ML14063A495

Units 1 & 2, Response to NRC Requests for Additional Information, Set 13, Dated February 7, 2014 Re License Renewal Application

ML14063A495
Person / Time
Site:	Byron, Braidwood
Issue date:	03/04/2014
From:	Gallagher M P Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-14-052
Download: ML14063A495 (49)

Text

10 CFR 50 10 CFR 51 10 CFR 54 RS-14-052 March 04,2014 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

References:

Response to NRC Requests for Additional Information, Set 13, dated February 7, 2014 related to the Braidwood Station, Units 1 and 2 and Byron Station, Units 1 and 2 License Renewal Application 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 February 7,2014,"Requests 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 13 (TAC NOS.MF1879, MF1880, MF1881, AND MF1882)In the Reference 1 letter, 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 the Reference 2 letter, the NRC requested additional information to support the staffs'review of the LRA.Enclosure A contains the responses to this request for additional information.

Enclosure B contains updates to sections of the LRA (except for the License Renewal Commitment List)affected by the responses.

March 04, 2014 U.S.Nuclear Regulatory Commission Page 2 Enclosure C provides an update to the License Renewal Commitment List (LRA Appendix A, Section A.5).There are no other new or revised regulatory commitments contained in this letter.If you have any questions, please contact Mr.AI 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 C: License Renewal Commitment List Changes cc: Regional Administrator

-NRC Region III 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 EmergencyManagementAgency

-Division of Nuclear Safety RS-14-052 Enclosure A Page 1 of 17 Enclosure A Byron and Braidwood Stations, Units 1 and 2 License Renewal Application (LRA) updates resulting from the responses to the following RAIs:

RAI B.2.1.31-1 RAI B.2.1.31-2 RAI B.2.1.31-3 RAI B.2.1.27.1 RAI B.2.1.10-1

Note: To facilitate understanding, the original LRA pages have been repeated in this Enclosure, with revisions indicated. Existing LRA text is shown in normal font. Changes are highlighted with bold italics for inserted text and strikethroughs for deleted text.

RS-14-052 Enclosure A Page 2 of 17 RAI B.2.1.31-1 Applicability

Byron Station (Byron) and Braidwood Station (Braidwood)

Background

The Generic Aging Lessons Learned (GALL) Report aging management program (AMP) XI.S3 recommends that the ASME Section XI, Subsection IWF AMP augment the requirements of the existing ASME Section XI, Subsection IWF program (required in accordance with 10 CFR 50.55a) to include monitoring of high-strength structural bolting with actual measured yield strength greater than or equal to 150 ksi or 1,034 MPa for cracking. Several program elements of the GALL Report AMP XI.S3 specify recommendations for aging management of high-strength structural bolting:

1. The "Preventive Actions" program element of the GALL Report AMP XI.S3 recommends (1) using bolting material that has an actual measured yield strength less than 150 ksi; and (2) for structural bolting consisting of ASTM A325, ASTM F1852, and/or ASTM A490 bolts, the preventive actions for storage, lubricants, and stress corrosion cracking (SCC) potential discussed in Section 2 of Research Council for Structural Connections (RCSC) publication "Specification for Structural Joints Using ASTM A325 or A490 Bolts" need to be used. 2. The "Parameters Monitored or Inspected" program element recommends that high-strength structural bolting susceptible to SCC be monitored for SCC. 3. The "Detection of Aging Effects" program element recommends that, for high-strength structural bolting in sizes greater than 1" nominal diameter, volumetric examination should be performed in addition to the VT-3 examination to detect cracking and that this volumetric examination may be waived with adequate plant-specific justification.

Issue:

During an on site audit and review of the license renewal application (LRA) AMP, "ASME Section XI, Subsection IWF," the staff noted that the AMP states IWF supports at Byron and Braidwood do not use high-strength bolts susceptible to SCC. However, in discussions with the applicant during its onsite audit, the staff noted that there may be high-strength bolting (i.e., ASTM A490) in sizes greater than 1" diameter and actual yield strength greater than 150 ksi that is applicable to the IWF program but that was not considered for SCC potential, as recommended in the GALL Report AMP XI.S3. Specifically:

1. The AMP does not state whether the applicant plans to discontinue use of high-strength structural bolting (actual yield strenglh greater lhan 150 ksi).

2. If there are structural bolts that are high-strength and greater than 1" diameter, it is not clear if or how the applicant plans to manage cracking due to SCC in accordance with the recommendations of the GALL Report AMP XI.S3.

RS-14-052 Enclosure A Page 3 of 17 Request: 1. Identify whether there are high-strength structural bolts (i.e., ASTM A490) that were not previously identified for aging management of cracking due to SCC in accordance with the GALL Report AMP XI.S3. If ASTM A490 bolts are used but are not considered for SCC potential, provide technical justification for this exception to the recommendations of the GALL Report.

2. Describe how the recommendations in the "Preventive Actions," "Parameters Monitored or Inspected," and "Detection of Aging Effects" program elements are addressed, including the use of high-strength bolting materials, preventive actions for storage, lubricants, and SCC in accordance with Section 2 of the RCSC document and the VT-3 utilized to manage aging for SCC potential. If the program will not address the recommendations in the above-mentioned program elements for high-strength bolting or does not manage aging for these components, provide the associated technical justification.

Exelon Response:

1. There are high strength structural bolts (i.e., ASTM A490) that were not previously identified in the LRA for aging management of cracking due to SCC in accordance with GALL Report Section XI.S3. ASTM A490 bolts are used for ASME Class 1 component supports as listed in UFSAR Table B.9-1, Material for NSSS Component Supports. Some of these bolts are greater than one-inch in diameter. No special aging management methods were originally identified in the LRA for aging management of cracking due to SCC for ASTM A490 bolts, beyond current ASME Code requirements, based upon our original understanding of the GALL Report AMP XI.S3 and NUREG-1950, Disposition of Public Comments and Technical Bases for Changes in the License Renewal Guidance Documents NUREG-1801 and NUREG-1800. Upon further review, the LRA is revised to include aging management of cracking due to SCC for ASTM A490 bolts in accordance with GALL Report Section XI.S3, as described in the response to

Request #2. ASME SA 540 bolting materials were also used for ASME Class 1 component supports and are separately addressed in the response to RAI B.2.1.31-3. In order to identify in the LRA that there are high strength structural bolts (i.e., ASTM A490) that were not previously identified for aging management of cracking due to SCC in accordance with GALL Report Section XI.S3, LRA Item Number 3.5.1-68 in Table 3.5.1, Summary of Aging Management Evaluations for the Structures and Component Supports, is revised to include all high strength bolting, not just ASME SA 540 high strength structural bolting, which was the only bolting material originally mentioned in the Discussion for this Item Number, as shown in Enclosure B.

2. As described below, the ASME Section XI, Subsection IWF aging management program is revised to follow the recommendations of the GALL Report AMP XI.S3 as described in the "Preventive Actions," "Parameters Monitored or Inspected," and "Detection of Aging Effects" program elements.

RS-14-052 Enclosure A Page 4 of 17 Element #2, Preventive Actions The "Preventive Actions" program element of GALL Report AMP SI.X3 recommends (1) using bolting material that has an actual measured yield strength less than 150 ksi; and (2) for structural bolting consisting of ASTM A325, ASTM F 1852, and/or ASTM A490 bolts, using the preventive actions for storage, lubricants, and stress corrosion cracking potential discussed in Section 2 of Research Council for Structural Connections (RCSC) publication "Specification for Structural Joints Using ASTM A325 or A490 Bolts". Implementing documents within the scope of the ASME Section XI, Subsection IWF aging management program will be enhanced to include the recommendations in the "Preventive Actions" program element of the GALL Report AMP XI.S3 for high strength bolts, with respect to the use of high-strength bolting materials, preventive actions for storage, lubricants, and stress corrosion cracking in accordance with Section 2 of the RCSC document as follows:

Revise implementing documents to provide guidance for specification of bolting material, storage, lubricants and sealants, and installation torque or tension to prevent or mitigate degradation and failure of structural bolting. Bolting material with actual measured yield strength of 150 ksi or greater shall not be used in plant changes without engineering approval, due to consideration of stress corrosion cracking vulnerability.

Revise implementing documents to specify storage requirements for high strength bolts that include the recommendations of the Research Council for Structural Connections, "Specification for Structural Joints Using ASTM A325 or A490 Bolts",

Section 2.

Revise implementing documents to specify that lubricants that contain molybdenum disulfide (MoS

2) shall not be applied to high strength structural bolts within the scope of license renewal. This issue is also addressed in the response to RAI B.2.1.31-2. As a result of the response to RAI B.2.1.31-1, LRA Appendix A, Section A.2.1.31 and Appendix B, Section B.2.1.31 are revised as shown in Enclosure B, and LRA Table A.5, Item 31, is revised as shown in Enclosure C. This revision provides more detail in Enhancement #2 to describe how the enhancements regarding preventative actions will be implemented to address the recommendations of the GALL Report AMP XI.S3 as described in the "Preventive Actions" program element. The response to RAI B.2.1.31-2 also revises Enhancement #2 of this program.

Element #3, Parameters Monitored or Inspected The "Parameters Monitored or Inspected" program element of GALL Report AMP SI.X3 recommends that high-strength structural bolting susceptible to stress corrosion cracking (SCC) be monitored for SCC. A description, of how the recommendations in the "Parameters Monitored or Inspected" program element of the GALL Report AMP XI.S3 are addressed regarding stress corrosion cracking of high strength bolts, is as follows.

RS-14-052 Enclosure A Page 5 of 17 Line items that address cracking of high strength bolts within the scope of the ASME Section XI, Subsection IWF aging management program already exist in LRA Table 3.5.2-3, Component Supports Commodity Group- Summary of Aging Management Evaluation. Therefore, no additional line items or changes are required for LRA Table

3.5.2-3. LRA Item Number 3.5.1-68 in Table 3.5.1, Summary of Aging Management Evaluations for the Structures and Component Supports, is revised as discussed in the response to

Request #1 to include all high strength bolting, not just ASME SA 540 high strength structural bolting, which was the only bolting material originally mentioned in the Discussion section for this Item Number, as shown in Enclosure B. As discussed below in the discussion of Element #4, "Detection of Aging Effects", periodic visual examinations that include parameters and criteria to detect a corrosive environment that supports SCC potential for high strength bolting greater than one-inch nominal diameter will be included as Enhancement #5 to the ASME Section XI, Subsection IWF program. The periodic visual examinations for high strength bolting greater than one-inch nominal diameter will include parameters and criteria to identify if the bolting has been exposed to moisture or other contaminants by evidence of moisture, residue, foreign substance, or corrosion. Conditions identified during the periodic visual examinations that identify a potential corrosive environment that supports SCC will be entered into the corrective action program (CAP) and dispositioned as discussed below in the discussion of Element #4, "Detection of Aging Effects". As a result of the response to RAI B.2.1.31-1, LRA Appendix A, Section A.2.1.31 and Appendix B, Section B.2.1.31 are revised as shown in Enclosure B, and LRA Table A.5, Item 31, is revised as shown in Enclosure C. Enhancement #5 is added to describe how this program provides for periodic visual inspections and detection of aging effects to address the recommendations of the GALL Report AMP XI.S3 as described in the "Parameters Monitored or Inspected" program element. The response to RAI B.2.1.31-3 also addresses Enhancement #5 of this program. Element #4, Detection of Aging Effects The "Detection of Aging Effects" program element of GALL Report AMP SI.X3 recommends that, for high-strength structural bolting in sizes greater than one-inch nominal diameter, volumetric examination should be performed in addition to a VT-3 examination to detect cracking, and that this volumetric examination may be waived with adequate plant-specific justification. The element goes on to add that other structural bolting (ASTM A-325, ASTM F1852, and ASTM A490 bolts) and anchor bolts are monitored for loss of material, loose or missing nuts, and cracking of concrete around the anchor bolts. Details of how the recommendations in the "Detection of Aging Effects" program element of the GALL Report AMP XI.S3 are addressed regarding stress corrosion cracking of high strength bolts and how periodic visual examinations will be utilized to manage aging for SCC potential of high strength bolts is as follows. Plant-specific history, on volumetric examination of high strength bolts greater than one-inch nominal diameter and periodic visual examinations to detect a corrosive RS-14-052 Enclosure A Page 6 of 17 environment with supplemental volumetric examinations if warranted, is used to justify taking an exception to the GALL Report recommendation that periodic volumetric examinations be performed. One-time volumetric examinations will be performed on a sample of ASTM A490 bolts, greater than one-inch nominal diameter, for the detection of stress corrosion cracking prior to the period of extended operation. These volumetric examinations together with

the extensive volumetric examinations that have been performed on the ASME SA 540 reactor head closure studs discussed in the response to RAI B.2.1.31-3 and periodic visual examinations discussed below are used to justify taking an exception to the GALL Report recommendation that periodic volumetric examination be performed to manage SCC. Volumetric examinations will be performed in accordance with the requirements of ASME Code Section XI, Appendix VIII, Supplement 8. The sample will consist of bounding and representative A490 bolt sizes, joint configurations, and environmental exposure conditions. The sample will consist of 20% of the ASTM A490 bolts greater than one-inch nominal diameter or a maximum of 25 ASTM A490 bolts total for both Byron and Braidwood stations. The selection of the samples will consider susceptibility to stress corrosion cracking (e.g., actual measured yield strength) and ALARA principles.

Any adverse results of the volumetric examinations will be entered into the corrective action program and will be evaluated by engineering to determine if additional actions are warranted such as expansion of sample size, scope, and frequency of any additional

supplemental visual or volumetric examinations, as well as any code requirements specified by ASME Section XI, Subsection IWF. The performance of the volumetric examinations of the ASTM A490 bolts prior to PEO is Enhancement #4 to the ASME Section XI, Subsection IWF program and is used to support the justification for Exception 2 in Appendix B, Section B.2.1.31, ASME Section XI, Subsection IWF as shown in Enclosure B.

The sample will consist of 20% of the ASTM A490 bolts greater than one-inch nominal diameter or a maximum of 25 ASTM A490 bolts total, for both Byron and Braidwood stations. A single population for both stations is considered adequate for sampling ASTM A490 bolts for the following reasons: Common specifications and drawings were used for the construction of both stations and for all 4 units. The stations were constructed as part of a continuous construction effort. The high strength bolts used for supports within the scope of the ASME Section XI, Subsection IWF aging management program are carbon steel bolts, so SCC would also exhibit surface corrosion that can be detected through visual examinations. ASTM A490 bolts in civil structures are not prone to SCC. Plant specific OE did not reveal any broken bolts due to SCC of ASTM A490 bolts at Byron and Braidwood. Periodic visual examinations that include parameters and criteria to detect a corrosive environment that supports SCC potential for all high strength bolting greater than one-inch nominal diameter will be included as Enhancement #5 to the ASME Section XI, Subsection IWF program. The periodic visual examinations for high strength bolting greater than one-inch nominal diameter will include parameters and criteria to identify if the bolting has been exposed to moisture or other contaminants by evidence of moisture, residue, foreign substance, or corrosion. The periodic visual examinations will RS-14-052 Enclosure A Page 7 of 17 be performed such that 100% of the accessible high strength bolting greater than one-inch nominal diameter within the scope of the ASME Section XI, Subsection IWF program, will be examined prior to the period of extended operation, and then each inspection interval of 10 years thereafter. Conditions identified during the periodic visual examinations that identify a potential corro sive environment that supports SCC will be entered into the corrective action program (CAP) and dispositioned as discussed below. Adverse conditions identified during the periodic visual examinations will be entered into CAP and will be evaluated by engineering to determine if the bolt has been exposed to a corrosive environment with the potential to cause SCC. The conditions will be subjected to supplemental visual examination or analysis of residue for additional information to determine if there is a potential for SCC. The bolts determined to have been exposed to an environment with the potential to cause SCC will be included in a sample population for each specific bolt material where SCC is a concern. A sample size equal to 20 percent (rounded up to the nearest whole number) of the bolts in the sample population, with a maximum sample size of 25 bolts will be subject to supplemental volumetric examination to determine if SCC is present. The selection of the samples will consider susceptibility to stress corrosion cracking (e.g., actual measured yield strength) and ALARA principles. These supplemental volumetric examinations will be performed in accordance with the requirements of ASME Code Section XI, Appendix VIII, Supplement 8. The results of the volumetric examinations will be evaluated by engineering to determine if additional actions are warranted such as expanding the sample population, scope, and frequency of any additional supplemental visual or volumetric examinations, as well as any code requirements specified by ASME Section XI, Subsection IWF. In addition to the above, evaluations will be performed utilizing CAP to determine if any other corrective actions may be required, such as identifying and correcting the source of the condition, cleansing or cleanup that may be needed to contain and/or eliminate the corrosive environment, and any actions that could be taken to prevent a recurrence of the condition. The periodic visual examinations and follow-up actions discussed above are used to support the justification for Exception 2 in Appendix B, Section B.2.1.31, ASME Section XI, Subsection IWF as shown in Enclosure B. Since all of the ASME Section XI, Subsection IWF program components utilizing high strength bolting are located within the same confined area of the secondary shield wall, they share a common environment and have a low potential to be exposed to a corrosive environment due to the limited components contained in the area. In addition, other programs such as the Boric Acid Corrosion (B.2.1.4), ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD (B.2.1.1), Closed Treated Water Systems (B.2.1.12), and External Surfaces Monitoring of Mechanical Components (B.2.1.23) aging management programs, as well as leakage monitoring required by Technical Specifications provide additional assurance that any changes to current environmental conditions, should they occur, will be identified and appropriate actions taken throughout the IWF interval and period of extended operation. SCC has not been observed as part of the past IWF examinations, and no cracked or broken ASTM A490 bolts, within the scope of the ASME Section XI, Subsection IWF program, have been identified at Byron and Braidwood. The periodic visual examinations, as part of the ASME Section XI, Subsection IWF program discussed above, that are conducted prior to the period of extended operation, will identify any such conditions so they will be evaluated prior to entering the period of extended operation. The periodic visual examinations, combined with the volumetric examinations of a sample of ASTM A490 bolts performed prior to RS-14-052 Enclosure A Page 8 of 17 PEO, will be effective in detecting SCC or a corrosive environment with the potential for SCC throughout the period of extended operation. The Enhancements 4 and 5 to the ASME Section XI, Subsection IWF program, discussed above, provide reasonable assurance that the potential for cracking due to SCC and exposure to a corrosive environment that supports SCC for ASTM A490 bolts will be adequately managed by the ASME Section XI, Subsection IWF program so that the potential for SCC of ASTM A490 bolts will be detected and corrected prior to a loss of function. As a result of the response to RAI B.2.1.31-1, LRA Appendix A, Section A.2.1.31 and Appendix B, Section B.2.1.31 are revised as shown in Enclosure B, and LRA Table A.5, Item 31, is revised as shown in Enclosure C. Enhancements 4 and 5 are added to describe how plant specific volumetric examination of ASTM A490 bolts and how periodic visual inspections and detection of aging effects will be implemented to address the recommendations of the GALL Report AMP XI.S3 as described in the "Detection of Aging Effects" program element. The response to RAI B.2.1.31-3 also addresses Enhancement #5 of this program.

RS-14-052 Enclosure A Page 9 of 17 RAI B.2.1.31-2 Applicability

Byron and Braidwood

Background

The "Preventive Actions" program element of the GALL Report AMP XI.S3 states that the use of molybdenum disulfide (MoS

2) as a lubricant is a potential contributor to SCC, especially when applied to high-strength bolting. The applicant's ASME Section XI, Subsection IWF AMP basis

document states that MoS 2 was used as a lubricant for faying surfaces of NSSS supports but not as a thread lubricant.

Issue:

There is no enhancement to the program to specifically prohibit the use of MoS 2 lubricants on structural bolting. It is not clear to the staff whether the applicant plans to prohibit the use of MoS 2 lubricant for structural bolting in the future.

Request:

State whether the program will be enhanced to specifically prohibit the use of MoS 2 on structural bolting. If so, update the LRA and updated final safety analysis report supplement to include this enhancement. If not, state how the program will ensure that MoS 2 lubricant is not used or that it will not be a potential contributor to SCC.

Exelon Response

As stated in the Section XI, Subsection IWF AMP basis document, molybdenum disulfide (MoS 2) is not used as a thread lubricant at Byron and Braidwood. Enhancement #2 of the program specifies that additional guidance for selection of proper lubricants will be provided. As a result of the response to this RAI, Enhancement #2 is revised to clarify that the use of lubricants containing MoS 2 for structural bolting is prohibited. As a result of this change, LRA Appendix A, Section A.2.1.31 and Appendix B, Section B.2.1.31 are revised as shown in Enclosure B of this response. In addition, the Byron and Braidwood LRA Table A.5 Commitment List, Item 31, is revised as shown in Enclosure C. The response to RAI B.2.1.31-1 also revises Enhancement #2 of this program.

RS-14-052 Enclosure A Page 10 of 17 RAI B.2.1.31-3 Applicability

Byron and Braidwood

Background

The "Detection of Aging Effects" program element of the GALL Report AMP XI.S3 recommends that for high-strength structural bolting (actual measured yield strength greater than 150 ksi) in sizes greater than one-inch diameter, volumetric examination should be performed in addition to VT-3 examination. The GALL Report also states that this volumetric examination may be waived with adequate plant-specific justification.

LRA Section B.2.1.31 states that for the 5" diameter high strength reactor coolant pump (RCP) hold-down bolts at Byron and Braidwood and the 1-1/2" diameter pressurizer hold-down bolts at Braidwood, the applicant takes exception to the GALL Report recommendation that periodic volumetric examinations be performed. The staff reviewed LRA Section B.2.1.31 ASME Section XI, Subsection IWF AMP supporting documentation during the onsite audit and noted that the applicant does not consider cracking due to SCC applicable to these bolts. The applicant uses the following plant-specific justification to waive the GALL Report-recommended volumetric examinations in addition to VT-3 visual examination:

The bolt design is in a configuration that precludes water from penetrating the interface between the bolt head and support surface and seeping beneath the bolt head, which prevents the potential initiation of corrosion. The bolts were torqued to bear tightly on the support surface. Metal-plated stud bolting is not used, which could cause degradation due to corrosion or hydrogen embrittlement. An approved lubricant was applied to the bolts; this lubricant did not contain MoS

2. There have been no recordable indications of degradation identified by ASME Section XI, Subsection IWF program examinations that would indicate the potential for SCC to occur.

Issue: The staff reviewed the applicant's plant-specific justification to waive volumetric examinations of the RCP hold-down bolts and pressurizer hold-down bolts, and the applicant's plan to use visual examinations only to manage aging of these components. The staff identified the following concerns:

The ASME Section XI, Subsection IWF AMP basis documents state that the RCP hold-down bolts are located in an "air with borated water leakage" environment. Since there is a potentially moist environment, susceptible material, and stress present to cause SCC, the GALL Report AMP XI.S3 recommends that high-strength bolting in sizes greater than 1" should be managed for SCC. An onsite review of the design drawings for the bolt configuration determined that there is no physical seal preventing water intrusion beneath the bolt head. The staff does not have enough information to confirm that the surface between the bolt head and support surface is watertight.

RS-14-052 Enclosure A Page 11 of 17 The AMP basis document states that the applicant examines the bolts using ASME Section XI, Subsection IWF Table IWF-2500-1, which states that for supports other than piping supports (class 1, 2, 3 or MC), VT-3 examination of 100 percent of the bolts should be performed each inspection interval of 10 years. The staff needs more information on how the VT-3 examination will ensure that SCC will be detected and that any effects of cracking due to SCC will be managed. The AMP does not indicate what parameters or criteria would be used to detect SCC, and how they would be effective in identifying SCC potential. The program does not identify actions to be taken (i.e., use ASME IWF criteria for expansion of scope, increase in inspection frequency, or perform volumetric examinations) if there are indications that SCC could be occurring. The applicant's previous experience with the IWF program indicates that cracking due to SCC has not been found to be a degradation mechanism. However, since the IWF examination does not include volumetric examination for cracking beneath the bolt head for high-strength structural bolts greater than 1" diameter, the operating experience referenced by the applicant does not preclude the potential for SCC for these components. During the onsite audit, the applicant stated that it does not have a history of volumetric examinations of similar bolting to show that there is no evidence of SCC.

Request:

Provide further technical justification to support a plant-specific waiver for periodic volumetric examination of high strength RCP and pressurizer hold-down bolts. Discuss how the ASME Section XI, Subsection IWF program will verify the absence of cracking due to SCC for the 5" SA540 high strength RCP hold-down bolts and the 11/2" pressurizer hold-down bolts.

Specifically:

1. For both plants, provide results of any plant-specific history of volumetric examination of high strength bolts in a similar environment to support a plant-specific justification to waive future volumetric examinations as recommended in the GALL Report. If there is no history of volumetric examination of the referenced bolts, state whether any volumetric examinations (or alternative method) will be conducted prior to period of extended operation (PEO) to confirm that cracking due to SCC has not affected the bolt threads.

2. State what parameters or criteria will be used to detect SCC or a corrosive environment and how visual inspections will be effective in detecting future SCC or corrosive environment.

State how the program will ensure that a noncorrosive environment is maintained throughout the IWF interval.

3. State what actions will be taken with respect to augmented examinations if inspections result in indications that there is degradation or a corrosive environment that could lead to SCC, including any plans for supplemental volumetric examination or evaluations.

RS-14-052 Enclosure A Page 12 of 17 Exelon Response:

1. Byron and Braidwood have an extensive history of volumetric examinations of the unpainted reactor head closure studs as described under the Reactor Head Closure Stud Bolting (B.2.1.3) aging management program. The reactor head closure stud material at Byron and Braidwood is ASME SA 540. The material of the ASME Section XI, Subsection IWF high strength five-inch RCP hold-down bolts at Byron and Braidwood and the 1.5-inch pressurizer hold-down bolts at Braidwood is also ASME SA 540. One hundred percent of the tensioned reactor head closure stud population has been subject to volumetric examination each ten-year inservice inspection interval (more than 500 volumetric examinations total for both stations), with no evidence of SCC identified. Because of the similar materials and environmental conditions, the numerous reactor head closure stud volumetric examinations with no evidence of SCC identified can be used to support a plant-specific justification to waive performing periodic volumetric examinations of ASME SA 540 bolting material within the scope of the ASME Section XI, Subsection IWF program as proposed in the GALL Report. This information is used to support the justification for Exception 2 in Appendix B, Section B.2.1.31, ASME Section XI, Subsection IWF as shown in Enclosure B.

2. Periodic visual examinations that include parameters and criteria to detect a corrosive environment that supports SCC potential for all high strength bolting greater than one-inch nominal diameter will be included as Enhancement #5 to the ASME Section XI, Subsection IWF program. The periodic visual examinations for high strength bolting greater than one-inch nominal diameter will include parameters and criteria to identify if the bolting has been exposed to moisture or other contaminants by evidence of moisture, residue, foreign substance, or corrosion. Conditions identified during the periodic visual examinations that identify a potential corrosive environment that supports SCC will be entered into the corrective action program (CAP) and dispositioned as discussed below in response to Request #3. The periodic visual examinations will be performed such that 100% of the accessible high strength bolting greater than one-inch nominal diameter within the scope of the ASME Section XI, Subsection IWF program, will be examined prior to the period of extended operation, and then each inspection interval of 10 years thereafter. Since all of the ASME Section XI, Subsection IWF program components utilizing high strength bolting are located within the same confined area of the secondary shield wall, they share a common environment and have a low potential to be exposed to a corrosive environment due to the limited components contained in the area. In addition, other programs such as the Boric Acid Corrosion (B.2.1.4), ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD (B.2.1.1), Closed Treated Water Systems (B.2.1.12), and External Surfaces Monitoring of Mechanical Components (B.2.1

.23) aging management programs, as well as leakage monitoring required by Technical Specifications provide additional assurance that any changes to current environmental conditions, should they occur, will be identified and appropriate actions taken throughout the IWF interval and period of extended operation.

SCC has not been observed as part of past IWF examinations, and no cracked or broken ASME SA 540 bolts, within the scope of the ASME Section XI, Subsection IWF program, have been identified at Byron and Braidwood. The periodic visual examinations, as part of the ASME Section XI, Subsection IWF program discussed above, that are conducted prior to the period of extended operation, will identify any such conditions so they will be evaluated prior to entering the period of extended operation. The periodic visual RS-14-052 Enclosure A Page 13 of 17 examinations, combined with the actions described below in response to Request 3, will be effective in detecting SCC or a corrosive environment with the potential for SCC throughout the period of extended operation.

3. Adverse conditions identified during the periodic visual examinations will be entered into CAP and will be evaluated by engineering to determine if the bolt has been exposed to a corrosive environment with the potential to cause SCC. The conditions will be subjected to supplemental visual examination or analysis of residue for additional information to determine if there is a potential for SCC. The bolts determined to have been exposed to an environment with the potential to cause SCC will be included in a sample population for each specific bolt material where SCC is a concern. A sample size equal to 20 percent (rounded up to the nearest whole number) of the bolts in the sample population, with a maximum sample size of 25 bolts will be subject to supplemental volumetric examination to determine if SCC is present. The selection of the samples will consider susceptibility to stress corrosion cracking (e.g., actual measured yield strength) and ALARA principles.

These supplemental volumetric examinations will be performed in accordance with the requirements of ASME Code Section XI, Appendix VIII, Supplement 8. The results of the volumetric examinations will be evaluated by engineering to determine if additional actions are warranted such as expanding the sample population, scope, and frequency of any additional supplemental visual or volumetric examinations, as well as any code requirements specified by ASME Section XI, Subsection IWF. In addition to the above, evaluations will be performed utilizing CAP to determine if any other corrective actions may be required, such as identifying and correcting the source of the condition, cleansing or cleanup that may be needed to contain and/or eliminate the corrosive environment, and any actions that could be taken to prevent a recurrence of the condition. The periodic visual examinations and follow-up actions discussed above are used to support the justification for Exception 2 in Appendix B, Section B.2.1.31, ASME Section XI, Subsection IWF as shown in Enclosure B.

The enhancement to the ASME Section XI, Subsection IWF program as discussed in items 2 and 3 above provides reasonable assurance that the periodic visual examinations will identify a corrosive environment that supports SCC and the supplemental volumetric examinations performed will detect SCC such that the potential for cracking due to SCC will be adequately managed by the ASME Section XI, Subsection IWF program.

As a result of the response to RAI B.2.1.31-3, LRA Appendix A, Section A.2.1.31 and Appendix B, Section B.2.1.31 are revised as shown in Enclosure B, and LRA Table A.5, Item 31, is revised as shown in Enclosure C, to add information related to plant specific operating experience for volumetric examinations, and to add Enhancement #5, describing the periodic visual inspections for the detection of aging effects. The response to RAI B.2.1.31-1 also addresses Enhancement #5 of this program.

RS-14-052 Enclosure A Page 14 of 17 RAI B.2.1.27-1 Applicability

Byron and Braidwood

Background

Both Byron and Braidwood have operating experience where the coupon tree holding the Boral sample coupons was not surrounded by freshly discharged fuel in accordance with the original equipment manufacturer's recommendations.

Issue:

In order to have an effective coupon monitoring program, the coupons should be the leading indicators of material degradation as compared to the neutron absorber material in the spent fuel storage racks. That is, the dose received and/or long-term exposure to the wet pool environment by the coupons should be bounding of the material in the racks. Allowing the coupons to lead the neutron absorber material in the racks provides reasonable assurance that the applicant will detect any material degradation in the coupons before the material in the spent fuel pool racks starts to degrade.

Request:

Please discuss how the coupon exposure (i.e., coupon tree location) will provide reasonable assurance that Boral degradation is identified prior to potential loss of neutron-absorbing capability of the material in the spent fuel racks. If the coupon exposure to the environment is not bounding of the material in the racks, discuss how the aging effects of the Boral material will be managed for the unbounded racks.

Exelon Response

Procedural control of the location and the loading of freshly discharged fuel around the spent fuel rack Boral coupon tree will provide reasonable assurance that Boral degradation is identified prior to potential loss of neutron-absorbing capability of the spent fuel rack material.

Reasonable assurance will be achieved by ensuring that the environment of the Boral coupon tree will be bounding of the Boral material in the spent fuel racks. The coupon exposure practice at Byron and Braidwood Stations was established to meet the spent fuel rack manufacturer recommendations. The original recommendation from the rack manufacturer was to place the coupon tree in a Region 2 rack and surround it with freshly discharged fuel assemblies following each of the first five (5) operating cycles of a given unit after rack installation. Region 2 racks contain only one (1) Boral panel separating any two (2) adjacent storage cells, whereas Region 1 racks are constructed such that two (2) Boral panels exist between any two (2) adjacent storage cells. This accelerated irradiation schedule was intended to ensure that the Boral coupons experienced a higher radiation dose than the Boral panels in the storage racks. Following the fifth accelerated exposure cycle, the fuel assemblies surrounding the test coupon tree could remain in place for the remaining life of the racks.

According to the licensing report for installation of the racks, which contains the RS-14-052 Enclosure A Page 15 of 17 recommendation for accelerated irradiation, it is stated that, "Over the duration of the coupon testing program, the coupons will have accumulated more radiation dose than the expected

lifetime dose for normal storage cells."

At Byron, the coupon tree was initially surrounded with freshly discharged fuel following the first Unit 2 refueling outage after rerack. The coupon tree was surrounded on four (4) sides for approximately 10 months in a Region 1 rack. Following the second Unit 2 refueling outage after rerack, the coupon tree was surrounded with freshly discharged Unit 2 fuel on all eight (8) sides for approximately 11 months, also in a Region 1 rack. For the remainder of the second cycle, the coupon tree was surrounded on five (5) sides with more recently discharged Unit 1 fuel in a Region 1 rack. The inappropriate placement of the coupon tree in a Region 1 rack, rather than a Region 2 rack as contained in the manufacturer's recommendations, was identified and entered into the corrective action program. Following the third, fourth, fifth, and sixth Unit 2 refueling cycles after rerack, the coupon tree was surrounded on all eight (8) sides in Region 2 racks by freshly discharged fuel from Unit 2.

At Braidwood, the coupon tree was surrounded with freshly discharged fuel following the first Unit 1 refueling outage after rerack. The coupon tree was surrounded on all eight (8) sides in a Region 2 rack for approximately three (3) months. After three (3) months, three (3) assemblies were removed, with the other five (5) remaining in place for approximately one (1) additional year. The condition was identified and entered into the corrective action program. As a result, freshly discharged fuel from the more recent Unit 2 outage was placed on all eight (8) sides of the coupon tree until the next scheduled Unit 1 refueling outage. It was determined that placement of the more recently discharged Unit 2 fuel around the coupon tree for the remainder of the cycle compensated for the absence of the three (3) assemblies. Beginning with the second Unit 1 refueling outage after rerack, the coupon tree was surrounded on all eight (8) sides by freshly discharged fuel from Unit 1 for four (4) consecutive refueling cycles.

The coupon tree was relocated to different rack locations throughout the accelerated irradiation schedule at both stations. This was done to avoid repetitive placement of freshly discharged fuel assemblies in the same cells, which would have allowed the cells to receive a similar exposure as the coupon. Since the 2005-2006 timeframe, in order to levelize the heat load of the spent fuel pool, freshly discharged fuel is scatter loaded throughout the spent fuel pool Region 2 racks after each refueling outage. Freshly discharged fuel is loaded such that only one (1) freshly discharged assembly is face-adjacent and one (1) freshly discharged assembly is diagonal-adjacent to a particular spent fuel rack cell. Although not in strict compliance with the manufacturer's recommendations, the accelerated irradiation schedules implemented at Byron and Braidwood Stations are considered to have met the intent of the manufacturer's recommendations. Therefore, it is reasonable to conclude that the coupons have obtained a radiation exposure condition currently bounding of any other storage rack cell locations.

The program will be enhanced to ensure that the Boral coupon exposure is bounding of the racks prior to testing of the coupons through the end of the period of extended operation. Boral coupon exposure will be maintained bounding of all rack locations by ensuring that the coupons have been surrounded with a greater number of freshly discharged fuel assemblies than that of any other cell location.

RS-14-052 Enclosure A Page 16 of 17 LRA Appendix A, Section A.1.1 and A.2.1.27, Appendix B, Section B.1.5 and B.2.1.27, are revised as shown in Enclosure B to include the new enhancement. LRA Table A.5, Item 27, is

revised as shown in Enclosure C. RAI B.2.1.10-1 Applicability

Byron and Braidwood

=

Background===

LRA Section B.2.1.10 Enhancement 1 provides three options the applicant may take to disposition potential primary water stress corrosion cracking (PWSCC) of the Byron and Braidwood Units 1 and 2 steam generator divider plate welds to the primary head and tubesheet cladding. The second option for Enhancement 1 indicates that an analytical evaluation will be performed to establish a technical basis to disposition the potential degradation mechanism.

Option 2: Analysis Perform an analytical evaluation of the steam generator divider plate welds in order to establish a technical basis which concludes that the steam generator reactor coolant pressure boundary is adequately maintained with the presence of steam generator divider plate weld cracking. The analytical evaluation will be submitted to the NRC for review and approval prior to entering associated PEO.

Option 2: Analysis - Susceptibility Perform an analytical evaluation of the steam generator tube-to-tubesheet welds to determine that the welds ar e not susceptible to primary water stress corrosion cracking. The evaluat ion for determining that the tube-to-tubesheet welds are not susceptible to primary water stress corrosion cracking will be submitted to the NRC for review and approval prior to entering the associated PEO.

Option 3: Analysis - Pressure Boundary Perform an analytical evaluation of the steam generator tube-to-tubesheet welds redefining the reactor coolant pressure boundary of the tubes, where the steam generator tube-to-tubesheet welds are not required to perform a reactor coolant pressure boundary function. The redefinition of the reactor coolant pressure boundary will be submitted to the NRC for review and approval prior to entering the associated PEO.

In the case of the applicant choosing Option 2 for Enhancement 1 and Option 2 or 3 for Enhancement 2, the staff is to review and approve the analysis prior to the Byron and Braidwood Units entering its respective PEO.

Issue:

The applicant did not provide a period when the analysis will be provided to the staff for review and approval. The LRA states that the analysis will be provided before the PEO. In order for the RS-14-052 Enclosure A Page 17 of 17 staff to complete its review of such an analysis before the PEO, adequate time needs to be provided for the review.

Request:

Please provide a period by which the analytical evaluation will be provided to the staff such that the staff will have adequate time to review it before Byron and Braidwood enters PEO.

Exelon Response:

Steam Generators aging management program Enhancement 1, Option 2 and Enhancement 2, Options 2 and 3 are revised to specify that analyses requiring NRC review and approval will be submitted two (2) years prior to entering the associated period of extended operation.

Changes to LRA Appendix A section A.2.1.10, Appendix B section B.2.1.10, and Appendix A.5, commitment 10, are included in Enclosures B and C.

RS-14-052 Enclosure B Page 1 of 22 Enclosure B Byron and Braidwood Stations, Units 1 and 2 License Renewal Application (LRA) updates resulting from the responses to the following RAIs:

RAI B.2.1.31-1 RAI B.2.1.31-2 RAI B.2.1.31-3 RAI B.2.1.27.1 RAI B.2.1.10-1

Note: To facilitate understanding, the original LRA pages have been repeated in this Enclosure, with revisions indicated. Existing LRA text is shown in normal font. Changes are highlighted with bold italics for inserted text and strikethroughs for deleted text.

RS-14-052 Enclosure B Page 2 of 22 As a result of the response to RAI B.2.1.31-1 provided in Enclosure A of this letter, the Discussion for Item Number 3.5.1-68 i n LRA Table 3.5.1, Summary of Aging Management Evaluations for the Structures and Component Supports, page 3.5-67, is revised as shown below.

Deletions are shown with strikethroughs. Table 3.5.1 Summary of Aging Management Evaluations for the Structures and Component Supports (Continued)

Item Number Component Aging Effect/Mechanis

m Aging Management Programs Further Evaluation

Recommended Discussion 3.5.1-68 High-strength structural bolting Cracking due to stress

corrosion cracking Chapter XI.S3, "ASME Section XI, Subsection IWF" No Consistent with NUREG-1801 with exceptions. The ASME Section XI, Subsection IWF (B.2.1.31) program will be used to manage cracking of SA540 high strength structural bolting for NSSS component supports exposed to an air with borated leakage environment.

Exceptions apply to the NUREG-1801 recommendations for ASME Section XI, Subsection IWF (B.2.1.31) implementation.

RS-14-052 Enclosure B Page 3 of 22 As a result of the response to RAI B.2.1.27-1 provided in Enclosure A of this letter, LRA Appendix A, Sections A.1.1, page A-6, and A.2.1.27, page A-30, are revised as shown below.

Additions are indicated with bolded italics.

A.1.1 NUREG-1801 Chapter XI Aging Management Programs The Byron and Braidwood NUREG-1801 Chapter XI Aging Management Programs (AMPs) are described in this section. The AMPs are either existing, existing with enhancements (enhanced) or new. The following list reflects the status of these programs at the time of the License Renewal Application (LRA) submittal. Commitments for program additions and enhancements are identified in the Appendix A.5 License Renewal Commitment List. 27. Monitoring of Neutron-Absorbing Materials Other than Boraflex (Section A.2.1.27) [Existing

- Requires Enhancement

] A.2.1.27 Monitoring of Neutron-Absorbing Materials Other than Boraflex The Monitoring of Neutron-Absorbing Materials Other than Boraflex aging management program is an existing condition monitoring program that periodically inspects and analyzes test coupons of the Boral material in the spent fuel storage racks to determine if the neutron-absorbing capacity of the material has degraded over time. This program ensures that a five (5) percent sub-criticality margin in the spent fuel pool is maintained during the period of extended operation by monitoring for loss of material, changes in dimension, and loss of neutron-absorption capacity of the Boral material. The existing coupon inspection frequency ensures at least one (1) coupon is examined during each 10 year period, beginning 10 years prior to the period of extended operation.

The Monitoring of Neutron-Absorbing Materials Other than Boraflex aging

management program will be enhanced to:

1. Maintain the coupon exposure such that it is bounding for the Boral material in all spent fuel racks prior to coupons being examined, by ensuring that the coupons have been surrounded with a greater number of freshly discharged fuel assemblies than that of any other cell

location.

This enhancement will be implemented prior to the period of extended operation.

RS-14-052 Enclosure B Page 4 of 22 As a result of changes to the Steam Generat ors aging management program identified in the response to B.2.1.10-1, LRA Appendix A, Section A.2.1.10, pages A-16 and A-17, Enhancements are revised as shown below. Revisions are indicated with bold italics for inserted text:

A.2.1.10 Steam Generators The Steam Generators aging management pr ogram is an existing preventive, mitigative, condition monitoring, and per formance monitoring program. The program establishes the operation, maintenance, testing, inspection, and repair requirements for the steam generators to ensure that plant technical specification surveillance requirements, ASME Code requirements, the Maintenance Rule performance criteria are met, thereby adequately managing the aging effects of steam generator tubes, plugs, and secondary side internal components. The aging effects include cracking, loss of material, reduction of heat transfer, and wall thinning. The program identifies and maintains the steam generator design and licensing bases and implements NEI 97-06, "Steam Generator Program Guidelines." NEI 97-06 establishes a framework for prevention, inspection, evaluation, repair and leakage monitoring measures.

Tube sleeve repair is currently not allowed by plant technical specifications for Byron and Braidwood Stations, Unit 1 and Unit 2 nor are there any sleeves currently installed. If BBS were to implement sleeving repair methods in the future, a Technical Specification change would be required and the sleeving would be incorporated into the Steam Generators aging management program. The Steam Generators aging management program will be enhanced to: 1. Validate that primary water stress corrosion cracking of the divider plate welds to the primary head and tubesheet cladding is not occurring. BBS commits to perform one (1) of the following three (3) resolution options for Units 1 and 2:

Option 1: Inspection Perform a one-time inspection, under the Steam Generators program, of each steam generator to assess the condition of the divider plate welds and the effectiveness of the Water Chemistry (A.2.1.2) program. For the Byron and Braidwood, Unit 1 steam generators which were replaced in 1998, the inspection will be performed between 2018 and the start of the period of extended operation to allow the steam generators to acquire at least twenty years of service. For the Byron and Braidwood, Unit 2 steam generators, which currently have at least twenty years of service, the inspection will be performed prior to entering the period of extended operation. The examination technique(s) will be capable of detecting primary water stress corrosion cracking (PWSCC) in the divider plate assemblies and associated

welds. or RS-14-052 Enclosure B Page 5 of 22 Option 2: Analysis

Perform an analytical evaluation of the steam generator divider plate welds in order to establish a technical basis which concludes that the steam generator reactor coolant pressure boundary is adequately maintained with the presence of steam generator divider plate weld cracking. The analytical evaluation will be submitted to the NRC for review and approval two (2) years prior to entering the associated period of extended operation.

or Option 3: Industry/NRC Studies If results of industry and NRC studies and operating experience document that potential failure of the steam generator reactor coolant pressure boundary due to PWSCC of the steam generator divider plate welds is not a credible concern, this commitment will be revised to reflect that conclusion. 2. Validate that primary water stress corrosion cracking of the tube-to-tubesheet welds is not occurring on BBS Unit 1. BBS commit to perform one (1) of the following three (3) resolution options for Unit 1:

Option 1: Inspection Perform a one-time inspection, under the Steam Generators program, of a representative number of tube-to-tubesheet welds in each steam generator to determine if PWSCC cracking is present. Since the Byron and Braidwood, Unit 1 steam generators were replaced in 1998, the inspection will be performed between 2018 and the start of the period of extended operation to allow the steam generators to acquire at least twenty years of service. The examination technique(s) will be capable of detecting primary water stress corrosion cracking (PWSCC) in the tube-to-tubesheet welds. If cracking is identified, the condition will be resolved through repair or engineering evaluation to justify continued service, as appropriate, and a periodic monitoring program will be established to perform routine tube-to-tubesheet weld inspections for the remaining life of the steam generators.

or Option 2: Analysis - Susceptibility Perform an analytical evaluation of the steam generator tube-to-tubesheet welds to determine that the welds are not susceptible to primary water stress corrosion cracking. The evaluation for determining that the tube-to-tubesheet welds are not susceptible to primary water stress corrosion cracking will be submitted to the NRC for review and approval two (2) years prior to entering the associated period of extended operation.

or RS-14-052 Enclosure B Page 6 of 22 Option 3: Analysis - Pressure Boundary Perform an analytical evaluation of the steam generator tube-to-tubesheet welds redefining the reactor coolant pressure boundary of the tubes, where the steam generator tube-to-tubesheet welds are not required to perform a reactor coolant pressure boundary function. The redefinition of the reactor coolant pressure boundary will be submitted to the NRC for review and approval two (2) years prior to entering the associated period of extended operation.

These enhancements will be implemented prior to entering the period of extended

operation.

RS-14-052 Enclosure B Page 7 of 22 As a result of the responses to RAI B.2.1.31-1, RAI B.2.1.31-2 and RAI B.2.1.31-3 provided in Enclosure A of this letter, LRA Section A.2.1.31, page A-34, is revised as shown below.

Additions are indicated with bolded italics; deletions are shown with strikethroughs. A.2.1.31 ASME Section XI, Subsection IWF The ASME Section XI, Subsection IWF aging management program is an existing program that consists of periodic visual examinations of component supports, evaluation, and corrective actions. The scope of the program includes ASME Class 1, 2, 3, and MC piping and component supports and high-strength structural bolting. The supports are examined for signs of degradation such as loss of material, loss of mechanical function, and loss of pre-load. The program is implemented through corporate and station procedures, which provide inspection and acceptance criteria consistent with the requirements of the ASME Code,Section XI, Subsection IWF as approved in 10 CFR 50.55a. This program is in accordance with ASME Section XI, Subsection IWF, 2001 Edition through the 2003 Addenda. The monitoring methods are effective in detecting the applicable aging effects and the frequency of monitoring is adequate to prevent significant degradation. The ASME Section XI, Subsection IWF aging management program will be enhanced to: 1. Add the MC supports for the transfer tube in the refueling cavity in the Containment Structure and refueling canal in the Fuel Handling Building to the scope of the program.

2. Revise implementing documents to P provide guidance for proper specification of bolting material, storage, lubricant s and sealants, and installation torque or tension to prevent or mitigate degradation and failure of structural bolting. Bolting material with actual measured yield strength of 150 ksi or greater shall not be used in plant changes without engineering approval, due to consideration of stress corrosion cracking vulnerability. Storage requirements for high strength bolts shall include the recommendations of the Research Council for Structural Connections, "Specification for Structural Joints Using ASTM A325 or A490 Bolts", Section 2. Lubricants that contain molybdenum disulfide (MoS 2) shall not be applied to high strength structural bolts within the scope of license renewal. 3. Provide procedural guidance, regarding the selection of supports to be inspected on subsequent inspections, when a support is repaired in accordance with the corrective action program. The enhanced guidance will ensure that the supports inspected on subsequent inspections are representative of the general population.

4. Perform one-time volumetric examinations on a sample of ASTM A490 bolts, greater than one-inch nominal diameter for the detection of stress corrosion cracking prior to the period of extended operation.

RS-14-052 Enclosure B Page 8 of 22 Volumetric examinations will be performed in accordance with the requirements of ASME Code Section XI, Appendix VIII, Supplement 8.

The sample will consist of bounding and representative A490 bolt sizes, joint configurations, and environmental exposure conditions. The sample will consist of 20% of the ASTM A490 bolts greater than one-inch nominal diameter or a maximum of 25 ASTM A490 bolts total for both Byron and Braidwood stations.

The selection of the samples will consider susceptibility to stress corrosion cracking (e.g., actual measured yield strength) and ALARA principles. Any adverse results of the volumetric examinations will be entered into the corrective action program and will be evaluated by engineering to determine if additional actions are warranted such as expansion of sample size, scope, and frequency of any additional supplemental visual or volumetric

examinations, as well as any code requirements specified by ASME

Section XI, Subsection IWF.

5. Revise implementing documents to perform periodic visual examinations to detect a corrosive environment that supports SCC potential for all (100%) of high strength bolting greater than one-inch nominal diameter prior to the period of extended operation, and then each inspection interval of 10 years thereafter. The periodic visual

examinations will include criteria to identify if the bolting has been

exposed to moisture or other contaminants by evidence of moisture, residue, foreign substance, or corrosion. Adverse conditions identified

during the examinations will be evaluated by engineering to determine if the bolt has been exposed to a corrosive environment with the potential to cause SCC. The bolts determined to have been exposed to corrosive

environment with the potential to cause SCC will be included in a

sample population for each specific bolt material where SCC is a concern. A sample size equal to 20 percent (rounded up to the nearest whole number) of the bolts in the sample population, with a maximum sample size of 25 bolts will be subject to supplemental volumetric examination to determine if SCC is present. The selection of the

samples will consider susceptibility to stress corrosion cracking (e.g., actual measured yield strength) and ALARA principles. Volumetric examinations will be performed in accordance with the requirements of ASME Code Section XI, Appendix VIII, Supplement 8. The results of the

volumetric examinations will be evaluated by engineering to determine if additional actions are warranted such as expansion of sample size, scope, and frequency of any additional supplemental visual or

volumetric examinations, as well as any code requirements specified by ASME Section XI, Subsection IWF. These enhancements will be implemented prior to the period of extended operation.

RS-14-052 Enclosure B Page 9 of 22 As a result of changes to the Steam Generat ors aging management program identified in the response to B.2.1.10-1, LRA Appendix B, Section B.2.1.10, pages B-76 and B-77, are revised as shown below. Revisions are indicated with bold italics for inserted text:

Enhancements Prior to the period of extended operation, the following enhancements will be implemented in the following program elements: 1. Validate that primary water stress corrosion cracking of the divider plate welds to the primary head and tubesheet cladding is not occurring. BBS commits to perform one (1) of the following three (3) resolution options for Units 1 and 2:

Option 1: Inspection Perform a one-time inspection, under the Steam Generators (B.2.1.10) program, of each steam generator to assess the condition of the divider plate welds and the effectiveness of the Water Chemistry (B.2.1.2) program. For the Byron and Braidwood, Unit 1 steam generators which were replaced in 1998, the inspection will be performed between 2018 and the start of the period of extended operation to allow the steam generators to acquire at least twenty years of service. For the Byron and Braidwood, Unit 2 steam generators, which currently have at least twenty years of service, the inspection will be performed prior to entering the period of extended operation. The examination technique(s) will be capable of detecting primary water stress corrosion cracking (PWSCC) in the divider plate assemblies and associated

welds. or Option 2: Analysis Perform an analytical evaluation of the steam generator divider plate welds in order to establish a technical basis which concludes that the steam generator reactor coolant pressure boundary is adequately maintained with the presence of steam generator divider plate weld cracking. The analytical evaluation will be submitted to the NRC for review and approval two (2) years prior to entering the associated period of extended operation.

or Option 3: Industry/NRC Studies If results of industry and NRC studies and operating experience document that potential failure of the steam generator reactor coolant pressure boundary due to PWSCC of the steam generator divider plate welds is not a credible concern, this commitment will be revised to reflect that conclusion. Program Element Affected: Parameters Monitored/Inspected (Element 3)

RS-14-052 Enclosure B Page 10 of 22

2. Validate that primary water stress corrosion cracking of the tube-to-tubesheet welds is not occurring on BBS Unit 1. BBS commit to perform one (1) of the following three (3) resolution options for Unit 1:

Option 1: Inspection Perform a one-time inspection, under the Steam Generator (B.2.1.10) program, of a representative number of tube-to-tubesheet welds in each steam generator to determine if PWSCC cracking is present. Since the Byron and Braidwood, Unit 1 steam generators were replaced in 1998, the inspection will be performed between 2018 and the start of the period of extended operation to allow the steam generators to acquire at least twenty years of service. The examination technique(s) will be capable of detecting primary water stress corrosion cracking (PWSCC) in the tube-to-tubesheet welds. If cracking is identified, the condition will be resolved through repair or engineering evaluation to justify continued service, as appropriate, and a periodic monitoring program will be established to perform routine tube-to-tubesheet weld inspections for the remaining life of the steam generators.

or Option 2: Analysis - Susceptibility Perform an analytical evaluation of the steam generator tube-to-tubesheet welds to determine that the welds are not susceptible to primary water stress corrosion cracking. The evaluation for determining that the tube-to-tubesheet welds are not susceptible to primary water stress corrosion cracking will be submitted to the NRC for review and approval two (2) years prior to entering the associated period of extended operation.

or Option 3: Analysis - Pressure Boundary Perform an analytical evaluation of the steam generator tube-to-tubesheet welds redefining the reactor coolant pressure boundary of the tubes, where the steam generator tube-to-tubesheet welds are not required to perform a reactor coolant pressure boundary function. The redefinition of the reactor coolant pressure boundary will be submitted to the NRC for review and approval two (2) years prior to entering the associated period of extended operation. Program Element Affected: Parameters Monitored/Inspected (Element 3) A license amendment (Adams Accession Number: ML12262A360), approved by the NRC for BBS Unit 2, redefined the pressure boundary in which the tube-to-tubesheet weld is no longer included; therefore a plant specific program to verify the effectiveness of the Water Chemistry (B.2.1.2) program is not required.

RS-14-052 Enclosure B Page 11 of 22 As a result of the response to RAI B.2.1.27-1 provided in Enclosure A of this letter, LRA Appendix B, Sections B.1.5, page B-10, and B.2.1.27, page B-169, are revised as shown below.

Additions are indicated with bolded italics; deletions are shown with strikethroughs. B.1.5 NUREG-1801 Chapter XI Aging Management Programs The following NUREG-1801 Chapter XI AMPs are described in Section B.2 of this appendix as indicated. Programs are identified as either existing or new to Byron and Braidwood. All programs are or will be consistent with programs discussed in NUREG-1801.

27. Monitoring of Neutron-Absorbing Materials Other than Boraflex (Section B.2.1.27)

[Existing - Requires Enhancement

]

B.2.1.27 Monitoring of Neutron-Absorbing Materials Other than Boraflex Enhancements None. Prior to the period of extended operation, the following enhancement will be implemented in the following program elements:

1. Maintain the coupon exposure such that it is bounding for the Boral material in all spent fuel racks prior to coupons being examined, by ensuring that the coupons have been surrounded with a greater number

of freshly discharged fuel assemblies than that of any other cell location. Program Element Affected: Monitoring and Trending (Element

5)

RS-14-052 Enclosure B Page 12 of 22 As a result of the responses to RAI B.2.1.31-1, B.2.1.31-2 and B.2.1.31-3 provided in Enclosure A of this letter, LRA Section B.2.1.31, pages B-204 through B-209, are revised as shown below. Additions are indicated with bolded italics; deletions are shown with strikethroughs. B.2.1.31 ASME Section XI, Subsection IWF Program Description The ASME Section XI, Subsection IWF aging management program is an existing condition monitoring program that consists of periodic visual examination of ASME Section XI Class 1, 2, 3, and MC piping and component support members for signs of degradation such as loss of material, loss of mechanical function, and loss of pre-load in the following environments: air-indoor uncontrolled, air-outdoor, air with borated water leakage, and treated borated water. Bolting for component supports is also included with these component supports and inspected for loss of material and for loss of preload by inspecting for missing, detached, or loosened bolts and nuts in the following environments: air indoor, air outdoor and treated water. The program utilizes procedures that are consistent with industry guidance to ensure proper specification of bolting material, lubricant, and installation torque to prevent or minimize loss of bolting preload or other loss of structural integrity. Indications of degradation are entered in the corrective action program for evaluation or correction to ensure the intended function of the component support is maintained. The current ASME Section XI, Subsection IWF program is implemented through corporate and station procedures, which provide inspection and acceptance criteria, and complies with ASME, Boiler and Pressure Vessel Code,Section XI, Subsection IWF 2001 Edition through the 2003 Addenda as approved in 10 CFR 50.55(a). In accordance with 10 CFR 50.55a(g)(4)(ii), the ISI program is updated each successive 120-month inspection interval to comply with the requirements of the latest edition of the ASME Code specified twelve months before the start of the inspection interval.

The monitoring methods are effective in detecting the applicable aging effects and the frequency of monitoring is adequate to prevent significant degradation. The ASME Section XI, Subsection IWF aging management program utilizes examinations that detect degradation before loss of intended function. Preventive measures associated with structural bolts are addressed in implementing procedures. The program will be enhanced, as noted below to provide reasonable assurance that the ASME Section XI, Subsection IWF program aging effects will be adequately managed during the period of extended operation. NUREG-1801 Consistency The ASME Section XI, Subsection IWF aging management program will be consistent with the ten elements of aging management program XI.M1, "ASME Section XI, Subsection IWF," specified in NUREG-1801 with the following exceptions:

RS-14-052 Enclosure B Page 13 of 22 Exceptions to NUREG-1801

1. NUREG-1801 requires, as a preventive measure that can reduce the potential for SSC or IGSCC, using bolting material for high strength structural applications that have an actual measured yield strength limited to less than 1,034 megapascals (MPa) (150 kilo-pounds per square inch) (NUREG-1339).

Site documentation indicates high strength bolts, consisting of ASME SA 540, which exceed this limit, and ASTM A490 materials, which may exceed this limit, were used as part of the original design. The site

documentation indicates that the originally installed five-inch diameter ASME SA 540 reactor coolant pump hold-down bolts at both Byron and Braidwood and the 1-1/2" diameter ASME SA 540 pressurizer hold-down bolts at only Braidwood have actual measured yield strength that is greater

than 150 ksi. In addition, and the originally installed five-inch diameter ASME SA 540 reactor coolant pump hold-down bolts at both Byron and Braidwood have actual measured tensile strength that is greater than 170 ksi. ASTM A490 bolts were used at connections between structural steel members of the steam generator, reactor coolant pump, and pressurizer

supports.

Program Element Affected: Preventive Measures (Element 2)

Justification for Exception NUREG-1801 provides guidance to use bolting material, for high strength structural applications, that has an actual measured yield strength limited to less than 150 ksi as delineated in NUREG-1339 and Reg Guide 1.65 Revision 1. SA 540, Class 1, Grade B24 and SA 540, Class 2, Grade B23 materials are described in these documents as high-strength, low alloy materials, which when tempered to a maximum tensile strength of less than 170 ksi, are relatively immune to stress corrosion cracking. The originally installed reactor coolant pump hold-down bolts, at both Byron and Braidwood, and the pressurizer hold-down bolts, at only Braidwood, material and quality control requirements were in accordance with the requirements of the 1974 edition of Subsection NF of the ASME Boiler and Pressure Vessel Code,Section III, with the Summer 1975 Addenda. The five-inch diameter reactor coolant pump hold-down bolts at both Byron and Braidwood were fabricated from SA 540, Class 1, Grade B24 low alloy steel with a minimum yield strength of 150 ksi and a minimum tensile strength of 165 ksi. The 1-1/2" diameter pressurizer hold-down bolts at Braidwood were fabricated from SA 540, Class 2, Grade B23 low alloy steel with a minimum yield strength of 140 ksi and a minimum tensile strength of 155 ksi. The installed bolts were consistent with the existing Code design guidance when installed and are relatively immune to stress corrosion cracking. Other preventive measures listed in NUREG-1801 program XI.S3, "ASME Section XI, Subsection IWF" that can reduce the potential for cracking are met by the ASME Section XI, Subsection IWF program. These include: a) Metal-plated stud bolting is not used, which could cause degradation due to corrosion or hydrogen embrittlement.

RS-14-052 Enclosure B Page 14 of 22 b) An approved stable lubricant was applied to the bolts. The lubricant used during original installation does not contain molybdenum disulfide. Procedures within the scope of the ASME Section XI, Subsection IWF aging management program will be enhanced to include the recommendations of the GALL Report AMP SI.X3 Preventive Actions" for ASTM A490 bolts. The enhancements follow the preventive actions

for storage, lubricants, and stress corrosion cracking potential discussed in Section 2 of Research Council for Structural Connections (RCSC) publication "Specification for Structural Joints Using ASTM A325 or A490 Bolts" The BBS hold down bolt design configuration at the reactor coolant pump and pressurizer supports prevents SSC from occurring at the portion of t he bolt below the bolt head where the bolt is in tension, since this portion of the bolt is not exposed to an environment that would initiate SCC. Therefore, volumetric examinations are not required to detect SCC in these hold down bolts. The hold down bolts for the reactor coolant pumps and pressurizer firmly connect the components to the component supports. Below the bolt head, the bolting materials and holes are not exposed to borated water leakage. The bolts were not installed in oversized holes with no initial bolt tension such as would be found at a sliding connection. The bolt heads bear tightly on the support surface, in standard holes, and were tightened to prevent sliding between the adjacent surfaces. The original installation torque used when installing the reactor coolant pump hold down bolts was designed to result in about 56% of the minimum tensile strength of the bolt material. The original installation of torque used when installing the pressurizer hold down bolts was designed to result in about 27% of the minimum tensile strength. This prevents borated water from seeping beneath the bolt head, which prevents the potential initiation of corrosion under the bolt head. This prevents the initiation of SCC beneath the bolt head since a borated water leakage environment will not exist below the bolt head. The top of the bolt head is exposed to an air with borated water leakage environment and potential losses of material due to corrosion would be readily identified during examinations that are currently performed as part of the ASME Section XI, Subsection IWF program.

Regarding ASTM A490 bolting material, Operating Experience cited in NUREG-1801 stated "SCC has occurred in high strength bolts used for nuclear steam supply system component supports (EPRI NP-5769)."

The OE cited in NUREG-1801 refers to NP-5769 (issued in 1988) and SCC was found only in certain specific materials. While EPRI NP-5769 Volume 1, Table 11-1 does list A490 bolts for ductile failures and failure due improper torque, no SCC failures were noted for A490 bolt materials. One failure of a special 4140 material with 200 ksi minimum yield strength due to SCC was noted and associated with a high preload and borated water environment. This last example describes where the A490 specification was used for heat treatment requirements but this was not an A490 bolt material. This information was reviewed under comment # 906 during the development of NUREG-1950, Disposition of RS-14-052 Enclosure B Page 15 of 22 Public Comments and Technical Bases for Changes in the License Renewal Guidance Documents NUREG-1801 and NUREG-1800. As a result, it was concluded that ASTM A490 bolting is not prone to SCC. Since the actual measured yield strength of some installed bolts may be greater than 150 ksi, the aging management review identified the bolt material as "High Strength Low Alloy Steel Bolting with Yield Strength of 150 ksi or Greater" and identified loss of material and potential cracking as an aging effect requiring management. There have been no recordable indications of degradation identified by ASME Section XI, Subsection IWF

program examination of reactor coolant pump and pressurizer support bolting components. The steam generator, reactor coolant pump and pressurizer supports, and the equipment hold down bolt heads, are examined per ASME Code,Section XI, Table IWF-2500-1. The current examination parameters include indications of corrosion and a loss of material at the bolt head which would indicate a potential for SSC to occur at the top of the bolt head due the presence of an air with borated water leakage environment. As a result, the current enhanced ASME Section XI, Subsection IWF program examination techniques, which include performing VT

-3 visual examinations, are appropriate for identifying degradation of these bolts without replacing the originally installed bolts.

given t The bolts were designed in accordance with the original design Code, the preventative measures described above were used during original design, fabrication, and installation thereby reducing the potential for SCC. , for the specific bolting materials used, and the support configuration prevents water from seeping beneath bolt head.

Therefore, the enhanced ASME Section XI, Subsection IWF program will provide reasonable assurance that the high strength bolts will perform their intended functions and will be effective in managing the degradation and subsequent potential cracking aging effect during the period of extended operation.

2. NUREG-1801 recommends, as a method of detecting aging effects, volumetric examination of high strength bolting material, with a diameter of greater than 1" and used in structural applications, which have actual measured yield strength greater than or equal to 150 ksi. Site documentation indicates high strength bolts, consisting of ASME SA 540, which exceed this limit, and ASTM A490 materials, which may exceed this limit, were used as part of the original design. The site documentation indicates that the originally installed five-inch diameter ASME SA 540 reactor coolant pump hold-down bolts at both Byron and Braidwood and the 1-1/2" diameter ASME SA 540 pressurizer hold-down bolts at only Braidwood have actual measured yield strength that is greater than 150 ksi.

Currently T t here are no qualified standards to perform volumetric examinations on these high strength bolts at BBS. The five-inch diameter ASME SA 540 hold down bolts for the reactor coolant pumps at BBS consist of cap screws where the bolt head is machined to allow for the insertion of a socket to tighten the bolt. This bolt head configuration currently does not allow for a recognized volumetric examination of the bolt.

The ASTM A490 bolts were used at connections between structural steel members of the steam generator, RS-14-052 Enclosure B Page 16 of 22 reactor coolant pump, and pressurizer supports.

Periodic volumetric examinations of high strength bolting material, with a diameter of greater than one-inch nominal and used in structural applications, which have actual measured yield strength greater than or

equal to 150 ksi will not be performed.

The following elements provide the bases to justify taking an exception to the GALL report recommendation that periodic volumetric examinations be performed. An extensive plant-specific history of volumetric examinations exists on ASME SA 540 bolting material with no evidence of SCC being identified.

Volumetric examinations will be performed on a sample of ASTM A490 bolts, greater than one-inch nominal diameter for the detection of stress corrosion cracking to establish plant-specific history for the ASTM A490 bolting materials.

Periodic visual examinations will be performed to detect a corrosive environment that supports SCC potential for high strength bolting greater than one-inch nominal diameter. Supplemental volumetric examinations will be performed on a sample of bolts determined to have been exposed to corrosive

environment with the potential to cause SCC.

Program Elements Affected: Parameters Monitored/Inspected (Element 3), Detection of Aging Effects (Element 4), Monitoring and trending (Element 5), Acceptance Criteria (Element 6), Corrective Actions (Element 7)

Justification for Exception NUREG-1801 provides guidance to use bolting material, for high strength structural applications, that has an actual measured yield strength limited to less than 150 ksi as delineated in NUREG-1339 and Reg Guide 1.65 Revision 1. The originally installed reactor coolant pump hold-down bolts at both Byron and Braidwood and the pressurizer hold-down bolts at only Braidwood material and quality control requirements were in accordance with the requirements of the 1974 edition of Subsection NF of the ASME Boiler and Pressure Vessel Code,Section III, with the Summer 1975 Addenda. The five-inch diameter reactor coolant pump hold-down bolts at both Byron and Braidwood were fabricated from SA 540, Class 1, Grade B24 low alloy steel with a minimum yield strength of 150 ksi and a minimum tensile strength of 165 ksi. The 1-1/2" diameter pressurizer hold-down bolts at Braidwood were fabricated from SA 540, Class 2, Grade B23 low alloy steel with a minimum yield strength of 140 ksi and a minimum tensile strength of 155 ksi.

The ASTM A490 bolts were used at connections between structural steel members of the steam generator, reactor coolant pump, and pressurizer supports. Therefore, the installed bolts were consistent with the existing Code design when installed.

RS-14-052 Enclosure B Page 17 of 22 Other preventive measures listed in NUREG-1801 program XI.S3, "ASME Section XI, Subsection IWF" that can reduce the potential for cracking are met by the ASME Section XI, Subsection IWF program. These include: a) Metal-plated stud bolting is not used, which could cause degradation due to corrosion or hydrogen embrittlement. b) An approved stable lubricant was applied to the bolts. The lubricant used does not contain molybdenum disulfide. The BBS hold down bolt design configuration at the reactor coolant pump and pressurizer supports prevents SSC from occurring at the portion of the bolt below the bolt head where the bolt is in tension, since this portion of the bolt is not exposed to an environment that would initiate SCC. Therefore, volumetric examinations are not required to detect SCC in these hold down bolts. The hold down bolts for the reactor coolant pumps and pressurizer firmly connect the components to the component supports. Below the bolt head, the bolting materials and holes are not exposed to borate d water leakage. The bolts were not installed in oversized holes with no initial bolt tension such as would be found at a sliding connection. The bolt heads bear tightly on the support surface, in standard holes, and were tightened to prevent sliding between the adjacent surfaces. The original installation torque used when installing the reactor coolant pump hold down bolts was designed to result in about 56% of the minimum tensile strength of the bolt material. The original installation of torque used when installing the pressurizer hold down bolts was designed to result in about 27% of the minimum tensile strength. This prevents borated water from seeping beneath the bolt head, which prevents the potential initiation of corrosion under the bolt head.

This prevents the initiation of SCC beneath the bolt head since a borated water leakage environment will not exist below the bolt head. The top of the bolt head is exposed to an air with borated water leakage environment and potential losses of material due to corrosion would be readily identified during examinations that are currently performed as part of the ASME Section XI, Subsection IWF program.

Since the actual measured yield strength of some installed bolts may be greater than 150 ksi, the aging management review identified the bolt material as "High Strength Low Alloy Steel Bolting with Yield Strength of 150 ksi or Greater" and identified loss of material and potential cracking as an aging effect requiring management. There have been no recordable indications of degradation identified by ASME Section XI, Subsection IWF

program examination of reactor coolant pump and pressurizer support bolting components. The reactor coolant pump and pressurizer supports, and the equipment hold down bolt heads, are examined per ASME Code,Section XI, Table IWF-2500-1. The current examination parameters include indications of corrosion and a loss of material at the bolt head which would indicate a potential for SSC to occur at the top of the bolt head due the presence of an

air with borated water leakage environment.

RS-14-052 Enclosure B Page 18 of 22 Periodic volumetric examinations of high strength bolting material, with a diameter of greater than one-inch nominal and used in structural applications, which have actual measured yield strength greater than or

equal to 150 ksi will not be performed. Plant-specific volumetric examinations of ASME SA 540 and ASTM A490 high strength bolts together with periodic visual examinations to detect a corrosive environment with supplemental volumetric examinations if warranted provide the justification to take an exception to the periodic

volumetric examinations.

The following provides details on the elements that provide the bases to justify taking an exception to the GALL report recommendation that periodic volumetric examinations be performed.

Byron and Braidwood have an extensive history of volumetric examinations of the reactor head closure studs. The reactor head closure stud material at Byron and Braidwood is ASME SA

540 and the reactor head closure studs have been identified as high strength low allow bolting with measured yield strength of 150 ksi or greater. The material of the ASME Section XI, Subsection IWF high strength five-inch RCP hold-down bolts at Byron and Braidwood and the 1.5-inch pressurizer hold-down bolts at Braidwood is ASME SA 540. One hundred percent of the

tensioned reactor head closure stud population has been subject to volumetric examination each ten-year inservice inspection interval (more than 500 volumetric examinations total for both stations), with no evidence of SCC identified. Because of the

similar materials and environmental conditions, the numerous

reactor head closure stud volumetric examinations with no evidence of SCC identified can be used to support a plant-specific justification to waive performing periodic volumetric examinations of the ASME SA 540 high strength bolts as proposed in the GALL Report.

One-time volumetric examinations will be performed on a sample of ASTM A490 bolts, greater than one-inch nominal diameter for

the detection of stress corrosion cracking prior to the period of extended operation. These volumetric examinations together

with the extensive volumetric examinations that have been performed on the ASME SA 540 reactor head closure studs are used to justify taking an exception to the GALL report

recommendation that periodic volumetric examination be performed to manage SCC. The volumetric examinations will be performed in accordance with the requirements of ASME Code

Section XI, Appendix VIII, Supplement 8.

The sample will consist of bounding and representative A490 bolt sizes, joint

configurations, and environmental exposure conditions. The sample will consist of 20% of the ASTM A490 bolts greater than RS-14-052 Enclosure B Page 19 of 22 one-inch nominal diameter or a maximum of 25 ASTM A490 bolts total for both Byron and Braidwood stations.

The selection of the samples will consider susceptibility to stress corrosion

cracking (e.g., actual measured yield strength) and ALARA

principles. Any adverse results of the volumetric examinations will be entered into the corrective action program and will be evaluated by engineering to determine if additional actions are warranted such as expansion of sample size, scope, and

frequency of any additional supplemental visual or volumetric

examinations, as well as any code requirements specified by ASME Section XI, Subsection IWF. This activity will be an enhancement to the ASME Section XI, Subsection IWF program and be implemented prior to the period of extended operation.

Periodic visual examinations to detect a corrosive environment that supports SCC potential for all (100%) accessible high strength bolting greater than one-inch nominal diameter will be performed prior to the period of extended operation, and then each inspection interval of 10 years thereafter. The periodic visual examinations will include criteria to identify if the bolting

has been exposed to moisture or other contaminants by

evidence of moisture, residue, foreign substance, or corrosion.

Adverse conditions identified during the examinations will be

evaluated by engineering to determine if the bolt has been exposed to a corrosive environment with the potential to cause SCC. The bolts determined to have been exposed to corrosive

environment with the potential to cause SCC will be included in a

sample population for each specific bolt material where SCC is a concern. A sample size equal to 20 percent (rounded up to the nearest whole number) of the bolts in the sample population, with a maximum sample size of 25 bolts will be subject to

supplemental volumetric examination to determine if SCC is

present. The selection of the samples will consider

susceptibility to stress corrosion cracking (e.g., actual measured yield strength) and ALARA principles. Volumetric examinations will be performed in accordance with the requirements of ASME

Code Section XI, Appendix VIII, Supplement 8. The results of the

volumetric examinations will be evaluated by engineering to determine if additional actions are warranted such as expansion of sample size, scope, and frequency of any additional supplemental visual or volumetric examinations, as well as any code requirements specified by ASME Section XI, Subsection IWF. This activity will be an enhancement to the ASME Section

XI, Subsection IWF program and be implemented prior to the

period of extended operation.

In summary, the extensive plant-specific history on volumetric examination of ASME SA 540 reactor head studs, the plant-specific volumetric examinations that will be performed on a sample of ASTM A490 bolts, greater than one-inch nominal diameter, and the periodic RS-14-052 Enclosure B Page 20 of 22 visual examinations to detect a corrosive environment with supplemental volumetric examinations if warranted, are used to justify

taking an exception to the GALL report recommendation that periodic

volumetric examinations be performed.

Specifically, periodic volumetric examination in addition to VT-3 examinations for high strength structural bolting (actual measured yield strength greater than 150 Ksi) in sizes greater than one-inch nominal diameter will not be performed.

As a result, of the above, the plant-specific history, of volumetric examinations performed on high strength bolting together with the ongoing periodic visual examinations to detect a corrosive environment with supplemental volumetric examinations if warranted, and the current ASME Section XI, Subsection IWF program examination techniques, which include performing VT

-3 visual examinations, are appropriate for identifying degradation of these high strength bolts. In addition, given the bolts were designed in accordance with the original design Code, the preventative measures described above were used during original design, fabrication, and installation, thereby, reducing the potential for SCC. , the specific bolting materials used, and the support configuration prevents water from seeping beneath bolt head. Therefore, the enhanced ASME Section XI, Subsection IWF program will provide reasonable assurance that the high strength bolts will perform their intended functions and will be effective in managing the degradation and subsequent potential cracking aging effect during the period of extended operation.

Enhancements Prior to the period of extended operation, the following enhancements will be implemented in the following program elements: 1. Add the MC supports for the transfer tube in the refueling cavity in the Containment Structure and refueling canal in the Fuel Handling Building to the scope of the program. Program Elements Affected: Scope of Program (Element 1)

2. Revise implementing documents to P provide guidance for proper specification of bolting material, storage, lubricant s and sealants, and installation torque or tension to prevent or mitigate degradation and failure of structural bolting. Bolting material with actual measured yield strength of 150 ksi or greater shall not be used in plant changes without engineering approval, due to consideration of stress corrosion cracking vulnerability. Storage requirements for high strength bolts shall include the recommendations of the Research Council for Structural Connections, "Specification for Structural Joints Using ASTM A325 or A490 Bolts", Section 2. Lubricants that contain molybdenum disulfide (MoS 2) shall not be applied to high strength structural bolts within the scope of license renewal. Program Elements Affected: Preventive Actions (Element 2)

RS-14-052 Enclosure B Page 21 of 22

3. Provide procedural guidance, regarding the selection of supports to be inspected on subsequent inspections, when a support is repaired in accordance with the corrective action program. The enhanced guidance will ensure that the supports inspected on subsequent inspections are representative of the general population.

Program Elements Affected: Monitoring and Trending (Element 5)

4. Perform one-time volumetric examinations on a sample of ASTM A490 bolts, greater than one-inch nominal diameter for the detection of stress

corrosion cracking prior to the period of extended operation.

Volumetric examinations will be performed in accordance with the requirements of ASME Code Section XI, Appendix VIII, Supplement 8.

The sample will consist of bounding and representative A490 bolt sizes, joint configurations, and environmental exposure conditions. The sample will consist of 20% of the ASTM A490 bolts greater than one-inch nominal diameter or a maximum of 25 ASTM A490 bolts total for both Byron and Braidwood stations.

The selection of the samples will consider susceptibility to stress corrosion cracking (e.g., actual measured yield strength) and ALARA principles. Any adverse results of the volumetric examinations will be entered into the corrective action program and will be evaluated by engineering to determine if additional actions are warranted such as expansion of sample size, scope, and

frequency of any additional supplemental visual or volumetric examinations, as well as any code requirements specified by ASME Section XI, Subsection IWF. Program Elements Affected: Detection of

Aging Effects (Element 4), Monitoring and trending (Element 5), Acceptance Criteria (Element 6), Corrective Actions (Element 7)

5. Revise implementing documents to perform periodic visual examinations to detect a corrosive environment that supports SCC potential for all (100%) high strength bolting greater than one-inch nominal diameter prior to the period of extended operation, and then each inspection interval of 10 years thereafter. The periodic visual

examinations will include criteria to identify if the bolting has been

exposed to moisture or other contaminants by evidence of moisture, residue, foreign substance, or corrosion. Adverse conditions identified during the examinations will be evaluated by engineering to determine if the bolt has been exposed to a corrosive environment with the potential

to cause SCC. The bolts determined to have been exposed to corrosive

environment with the potential to cause SCC will be included in a

sample population for each specific bolt material where SCC is a concern. A sample size equal to 20 percent (rounded up to the nearest whole number) of the bolts in the sample population, with a maximum sample size of 25 bolts will be subject to supplemental volumetric examination to determine if SCC is present. The selection of the

samples will consider susceptibility to stress corrosion cracking (e.g., actual measured yield strength) and ALARA principles. Volumetric examinations will be performed in accordance with the requirements of ASME Code Section XI, Appendix VIII, Supplement 8. The results of the

volumetric examinations will be evaluated by engineering to determine RS-14-052 Enclosure B Page 22 of 22 if additional actions are warranted such as expansion of sample size, scope, and frequency of any additional supplemental visual or

volumetric examinations, as well as any code requirements specified by ASME Section XI, Subsection IWF. Program Elements Affected:

Parameters Monitored/Inspected (Element 3), Detection of Aging Effects (Element 4), Monitoring and trending (Element 5), Acceptance Criteria (Element 6), Corrective Actions (Element 7)

RS-14-052 Enclosure C Page 1 of 8 Enclosure C Byron and Braidwood Stations (BBS) Units 1 and 2 License Renewal Commitment List Changes This Enclosure identifies commitments made in this document and is an update to the Byron and Braidwood Station (BBS) LRA Appendix A, Table A.5, License Renewal Commitment List.

Any other actions discussed in the submittal represent intended or planned actions and are described to the NRC for the NRC's information and are not regulatory commitments. Changes to the BBS LRA Appendix A, Table A.5 License Renewal Commitment List are as a result of the Exelon response to the following RAIs:

RAI B.2.1.10-1

RAI B.2.1.27-1

RAI B.2.1.31-1 RAI B.2.1.31-2 RAI B.2.1.31-3

Notes: To facilitate understanding, portions of the original License Renewal Commitment List have been repeated in this Enclosure, with revisions indicated. Existing LRA text is shown in normal font. Changes are highlighted with bold italics for inserted text and strikethroughs for deleted text.

RS-14-052 Enclosure C Page 2 of 8 As a result of the response to RAI B.2.1.10-1 provided in Enclosure A of this letter, LRA Appendix A, Table A.5 License Renewal Commitment List, line item 10 on pages A-72 and A-74, is revised as shown below. The RAI that led to this commitment modification is listed in the "SOURCE" column. Any other actions described in this submittal represent intended or planned actions. They a re described for the NRC's information and are not regulatory commitments.

A.5 License Renewal Commitment List NO. PROGRAM OR TOPIC COMMITMENT IMPLEMENTATION SCHEDULESOURCE 10 Steam Generators Steam Generators is an existing program that will be enhanced to:

1. Validate that primary water stress corrosion cracking of the divider plate welds to the primary head and tubesheet cladding is not occurring. BBS commits to perform one (1) of the following three (3) resolution options for Units 1 and 2:

Option 1: Inspection Perform a one-time inspection, under the Steam Generators program, of each steam generator to assess the condition of the divider plate welds and the effectiveness of the Water Chemistry (A.2.1.2) program. For the Byron and Braidwood, Unit 1 steam generators which were replaced in 1998, the inspection will be performed between 2018 and the start of the period of extended operation to allow the steam generators to acquire at least twenty years of service. For the Byron and Braidwood, Unit 2 steam generators which currently have at least twenty years of service, the inspection will be performed prior to entering the period of extended operation. The examination technique(s) will be capable of detecting primary water stress corrosion cracking (PWSCC) in the divider plate assemblies and associated welds.

or Option 2: Analysis Perform an analytical evaluation of the steam generator Program to be enhanced prior to the period of extended operation.

Schedule for submittal of analysis, if applicable, identified in Commitment.

Section A.2.1.10 Exelon Letter RS-14-052 03/04/2014 RAI B.2.1.10-1

RS-14-052 Enclosure C Page 3 of 8 NO. PROGRAM OR TOPIC COMMITMENT IMPLEMENTATION SCHEDULESOURCE divider plate welds in order to establish a technical basis which concludes that the steam generator reactor coolant pressure boundary is adequately maintained with the presence of steam generator divider plate weld cracking. The analytical evaluation will be submitted to the NRC for review and approval two (2) years prior to entering the associated period of extended operation.

or Option 3: Industry/NRC Studies If results of industry and NRC studies and operating experience document that potential failure of the steam generator reactor coolant pressure boundary due to PWSCC of the steam generator divider plate welds is not a credible concern, this commitment will be revised to reflect that conclusion.

2. Validate that primary water stress corrosion cracking of the tube-to-tubesheet welds is not occurring on BBS Unit 1. BBS commit to perform one (1) of the following three (3) resolution options for Unit 1:

Option 1: Inspection Perform a one-time inspection, under the Steam Generators (A.2.1.10) program, of a representative number of tube-to-tubesheet welds in each steam generator to determine if PWSCC cracking is present. Since the Byron and Braidwood Unit 1 steam generators were replaced in 1998, the inspection will be performed between 2018 and the start of the period of extended operation to allow the steam generators to acquire at least twenty years of service. The examination technique(s) will be capable of detecting primary water stress corrosion cracking (PWSCC) in the tube-to-tubesheet welds. If cracking is identified, the condition will be resolved through repair or engineering evaluation to justify continued service, as appropriate, and RS-14-052 Enclosure C Page 4 of 8 NO. PROGRAM OR TOPIC COMMITMENT IMPLEMENTATION SCHEDULESOURCE a periodic monitoring program will be established to perform routine tube-to-tubesheet weld inspections for the remaining life of the steam generators.

or Option 2: Analysis - Susceptibility Perform an analytical evaluation of the steam generator tube-to-tubesheet welds to determine that the welds are not susceptible to primary water stress corrosion cracking. The evaluation for determining that the tube-to-tubesheet welds are not susceptible to primary water stress corrosion cracking will be submitted to the NRC for review and approval two (2) years prior to entering the associated period of extended operation.

or Option 3: Analysis - Pressure Boundary Perform an analytical evaluation of the steam generator tube-to-tubesheet welds redefining the reactor coolant pressure boundary of the tubes, where the steam generator tube-to-tubesheet welds are not required to perform a reactor coolant pressure boundary function. The redefinition of the reactor coolant pressure boundary will be submitted to the NRC for review and approval two (2) years prior to entering the associated period of extended operation

RS-14-052 Enclosure C Page 5 of 8 As a result of the response to RAI B.2.1.27-1 provided in Enclosure A of this letter, LRA Appendix A, Table A.5 License Renewal Commitment List, Item 27 on page A-80, is revised as shown below. The RAI that led to this commitment modification is listed in the "SOURCE" column. Any other actions described in this submittal represent intended or planned actions. They are described for the NRC's information and are not regulatory commitments. Additions are indicated with bolded italics; deletions are shown with strikethroughs. NO. PROGRAM OR TOPIC COMMITMENT IMPLEMENTATION SCHEDULE SOURCE 27 Monitoring of Neutron-Absorbing Materials Other than Boraflex Existing program is credited.

Monitoring of Neutron-Absorbing Materials Other than Boraflex is an existing program that will be enhanced to:

1. Maintain the coupon exposure such that it is bounding for the Boral material in all spent fuel racks prior to coupons

being examined, by ensuring that the coupons have been surrounded with a greater number of freshly discharged fuel assemblies than that of any other cell location.

Ongoing Program to be enhanced prior to the period of extended operation. Section A.2.1.27 Exelon letter RS-14-052 03/04/2014 RAI B.2.1.27-1

RS-14-052 Enclosure C Page 6 of 8 As a result of the response to RAI B.2.1.31-1, B.2.1.31-2 and B.2.1.31-3 provided in Enclosure A of this letter, LRA Table A.5, Item 31, page A-84, is revised as shown below. Additions are indicated with bolded italics and strikethroughs for deleted text.

A.5 License Renewal Commitment List NO. PROGRAM OR TOPIC COMMITMENT IMPLEMENTATION SCHEDULE SOURCE 31 ASME Section XI, Subsection IWF ASME Section XI, Subsection IWF is an existing program that will be enhanced to:

1. Add the MC supports for the transfer tube in the refueling cavity in the Containment Structure and refueling canal in the Fuel Handling Building to the scope of the program.

2. Revise implementing documents to Pprovide guidance for proper specification of bolting material, storage, lubricant s and sealants, and installation torque or tension to prevent or mitigate degradation and failure of structural bolting. Bolting material with actual measured yield strength of 150 ksi or greater shall not be used in plant changes without engineering approval, due to consideration of stress corrosion cracking vulnerability. Storage requirements for high strength bolts shall include the recommendations of the Research Council for Structural Connections, "Specification for Structural Joints Using ASTM A325 or A490 Bolts", Section 2. Lubricants that contain molybdenum disulfide (MoS

2) shall not be applied to high strength structural bolts within the scope of license renewal.

3. Provide procedural guidance, regarding the selection of supports to be inspected on subsequent inspections, when a support is repaired in accordance with the corrective action program. The enhanced guidance will ensure that the supports inspected on subsequent inspections are representative of the general population.

4. Perform one-time volumetric examinations on a sample of ASTM A490 bolts, greater than one-inch nominal diameter Program to be enhanced and one-time volumetric examinations to be performed prior to the period of extended operation. Section A.2.1.31 Exelon Letter RS-14-052 03/04/2014 RAIs B.2.1.31-1 B.2.1.31-2 B.2.1.31-3

RS-14-052 Enclosure C Page 7 of 8 NO. PROGRAM OR TOPIC COMMITMENT IMPLEMENTATION SCHEDULE SOURCE for the detection of stress corrosion cracking prior to the period of extended operation. Volumetric examinations will be performed in accordance with the requirements of ASME Code Section XI, Appendix VIII, Supplement 8. The sample will consist of bounding and representative A490 bolt sizes, joint configurations, and environmental exposure conditions. The sample will consist of 20% of the ASTM A490 bolts greater than one-inch nominal diameter or a maximum of 25 ASTM A490 bolts total for both Byron and Braidwood stations. The selection of the samples will consider susceptibility to stress corrosion cracking (e.g., actual measured yield strength) and ALARA principles. Any adverse results of the volumetric examinations will be entered into the corrective action program and will be evaluated by engineering to determine if additional actions are warranted such as expansion of sample size, scope, and frequency of any additional supplemental visual or volumetric examinations, as well as any code requirements specified by ASME Section XI, Subsection IWF.

5. Revise implementing documents to perform periodic visual examinations to detect a corrosive environment that supports SCC potential for all (100%) of high strength bolting greater than one-inch nominal diameter prior to the period of extended operation, and then each inspection interval of 10 years thereafter. The periodic visual examinations will include criteria to identify if the bolting has been exposed to moisture or other contaminants by evidence of moisture, residue, foreign substance, or corrosion. Adverse conditions identified during the examinations will be evaluated by engineering to determine if the bolt has been exposed to a corrosive environment with the potential to cause SCC. The bolts determined to have been exposed to corrosive environment with the potential to cause SCC will be included in a sample population for each specific bolt material where SCC is a concern. A sample size equal to 20 percent (rounded up to the nearest whole number) of the bolts in the sample population, with a maximum sample size of 25 bolts will be subject to RS-14-052 Enclosure C Page 8 of 8 NO. PROGRAM OR TOPIC COMMITMENT IMPLEMENTATION SCHEDULE SOURCE supplemental volumetric examination to determine if SCC is present. The selection of the samples will consider susceptibility to stress corro sion cracking (e.g., actual measured yield strength) and ALARA principles. Volumetric examinations will be perfor med in accordance with the requirements of ASME Code Section XI, Appendix VIII, Supplement 8. The results of the volumetric examinations will be evaluated by e ngineering to determi ne if additional actions are warranted such as expansion of sample size, scope, and frequency of any additional supplemental visual or volumetric examinations, as well as any code requirements specified by ASME Section XI, Subsection IWF.