fatigue TLAA include any component subject to the fatigue analysis per ASME Section VIII, Division 2. If so, discuss the following:

(1) components subject to the fatigue analysis per ASME Code Section VIII, Division 2; and (2)

TLAA evaluation and disposition for the components, including the specific Code provisions and the associated transient cycles and 80-year projected cycles.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA Section 4.3.10, BFN Unit 3 Core Spray Lower Line Section Replacement Fatigue Evaluation Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.3.10 4-94 SLRA Section 4.3.10 addresses the fatigue analysis for BFN Unit 3 core spray lower line replaced section. The applicant dispositioned this fatigue TLAA in accordance with 10 CFR 54.21(c)(1)(iii) by proposing to use the Fatigue Monitoring AMP and BWR Vessel Internals AMP for aging management.

SLRA Section 4.3.10 explains that a comparison of Unit 3 projected 80-year cycles to the design cycles used in the fatigue analysis indicates that the 80 year-projected cycles do not exceed 2 times the number of the design cycles assumed in the fatigue analysis.

However, SLRA Section 4.3.10 does not describe the design cycles assumed in the fatigue analysis and the associated 80-year projected cycles.

1. Describe the 40-year design cycles assumed in the fatigue analysis and the associated 80-year projected cycles to demonstrate that the 80 year-projected cycles do not exceed 2 times the number of the design cycles assumed in the fatigue analysis. In this discussion, describe what the relevant design transients are.

2. Describe how the applicant determined the 80-year projected cycles and the technical basis for the applicants cycle projections.

3. Clarify whether the TLAA is only applied to Unit 3 and its technical basis.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 2

4.3.10 4-94 SLRA Section 4.3.10 addresses the fatigue analysis for BFN Unit 3 core spray lower line replaced section. The applicant dispositioned this fatigue TLAA in accordance with 10 CFR 54.21(c)(1)(iii) by proposing to use the Fatigue Monitoring AMP and BWR Vessel Internals AMP for aging management.

However, SLRA Section 4.3.10 does not clearly discuss the method and frequency of the inspections for the core spray lower line replaced section. In addition, the staff needs clarification on whether the inspections for the core spray lower line replaced section need to be identified as an enhancement to the existing BWR Vessel Internals AMP.

1. Describe the method and frequency of the inspections for the core spray lower line replaced section for the aging management of the effect of fatigue. In addition, discuss why these inspections are sufficient to manage the effects of fatigue for the repaired T-box.

2. Clarify whether the inspections discussed in Request 1 need to be identified as an enhancement to the existing BWR Vessel Internals AMP (SLRA Section B.2.1.7).

SLRA Section 4.3.11, Jet Pump to Core Shroud Support Plate Fatigue Evaluation Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.3.11 4-95 SLRA Section 4.3.11 addresses the fatigue analysis for the jet pump diffuser to core shroud support weld location.

SLRA Table 4.3.11-1 describes that the transients, which contribute to the fatigue of the jet pump weld location, and the 80-year projected cycles for the transient.

1. Discuss how the 80-year projected cycles listed in SLRA Table 4.3.11-1 were determined. As part of the discussion, clarify whether the startup/major

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions However, the SLRA does not clearly discuss how the projected cycles were determined. For example, the HPCI-startup transient (high pressure coolant injection startup transient) is not included in SLRA Table 4.3.1-1, 4.3.1-2 or 4.3.1-3.

In addition, the staff needs clarification on the following items: (1) which transient listed in SLRA Tables 4.3.1-1, 4.3.1-2 and 4.3.1-3 (cycle projection tables for fatigue analyses) corresponds to the sudden start of cold pump transient listed in SLRA Table 4.3.11-1; (2) the definition of the DBA [design basis accident] transient listed in SLRA Table 4.3.11-1; and (3) whether the 80-year projected cycles listed in SLRA Table 4.3.11-1 are bounding for the projected cycles for BFN Units 1, 2 and 3.

heatup transient cycles listed in SLRA Tables 4.3.1-1, 4.3.1-2 and 4.3.1-3 consist of the HPCI-startup transient cycles and startup-shutdown transient cycles listed in SLRA Table 4.3.11-1.

2. Clarify which transient in SLRA Tables 4.3.1-1, 4.3.1-2 and 4.3.1-3 (cycle projection tables for fatigue analyses) corresponds to the sudden start of cold pump transient listed in SLRA Table 4.3.11-1.

3. Describe the definition of the DBA

[design basis accident] transient listed in SLRA Table 4.3.11-1.

4. Clarify whether the 80-year projected cycles listed in SLRA Table 4.3.11-1 are bounding for the projected cycles of

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions BFN Units 1, 2 and

3.

2 4.3.11 4-95 SLRA Section 4.3.11 addresses the fatigue analysis for the jet pump diffuser to core shroud support weld location.

SLRA Table 4.3.11-1 describes the allowable cycles for each of the transients that contribute to the fatigue of the jet pump weld location.

However, SLRA Section 4.3.11 does not clearly describe how the applicant determined the allowable cycles for each transient.

Discuss how the applicant determined the allowable cycles for each transient listed in SLRA Table 4.3.1-11 (e.g.,

Code provisions such as fatigue design curves used in the fatigue analysis, and the assumption associated with the allowable cycle determination).

3 4.3.11 A.4.3.11 4-95 A-82 SLRA Section 4.3.11 addresses the fatigue analysis for the jet pump diffuser to core shroud support weld location.

SLRA Section 4.3.11 indicates that the evaluated jet pump location is a location where susceptibility to fatigue is a concern due to dynamic forces from jet pump flow and thermal stresses of the interfacing liquid temperature differences.

A similar discussion regarding the susceptibility to flow-induced dynamic forces (i.e., flow induced vibration) is also discussed in the supplement for this TLAA in SLRA Section A.4.3.11.

However, SLRA Sections 4.3.11 and A.4.3.11 do not address the cumulative usage due to flow-induced vibration.

In light of the discussion on the susceptibility to flow-induced vibration, explain why SLRA Sections 4.3.11 and A.4.3.11 do not address the cumulative usage resulting from flow-induced vibration. If flow-induced vibration does not cause the effect of fatigue on the jet pump location, provide the technical basis of the applicants determination. As part of the discussion, clarify whether the cycle stresses due to flow-induced vibration for the jet pump location and

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions other jet pump locations are less than the fatigue endurance limit.

SLRA Section 4.3.2, Metal Fatigue of Class 1 Components Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.3.2 4-76 SLRA Section 4.3.2 addresses the metal fatigue TLAA for ASME Code Section III Class 1 piping and components. In addition, SLRA Table 4.3.1-4 describes the 80-year CUF and CUFen values for the limiting Class 1 locations for 80 years of operation.

However, SLRA Section 4.3.2 and Table 4.3.1-4 do not clearly discuss the CUF for the recirculation pumps and valves as Class 1 components. The staff needs clarification on whether the recirculation pumps and valves are bounded by other Class 1 components in terms of CUF.

1. Explain why SLRA Section 4.3.2 and Table 4.3.1-4 do not clearly discuss the CUF for the recirculation pumps and valves as Class 1 components. If the recirculation pumps and valves are bounded by the other Class 1 components in terms of CUF, identify the limiting components and the CUF values for the bounding locations and bounded pumps and valves.

2. In addition, clarify whether the scope of the fatigue analysis in SLRA Section 4.3.2 includes all the Class 1 piping and components. If not, identify the Class 1 piping and

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions components omitted in the scope of the fatigue analysis and provide justification for the omission of those piping and components (this discussion does not need to include the components waived from fatigue analyses separately discussed in SLRA Section 4.3.3).

2 4.3.2 4-76 SLRA Section 4.3.2 addresses the metal fatigue TLAA for ASME Code Section III Class 1 piping and components. In addition, SLRA Table 4.3.1-4 describes the CUF and CUFen values for the limiting Class 1 locations for 80 years of operation.

SLRA Table 4.3.1-4 indicates that the 80-year CUFen value for the reactor vessel main closure studs is 0.726 and that the 80-year CUF value for the studs is bounded by the CUFen value. However, it is not clear to the staff why environmental effects need to be considered in the fatigue analysis for the main closure studs given that the main closure studs are not exposed to reactor coolant during the normal operation.

In addition, the staff noted that the following reference describes the 40-year CUF value for the main closure studs prior to the extended power uprate is 0.726, which is identical to the 80-year CUFen value listed in SLRA Table 4.3.1-4 (

Reference:

NEDO-33860, Revision 1, Safety Analysis Report For Browns Ferry Nuclear Plant Units 1, 2, And 3 Extended Power Uprate, October 2016, Table 2.2-6, ADAMS Accession No. ML16302A441).

1. Clarify whether the cumulative usage value for the main closure studs listed in SLRA Table 4.3.1-4 is the CUF value rather than CUFen value. If it is the CUFen value, explain why CUFen is calculated for the main closure studs even though the studs are not exposed to the reactor coolant during the normal operation.

2. Given that the reference document indicates that the 40-

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Given that the reference document indicates that the 40-year CUF value for the main closure studs prior to the extended power uprate is 0.726, the staff needs justification for why SLRA Table 4.3.1-4 indicates that the 80-year CUFen value of the main closure studs is 0.726 (i.e., the same value as the 40-year CUF before the extended power uprate discussed in the reference).

year CUF value for the main closure studs prior to the extended power uprate is 0.726, provide justification for why SLRA Table 4.3.1-4 indicates that the 80-year CUFen value of the main closure studs is 0.726, too (i.e., the same value as the 40-year CUF before the extended power uprate described in the reference).

SLRA Section 4.3.3, Class 1 Fatigue Waivers Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.3.3 4-76 SLRA Section 4.3.3 addresses the fatigue waiver TLAA for ASME Code Section III Class 1 components. The SLRA dispositions the fatigue waiver TLAA in accordance with 10 CFR 54.21(c)(1)(ii) to demonstrate that the fatigue waiver TLAA has been projected to the end of the subsequent period of extended operation.

SLRA Table 4.3.3-1 also describes the transient groups evaluated in the 80-year fatigue waiver analysis (e.g., startup and shutdown transient group and significant temperature fluctuation transient group) and the 80-year projected cycles for the transient groups. The comparison is intended to confirm that the 80-year projected cycles do not exceed the cycles evaluated in the 80-year fatigue waiver analysis.

1. Given that the number of some transient cycles evaluated in the 80-year fatigue analysis is the same as the 80-year projected cycles without a cycle margin, provide justification for (1) why the TLAA is not dispositioned in accordance with 10 CFR 54.21(c)(1)(iii) and (2) why the

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions SLRA Table 4.3.3-1 indicates that the number of cycles for the startup and shutdown transient group evaluated in the 80-year fatigue waiver analysis is the same as the 80-year projected cycles of BFN Unit 3 for the transient group (i.e., 266 cycles), which are bonding for the projected cycles of Units 1 and 2.

SLRA Table 4.3.3-1 also indicates that the number of cycles of the significant temperature fluctuation transient group evaluated for the liquid control nozzle in the 80-year fatigue waiver analysis is the same as the 80-year projected cycles of BFN Unit 2 for the transient group (i.e., 94 cycles), which are bounding for the projected cycles of Units 1 and 3.

In addition, SLRA Table 4.3.3-1 indicates that the number of cycles of the significant temperature fluctuation transient group evaluated for the fatigue-waived components other than the liquid control nozzle in the 80-year fatigue waiver analysis is the same as the 80-year projected cycles of BFN Unit 2 for the transient group (i.e., 93 cycles),

which are bounding for the projected cycles of Units 1 and 3.

Even though the number of some transient cycles evaluated in the 80-year fatigue analysis is the same as the 80-year projected cycles without a cycle margin, the SLRA does not indicate the use of the Fatigue Monitoring AMP to ensure that the actual transient cycles do not exceed the transient cycles assumed in the 80-year fatigue waiver analysis for the subsequent period of extended operation.

Fatigue Monitoring AMP is not used to ensure that the actual transient cycles do not exceed the transient cycles assumed in the 80-year fatigue waiver analysis for the subsequent period of extended operation.

2. If the Fatigue Monitoring AMP is not needed to manage the effect of fatigue on the validity of the 80-year fatigue waiver analysis, clarify whether a corrective action will be initiated as the actual cycles approach the transient cycles assumed in the 80-year fatigue waiver analysis.

SLRA Section 4.3.4, Metal Fatigue of Non-Class 1 Components Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.3.4 4-80 SLRA Section 4.3.4 addresses the fatigue TLAA for the non-Class 1 piping systems (i.e., ASME Section III Class 2 and 3 and ANSI B31.1

1. Discuss how the 80-year projected cycles were determined

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions piping systems). The TLAA is also related to the allowable stress analyses for the piping systems.

The TLAA regarding allowable stress analyses relies on the implicit fatigue analysis provisions in the ANSI B31.1 code. These provisions allow no reduction in the allowable stress range for thermal expansion stresses if the number of equivalent full temperature cycles does not exceed 7000 cycles.

In addition, SLRA Tables 4.3.4-3 and 4.3.4-4 describe the 80-year projected cycles for the non-Class 1 piping systems other than the high temperature process sample system lines and the 80-year projected cycles for the high temperature process sample system lines, respectively.

However, SLRA Section 4.3.3 does not clearly describe how the 80-year projected cycles were determined (e.g., based on piping system design information, plant operation procedures, test requirements, FSAR information and specific system-level knowledge).

(e.g., based on piping system design information, plant operation procedures, test requirements, FSAR information and specific system-level knowledge). As part of the discussion, clarify whether transient cycles per a unit time period (e.g.,

annual cycles) are used in the 80-year cycle projections.

2 4.3.4 4-80 SLRA Section 4.3.4 addresses the fatigue TLAA for the non-Class 1 piping systems (i.e., ASME Section III Class 2 and 3 and ANSI B31.1 piping systems). The TLAA is also related to the allowable stress analyses for the piping systems.

SLRA Section 4.3.4 indicates that the non-Class 1 piping portions of the following piping systems are only affected by the same pressure and temperature transients as the reactor coolant system (RCS) transients that are listed in SLRA Table 4.3.4-2: (1) control rod drive, (2) core spray, (3) feedwater, (4) main steam, (5) containment atmosphere dilution, (6) residual heat removal (RHR) (including residual heat removal service water (RHRSW) since transients are bounded the RHR transients), and (7) standby liquid control piping systems.

1. Clarify why the non-Class 1 piping portions of the following piping systems are not subject to piping-specific transients other than the RCS transients listed in SLRA Table 4.3.4-2:

(1) control rod drive, (2) core spray, (3) feedwater, (4) main steam, (5) containment atmosphere dilution,

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions However, SLRA Section 4.3.4 does not clearly discuss why the non-Class 1 piping portions of these piping systems are not subject to piping-specific transients other than the RCS transients listed in SLRA Table 4.3.4-2.

(6) RHR (including RHRSW), and (7) standby liquid control piping systems.

SLRA Section 4.3.5, Environmental Fatigue Analyses for Reactor Vessel and Class 1 Piping Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.3.5 4-87 SLRA Section 4.3.5 addresses the environmentally-assisted fatigue (EAF) time-limited aging analysis (TLAA). EAF is also called environmental fatigue. In addition, SLRA Table 4.3.1-4 describes the 80-year CUF and CUFen values for the limiting Class 1 locations for 80 years of operation.

However, SLRA Section 4.3.5 and Table 4.3.1-4 do not clearly discuss the CUFen for the recirculation pumps and valves as Class 1 components. The staff needs clarification on whether the recirculation pumps and valves are bounded by other Class 1 components in terms of CUFen.

1. Explain why SLRA Section 4.3.5 and Table 4.3.1-4 do not clearly discuss the CUFen for the Class 1 pumps and valves as Class 1 components.

If the Class 1 pumps and valves are bounded by the other Class 1 components in terms of CUFen, identify the limiting components.

2. In addition, clarify whether the term in the title of Section 4.3.5, Class 1 Piping means the Class piping systems including Class 1 components. If not, provide justification for why the EAF analysis does not include Class 1 components (other

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions than the reactor vessel) in the scope of EAF.

2 4.3.5 4-87 SLRA Section 4.3.5 addresses the environmentally-assisted fatigue (EAF) TLAA. EAF is also called environmental fatigue.

SLRA Section 4.3.5 indicates that the locations in contact with steam were excluded from the EAF screening. However, SLRA Section 4.3.5 does not provide a technical basis for excluding the Class 1 piping and components exposed to steam from the EAF analysis.

The staff needs additional information on why the Class 1 piping and components exposed to steam (e.g., reactor vessel steam outlet nozzle and main steam piping line) do not need to be included in the EAF analysis.

1. Provide the technical basis for why the Class 1 piping and components exposed to steam are not included in the EAF analysis.

2. Describe the 80-year limiting CUF locations and associated CUF values for the Class 1 piping and components exposed to steam, including the 80-year CUF value for the reactor vessel steam outlet nozzle. In addition, clarify whether the CUF values of the limiting steam-exposed Class 1 locations are bounded by the CUF values of the limiting Class 1 locations exposed to the liquid coolant.

3. Describe the inspections

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions performed on the limiting CUF locations of the steam-exposed Class 1 piping and components and the inspection results, including the inspection results for the steam outlet nozzle. In addition, discuss the industry operating experience regarding the fatigue crack initiation and growth in the BWR Class 1 piping and components exposed to steam to support the applicants approach.

3 4.3.5 4-87 SLRA Section 4.3.5 addresses the environmentally-assisted fatigue (EAF) TLAA. The SLRA section indicates that the limiting locations described in NUREG/CR-6260 for BWR plants are evaluated in the EAF analysis. The staff noted that the locations in NUREG/CR-6260 include the reactor recirculation piping and core spray line reactor vessel nozzle.

In comparison, SLRA Table 4.3.1-4 describes the limiting EAF locations of the applicants analysis, including the applicable NUREG/CR-6260 locations, and the 80-year CUFen values for the limiting locations (also called sentinel locations).

However, SLRA Table 4.3.1-4 does not include the reactor recirculation piping and core spray line reactor vessel nozzle, which are listed in NUREG/CR-6260, as limiting EAF locations.

1. Discuss why the following locations described in NUREG/CR-6260 are not included in SLRA Table 4.3.1-4: (1) reactor recirculation piping and (2) core spray line reactor vessel nozzle. As part of the discussion, clarify the following items:(1) whether these locations are

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions bounded by other limiting EAF locations and (2) if so, the CUFen and Fen values and material of fabrication for these locations and the bounding EAF locations.

4 4.3.5 4-88 SLRA Section 4.3.5 addresses the environmentally-assisted fatigue (EAF) TLAA.

SLRA Section 4.3.5 explains that, during the EAF screening evaluation, a maximum Fen (environmental fatigue correction factor) was calculated based on material type, maximum temperature, and chemistry and a screening 80-year CUFen (environmentally adjusted CUF) value was calculated.

However, SLRA does not clearly discuss how the maximum Fen value was calculated in the EAF screening evaluation (e.g., how the temperature and strain rate were determined).

In addition, SLRA Section 4.3.5 indicates that the threshold screening CUFen is 1.0 for the conservative screening CUFen. Therefore, the staff needs clarification on how the screening CUFen values were further refined to determine the 80-year CUFen values after the screening evaluation.

1. Clarify how the applicant calculated the maximum Fen value in the EAF screening evaluation (e.g., how the temperature and strain rate were determined).

2. Discuss how the screening CUFen values were further refined to determine the 80-year CUFen values after the screening evaluation.

SLRA Section 4.3.6, Replacement Steam Dryer Stress Report and Fatigue Evaluation Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.3.6 4-89 SLRA Section 4.3.6 addresses the flow-induced vibration and related time-limited aging analysis (TLAA) for the replacement steam dryers.

1.

Clarify whether the TLAA addressed in SLRA Section 4.3.6 is based on the flow-induced vibration

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions In comparison, the title of SLRA Section 4.3.6, Replacement Steam Dryer Stress Report and Fatigue Evaluation is more general and does not specifically refer to the flow-induced vibration or high cycle fatigue as a topic for the fatigue TLAA section.

Therefore, the staff needs clarification on whether the TLAA addressed in SLRA Section 4.3.6 is based on the flow-induced vibration and related high cycle fatigue analysis but is not related to a fatigue analysis due to low cycle fatigue (resulting from thermal transients) or non-cyclic time-dependent stress analysis.

and related high cycle fatigue analysis such that the TLAA is not related to a fatigue analysis due to low cycle (resulting from thermal transients) or non-cyclic time-dependent stress analysis. If the TLAA in the section is related to a low cycle fatigue or non-cyclic time-dependent stress analysis, provide information regarding those analyses (e.g., CUF values due to thermal transients and time-dependent stress values against allowable stresses).

2 4.3.6 4-89 SLRA Section 4.3.6 addresses the flow-induced vibration and related fatigue TLAA for the replacement steam dryers.

The SLRA section also refers to the following reference as the technical basis for the high cycle fatigue TLAA addressing the effects of fatigue due to flow-induced vibration (

Reference:

GE Report NEDC-33824P, Revision 0, Browns Ferry Replacement Steam Dryer Stress Analysis, August 2015). This reference is part of the applicants submittal for the extended power uprate project, which was approved by the NRC staff in 2016 (ML16144A645).

1.

Clarify whether the evaluation of flow-induced vibration and related fatigue analysis in the refence document (i.e., negligible effect of flow-induced vibration based on the fatigue

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions The staff noted that the above reference (e.g., Sections 2.1 and 4.3) discusses that the cyclic stresses applied to the replacement steam dryers do not exceed the fatigue endurance limit for the components with margins. Therefore, the reference document concludes that the effect of fatigue due to flow-induced vibration is not a concern for the structural integrity of the replacement steam dryers.

However, SLRA Section 4.3.6 does not clearly discuss this aspect of the flow-induced vibration and related fatigue analysis described in the reference document.

endurance limit analysis) is applied to the fatigue TLAA described in SLRA Section 4.3.6. If not, provide the following information regarding the fatigue analysis in SLRA Section 4.3.6: (1) 80-year CUF value for the replacement steam dryers; (2) transient cycles used in the CUF analysis; and (3)

Code provisions used in the fatigue analysis.

SLRA Section 4.3.7, Emergency Equipment Cooling Water System Weld Flaws Evaluation Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.3.7 4-93 SLRA Section 4.3.7 addresses the fatigue crack growth analysis for emergency equipment cooling water system (EECW) weld flaws.

SLRA Section 4.3.7 indicates that the inspections performed in 1987 found 27 flaws that needed crack growth analyses. The SLRA also indicates that subsequently the applicant performed fatigue crack growth analyses on these flaws. The SLRA further describes the reevaluation of the crack growth calculations for the subsequent period of extended operation.

However, SLRA Section 4.3.7 does not clearly discuss the following information related to the operating experience and crack growth

1. Clarify the following items related to the operating experience and crack growth analyses: (1) aging mechanism that caused the cracks in the EECW welds; (2) whether the crack growth analyses need to consider degradation mechanisms other

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions analyses: (1) aging mechanism that caused the cracks in the EECW; (2) whether the crack growth analyses need to consider degradation mechanisms other than fatigue (e.g. stress corrosion cracking); (3)

ASME Code,Section III Class of the piping and welds; (4) Code provisions used in the fatigue crack growth analyses; and (5) inspections on these welds required in accordance with ASME Code,Section XI, other applicable Codes, or applicants procedures.

than fatigue (e.g.,

stress corrosion cracking); (3) ASME Code Section III Class of the piping and welds; (4) Code provisions used in the fatigue crack growth analyses; and (5) inspections on these welds required in accordance with ASME Code,Section XI, other applicable Codes, or applicants procedures.

2 4.3.7 4-93 SLRA Section 4.3.7 addresses the fatigue crack growth analysis for emergency equipment cooling water system (EECW) weld flaws.

SLRA Section 4.3.7 indicates that the inspections performed in 1987 found 27 flaws that needed crack growth analyses. The SLRA indicates that the crack growth analyses determined a cycle limit of 125 for these welds.

The SLRA also explained that 17 flaws of the 27 flaws were re-evaluated and that the number of cycles to exceed the allowable crack depth was increased from 125 to 2600 cycles.

However, SLRA Section 4.3.7 does not clearly address the following items: (1) which weld flaws were reevaluated to have an increased acceptable cycle of 2600; (2) whether the reevaluation of the 17 flaws is also applied to the other 10 flaws (i.e., whether the increased acceptable cycle number of 2600 is applied to all the 27 flaws); and (3) what transients are evaluated in the crack growth analyses (i.e., what transients are associated with the 125 and 2600 cycles evaluated in the crack growth analyses).

1. Clarify the following items: (1) which weld flaws were reevaluated to have an increased acceptable cycle of 2600 (in this discussion, identify the relevant flaws using the flaw identification numbers listed in SLRA Table 4.3.7-1);

(2) whether the reevaluation of the 17 flaws is also applied to the other 10 flaws (i.e., whether the increased acceptable cycle number of 2600

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions is applied to all the 27 flaws); and (3) what transients are evaluated in the crack growth analyses (i.e., what transients are associated with the 125 and 2600 cycles evaluated in the crack growth analyses).

SLRA Section 4.3.8, Core Shroud Support Fatigue Analysis Reevaluation Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.3.8 4-93 SLRA Section 4.3.8 addresses the 80-year CUF and CUFen calculations for the core shroud support.

However, SLRA Section 4.3.8 does not clearly describe the following items: (1) the material of the core shroud location; (2) whether the analysis is bounding for BFN Units 1, 2 and 3; (3) whether the evaluated core shroud location is a reactor coolant pressure boundary location; and (4) whether the CUFen calculations are performed in accordance with the guidance in NUREG/CR-6909, Revision 1.

The staff also noted that that the 80-year CUFen value (0.256) described in SLRA Section 4.3.8 is identical to the CUFen value for the location 21 (reactor vessel Region C/shroud support low-alloy steel location) described in SLRA Table 4.3.1-4. Therefore, the staff needs to confirm that the location 21 in SLRA Table 4.3.1-4 is identical to the core shroud location evaluated in SLRA Sections 4.3.1 and 4.3.5 (environmentally assisted fatigue (EAF) analysis).

In comparison with SLRA Section 4.3.8, SLRA Section 4.3.5 indicates that the aging effects of EAF will be managed by using the Fatigue

1. Clarify the following items in relation to the CUFen calculation in SLRA Section 4.3.8: (1) the material of the core shroud location; (2) whether the analysis is bounding for BFN Units 1, 2 and 3; (3) whether the evaluated core shroud location is a reactor coolant pressure boundary location; and (4) whether the CUFen calculations are performed in accordance with the

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Monitoring AMP in accordance with 10 CFR 54.21(c)(1)(iii).

Therefore, the staff also needs resolution of the potential inconsistency between the TLAA dispositions for the core shroud location (i.e., 10 CFR 54.21(c)(1)(ii) in SLRA Section 3.4.8 and 10 CFR 54.21(c)(1)(iii) in SLRA Section 3.4.5) regarding the CUFen analysis for the core shroud support.

guidance in NUREG/CR-6909, Revision 1.

2. Clarify whether the core shroud location evaluated in SLRA Section 4.3.8 is the limiting EAF location 21 listed in SLRA Table 4.3.1-4 and discussed in SLRA Section 4.3.5 (EAF analysis).

3. Resolve the potential inconsistency between the EAF TLAA disposition between SLRA Sections 4.3.5 and 4.3.8 (i.e., (iii) disposition and (ii) disposition).

Similarly, resolve the potential inconsistency between the Class 1 fatigue TLAA disposition between SLRA Sections 4.3.2 and 4.3.8 (i.e., (iii) disposition and (iii) disposition)

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions

4. If the fatigue TLAA for the core shroud location evaluated in SLRA Section 4.3.8 is dispositioned in accordance with 10 CFR 54.21(c)(1)(ii),

provide the following information to confirm that the use of the Fatigue Monitoring AMP is not necessary for the core shroud location:

(1) the transient cycles used in the CUFen calculation; (2) whether the cycles are 80-year projected cycles or are based on conservative design cycles; and (3) if 80-year projected cycles are used in the CUFen calculation, the technical basis for why the fatigue monitoring is not necessary.

SLRA Section 4.3.9, BFN Unit 3 Core Spray T-box Repair Fatigue Evaluation Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

4.3.9 4-93 SLRA Section 4.3.9 addresses the fatigue analysis for the repaired T-box of BFN Unit 3 core spray piping. The applicant dispositioned this

1. Describe the design cycles assumed in

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions fatigue TLAA in accordance with 10 CFR 54.21(c)(1)(iii) by proposing to use the Fatigue Monitoring AMP and BWR Vessel Internals AMP for aging management.

SLRA Section 4.3.9 explains that a comparison of Unit 3 projected 80-year cycles to the design cycles used in the fatigue analysis indicates that the 80 year-projected cycles do not exceed 2 times the number of the design cycles assumed in the fatigue analysis.

SLRA Section 4.3.9 also explains that the fatigue analysis considers two configurations: (1) configuration where the piping remains attached to the T-box and (2) configuration where the piping has become detached from the T-box. For the detached configuration, SLRA Section 4.3.9 indicates that the initially calculated cumulative usage factor (CUF) was greater than 1.0 and was subsequently reduced to 0.9 based on an assumption of reduced cycles associated with the remaining life of the plant.

However, SLRA Section 4.3.9 does not describe the design cycles assumed in the fatigue analysis and the associated 80-year projected cycles for the attached and detached configurations.

In addition, the staff needs the following items as basis information of the analysis: (1) Code provisions used in the CUF calculations; (2) specific location evaluated in the fatigue analysis; (3) the specific operating time associated with the CUF value of 0.9 for the detached configuration; and (3) aging mechanisms that can cause the detachment of the T-box from the piping and the extent of the detachment assumed in the fatigue analysis.

the fatigue analysis and the associated 80-year projected cycles to demonstrate that the 80 year-projected cycles do not exceed 2 times the number of the 40-year design cycles assumed in the fatigue analysis.

In this discussion, describe the design cycles and 80-year projected cycles for both the attached configuration and detached configuration and what the relevant design transients are.

2. Describe how the applicant determined the 80-year projected cycles and the technical basis for the applicants cycle projections.

3. Discuss how the applicant determined the CUF values (e.g.,

Code provisions such as fatigue design curves, and the

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions associated assumptions for the CUF calculations).

4. Describe the specific locations evaluated in the fatigue analysis for the repaired T-box.

5. Clarify the operating time associated with the CUF value of 0.9 for the detached configuration (e.g.,

40 years, 60 years or 80 years of operation)

6. In relation to the detached configuration evaluated in the fatigue analysis, describe the likely cause of the detachment and the extent of the detachment assumed in the fatigue analysis (e.g., what aging mechanisms can cause the detachment of the T-box from the core spray piping and the

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions extent of detachment assumed in the fatigue analysis).

7. Clarify whether this TLAA is only applied to Unit 3 and its technical basis.

2 4.3.9 4-93 SLRA Section 4.3.9 addresses the fatigue analysis for the repaired T-box of BFN Unit 3 core spray piping line. The applicant dispositioned this fatigue TLAA in accordance with 10 CFR 54.21(c)(1)(iii) by proposing to use the Fatigue Monitoring AMP and BWR Vessel Internals AMP for aging management.

However, SLRA Section 4.3.9 does not clearly discuss the method and frequency of the inspections for the repaired T-box and adjacent core spray piping. In addition, the staff needs clarification on whether the inspections for the repaired T-box need to be identified as an enhancement to the existing BWR Vessel Internals AMP.

1. Describe the method and frequency of the inspections for the repaired T-box and adjacent core spray piping for the aging management of the effect of fatigue. In addition, discuss why these inspections are sufficient to manage the effects of fatigue for the repaired T-box.

2. Clarify whether the inspections discussed in Request 1 need to be identified as an enhancement to the existing BWR Vessel Internals Program (SLRA Section B.2.1.7).

SLRA Section 4.7.3 Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 1

4.7.3 4-107 to 4-108 Topic: SLRA Section 4.7.3, BFN Unit 2 Reactor Vessel Axial Weld Flaw Section 4.7.3 states that Current procedures establish a minimum temperature of 200.6°F up to normal operating pressure, which is consistent with the P-T Limits Curve (Reference 4.8.36, Figure 3.4.9-2) at 38 EFPY The staff cannot find 200.6°F in the P-T limits in Figure 3.4.9-2 of Reference 4.8.36 (ML15065A049) as being the minimum temperature.

The minimum temperature in Figure 3.4.9-2 is not 200.6°F.

a. Discuss what is meant by a minimum temperature of 200.6°F up to normal operating pressure.

b. Discuss why 200.6°F is consistent with the P-T limit curves in Figure 3.4.9-2 of Reference 4.8.36.

2 4.7.3 4-107 to 4-108 Section 4.7.3 states that The peak fluence value was calculated to occur at the lowest elevation of the weld flaw and had a value of 1.92E+15 n/cm2 which is well below the established embrittlement fluence threshold of 1.0E+17 n/cm2.

The staff is not clear how 1.92E+15 n/cm2 is derived. The staff could not find this neutron fluence in SLRA Sections 4.2.1.1 and 4.2.1.2.

These two sections discuss neutron fluence calculations.

a. Discuss briefly how a neutron fluence of 1.92E+15 n/cm2 was derived, e.g., where in SLRA is this neutron fluence discussed besides Section 4.7.3.

b. Confirm that neutron fluence of 1.92E+15 n/cm2 is the maximum neutron fluence that the axial weld will experience.

3 4.7.3 4-107 to 4-108 The Browns Ferry ePortal contains a report from Structural Integrity Associates (SIA), Inc., titled Browns Ferry Unit 2 Reactor Pressure Vessel Vertical Weld Flaw Evaluation for 80-Year SLR, File No.

2200107.307.

By letter dated October 4, 2021 (ML21277A123), TVA submitted an analysis of the same axial weld, V-3-A, at Unit 2 that was performed by SIA.

Discuss/explain the differences in methodology/assumptions in the SIA analysis attached to TVAs letter dated October 4, 2021 compared to the one documented on the

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions It appears that the October 4, 2021 submittal is similar to the SIA report in the ePortal.

The staff noted the neutron fluence appears to be much lower at 64 EFPY compared to 48 EFPY between the two calculations for this Unit 2 axial weld. The staff is seeking to understand if there are any additional differences between these evaluations.

eportal (SIA Calc 2200107.307).

Scoping and Screening - Various Mechanical Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question /

Request 1

2.3.3.9 pdf page 134 Drawing 0-47E865-2-SLR - shows two trains of the drywell cooler assembly. The two trains have different segments classified as in-scope.

Discuss why the two trains are treated differently.

2 2.3.3.11 pdf page 139 Drawings 0-47E847-2, -3, -4, and -5 shows changes between in-scope and out of scope in the middle of pipe runs.

Discuss how function is preserved and pressure boundary maintained.

3 2.3.3.12 pdf page 142 Drawing 0-47E845-1 shows components within scope and components out of scope without a clear boundary (isolation) between portions.

Discuss how function is preserved and pressure boundary maintained.

4 2.3.3.31 pdf page 173 Drawing 1-47E831-1 shows change in boundary without clear justification. Drawing 2-47E831-1 and 3-47E831-1, included within scope equipment that the other units excluded.

Provide justification for this difference.

5 2.3.4.6 pdf page 198 Drawings 1-47E831-1, 2-47E831-1 and 3-47E831-1 shows change in boundary without clear boundary between portions.

Discuss how function is preserved and pressure boundary maintained.

6 2.3.3.5 pdf page 127 Drawings 1-47E844-1-SLR, 3-47E844-1-SLR, 1-47E844-3-SLR, 2-47E844-3-SLR, and 3-47E844-3-SLR identify the boundary between within scope and out of scope sometimes occur in the middle of pipe segments and/or without isolation capability.

Discuss how these boundaries and the methodology are used to identify them.

7 2.3.3.6 pdf page 129 Drawings 1-47E836-1-2-SLR, 2-47E836-1-SLR, and 3-47E836-1-SLR identify the boundary between within scope and out of scope sometimes occur in the middle of pipe segments and/or without isolation capability.

Discuss how these boundaries and the methodology are used to identify them.

8 2.3.3.26 pdf page 164 Drawing 1-47E852-1-SLR, locations A3 and A6, shows piping segments out of scope, but these segments are connected to a piping Discuss the presented boundary configuration.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions segment that is within scope. Drawing 2-47E852-1-SLR, locations A3 and A6, shows piping segments out of scope, but these segments are connected to a piping segment that is within scope. Drawing 3-47E852-1-SLR, locations A3 and A6, shows piping segments out of scope, but these segments are connected to a piping segment that is within scope.

9 2.3.4.1 pdf page 187 Drawings 1-47E801-2-SLR, 2-47E801-2-SLR, 3-47E801-2-SLR identify the boundary between within scope and out of scope sometimes occur in the middle of pipe segments and/or without isolation capability.

Discuss how these boundaries and the methodology are used to identify them.

10 2.3.4.2 pdf page 191 Drawings 0-47E856-1-SLR and 0-47E830-2-SLR identify the boundary between within scope and out of scope sometimes occur in the middle of pipe segments and/or without isolation capability.

Discuss how these boundaries and the methodology are used to identify them.

2.2 Plant Level Scoping Results 2.3.3.7 High Pressure Fire Protection (Diesel Driven Pump) System 2.3.3.13 CO2 Storage, Fire Protection, Purge System 2.4 Scoping and Screening Results: Structures 2.4.36 Structural Commodities Question Number SLRA Section SLRA Page Background / Issue (As applicable/needed)

Discussion Question / Request 1

2.2 2-30 SLRA Section 2.2, Plant Level Scoping Results, provides the scoping and screening results of structures and commodities within the scope of subsequent license renewal and subject to an AMR. Further, SLRA Table 2.2-1 provides the results of scoping and screening of structures and commodities. However, scoping and screening results do not include diesel fire pump building, within the scope of subsequent license renewal and subject to an AMR.

Fire protection licensing basis documents, Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Updated Final Safety Analysis Report (FSAR) Amendment 30, ML23278A013 and Browns Ferry Nuclear Plant, Units 1, 2, and 3, NFPA 805 Safety Evaluation Report ML15212A796, includes Verify whether the diesel fire pump building structure is within the scope of subsequent license renewal in accordance with 10 CFR 54.4(a) and whether the building structure is subject to an AMR in accordance with 10 CFR 54.21(a)(1). If the diesel fire pump building is not within the scope of license renewal and is not subject to an aging management review (AMR), the staff requests that the applicant justify the exclusion.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions diesel fire pump building in power block structures that have the equipment required for nuclear plant operations.

2 Section 2.3.3.7 2-88 Table 2.3.3-7 of the SLRA does not include the following fire protection components in the scope of subsequent license renewal and subject to an AMR:

Passive components in incipient fire detection system installed in Relay and Upper Cable Spreading Rooms Halon bottles Flame arrestor Spray nozzle Sight glass and body Filter housing Stainer housing and element Orifice Hose reel (fitting), fire hose connections, hose racks flexible hose standpipe risers Pump casing (jockey and main fire pump)

Aqueous foam system tank Diesel fire pump fuel oil tank Intake traveling screen/trash rack Floor drains for removal of fire-fighting water Curbs and dike for oil spill confinement Station transformer fire suppression system and components Valve body Seismic support for standpipes system piping Passive components in aqueous foam system Passive components in diesel driven fire pump engine, heat exchanger channel, Muffler/exhaust silencer shell, tube, and jacket water Verify whether the listed system/components are within the scope of subsequent license renewal in accordance with 10 CFR 54.4(a) and whether they are subject to an AMR in accordance with 10 CFR 54.21(a)(1). If they are not within the scope of license renewal and are not subject to an AMR, the staff requests that the applicant justify the exclusion.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions 3

Section 2.3.3.13 2-101 Table 2.3.3-13 of the SLRA does not include the following fire protection components in the scope of subsequent license renewal and subject to an AMR:

Odorizer Orifice Spray nozzle Verify whether the fire protection components listed are within the scope of subsequent license renewal in accordance with 10 CFR 54.4(a) and whether they are subject to an AMR in accordance with 10 CFR 54.21(a)(1). If they are not within the scope of license renewal and are not subject to an AMR, the staff requests that the applicant justify the exclusion.

4 2.4.36 2-216 SLRA Table 2.4-36 of the SLRA does not include the following fire barriers.

Fire wrap/electric raceway fire barrier system (ERFBS)

Radiant energy shield Seismic gap cover Verify whether the passive fire protection features/fire barriers listed are within the scope of subsequent license renewal in accordance with 10 CFR 54.4(a) and whether they are subject to an aging management review (AMR) in accordance with 10 CFR 54.21(a)(1). If they are not within the scope of subsequent license renewal and are not subject to an AMR, the staff requests that the applicant provide justification for the exclusion.

5 2.4 2-160 SLRA Section 2.4, Scoping and Screening Results:

Structures, provides the scoping and screening results of various structures within the scope of subsequent license renewal and subject to an AMR. Further, SLRA Section 2.4 does not provide the results of scoping and screening of fire barriers in power block structures listed in Browns Ferry Nuclear Plant, Units 1, 2, and 3, NFPA 805 license amendment request ML13092A392, Attachment I: Unit 1 Reactor Building, Unit 2 Reactor Building, Unit 3 Reactor Building, Control Building, Turbine Building, Radwaste Building, Unit1/2 Diesel Generator Building, Unit 3 Diesel Generator Building, Intake Pumping Station, Off-Gas Building & Off-Gas Stack, Standby Gas Treatment Building, Cooling Towers & Channel Diesel Fire Pump The staff requested that the applicant provide details of the in-scope power block building structure of the plant in accordance with 10 CFR 54.4(a), and subject to an AMR, in accordance with 10 CFR 54.21(a)(1). For example, the staff requested that the applicant provide a list of power block buildings within the scope of license renewal with passive fire protection features.

If any of the power block building passive fire protection features are not within the scope of subsequent license renewal and are not subject to an AMR, the staff

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application (SLRA) Breakout Audit Questions Building, and Reactor Building Air Intake Plenum - Units 1, 2, 3.

requests that the applicant provide justification for the exclusion.