View source for NL-14-093, Indian Point, Units 2 & 3, Reply to Request for Additional Information Regarding License Renewal Application

{{Adams
| number = ML14225A777
| issue date = 08/05/2014
| title = Indian Point, Units 2 & 3, Reply to Request for Additional Information Regarding License Renewal Application
| author name = Dacimo F
| author affiliation = Entergy Nuclear Northeast
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000247, 05000286
| license number = DPR-026, DPR-064
| contact person = 
| case reference number = NL-14-093, TAC MD5407, TAC MD5408
| document type = Letter type:NL
| page count = 31
| project = TAC:MD5407, TAC:MD5408
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:S'EEntergy Enteray Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel (914) 254-2055 Fred Dacimno Vice President Operations License Renewal NL-14-093 August 05, 2014 U.S. Nuclear Regulatory Commission Document Control Desk 11545 Rockville Pike, TWFN-2 F1 Rockville, MD 20852-2738

==SUBJECT:==

==REFERENCES:==

Reply to Request for Additional Information Regarding the License Renewal Application Indian Point Nuclear Generating Unit Nos. 2 & 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64 1. Entergy letter (NL-14-067), "Reply to Request for Additional Information Regarding the License Renewal Application," dated June 9, 2014.2. NRC letter, "Request for Additional Information for the Review of the Indian Point Nuclear Generating Unit Nos. 2 and 3, License Renewal Application, SET 2014-02 (TAC Nos. MD5407 and MD5408)" dated April 9, 2014.3. NRC letter, "Summary of Conference Calls Held on June 19 and July 16, 2014, Between the U.S. Regulatory Commission and Entergy Nuclear Operations, Inc., Entergy's Response to a Request for Additional Information on Applicant/Licensee Action Item 7 from MRP-227-A," dated August 1, 2014.

==Dear Sir or Madam:==
Entergy Nuclear Operations, Inc. submitted in Reference 1 a response to NRC RAIs 11-C, 16-B and 17-A (Reference 2). Following that submittal Entergy and the NRC held two clarification calls concerning the response to RAI 11-C (Reference 3). As a result of those calls Entergy is submitting, in Attachment 1, a revised response to RAI 11-C. The responses to RAIs 16-B and 17-A remain unchanged.
The revised response to RAI 11-C includes the deletion of Commitment 
: 47. This commitment change is depicted in the list of regulatory commitments provided in Attachment 2.If you have any questions, or require additional information, please contact Mr. Robert Walpole at 914-254-6710.
Docket Nos. 50-247 & 50-286 NL-14-093 Page 2 of 2 I declare uder penalty of perjury that the foregoing is true and correct. Executed on$ ,2014.FRD/rw 

==Attachment:==

1.Reply to NRC Request for Additional Information Regarding the License Renewal Application
: 2. License Renewal Application IPEC List of Regulatory Commitments Revision 24 cc: Mr. William Dean, Regional Administrator, NRC Region I Mr. Sherwin E. Turk, NRC Office of General Counsel, Special Counsel Mr. Dave Wrona, NRC Branch Chief, Engineering Review Branch I Ms. Kimberly Green, NRC Sr. Project Manager, Division of License Renewal Mr. Douglas Pickett, NRR Senior Project Manager Ms. Bridget Frymire, New York State Department of Public Service NRC Resident Inspector's Office Mr. John B. Rhodes, President and CEO NYSERDA ATTACHMENT 1 TO NL-14-093 REPLY TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE LICENSE RENEWAL APPLICATION ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286 NL-14-093 Attachment 1 Page 1 of 6 REQUEST FOR ADDITIONAL INFORMATION, SET 2014-02 RELATED TO INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 LICENSE RENEWAL APPLICATION REACTOR VESSEL INTERNALS PROGRAM AND INSPECTION PLAN RAI 11-C Applicant/Licensee Action Item 7 from the staff's final safety evaluation (SE) of MRP-227 requires the applicants/licensees of Babcock & Wilcox (B&W), Combustion Engineering (CE), and Westinghouse reactors to develop plant-specific analyses to be applied for their facilities to demonstrate that B&W In-Core Monitoring Instrumentation (IMI) guide tube assembly spiders and control rod guide tube (CRGT) assembly spacer castings, CE lower support columns, and Westinghouse lower support column bodies will maintain their functionality during the period of extended operation, and states that these analyses should also consider the possible loss of fracture toughness in these components due to thermal embrittlement (TE) and irradiation embrittlement (IE). For Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3), the equivalent component to the lower support column bodies are the lower internals assembly -column caps (column caps).By letter dated January 28, 2014, Entergy provided plant specific information on the ferrite content and susceptibility to TE for the Indian Point Nuclear Generating Unit Nos. 2 and 3 lower internals assembly -column caps. Based on its evaluation of the plant-specific material information for the column caps, Entergy concluded that the IP2 and IP3 column caps are not susceptible to TE.Entergy's conclusion related to TE notwithstanding, the column caps remain susceptible to IE.The staff is concerned that the linked Primary component for the column caps, the CRGT assembly lower flange welds, is not a good predictor of IE for the column caps since the CRGT lower flange welds receive substantially lower neutron fluence than the column caps (based on the estimated neutron fluence tabulated in MRP-191 for the two components).
Irradiation assisted stress corrosion cracking (IASCC) is the only mechanism of cracking that screened in for the column caps. The CRGT lower flange welds are also not a good predictor for IASCC of the column caps, because the lower flange welds are susceptible to stress corrosion cracking (SCC) and fatigue cracking, but not IASCC.The staff, therefore, requests that Entergy modify its Reactor Vessel Internals Inspection Plan (RVI Inspection Plan) to provide a link to a Primary component or components that is an appropriate predictor of IE and IASCC of the column caps.Response to RAI 11-C The following text will be incorporated into the reactor vessel internals (RVI) Inspection Plan for Indian Point Units 2 and 3. This text regards a link to a component that is an appropriate predictor of irradiation embrittlement (IE) and irradiation-assisted stress corrosion cracking (IASCC) of the column caps (i.e., lower internals assembly -column caps).
NL-14-093 Attachment 1 Page 2 of 6"The cast austenitic stainless steel lower support column bodies (column caps) are listed in MRP-227-A 
[5, Table 4-6] as "Expansion" components, with the inspection triggered by the observation of cracking in the control rod guide tube (CRGT) lower flange weld, which is a"Primary" component.
The degradation mechanisms that were screened in for these components are identified in MRP-191 [2], a supporting report for MRP-227-A, and are summarized in Table 1, below. IE, which has no demonstrated non-destructive examination (NDE) technique, is identified in Table 1 below as a degradation mechanism for the lower support column bodies (column caps) and CRGT lower flange weld. The "Primary" to"Expansion" linkage in MRP-227-A was based on consideration of the total relative susceptibility of the components due to the combined application of all screened in degradation mechanisms.
Priority is given to the leading indicator and the inspection technique and timing focus on initial onset detectability of the effects of the degradation using proven NDE techniques.
The relevant degradation mechanisms that could drive cracking in the lower support column bodies (column caps) are IE and IASCC.The "Primary" inspection component for the cast section of the lower support column body (column cap) was originally the cast CRGT lower flange weld. Both of these components were identified as potentially subject to thermal embrittlement (TE) in MRP-191 [2]. However, plant-specific evaluations indicate that TE is not a concern for Indian Point Units 2 and 3 because the Hull's ferrite number is less than the industry screening criterion of 20%. The ferrite content is below 15% for Indian Point Unit 2 and below 12% for Indian Point Unit 3 [4]. Although the low ferrite content eliminates the TE concerns in the lower support column bodies (column caps) at Indian Point Units 2 and 3, the MRP-227-A link to the "Primary" CRGT lower flange weld inspection was maintained to be consistent with the NRC direction in the Safety Evaluation 
[5]for MRP-227-A.
An additional "Primary" inspection link for the lower support column bodies (column caps) is being added at both Indian Point units because the 60-year fluence in the top few inches of the lower support column body, which is where the column cap is located, is projected to exceed the threshold criteria for IE and IASCC of austenitic stainless steel. Therefore, regardless of the material type or ferrite content, IE and IASCC must be considered as damage mechanisms in the lower support column bodies (column caps).A review of the screening results from MRP-1 91 [2] indicates multiple near-core components that potentially experience neutron fluences and stresses over the IE and IASCC threshold/screening criterion, similar to the lower support column bodies (column caps). From this set of components, as explained below, the lower core barrel cylinder girth weld is an appropriate leading indicator of IE and IASCC in the lower support column bodies (column caps).MRP-227-A 
[5] requires an EVT-1 inspection of welds in both the upper and lower core barrel sections.
Although many of these core barrel welds do not exceed the neutron fluence threshold for either IE or IASCC, the lower core barrel cylinder girth weld has been identified in MRP-191 as being subject to both mechanisms.
Selection of the lower core barrel cylinder girth weld, which incurs the highest fluence, as a leading indicator of IE and IASCC in the lower support columns is based on consideration of material, temperature, fluence, and stress factors.
NL-14-093 Attachment 1 Page 3 of 6 1. Material:
The process of melting and solidification that occurs during the welding process produces a steel microstructure that is similar to an austenitic casting with low ferrite content.2. Temperature:
Reactor coolant at the core barrel exterior and the lower core support columns is at core inlet temperatures.
Gamma heating effects in these components are expected to be small. The analysis provided in MRP-230 [6] indicates relatively small temperature increases in the core barrel and lower core support structure.
: 3. Fluence: Examination of neutron fluence data from the Indian Point units indicates that there is a general correspondence between the peak neutron fluence in the core barrel [7]and the fluence range in the top four centimeters of the lower support columns [8]. The 60-year peak fast neutron fluence projected in the lower core barrel after 48 effective full power years of operation will be approximately 2.5x1 021 n/cm 2 (E > 1 MeV) = 4dpa. After the same period of operation, the peak fast neutron fluence at the top surface of the lower support columns near the core centerline will be approximately, 4.1 lx1021 n/cm 2 (E >1 MeV) = 6 dpa. However, the attenuation of fluence below the core is steep and the peak fast neutron fluence in the lower support columns at a distance of four centimeters below the lower core plate will be 2.07x1 021 n/cm 2 (E > 1MeV) = 3 dpa.4. Stress: Although it is difficult to quantify the stresses in the lower core barrel cylinder girth weld and the lower support column, it is reasonable to anticipate large surface stresses in the lower core barrel cylinder girth weld. Under typical welding procedures, large tensile stresses approaching the unirradiated yield stress can be generated at the weld surfaces as the metal cools. Additional tensile stresses may be generated on the outer surface of the core barrel due to temperature gradients caused by the attenuation of gamma heating through the barrel thickness.
Axial loads associated with the core support function of the barrel may also contribute to the tensile stress on the core barrel weld. In contrast, the core support loads on the lower support columns are primarily compressive and there are no anticipated residual stresses.
Thermal deflections caused by heating of the lower core plate may impose a surface stress due to bending of the lower support columns. However, these stresses are expected to be small compared to the stress in the lower core barrel cylinder girth weld.Given the similarities between material, temperature, and fluence in the lower support columns and the lower core barrel cylinder girth weld, stress is the leading factor that determines the relative susceptibility to degradation effects caused by IE and IASCC. Given the significantly higher stress, the lower core barrel cylinder girth weld is a leading indicator of IE and IASCC in the lower support columns.The MRP-227-A 
[5] requirements are augmented by including the MRP-227-A required EVT-1 inspection of the lower core barrel cylinder girth weld as an additional "Primary" link for an"Expansion" inspection of the Indian Point lower support column bodies (column caps) for managing IE and IASCC. Consistent with existing guidance in MRP-227-A, confirmation of a surface crack greater than two inches in length will be taken as an indication of active cracking in the lower core barrel cylinder girth weld. This confirmation will require expansion of the inspection to the Indian Point lower support column bodies (column caps). The link between the CRGT lower flange welds and the lower support column bodies (column caps) provided in MRP-227-A remains unchanged.
NL-14-093 Attachment 1 Page 4 of 6 The expansion inspection will be conducted within two refueling outages of the original observation of cracking in the lower core barrel cylinder girth weld. It is reasonable to expect a significant delay in the onset of cracking in the lower support column bodies (column caps)based on the difference in stress. Existing analysis of laboratory data [3] indicates that large stresses are required to initiate IASCC at fluences below 10 dpa. The lower support column bodies (column caps) are projected to be primarily in a compressive stress state; therefore, they are expected to have a higher threshold fluence for IE and IASCC than the core barrel.Completion of the "Expansion" inspection within two refueling outages (each fuel cycle is 24 months in duration) represents only 13% more than the approximately 32 effective full power years of service experienced prior to the inspection of the core barrel cylinder girth weld. This provides reasonable assurance that IE and IASCC degradation, if any, will be detected in a timely fashion." The intent of Commitment 47 was to demonstrate that the lower support column bodies (column caps) will maintain their functionality during the period of extended operation considering the possible loss of fracture toughness due to thermal and irradiation embrittlement.
As the staff noted in this RAI 11-C, Entergy has concluded that the IP2 and IP3 lower support column bodies (column caps) are not susceptible to TE and will modify the Reactor Vessel Internals (RVI)Inspection Plan to credit inspection of the lower core barrel cylinder girth weld as an appropriate predictor of IE and IASCC in the lower support column bodies (column caps). Therefore, the functionality of the lower support column bodies (column caps) is assured during the period of extended operation without the need for a functionality analysis.
Therefore, Entergy is deleting Commitment 47 (additions are underlined, deletions lined out).Commitment 47 IPE-C- Will pe~form and submit analyses that de.monsrtrate-that the lower support coILumn bodiesA PwRll Rmitain their fUnctigRality drlirg the porild of exteRded operation conRidering the possible los~s of fracture toughness due to ther~m.al and- irrad-iationA embrittle-mont.
The analyses will be consistent With the 1P2/!P3 icensing basis. Deleted.References
: 1. U.S. Nuclear Regulatory Commission, "Request for Additional Information for the Review of the Indian Point Nuclear Generating Unit Nos. 2 and 3, License Renewal Application, Set 2014-02 (TAC Nos. MD5407 and MD5408)," April 9, 2014. (ADAMS ML14094A173).
: 2. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191).
EPRI, Palo Alto, CA: 2006. 1013234 (ADAMS ML12335A503).
: 3. Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175).
EPRI, Palo Alto. CA: 2005. 1012081 (ADAMS ML12335A503).
: 4. Entergy Letter, NL-14-013, "Additional Information Regarding the License Renewal Application 
-Action Item 7 from MRP-227-A, Indian Point Nuclear Generating Unit Nos.2 & 3, Docket Nos. 50-247 and 50-286, License Nos. DPR-26 and DPR-64," January 28, 2014 (ADAMS ML14038A150).
NL-14-093 Attachment 1 Page 5 of 6 5. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A).
EPRI, Palo Alto, CA: 2011. 1022863 (ADAMS ML1 20170453).
: 6. Materials Reliability Program: Functionality Analysis for Westinghouse and Combustion Engineering Representative PWR Internals (MRP-230-Rev.
1).EPRI, Palo Alto, CA: 2009.1019091.7. Westinghouse Report, WCAP-17780-P, Rev. 0, "Reactor Internals Aging Management MRP-227-A Applicability for Combustion Engineering and Westinghouse Pressurized Water Reactor Designs," June 2013 (ADAMS ML13322A454). (Westinghouse Proprietary Class 2).8. Westinghouse Calculation Note, CN-REA-1 3-38, Rev. 1, "Lower Core Support Fluence and Gamma Heat Generation Rate Evaluation for Aging Management at Indian Point Units 2 &3," October 9, 2013. (Westinghouse Proprietary Class 2).
NL-14-093 Attachment 1 Page 6 of 6 Table 1 MRP-191 Screening Table for Westinghouse Reactor Internals 
[21 Assembly Subassembly Component Material None SCC IASCC Wear Fatigue TE IE VS ISRIIC Control Rod Guide Flanges -304 SS Weld Fat Upper Internals Tube Assemblies Lower Assembly and Flow Flanges -Downcomers Lower CF8 SCC Fat TE IE Core barrel Core Barrel Flange 3 S Lower Lower Core 304SS Weld IASCC IE Internals Barrel (Weld) I Assembly Lower Support Lower Support CF8 IASCC TE IE VS Column -Column Bodies Assemblies Lower Support 304 SS IASCC IE VS_ Notes:_ Column Bodies Notes: SCC = stress corrosion cracking VS = void swelling ISR/IC = irradiation-induced stress relaxation and irradiation creep ATTACHMENT 2 TO NL-14-093 LICENSE RENEWAL APPLICATION IPEC LIST OF REGULATORY COMMITMENTS Rev. 24 ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286 NL-14-093 Attachment 2 Page 1 of 21 List of Regulatory Commitments Rev. 24 The following table identifies those actions committed to by Entergy in this document.Changes are shown as strikethroughs for deletions and underlines for additions.
# COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION/ AUDIT ITEM 1 Enhance the Aboveground Steel Tanks Program for P2: NL-07-039 A.2.1.1 IP2 and IP3 to perform thickness measurements of Complete A.3.1.1 NL-13-122 B.1.1 the bottom surfaces of the condensate storage tanks, IP3: city water tank, and fire water tanks once during the December 12, first ten years of the period of extended operation.
2015 Enhance the Aboveground Steel Tanks Program for IP2 and IP3 to require trending of thickness measurements when material loss is detected.2 Enhance the Bolting Integrity Program for IP2 and IP3 IP2: NL-07-039 A.2.1.2 to clarify that actual yield strength is used in selecting Complete A.3.1.2 carifyB.1.2 materials for low susceptibility to SCC and clarify the 1P3: prohibition on use of lubricants containing MoS 2 for omplete NL-07-153 Audit Items bolting. 201,241, The Bolting Integrity Program manages loss of NL-1 3-122 270 preload and loss of material for all external bolting. I I NL-14-093 Attachment 2 Page 2 of 21 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I I / IAUDIT ITEM 3 Implement the Buried Piping and Tanks Inspection Program for IP2 and IP3 as described in LRA Section B.1.6.This new program will be implemented consistent with the corresponding program described in NUREG-1801 Section XI.M34, Buried Piping and Tanks Inspection.
Include in the Buried Piping and Tanks Inspection Program described in LRA Section B.1.6 a risk assessment of in-scope buried piping and tanks that includes consideration of the impacts of buried piping or tank leakage and of conditions affecting the risk for corrosion.
Classify pipe segments and tanks as having a high, medium or low impact of leakage based on the safety class, the hazard posed by fluid contained in the piping and the impact of leakage on reliable plant operation.
Determine corrosion risk through consideration of piping or tank material, soil resistivity, drainage, the presence of cathodic protection and the type of coating. Establish inspection priority and frequency for periodic inspections of the in-scope piping and tanks based on the results of the risk assessment.
Perform inspections using inspection techniques with demonstrated effectiveness.
P2:-omplete P3: December 12, 2015 NL-07-039 NL-13-122 NL-07-153 NL-09-106 NL-09-111 A.2.1.5 A.3.1.5 B.1.6 Audit Item 173 NL-1 1-101 NL-14-093 Attachment 2 Page 3 of 21 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I__ I_ I AUDIT ITEM 4 Enhance the Diesel Fuel Monitoring Program to include cleaning and inspection of the IP2 GT-1 gas turbine fuel oil storage tanks, IP2 and IP3 EDG fuel oil day tanks, IP2 SBO/Appendix R diesel generator fuel oil day tank, and IP3 Appendix R fuel oil storage tank and day tank once every ten years.Enhance the Diesel Fuel Monitoring Program to include quarterly sampling and analysis of the IP2 SBO/Appendix R diesel generator fuel oil day tank, IP2 security diesel fuel oil storage tank, IP2 security diesel fuel oil day tank, and IP3 Appendix R fuel oil storage tank. Particulates, water and sediment checks will be performed on the samples. Filterable solids acceptance criterion will be less than or equal to 10mg/l. Water and sediment acceptance criterion will be less than or equal to 0.05%.Enhance the Diesel Fuel Monitoring Program to include thickness measurement of the bottom of the following tanks once every ten years. IP2: EDG fuel oil storage tanks, EDG fuel oil day tanks, SBO/Appendix R diesel generator fuel oil day tank, GT-1 gas turbine fuel oil storage tanks, and diesel fire pump fuel oil storage tank; IP3: EDG fuel oil day tanks, EDG fuel oil storage tanks, Appendix R fuel oil storage tank, and diesel fire pump fuel oil storage tank.Enhance the Diesel Fuel Monitoring Program to change the analysis for water and particulates to a quarterly frequency for the following tanks. IP2: GT-1 gas turbine fuel oil storage tanks and diesel fire pump fuel oil storage tank; IP3: Appendix R fuel oil day tank and diesel fire pump fuel oil storage tank.Enhance the Diesel Fuel Monitoring Program to specify acceptance criteria for thickness measurements of the fuel oil storage tanks within the scope of the program.Enhance the Diesel Fuel Monitoring Program to direct samples be taken and include direction to remove water when detected.Revise applicable procedures to direct sampling of the onsite portable fuel oil contents prior to transferring the contents to the storage tanks.Enhance the Diesel Fuel Monitoring Program to direct the addition of chemicals including biocide when the presence of biological activity is confirmed.
IP2: Complete I P3: December 12, 2015 NL-07-039 NL-13-122 NL-07-153 NL-08-057 A.2.1.8 A.3.1.8 B.1.9 Audit items 128, 129, 132, 491,492, 510 NL-14-093 Attachment 2 Page 4 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM 5 Enhance the External Surfaces Monitoring Program lP2: NL-07-039 A.2.1.10 for IP2 and IP3 to include periodic inspections of Complete A.3.1.10 systems in scope and subject to aging management IP3:NL-13-122 B.1.11 review for license renewal in accordance with 10 CFR DP e1 December 12, 54.4(a)(1) and (a)(3). Inspections shall include areas surrounding the subject systems to identify hazards to those systems. Inspections of nearby systems that could impact the subject systems will include SSCs that are in scope and subject to aging management review for license renewal in accordance with 10 CFR 54.4(a)(2).
6 Enhance the Fatigue Monitoring Program for IP2 to IP2: NL-07-039 A.2.1.11 monitor steady state cycles and feedwater cycles or omplete A.3.1.11 NL-13-122 B.1.12, perform an evaluation to determine monitoring is not NL-03-122 Ad Item required.
Review the number of allowed events and 164 resolve discrepancies between reference documents and monitoring procedures.
IP3: Enhance the Fatigue Monitoring Program for IP3 to December 12, include all the transients identified.
Assure all fatigue 2015 analysis transients are included with the lowest limiting numbers. Update the number of design transients accumulated to date.
NL-14-093 Attachment 2 Page 5 of 21 COMMITMENT IMPLEMENTATION SOURCE I RELATED SCHEDULE LRA SECTION 4 I /I AUDIT ITEM 7 Enhance the Fire Protection Program to inspect external surfaces of the IP3 RCP oil collection systems for loss of material each refueling cycle.Enhance the Fire Protection Program to explicitly state that the IP2 and IP3 diesel fire pump engine sub-systems (including the fuel supply line) shall be observed while the pump is running. Acceptance criteria will be revised to verify that the diesel engine does not exhibit signs of degradation while running;such as fuel oil, lube oil, coolant, or exhaust gas leakage.Enhance the Fire Protection Program to specify that the IP2 and IP3 diesel fire pump engine carbon steel exhaust components are inspected for evidence of corrosion and cracking at least once each operating cycle.Enhance the Fire Protection Program for IP3 to visually inspect the cable spreading room, 480V switchgear room, and EDG room C02 fire suppression system for signs of degradation, such as corrosion and mechanical damage at least once every six months.P2: ,omplete NL-07-039 NL-13-122 A.2.1.12 A.3.1.12 B.1.13 P3: December 12, 2015 NL-14-093 Attachment 2 Page 6 of 21 COMMITMENT IMPLEMENTATION SOURCE I RELATED SCHEDULE LRA SECTION I I /I AUDIT ITEM 8 Enhance the Fire Water Program to include inspection of IP2 and IP3 hose reels for evidence of corrosion.
Acceptance criteria will be revised to verify no unacceptable signs of degradation.
Enhance the Fire Water Program to replace all or test a sample of IP2 and IP3 sprinkler heads required for 10 CFR 50.48 using guidance of NFPA 25 (2002 edition), Section 5.3.1.1.1 before the end of the 50-year sprinkler head service life and at 10-year intervals thereafter during the extended period of operation to ensure that signs of degradation, such as corrosion, are detected in a timely manner.Enhance the Fire Water Program to perform wall thickness evaluations of IP2 and IP3 fire protection piping on system components using non-intrusive techniques (e.g., volumetric testing) to identify evidence of loss of material due to corrosion.
These inspections will be performed before the end of the current operating term and at intervals thereafter during the period of extended operation.
Results of the initial evaluations will be used to determine the appropriate inspection interval to ensure aging effects are identified prior to loss of intended function.Enhance the Fire Water Program to inspect the internal surface of foam based fire suppression tanks.Acceptance criteria will be enhanced to verify no siqnificant corrosion.
I P2: Complete NL-07-039 NL-13-122 NL-07-153 NL-08-014 A.2.1.13 A.3.1.13 B.1.14 Audit Items 105, 106 IP3: December 12, 2015 NL-14-093 Attachment 2 Page 7 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION: I_ I I jI /AUDIT ITEM 9 Enhance the Flux Thimble Tube Inspection Program for IP2 and IP3 to implement comparisons to wear rates identified in WCAP-12866.
Include provisions to compare data to the previous performances and perform evaluations regarding change to test frequency and scope.Enhance the Flux Thimble Tube Inspection Program for IP2 and IP3 to specify the acceptance criteria as outlined in WCAP-12866 or other plant-specific values based on evaluation of previous test results.Enhance the Flux Thimble Tube Inspection Program for IP2 and IP3 to direct evaluation and performance of corrective actions based on tubes that exceed or are projected to exceed the acceptance criteria.
Also stipulate that flux thimble tubes that cannot be inspected over the tube length and cannot be shown by analysis to be satisfactory for continued service, must be removed from service to ensure the integrity of the reactor coolant system pressure boundary.P2: Zomplete NL-07-039 A.2.1.15 A.3.1.15 B.1.16 NL-13-122 IP3: December 12, 2015 NL-14-093 Attachment 2 Page 8 of 21 COMMITMENT IMPLEMENTATIONI SOURCE I RELATED SCHEDULE LRA SECTION I I /I AUDIT ITEM 10 Enhance the Heat Exchanger Monitoring Program for IP2: NL-07-039 A.2.1.16 IP2 and IP3 to include the following heat exchangers Complete A.3.1.16 NL-13-122 B.1.17, in the scope of the program. IP3: NL-07-153 Audit Item* Safety injection pump lube oil heat exchangers December 12, 52* RHR heat exchangers 2015* RHR pump seal coolers* Non-regenerative heat exchangers
* Charging pump seal water heat exchangers
* Charging pump fluid drive coolers* Charging pump crankcase oil coolers* Spent fuel pit heat exchangers
* Secondary system steam generator sample coolers* Waste gas compressor heat exchangers
* SBO/Appendix R diesel jacket water heat exchanger (IP2 only)Enhance the Heat Exchanger Monitoring Program for IP2 and IP3 to perform visual inspection on heat exchangers where non-destructive examination, such as eddy current inspection, is not possible due to heat exchanger design limitations.
Enhance the Heat Exchanger Monitoring Program for IP2 and IP3 to include consideration of material-environment combinations when determining sample population of heat exchangers.
Enhance the Heat Exchanger Monitoring Program for IP2 and IP3 to establish minimum tube wall thickness for the new heat exchangers identified in the scope of the program. Establish acceptance criteria for heat exchangers visually inspected to include no indication of tube erosion, vibration wear, corrosion, pitting, NL-09-018 fouling, or scaling.NL-09-056 11 Deleted NL-1 1-101 NL-14-093 Attachment 2 Page 9 of 21 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM 12 Enhance the Masonry Wall Program for IP2 and IP3 IP2: NL-07-039 A.2.1.18 to specify that the IP1 intake structure is included in Complete A.3.1.18 the program. IP3: NL-13-122 8.1.19 Complete 13 Enhance the Metal-Enclosed Bus Inspection Program IP2: NL-07-039 A.2.1.19 for IP2 and IP3 to visually inspect the external surface Complete A.3.1.19 NL-13-122 8.1.20 of MEB enclosure assemblies for loss of material at NL-13-122 Aui.tm least once every 10 years. The first inspection will occur prior to the period of extended operation and ecember 12, 124 the acceptance criterion will be no significant loss of 015 NL-08-057 133,519 material.
NL-1 3-077 Enhance the Metal-Enclosed Bus Inspection Program to add acceptance criteria for MEB internal visual inspections to include the absence of indications of dust accumulation on the bus bar, on the insulators, and in the duct, in addition to the absence of indications of moisture intrusion into the duct.Enhance the Metal-Enclosed Bus Inspection Program for IP2 and IP3 to inspect bolted connections at least once every five years if performed visually or at least once every ten years using quantitative measurements such as thermography or contact resistance measurements.
The first inspection will occur prior to the period of extended operation.
The plant will process a change to applicable site procedure to remove the reference to "re-torquing" connections for phase bus maintenance and bolted connection maintenance.
14 Implement the Non-EQ Bolted Cable Connections 1P2: NL-07-039 A.2.1.21 Program for IP2 and IP3 as described in LRA Section Complete A.3.1.21 NL-13-122 B.1.22 B.1.22. IP3: December 12, 2015 NL-14-093 Attachment 2 Page 10 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION/ AUDIT ITEM 15 Implement the Non-EQ Inaccessible Medium-Voltage lP2: NL-07-039 A.2.1.22 Cable Program for IP2 and IP3 as described in LRA Complete A.3.1.22 CeProg for.23. NL-13-122 B.1.23 Section B.1.23. IP3: NL-07-153 Audit item This new program will be implemented consistent with December 12, 173 the corresponding program described in NUREG- 2015 NL-11-032 1801 Section XI.E3, Inaccessible Medium-Voltage Cables Not Subject To 10 CFR 50.49 Environmental NL-1 1-096 Qualification Requirements.
NL-1 1-101 16 Implement the Non-EQ Instrumentation Circuits Test P2: NL-07-039 A.2.1.23 Review Program for IP2 and IP3 as described in LRA &#xfd;omplete A.3.1.23 Section B.1.24. NL-13-122 B.1.24 SP3: NL-07-153 Audit item This new program will be implemented consistent with December 12, 173 the corresponding program described in NUREG- 2015 1801 Section XI.E2, Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits.17 Implement the Non-EQ Insulated Cables and P2: NL-07-039 A.2.1.24 Connections Program for IP2 and IP3 as described in 3omplete A.3.1.24 LRA Section B.1.25. NL-13-122 B.1.25 LP3: NL-07-153 Audit item This new program will be implemented consistent with December 12, 173 the corresponding program described in NUREG- 2015 1801 Section XI.E1, Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements.
NL-14-093 Attachment 2 Page 11 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM 18 Enhance the Oil Analysis Program for IP2 to sample 1P2: NL-07-039 A.2.1.25 and analyze lubricating oil used in the SBO/Appendix Complete A.3.1.25 NL-13-122 B.1.26 R diesel generator consistent with the oil analysis for NL-11-101 other site diesel generators.
3ecember 12, Enhance the Oil Analysis Program for IP2 and IP3 to 2015 sample and analyze generator seal oil and turbine hydraulic control oil.Enhance the Oil Analysis Program for IP2 and IP3 to formalize preliminary oil screening for water and particulates and laboratory analyses including defined acceptance criteria for all components included in the scope of this program. The program will specify corrective actions in the event acceptance criteria are not met.Enhance the Oil Analysis Program for IP2 and IP3 to formalize trending of preliminary oil screening results as well as data provided from independent laboratories.
19 Implement the One-Time Inspection Program for IP2 P2: NL-07-039 A.2.1.26 and IP3 as described in LRA Section B.1.27. omplete A.3.1.26 NL-13-122 B.1.27 This new program will be implemented consistent with IP3: NL-07-153 Audit item the corresponding program described in NUREG- ecember 12, 173 1801, Section Xl.M32, One-Time Inspection.
015 20 Implement the One-Time Inspection 
-Small Bore P2: NL-07-039 A.2.1.27 Piping Program for IP2 and IP3 as described in LRA omplete A.3.1.27 Secio B1.8.NL-13-122 B.1.28 Section B.1.28. IP3: NL-07-153 Audit item This new program will be implemented consistent with December 12, 173 the corresponding program described in NUREG- 2015 1801, Section XI.M35, One-Time Inspection of ASME Code Class I Small-Bore Piping.21 Enhance the Periodic Surveillance and Preventive IP2: NL-07-039 A.2.1.28 Maintenance Program for iP2 and IP3 as necessary Complete A.3.1.28 to assure that the effects of aging will be managed IP3: such that applicable components will continue to perform their intended functions consistent with the ee r1 current licensing basis through the period of extended operation.
NL-14-093 Attachment 2 Page 12 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM 22 Enhance the Reactor Vessel Surveillance Program for IP2: NL-07-039 A.2.1.31 IP2 and IP3 revising the specimen capsule withdrawal Complete A.3.1.31 NL-13-122 B.1.32 schedules to draw and test a standby capsule to IP3: cover the peak reactor vessel fluence expected December 12, through the end of the period of extended operation.
2015 Enhance the Reactor Vessel Surveillance Program for IP2 and IP3 to require that tested and untested specimens from all capsules pulled from the reactor vessel are maintained in storage.23 Implement the Selective Leaching Program for IP2 P2: NL-07-039 A.2.1.32 and IP3 as described in LRA Section B.1.33. Complete A.3.1.32 NL-13-122 B.1.33 This new program will be implemented consistent with IP3: NL-07-153 Audit item the corresponding program described in NUREG- December 12, 173 1801, Section XI.M33 Selective Leaching of Materials.
015 24 Enhance the Steam Generator Integrity Program for IP2: NL-07-039 A.2.1.34 IP2 and IP3 to require that the results of the condition omplete A.3.1.34 monitoring assessment are compared to the P3:NL-13-122 B.1.35 operational assessment performed for the prior Complete operating cycle with differences evaluated.
25 Enhance the Structures Monitoring Program to JP2: NL-07-039 A.2.1.35 explicitly specify that the following structures are Complete A.3.1.35 included in the program. NL-13-122 B.1.36" Appendix R diesel generator foundation (IP3) IP3: NL-07-153* Appendix R diesel generator fuel oil tank vault December 12, Audit items (IP3) 2015 86, 87, 88,* Appendix R diesel generator switchgear and NL-08-057 417 enclosure (IP3)* city water storage tank foundation
* condensate storage tanks foundation (IP3) NL-13-077" containment access facility and annex (IP3)" discharge canal (IP2/3)" emergency lighting poles and foundations (IP2/3)* fire pumphouse (IP2)* fire protection pumphouse (IP3)* fire water storage tank foundations (IP2/3)* gas turbine 1 fuel storage tank foundation
* maintenance and outage building-elevated passageway (IP2)* new station security building (IP2)* nuclear service building (IP1)" primary water storage tank foundation (iP3).6 refueling water storage tank foundation (IP3)
NL-14-093 Attachment 2 Page 13 of 21 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION II I I I/ AUDIT ITEM" security access and office building (IP3)* service water pipe chase (IP2/3)* service water valve pit (IP3)" superheater stack* transformer/switchyard support structures (IP2)* waste holdup tank pits (IP2/3)Enhance the Structures Monitoring Program for IP2 and IP3 to clarify that in addition to structural steel and concrete, the following commodities (including their anchorages) are inspected for each structure as applicable.
* cable trays and supports* concrete portion of reactor vessel supports" conduits and supports* cranes, rails and girders" equipment pads and foundations
* fire proofing (pyrocrete)
* HVAC duct supports" jib cranes" manholes and duct banks* manways, hatches and hatch covers* monorails" new fuel storage racks* sumps NL-13-077 Enhance the Structures Monitoring Program for IP2 and IP3 to inspect inaccessible concrete areas that are exposed by excavation for any reason. IP2 and IP3 will also inspect inaccessible concrete areas in environments where observed conditions in accessible areas exposed to the same environment indicate that significant concrete degradation is occurring.
Enhance the Structures Monitoring Program for IP2 and IP3 to perform inspections of elastomers (seals, gaskets, seismic joint filler, and roof elastomers) to identify cracking and change in material properties and for inspection of aluminum vents and louvers to identify loss of material.Enhance the Structures Monitorina Proaram for IP2 Enhance the Structures Monitorina Proaram for 1P2 .- .- I I NL-14-093 Attachment 2 Page 14 of 21 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I_ I / AUDIT ITEM and IP3 to perform an engineering evaluation of groundwater samples to assess aggressiveness of groundwater to concrete on a periodic basis (at least once every five years). IPEC will obtain samples from at least 5 wells that are representative of the ground water surrounding below-grade site structures and perform an engineering evaluation of the results from those samples for sulfates, pH and chlorides.
Additionally, to assess potential indications of spent fuel pool leakage, IPEC will sample for tritium in groundwater wells in close proximity to the IP2 spent fuel pool at least once every 3 months.Enhance the Structures Monitoring Program for IP2 and IP3 to perform inspection of normally submerged concrete portions of the intake structures at least once every 5 years. Inspect the baffling/grating partition and support platform of the IP3 intake structure at least once every 5 years.Enhance the Structures Monitoring Program for IP2 and IP3 to perform inspection of the degraded areas of the water control structure once per 3 years rather than the normal frequency of once per 5 years during the PEO.Enhance the Structures Monitoring Program to include more detailed quantitative acceptance criteria for inspections of concrete structures in accordance with ACI 349.3R, "Evaluation of Existing Nuclear Safety-Related Concrete Structures" prior to the period of extended operation.
NL-08-127 NL-1 1-032 NL- 11-101 Audit Item 360 Audit Item 358 26 Implement the Thermal Aging Embrittlement of Cast I P2: NL-07-039 A.2.1.36 Austenitic Stainless Steel (CASS) Program for IP2 omplete A.3.1.36 NL-13-122 B.1.37 and IP3 as described in LRA Section B.1.37. 1P3: NL-07-153 Audit item This new program will be implemented consistent with December 12, 173 the corresponding program described in NUREG- 2015 1801, Section XI.M12, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) Program.
NL-14-093 Attachment 2 Page 15 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM 27 Implement the Thermal Aging and Neutron Irradiation IP2: NL-07-039 A.2.1.37 Embrittlement of Cast Austenitic Stainless Steel Complete A.3.1.37 NL-13-122 B.1.38 (CASS) Program for IP2 and IP3 as described in LRA IP3: NL-07-153 Audit item Section B.1.38.Complete 173 This new program will be implemented consistent with the corresponding program described in NUREG-1801 Section XI.M13, Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS) Program.28 Enhance the Water Chemistry Control -Closed IP2: NL-07-039 A.2.1.39 Cooling Water Program to maintain water chemistry of Complete A.3.1.39 NL-13-122 B.1.40 the IP2 SBO/Appendix R diesel generator cooling IP3: NL-08-057 Audit item system per EPRI guidelines.
Complete 509 Enhance the Water Chemistry Control -Closed Cooling Water Program to maintain the IP2 and IP3 security generator and fire protection diesel cooling water pH and glycol within limits specified by EPRI guidelines.
29 Enhance the Water Chemistry Control -Primary and IP2: NL-07-039 A.2.1.40 Secondary Program for IP2 to test sulfates monthly in Complete B.1.41 the RWST with a limit of <150 ppb. NL-13-122 IP2: NL-07-039 A.2.1.41 30 For aging management of the reactor vessel internals, IPEC will (1) participate in the industry programs for omplete A.3.1.41 NL-13-122 investigating and managing aging effects on reactor P3: internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor omplete internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.1P2: NL-07-039 A.2.2.1.2 31 Additional P-T curves will be submitted as required per 10 CFR 50, Appendix G prior to the period of 3omplete A.3.2.1.2 NL-13-122 

====4.2.3 extended====
operation as part of the Reactor Vessel P3: Surveillance Program. December 12, 2015 32 As required by 10 CFR 50.61(b)(4), IP3 will submit a IP3: NL-07-039 A.3.2.1.4 plant-specific safety analysis for plate B2803-3 to the December 12, 4.2.5 NRC three years prior to reaching the RTPTS 2015 NL-08-127 screening criterion.
Alternatively, the site may choose to implement the revised PTS rule when approved.
NL-14-093 Attachment 2 Page 16 of 21 COMMITMENT IMPLEMENTATION1 SOURCE RELATED SCHEDULE LRA SECTION II I / AUDIT ITEM 33 At least 2 years prior to entering the period of IP2: NL-07-039 A.2.2.2.3 extended operation, for the locations identified in LRA Complete A.3.2.2.3 NL-13-122 

====4.3.3 Table====
4.3-13 (IP2) and LRA Table 4.3-14 (IP3), under NL-13-122 Adtie the Fatigue Monitoring Program, IP2 and IP3 will p 146 implement one or more of the following:
Complete L-08-02 146 (1) Consistent with the Fatigue Monitoring Program, Detection of Aging Effects, update the fatigue usage calculations using refined fatigue analyses to determine valid CUFs less than 1.0 when accounting NL-1 0-082 for the effects of reactor water environment.
This includes applying the appropriate Fen factors to valid CUFs determined in accordance with one of the following:
: 1. For locations in LRA Table 4.3-13 (IP2) and LRA Table 4.3-14 (1P3), with existing fatigue analysis valid for the period of extended operation, use the existing CUF.2. Additional plant-specific locations with a valid CUF may be evaluated.
In particular, the pressurizer lower shell will be reviewed to ensure the surge nozzle remains the limiting component.
: 3. Representative CUF values from other plants, adjusted to or enveloping the IPEC plant specific external loads may be used if demonstrated applicable to IPEC.4. An analysis using an NRC-approved version of the ASME code or NRC-approved alternative (e.g., NRC-approved code case) may be performed to determine a valid CUF.(2) Consistent with the Fatigue Monitoring Program, Corrective Actions, repair or replace the affected locations before exceeding a CUF of 1.0.34 IP2 SBO / Appendix R diesel generator will be NL-13-122 2.1.1.3.5 installed and operational by April 30, 2008. This Complete NL-07-078 committed change to the facility meets the NL-08-074 requirements of 10 CFR 50.59(c)(1) and, therefore, a license amendment pursuant to 10 CFR 50.90 is not NL-1 1-101 required.
NL 10 NL-14-093 Attachment 2 Page 17 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION SIAUDIT ITEM 35 Perform a one-time inspection of representative IP2: NL-08-127 Audit Item sample area of IP2 containment liner affected by the Complete 27 1973 event behind the insulation, prior to entering the period of extended operation, to assure liner degradation is not occurring in this area. IP3: NL- 1-101 Perform a one-time inspection of representative December 12, sample area of the IP3 containment steel liner at the 2015 juncture with the concrete floor slab, prior to entering the period of extended operation, to assure liner degradation is not occurring in this area.Any degradation will be evaluated for updating of the NL-09-018 containment liner analyses as needed.36 Perform a one-time inspection and evaluation of a IP2: NL-08-127 Audit Item sample of potentially affected IP2 refueling cavity omplete NL-11-101 359 concrete prior to the period of extended operation.
The sample will be obtained by core boring the refueling cavity wall in an area that is susceptible to exposure to borated water leakage. The inspection will include an assessment of embedded reinforcing steel.Additional core bore samples will be taken, if the NL-09-056 leakage is not stopped, prior to the end of the first ten years of the period of extended operation.
A sample of leakage fluid will be analyzed to NL-09-079 determine the composition of the fluid. If additional core samples are taken prior to the end of the first ten years of the period of extended operation, a sample of leakage fluid will be analyzed.37.I. P2: NL-08-127 Audit Item 37 Enhance the Containment Inservice Inspection (Cl- , N0omplete IWL) Program to include inspections of the NL-13-122 containment using enhanced characterization of IP3: degradation (i.e., quantifying the dimensions of noted Cplt indications through the use of optical aids) during the omplete period of extended operation.
The enhancement includes obtaining critical dimensional data of degradation where possible through direct measurement or the use of scaling technologies for photographs, and the use of consistent vantage points for visual inspections.
NL-14-093 Attachment 2 Page 18 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM lP2: NL-08-143 4.2.1 38 For Reactor Vessel Fluence, should future core loading patterns invalidate the basis for the projected Complete NL-13-122 values of RTpts or CVUSE, updated calculations will lP3: be provided to the NRC. ecember 12, 2015 39 Deleted NL-09-079 40 Evaluate plant specific and appropriate industry 1P2: NL-09-106 B.1.6 operating experience and incorporate lessons learned -omplete B.1.22 NL-13-122 8.1.23 in establishing appropriate monitoring and inspection lP3: B.1,24 frequencies to assess aging effects for the new aging December 12, B.1.25 management programs.
Documentation of the 015 B.1.27 operating experience evaluated for each new program B.1.28 will be available on site for NRC review prior to the B.1.33 period of extended operation.
B.1.37 I _B.1.38 1P2: NL-11-032 N/A 41 IPEC will inspect steam generators for both units to fte the assess the condition of the divider plate assembly.
egn t The examination technique used will be capable of beginning of the detecting PWSCC in the steam generator divider plate e'Em and prior to assembly.
The IP2 steam generator divider plate S02 eNL1 1-074 inspections will be completed within the first ten years 023 of the period of extended operation (PEO). The IP3 P& NL-1 1-090 steam generator divider plate inspections will be completed within the first refueling outage following of the first NL-11-101 the beginning of the PEO. refueling outage ollowing the eginning of the PEQ.
NL-14-093 Attachment 2 Page 19 of 21 COMMITMENT IMPLEMENTATION SOURCE I RELATED SCHEDULE LRA SECTION I I_ I AUDIT ITEM 42 IPEC will develop a plan for each unit to address the potential for cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds using one of the following two options.Option 1 (Analysis)
IPEC will perform an analytical evaluation of the steam generator tube-to-tubesheet welds in order to establish a technical basis for either determining that the tubesheet cladding and welds are not susceptible to PWSCC, or redefining the pressure boundary in which the tube-to-tubesheet weld is no longer included and, therefore, is not required for reactor coolant pressure boundary function.
The redefinition of the reactor coolant pressure boundary must be approved by the NRC as a license amendment request.Option 2 (Inspection)
IPEC will perform a one-time inspection of a representative number of tube-to-tubesheet welds in each steam generator to determine if PWSCC cracking is present. If weld cracking is identified:
: a. The condition will be resolved through repair or engineering evaluation to justify continued service, as appropriate, and b. An ongoing monitoring program will be established to perform routine tube-to-tubesheet weld inspections for the remaining life of the steam generators.
NL-1 1-032 NL-1 1-074 NL-1 1-090 NL-1 1-096 N/A I P2: Prior to March 2024 IP3: Prior to the and of the first refueling outage following the beginning of the PEO.IP2: Between March 2020 and March 2024 1P3: Prior to the end of the first refueling outage following the beginning of the PEO.43 IPEC will review design basis ASME Code Class 1 fatigue evaluations to determine whether the NUREG/CR-6260 locations that have been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting locations for the IP2 and IP3 configurations.
If more limiting locations are identified, the most limiting location will be evaluated for the effects of the reactor coolant environment on fatigue usage.IPEC will use the NUREG/CR-6909 methodology in the evaluation of the limiting locations consisting of nickel alloy, if any.P2:.omplete P3: Prior to December 12, 0115 NL-1 1-032 NL-13-122 NL-1 1-101 4.3.3 NL-14-093 Attachment 2 Page 20 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM 44 IPEC will include written explanation and justification oP2: NL-11-032 N/A of any user intervention in future evaluations using the omplete NL-1 1-101 WESTEMS "Design CUF" module. lP3: Prior to NL-13-122 December 12, 2015 45 IPEC will not use the NB-3600 option of the oP2: NL-1 1-032 N/A WESTEMS program in future design calculations until omplete NL-1 1-101 the issues identified during the NRC review of the IP3: Prior to NL-13-122 program have been resolved.
December 12, 2015 46 Include in the IP2 ISI Program that IPEC will perform IP2: NL-1 1-032 N/A twenty-five volumetric weld metal inspections of Complete socket welds during each 10-year ISI interval NL-11-074 scheduled as specified by IWB-2412 of the ASME NL-13-122 Section Xl Code during the period of extended operation.
In lieu of volumetric examinations, destructive examinations may be performed, where one destructive examination may be substituted for two volumetric examinations.
47 IPEC" will peform and ,-,,it anay6ec that R21. NL 12 089 N/A demoecr.ate that the loWe.r SWP199F column bod8eFlF t9I ma.intain their fuctionality during the period of ,L 13 522 evtenlded operatn GQcnideFrig the pocibilIoiloss ot fracture toughness due to thenrmal anRd irradoWiAtiRon 
....... , 4 6 embrittlement.
The analyspcs will be consistent with the 12!ID3 l.cen;ing basic. Deleted. NL-14-093 NL-14-093 Attachment 2 Page 21 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM 48 Entergy will visually inspect IPEC underground piping oP2: NL-12-174 N/A within the scope of license renewal and subject to omplete aging management review prior to the period of IP3: Prior to NL-13-122 extended operation and then on a frequency of at ecember 12, least once every two years during the period of extended operation.
This inspection frequency will be maintained unless the piping is subsequently coated in accordance with the preventive actions specified in NUREG-1801 Section XI.M41 as modified by LR-ISG-2011-03. Visual inspections will be supplemented with surface or volumetric non-destructive testing if indications of significant loss of material are observed.
Consistent with revised NUREG-1 801 Section XI.M41, such adverse indications will be entered into the plant corrective action program for evaluation of extent of condition and for determination of appropriate corrective actions (e.g., increased inspection frequency, repair, replacement).
49 Recalculate each of the limiting CUFs provided in P2: NL-13-052 A.2.2.2 section 4.3 of the LRA for the reactor vessel internals Complete A.3.2.2 to include the reactor coolant environment effects IP3: Prior to NL-13-122 (Fen) as provided in the IPEC Fatigue Monitoring December 12, Program using NUREG/CR-5704 or NUREG/CR-emr1 6909. In accordance with the corrective actions specified in the Fatigue Monitoring Program, corrective actions include further CUF re-analysis, and/or repair or replacement of the affected components prior to the CUFen reaching 1.0.50 Replace the IP2 split pins during the 2016 IP2: NL-13-122 A.2.1.41 refueling outage (2R22). Prior to B. 1.42 completion of NL-14-067 2R22 IP3: N/A}}