RS-14-165, Responses to NRC Requests for Additional Information, Set 24, Dated May 19, 2014, Related to the Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, License Renewal Application

Responses to NRC Requests for Additional Information, Set 24, Dated May 19, 2014, Related to the Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, License Renewal Application

ML14160A871
Person / Time
Site:	Byron, Braidwood
Issue date:	06/09/2014
From:	Gallagher M Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-14-165
Download: ML14160A871 (59)

Text

10 CFR 50 10 CFR 51 10 CFR 54 RS-14-165 June 9, 2014 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

Subject:

Responses to NRC Requests for Additional Information, Set 24, dated May 19, 2014, related to the Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, License Renewal Application

References:

1. Letter from Michael P. Gallagher, Exelon Generation Company LLC (Exelon) to NRC Document Control Desk, dated May 29, 2013, "Application for Renewed Operating Licenses"

2. Letter from Lindsay R Robinson, US NRC to Michael P. Gallagher, Exelon, dated May 19, 2014, "Request for Additional Information for the Review of the Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, License Renewal Application, Set 24 (TAC NOS. MF1879, MF1880, MF1881, and MF1882)"

In Reference 1, Exelon Generation Company, LLC (Exelon) submitted the License Renewal Application (LRA) for the Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 (BBS). In Reference 2, the NRC requested additional information to support staff review of the LRA.

Enclosure A contains the responses to these requests for additional information.

Enclosure B contains updates to sections of the LRA affected by the response.

There are no new or revised regulatory commitments contained in this letter.

June 9, 2014 U.S. Nuclear Regulatory Commission Page 2 If you have any questions, please contact Mr. Al Fulvio, Manager, Exelon License Renewal, at 610-765-5936.

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

Executed on Respectfully,

~---

Vice President - License Renewal Projects Exelon Generation Company, LLC

Enclosures:

A Responses to Requests for Additional Information B. Updates to affected LRA sections cc: Regional Administrator - NRC Region Ill NRC Project Manager (Safety Review), NRR-DLR NRC Project Manager (Environmental Review), NRR-DLR NRC Senior Resident Inspector, Braidwood Station NRC Senior Resident Inspector, Byron Station NRC Project Manager, NRR-DORL-Braidwood and Byron Stations Illinois Emergency Management Agency - Division of Nuclear Safety

RS-14-165 Enclosure A Page 1 of 23 Enclosure A Byron and Braidwood Stations (BBS), Units 1 and 2 License Renewal Application Responses to Requests for Additional Information RAI 3.5.2.2.1-1 RAI 3.5.2-6 RAI B.2.1.12-1a RAI B.2.1.3-4 RAI B.2.1.24-1 RAI 3.5.1-1

RS-14-165 Enclosure A Page 2 of 23 RAI 3.5.2.2.1-1 Applicability:

Byron Station (Byron) and Braidwood Station (Braidwood), all units.

Background:

Subsequent to the issuance of Revision 2 of the Generic Aging Lessons Learned (GALL) Report (December 2010), NRC Information Notice (IN) 2011-20, Concrete Degradation by Alkali-Silica Reaction (ASR), was issued on November 01, 2011 to inform industry of operating experience related to concrete degradation due to ASR in seismic category 1 structures at Seabrook Station. IN 2011-20 states that American Society for Testing and Materials (ASTM) updated standards ASTM C1260 and ASTM C1293 and guidance provided in the appendices of ASTM C289 and ASTM C1293 caution that the tests described in ASTM C227 and ASTM C289 may not accurately predict aggregate reactivity when dealing with late-expanding or slow-expanding aggregates containing strained quartz or microcrystalline quartz. Therefore, licensees that tested using ASTM C227 and ASTM C289 could have concrete that is susceptible to ASR-induced degradation. Further, presence of ASR-induced degradation can be positively confirmed or refuted only by optical microscopy performed as part of petrographic examination of concrete core samples. As stated in NRC Inspection Report 05000443/2012010, dated August 9, 2013 (ADAMS Accession No. ML13221A172), the two root causes related to the occurrence of ASR degradation at Seabrook Station were: (1) the concrete mix design unknowingly utilized a coarse aggregate that would, in the long term, contribute to ASR and (2) the long-standing organizational belief that ASR was not a credible degradation mechanism due to the concrete mix design meeting industry standards and reactivity testing at the time of construction.

Section 3.5 in NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (SRP-LR), includes several subsections (e.g.,

3.5.2.2.1.8, 3.5.2.2.2.1.2, 3.5.2.2.2.3.2) which identify the aging effect and mechanism of concrete cracking due to expansion from reaction with aggregates that do not require additional plant-specific aging management for inaccessible concrete areas if certain conditions identified in the GALL Report can be met. Table 3.5-1 in the SRP-LR includes line items for this aging effect and mechanism for accessible concrete areas in any environment and recommends the GALL Report aging management programs (AMPs) (e.g., XI.S2, ASME Section XI, Subsection IWL, and XI.S6, Structures Monitoring) to manage the effects of aging, with no associated conditions that can be met to consider the aging effect not applicable. The parameters monitored program element of the ASME Section XI, Subsection IWL and the Structures Monitoring AMPs in the GALL Report include this aging effect and mechanism. The NRC staff expects this aging effect and mechanism to be included within the recommended structural AMP.

Issue:

License renewal application (LRA) Table 3.5.1, items 3.5.1-12, 3.5.1-19, 3.5.1-43, 3.5.1-50, and 3.5.1-54; and LRA Sections 3.5.2.2.1.8, 3.5.2.2.2.1.2, and 3.5.2.2.2.3.2, address cracking from expansion due to reaction with aggregates in concrete elements in accessible and inaccessible areas. The Discussion column of the aging management review (AMR) line items noted above states that this aging effect and mechanism does not apply to the respective Byron and

RS-14-165 Enclosure A Page 3 of 23 Braidwood concrete structures because the fine and coarse aggregates used conform to ASTM C33, petrographic examinations of aggregates were performed in accordance with ASTM C295 and ASTM C289, and the concrete structures were constructed in accordance with American Concrete Institute (ACI) 318. The staff does not agree, since adequate plant-specific technical basis to support that statement has not been provided and the tests described in ASTM C227 and ASTM C289 may not accurately predict aggregate reactivity when dealing with late-expanding or slow-expanding aggregates containing strained quartz or microcrystalline quartz. In light of the industry operating experience at Seabrook Station, unless positively justified, the staff position is that cracking due to expansion from reaction with aggregates for concrete in accessible and inaccessible areas could occur and should be managed through the period of extended operation. The discussion in the LRA sections referenced above states that although cracking associated with expansion due to reaction with aggregates has not been observed on Byron and Braidwood concrete structures, the respective structural AMPs mentioned, therein, will continue to inspect and monitor concrete structures for cracking due to any mechanism. However, the associated line items for concrete cracking from expansion due to reaction with aggregates do not appear in any of the LRA Table 2s.

Request:

Provide technical justification why cracking due to expansion from reaction with aggregates (i.e., alkali-aggregate reaction) does not require management for concrete in accessible and inaccessible areas or identify applicable program(s) to manage this aging effect. If a program is identified to manage this aging effect, update the applicable LRA sections accordingly.

Exelon Response:

The aging effect of cracking of reinforced concrete is managed by the ASME Section XI, Subsection IWL (B.2.1.30) program, the Structures Monitoring (B.2.1.34) program, and Regulatory Guide 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35) aging management program. This aging effect is managed regardless of the mechanism that caused the aging effect. Therefore, the applicable aging management programs also address cracking due to reaction with aggregates. The information in Section 3.5 of the LRA was prepared to provide information on why this particular mechanism will not cause cracking of reinforced concrete at BBS. In order to clarify that the aging effect and mechanism, cracking due to reaction with aggregates, are included in the applicable aging management programs, the changes below are made to the LRA regarding concrete areas.

LRA Table 3.5.1, Item 19 discusses the aging effect, "Cracking due to expansion from reaction with aggregates" for Component "Concrete (accessible areas): dome; wall; basemat; ring girders; buttresses, Concrete (accessible areas): basemat, Concrete (accessible areas):

containment; wall; basemat, Concrete (accessible areas): basemat, concrete fill-in annulus."

The aging effect, "Cracking due to expansion from reaction with aggregates" for this component in the Containment Structure at Byron and Braidwood will be managed by the ASME Section XI, Subsection IWL (B.2.1.30) aging management program.

LRA changes include revision to Table 3.5.1, Item 19, and revision to LRA Table 3.5.2-4 for the Containment Structure.

RS-14-165 Enclosure A Page 4 of 23 LRA changes also include revision to Section 3.5.2.2.1.8, and Table 3.5.1, Item 12 to address the inaccessible areas for this Component in the Containment Structure. These changes are to ensure LRA consistency on the applicability of the aging effect.

LRA Table 3.5.1, Item 54 discusses the aging effect, "Cracking due to expansion from reaction with aggregates" for component "All groups except 6: concrete (accessible areas): all. "The aging effect, "Cracking due to expansion from reaction with aggregates" for this component at Byron and Braidwood will be managed by the Structures Monitoring (B.2.1.34) aging management program.

In addition, accessible group 6 concrete is also monitored for cracking due to expansion from reaction with aggregates under LRA Table 3.5.1, Item 54, as there is no associated GALL line item for accessible group 6 concrete. The Regulatory Guide 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35) program is substituted for the Structures Monitoring (B.2.1.34) aging management program for the accessible group 6 concrete.

LRA changes include revision to Table 3.5.1, Item 54, and revision to LRA Tables 3.5.2-1, 3.5.2-2, 3.5.2-4, 3.5.2-5, 3.5.2-7 through 3.5.2-14, and 3.5.2-16 through 3.5.2-18 for the structures at Byron and Braidwood.

LRA changes also include revision to Section 3.5.2.2.2.1 Item 2, Section 3.5.2.2.2.3 Item 2, and Table 3.5.1, Items 43 and 50, to address the inaccessible areas for this Component for the structures at Byron and Braidwood. These changes are to ensure LRA consistency on the applicability of the aging effect.

The LRA changes discussed above are provided in Enclosure B of this submittal.

RS-14-165 Enclosure A Page 5 of 23 RAI 3.5.2-6 Applicability:

Byron and Braidwood

Background:

Item 24 located in SRP-LR Table 3.5-1 references the GALL Report item II.A1.CP-100. The AMP recommended for item II.A1.CP-100 in the GALL Report is XI.S2, ASME Section XI, Subsection IWL, or XI.S6, Structures Monitoring.

The GALL Report AMP XI.S2, ASME Section XI, Subsection IWL, Program Description states that 10 CFR 50.55a imposes the examination requirements of ASME Code,Section XI, Subsection IWL, for Class CC reinforced and prestressed concrete containments. The GALL Report AMP Scope of Program (Program Element 1) states that the components within the scope of Subsection IWL are reinforced concrete and unbonded post-tensioning systems of Class CC containments. Subsection IWL exempts from examination portions of the concrete containment that are inaccessible such as concrete covered by liner, foundation material, or backfill or obstructed by adjacent structures or other components. However, 10 CFR 50.55a(b)(2)(viii) specifies additional requirements for inaccessible areas that requires the licensee to evaluate the acceptability of concrete in inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation of inaccessible areas.

Issue:

The corresponding LRA Table 3.5.1, Item 3.5.1-24, states in the Discussion column that this item is consistent with the GALL Report and cites the Structures Monitoring (LRA Section B.2.1.34) program as the AMP for managing this aging effect and mechanism for inaccessible areas of containment concrete exposed to groundwater and soil environments, including groundwater chemistry. In the context of the LRA, the description for Item II.A1.CP-100 in the GALL Report does include containment pressure-resisting boundary concrete components in inaccessible areas above grade in an Air - Outdoor environment as well as below-grade areas in a Groundwater/Soil environment. The applicant has not addressed inaccessible components in the Air - Outdoor environment for this AMR line item. Further, the ASME Section XI, Subsection IWL program, mandated by the GALL Report and 10 CFR 50.55a for concrete containment pressure-resting boundary components in both accessible and inaccessible areas, is not included as an applicable AMP for AMR line item 3.5.1-24 and the corresponding line items in LRA Table 3.5.2-4.

Request:

With regard to AMR line Item 3.5.1-24 in LRA Table 3.5-1 that corresponds to Item II.A1.CP-100 in the GALL Report, provide the technical basis to justify why the ASME Code,Section XI, Subsection IWL program, recommended by the GALL Report and required by 10 CFR 50.55a for concrete containment pressure-resisting boundary components in both accessible and inaccessible areas, is not included as an applicable AMP for the line item and corresponding line items in LRA Table 3.5.2-4. Update the LRA, as necessary, based on the response to this request.

RS-14-165 Enclosure A Page 6 of 23 Exelon Response:

ASME Section XI, Subsection IWL program was not listed for aging management of concrete containment pressure-resisting boundary components in accessible and inaccessible areas for AMR line item 3.5.1-24 in LRA Table 3.5-1 due to a different interpretation of the intent of the GALL line item, which specifies the use of either Chapter XI.S2, ASME Section XI, Subsection IWL or Chapter XI.S6, "Structure Monitoring as the aging management program.

ASME Section XI, Subsection IWL program will be used for aging management of accessible and inaccessible concrete containment pressure-resisting boundary component areas for AMR line item 3.5.1-24 in LRA Table 3.5.1.

LRA changes include revision to Table 3.5.1, Item 3.5.1-24, and revision to Table 3.5.2-4, which are provided in Enclosure B of this submittal.

RS-14-165 Enclosure A Page 7 of 23 RAI B.2.1.12-1a Applicability:

Byron and Braidwood

Background:

The response to RAI B.2.1.12-1, dated February 27, 2014, stated that existing station procedures require a general visual inspection of internal surfaces of components within the scope of the Closed Treated Water Systems program when the systems are opened. In addition, the personnel performing the inspections are qualified to Exelon job qualifications and in accordance with the Institute of Nuclear Power Operations (INPO) National Academy for Nuclear Training accredited training program.

The staff notes that, similar to the Closed Treated Water Systems program, the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program uses opportunistic visual inspections to monitor aging effects of component internal surfaces. However, during its AMP audit, the staff noted that the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program does not rely on the use of existing station procedures to identify age-related degradation. Rather, a new procedure was proposed to inspect for evidence of loss of material, leakage, cracking, and reduction of heat transfer when the internal surfaces of metallic components were made accessible.

Issue:

It is unclear to the staff how the existing station opportunistic inspections will be capable of detecting the specific applicable aging effects of components internal surfaces in the Closed Treated Water Systems program. The RAI response did not provide sufficient information regarding:

1. The details within the INPO training program and Exelon job qualifications that demonstrate that personnel performing the opportunistic inspections are qualified to identify the applicable aging effects, and

2. The details within the INPO training program, Exelon job qualifications, or the existing station procedures that demonstrate that, when piping internal surfaces are made accessible, personnel will be inspecting for parameters that are capable of detecting the presence and extent of aging effects.

Request:

1. State the details within the INPO training program and Exelon job qualifications that demonstrate that personnel performing the opportunistic inspections are qualified to identify loss of material due to general, pitting, crevice, and galvanic corrosion; and cracking due to stress corrosion cracking.

2. State the details within the INPO training program and Exelon job qualification that will ensure that, when component internal surfaces are made accessible, personnel will be inspecting for parameters that are capable of detecting the presence and extent of loss of

RS-14-165 Enclosure A Page 8 of 23 material due to general, pitting, crevice, and galvanic corrosion; and cracking due to stress corrosion cracking. Alternatively, state the process-based controls in existing station procedures (e.g., prejob brief details, checklists within the work order) that will ensure personnel will be inspecting for the appropriate parameters.

Exelon Response:

1. The Closed Treated Water Systems (B.2.1.12) aging management program includes opportunistic inspections, as well as periodic visual inspections and non-destructive examinations of the internal surfaces of components within the scope of the program.

These inspections verify the effectiveness of the water chemistry control of the closed-cooling water systems. Station maintenance personnel who will perform opportunistic visual inspections are trained on the various methods of corrosion control in the closed-cooling water systems, and are knowledgeable about the expected conditions of the piping and components.

In accordance with the Institute of Nuclear Power Operations (INPO) National Academy for Nuclear Training accredited training program, maintenance personnel receive training on fundamentals, general systems and components, and applicable on-the-job training. The mechanical technicians receive an introduction to the various aspects of material science.

To ensure personnel are familiar with and capable of detecting various forms of age-related degradation, the program includes computer-based training modules on the following topics:

introduction to plant materials, metals and alloys, basic properties of metals, manufacturing and fabrication processes, typical application, common failure mechanisms, identifying coatings and linings, and non-destructive examination (NDE) techniques. The modules include color photographs of various corrosion types, which could be encountered on the plant components and systems, as well as the limitations of the various examination methods. Familiarization with the photographs of various corrosion types ensures that the qualified personnel performing the opportunistic examinations are capable of distinguishing the different types of corrosion and determining the relative condition of the equipment.

These qualified personnel can identify loss of material due to general, pitting, crevice, and galvanic corrosion; and cracking due to stress corrosion cracking during the opportunistic examinations of the closed-cooling water systems. Therefore, there is reasonable assurance that the maintenance personnel will be able to identify the applicable aging effects during the opportunistic inspections of the closed-cooling water systems.

2. Exelon procedures require maintenance personnel to enter any inspection results that reveal more than the expected amount of age-related degradation in the corrective action program. In addition, when piping or components are open for inspection and the condition is not as expected, the first-line supervisor or engineer is notified of the abnormal condition.

Due to the chemistry controls utilized for the closed-cooling water systems managed by the Closed Treated Water Systems (B.2.1.12) aging management program, no age-related degradation is expected on the piping and components. Therefore, any detectable loss of material or cracking identified by maintenance personnel during these opportunistic visual inspections will be entered, evaluated, and documented in the corrective action program.

The 10 CFR Part 50 Appendix B corrective action program ensures that appropriate corrective actions, including evaluation of the condition by engineering personnel and/or additional inspections to determine the extent of degradation, are performed, as applicable.

Additional inspections to determine the extent of degradation will be performed utilizing the

RS-14-165 Enclosure A Page 9 of 23 same procedural guidance as the new periodic inspections required by the first enhancement of the Closed Treated Water Systems (B.2.1.12) aging management program.

This includes the use of ASME Code requirements, where applicable. For non-ASME Code components, plant-specific inspection procedures are used that are capable of detecting the loss of material, cracking, and reduction of heat transfer. Since there is sufficient procedural guidance to ensure that the appropriate parameters are examined on the piping and components, there is reasonable assurance that age-related degradation will be identified and appropriate corrective actions will be taken prior to loss of component intended function.

RS-14-165 Enclosure A Page 10 of 23 RAI B.2.1.3-4 Applicability:

Braidwood

Background:

During the audit of the operating experience program element for Braidwood Units 1 and 2, the staff found that operating experience provided by the applicant in the LRA was incomplete.

Specifically, the applicants onsite database contained information related to a stuck reactor vessel closure stud at the Braidwood Unit 2. Based on the information provided by the applicant during the audit, Stud No. 35 became stuck during the 1991 outage, but it had enough thread engagement to be tensioned. The applicant was able to tension the stuck stud. The stuck stud was cut at the flange level in May of 1995 in order to facilitate safer fuel transfer during refueling outages. In an effort to repair Stud Hole No. 35, the remnant of the stud was bored out in 2002.

Due to human error, the stud bore hole was over bored, and the applicant decided to abandon its repair efforts. Currently, Braidwood Unit 2 has only 53 of 54 studs operable.

Issue:

The LRA does not provide any information regarding the significant plant-specific operating experience relative to Stud No. 35 for Braidwood Unit 2. In addition, no information was provided in the LRA or during the audit on the root cause of the failure. Without a root cause, the staff is concerned that similar failures could reoccur and further challenge the integrity of the reactor vessel head.

Request:

1. Perform a comprehensive plant-specific operating experience search for Braidwood Units 1 and 2. In addition to Stud No. 35, provide search results that include all instances of stuck studs, missing threads, damaged threads, or any form of degradation in reactor pressure vessel studs, guide studs, washers, vessel flange threads, and nuts.

2. Provide a detailed chronology of the events related to Braidwood Unit 2, Stud No. 35.

3. Provide a root cause analysis related to the failure of Stud No. 35. Include corrective actions, inspection results, engineering changes, repair replacement activities related to Stud No. 35 and its respective flange hole.

4. Provide details of the current configuration of Stud Hole No. 35 and inspection results from 2002 to present.

5. Provide inspection results for Stud and Stud Hole Nos. 33, 34, 36 and 37 for Braidwood Unit No. 2 from 1995 to present.

RS-14-165 Enclosure A Page 11 of 23 Exelon Response:

1. A thorough operating experience review has been performed as requested. The operating experience review involved key word searches of the Braidwood Station Passport Action Request (AR) database, Exelons Electronic Document Management System (EDMS) regulatory correspondence database, and the NRCs LER databases.

In addition, Braidwood Unit 1 and 2 ISI Inspection Summary Reports and applicable inspection reports were reviewed.

Below are summaries of applicable events and conditions (i.e., instances of stuck studs, missing threads, damaged threads, or any form of degradation in reactor pressure vessel (RPV) head studs, guide studs, washers, vessel flange threads, and nuts at Braidwood Unit 1 and Unit 2); except for operating experience results associated with the Braidwood Unit 2 stud 35 which is discussed in response to Request 4 below. The operating experience review also identified events involving minor degradation of RPV flange O-rings or O-ring seating surfaces. These events are summarized below in chronological order.

During the fall 2006 refueling outage, prior to commencing the Braidwood Unit 2 RPV reassembly, site personnel inspected newly installed RPV flange O-rings. The examination found scratches across the inner O-ring pressure retaining surface face and numerous other small blemishes and scratches on the pressure retaining faces of both the inner and outer O-rings. The O-rings were damaged during refueling outage activities while installing a radiation barrier in the reactor vessel. This had been the first time this particular radiation barrier had been installed. The damaged O-rings were replaced prior to RPV head reassembly. Corrective actions included procedures changes to minimize the potential for O-ring damage during future installation of radiation barriers.

During the fall 2010 Braidwood Unit 1 refueling outage, an area of boric acid accumulation was observed on the RPV head between the inner and outer O-rings near studs 48 and 49. Inspection of the RPV head revealed small indications in the vicinity of RPV head studs 40 through 50, which had previously been reported in 2006, 2007, and 2009 and were dispositioned as meeting acceptance criteria. There was no evidence of steam cutting across the RPV head inner O-ring sealing surfaces, which would have explained the area of boric acid accumulation between the inner and outer O-rings near studs 48 and 49. However, after cleaning, inspection of the RPV flange O-ring seating surface in the area near studs 48 and 49 revealed a horizontal steam cut of 0.003 inches laterally across the RPV flange inner O-ring seating surface. The size of the steam cut exceeded acceptance criteria. The inspection showed no steam cut indications across the RPV flange outer O-ring seating surface. The steam cut area on the RPV flange inner O-ring seating surface was repaired and returned to original surface conditions. No other flaws were detected on the flange surface. No reactor coolant leakage has been observed through the RPV flange inner O-ring at Braidwood Unit 1 before or after this event in 2010.

An apparent root cause evaluation concluded that this event was attributed to less than adequate RPV flange monitoring practices. Although the RPV head surfaces

RS-14-165 Enclosure A Page 12 of 23 were inspected in 2006, 2007, and 2009; the RPV flange surfaces were not thoroughly inspected. As a result, the RPV flange flaw may have gone unaddressed in previous refueling outages, and was allowed to grow to a point that the sealing capability of the inner O-ring was impacted. The apparent root cause evaluation concluded that this event could have been prevented by performing more thorough RPV flange inspections during previous outages. As a result, site procedures were revised to perform a visual inspection of the RPV flange surface immediately following the removal of the RPV head and prior to O-ring removal to determine proper sealing of the RPV head O-rings.

During RPV disassembly in the spring 2011 Braidwood Unit 2 refueling outage, the installation of the two (2) RPV head temporary guide studs at stud locations 12 and 44 took longer than expected. Inspection of the guide studs after they were removed from the RPV flange revealed minor wear on the male guide stud threads. The minor wear was removed by cleaning the threads and removing burrs and imperfections.

Review of the completed work orders from the subsequent refueling outage showed:

no documented wear or damage to the female threads of the associated RPV flange stud holes at stud locations 12 and 44; no documented difficulty during the installation of the RPV head temporary guide studs at stud locations 12 and 44; and no documented difficulty during removal and installation of RPV head studs 12 and 44.

2. Braidwood Unit 2 began commercial operation in the fall of 1988. In 1991, during the second Braidwood Unit 2 refueling outage, RPV head closure stud 35 (stud 35) became stuck during RPV disassembly. Attempts to remove the stud included soaking the stud with an approved lubricant. However, the stud could not be removed without excessive or destructive methods. Since the stud was only withdrawn 15/32 inches (4 turns), it was decided to leave the stud in place during the refueling outage and protect it from borated water when the reactor cavity was flooded. An engineering change was developed in 1991 which justified operating Braidwood Unit 2 with stud 35 tensioned and withdrawn 15/32 of an inch from the RPV flange. From the fall of 1991 until the spring of 1994, Braidwood Unit 2 stud 35 was tensioned during plant operation and remained in the RPV flange during refueling outages. Throughout this time period, stud 35 and the associated stud hole were protected from borated water during refueling outages.

However, the protruding portion of stud 35 was an obstacle during refueling outage activities. Therefore, in the spring of 1994 an evaluation was developed that demonstrated the Braidwood Unit 2 RPV could be placed in service without stud 35 tensioned. The evaluation concluded that the increased stresses and flange separation in the area corresponding to stud 35 and the adjacent studs were not significant, and the O-ring configuration would ensure the RPV flange remain sealed during reactor operation. The evaluation confirmed that without stud 35 tensioned the structural integrity of the RPV satisfied the 1971 Edition of ASME Section III, with addenda, through summer 1973. An engineering change was performed authorizing a new configuration without RPV stud 35 tensioned during power operation. In the spring 1994 refueling outage, the portion of stud 35 that protruded above the RPV flange was removed and Braidwood Unit 2 started up from the refueling outage with 53 studs tensioned. UFSAR Section 5.3.1.7 and Table 5.3-2 were updated to reflect the new configuration, with the periodic UFSAR update submitted to the NRC in accordance with 10 CFR 50.71(e).

RS-14-165 Enclosure A Page 13 of 23 Braidwood then developed plans to restore the capability of stud 35. The plans included destructively removing the remaining portion of stud 35 and a contingency modification in case the RPV flange threads were damaged to the extent in which the threads could not be reused. The contingency modification would require the installation of a larger diameter sleeve in the RPV vessel flange hole. The outer male threads of the sleeve would thread into new female threads that would be machined into the new larger RPV flange hole. A new stud would then be threaded into the inner female threads of the sleeve. During the 2002 refueling outage the plan was implemented and the remaining portion of stud 35 was destructively removed from the RPV flange hole. Inspection of the RPV flange hole threads showed significant damage and it was concluded that the RPV flange hole could not be reused as found. Therefore, the site commenced the contingency modification, which first required boring a larger hole in the RPV flange hole and then machining new threads in the RPV flange hole. However the vendors equipment malfunctioned and, as a result, the station decided not to continue the repair in the refueling outage and to continue operating Braidwood Unit 2 with 53 studs tensioned during operation as previously evaluated. An engineering change was performed authorizing the new configuration of the RPV flange hole in stud location 35.

Braidwood Unit 2 has operated in this configuration from that time to the present. No reactor coolant leakage has been observed through the RPV flange at Braidwood Unit 2.

As documented in response to RAI B.2.1.3-2 in letter Exelon RS-13-285, dated December 19, 2013, Exelon has made the following commitment:

Braidwood Unit 2 reactor head closure stud location 35 will be repaired so that all 54 reactor head closure studs are tensioned during the period of extended operation.

This commitment will be implemented no later than 6 months prior to the period of extended operation.

In August 2013 a non-conservative input was identified involving Westinghouse WCAP-16143-P, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2, approved in 2003, which justified removing the 10 CFR 50, Appendix G flange requirements when determining reactor pressure-temperature limits. The technical basis documented in this report assumed 54 reactor head closure studs were in service for Braidwood Units 1 and 2. In 2006, the NRC approved use of WCAP-16143-P to support the exemption. The NRC subsequently approved a license amendment to implement the Pressure Temperature Limit Report (PTLR) using the methodology in WCAP-16143-P as one of the basis documents for the current PTLR reports for BBS Units 1 and 2. WCAP-16143-P is listed in Technical Specification 5.6.6 as one of the analytical methods previously approved for use to determine the RCS pressure and temperature limits. Given that the P-T limits minimum temperature requirement methodology in WCAP-16143-P was not based on the configurations of RPV closure flange assemblies at Braidwood Unit 2, the issue was entered into the corrective action program. Corrective actions, which includes the revision of WCAP-16143-P to reflect the Braidwood Unit 2 configuration of 53 RPV head bolts, have been identified in an Exelon letter to the NRC dated December 13, 2013. The revision of WCAP-16143-P will bring the methodology in agreement with the current configuration.

RS-14-165 Enclosure A Page 14 of 23 In October 2013, a non-conservative input related to the original calculation that justified operating Braidwood Unit 2 with 53 tensioned reactor head closure studs was identified and entered into the correct action program. The calculation incorrectly used a larger washer bearing surface area between the RPV closure stud washers and RPV head.

This resulted in lower calculated applied stresses that met ASME stress allowables. If the calculation had used the correct washer bearing surface area, the resulting calculated bearing stress on the RPV head flange increases to values slightly greater than ASME allowables. Short term corrective actions included the development and approval of an operability evaluation that demonstrated that the stresses on the RPV flange head were acceptable based on the as-tested mechanical properties for the RPV head material. Long term corrective actions included the revision of RPV assembly and disassembly procedures to provide revised stud elongation limits so that applied bearing stresses under the washers remain below ASME allowables, while ensuring that the RPV O-ring configuration would remain sealed during reactor operation. An engineering change, which authorized the new stud elongation limits, was prepared and approved by Braidwood Engineering.

3. Based on the review of the Braidwood Unit 2 operating experience described above, RPV disassembly and assembly procedures, and personnel interviews, the most likely potential causes of stud 35 becoming stuck in 1991 was: 1) undetected mechanical damage or galled threads during the handling process, 2) undetected improper thread lubrication during installation of the stud during the previous refueling outage or, 3) the introduction of undetected foreign material in the RPV flange hole during the previous refueling outage. The review did not reveal any evidence that stud 35 became stuck due to age-related degradation.

EPRI bolting good practices and guidelines indicate that the credible potential causes for a stuck RPV stud are: (a) flange to bolt misalignment, (b) foreign material, (c) improper or no thread lubrication, (d) damaged or galled threads, (e) corrosion byproduct buildup on the stud or flange threads, and (f) stud to RPV flange hole cross threading. Corrosion byproduct buildup, item e, is the only one of the above potential causes that is age-related. The following evaluation considers each of these potential causes in turn, and explains why it is not credible that Braidwood Unit 2 stud 35 became stuck due to age-related degradation.

(a) Flange to bolt misalignment: The RPV assembly procedure provides specific guidance to prevent RPV head closure stud to flange misalignment by tensioning the RPV head closure studs in a proceduralized pattern and sequence. The documented as found configuration of stud 35 and the RPV flange in 1991 provided no evidence of misalignment. Therefore, it is concluded that it is highly unlikely that stud 35 became stuck due to misalignment during installation in the previous refueling outage.

(b) Foreign material: The RPV disassembly and assembly procedures provide specific guidance to minimize the likelihood of foreign material entering the RPV flange hole or RPV head closure stud threads. This includes requiring that each stud and vessel flange hole are inspected and cleaned and that each vessel flange hole is plugged after removal of the stud and prior to flood up of the reactor cavity. In addition, after the reactor cavity is drained, borated water is removed from the stud plug relief areas and the areas are cleaned. The plugs are then removed and the vessel flange holes

RS-14-165 Enclosure A Page 15 of 23 are again inspected. If any borated water has leaked past the plugs, the borated water is removed and the holes are cleaned. Although these steps minimize the potential of foreign material intrusion into the stud and vessel flange threads, they may not eliminate the possibility of this occurrence. If foreign material was missed during the inspection and cleaning of stud 35 and the associated vessel flange hole during the previous outage, it is plausible that stud 35 could have been installed without any problems and become stuck only upon removal.

(c) Improper or no thread lubrication: The disassembly and assembly procedures provide specific guidance to ensure proper lubrication of the RPV flange hole and stud threads. Although these steps minimize the potential of damage to the stud and vessel flange threads, they may not eliminate the possibility of this occurrence. It is plausible that stud 35 could have been improperly lubricated, without detection, during the 1989 refueling outage causing the stud to become stuck upon removal during the 1991 refueling outage.

(d) Damaged or galled threads: The RPV disassembly and assembly procedures provide specific guidance to minimize the likelihood of damaged or galled threads.

Although these steps minimize the potential of damage to the stud and vessel flange threads, they may not eliminate the possibility of this occurrence. It is plausible that stud 35 could have been damaged without detection during the 1989 refueling outage or during removal in 1991 causing the stud to become stuck.

(e) Corrosion byproduct buildup on the stud or flange threads: The corrosion byproduct buildup potential cause is age-related and requires long term exposure to a corrosive environment to manifest into a stuck stud. The RPV disassembly and assembly procedures provide specific guidance to prevent corrosion by minimizing exposure of the RPV head closure studs and associated components to a corrosive environment.

This includes plugging the RPV flange stud holes and removing the studs, nuts, and washers from the reactor cavity and storing these components away from the borated water environment. In addition, the procedures require borated water to be removed from the stud plug relief areas and the stud holes to be inspected to ensure that no borated water migrated past the plugs while the reactor cavity is flooded.

Finally, the RPV assembly procedure requires specific stud preload elongation to ensure the RPV vessel to RPV head flange is leak tight, thereby minimizing the potential for exposure of the RPV head stud area to reactor coolant during plant operation. The RPV disassembly and assembly procedures also provide specific guidance to detect corrosion of the RPV head closure studs and RPV flange stud holes. This includes requiring the inspection and cleaning of each RPV head closure stud, nut, washer, and RPV flange stud hole.

The above procedure instructions are intended to prevent long term corrosion byproduct buildup and are performed every outage and therefore, provide multiple opportunities to prevent a stuck stud that could theoretically be caused by corrosion byproduct buildup. Since stud 35 was only in service for three (3) years with only one (1) refueling outage prior to becoming stuck, it is not credible that long term corrosion byproduct buildup occurred. Based on the above, it is highly unlikely that Braidwood Unit 2 stud 35 became stuck due to corrosion byproduct buildup on the stud and flange threads.

RS-14-165 Enclosure A Page 16 of 23 (f) Stud to RPV flange hole cross threading: The RPV disassembly and assembly procedures provide specific guidance to minimize the likelihood of stud to RPV vessel cross threading. Although these steps minimize the potential of damage between the stud and vessel flange threads, they may not eliminate the possibility of this occurrence. However, should cross threading have occurred during RPV assembly in 1989, stud 35 would have become stuck during installation rather than removal in the subsequent outage. It is not plausible that stud 35 could have been cross threaded and installed without any problems during the 1989 refueling outage.

Based on the above, it is highly unlikely that Braidwood Unit 2 stud 35 became stuck due to stud to RPV flange hole cross threading.

The corrosion byproduct buildup is the only potential cause described above which is age-related. This potential cause would require long term exposure to the borated water environment to manifest as a stuck stud. As explained above, it is not credible to conclude stud 35 became stuck due to corrosion byproduct buildup in a three (3) year period with only one (1) refueling outage. An operating experience review of industry PWRs which have also experienced stuck studs revealed no reported instances where the studs became stuck due to age-related corrosion byproduct buildup. Therefore, there is confidence that stud 35 did not become stuck due to age-related degradation.

Most likely, the stud became stuck due to: 1) undetected mechanical damage or galled threads during the handling process during the previous refueling outage, 2) undetected improper thread lubrication during installation of the stud during the previous refueling outage or, 3) the introduction of undetected foreign material in the RPV flange hole during the previous refueling outage, as outlined above.

The RPV disassembly and assembly procedures at Braidwood are periodically revised to ensure that best practices are utilized to eliminate or mitigate the above potential causes. For example, the Braidwood station RPV disassembly and assembly procedures have been revised in excess of 25 times each, since stud 35 became stuck in 1991. These revisions include requiring the use of the Biach electrical stud drive tool (ESDT), which is capable of supporting the weight of the stud during installation and removal, thereby minimizing the stress on the RPV flange and stud threads, and requiring inspection for foreign material that could be concealed under the RPV head prior to installation. Given that Braidwood has removed, handled, stored, and reinstalled individual RPV head closure studs over 1750 times since commercial operation began, one (1) stuck stud, although not desired, provides a positive indication that these procedures have significantly minimized the above potential causes. There is confidence that these procedures provide an effective process to eliminate or mitigate the above potential causes, including those that are aging-related, and that ongoing improvements implemented through the operating experience and corrective action program will ensure that best practices are utilized to minimize the occurrence of stuck studs during the period of extended operation.

RPV pressure tests, in accordance with ASME Section XI, were performed during the refueling outages since the 1994 spring outage, resulting in no observed leakage from the RPV flange.

A formal root cause evaluation of the 1991 refueling outage event has not been performed. A detailed visual inspection of the threads on the stud would have provided important information necessary to determine the root cause but it was not possible to

RS-14-165 Enclosure A Page 17 of 23 perform this inspection in 2002 because the stud was destructively removed and the threads on the stud were damaged by the stud removal process. Nevertheless, there is confidence that the root cause was not age-related degradation and steps have been taken to eliminate or mitigate the likely potential causes.

Engineering changes and repair and replacement activities related to stud 35 and its respective flange hole include:

A 1991 engineering change authorizing operation of Braidwood Unit 2 with stud 35 tensioned and withdrawn 15/32 of an inch from the RPV flange, as discussed in the response to Request 2 above.

A 1994 engineering change authorizing operation of Braidwood Unit 2 without RPV head closure stud 35 tensioned, as discussed in the response to Request 2 above.

A 2002 contingency engineering change that would have required the installation of a larger diameter sleeve in the RPV vessel flange hole, as discussed in the response to Request 2 above. This engineering change was not implemented.

A 2002 engineering change authorizing operation of Braidwood Unit 2 with an enlarged RPV flange hole associated with stud 35, as discussed in the response to Request 2 above.

A 2013 engineering change authorizing the implementation of new RPV stud elongation limits on Braidwood Unit 2 so that applied bearing stresses under the washers remain below ASME allowables, while ensuring that the RPV O-ring configuration would remain sealed during reactor operation, as discussed in the response to Request 2 above.

4. During the 2002 contingency modification the full depth of the RPV flange hole associated with Braidwood Unit 2 RPV stud 35 was enlarged. The original diameter of RPV flange hole was approximately 7 inches with internal threads and was designed to accommodate a 7-inch diameter stud with 7-8N modified threads. The diameter of the top 1.45 inches (from the face of the flange) of the RPV flange hole was enlarged to approximately 8.368 inches to accommodate the top unthreaded portion of the insert.

The diameter of remaining depth of the RPV flange hole (starting approximately 1.45 inches from the face of the flange) was enlarged to 7.610 through 7.615 inches. The depth of the RPV flange hole into the flange RPV is approximately 14.313 inches. New internal threads were not machined into the RPV flange hole. The configuration of RPV flange hole associated with Braidwood Unit 2 RPV stud 35 remains in this configuration today.

During the fall 1997 refueling outage, all RPV flange stud holes, including the one associated with stud 35, were volumetrically examined in accordance with ASME Section XI, Table IWB-2500-1 resulting in no recordable indications. During the fall 2000 refueling outage, all RPV flange stud holes, including the one associated with stud 35, were again volumetrically examined in accordance with ASME Section XI, Table IWB-2500-1 resulting in no recordable indications. These inspections satisfied the ASME

RS-14-165 Enclosure A Page 18 of 23 Section XI, Table IWB-2500-1 requirements for all RPV flange stud holes in the first and second ISI Inspection Ten-Year Intervals. During the spring 2002 refueling outage, after the diameter of the RPV flange hole associated with stud 35 was enlarged, this RPV flange hole was volumetrically examined in accordance with ASME Section XI, Table IWB-2500-1, to ensure the RPV flange ligaments in the vicinity of stud hole 35 were not damaged. The inspection resulted in no recordable indications.

In addition, the RPV flange stud hole 35 is cleaned and inspected prior to reactor vessel flood-up, and reactor vessel flange stud hole 35 is cleaned, inspected, and borated water is removed after the reactor cavity is drained. These specific aging management requirements are clearly identified in the reactor closure head removal and installation procedures and require signoff. These procedures steps apply to each reactor vessel flange hole, including stud hole 35. Review of completed work orders and corrective action program database shows that loss of material, cracking, or foreign material has not been observed for any reactor vessel flange stud hole, including stud location 35, during the two most recent (fall 2012 and spring 2014) refueling outages at Braidwood Unit 2.

5. During the fall 1997 refueling outage, RPV head closure studs and RPV flanges stud holes associated with studs 33, 34, 36, and 37 were examined in accordance with ASME Section XI, Table IWB-2500-1, resulting in no recordable indications. During the fall 2000 refueling outage RPV flange stud holes associated with studs 33, 34, 36, and 37 were again examined in accordance with ASME Section XI, Table IWB-2500-1 resulting in no recordable indications. During the fall 2003 refueling outage RPV head closure studs 33, 34, 36, and 37 were examined in accordance with ASME Section XI, Table IWB-2500-1, resulting in no recordable indications. These inspections satisfied the ASME Section XI, Table IWB-2500-1, requirements for RPV head closure studs and RPV flanges stud holes associated with studs 33, 34, 36, and 37 for the first and second ISI Inspection Ten-Year Intervals. Volumetric examinations in accordance with ASME Section XI, Table IWB-2500-1 of RPV head closure studs associated with studs 33, 34, 36, and 37, for the third ISI Inspection Ten-Year Interval, were performed in the spring 2014 refueling outage resulting in no recordable indications.

RS-14-165 Enclosure A Page 19 of 23 RAI B.2.1.24-1 Applicability:

Braidwood

Background:

In LRA Section B.2.1.24, the applicant provided brief discussions covering the operating experience of the Byron and Braidwood Units 1 and 2. In these discussions, it was noted that the Braidwood Unit 1 and 2 flux thimbles have experienced more wear than the Byron Unit 1 and 2 flux thimbles. Due to the observed higher wear rates, the examination frequency for both Braidwood Units was changed to every refueling outage. In addition, the operating experience provided in the LRA indicated that there have been instances when, either due to an obstruction or due to other outage related work, all the Braidwood flux thimbles were not examined.

Furthermore, the staffs review of operating experience data base for Braidwood also revealed that eddy current examinations were not performed for certain flux thimbles, due to the presence of moisture in the flux thimble tubes.

Issue:

The applicants plant-specific operating experience discussion in the LRA section states that Braidwood Units 1 and 2 flux thimble tubes examination frequency is every outage, due to higher than anticipated wear rates. The staff noted that the applicants operating experience discussion in the LRA did not fully address the reasons for the unexpected high wear rates observed for Braidwood Units 1 and 2 address all the issues during eddy current testing which precluded the testing of all flux thimbles. The staff is concerned about the sufficiency of the proposed AMP if these issues are not properly addressed and corrected.

Request:

1. Provide information in terms of root cause analyses and corrective actions which can explain and account for the higher than anticipated observed wear rates for Braidwood Units 1 and 2 flux thimble tubes.

2. Explain what root cause analyses and corrective actions have been performed to correct the occurrences of moisture in the thimble tubes given that these occurrences interfere in eddy current examinations of the flux thimble tubes.

3. Justify the adequacy of the program if the unexpected high wear rates are not accounted for and mitigated, given that there are issues related to the eddy current examinations of all flux thimble tubes (i.e., conflicting outage schedule, tube blockage, and the presence of moisture in the flux thimbles).

Exelon Response:

1. The frequency of flux thimble tube eddy current testing at Braidwood Station, Units 1 and 2, was conservatively increased to every refueling outage based on higher than anticipated flux thimble wear rates on two (2) of 58 flux thimble tubes on Unit 1 and two (2) of 58 flux thimble tubes on Unit 2.

RS-14-165 Enclosure A Page 20 of 23 For Unit 1, higher than anticipated wear rates of 37% per cycle and 27% per cycle were observed on two (2) flux thimble tubes during the Fall 2010 Refueling Outage. These two (2) flux thimble tubes were installed in the spring 2009 Refueling Outage and replaced original equipment flux thimble tubes. The two (2) original flux thimble tubes had been capped during the Fall 2007 Refueling Outage due to the flux thimble tubes being restricted (i.e., full length could not be tested). Subsequent eddy current test results showed that the location with 27% wear in 2010 had 26% wear in 2012, or no distinguishable wear during the operating cycle. The flux thimble tube in the location that experienced 37% wear in 2010 had to be replaced in 2012 due to a neutron detector becoming stuck while the unit was online; therefore, eddy current testing was not performed.

For Unit 2, a higher than anticipated wear rate of 35% per cycle was observed on one (1) flux thimble tube during the Spring 2011 Refueling Outage. This flux thimble tube was installed in the Fall 2009 Refueling Outage and replaced an original equipment flux thimble tube. The original flux thimble tube had been capped during the Spring 2008 Refueling Outage due to wear. Subsequent eddy current test results showed that the location with 35% wear in 2011 had 41% wear in 2012. In addition, a higher than expected wear rate was observed on another original equipment flux thimble tube in which the wear went from 36% in the Spring 2008 Refuel Outage to 57% in the Spring 2011 Refuel Outage, an increase in wear of 21% in two (2) operating cycles. Previous testing indicated a 3% per cycle wear rate. This flux thimble tube was tested during the Fall 2012 Refueling Outage. The results indicate that the wear increased from 57% to 60%, or a 3% change during the operating cycle which is consistent with previous test results. This tube was removed from service (capped) based on the greater than 60%

wear criterion. This demonstrates that actions are being taken under the existing program to ensure continued operability of the flux thimble tubes.

Industry operating experience indicates that the highest flux thimble tube wear rate occurs during the first operating cycle and then decreases with subsequent cycles, which is consistent with tube wear observed at Braidwood Station. The cause of the higher than anticipated wear rates has not been determined but it can be seen from the discussion above, the higher than anticipated wear rates that resulted in performing flux thimble tube eddy current testing each outage are not excessive or wide spread (i.e., two (2) locations), and these issues are being properly addressed and corrected by the Flux Thimble Tube Inspection aging management program. The highest observed wear of 37% that occurred during one cycle of operation is below the criteria for corrective action (60% wall loss) and well below the limit for structural integrity (85% wall loss); therefore the wear is manageable within the defined test frequency. Observed wear rates were found significantly lower on each affected flux thimble during the second cycle of operation and additional similar test results could make increasing the time between tests appropriate. Performing eddy current testing each outage may not be justifiable long term due to the radiological dose concerns, cost, and station resources. If subsequent eddy current testing does not support increasing the time between tests due only to these specific locations, then these locations could either be abandoned or the flux thimble tubes replaced each outage since only a few flux thimble tube locations have experienced higher than anticipated wear.

RS-14-165 Enclosure A Page 21 of 23

2. The cause of the moisture in the flux thimble tubes was determined to be condensation due to changes in containment temperature during the time period between when the flux thimble tubes are cleaned and dried and the performance of eddy current testing.

Flux thimble cleaning is typically performed at the beginning of the refueling outage. All accessible flux thimble tubes are cleaned using alcohol and water followed by forced air drying. After cleaning, a dummy neutron probe is inserted into each flux thimble tube to gauge the flux thimble tube, with the objective of ensuring a clear pathway. The flux thimble tubes are then withdrawn to support fuel offload. Once the fuel is reloaded into the reactor vessel, approximately 16 days later, the flux thimble tubes are re-inserted and eddy current testing is performed. In order to reduce the potential for condensation build-up, a corrective action to perform eddy current testing immediately after cleaning and drying is being implemented. Another corrective action of only cleaning approximately half the flux thimble tubes each refueling outage is being implemented, since the cleaning process has the potential to leave residual water in the flux thimble tubes. The effectiveness of these corrective actions is still under evaluation.

The issue of moisture hindering the ability to collect eddy current data is relatively recent, since eddy current data for all accessible flux thimble tubes was able to be obtained during the Unit 1 Fall 2010 Refueling Outage and the Unit 2 Spring 2011 Refueling Outage. Full length eddy current data for all accessible flux thimble tubes at both units has been collected in the past and no modifications to the flux thimble tubes have been made. A dummy probe was capable of being inserted during the cleaning process and the flux thimble tubes are capable of being withdrawn and inserted indicating the flux thimble tubes are not damaged or deformed. This issue has been entered into the corrective action program and is being investigated. Additional corrective actions will focus on any changes made to work practices, cleaning and testing procedures and equipment since the last successful performance of flux thimble tube eddy current testing at Braidwood Station, Units 1 and 2.

3. The Flux Thimble Tube Inspection aging management program accounts for unexpected wear rates by imposing a low threshold for corrective action. The program requires that corrective action (i.e., replacement, re-positioning, or isolation) be taken if wall loss greater than 60% is identified. The program also requires that action be taken if the wall loss is less than 60% but the projected wall loss prior to the next scheduled test exceeds 80%. Analysis determined that a flux thimble tube is capable of performing its intended function with up to 85% wall loss. The Flux Thimble Tube Inspection aging management program also directs that if full length eddy current test data for each flux thimble tube is not obtained, further review is required. This review will determine additional actions that can include replacement, isolation (capping), or a conservative projection of wear based historical test data. Projections of wear are performed using two (2) methods; a linear projection and the method described in WCAP-12866, Bottom Mounted Instrumentation Flux Thimble Wear, which is an exponentially decreasing projection.

The more severe wear projection from these two (2) methods is then evaluated against a more conservative wall loss criteria (i.e., 50% wall loss) to determine if further action is required prior to the next scheduled eddy current test. Based on this information, the aging management program is managing, monitoring, and maintaining the system and components for this program and will resolve the issues related to eddy current examination of all flux thimble tubes to ensure that the intended function will be maintained throughout the period of extended operation.

RS-14-165 Enclosure A Page 22 of 23 RAI 3.5.1-1 Applicability:

Byron and Braidwood

Background:

The LRA states, in Sections B.2.1.34 and B.2.1.30, respectively, that the Structures Monitoring and ASME Section XI, Subsection IWL programs are consistent, with enhancements, with the GALL Report Chapters XI.S6, Structures Monitoring, and XI.S2, ASME Section XI, Subsection IWL. The GALL Report XI.S6 Scope of Program program element states that the program includes all structures, structural components, component supports, and structural commodities in the scope of license renewal that are not covered by other structural AMPs (i.e., ASME Section XI, Subsection IWE (AMP X1.S1) and ASME Section XI, Subsection IWL (AMP X1.S2)).

The Parameters Monitored or Inspected program element states: If necessary for managing settlement and erosion of porous concrete sub-foundations, the continued functionality of a site de-watering system is monitored.

The GALL Report AMP XI.S2 Program Description section states that 10 CFR 50.55a imposes the examination requirements of ASME Code,Section XI, Subsection IWL, for Class CC reinforced and prestressed concrete containments. The Scope of Program program element states that the components within the scope of Subsection IWL are reinforced concrete and unbonded post-tensioning systems of Class CC containments. Subsection IWL exempts from examination portions of the concrete containment that are inaccessible such as concrete covered by liner, foundation material, or backfill or obstructed by adjacent structures or other components. However, 10 CFR 50.55a(b)(2)(viii) specifies additional requirements for inaccessible areas that require the licensee to evaluate the acceptability of concrete in inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation of inaccessible areas.

LRA Table 3.5.1, Item 3.5.1-1 corresponds to Item II.A1.CP-101 in the GALL Report for containment concrete components (Concrete: dome; wall; basemat; ring girders; buttresses; Concrete elements, all) for the aging effect of concrete cracking and distortion due to the aging mechanism of increased stress levels from settlement in a soil environment. For this item, the LRA states in the Discussion column: Consistent with NUREG-1801. The Structures Monitoring (B.2.1.34) program will be used to manage cracking and distortion of the concrete dome, wall, basemat, and buttresses in inaccessible areas of the Containment Structure exposed to a groundwater and soil environment. [Byron and Braidwood Stations] BBS do not rely upon a de-watering system to control settlement. See subsection 3.5.2.2.1.1. LRA Section 3.5.2.2.1.1 states that Item 3.5.1-1 is applicable to Byron and Braidwood and that inaccessible below grade containment concrete surfaces will be examined by the Structures Monitoring (B.2.1.34) program when excavated for any reason.

Issue:

For the aging effect of cracking and distortion due to increase in stress levels from settlement, although settlement can occur in a soil environment, the symptoms can be manifested in either an air-indoor uncontrolled or air-outdoor environment (see Table IX.E of the GALL Report for the aging effect term Cracks; distortion; increase in component stress level). NUREG-1800,

RS-14-165 Enclosure A Page 23 of 23 Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (SRP-LR), Section 3.5.2.2.1.1 acceptance criterion for the aging effect or mechanism of cracking and distortion due to increased stress levels from settlement states that the existing program relies on ASME Section XI, Subsection IWL to manage these aging effects on concrete components of the containment pressure-resisting boundary in both accessible and inaccessible areas.

Contrary to the scope and program descriptions for the GALL Report AMP XI.S2 and GALL Report AMP XI.S6 and the SRP-LR Section 3.5.2.2.1.1 acceptance criterion, LRA Table 3.5.1-1 does not identify the ASME Section XI, Subsection IWL as an applicable AMP. Also, LRA Table 3.5.2-4 does not identify the aging effect or mechanism corresponding to LRA Table 3.5-1, Item 3.5.1-1, for containment pressure boundary concrete component (Concrete: dome; wall; basemat; ring girders; buttresses; reinforcing steel) as an aging effect requiring management for accessible areas and for inaccessible areas; it identifies the Structures Monitoring Program as the only applicable AMP. The applicant has not provided the technical basis justifying these determinations and consistency with NUREG-1801 in the LRA for LRA Table 3.5.1, Item 3.5.1-1, and corresponding line items in LRA Table 3.5.2-4.

Request:

With regard to AMR line item 3.5.1-1 in LRA Table 3.5-1 that corresponds to Item II.A1.CP-101 in the GALL Report, provide the technical basis to justify: (a) why the ASME Section XI, Subsection IWL program is not listed for aging management of concrete containment pressure-resisting boundary components in accessible and inaccessible areas for this AMR line item and corresponding items in LRA Table 3.5.2-4 and (b) why the aging effect or mechanism corresponding to the AMR line item is not identified as an aging effect or mechanism requiring management in LRA Table 3.5.2-4 for containment pressure boundary concrete components in accessible areas.

Exelon Response:

ASME Section XI, Subsection IWL program was not listed for aging management of concrete containment pressure-resisting boundary components in accessible and inaccessible areas for AMR line item 3.5.1-1 in LRA Table 3.5.1 due to a different interpretation of the intent of the GALL line item, which specifies the use of either Chapter XI.S2, ASME Section XI, Subsection IWL or Chapter XI.S6, "Structure Monitoring as the aging management program.

ASME Section XI, Subsection IWL program will be used for aging management of accessible and inaccessible concrete containment pressure-resisting boundary component areas for AMR line item 3.5.1-1 in LRA Table 3.5.1.

LRA changes include revision to Section 3.5.2.2.1.1, revision to Table 3.5.1, Item 3.5.1-1, and revision to Table 3.5.2-4, which are provided in Enclosure B of this submittal.

RS-14-165 Enclosure B Page 1 of 34 Enclosure B Byron and Braidwood Stations, Units 1 and 2 License Renewal Application (LRA) updates resulting from the responses to RAIs contained in Enclosure A of this letter Note: Provided below is a list of the RAI numbers for which the responses in Enclosure A have resulted in changes to the LRA. LRA mark-ups are grouped in accordance with the RAI response by which they are affected, and provided in the order that the responses appear in Enclosure A. The corresponding page numbers listed below identify the page of this Enclosure where the first LRA changes associated with each respective RAI response are shown. To facilitate understanding, portions of the original LRA have been repeated in this Enclosure, with revisions indicated. Existing LRA text is shown in normal font. Changes are highlighted with bolded italics for inserted text and strikethroughs for deleted text.

RAI Number Enclosure B Page RAI 3.5.2.2.1-1. 2 RAI 3.5.2-6. 28 RAI 3.5.1-1. 31

RS-14-165 Enclosure B Page 2 of 34 As a result of the response to RAI 3.5.2.2.1-1 provided in Enclosure A of this letter, the following LRA Sections are revised as follows:

LRA Section LRA Page Enclosure B Page 3.5.2.2.1.8 3.5-31 3 3.5.2.2.2.1, Item 2 3.5-34 5 3.5.2.2.2.3, Item 2 3.5-39 6 Table 3.5.1 Item 3.5.1-12 3.5-46 7 Table 3.5.1 Item 3.5.1-19 3.5-49 8 Table 3.5.1 Item 3.5.1-43 3.5-56 9 Table 3.5.1 Item 3.5.1-50 3.5-59 10 Table 3.5.1 Item 3.5.1-54 3.5-60 11 Table 3.5.2-1 Auxiliary Building* 3.5-84 12 Table 3.5.2-2 Circulating Water Pump House (Byron)* 3.5-99 13 Table 3.5.2-4 Containment Structure* 3.5-131 14 Table 3.5.2-5 Deep Well Enclosures (Byron)* 3.5-169 16 Table 3.5.2-7 Essential Service Water Cooling Towers 3.5-175 17 (Byron)*

Table 3.5.2-8 Fuel Handling Building* 3.5-190 18 Table 3.5.2-9 Lake Screen Structures (Braidwood)* 3.5-202 19 Table 3.5.2-10 Main Steam & Auxiliary Feedwater Tunnels 3.5-211 20 and Isolation Valve Rooms*

Table 3.5.2-11 Natural Draft Cooling Towers (Byron)* 3.5-219 21 Table 3.5.2-12 RWST Foundation and Tunnel* 3.5-223 22 Table 3.5.2-13 Radwaste and Service Building Complex* 3.5-230 23 Table 3.5.2-14 River Screen House (Byron)* 3.5-238 24 Table 3.5.2-16 Switchyard Structures* 3.5-264 25 Table 3.5.2-17 Turbine Building Complex* 3.5-270 26 Table 3.5.2-18 Yard Structures* 3.5-277 27

• For clarity purposes, the new line items shown in Enclosure B for the Summary of Aging Management Evaluation Table 2s are presented by themselves without any of the existing line items. The new additions are shown as individual line item additions, with intended functions grouped together; therefore the appearance is different than the cascading table approach in the LRA.

RS-14-165 Enclosure B Page 3 of 34 3.5.2.2.1.8 Cracking due to Expansion from Reaction with Aggregates Cracking due to expansion from reaction with aggregates could occur in inaccessible areas of concrete elements of PWR and BWR concrete and steel containments. The GALL Report recommends further evaluation to determine if a plant-specific aging management program is required to manage this aging effect.

Acceptance criteria are described in Branch Technical Position RLSB-1.

Item Number 3.5.1-12 is not applicable to BBS. This aging effect and mechanism combination does not apply to BBS concrete Containment Structures. Concrete fine and course aggregates conform to ASTM C33. Petrographic examinations of aggregates used in concrete were performed in accordance with ASTM C295, Petrographic Examination of Aggregates for Concrete, and ASTM C289, Potential Reactivity of Aggregates, to demonstrate that the aggregates do not adversely react within the concrete. In addition, concrete structures were constructed in accordance with ACI 318 per UFSAR Table 3.8-2. Thus, cracking due to expansion and reaction with aggregates is not expected, however, this potential aging effect is managed regardless of the mechanism that caused the aging effect.

Cracking associated with expansion due to reaction with aggregates has not been observed on BBS concrete structures, including Containment Structures.

Nevertheless, the ASME Section XI, Subsection IWL (B.2.1.30) program and the Structures Monitoring (B.2.1.34) program will continue to inspect and monitor the concrete Containment Structures for cracking due to any mechanism. BBS will also examine exposed portions of the below-grade concrete, when excavated for any reason, in accordance with the Structures Monitoring (B.2.1.34) program.

Accessible concrete surfaces of the Containment Structures, that are part of the containment pressure boundary, are monitored for cracking due to expansion from reaction with aggregates by the ASME Section XI, Subsection IWL (B.2.1.30) program and addressed under Item Number 3.5.1-19. The ASME Section XI, Subsection IWL (B.2.1.30) program requires evaluation of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas. In addition, other accessible containment concrete, not part of the containment pressure boundary, is monitored for cracking due to expansion from reaction with aggregates by the Structures Monitoring (B.2.1.34) program and addressed under Item Number 3.5.1-54.

The Structures Monitoring (B.2.1.34) program also requires evaluation of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas. The condition of accessible and above grade concrete is used as an indicator for the condition of the inaccessible and below grade structural components and provides reasonable assurance that degradation of inaccessible structural components will be detected before a loss of an intended function.

The BBS structural concrete was constructed as recommended to preclude cracking due to this mechanism; therefore, no plant specific aging management program or further evaluation of inaccessible below grade concrete for this

RS-14-165 Enclosure B Page 4 of 34 mechanism is required. The ASME Section XI, Subsection IWL program (B.2.1.30) and the Structures Monitoring (B.2.1.34) program are described in Appendix B.

RS-14-165 Enclosure B Page 5 of 34 3.5.2.2.2.1 Aging Management of Inaccessible Areas

2. Cracking due to expansion and reaction with aggregates could occur in below-grade inaccessible concrete areas for Groups 1-5 and 7-9 structures. The GALL Report recommends further evaluation of inaccessible areas of these Groups of structures if concrete was not constructed in accordance with the recommendations in the GALL Report.

Item Number 3.5.1-43 is not applicable to BBS. Structures other than Containment at BBS consist of Groups 3 through 8. This aging effect and mechanism combination does not apply to BBS concrete structures. Concrete fine and course aggregates conform to ASTM C33. Petrographic examinations of aggregates used in concrete were performed in accordance with ASTM C295, Petrographic Examination of Aggregates for Concrete, and ASTM C289, Potential Reactivity of Aggregates, to demonstrate that the aggregates do not adversely react within the concrete. In addition, concrete structures were constructed in accordance with ACI 318, per UFSAR Table 3.8-2. Thus, cracking due to expansion and reaction with aggregates is not expected, however, this potential aging effect is managed regardless of the mechanism that caused the aging effect.Thus, cracking due to expansion and reaction with aggregates is not applicable and requires no aging management.

Cracking associated with expansion due to reaction with aggregates has not been observed on BBS concrete structures. Nevertheless, the Structures Monitoring (B.2.1.34) program continues to inspect and monitor concrete structures for cracking due to any mechanism. Accessible concrete is monitored for cracking due to expansion from reaction with aggregates by the Structures Monitoring (B.2.1.34) program and addressed under Item Number 3.5.1-54. The Structures Monitoring (B.2.1.34) program requires evaluation of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas. BBS will also examine exposed portions of the below-grade concrete, when excavated for any reason, in accordance with the Structures Monitoring (B.2.1.34) program. The condition of accessible and above grade concrete is used as an indicator for the condition of the inaccessible and below grade structural components and provides reasonable assurance that degradation of inaccessible structural components will be detected before a loss of an intended function. The BBS structural concrete was constructed as recommended to preclude cracking due to this mechanism; therefore, no plant specific aging management program or further evaluation of inaccessible below grade concrete for this mechanism is required. The Structures Monitoring (B.2.1.34) program is described in Appendix B.

RS-14-165 Enclosure B Page 6 of 34 3.5.2.2.2.3 Aging Management of Inaccessible Areas for Group 6 Structures

2. Cracking due to expansion and reaction with aggregates could occur in below-grade inaccessible reinforced concrete areas of Group 6 structures. The GALL Report recommends further evaluation to determine if a plant-specific aging management program is required to manage this aging effect. Acceptance criteria are described in Branch Technical Position RLSB-1.

Item Number 3.5.1-50 is not applicable to BBS. This aging effect and mechanism combination does not apply to BBS Group 6 concrete structures.

Concrete fine and course aggregates conform to ASTM C33. Petrographic examinations of aggregates used in concrete were performed in accordance with ASTM C295, Petrographic Examination of Aggregates for Concrete, and ASTM C289, Potential Reactivity of Aggregates, to demonstrate that the aggregates do not adversely react within the concrete. In addition, concrete structures were constructed in accordance with ACI 318, per UFSAR Table 3.8-2. Thus, cracking due to expansion and reaction with aggregates is not expected, however, this potential aging effect is managed regardless of the mechanism that caused the aging effect.

Cracking associated with expansion due to reaction with aggregates has not been observed on BBS Group 6 concrete structures. Nevertheless, the Regulatory Guide 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35) aging management program continues to inspect and monitor Group 6 concrete structures for cracking due to any mechanism. Accessible group 6 concrete is monitored for cracking due to expansion from reaction with aggregates by the Regulatory Guide 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35) program and addressed under Item Number 3.5.1-54.

The Regulatory Guide 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35) program requires evaluation of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas. The condition of accessible and above grade concrete is used as an indicator for the condition of the inaccessible and below grade structural components and provides reasonable assurance that degradation of inaccessible structural components will be detected before a loss of an intended function. BBS will also examine exposed portions of the below grade concrete, when excavated for any reason, in accordance with the Structures Monitoring (B.2.1.34) program. The BBS structural concrete was constructed as recommended to preclude cracking due to this mechanism. Therefore, no plant specific aging management program is required offor inaccessible, below grade, Group 6 concrete structures for this mechanism. The RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35) program, and the Structures Monitoring (B.2.1.34) program are described in Appendix B.

RS-14-165 Enclosure B Page 7 of 34 Table 3.5.1 Summary of Aging Management Evaluations for the Structures and Component Supports Further Item Aging Aging Management Component Evaluation Discussion Number Effect/Mechanism Programs Recommended 3.5.1-12 Concrete (inaccessible Cracking Further evaluation is Yes, if concrete is Not Applicable areas): dome; wall; due to expansion from required to determine if not constructed as basemat; ring girders; reaction with a plant-specific aging stated function This aging effect/mechanism does not apply to BBS buttresses, Concrete aggregates management program Containment Structures. The aging mechanism of (inaccessible areas): is needed. concrete expansion from reaction with aggregates is basemat, Concrete not expected at BBS Containment Structures. Fine and (inaccessible areas): course aggregates conform to ASTM C33. Petrographic containment; wall; examinations of aggregates were performed in accordance basemat, Concrete with ASTM C295 and ASTM C289. In addition, concrete (inaccessible areas): structures were constructed in accordance with ACI 318.

basemat, concrete fill-in Cracking associated with expansion due to reaction with annulus aggregates has not been observed on accessible portions of the BBS Containment Structures.

Accessible concrete surfaces of the Containment Structures that are part of the containment pressure boundary are monitored for cracking due to expansion from reaction with aggregates by the ASME Section XI, Subsection IWL (B.2.1.30) program and addressed under Item Number 3.5.1-19. The ASME Section XI, Subsection IWL (B.2.1.30) program requires evaluation of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas. In addition, other accessible containment concrete, not part of the containment pressure boundary, is monitored for cracking due to expansion from reaction with aggregates by the Structures Monitoring (B.2.1.34) program and addressed under Item Number 3.5.1-54. The Structures Monitoring (B.2.1.34) program also requires evaluation of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas.

See subsection 3.5.2.2.1.8.

RS-14-165 Enclosure B Page 8 of 34 Table 3.5.1 Summary of Aging Management Evaluations for the Structures and Component Supports Further Item Aging Aging Management Component Evaluation Discussion Number Effect/Mechanism Programs Recommended 3.5.1-19 Concrete (accessible Cracking Chapter XI.S2, ASME No Not Applicable.

areas): dome; wall; due to expansion from Section XI, Subsection basemat; ring girders; reaction with IWL Consistent with NUREG-1801.

buttresses, Concrete aggregates This aging effect/mechanism does not apply to BBS (accessible areas): Containment Structures. The aging mechanism of basemat, Concrete concrete expansion from reaction with aggregates is (accessible areas): not expected at BBS Containment Structures. Fine and containment; wall; course aggregates conform to ASTM C33. Petrographic basemat, Concrete examinations of aggregates were performed in accordance (accessible areas): with ASTM C295 and ASTM C289. In addition, concrete basemat, concrete fill-in structures were constructed in accordance with ACI 318.

annulus Cracking associated with expansion due to reaction with aggregates has not been observed on accessible portions of the BBS Containment Structures.

The ASME Section XI, Subsection IWL (B.2.1.30) program will be used to manage cracking due to expansion from reaction with aggregates in concrete:

dome; wall; basemat; ring girders; buttresses; reinforcing steel (accessible areas) exposed to indoor air, air with borated water leakage, outdoor air, and water flowing environments.

RS-14-165 Enclosure B Page 9 of 34 Table 3.5.1 Summary of Aging Management Evaluations for the Structures and Component Supports Further Item Aging Aging Management Component Evaluation Discussion Number Effect/Mechanism Programs Recommended 3.5.1-43 All Groups except Cracking Further evaluation is Yes, if concrete is Not Applicable.

Group 6:Concrete due to expansion from required to determine if not constructed as (inaccessible areas): all reaction with a plant-specific aging stated The aging effect/mechanism does not apply to BBS aggregates management program concrete structures. The aging mechanism of concrete is needed. expansion from reaction with aggregates is not expected at BBS structures. Fine and course aggregates conform to ASTM C33. Petrographic examinations of aggregates were performed in accordance with ASTM C295 and ASTM C289. In addition, concrete structures were constructed in accordance with ACI 318. Cracking associated with expansion due to reaction with aggregates has not been observed on accessible portions of BBS concrete structures.

Accessible concrete is monitored for cracking due to expansion from reaction with aggregates by the Structures Monitoring (B.2.1.34) program and addressed under Item Number 3.5.1-54. The Structures Monitoring (B.2.1.34) program requires evaluation of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas.

See subsection 3.5.2.2.2.1.2.

RS-14-165 Enclosure B Page 10 of 34 Table 3.5.1 Summary of Aging Management Evaluations for the Structures and Component Supports Further Item Aging Aging Management Component Evaluation Discussion Number Effect/Mechanism Programs Recommended 3.5.1-50 Groups 6: concrete Cracking Further evaluation is Yes, if concrete is Not Applicable.

(inaccessible areas): all due to expansion from required to determine if not constructed as reaction with a plant-specific aging stated The aging effect/mechanism does not apply to BBS Group aggregates management program 6 concrete structures. The aging mechanism of concrete is needed. expansion from reaction with aggregates is not expected at BBS structures. Fine and course aggregates conform to ASTM C33. Petrographic examinations of aggregates were performed in accordance with ASTM C295 and ASTM C289. In addition, concrete structures were constructed in accordance with ACI 318.

Accessible group 6 concrete is monitored for cracking due to expansion from reaction with aggregates by the Regulatory Guide 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35) program and addressed under Item Number 3.5.1-54. The Regulatory Guide 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35) program requires evaluation of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas.

See Subsection 3.5.2.2.2.3.2.

RS-14-165 Enclosure B Page 11 of 34 Table 3.5.1 Summary of Aging Management Evaluations for the Structures and Component Supports Further Item Aging Aging Management Component Evaluation Discussion Number Effect/Mechanism Programs Recommended 3.5.1-54 All groups except 6: Cracking Chapter XI.S6, No Not Applicable.

concrete (accessible due to expansion from Structures Monitoring areas): all reaction with Consistent with NUREG-1801.

aggregates The aging effect/mechanism does not apply to BBS Group 6 concrete structures. The aging mechanism of concrete expansion from reaction with aggregates is not expected at BBS structures. Fine and course aggregates conform to ASTM C33. Petrographic examinations of aggregates were performed in accordance with ASTM C295 and ASTM C289. In addition, concrete structures were constructed in accordance with ACI 318.

The Structures Monitoring (B.2.1.34) program will be used to manage cracking due to expansion from reaction with aggregates in accessible areas of all groups except group 6 concrete exposed to indoor air, air with borated water leakage, outdoor air, and water flowing environments.

The Regulatory Guide 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35) program will be substituted to manage cracking due to expansion from reaction with aggregates in accessible areas of group 6 concrete exposed to indoor air, outdoor air, and water flowing environments.

RS-14-165 Enclosure B Page 12 of 34 Table 3.5.2-1 Auxiliary Building Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete Curbs Direct Flow Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A concrete (B.2.1.34)

Concrete Curbs Direct Flow Reinforced Air with Borated Water Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A concrete Leakage (B.2.1.34)

Concrete: Above- Flood Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A grade exterior Missile Barrier concrete (B.2.1.34)

(accessible areas) Shelter, Protection Structural Pressure Barrier Structural Support Concrete: Interior Flood Barrier Reinforced Air with Borated Water Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A HELB Shielding concrete Leakage (B.2.1.34)

Missile Barrier Shelter, Protection Shielding Structural Pressure Barrier Structural Support Water retaining boundary Concrete: Interior Water retaining Reinforced Water - Flowing Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A boundary concrete (B.2.1.34)

Hatches/Plugs Missile Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A concrete (B.2.1.34)

Hatches/Plugs Flood Barrier Reinforced Air with Borated Water Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Missile Barrier concrete Leakage (B.2.1.34)

Shelter, Protection Shielding

RS-14-165 Enclosure B Page 13 of 34 Table 3.5.2-2 Circulating Water Pump House (Byron)

Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Above- Shelter, Protection Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A grade exterior Structural Support concrete (B.2.1.34)

(accessible areas)

Concrete: Interior Shelter, Protection Reinforced Air - Indoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Structural Support concrete Uncontrolled (B.2.1.34)

Precast Panel Shelter, Protection Reinforced Air - Indoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Structural Support concrete Uncontrolled (B.2.1.34)

Precast Panel Shelter, Protection Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Structural Support concrete (B.2.1.34)

RS-14-165 Enclosure B Page 14 of 34 Table 3.5.2-4 Containment Structure Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete Curbs Direct Flow Reinforced Air with Borated Water Cracking Structures Monitoring III.A1.TP-25 3.5.1-54 A concrete Leakage (B.2.1.34)

Concrete: Above- Flood Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A1.TP-25 3.5.1-54 A grade exterior Missile Barrier concrete (B.2.1.34)

(accessible areas Shelter, Protection of exterior Shielding features) Structural Support Concrete: Dome; Flood Barrier Reinforced Air - Indoor Cracking ASME Section XI, II.A1.CP-33 3.5.1-19 A wall; basemat; HELB Shielding concrete Uncontrolled Subsection IWL ring girders; Missile Barrier (B.2.1.30) buttresses; Pressure Boundary reinforcing steel Shelter, Protection (accessible areas) Shielding Structural Support Concrete: Dome; Flood Barrier Reinforced Air - Outdoor Cracking ASME Section XI, II.A1.CP-33 3.5.1-19 A wall; basemat; HELB Shielding concrete Subsection IWL ring girders; Missile Barrier (B.2.1.30) buttresses; Pressure Boundary reinforcing steel Shelter, Protection (accessible areas) Shielding Structural Support Concrete: Dome; Flood Barrier Reinforced Air with Borated Water Cracking ASME Section XI, II.A1.CP-33 3.5.1-19 A wall; basemat; HELB Shielding concrete Leakage Subsection IWL ring girders; Missile Barrier (B.2.1.30) buttresses; Pressure Boundary reinforcing steel Shelter, Protection (accessible areas) Shielding Structural Support Concrete: Dome; Flood Barrier Reinforced Water - Flowing Cracking ASME Section XI, II.A1.CP-33 3.5.1-19 A wall; basemat; HELB Shielding concrete Subsection IWL ring girders; Missile Barrier (B.2.1.30) buttresses; Pressure Boundary reinforcing steel Shelter, Protection (accessible areas) Shielding Structural Support

RS-14-165 Enclosure B Page 15 of 34 Table 3.5.2-4 Containment Structure (Continued)

Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Interior HELB Shielding Reinforced Air with Borated Water Cracking Structures Monitoring III.A4.TP-25 3.5.1-54 A Missile Barrier concrete Leakage (B.2.1.34)

Pipe Whip Restraint Shelter, Protection Shielding Structural Support Concrete: Interior Flood Barrier Reinforced Air - Indoor Cracking Structures Monitoring III.A4.TP-25 3.5.1-54 A (Exterior Missile Barrier concrete Uncontrolled (B.2.1.34) structural Shelter, Protection features) Shielding Structural Support Hatches/Plugs HELB Shielding Reinforced Air with Borated Water Cracking Structures Monitoring III.A4.TP-25 3.5.1-54 A Missile Barrier concrete Leakage (B.2.1.34)

Shelter, Protection Shielding Tunnel (Tendon Shelter, Protection Reinforced Air - Indoor Cracking Structures Monitoring III.A4.TP-25 3.5.1-54 A access gallery) concrete Uncontrolled (B.2.1.34)

Tunnel (Tendon Shelter, Protection Reinforced Water - Flowing Cracking Structures Monitoring III.A4.TP-25 3.5.1-54 A access gallery) concrete (B.2.1.34)

RS-14-165 Enclosure B Page 16 of 34 Table 3.5.2-5 Deep Well Enclosures (Byron)

Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Above- Missile Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A grade exterior Shelter, Protection concrete (B.2.1.34)

(accessible areas) Structural Support Concrete: Interior Missile Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Shelter, Protection concrete (B.2.1.34)

Structural Support

RS-14-165 Enclosure B Page 17 of 34 Table 3.5.2-7 Essential Service Water Cooling Towers (Byron)

Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Above- Missile Barrier Reinforced Air - Outdoor Cracking RG 1.127, Inspection of III.A3.TP-25 3.5.1-54 E, 2 grade exterior Shelter, Protection concrete Water-Control (accessible areas) Structural Support Structures Associated Water retaining with Nuclear Power boundary Plants (B.2.1.35)

Concrete: Above- Missile Barrier Reinforced Water - Flowing Cracking RG 1.127, Inspection of III.A3.TP-25 3.5.1-54 E, 2 grade exterior Shelter, Protection concrete Water-Control (accessible areas) Structural Support Structures Associated Water retaining with Nuclear Power boundary Plants (B.2.1.35)

Concrete: Interior Flood Barrier Reinforced Air - Indoor Cracking RG 1.127, Inspection of III.A3.TP-25 3.5.1-54 E, 2 Missile Barrier concrete Uncontrolled Water-Control Shelter, Protection Structures Associated Structural Support with Nuclear Power Plants (B.2.1.35)

Plant Specific Notes:

2. The RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35) program is used to manage the aging effect(s) applicable to this component type, material, and environment combination.

RS-14-165 Enclosure B Page 18 of 34 Table 3.5.2-8 Fuel Handling Building Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete Curbs Direct Flow Reinforced Air with Borated Water Cracking Structures Monitoring III.A5.TP-25 3.5.1-54 A concrete Leakage (B.2.1.34)

Concrete: Above- Flood Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A5.TP-25 3.5.1-54 A grade exterior Missile Barrier concrete (B.2.1.34)

(accessible areas) Shelter, Protection Shielding Structural Pressure Barrier Structural Support Concrete: Interior Flood Barrier Reinforced Air with Borated Water Cracking Structures Monitoring III.A5.TP-25 3.5.1-54 A Shelter, Protection concrete Leakage (B.2.1.34)

Shielding Structural Pressure Barrier Structural Support

RS-14-165 Enclosure B Page 19 of 34 Table 3.5.2-9 Lake Screen Structures (Braidwood)

Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Above- Shelter, Protection Reinforced Air - Outdoor Cracking RG 1.127, Inspection of III.A3.TP-25 3.5.1-54 E, 1 grade exterior Structural Support concrete Water-Control (accessible areas) Structures Associated with Nuclear Power Plants (B.2.1.35)

Concrete: Below- Shelter, Protection Reinforced Air - Outdoor Cracking RG 1.127, Inspection of III.A3.TP-25 3.5.1-54 E, 1 grade exterior Structural Support concrete Water-Control (accessible areas) Structures Associated with Nuclear Power Plants (B.2.1.35)

Concrete: Interior Shelter, Protection Reinforced Air - Indoor Cracking RG 1.127, Inspection of III.A3.TP-25 3.5.1-54 E, 1 Structural Support concrete Uncontrolled Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35)

Precast Panel Shelter, Protection Reinforced Air - Indoor Cracking RG 1.127, Inspection of III.A3.TP-25 3.5.1-54 E, 1 Structural Support concrete Uncontrolled Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35)

Precast Panel Shelter, Protection Reinforced Air - Outdoor Cracking RG 1.127, Inspection of III.A3.TP-25 3.5.1-54 E, 1 Structural Support concrete Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35)

Plant Specific Notes:

1. The RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35) program is used to manage the aging effect(s) applicable to this component type, material, and environment combination.

RS-14-165 Enclosure B Page 20 of 34 Table 3.5.2-10 Main Steam & Auxiliary Feedwater Tunnels and Isolation Valve Rooms Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Above- Flood Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A grade exterior Missile Barrier concrete (B.2.1.34)

(accessible areas) Shelter, Protection Structural Support Concrete: Interior Flood Barrier Reinforced Air - Indoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A HELB Shielding concrete Uncontrolled (B.2.1.34)

Missile Barrier Shelter, Protection Structural Support Concrete: Interior Flood Barrier Reinforced Water - Flowing Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A HELB Shielding concrete (B.2.1.34)

Missile Barrier Shelter, Protection Structural Support Hatches/Plugs Missile Barrier Reinforced Air - Indoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Shelter, Protection concrete Uncontrolled (B.2.1.34)

Hatches/Plugs Missile Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Shelter, Protection concrete (B.2.1.34)

RS-14-165 Enclosure B Page 21 of 34 Table 3.5.2-11 Natural Draft Cooling Towers (Byron)

Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Above- Structural Support Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A grade exterior Water retaining concrete (B.2.1.34)

(accessible areas) boundary

RS-14-165 Enclosure B Page 22 of 34 Table 3.5.2-12 RWST Foundation and Tunnel Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete Curbs Flood Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Missile Barrier concrete (B.2.1.34)

Shelter, Protection Structural Support Concrete: Above- Missile Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A grade exterior Shelter, Protection concrete (B.2.1.34)

(accessible areas) Structural Support Concrete: Structural Support Reinforced Air with Borated Water Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Foundation, concrete Leakage (B.2.1.34) subfoundation (accessible areas)

Concrete: Structural Support Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Foundation, concrete (B.2.1.34) subfoundation (accessible areas)

Concrete: Interior Flood Barrier Reinforced Air with Borated Water Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Missile Barrier concrete Leakage (B.2.1.34)

Shelter, Protection Structural Support

RS-14-165 Enclosure B Page 23 of 34 Table 3.5.2-13 Radwaste and Service Building Complex Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Above- Flood Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A grade exterior Shelter, Protection concrete (B.2.1.34)

(accessible areas) Structural Support Concrete: Flood Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Foundation, Shelter, Protection concrete (B.2.1.34) subfoundation Structural Support (accessible areas)

Concrete: Interior Flood Barrier Reinforced Air with Borated Water Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Shelter, Protection concrete Leakage (B.2.1.34)

Structural Support Precast Panel Shelter, Protection Reinforced Air with Borated Water Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Structural Support concrete Leakage (B.2.1.34)

Precast Panel Shelter, Protection Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Structural Support concrete (B.2.1.34)

RS-14-165 Enclosure B Page 24 of 34 Table 3.5.2-14 River Screen House (Byron)

Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Above- Shelter, Protection Reinforced Air - Outdoor Cracking RG 1.127, Inspection of III.A3.TP-25 3.5.1-54 E, 2 grade exterior Structural Support concrete Water-Control (accessible areas) Structures Associated with Nuclear Power Plants (B.2.1.35)

Concrete: Below- Shelter, Protection Reinforced Air - Outdoor Cracking RG 1.127, Inspection of III.A3.TP-25 3.5.1-54 E, 2 grade exterior Structural Support concrete Water-Control (accessible areas) Structures Associated with Nuclear Power Plants (B.2.1.35)

Concrete: Interior Flood Barrier Reinforced Air - Indoor Cracking RG 1.127, Inspection of III.A3.TP-25 3.5.1-54 E, 2 Shelter, Protection concrete Uncontrolled Water-Control Structural Support Structures Associated with Nuclear Power Plants (B.2.1.35)

Plant Specific Notes:

2. The RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.1.35) program is used to manage the aging effect(s) applicable to this component type, material, and environment combination.

RS-14-165 Enclosure B Page 25 of 34 Table 3.5.2-16 Switchyard Structures Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Above- Shelter, Protection Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A grade exterior Structural Support concrete (B.2.1.34)

(accessible areas)

Concrete: Interior Shelter, Protection Reinforced Air - Indoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Structural Support concrete Uncontrolled (B.2.1.34)

Equipment Structural Support Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A supports and concrete (B.2.1.34) foundations Hatches/Plugs Shelter, Protection Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A concrete (B.2.1.34)

RS-14-165 Enclosure B Page 26 of 34 Table 3.5.2-17 Turbine Building Complex Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Above- Shelter, Protection Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A grade exterior Structural Support concrete (B.2.1.34)

(accessible areas)

Concrete: Interior Flood Barrier Reinforced Air - Indoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Missile Barrier concrete Uncontrolled (B.2.1.34)

Shelter, Protection Structural Support Hatches/Plugs Shelter, Protection Reinforced Air - Indoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Structural Support concrete Uncontrolled (B.2.1.34)

Hatches/Plugs Shelter, Protection Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Structural Support concrete (B.2.1.34)

Precast Panel Shelter, Protection Reinforced Air - Indoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Structural Support concrete Uncontrolled (B.2.1.34)

Precast Panel Shelter, Protection Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Structural Support concrete (B.2.1.34)

RS-14-165 Enclosure B Page 27 of 34 Table 3.5.2-18 Yard Structures Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Above- Missile Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A grade exterior Shelter, Protection concrete (B.2.1.34)

(accessible areas) Structural Support Concrete: Structural Support Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Foundation, concrete (B.2.1.34) subfoundation (accessible areas)

Concrete: Interior Shelter, Protection Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Structural Support concrete (B.2.1.34)

Equipment Structural Support Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A supports and concrete (B.2.1.34) foundations Hatches/Plugs Missile Barrier Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Shelter, Protection concrete (B.2.1.34)

Manholes, Shelter, Protection Reinforced Water - Flowing Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Handholes & Duct Structural Support concrete (B.2.1.34)

Banks Manholes, Shelter, Protection Reinforced Air - Outdoor Cracking Structures Monitoring III.A3.TP-25 3.5.1-54 A Handholes & Duct Structural Support concrete (B.2.1.34)

Banks

RS-14-165 Enclosure B Page 28 of 34 As a result of the response to RAI 3.5.2-6 provided in Enclosure A of this letter, LRA Table 3.5.1 Item 3.5.1-24, page 3.5-51 is revised as shown below.

Existing LRA text is shown in normal font. Changes are highlighted with bolded italics for inserted text.

Table 3.5.1 Summary of Aging Management Evaluations for the Structures and Component Supports Further Item Aging Aging Management Component Evaluation Discussion Number Effect/Mechanism Programs Recommended 3.5.1-24 Concrete (inaccessible Increase in porosity Chapter XI.S2, ASME No Consistent with NUREG-1801.

areas): dome; wall; and permeability; Section XI, Subsection basemat; ring girders; cracking; loss of IWL, or Chapter XI.S6, The ASME Section XI, Subsection IWL buttresses, Concrete material (spalling, "Structures Monitoring" (B.2.1.30) program will be used to (inaccessible areas): scaling) manage increase in porosity and basemat, Concrete due to aggressive permeability, cracking, and loss of (accessible areas): chemical attack material (spalling, scaling) of the dome; wall; basemat concrete dome, wall, basemat, and buttresses in accessible and inaccessible areas of the Containment Structure exposed to groundwater and soil environments.

The Structures Monitoring (B.2.1.34) program will also be used to manage increase in porosity and permeability, cracking, and loss of material (spalling, scaling) of the concrete dome, wall, basemat, and buttresses in inaccessible areas of the Containment Structure exposed to groundwater and soil environments.

The Structures Monitoring (B.2.1.34) program will select a structure based on groundwater chemistry results and perform inspections to be used as a leading indicator for the condition of the below grade inaccessible concrete exposed to ground water and soil environments.

RS-14-165 Enclosure B Page 29 of 34 As a result of the response to RAI 3.5.2-6 provided in Enclosure A of this letter, LRA Table 3.5.2-4, which starts on page 3.5-131, is revised as shown below. For clarity purposes, the new line items are presented by themselves without any of the existing line items. The new additions are shown in bolded italics as individual line item additions and, therefore, appear different than the cascading table approach in the LRA.

Table 3.5.2-4 Containment Structure Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Dome; Flood Barrier Reinforced Groundwater/Soil Increase in Porosity and ASME Section XI, II.A1.CP-100 3.5.1-24 A wall; basemat; concrete Permeability, Cracking, Subsection IWL ring girders; Loss of Material (B.2.1.30) buttresses; (Spalling, Scaling) reinforcing steel Concrete: Dome; HELB Shielding Reinforced Groundwater/Soil Increase in Porosity and ASME Section XI, II.A1.CP-100 3.5.1-24 A wall; basemat; concrete Permeability, Cracking, Subsection IWL ring girders; Loss of Material (B.2.1.30) buttresses; (Spalling, Scaling) reinforcing steel Concrete: Dome; Missile Barrier Reinforced Groundwater/Soil Increase in Porosity and ASME Section XI, II.A1.CP-100 3.5.1-24 A wall; basemat; concrete Permeability, Cracking, Subsection IWL ring girders; Loss of Material (B.2.1.30) buttresses; (Spalling, Scaling) reinforcing steel Concrete: Dome; Reinforced Groundwater/Soil Increase in Porosity and ASME Section XI, II.A1.CP-100 3.5.1-24 A wall; basemat; Pressure Boundary concrete Permeability, Cracking, Subsection IWL ring girders; Loss of Material (B.2.1.30) buttresses; (Spalling, Scaling) reinforcing steel Concrete: Dome; Shelter, Protection Reinforced Groundwater/Soil Increase in Porosity and ASME Section XI, II.A1.CP-100 3.5.1-24 A wall; basemat; concrete Permeability, Cracking, Subsection IWL ring girders; Loss of Material (B.2.1.30) buttresses; (Spalling, Scaling) reinforcing steel

RS-14-165 Enclosure B Page 30 of 34 Table 3.5.2-4 Containment Structure (Continued)

Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Dome; Shielding Reinforced Groundwater/Soil Increase in Porosity and ASME Section XI, II.A1.CP-100 3.5.1-24 A wall; basemat; concrete Permeability, Cracking, Subsection IWL ring girders; Loss of Material (B.2.1.30) buttresses; (Spalling, Scaling) reinforcing steel Concrete: Dome; Structural Support Reinforced Groundwater/Soil Increase in Porosity and ASME Section XI, II.A1.CP-100 3.5.1-24 A wall; basemat; concrete Permeability, Cracking, Subsection IWL ring girders; Loss of Material (B.2.1.30) buttresses; (Spalling, Scaling) reinforcing steel

RS-14-165 Enclosure B Page 31 of 34 As a result of the response to RAI 3.5.1-1 provided in Enclosure A of this letter, LRA Section 3.5.2.2.1.1, page 3.5-24 is revised as shown below. Existing LRA text is shown in normal font.

Changes are highlighted with bolded italics for inserted text and strikethroughs for deleted text.

3.5.2.2.1.1 Cracking and Distortion due to Increased Stress Levels from Settlement; Reduction of Foundation Strength, and Cracking due to Differential Settlement and Erosion of Porous Concrete Subfoundations Cracking and distortion due to increased stress levels from settlement could occur in PWR and BWR concrete and steel containments. The existing program relies on ASME Section XI, Subsection IWL to manage these aging effects. Also, reduction of foundation strength and cracking, due to differential settlement and erosion of porous concrete subfoundations could occur in all types of PWR and BWR containments. The existing program relies on the structures monitoring program to manage these aging effects. However, some plants may rely on a de-watering system to lower the site ground water level. If the plants current licensing basis (CLB) credits a de-watering system to control settlement, the GALL Report recommends further evaluation to verify the continued functionality of the de-watering system during the period of extended operation.

Item Number 3.5.1-1 is applicable for BBS. The foundations of the Containment Structures at Byron and Braidwood Stations are supported on the underlying bedrock, as described in UFSAR Section 2.5.4.10.2.3 for Byron and UFSAR Section 2.5.4.10.1.1 for Braidwood. A settlement monitoring program was implemented during construction, and shortly thereafter, to monitor for settlement of Category I structure foundations at Braidwood. Predicted and measured values, for settlement of the Braidwood Containment Structures showed that plant settlement is complete and less than the values considered in the design of the structures. The results of calculations for settlement of the Byron Containment Structures showed negligible total and differential settlement. Therefore, cracking and distortion due to increased stress levels from settlement is not a significant aging effect expected to occur. The ASME Section XI, Subsection IWL (B.2.1.30) program will be used to manage cracking and distortion of the concrete dome, wall, basemat, and buttresses in accessible and inaccessible areas of the Containment Structure exposed to a groundwater and soil environment. HoweverIn addition, inaccessible below-grade Containment concrete surfaces will be examined by the Structures Monitoring (B.2.1.34) program when excavated for any reason.

Item Number 3.5.1-2 is not applicable to Byron and Braidwood Stations. Byron and Braidwood Stations do not have any porous concrete subfoundations, therefore, this aging effect and mechanism is not applicable.

RS-14-165 Enclosure B Page 32 of 34 As a result of the response to RAI 3.5.1-1 provided in Enclosure A of this letter, LRA Table 3.5.1 Item 3.5.1-1, page 3.5-43 is revised as shown below.

Existing LRA text is shown in normal font. Changes are highlighted with bolded italics for inserted text.

Table 3.5.1 Summary of Aging Management Evaluations for the Structures and Component Supports Further Item Aging Aging Management Component Evaluation Discussion Number Effect/Mechanism Programs Recommended 3.5.1-1 Concrete: dome; wall; Cracking and Chapter XI.S2, ASME Yes, if a de-watering Consistent with NUREG-1801.

basemat; ring girders; distortion Section XI, Subsection system is relied upon buttresses, Concrete due to increased IWL or Chapter XI.S6, to control settlement The ASME Section XI, Subsection IWL elements, all stress levels from "Structure Monitoring" (B.2.1.30) program will be used to settlement manage cracking and distortion of the If a de-watering system is concrete dome, wall, basemat, and relied upon for control of buttresses in accessible and settlement, then the inaccessible areas of the Containment licensee is to ensure Structure exposed to a groundwater and proper functioning of the soil environment.

de-watering system The Structures Monitoring (B.2.1.34) through the period of program will also be used to manage extended operation. cracking and distortion of the concrete dome, wall, basemat, and buttresses in inaccessible areas of the Containment Structure exposed to a groundwater and soil environment.

BBS do not rely upon a de-watering system to control settlement.

See subsection 3.5.2.2.1.1.

RS-14-165 Enclosure B Page 33 of 34 As a result of the response to RAI 3.5.1-1 provided in Enclosure A of this letter, LRA Table 3.5.2-4, which starts on page 3.5-131, is revised as shown below. For clarity purposes, the new line items are presented by themselves without any of the existing line items. The new additions are shown in bolded italics as individual line item additions and, therefore, appear different than the cascading table approach in the LRA.

Table 3.5.2-4 Containment Structure Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Dome; Flood Barrier Reinforced Groundwater/Soil Cracking and Distortion ASME Section XI, II.A1.CP-101 3.5.1-1 A wall; basemat; concrete Subsection IWL ring girders; (B.2.1.30) buttresses; reinforcing steel Concrete: Dome; HELB Shielding Reinforced Groundwater/Soil Cracking and Distortion ASME Section XI, II.A1.CP-101 3.5.1-1 A wall; basemat; concrete Subsection IWL ring girders; (B.2.1.30) buttresses; reinforcing steel Concrete: Dome; Missile Barrier Reinforced Groundwater/Soil Cracking and Distortion ASME Section XI, II.A1.CP-101 3.5.1-1 A wall; basemat; concrete Subsection IWL ring girders; (B.2.1.30) buttresses; reinforcing steel Concrete: Dome; Reinforced Groundwater/Soil Cracking and Distortion ASME Section XI, II.A1.CP-101 3.5.1-1 A wall; basemat; Pressure Boundary concrete Subsection IWL ring girders; (B.2.1.30) buttresses; reinforcing steel Concrete: Dome; Shelter, Protection Reinforced Groundwater/Soil Cracking and Distortion ASME Section XI, II.A1.CP-101 3.5.1-1 A wall; basemat; concrete Subsection IWL ring girders; (B.2.1.30) buttresses; reinforcing steel

RS-14-165 Enclosure B Page 34 of 34 Table 3.5.2-4 Containment Structure (Continued)

Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Concrete: Dome; Shielding Reinforced Groundwater/Soil Cracking and Distortion ASME Section XI, II.A1.CP-101 3.5.1-1 A wall; basemat; concrete Subsection IWL ring girders; (B.2.1.30) buttresses; reinforcing steel Concrete: Dome; Structural Support Reinforced Groundwater/Soil Cracking and Distortion ASME Section XI, II.A1.CP-101 3.5.1-1 A wall; basemat; concrete Subsection IWL ring girders; (B.2.1.30) buttresses; reinforcing steel