RS-14-150, Responses to NRC Requests for Additional Information, Set 23, Dated April 24, 2014, Related to the Byron Station, Units 1 and 2, and Braidwood Station, Units 1, and 2, License Renewal Application

Responses to NRC Requests for Additional Information, Set 23, Dated April 24, 2014, Related to the Byron Station, Units 1 and 2, and Braidwood Station, Units 1, and 2, License Renewal Application

ML14143A313
Person / Time
Site:	Byron, Braidwood
Issue date:	05/23/2014
From:	Gallagher M Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-14-150
Download: ML14143A313 (39)

Text

10 CFR 50 10 CFR 51 10 CFR 54 RS-14-150 May 23, 2014 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

Subject:

Responses to NRC Requests for Additional Information, Set 23, dated April 24, 2014, related to the Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, License Renewal Application

References:

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

2. Letter from Lindsay R. Robinson, US NRC to Michael P. Gallagher, Exelon, dated April 24, 2014, "Request for Additional Information for the Review of the Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, License Renewal Application, Set 23 (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 staff review of the LRA.

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

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

May 23, 2014 U.S. Nuclear Regulatory Commission Page 2 There are no new or revised regulatory commitments contained in this letter.

If you have any questions, please contact Mr. Al Fulvio, Manager, Exelon License Renewal, at 610-765-5936.

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

Executed on Respectfully,

((!.~@~

Vice President - License Renewal Projects Exelon Generation Company, LLC

Enclosures:

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

RS-14-150 Enclosure A Page 1 of 25 Enclosure A Byron and Braidwood Stations (BBS), Units 1 and 2 License Renewal Application Responses to Requests for Additional Information RAI 3.5.2.2.2.2-1 RAI 4.3.8-1 RAI 4.3.9-1 RAI 4.6.1-1 RAI 4.6.4-1 RAI 4.6.5-1 RAI 4.6.6-1

RS-14-150 Enclosure A Page 2 of 25 RAI 3.5.2.2.2.2-1 Applicability:

Byron Station (Byron) and Braidwood Station (Braidwood)

Background:

Standard Review Plan - License Renewal (SRP-LR) Section 3.5.2.2.2.2, Reduction of Strength and Modulus due to Elevated Temperature, recommends further evaluation for any concrete elements of safety-related structures and other concrete structures that exceed temperature limits of 66°C (150°F) for general areas and 93°C (200°F) for local areas. The SRP-LR also states that higher temperatures may be allowed if tests or calculations are provided to evaluate the reduction in strength and modulus of elasticity and these reductions are applied to the design calculations.

Issue:

License renewal application (LRA) Section 3.5.2.2.2.2 states, in part, that:

High energy line penetrations have been designed to limit surrounding concrete surfaces to temperatures less than 200°F, except for the special pipe whip restraints that are located around each feedwater and main steam pipe as it passes through the concrete wall separating the main steam isolation valve room from the main steam tunnel. The design documents for the concrete at these pipe whip restraints include an evaluation for elevated temperatures, which determined that the concrete temperature up to 300°F at the local areas of the pipes was acceptable.

LRA Table 3.5-1, item 3.5.1-48, which references LRA section 3.5.2.2.2.2, states that the aging effect and mechanism of reduction of concrete strength and modulus due to elevated temperature is not applicable to Byron and Braidwood. It further states that the main steam (MS) tunnel and main steam isolation valve (MSIV) room walls have been evaluated and found acceptable for temperatures up to 300°F. It is not clear if the elevated temperatures experienced by the concrete walls of the MS tunnel and the MSIV room near the special pipe whip restraints remain below the 300°F used in the evaluation and why the AMR line item would not be applicable to these MSIV room and MS tunnel concrete walls. The staff reviewed the Byron and Braidwood updated final safety analyses report (UFSAR), specifically Section 3.8.4, and did not find any discussion of an engineering evaluation that accounted for possible reductions in concrete strength or modulus of elasticity due to elevated temperatures.

Request:

1. Provide the maximum temperature that is experienced by the concrete walls of the MSIV room and MS tunnel near the special feedwater and main steam pipe whip restraints.

2. If the maximum temperature experienced is greater than 200oF, provide a discussion of the engineering evaluation that was conducted to demonstrate the concrete would be able to perform its intended functions while being exposed to elevated temperatures above the Generic Aging Lessons Learned (GALL) Report recommended limits. Include any

RS-14-150 Enclosure A Page 3 of 25 reductions in strength or modulus of elasticity that were applied to the design calculations or any test results used in the evaluation.

3. If the maximum temperature experienced is greater than 200oF, justify why aging management review (AMR) line item 3.5.1-48 is not applicable to the MSIV room and MS tunnel concrete walls that experience elevated temperatures above 200°F and the required evaluation.

Exelon Response:

1. The maximum local temperature recorded on the surface of the concrete walls of the MSIV room and MS tunnel near the special feedwater and main steam pipe whip restraints was 166oF. Thus, there are no reductions in concrete strength or modulus of elasticity due to elevated temperatures to consider. As a result, Request items 2 and 3 are not applicable since the maximum local temperatures experienced on the concrete walls is not greater than 200oF.

In order to clarify the AMR further evaluation, LRA Section 3.5.2.2.2.2 and Table 3.5.1, Item Number 3.5.1-48, are revised as shown in Enclosure B of this letter to explicitly state that the normal operating temperature of the concrete walls near the special feedwater and main steam pipe whip restraints is less than 200oF.

RS-14-150 Enclosure A Page 4 of 25 RAI 4.3.8-1 Applicability:

Byron and Braidwood

Background:

LRA Section 4.3.8, ASME Section III, Subsection NF, Class 1 Component Supports Allowable Stress Analyses, references NRC-approved Westinghouse Owners Group (WOG) Topical Report WCAP-14422, Revision 2-A, License Renewal Evaluation: Aging Management of Reactor Coolant System (RCS) Supports, as a bounding evaluation in the fatigue time-limited aging analyses (TLAA) for the Subsection NF Class 1 component supports for the reactor vessel, steam generator, reactor coolant pump, and pressurizer.

In its final Safety Evaluation Report (FSER), dated November 17, 2000, the staff found Topical Report WCAP-14422, Revision 2-A, acceptable for member plants to reference in a LRA to the extent specified and under the limitations delineated in the staff FSER, which includes completing the renewal applicant action items described in Section 4.1 of the FSER. In FSER Section 4.1, Renewal Applicant Action Item 6 Fatigue (Section 3.3.1.7), it states that a license renewal applicant will have to justify differences between the materials used for its RCS supports and the values listed in Table 2-4 of the topical report.

Issue:

LRA Section 4.3.8, in the first paragraph under TLAA Evaluation, states in part that: The number of transients analyzed in the report [WOG Report WCAP-14422, Revision 2-A] are bounded by the transient limits shown in Section 4.3.1.

Further, LRA Section 4.3.8, in the second paragraph under TLAA Evaluation states in part that:

Even though BBS Class 1 component supports were designed to ASME Section III, Subsection NF 1974 through the 1975 Summer Addenda requirements and not the AISC 1963 Edition, design documents were reviewed and the majority of installed Class 1 component support materials were found in the WCAP Table 2-4. Several materials were identified that were not documented in the table. Evaluation of these different materials showed that their yield strength and fatigue resistance properties are consistent with materials in WCAP-14422, Table 2-4 or the materials are used in bearing plates which do not experience cyclical tensile stresses.

It is not clear to the NRC staff which specific transient limits in LRA Section 4.3.1 were considered by the applicant in making the comparison to the number of transients analyzed in the WOG Report WCAP-14422. Further, the information provided above, in LRA Section 4.3.8 with regard to License Renewal Applicant Action Item 6 in the NRC FSER for Topical Report WCAP-14422, Revision 2-A, is not sufficient for the staff to verify that the fatigue evaluation in the topical report WCAP-1422 remains bounded for materials used in the Byron and Braidwood RCS supports that are not listed in Table 2-4 of the topical report.

RS-14-150 Enclosure A Page 5 of 25 Request:

1. Identify the specific transient (cycle) limits in LRA Section 4.3.1 that were used in LRA Section 4.3.8 to make the comparison with the number of transients analyzed in the WOG Topical Report WCAP-14422, Revision 2-A.

2. Provide a list of the materials used in the Byron and Braidwood RCS Class 1 component supports that are not documented in Table 2-4 of Topical Report WCAP-14422, Revision 2-A. For these materials, include information such as the yield strength, fatigue resistance properties (with number of loading cycles implicit in the design allowable using these materials), the component, and the function of the component used that would allow the NRC staff to verify that the fatigue evaluation in the topical report remains bounding for the components using these materials.

Exelon Response:

1. There are three specific transients and associated cycle limits in LRA Section 4.3.1 that were considered in LRA Section 4.3.8. These are the same transients and cycle limits that were analyzed in the WOG Topical Report WCAP-14422, Revision 2-A. These transients are identified in LRA Tables 4.3.1-1 and 4.3.1-4 as Transient 1, Plant Heatup at 100°F/hr and Transient 2, Plant Cooldown at 100°F/hr each with CLB cycle limits of 200; and Transient 34 Operating Basis Earthquake (OBE) with a CLB cycle limit of 20 OBE seismic events each with 20 sub-cycles for a total of 400 cycles. Overall this results in a combined limit of 600 cycles (i.e., 200 heatup and cooldown cycles plus 400 OBE cycles) that were considered in the LRA.

WCAP-14422, Revision 2-A concludes that the transients that affect RCS Class 1 component supports are thermal cycles due to plant heatup and cooldown transients and cycles during Operating Basis Earthquake (OBE) seismic events. The number of representative cycles identified in the WCAP-14422, Revision 2-A, Table 3-2, are 200 thermal cycles for heatup and cooldown transients (equivalent to 200 Plant Heatup at 100°F/hr cycles and 200 Plant Cooldown at 100°F/hr cycles as accounted in LRA Tables 4.3.1-1 and 4.3.1-4) and 400 cycles for OBE seismic events; which have a combined limit of 600 cycles.

Therefore, the LRA combined cycle limit of 600 is the same as the number of transients analyzed in the WOG Topical Report WCAP-14422, Revision 2-A.

2. The Byron and Braidwood Stations, Units 1 and 2, (BBS) RCS Class 1 component supports were designed, fabricated, and installed in accordance with the requirements of ASME Boiler and Pressure Vessel Code Section III, Division I, Subsection NF, 1974 Edition with Summer 1975 addendum (ASME Section III). RCS Class 1 component support materials conform to ASME material requirements, or ASTM materials meeting the requirements of ASME Code Case 1644, Additional Materials for Component Supports and Alternate Design Requirements for Bolted Joints Section Ill Division 1, Subsection NF Class, 1, 2, 3, and MC Construction. The references to ASME or ASTM designations in the following sections are based upon the designation shown in design documents.

Section 3.3.1.7 of the NRC Final Safety Evaluation Report (FSER) associated with WCAP-14422, Revision 2-A, which was approved in 2000, concludes that fatigue is not a concern

RS-14-150 Enclosure A Page 6 of 25 for RCS Class 1 component support materials documented on Table 2-4 of WCAP-14422, Revision 2-A, (WCAP Table 2-4). However, some plant specific BBS RCS Class 1 component support materials are not included in WCAP Table 2-4, which only lists the most commonly specified materials. For the materials not listed in WCAP Table 2-4, action item 6 of the WCAP requires that the license renewal applicant must justify that the fatigue properties of the materials not in the WCAP Table 2-4 are consistent with the fatigue properties of the materials listed in WCAP Table 2-4. This justification was performed during the development of LRA Section 4.3.8 and the results are described below.

At Byron and Braidwood Stations, Units 1 and 2, four subcomponent types associated with RCS Class 1 component supports are made of materials that are not specifically documented in WCAP Table 2-4. These subcomponents may experience stresses during the heatup and cooldown or during OBE transients, which have a combined limit of 600 cycles, as described above in the response to Request 1. The material specification, yield strengths, associated subcomponents, associated subcomponent functions, fatigue resistance properties, and consistency with material properties in WCAP Table 2-4 are discussed below, as appropriate, for each subcomponent.

Steam Generator Lower Lateral Support Inner Frame Structural Plates The steam generator lower lateral support inner frame structural plates are made of ASME SA533 Class 2 steel, 6-1/2 thick plate (shown as inner frame on UFSAR figure 3.9-7a). The function of this subcomponent is to support the steam generators during plant operation and under design basis conditions, including an OBE. ASME SA533 Class 2 is approved for use in RCS Class 1 component supports as documented in ASME Section III, in Table I-1.1. This material has a yield strength of 70 ksi.

WCAP Table 2-4 documents other materials used as structural plates, such as ASTM A588 Grade A or B and ASTM A572 Grade 42 both with yield strengths of 42 ksi.

With respect to fatigue resistance properties, the application of Figure I-9.1 in ASME Section III 1974 edition, Division I, in Appendix I, results in ASME SA533 Class 2 material having essentially the same number of loading cycles, as the other material documented in WCAP Table 2-4 (e.g., ASTM A588 Grade A or B and ASTM A572 Grade 42). Therefore, the use of ASME SA533 Class 2 for structural plates in RCS Class 1 component supports is consistent with WCAP Table 2-4 and fatigue of this material is not a concern.

Steam Generator Upper Lateral Support Snubber End-Blocks The Byron and Braidwood steam generator upper lateral support snubber end-blocks are made of ASME SA533 Grade B, Class 1 steel, ASME SA516 Grade 70 steel, or ASTM A36 steel (illustrated in UFSAR figures 3.9-6 and 3.9-8). The function of this subcomponent is to anchor the steam generator upper lateral support snubber to the containment structure during transient conditions and under design basis conditions, including an OBE. ASME SA533 Grade B, Class 1 material has a yield strength of 50 ksi, and is approved for use in RCS Class 1 component supports as documented in ASME Section III, in Table I-1.1. ASME SA516 Grade 70 is a carbon steel material with a yield strength of 36 ksi, and is approved for use in RCS Class 1 component supports as documented in ASME Section III, Table I-1.1. ASTM A36 is a carbon steel material

RS-14-150 Enclosure A Page 7 of 25 with a yield strength of 36 ksi, and is approved for in RCS Class 1 component supports as documented in ASME Section III, Table I-11.1.

WCAP Table 2-4 documents another material used as structural members, ASTM A572 Grade 42, which has a yield strength of 42 ksi.

Although the yield strength of material ASTM A572 Grade 42 in WCAP Table 2-4 (42 ksi) is greater than the yield strengths of ASME SA516 Grade 70 (36 ksi) and ASTM A36 (36 ksi), the fatigue resistance properties are the same. The fatigue resistance properties were determined based on the application of Figure I-9.1 in ASME Section III 1974 edition, Division I, in Appendix I, results in ASME SA533 Grade B, Class 1, ASME SA516 Grade 70 steel, and ASTM A36 steel materials having the same number of loading cycles, as other material documented in WCAP Table 2-4 (e.g., ASTM A572 Grade 42). Therefore, the use of these materials in RCS Class 1 component support, snubber end-blocks is consistent with Table 2.4 and fatigue of this material in RCS Class 1 component supports is not a concern.

ASME SA193 Grade B7 Bolting Various BBS RCS Class 1 component support bolting is made of ASME SA193 Grade B7 material. The function of these subcomponents is to connect steel elements under normal and transient conditions and under design basis conditions, including an OBE.

ASME SA193 Grade B7 is approved for use in RCS Class 1 component supports as documented in ASME Section III, Table I-13.3. The material has yield strength of 75 to 105 ksi, depending on the size of the bolt.

WCAP Table 2-4 documents other bolting materials, such as ASTM A354 Grade BC which has yield strength of 99 to 109 ksi, depending on the size of the bolt, and ASTM A540 Grade B-23 Class 4, which has yield strength of 120 ksi.

Although the yield strengths of the material in the WCAP Table 2-4 are greater than the yield strength of the ASTM A193 Grade B7 material, the fatigue resistance properties are the same as can be determined from the application of Figure I-9.4. The fatigue resistance properties were determined based on the application of Figure I-9.4 in ASME Section III 1974 edition, Division I, in Appendices I, which resulted in concluding that all these material have the same number of loading cycles. Therefore, the use of ASTM A193 Grade B7 for bolting in RCS Class 1 component supports is consistent with WCAP Table 2-4 and fatigue of this material in RCS Class 1 component support bolting is not a concern.

Shim Plate and Spacer Materials Various carbon steel materials are used as shim plates or spacers for BBS RCS Class 1 component supports. Of these, ASTM/ASME A36, A53, A366, A414, A569, A570, A606 Type 4, A607, and A1008 CS Type B are not in WCAP Table 2-4 of WCAP-14422, Revision 2-A. These shim plates and spacers are designed for compression loads and are not subject to tension or bending loads and thus would not experience cyclic tensile stresses and fatigue.

RS-14-150 Enclosure A Page 8 of 25 Conclusion The BBS RCS Class 1 component support design information provided above demonstrate that the fatigue properties of the materials not explicitly listed in WCAP Table 2-4 are consistent with the fatigue properties of the materials listed on WCAP Table 2-4. As a result, the BBS RCS Class 1 component supports made from these materials are inherently designed for a minimum of 20,000 stress cycles as discussed in LRA Section 4.3.8, while the Fatigue Monitoring program has a combined cycle limit of 600 for the applicable transients. Therefore, the fatigue evaluation in the WCAP is bounding for the RCS Class 1 component supports using these materials at BBS.

RS-14-150 Enclosure A Page 9 of 25 RAI 4.3.9-1 Applicability:

Byron and Braidwood

Background:

The TLAA evaluation for spent fuel pool liner and spent fuel storage racks in LRA Section 4.3.9, "Fatigue Design of Spent Fuel Pool Liner and Spent Fuel Storage Racks for Seismic Events,"

states, in part, that:

The analyses include a fatigue evaluation of the replacement spent fuel storage racks and the spent fuel pool liner for the cyclic loads imposed by twenty (20) operating basis earthquake (OBE) plus one (1) safe shutdown earthquake (SSE) event...OBE events are monitored by the Fatigue Monitoring (B.3.1.1) program ... The Fatigue Monitoring (B.3.1.1) program will continue to be used to manage fatigue of these components through the period of extended operation by monitoring Operating Basis Earthquake (OBE) and (Safe Shutdown Earthquake) SSE events.

Issue:

The program descriptions in LRA Section B.3.1.1, "Fatigue Monitoring;" LRA Section A.3.1.1, "Fatigue Monitoring;" and the GALL Report Section X.M1, "Fatigue Monitoring," appear focused on monitoring and tracking critical thermal and pressure transients for selected components. It is not clear to the staff whether the LRA Section B.3.1.1, "Fatigue Monitoring" program, credited in the disposition of the TLAA in LRA Section 4.3.9 in accordance with 10 CFR 54.21(c)(1 )(iii),

includes under its scope: (a) the spent fuel pool liner and spent fuel storage racks as components and (b) load cycles from OBE and SSE events as parameters monitored and tracked.

Further, it is not clear as to the number of specific load cycles considered in the fatigue evaluation for each OBE event and the SSE event that would define the total bounding limit of seismic load cycles that should not be exceeded when monitored by the credited Fatigue Monitoring program in LRA Sections B.3.1.1 and A.3.1.1.

Request:

1. Clarify whether the LRA Section B.3.1.1, "Fatigue Monitoring" program, credited in the disposition of the TLAA in LRA Section 4.3.9 in accordance with 10 CFR 54.21(c)(1)(iii),

includes under its scope: (a) the spent fuel pool liner and spent fuel storage racks as components and (b) load cycles from OBE and SSE events as parameters monitored and tracked. Update LRA Sections B.3.1.1 and A.3.1.1, as necessary, based on this request.

2. Identify the number of specific load cycles, considered in the fatigue evaluation of the spent fuel storage racks and spent fuel pool liner in LRA Section 4.3.9, for each OBE event and the SSE event. Update LRA Section A.4.3.9, as necessary, based on the response to this request

RS-14-150 Enclosure A Page 10 of 25 Exelon Response:

1. The Fatigue Monitoring program described in LRA Section B.3.1.1 includes within its scope:

(a) The spent fuel pool liner and the replacement spent fuel storage rack components, which are analyzed for fatigue in accordance with ASME Section III. LRA Sections A.3.1.1 and B.3.1.1 are revised as shown in Enclosure B to add the words other components which includes the spent fuel pool liner and the replacement spent fuel storage racks, as well as other components.

(b) The specific load cycles, which define the OBE and SSE seismic events, are parameters which confirm the transient event has occurred, and which are monitored and tracked.

The event occurrence, either an OBE or SSE, is monitored and tracked by the Fatigue Monitoring program. The number of cycles occurring per event are evaluated as part of the event analysis. The program procedures evaluate seismic events using the parameters of duration, magnitude, and cycles of the event. Since one-time, faulted events, are not normal inputs to fatigue monitoring, the SSE event is not in LRA Tables 4.3.1-1 through 4.3.1-6. However, the SSE is an event monitored by the current Fatigue Monitoring program.

LRA Sections 4.3.9, A.4.3.9, A.3.1.1, and B.3.1.1 are revised to provide further clarification as shown in Enclosure B.

2. The number of specific load cycles, considered in the fatigue evaluation of the spent fuel storage racks and the spent fuel pool liner in LRA Section 4.3.9, are 25 cycles for each OBE event and 20 cycles for the single SSE event. LRA section 4.3.9 is revised as presented in Enclosure B to include this additional information requested.

RS-14-150 Enclosure A Page 11 of 25 RAI 4.6.1-1 Applicability:

Byron and Braidwood

Background:

LRA Sections 4.6.1, "Containment Liner Plates Fatigue"; Section 4.6.2, "Containment Airlocks and Hatches Fatigue"; and Section 4.6.3, "Containment Electrical Penetrations Fatigue,"

describe analyses performed per Subparagraph NE-3222.4(d) of ASME Section Ill, Division 1, associated with the containment structures and not requiring analysis for cyclic operation.

These analyses have been identified as TLAAs because the original analysis involved 40-year design inputs. For each of the aforementioned sections, the applicant dispositioned the analysis in accordance with 10 CFR 54.21(c)(1)(iii), indicating that the effects of aging on the intended function(s) will be adequately managed for the period of extended operation by LRA Section B.3.1.1, "Fatigue Monitoring," an enhanced AMP that the applicant claims to be consistent with the GALL Report AMP X.M1. The applicant's AMP monitors transient cycles to ensure that the considered design limits for the liner plates, airlocks and hatches, and electrical penetrations are not exceeded during the period of extended operation.

Issue:

For LRA Sections 4.6.1, 4.6.2, and 4.6.3, the applicant states that "[t]he Fatigue Monitoring (B.3.1.1) program will be used to monitor the 'applicable cycles' and ensure that the transient limits will not be exceeded during the period of extended operation." The LRA also states that the 60-year transient projections provided in LRA Section 4.3.1 show that the transient limits will not be exceeded during the period of extended operation.

The staff reviewed UFSAR Section 3.9.1.1, "Design Transients," and LRA Tables 4.3.1 4.3.1-6, but it was not clear which of the listed transient limits were considered in the original design analyses, which concluded that the conditions specified in ASME Section Ill, Division 1, Subparagraph NE-3222.4(d) for waiver were met and that no analysis for cyclic operation was required. The staff needs to verify that the transients considered in the ASME NE-3222.4(d) evaluations are being monitored by the Fatigue Monitoring program and that the "applicable cycles" are clearly identified to ensure that the limits considered in the original design analyses will not be exceeded during the period of extended operation.

Request:

With respect to the six design inputs meeting the conditions of ASME Section Ill, Division 1, Subparagraph NE-3222.4(d),

1. Indicate which transients were considered in each of the design analyses described in LRA Section 4.6.1, 4.6.2, and 4.6.3, which will be monitored by the Fatigue Monitoring program to support the original waiver evaluation that an analysis for cyclic operation is not required for a fatigue analysis of these components; and

RS-14-150 Enclosure A Page 12 of 25

2. Provide the number of stress cycles that was assumed in the original design analyses, as well as the number of additional cycles anticipated for LRA Sections 4.6.1, 4.6.2, and 4.6.3 during the period of extended operation.

Exelon Response:

1. LRA Sections 4.6.1, "Containment Liner Plates Fatigue"; Section 4.6.2, "Containment Airlocks and Hatches Fatigue"; and Section 4.6.3, "Containment Electrical Penetrations Fatigue, describe analyses performed per Subparagraph NE-3222.4(d) of ASME Section Ill, Division 1, associated with the containment structures and not requiring analysis for cyclic operation. The transients considered in each of the design analyses described in LRA Section 4.6.1, 4.6.2, and 4.6.3 are listed in Table 1 of this RAI response. The transients will be monitored by the Fatigue Monitoring program to support the original waiver evaluation to ensure that an analysis for cyclic operation is not required for a fatigue analysis of these components.

The original design analysis for these components determined that the design inputs (operating conditions) met the exemption conditions of ASME Section III, Subparagraph NE-3222.4(d) and that no fatigue analysis was required. The six conditions addressed are the following:

(1) Atmospheric-to-Operating Pressure Cycles (2) Normal Operation Pressure Fluctuations (3) Temperature Difference - Startup and Shutdown (4) Temperature Difference - Normal Operation (5) Temperature Difference - Dissimilar Materials (6) Mechanical Loads The verification that the conditions are met is dependent on the specified number and magnitude of pressure and temperature transients and mechanical load cycles. These assumed inputs are then used to assess the potential effect on the component and consideration of other limitations to determine if the fatigue waiver can be applied. The original design specifications provided the expected number and magnitude of pressure and temperature transients and mechanical load cycles to be considered for the fatigue waiver in the original design analysis. As part of the evaluation of the TLAAs, since the original basis for the transient inputs were not provided, correlation of the inputs to design basis transients was necessary.

Table 1 below provides the review of each of the exemption conditions, and the corresponding LRA transients representative of the transient inputs to the fatigue waiver in the original design analysis. Most of these inputs are directly related to the transients in LRA Tables 4.3.1-1 and 4.3.1-4. For condition 2 above, the Normal Operation Pressure Fluctuations condition is not considered in the LRA Section 4.3.1 tables and not monitored by the fatigue monitoring program. This is justified because the pressure fluctuations during normal operation are insignificant and the maximum number of cycles for a Type A leak test (15) are insignificant compared to the minimum number of cycles (2500) assumed in the design analysis. For conditions 4 and 5 above, it was conservatively interpreted that the temperature differences could be the result of not only heatups and cooldowns but also upset conditions. Therefore, as shown in Table 1, a number of transients were associated

RS-14-150 Enclosure A Page 13 of 25 with these two conditions. The total number of 60-year projected transients is still considerably lower than the values in the original design analysis.

The transients which are considered significant in each of the design analyses described in LRA Section 4.6.1, 4.6.2, and 4.6.3, are provided in Table 1 with this RAI response. The LRA transients listed in the table are currently monitored by the Fatigue Monitoring program.

Monitoring of these transients supports the original waiver evaluations that an analysis for cyclic operation is not required for a fatigue analysis of these components for the period of extended operation.

2. The number of stress cycles that was assumed in the original design analyses are tabulated in Table 1 below. There were no additional cycles, above those used as inputs for the exemption, anticipated for component analyses in LRA Sections 4.6.1, 4.6.2, and 4.6.3 during the period of extended operation. All 60-year cycle projections for the transients considered to be those associated with the fatigue exemptions were considerably below the design fatigue exemption analysis cycle inputs. The current license basis (CLB) Cycle Limits are equal to or bounded by the fatigue exemption cycles assumed in the original design analysis. The Fatigue Monitoring program will continue to monitor and track the transients of concern to ensure that the input values considered in the original design analyses will not be exceeded during the period of extended operation.

RS-14-150 Enclosure A Page 14 of 25 Table 1 Byron and Braidwood Units 1 and 2 Fatigue Exemption Inputs Assessment ASME Section Ill, Corresponding LRA Maximum 60- LRA LRA LRA LRA Table 4.3.1-1, and Division 1, LRA Transients Table year Cycle Section Section Section LRA Table 4.3.1-4 CLB Subparagraph Projection from 4.6.1 4.6.2 4.6.3 Cycle Limit NE-3222.4(d) LRA Tables Cycles Cycles Cycles Condition 4.3.1-1 and Assumed Assumed Assumed 4.3.1-4 in Original in Original in Original Design Design Design Analysis Analysis Analysis (1) Plant Heatup and 4.3.1-1 117 2500 2500 4000 200 Atmospheric to Plant Cooldown, and Operating Pressure Transients 1 and 2 4.3.1-4 Cycles (2) None, NA 15 (Note 1) 2500 2500 1,000,000 Not in the LRA Tables Normal Operation corresponding and based on the large Pressure pressure transient margins, monitoring is not Fluctuations would be associated required.

with the Appendix J, Type A leak test (3) Plant Heatup and 4.3.1-1 117 200 200 1,000,000 200 Temperature Plant Cooldown, and Difference - Start- Transients 1 and 2 4.3.1-4 up and Shutdown (4) Plant Heatup, 4.3.1-1 117 4000 4000 1,000,000 200 Temperature Transient 1 and Difference - 4.3.1-4 Normal Operation Plant Cooldown, 4.3.1-1 117 200 Transient 2 and 4.3.1-4

RS-14-150 Enclosure A Page 15 of 25 Table 1 Byron and Braidwood Units 1 and 2 Fatigue Exemption Inputs Assessment ASME Section Ill, Corresponding LRA Maximum 60- LRA LRA LRA LRA Table 4.3.1-1, and Division 1, LRA Transients Table year Cycle Section Section Section LRA Table 4.3.1-4 CLB Subparagraph Projection from 4.6.1 4.6.2 4.6.3 Cycle Limit NE-3222.4(d) LRA Tables Cycles Cycles Cycles Condition 4.3.1-1 and Assumed Assumed Assumed 4.3.1-4 in Original in Original in Original Design Design Design Analysis Analysis Analysis Loss of Load, 4.3.1-1 6 80 Transient 20 and 4.3.1-4 Loss of Power, 4.3.1-1 6 40 Transient 21 and 4.3.1-4 Reactor Trip from 4.3.1-1 38 160 Full Power: Case B - and with Cooldown and 4.3.1-4 no SI, Transient 24 Reactor Trip from 4.3.1-1 5 10 Full Power - Case and C - with Cooldown 4.3.1-4 and SI, Transient 25 Inadvertent Safety 4.3.1-1 9 60 Injection (ECCS) and Actuation, 4.3.1-4 Transient 29 Total NA 298 4000 4000 1,000,000 750

RS-14-150 Enclosure A Page 16 of 25 Table 1 Byron and Braidwood Units 1 and 2 Fatigue Exemption Inputs Assessment ASME Section Ill, Corresponding LRA Maximum 60- LRA LRA LRA LRA Table 4.3.1-1, and Division 1, LRA Transients Table year Cycle Section Section Section LRA Table 4.3.1-4 CLB Subparagraph Projection from 4.6.1 4.6.2 4.6.3 Cycle Limit NE-3222.4(d) LRA Tables Cycles Cycles Cycles Condition 4.3.1-1 and Assumed Assumed Assumed 4.3.1-4 in Original in Original in Original Design Design Design Analysis Analysis Analysis (5) Plant Heatup, 4.3.1-1 117 8000 8000 1,000,000 200 Temperature Transient 1 and Difference - 4.3.1-4 Dissimilar Materials Plant Cooldown, 4.3.1-1 117 200 Transient 2 and 4.3.1-4 Loss of Load, 4.3.1-1 6 80 Transient 20 and 4.3.1-4 Loss of Power, 4.3.1-1 6 40 Transient 21 and 4.3.1-4 Reactor Trip from 4.3.1-1 38 160 Full Power: Case B - and with Cooldown and 4.3.1-4 no SI, Transient 24 Reactor Trip from 4.3.1-1 5 10 Full Power - Case and C - with Cooldown 4.3.1-4 and SI, Transient 25

RS-14-150 Enclosure A Page 17 of 25 Table 1 Byron and Braidwood Units 1 and 2 Fatigue Exemption Inputs Assessment ASME Section Ill, Corresponding LRA Maximum 60- LRA LRA LRA LRA Table 4.3.1-1, and Division 1, LRA Transients Table year Cycle Section Section Section LRA Table 4.3.1-4 CLB Subparagraph Projection from 4.6.1 4.6.2 4.6.3 Cycle Limit NE-3222.4(d) LRA Tables Cycles Cycles Cycles Condition 4.3.1-1 and Assumed Assumed Assumed 4.3.1-4 in Original in Original in Original Design Design Design Analysis Analysis Analysis Inadvertent Safety 4.3.1-1 9 60 Injection (ECCS) and Actuation, 4.3.1-4 Transient 29 Total Cycles NA 298 4000 4000 1,000,000 750 (6) None. Mechanical NA NA 0 0 (Note 3) NA Mechanical Loads Load fluctuations (Note 2) (Note 2) are not a part of the Fatigue Monitoring program requirements.

Notes:

1. Normal operating pressure fluctuations are insignificant and assuming Type A tests are conducted at the maximum frequency of once every 48 months the maximum number of tests would be 15.

2. The original design specification states that there are no significant mechanical load fluctuations on these components.

3. 1,000,000 cycles were acceptable based on the sum of the stress values associated with the combination of a Seismic (SSE) and Jet Loads (Accident). The number of cycles justified for these faulted events does not require monitoring the events for these components.

RS-14-150 Enclosure A Page 18 of 25 RAI 4.6.4-1 Applicability:

Byron and Braidwood

Background:

LRA Section 4.6.4, Containment Piping Penetrations Fatigue, states that the containment structure instrument and process pipe penetrations were analyzed in accordance with design specifications, subject to Subparagraphs NE-3222.4(e) or NB-3222.4(e), Procedure for Analysis for Cyclic Loading, of ASME Section III, Division 1, which required fatigue evaluations of each containment structure piping penetration. The LRA further states that the design specifications for the containment piping penetrations define the transients applicable to penetration stress analyses and that these same transients are listed in Section 4.3.1. The applicant dispositioned this TLAA in accordance with 10 CFR 54.21(c)(1)(iii), indicating that the effects of aging on the intended function(s) will be adequately managed for the period of extended operation by LRA Section B.3.1.1, Fatigue Monitoring, an enhanced AMP that the applicant claims to be consistent with GALL Report AMP X.M1. The Fatigue Monitoring AMP states that it monitors and tracks critical thermal and pressure transients to ensure analyzed components do not exceed the number of allowable cycles and that the cumulative usage factor (CUF) for each component remain less than 1.0 through the period of extended operation.

Issue:

The staff reviewed UFSAR Section 3.8.2, Steel Containment and ASME Class MC Components, and noted that penetration sleeves are designed as Class MC components in accordance with Subsection NE of the ASME B&PV Code,Section III, except for their head fittings, directly exposed to worst case loading conditions of the process piping. These are designed in accordance with Subsection NB, NC, or ND of the ASME Code,Section III, as applicable. The staff also noted that fatigue loading conditions include thermal and pressure load transients as well those of OBE and other mechanical loads. It is not clear which of the transients listed in LRA Section 4.3.1 were considered for stress differences in the analyses of penetration sleeves, which will be monitored by the Fatigue Monitoring Program, to support the containment piping penetration fatigue analyses.

Request:

State the applicable transients, including the cycle limit for each transient that was assumed in the fatigue analysis for the containment piping penetrations, which will be monitored to demonstrate that the effects of aging on the intended functions of the containment piping penetrations will be adequately managed by the Fatigue Monitoring Program during the period of extended operation.

RS-14-150 Enclosure A Page 19 of 25 Exelon Response:

Tables 1 and 2 below, document the pressure and temperature transients and the numbers of cycles that were assumed in the fatigue analyses for the containment piping penetrations, and also document the corresponding transient numbers and CLB cycle limits from LRA Tables 4.3.1-1, 4.3.1-2, 4.3.1-4, and 4.3.1-5. These fatigue analyses also applied loads associated with Operating Basis Earthquakes (OBE) in combination with loads created by the pressure and temperature transients.

Table 1 Correlation of Containment Piping Penetration Analyses Assumed Transients and Limits to LRA RCS Transients and Limits Contained in LRA Tables 4.3.1-1 and 4.3.1-4 Containment Piping Most Limiting Number of LRA Tables LRA Tables Penetration Design Cycles Assumed in the 4.3.1-1 and 4.3.1-1 and Analyses Transient Containment Piping 4.3.1-4 4.3.1-4 CLB Description Penetrations Design Transient Cycle Limits Analyses Number(s)

RCS Heatup 200 (Note 1) 1 200 RCS Shutdown 200 (Note 1) 2 200 Loading 0% - 15% Power 200 3 330 (Note 2)

Unloading 15% - 0% 200 4 500 (Note 2)

Power Load @ 5%/min (15% to 200 5 13,200 100% power) (Note 2)

Unload @ 5%/min (100% 200 6 12,240 to 15% power) (Note 2)

Auxiliary Feedwater 700 (Note 3) 20, 21, 22, 23, 610 (Note 3)

System Initiation 25, 26, 27, 28, Following a Reactor Trip and 29 with Loss of Power Table 2 Correlation of Containment Piping Penetration Analyses Assumed Transients and Limits to LRA Auxiliary System Transients and Limits Contained in LRA Tables 4.3.1-2 and 4.3.1-5 Containment Piping Most Limiting Number of LRA Tables LRA Tables Penetration Design Cycles Assumed in the 4.3.1-2 and 4.3.1-2 and Analyses Transient Containment Piping 4.3.1-5 4.3.1-5 CLB Description Penetrations Design Transient Cycle Limits Analyses Number(s)

Letdown Flow Shutoff 200 5 20 Delayed Return to Service Letdown Flow Isolation 200 6 200 Charging Flow 50% 24,000 8 24,000 Increase and Return Letdown Flow 50% 2,000 9 2,000 Decrease and Return Letdown Flow 60% 24,000 10 24,000 Increase and Return

RS-14-150 Enclosure A Page 20 of 25 Containment Piping Most Limiting Number of LRA Tables LRA Tables Penetration Design Cycles Assumed in the 4.3.1-2 and 4.3.1-2 and Analyses Transient Containment Piping 4.3.1-5 4.3.1-5 CLB Description Penetrations Design Transient Cycle Limits Analyses Number(s)

Charging and Letdown 100 13 60 Flow Shutoff and Return to Service Sampling Line and Nozzles 46,000 14 NA (Note 4)

Transients RCP Thermal Barrier 40 NA (Note 5) NA (Note 5)

Failure Tables 1 and 2 Notes:

1. 200 cycles is the most limiting number (least number) of cycles assumed for this transient in the containment piping penetration fatigue analyses.

2. As described in LRA Section 4.6.4, Containment Piping Penetrations Fatigue, the cycle numbers of loading and unloading transients were underestimated in the original Main Steam and Feedwater containment piping penetration analyses and were not consistent with the governing Westinghouse transient design specification. Review of these analyses show that there is sufficient margin to accommodate the greater number of transients in the Westinghouse specification. Corrective action is being taken to revise the analyses to increase the number of cycles to the original CLB cycle limits presented in Tables 4.3.1-1 and 4.3.1-4 for these transients.

3. The feedwater and main steam line penetration fatigue analyses assumed a conservative case (umbrella case) in which auxiliary feedwater system is initiated following a reactor trip with loss of power. This umbrella case was assumed to occur 700 times in the fatigue analyses. The breakdown of specific transients which make up the umbrella case were specified in the Westinghouse design specification and are listed in LRA Tables 4.3.1-1 and 4.3.1-4 as transients 20, 21, 22, 23, 25, 26, 27, 28, and 29. The umbrella case also included transients which are emergency and faulted events. Since one time, emergency and faulted events are not normal inputs for fatigue monitoring, these events are not in LRA Tables 4.3.1-1 through 4.3.1-6. However, these events are monitored by the current Fatigue Monitoring program.

4. Refer to Exelons response to RAI 4.3.1-4 in letter RS-14-084, dated March 28, 2014.

Chemistry samples are taken at a much lower frequency than that which was assumed in the design, resulting in fewer thermal cycles. Also, samples are no longer taken from the RCS as had been specified in the original design, but are taken from the letdown system in which temperatures differences are much lower than those postulated in the original design, resulting in lower transient severity. As a result, fatigue is not a concern. As noted in Exelons response to RAI 4.3.1-4, the number of design transients for the sampling line piping was assumed to be 45,000 cycles over the 40-year plant life. However, the fatigue analysis for the associated containment piping penetration which contains the sampling line conservatively assumed 46,000 cycles over the 40-year plant life.

RS-14-150 Enclosure A Page 21 of 25

5. The fatigue analysis for the containment piping penetration associated with the reactor coolant pump (RCP) thermal barrier cooling coil Component Cooling System return line conservatively assumed 40 cycles. During this transient, a RCP thermal barrier cooling coil is postulated to fail and heat up the associated component cooling system return line and the associate containment piping penetration. A postulated failure of a RCP thermal barrier cooling coil would result in a small break LOCA which is an emergency condition.

Since onetime, emergency and faulted events are not normal inputs for fatigue monitoring, the small break LOCA event is not in LRA Tables 4.3.1-1 through 4.3.1-6. However, small break LOCA is an event monitored by the current Fatigue Monitoring program.

The transients and cycle limits documented in the above tables reflect what was assumed in the original fatigue analysis for the containment piping penetrations, are currently monitored by Fatigue Monitoring aging management program implementing procedures, and will continue to be monitored during the period of extended operation.

RS-14-150 Enclosure A Page 22 of 25 RAI 4.6.5-1 Applicability:

Byron and Braidwood

Background:

LRA Section 4.6.5, "Fuel Transfer Tube Bellows Fatigue," states that the design specification for the bellows includes design load cycles equal to 100 cycles and the design is based on ASME Section Ill, Subsection NE-3365.2(e)(2), through testing of duplicate bellows according to the 1974 Edition through Summer 1974 (Addenda). The LRA further states that the specification of 100 design load cycles along with maximum displacements are intended to envelope all postulated design basis conditions, including 1 SSE transient event. The applicant dispositioned this TLAA in accordance with 10 CFR 54.21 (c)(1 )(iii), indicating that the effects of aging on the intended function(s) will be adequately managed for the period of extended operation by LRA Section B.3.1.1, "Fatigue Monitoring," an enhanced AMP that the applicant claims to be consistent with the GALL Report AMP X.M1. The Fatigue Monitoring AMP states that it monitors and tracks critical thermal and pressure transients to ensure analyzed components do not exceed the number of allowable cycles.

Issue:

LRA Section 4.6.5 states that these bellows are limited to 100 design load cycles and maximum displacements, including 1 SSE event, which would cause deflection of the bellows, and that these transient cycle projections listed in Section 4.3.1 are less than the numbers of cycles which form the basis for the design requirement of the bellows. The transients listed in LRA Section 4.3.1 do not include the SSE event. It is not clear that the Fatigue Monitoring Program is monitoring the SSE event to support the applicant's claim that the effects of aging on the fuel transfer tube bellows will be adequately managed by the Fatigue Monitoring Program.

Request:

1. State what transients along with maximum displacements, other than those associated with SSE, have been considered in the fuel transfer tube bellows fatigue analysis, and provide the number of cycles assumed in the design analysis, along with the maximum displacements, to demonstrate that the effects of aging on the intended functions will be adequately managed by the Fatigue Monitoring Program.

2. Clarify why the SSE transients are not listed in the tables in LRA Section 4.3.1, or revise the LRA as necessary.

Exelon Response:

1. The SSE and LOCA events are the only transients considered in the analysis for the fuel transfer bellows. As described in LRA Section 4.6.5, TLAA Description, the 100 design load cycles envelope the postulated design basis conditions. Therefore, there are no other transients associated with the analysis. The maximum displacements specified were 1.75 inches axially and 0.5 inches laterally. The Fatigue Monitoring program monitors and tracks SSE and LOCA events. If a seismic event occurs, the program reviews the duration,

RS-14-150 Enclosure A Page 23 of 25 magnitude, and number of cycles of the event. LRA Sections A.3.1.1 and B.3.1.1 are revised as shown in Enclosure B to provide clarification that the Fatigue Monitoring program (LRA Section A.3.1.1) manages the cumulative fatigue damage of other components and monitors design basis events and counts them in the appropriate design transient category. Therefore, since the Fatigue monitoring program currently monitors the SSE and LOCA events and the seismic profile, the effects of aging due to cyclic loading on the intended functions of the fuel transfer tube bellows are adequately managed by the Fatigue Monitoring Program.

2. SSE and LOCA are one-time, faulted events. Since one-time, faulted events, are not normal inputs to fatigue monitoring, the SSE and LOCA event is not in LRA Tables 4.3.1-1 through 4.3.1-6. However, the SSE and LOCA are events monitored by the current Fatigue Monitoring program. LRA Sections 4.6.5 and A.4.6.5 are revised as shown in Enclosure B.

RS-14-150 Enclosure A Page 24 of 25 RAI 4.6.6-1 Applicability:

Byron and Braidwood

Background:

LRA Section 4.6.6, "Recirculation Sump Guard Piping Bellows Fatigue," states that the bellows were analyzed for fatigue in accordance with Expansion Joint Manufacturers Association, 4th Edition, 1975, and substantiated per ASME Section Ill, Subparagraph NE-3365.2(e)(1), 1977 Edition through Summer of 1977 Addenda , where cyclic life was obtained through analyses and testing. The applicant dispositioned this TLAA in accordance with 10 CFR 54.21(c)(1)(iii),

indicating that the effects of aging on the intended function(s) will be adequately managed for the period of extended operation by LRA Section B.3.1.1, "Fatigue Monitoring," an enhanced AMP that the applicant claims to be consistent with the GALL Report AMP X.M1. The Fatigue Monitoring AMP states that it monitors and tracks critical thermal and pressure transients to ensure analyzed components do not exceed the number of their allowable cycles.

Issue:

LRA Section 4.6.6 states the these bellows are affected by plant heatup and cooldown transients and other transients associated with accident conditions that would fill the containment recirculation sump; however, it is not clear which of the transients listed in LRA Section 4.3.1 are considered to contribute towards the 7000 cycle limit for which the bellows were designed.

Request:

State the applicable transients, including the cycle limit for each transient that was assumed in the design fatigue analysis for the recirculation sump guard piping bellows, to demonstrate that the effects of aging on the intended functions will be adequately managed by the Fatigue Monitoring Program.

Exelon Response:

The bellows on the recirculation sump guard piping are not required to maintain containment integrity as described in LRA Section 2.4.4. A seal ring exists between the guard pipe and the recirculation sump piping which serves as the containment boundary. The guard pipe is comprised of a 28-inch diameter sleeve that penetrates through the Containment Structure and Auxiliary Building walls. The guard pipe also includes two (2) sets of expansion joints (bellows),

with one bellows sealing between the containment sump piping and the guard pipe located inside the Containment Structure; and the other bellows sealing between the guard pipe and the sump suction valve protection chamber inside the Auxiliary Building Structure. There is another set of bellows inside the Auxiliary Building which seal between the sump suction valve protection chamber and the recirculation sump effluent piping. The details of this configuration in LRA Sections 4.6.6 and A.4.6.6 are revised as provided in Enclosure B to clarify the description of the configuration.

LRA Section 4.6.6, "Recirculation Sump Guard Piping Bellows Fatigue," states that the bellows were analyzed for fatigue. The cycle limit and the transients, which were assumed in the design

RS-14-150 Enclosure A Page 25 of 25 fatigue analysis for the recirculation sump guard piping bellows, are different for the three sets of bellows. The cycle limits and applicable transients for the different sets of bellows located both inside containment and in the auxiliary building are described separately in the following sections.

Containment Structure The analysis for the recirculation sump guard piping bellows inside containment was performed in accordance with ASME Section Ill, Subparagraph NE-3365.2(e)(1), 1977 Edition through Summer of 1977 Addenda to determine the appropriate numbers of fatigue test cycles required to support the design requirement of 10 cycles. The applicable transient associated with the analysis performed for the recirculation sump guard piping bellows is only that associated with the LOCA event. The Current License Basis (CLB) LOCA event limit for this transient is one (1). Since one-time, faulted events are not normal inputs to fatigue monitoring, the LOCA event is not in LRA Tables 4.3.1-1 through 4.3.1-6. The Fatigue Monitoring program currently includes tracking and monitoring of LOCA events. Therefore, the Fatigue Monitoring program will adequately manage the aging on the intended functions of the bellows.

Auxiliary Building Structure Inside the auxiliary building on the sump suction valve protection chamber there are two bellows assemblies. One bellows assembly is between the sump suction valve protection chamber and the recirculation sump guard pipe, where the recirculation sump guard pipe extends past the containment boundary. This bellows provides expansion capability between the sump suction valve protection chamber in the Auxiliary Building and the recirculation sump guard pipe that is in the Containment Structure. The other bellows assembly is on the other side of the sump suction valve protection chamber providing expansion capability between the sump suction valve protection chamber and the process pipe. The 7000 cycles are inputs to the analysis to qualify the bellows and are the cycle limits. The 7000 cycles input to the analysis are similar to those evaluated in LRA Section 4.3.3, since the process pipe which is attached to the bellows is ASME Section III, Class 2. Both of these bellows assemblies in the Auxiliary Building do not perform a containment pressure boundary function. Similar to the disposition in LRA Section 4.3.3, for Class 2 fatigue analysis, the Fatigue Monitoring program will monitor the transients provided in Tables 4.3.1-3 (Byron) and 4.3.1-6 (Braidwood), which are the transients that have the potential to impart differential movement intended for the bellows assemblies. OBE and SSE seismic events are other transients associated with the cyclic differential movement associated with seismic events. Monitoring and tracking of OBE and SSE seismic events is currently performed by the Fatigue Monitoring program. Therefore, with the combination of the monitoring of the plant transients in LRA Tables 4.3.1-3 and 4.3.1-6, and the seismic transients, the Fatigue Monitoring program will adequately manage the aging on the intended functions of the bellows and maintain the total cycles below the 7000 cycle input limit which was used to qualify the bellows.

LRA Sections A.3.1.1 and B.3.1.1 are revised to provide confirmation of the monitoring of LOCA and SSE (seismic) events by the Fatigue Monitoring program (A.3.1.1) as shown in Enclosure B of this submittal.

RS-14-150 Enclosure B Page 1 of 12 Enclosure B Byron and Braidwood Stations, Units 1 and 2 License Renewal Application (LRA) updates resulting from the responses to the following RAIs:

RAI 3.5.2.2.2.2-1 RAI 4.3.9-1 RAI 4.6.5-1 RAI 4.6.6-1 Note: To facilitate understanding, original LRA text has been repeated in this Enclosure, with revisions indicated. Existing LRA text is shown in normal font. Changes are highlighted with bolded italics for inserted text and strikethroughs for deleted text.

RS-14-150 Enclosure B Page 2 of 12 As a result of the response to RAI 3.5.2.2.2.2-1 provided in Enclosure A of this letter, LRA Section 3.5.2.2.2.2, last paragraph on page 3.5-37, is revised as shown below. Changes are highlighted with bolded italics for inserted text and strikethroughs for deleted text.

Plant operating experience has not identified elevated general and local area temperature as a concern for concrete structural components. The high High energy line penetrations have been designed to limit surrounding concrete surfaces to temperatures less than 200°F, except for the special pipe whip restraints that are located around each feedwater and main steam pipe as it passes through the concrete wall separating the main steam isolation valve room from the main steam tunnel. The design documents for the concrete at these pipe whip restraints include an evaluation for elevated temperatures, which determined that the concrete temperature up to 300°F at the local areas around the pipes was acceptable. However, the actual plant Plant operating experience has shown that the identified elevated general and local area temperature as a concern for concrete temperatures around the special pipe whip restraints are less than 200oF structural components. Therefore, the aging effects of reduction of strength and modulus of concrete due to elevated temperatures are not expected to occur at the BBS Group 3 structures of BBS.

RS-14-150 Enclosure B Page 3 of 12 As a result of the response to RAI 3.5.2.2.2.2-1 provided in Enclosure A of this letter, LRA Table 3.5.1, Item Number 3.5.1-48, page 3.5-58, is revised as shown below. Changes are highlighted with strikethroughs for deleted text.

Table 3.5.1 Summary of Aging Management Evaluations for the Structures and Component Supports Further Item Aging Aging Management Component Evaluation Discussion Number Effect/Mechanism Programs Recommended 3.5.1-48 Groups 1-5: concrete: all Reduction of strength A plant-specific aging Yes, if temperature Not Applicable.

and modulus management program is limits are exceeded BBS Group 3-5 structures are not subject to due to elevated to be evaluated. general area temperatures greater than temperature (>150°F 150°F or local areas >200°F, except for the general; >200°F local) main steam tunnel and main steam isolation valve room walls which have been evaluated and found acceptable for temperatures up to 300F.

See Subsection 3.5.2.2.2.2.

RS-14-150 Enclosure B Page 4 of 12 As a result of the response to RAI 4.3.9-1, LRA Section 4.3.9, page 4.3-39, is revised as shown below. Revisions are indicated with bold italics for inserted text and strikethroughs for deleted text:

4.3.9 FATIGUE DESIGN OF SPENT FUEL POOL LINER AND SPENT FUEL STORAGE RACKS FOR SEISMIC EVENTS TLAA

Description:

The replacement spent fuel storage racks are designed to the stress limits of, and analyzed in accordance with ASME Section III, Division 1, Subsection NF. These analyses are described in UFSAR Section 9.1.2, "Spent Fuel Storage." The BBS spent fuel storage racks were replaced in 2000 and 2001 and both the spent fuel storage rack and the spent fuel pool liner were analyzed for fatigue due to seismic events using methods similar to those for ASME Section III, Subsection NB, Class 1 components.

Therefore, these design analyses are a TLAA requiring evaluation for the period of extended operation.

TLAA Evaluation:

The analyses include a fatigue evaluation of the replacement spent fuel storage racks and the spent fuel pool liner for the cyclic loads imposed by twenty (20) OBE events plus one (1) SSE event. The number of specific load cycles utilized in the fatigue evaluation of the replacement spent fuel storage racks and the spent fuel pool liner are 25 cycles for each OBE event, and 20 cycles for the single SSE event.

The analyses calculated a cumulative usage factor (CUF) of 0.950 for the spent fuel storage racks. The analyses also includes a fatigue evaluation of the spent fuel pool liner for the loads imposed by the new racks, and uses the same input of twenty (20)

OBE events plus one (1) SSE event. The analyses calculated a CUF of 0.00052 for the spent fuel pool liner. Both are less than the CUF allowable of 1.0.

OBE and SSE events are monitored by the Fatigue Monitoring (B.3.1.1) program. No OBE or SSE events have occurred to date. SSE events are more severe than OBE events, and since no OBE events have occurred, no SSE events have occurred to-date.

The Fatigue Monitoring (B.3.1.1) program will continue to be used to manage fatigue of these components through the period of extended operation by monitoring OBE and SSE events.

TLAA Disposition: 10 CFR 54.21(c)(1)(iii) - The Fatigue Monitoring (B.3.1.1) program will manage fatigue of these components through the period of extended operation.

RS-14-150 Enclosure B Page 5 of 12 As a result of the response to RAI 4.6.5-1, LRA Section 4.6.5, page 4.6-6, is revised as shown below. Revisions are indicated with bold italics for inserted text and strikethroughs for deleted text:

4.6.5 FUEL TRANSFER TUBE BELLOWS FATIGUE TLAA

Description:

The fuel transfer tubes connect the refueling cavity (inside the Containment Structure) to the fuel transfer canal (inside the Fuel Handling Building). The fuel transfer tubes pass through the Containment Structure wall and through the wall of the Fuel Handling Building. Guard pipe assemblies, which also function as penetration sleeves, were utilized for the fuel transfer tubes. There are three expansion bellows in the penetration sleeve around each fuel transfer tube. The guard pipes for the fuel transfer tubes are comprised of 24-inch diameter penetration sleeves that penetrate through the refueling cavity, Containment Structure, and the Fuel Handling Building walls. They also include three (3) sets of expansion joints (bellows). The design specification for the bellows includes design load cycles equal to 100 and the design is based on ASME Section III, Subsection NE, 1974 Edition through Summer 1974. The specification of 100 design load cycles along with the maximum displacements intended to envelope all postulated design basis conditions, including one (1) SSE and one (1) LOCA transient event, for fatigue consideration. The bellows were qualified in accordance with ASME Section III, Subparagraph NE-3365.2(e)(2). Therefore, the transfer tube bellows design load cycles have been identified as a TLAA requiring evaluation for the period of extended operation.

TLAA Evaluation:

These bellows are affected by seismic transients (one (1) SSE event) and one (1)

LOCA that would cause deflection of the bellows. These transients are listed in Section 4.3.1 and have 60-year projections that are less than the numbers of cycles which The SSE and LOCA event load cycles form the basis for the design requirement of 100 design load cycles which areand used for the qualification of the bellows. Therefore, the qualification of the bellows is acceptable for the period of extended operation. The Fatigue Monitoring (B.3.1.1) program monitors SSE and LOCA transient events.

TLAA Disposition: 10 CFR 54.21(c)(1)(iii) - The effects of aging on the intended function(s) of the Fuel Transfer Tube Bellows will be adequately managed for the period of extended operation by the Fatigue Monitoring (B.3.1.1) program, which monitors transient events to ensure the CLB limits are not exceeded through the period of extended operation.

RS-14-150 Enclosure B Page 6 of 12 As a result of the response to RAI 4.6.6-1, LRA Section 4.6.6, page 4.6-7, is revised as shown below. Revisions are indicated with bold italics for inserted text and strikethroughs for deleted text:

4.6.6 RECIRCULATION SUMP GUARD PIPING BELLOWS FATIGUE TLAA

Description:

The guard pipe for the containment recirculation sump effluent piping extends from the recirculation sump, inside the Containment Structure, to the sump suction valve protection chamber inside the Auxiliary Building. The guard pipe is comprised of a 28-inch diameter sleeve that penetrates through the Containment Structure and Auxiliary Building walls. The guard pipe also includes two (2) ) three (3) sets of expansion joints (bellows), with one bellows sealing between the containment sump piping and the guard pipe located inside the Containment Structure;. The other two sets of bellows are in the Auxiliary Building on either side of the recirculation sump suction valve chamber. and the other bellows One of these bellows seals sealing between the containment recirculation sump effluent piping end of the guard pipe, where it extends outside the containment boundary, and the recirculation sump suction valve protection chamber inside the Auxiliary Building. A third bellows is located between the recirculation sump suction valve protection chamber and the containment recirculation sump effluent piping. The bellows were analyzed for fatigue in accordance with Expansion Joint Manufacturers Association, 4th Edition, 1975 and substantiated per ASME Section III, Subparagraph NE-3365.2(e)(1) 1977 Edition through Summer of 1977 addenda. The required design cycles are 10 cycles for the bellows in the Containment Structure and 7000 cycles for the bellows in the Auxiliary Building. Therefore, the design analysis for these bellows has been identified as a TLAA requiring evaluation for the period of extended operation.

TLAA Evaluation:

These bellows are affected by plant heatup and cooldown transients and other transient associated with accident conditions that would fill the containment recirculation sump, including OBE transients. These transients are listed in Section 4.3.1 and have 60-year projections that are less than the numbers of cycles analyzed for the bellows. Therefore, the design analysis of the bellows is acceptable for the period of extended operation.

The BBS Fatigue Monitoring (B.3.1.1) program monitors plant heatup and cooldown transients, as well as upset, emergency, and faulted conditions, including OBE and SSE events.

TLAA Disposition: 10 CFR 54.21(c)(1)(iii) - The effects of aging on the intended function(s) of the recirculation sump guard pipe bellows will be adequately managed for the period of extended operation by the Fatigue Monitoring (B.3.1.1) program, which monitors transient cycles and events to ensure they do not exceed their CLB cycle limits, validating the assumptions used in the analysis.

RS-14-150 Enclosure B Page 7 of 12 As a result of the response to RAI 4.3.9-1, RAI 4.6.5-1, and RAI 4.6.6-1, LRA, Appendix A, Section A.3.1.1, pages A-45 and A-46 is revised as shown below. Revisions are indicated with bold italics for inserted text and strikethroughs for deleted text:

A.3.1.1 Fatigue Monitoring The Fatigue Monitoring aging management program is an existing preventive program that manages cumulative fatigue damage of the reactor pressure vessel (RPV) components, reactor coolant pressure boundary piping components, and other components. The Fatigue Monitoring aging management program manages fatigue of piping, piping elements, piping components, bolting, reactor vessels, reactor vessel internals, supports, and heat exchangers and other components.

The Fatigue Monitoring aging management program monitors and tracks critical thermal, and pressure, and seismic transients to ensure each analyzed component does not exceed the number of allowable cycles, thus ensuring that the cumulative usage factor (CUF) for each analyzed component does not exceed the design limit of 1.0 through the period of extended operation. The Fatigue Monitoring program also monitors and tracks other design basis events such as LOCAs. The number of allowable cycles is based on the design fatigue analyses transient inputs. The program requires comparison of the actual operational transient parameters to the applicable design transient definitions to assure the actual operational transients are bounded. If an allowable cycle limit is approached or the severity of an actual operational transient is not bounded by the applicable design transient definition, then this condition is entered into and addressed within the corrective action program to ensure that the design CUF limit is not exceeded.

The Fatigue Monitoring aging management program will be enhanced to:

1. Address the cumulative fatigue damage effects of the reactor coolant environment on component life by evaluating the impact of the reactor coolant environment on critical components for the plant identified in NUREG/CR-6260. Additional plant-specific component locations in the reactor coolant pressure boundary will be evaluated if they are more limiting than those considered in NUREG/CR-6260.

2. Monitor and track additional plant transients that are significant contributors to component fatigue usage.

3. Evaluate the effects of the reactor coolant system water environment on the reactor vessel internal components with existing fatigue CUF analyses to satisfy the evaluation requirements of ASME Code,Section III, Subsection NG-2160 and NG-3121.

These enhancements will be implemented prior to the period of extended operation.

RS-14-150 Enclosure B Page 8 of 12 As a result of the response to RAI 4.3.9-1, LRA, Appendix A, Section A.4.3.9, pages A-58 and A-59, is revised as shown below. Revisions are indicated with bold italics for inserted text and strikethroughs for deleted text:

A.4.3.9 Fatigue Design of Spent Fuel Pool Liner and Spent Fuel Storage Racks for Seismic Events The new spent fuel storage racks, which were replaced in 2000 and 2001, are designed in accordance with ASME Section III, Division 1, Subsection NF, and were analyzed for fatigue due to seismic events using methods similar to those for ASME Section III, Division 1, Subsection NB. The analyses include a fatigue evaluation of the replacement spent fuel storage racks and for the spent fuel pool liner for the cyclic loads imposed by twenty (20) OBE events plus one (1) SSE event. The analyses calculated a cumulative usage factor (CUF) of less than 1.0 for the spent fuel storage racks. The analyses also includes a fatigue evaluation of and the spent fuel pool liner for the loads imposed by the new racks., and uses the same input of twenty (20) OBE events plus one (1) SSE event. The analyses calculated a CUF of less than 1.0 for the spent fuel pool liner. Therefore, these analyses were identified as TLAAs that require evaluation for the period of extended operation.

OBE and SSE events are monitored by the Fatigue Monitoring (A.3.1.1) program. No OBE or SSE events have occurred to date. SSE events are more severe than OBE events, and since no OBE events have occurred, no SSE events have occurred to date. The Fatigue Monitoring (A.3.1.1) program will continue to be used to manage fatigue of these components through the period of extended operation by monitoring OBE and SSE events. The Fatigue Monitoring (A.3.1.1) program will manage fatigue of these components through the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).

RS-14-150 Enclosure B Page 9 of 12 As a result of the response to RAI 4.6.5-1, LRA Section A.4.6.5, page A-63, is revised as shown below. Revisions are indicated with bold italics for inserted text and strikethroughs for deleted text:

A.4.6.5 Fuel Transfer Tube Bellows Fatigue The fuel transfer tubes connect the refueling cavity (inside the Containment Structure) to the fuel transfer canal (inside the Fuel Handling Building). The fuel transfer tubes pass through the Containment Structure wall and through the exterior wall of the Fuel Handling Building. Guard pipe assemblies, which also function as penetration sleeves, are utilized for the fuel transfer tubes. There are three expansion bellows in each penetration sleeve, designed to the requirements of the 1974 Edition of ASME Section III, Subsection NE through the Summer 1974 Addenda. These design inputs for these bellows include 1 SSE and 1 LOCA transient and that envelope all postulated design basis conditions. The original design transients bound the corresponding 60-year projections for these transients. Therefore, the bellows fatigue analyses remain valid through the period of extended operation. The effects of aging on the intended function(s) of the fuel transfer tube bellows will be adequately managed for the period of extended operation by the Fatigue Monitoring (A.3.1.1) program, which monitors transient cycles to ensure the transient limits are not exceeded through the period of extended operation. Therefore, this TLAA will be managed in accordance with 10 CFR 54.21(c)(1)(iii).

RS-14-150 Enclosure B Page 10 of 12 As a result of the response to RAI 4.6.6-1, LRA Section A.4.6.6, page A-63, is revised as shown below. Revisions are indicated with bold italics for inserted text and strikethroughs for deleted text:

A.4.6.6 Recirculation Sump Guard Piping Bellows Fatigue The guard pipe for the containment recirculation sump effluent piping extends from the recirculation sump, inside the Containment Structure, to the sump suction valve chamber inside the Auxiliary Building. The guard pipe is comprised of a 28-inch diameter sleeve that penetrates through the Containment Structure and Auxiliary Building walls and two (2) three (3) sets of expansion joints (bellows). One set of bellows seals is in the Containment Structure between the containment sump piping and the guard pipe located inside the Containment Structure. The other two sets of bellows are in the Auxiliary Building on either side of the recirculation sump suction valve chamber. and another One of these bellows set of bellows seals between the containment recirculation sump effluent piping end of the guard pipe where it extends outside the containment boundary, and the recirculation sump suction valve chamber, inside the Auxiliary Building. A third bellows is located between the recirculation sump suction valve protection chamber and the containment recirculation sump effluent piping. These bellows were analyzed in accordance with ASME Section III, subsection NE for fatigue, based upon design inputs including RCS heatup and cooldown transients, OBE transients, and other upset, emergency, and faulted conditions that would fill the containment recirculation sump. The original design transients bound the 60-year projections for these transients through the period of extended operation. The effects of aging on the intended function(s) of the recirculation sump guard pipe bellows will be adequately managed for the period of extended operation by the Fatigue Monitoring (A.3.1.1) program, which monitors transient cycles to ensure they do not exceed their design limits, validating the assumptions used in this analysis. Therefore, this TLAA will be managed in accordance with 10 CFR 54.21(c)(1)(iii).

RS-14-150 Enclosure B Page 11 of 12 As a result of the response to RAI 4.3.9-1, RAI 4.6.5-1, and RAI 4.6.6-1, LRA Appendix B, Section B.3.1.1, page B-279, is revised as shown below. Revisions are indicated with bold italics for inserted text and strikethroughs for deleted text:

B.3.1.1 Fatigue Monitoring Program Description The Fatigue Monitoring aging management program is an existing preventive program that manages cumulative fatigue damage of the reactor pressure vessel (RPV) components, reactor coolant pressure boundary (RCPB) piping components, and other components subject to air-indoor uncontrolled, air with borated water, condensation, diesel exhaust, neutron flux, reactor coolant, treated water, treated borated water, and steam. The Fatigue Monitoring aging management program manages cumulative fatigue damage of piping components, piping elements, bolting, reactor vessels, reactor vessel internals, supports, and heat exchangers and other components. The program reviews the temperature, and pressure, and seismic profiles of the actual operational transients and counts them in the appropriate design transient category.

The Fatigue Monitoring aging management program monitors and tracks critical thermal, and pressure, and seismic transients to ensure each analyzed component does not exceed the number of allowable cycles, thus ensuring that the cumulative usage factor (CUF) for each analyzed component does not exceed the design limit of 1.0 through the period of extended operation. The Fatigue Monitoring aging management program also monitors and tracks other design basis events such as LOCAs. The number of allowable cycles is based on the design fatigue analyses transient inputs. The program requires comparison of the actual operational transient parameters to the applicable design transient definitions to assure the actual operational transients are bounded. If an allowable cycle limit is approached or the severity of an actual operational transient is not bounded by the applicable design transient definition, then this condition is entered into and addressed within the corrective action program to ensure that the design CUF limit is not exceeded. The fatigue cycle monitoring data was used to project the numbers of cycles that will occur during 60 years. These projections show that the current 40 year allowable cycle limits will not be exceeded in 60 years. Therefore, the current 40 year cycle limits will be maintained for the period of extended operation. The Fatigue Monitoring aging management program will be enhanced to monitor additional plant transients that are significant contributors to cumulative fatigue damage.

Maintaining the number of cumulative cycles below the analyzed allowable cycle limits assures that the fatigue analyses remains valid. If a cycle limit is approached or the severity of an actual operational transient is not bounded by the applicable design transient definition, the condition is entered into the corrective action program. Fatigue analyses exists for BBS reactor pressure vessels (RPV) components, and reactor coolant pressure boundary (RCPB) piping components, and other components (e.g.

pumps, heat exchangers, spent fuel pool racks, spent fuel pool liner) in accordance with ASME Section III, Class 1 fatigue design requirements per the current licensing basis. This includes the analyses provided in the original stress reports as well as subsequent analyses developed to evaluate design changes, power rerates, and operational events. These Class 1 fatigue analyses have been identified as Time-

RS-14-150 Enclosure B Page 12 of 12 Limited Aging Analyses (TLAAs) that are evaluated in Section 4.0 of the Byron and Braidwood License Renewal Application. In addition, components designed in accordance with ASME Section III, Class 2 and 3 and ANSI B31.1 requirements have been identified as having implicit fatigue Time-Limited Aging Analyses (TLAAs).

In addition, the program will be enhanced to evaluate the cumulative fatigue damage effect of the reactor coolant environment on reactor pressure vessel (RPV) components and reactor coolant pressure boundary (RCPB) piping components by performing environmentally assisted fatigue analyses for critical locations selected accordance with NUREG/CR-6260 guidance. Additional plant-specific component locations in the reactor coolant pressure boundary will be evaluated if they are more limiting than those considered in NUREG/CR-6260. Environmentally-adjusted cumulative usage factors (CUFen) were computed for each wetted material within the analyzed component or system to assure the limiting case was analyzed. The resulting 60-year CUFen values did not exceed the limit of 1.0. If a CUFen allowable cycle limit is approached, the condition will be entered into and addressed within the corrective action program.

The program will also be enhanced to evaluate the effects of the reactor coolant system water environment on the reactor vessel internal components with existing fatigue CUF analyses to satisfy the evaluation requirements of ASME Code,Section III, Subsection NG-2160 and NG-3121.