ML14206A837

Response to Request for Additional Information Regarding Aging Management Program Description: Inservice Inspection - Reactor Vessel Internals, License Renewal Commitment #13 (MF3402)

ML14206A837
Person / Time
Site:	Millstone
Issue date:	07/21/2014
From:	Sartain M D Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
14-286, TAC MF3402
Download: ML14206A837 (16)

Text

Dominion Dominion Nuclear Connecticut, Inc.5000 Dominion Boulevard, Glen Allen, VA 23060 Web Address: www.dom.com U. S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555 July 21, 2014 Serial No.NSSL/MLC Docket No.License No.14-286 RO 50-336 DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING AGING MANAGEMENT PROGRAM DESCRIPTION:

INSERVICE INSPECTION

-REACTOR VESSEL INTERNALS, LICENSE RENEWAL COMMITMENT

1. 13 (MF3402)By letter dated July 31, 2013, Dominion Nuclear Connecticut, Inc. (DNC) submitted the"Aging Management Program

Description:

Inservice Inspection

-Reactor Vessel Internals" to address License Renewal Commitment

1. 13 for Millstone Power Station Unit 2 (MPS2). The submittal contains an updated Reactor Vessel Internals (RVI)Aging Management Program and RVI Inspection Plan in accordance with topical report,"Material Reliability Program: Pressurized Water Reactor Inspection and Evaluation Guidelines" (MRP-227-A).

In an e-mail dated May 14, 2014, the Nuclear Regulatory Commission transmitted a request for additional information (RAI) related to the submittal.

Attachment 1 to this letter contains DNC's response to the RAI.If you have any questions or require additional information, please contact Wanda Craft at (804) 273-4687.Sincerely, Mark D. Sartain Vice President

-Nuclear Engineering commonweaft of Vlkginil Reg. # 140542 My Commission Expires May 31, 2018 COMMONWEALTH OF VIRGINIA)))COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President

-Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.Acknowledged before me this -l day of ",,i 4 V ,2014.My Commission Expires: n -__ I -'Notary Public 1o~J Serial No.14-286 Docket No. 50-336 Page 2 of 2 Commitments made in this letter: None

Attachment:

1 .Response to Request for Additional Information Regarding License Renewal Commitment

1. 13 cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd Suite 100 King of Prussia, PA 19406-2713 Mohan C. Thadani Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 B1 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Serial No 14-286 Docket No. 50-336 Attachment 1 Response to Request for Additional Information Regarding License Renewal Commitment

1. 13 Dominion Nuclear Connecticut, Inc.Millstone Power Station Unit 2 Serial No 14-286 Docket No. 50-336 Attachment 1, Page 1 of 13 Response to Request for Additional Information Regarding License Renewal Commitment

1. 13 By letter dated July 31, 2013, Dominion Nuclear Connecticut, Inc. (DNC) submitted the"Aging Management Program

Description:

Inservice Inspection

-Reactor Vessel Internals" to address License Renewal Commitment

1. 13 for Millstone Power Station Unit 2 (MPS2).The submittal contains an updated Reactor Vessel Internals (RVI) Aging Management Program (AMP) and RVI Inspection Plan in accordance with topical report "Material Reliability Program: Pressurized Water Reactor Inspection and Evaluation Guidelines" (MRP-227-A).

In an e-mail dated May 14, 2014, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) related to the submittal.

The response to this RAI is as follows: RAI I As discussed in References 1 and 2, the staff has identified two additional questions that all CE and Westinghouse design plants referencing topical report "Material Reliability Program: Pressurized Water Reactor Inspection and Evaluation Guidelines" (MRP-227-A) must answer to close Applicant/Licensee Action Item (A/LAI) I related to plant-specific applicability of the topical report. If the answer to either or both questions is yes, then further evaluation will be necessary to demonstrate the applicability of MRP-227-A to Millstone Power Station, Unit 2 (MPS2). The staff therefore requests the following information:

1. Do the MPS2 RVI have non-weld or bolting austenitic stainless steel components with 20 percent cold work or greater, and if so do the affected components have operating stresses greater than 30 kilopounds per square inch?2. Has MPS2 ever utilized atypical design or fuel management that could make the assumptions of MRP-227-A regarding core loading/core design non-representative for that plant, including power changes/uprates?

DNC Response 1. DNC is working with Westinghouse to provide this information.

Based on vendor resource availability, DNC plans to respond to this question by December 19, 2014.2. MPS2 has not used atypical core design or fuel management.

Average core power density is based on licensed thermal power and the essentially fixed core geometry.For MPS2, the average core power density has been verified to be 83 Watts/cm 3 for the last six fuel cycles, and is less than the maximum value of 110 Watts/cm 3 recommended in the MRP-227-A applicability guidelines issued by Electric Power Research Institute (EPRI) (NRC Reference 2). For the peripheral fuel assembly Serial No 14-286 Docket No. 50-336 Attachment 1, Page 2 of 13 power density, the calculated Figure of Merit (F) for the last six fuel cycles has been F < 58 Watts/cm 3 , which is less than the maximum value of F = 68 Watts/cm 3 recommended in the EPRI applicability guidelines.

The minimum height of the active fuel to the fuel alignment plate in normal operation considering past and current fuel designs is 13.16 inches, which is greater than the 12.4 inch minimum value recommended in the EPRI applicability guidelines.

These core design and management parameters are expected to remain representative for future plant operations within the currently licensed thermal power limit.RAI 2 In Attachment 2 to the submittal, in its response to A/LAI 7, the licensee described its plant-specific evaluation of cast austenitic stainless steel (CASS) Reactor Vessel Internals (RVI). The licensee's evaluation used a screening approach using the criteria of U.S.Nuclear Regulatory Commission (NRC) Letter, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components," (Ref. 3).The result of the screening was that 63 of 68 core support columns were determined to be non-susceptible (screened out) for thermal embrittlement (TE). For 5 of 68 columns, the certified material test reports could not be located so the licensee conservatively assumed these 5 columns are susceptible to TE.Recently, the staff has developed interim guidance (Ref. 4, 5) for screening of CASS materials that are susceptible to both TE and irradiation embrittlement (IE). Under the interim guidance, low-molybdenum CASS materials (such as Type CF8) that receive neutron fluences greater than 0.45 displacements per atom (dpa) (3x1020 n/cm 2) are considered to be susceptible to TE and IE if the ferrite content is greater than 15 percent, while if the ferrite content is less than or equal to 15%, these materials are only susceptible to IE at neutron fluences greater than 1.5 dpa (lx1021 n/cm 2). The staff notes that all 63 MPS2 core support columns with CMTRs would screen out for TE based on the staff's interim guidance.The columns screen in for IE based on the peak neutron fluence to the columns. The staff's understanding is that although the core support column welds are required to be inspected as a "Primary" category component by MRP-227-A, the weld inspection is conducted from above the core support plate, therefore no portion of the core support columns would be viewed during the visual inspection of the core support column welds.1. Clarify the scope of the core support column weld inspection.

Specifically, is any portion of the columns below the core support plate viewed during the visual examination of the core support column welds? If not, can any information regarding the integrity of the columns be gained from the results of the core support column weld visual examination?

2. If the core support columns (other than the welds) are not inspected, Serial No 14-286 Docket No. 50-336 Attachment 1, Page 3 of 13 a. Provide a functionality evaluation of the lower support structure considering the effects of IE, and IE plus TE for those columns for which TE could not be screened out; or b. Modify the RVI inspection program to include inspections of the core support columns as a "Primary" or "Expansion" component.

Provide the schedule, scope examination method and acceptance criteria for these inspections.

Appropriate Primary link(s) to components with higher susceptibility to TE and IE would need to be identified if the columns are added as Expansion components.

DNC Response 1. Due to the small diameter of most of the flow holes (less than 2%-inch except for one 3-inch flow hole per quadrant) and relative thick plate (approximately 2-inch), the portions of the core support columns below the core support plate are not accessible for inspection in conjunction with the inspection of the core support column welds.During core support column weld visual examination, loss of integrity or material degradation would be detectable via evidence of changes to the surface at the weld location.

Cracking (stress corrosion cracking (SCC), irradiation assisted stress corrosion cracking (IASCC), and fatigue including damaged material) or structural distortion of the embrittled material at the weld would appear as a disruption of the normally smooth machined surface of the lower core support plate.2. The question suggests that DNC conservatively assumed the five support columns to be susceptible.

To clarify, the response to A/LAI 7 stated that the five core support columns were "potentially susceptible" to TE based on the range of element concentrations permitted by the CASS material specification.

The NRC statement represents the worst case possibility if no further information were available regarding typical CASS composition.

However, there is sufficient information available to conclude with reasonable assurance that the remaining five support columns were produced with a composition that would also result in ferrite contents within acceptable screening limits. A statistical analysis is presented to support this conclusion:

First, as noted in Table 1, note 3 of the July 31, 2013 submittal, the maximum ferrite content of the available 63 column compositions is 9.84%1. The ferrite contents of the known 63 columns are based on 25 separate material heats. The average of the ferrite contents for the 25 heats is 5.53%, the minimum is 3.43%, and the standard deviation is 1.43%. Thus, the maximum calculated ferrite content of 9.84% is three standard deviations above the average. At five standard deviations above the average, the ferrite content would be 12.68%. For a normal distribution, Ferrite contents are weight percents based on Hull's equivalency using material element composition taken from available certified material test reports (CMTR)

Serial No 14-286 Docket No. 50-336 Attachment 1, Page 4 of 13 only about six out of ten million measurements would exceed five standard deviations above the mean. On this basis, DNC concludes that the CASS support columns were consistently produced with acceptably low ferrite content, and it is statistically unlikely that any of the five support columns with unavailable composition data would have a ferrite content greater than 12.68%, which is well below the statically cast CF8 screening criteria of 20%2. Hence, DNC concludes with reasonable assurance that all 68 CASS core support columns meet the screening criteria and therefore, are not susceptible to TE.Since the core support columns screen out for TE, the CASS material in the upper portions of the columns may be considered equivalent to wrought austenitic material with respect to IE. Each arm of the core support columns has a peg that is inserted into and welded to the core support plate. This location exposes the CASS material to a bounding level of irradiation such that the Primary inspection of this weld area required by MRP-227-A constitutes adequate management for potential IE of the core support columns.RAI 3 For the components with material differences identified in the A/LAI 2 evaluation, identify the component, material type at MPS2, and the generic material from MRP-227-A.

Describe how the difference in material was evaluated and explain why no changes to the aging management requirements were needed for those components.

DNC Response There were only two material differences identified in the A/LAI 2 evaluation.

They are addressed below." The in-core instrumentation (ICI) guide tubes are listed in "Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design" (MRP-191) (DNC Reference

1) as Type 316 stainless steel, while for MPS2 they are fabricated from Type 304 stainless steel. The degradation mechanisms of concern listed for the ICI guide tubes are SCC and IE. Types 304 and 316 materials fall under the same austenitic stainless steel category and the two materials have equivalent degradation susceptibility.

Therefore, the same degradation mechanisms listed in MRP-191 are applicable to both materials and no change to aging management requirements for the ICI guide tubes is required." The control element assembly (CEA) shroud material is listed in MRP-191 as either Type 304 stainless steel, or cast austenitic grade CPF8 or CF8 stainless steel.Although not listed separately in MRP-191, the shaft retention pin and retention block are sub-part items of the CEA shroud, and are Type 304L stainless steel 2 Screening value for statically cast grade CF8 CASS as justified in DNC Reference 2

Serial No 14-286 Docket No. 50-336 Attachment 1, Page 5 of 13 material for MPS2. The MRP-191 degradation mechanism of concern for the Type 304 stainless steel CEA shroud material is SCC. Types 304 and 304L are in the same austenitic stainless steel category and have equivalent degradation susceptibility for SCC. Therefore, no additional degradation mechanisms are applicable for the Type 304L components and no change to aging management is required for the CEA shroud.RAI 4 Provide the MRP-191 equivalent component name for the following subcomponents that are listed in Table 3.1.2-2 of the MPS2 LRA: " Expansion Compensating Ring" Fuel Alignment Plate Guide Lugs and Guide Lug Inserts DNC Response" The MRP-1 91 equivalent component name for the Expansion Compensating Ring is the "hold-down ring" listed in Table 4-5, page 4-11 of MRP-1 91 (DNC Reference 1)." MRP-191 lists the Fuel Alignment Plate "Guide Lugs" and "Guide Lug Inserts" in Table 4-5, page 4-13.RAI 5 The in-core instrumentation (ICI) flux thimble tubes are not included in MPS2 License Renewal Application (LRA) Table 3.1.2-2, "Reactor Vessel, Internals, and Reactor Coolant System -Reactor Vessel Internals

-Aging Management Evaluation." This component is also not included in the Generic Aging Lessons Learned Report, Rev. 0. However, License Renewal Interim Staff Guidance (LR-ISG-2011-04): "Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors, (Ref. 6)," Table IVB3, "Reactor Vessel, Internals, and Reactor Coolant System -Reactor Vessel Internals

-Combustion Engineering," does have line item IV.B3.RP-357 for ICI: /CI Thimble Tubes -Lower. However, the applicable aging effect/mechanism is loss of material due to wear, not change in dimensions.

Were the ICI Flux Thimble Tubes identified as a component subject to aging management in the LRA for MPS2? If so, what aging effects and mechanisms were determined to require aging management and which program(s) were credited for managing aging of the/C/ flux thimble tubes? If not identified in the LRA as subject to AMR, what is the basis for including the /C/ flux thimbles as a component required to be addressed under MRP-227-A A/LA/ 3?

Serial No 14-286 Docket No. 50-336 Attachment 1, Page 6 of 13 DNC Response The 1C1 flux thimble tubes were not identified as a component subject to aging management in the license renewal application (LRA) for MPS2. Since the 101 design for MPS2 does not involve a pressure boundary function for the thimble tubes, the license renewal application did not identify any aging effects requiring management.

The flux thimble tubes are discussed in the submitted aging management program because MRP-227-A considers the flux thimble tubes to be subject to an Existing program and A/LAI 3 requires that the evaluation of any future aging management program requirements be described.

The requirements of the applicable Existing program for MPS2 addressed dimensional changes of the flux thimble tubes which were completed upon their replacement in 2009.By design, the replacement flux thimble tubes have sufficient margin for future dimensional changes and do not require further monitoring.

It is noted that loss of material due to wear of the 1C1 thimble tubes is listed as line item IV.B3.RP-357 in LR-ISG-2011-04, and that the applicable aging management program is Chapter XI.M16A in Revision 2 of the Generic Aging Lessons Learned report. However, the MPS2 design does not include the interface/features which caused significant flux thimble wear at the fuel alignment plate in a different Combustion Engineering design.Therefore, there is no Existing program at MPS2 to manage wear of the flux thimble tubes.The limited operating experience related to wear of replacement flux thimble tubes noted in the submitted AMP is for information only. Since the incore monitoring system has multiple degrees of redundancy and the flux thimble tubes only guide the insertion of the instrumentation, the wear discussed in the operating experience does not affect the functioning of the incore monitoring system. After the instrumentation is inserted, functioning of the system is not affected by flux thimble tube wear. The functional capability of the incore instrumentation system is monitored and maintained in accordance with the requirements specified in Section 3.3.2 of the MPS2 Technical Requirements Manual.RAI 6 The response to A/LAI 5 indicates for separation between the upper and lower core shroud sections, a maximum gap of 1/8 inch is acceptable at the innermost comers and a maximum gap of 1/16 inch is acceptable at the corners furthest from the innermost corners. The response also states the structural and functional effects associated with the presence of these gaps have been evaluated (Westinghouse letter is referenced), and are acceptable.

The staff requests the following additional information related to the evaluation that determined the acceptable gap size: 1. How are the acceptance criteria consistent with the licensing basis of MPS2?

Serial No 14-286 Docket No. 50-336 Attachment 1, Page 7 of 13 2. Other than distortion, what structural effects are expected to occur (for example, increased stresses), and how were these determined to be acceptable?

3. How is the function of the core shroud affected, if at all, by the maximum allowable swelling?4. How were the effects on the core shroud functions determined to be acceptable and what is the source of the functionality acceptance criteria for the core shroud?DNC Response 1. Gaps between the upper and lower core shroud (CS) subassemblies may result from (postulated) irradiation-induced void swelling.

The acceptance criteria for the visual examinations of the CS consist of maximum allowable values for these gaps.The potential adverse structural effects of these maximum gaps were identified and were determined to satisfy the stress criteria defined in the MPS2 final safety analysis report (FSAR) (see item 2 below). The potential adverse effects of these maximum gaps on the functions of the CS were identified and were determined to be within the limits prescribed in the MPS2 FSAR (see items 3 and 4 below). Thus, the acceptance criteria for the visual examinations of the CS are consistent with the MPS2 licensing basis.2. Potential adverse structural effects of the maximum CS gaps are identified and evaluated below: Identification of Potential Adverse Structural Effects of CS Gaps a. structural effect on the interfacing horizontal plates of the upper and lower CS subassemblies

b. structural effect on the tie rods joining the upper and lower CS subassemblies Evaluation of Potential Adverse Structural Effects of CS Gaps a. Stresses due to irradiation-induced void swelling have not been calculated for any of the CE plants. Based on the stress classifications defined in Section III, Subsection NG of the ASME B&PV Code (prescribed in the MPS2 FSAR), void swelling stresses are classified as secondary stresses since they are self-limiting.

In a structural evaluation, the main impact of secondary stresses is on the fatigue analysis.

The inclusion of void swelling stresses would increase the maximum primary plus secondary stress range. However, because void swelling stresses are not cyclical, they would not contribute to an increase in the cumulative usage factor. Therefore, it is reasonable to conclude that void swelling stresses need not be included in the structural evaluation of the CS, and thus would not adversely impact the satisfaction of the MPS2 licensing basis acceptance criteria.

Serial No 14-286 Docket No. 50-336 Attachment 1, Page 8 of 13 b. Irradiation-induced void swelling would increase the thicknesses of the interfacing horizontal plates of the upper and lower CS subassemblies.

This increase in plate thickness would produce an increase in the overall height of the CS assembly, and thus would increase loads in the tie rods which join the upper and lower CS subassemblies.

However, the increase in plate thickness is very small compared with the overall height of the CS assembly, and the resulting increase in tie rod loading is negligible.

Stresses in the tie rods would continue to satisfy stress limits defined in the MPS2 FSAR.3. As described in the MPS2 FSAR, the CS provides an envelope for the core and limits the amount of coolant bypass flow. The potential adverse effects of the maximum CS gaps on these functions are identified:

a. increase in CS-to-core support barrel (CSB) bypass coolant flow, b. inward deflection of CS plates encroaching on fuel space.4. The potential adverse effects of the maximum CS gaps on the functions of the CS, identified in item 3 above, are evaluated below: a. Coolant flow jetting through the gaps between the interfacing horizontal plates of the upper and lower CS subassemblies would increase the CS-to-CSB bypass coolant flow. This increase in CS-to-CSB bypass flow was conservatively estimated and added to the existing total bypass flow. The resulting increased total bypass flow was less than the allowable value prescribed in the MPS2 FSAR. Therefore, the effect of the CS gaps on core bypass flow was determined to be acceptable and bounded with regard to the MPS2 licensing basis.b. The maximum gaps between the interfacing plates of the CS upper and lower subassemblies result from the vertical deflection of one plate relative to the other. There is also a horizontal (inward) component to this deflection.

The maximum inward deflection is less than the dimensional tolerance on the lateral position of the CS plates relative to the core centerline.

In addition, this inward deflection, which would occur at core mid-height (approximately), could be accommodated by the lateral flexibility of the fuel assemblies.

Therefore, the inward deflection of the CS due to irradiation-induced void swelling would not have a significant adverse effect on the fuel assemblies.

RAI 7 With respect to the MPS2 plant-specific item "core support barrel assembly -crack stop holes at areas of prior fatigue cracking near thermal shield support bracket assemblies," Table 4-2 of the RVI Program Description lists cracking due to fatigue as the only applicable aging effect/mechanism.

The staff requests the licensee provide the following additional information:

Serial No 14-286 Docket No. 50-336 Attachment 1, Page 9 of 13 1. Describe how the applicable aging mechanisms and effects for the period of extended operation were determined.

Provide a justification for the process used if different than the process described in MRP-22 7-A, Section 2.2. Specifically, provide details of the evaluation that determined that fatigue cracking was the only aging effect requiring management for these locations.

Explain why the other aging effects/mechanisms generically evaluated in MRP-227-A are not applicable to these locations.

DNC Response 1 .DNC used the guidance of MRP-191 to determine the applicable aging mechanisms and effects. The repair areas, including crack stop holes at the two former locations of thermal shield support brackets, was considered a small scale feature of the core support barrel lower cylinder that was encompassed by the MRP-191 review of the larger structure and included other surface discontinuities, such as the remaining seven abandoned thermal shield support brackets.

The applicable aging mechanisms for the repair area were considered the same as those applicable to the core support barrel lower cylinder.

MRP-191 Table 5-2 lists the screened in aging mechanisms as SCC of welds, IASCC and IE. Since the welded support brackets at the two locations were removed, the SCC aging mechanism no longer screens in for the repair area.As noted in the AMP submittal, the repair was developed by Combustion Engineering and was within the scope of the applicability of MRP-227-A.

DNC has confirmed with Westinghouse that the repair area was known and evaluated as part of the core support barrel during the original MRP-191 screening and failure modes, effects, and criticality analysis (FMECA) evaluations that were performed during the development of MRP-227-A.

Therefore, the inclusion of the repair area in the Millstone AMP is an enhancement exceeding the requirements of MRP-227-A.

2. The remaining aging mechanisms evaluated for applicability but which did not screen in for the core support barrel lower cylinder (and by inclusion the crack stop holes)are (1) wear, (2) TE, (3) void swelling (VS), and (4) irradiation stress relief/irradiation creep (ISR/IC).

Wear is not applicable to this area because there are no mating parts in contact with each other. Thermal embrittlement is not applicable because the component is not cast austenitic stainless steel. Void swelling is not applicable for the core support barrel lower cylinder because, as shown in Table 4-7 of MRP-191, the core support barrel lower cylinder is considered a T-cold component.

Irradiation stress relief may occur to a slight degree at this location; however, the component function is not dependent upon maintaining a controlled preload. Therefore, the remaining aging mechanisms are not applicable to the repaired area of the core support barrel lower cylinder and do not create a need for additional aging management.

Serial No 14-286 Docket No. 50-336 Attachment 1, Page 10 of 13 RAI 8 For two of the "Primary" inspection category components applicable to MPS2, MRP-227-A permits a demonstration of fatigue life versus a time-limited aging analysis (TLAA) instead of inspection.

These components are the Core Support Barrel Assembly -Lower Flange Weld and the Lower Support Structure

-Core Support Plate. For each of these components, Note 6 to Table 2 of the RVI Program Description applies, which states this item originally screened in for fatigue by MRP-227-A, but as permitted by MRP-227-A for this item a plant specific fatigue evaluation has been performed and demonstrated acceptable fatigue life. Therefore, the listed enhanced visual examination (EVT-1) is not required and this item is subject to the normal American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section X1, Examination Category B-N-3 inspection requirements.

Details of these plant-specific fatigue evaluations were not provided in the RVI Program Description.

Further, these analyses appear to be new TLAAs since the analyses are not identified as such in the MPS2 LRA. Additionally, Title 10 of the Code of Federal Regulations (10 CFR) 54.37, "Additional records and recordkeeping requirements," (b)states that: After the renewed license is issued, the Final Safety Analysis Report (FSAR)update required by 10 CFR 50.71(e) must include any systems, structures, and components newly identified that would have been subject to an aging management review or evaluation of time-limited aging analyses in accordance with§ 54.21. This FSAR update must describe how the effects of aging will be managed such that the intended function(s) in § 54.4(b) will be effectively maintained during the period of extended operation.

The staff therefore requests: 1. Provide the plant-specific fatigue evaluations for the RVI components for which fatigue evaluations are being credited in lieu of inspections.

2. Discuss the need to update the MPS2 FSAR to reflect the fatigue analyses for the three components.

DNC Response 1. The vendor did not provide a plant-specific evaluation for this MRP-227-A activity for MPS2. The evaluation for MPS2 was developed based on comparisons and scaling of analyses performed by the vendor for another plant. The methodology of the evaluation is described briefly in the following:

Serial No 14-286 Docket No. 50-336 Attachment 1, Page 11 of 13 1. Identify a reference plant that is similar to MPS2 in terms of RVI design and operational parameters.

2. Review the RVI design drawings to confirm that there are no significant differences between MPS2 and the reference plant.3. Review the analysis for MPS2 to obtain design loads applied to the core support plate (CSP) and the core support barrel/lower support structure (CSB/LSS) flexure weld. Similarly, review the analysis for the reference plant to obtain the same information for that plant.4. Confirm that the design loads applied to the MPS2 CSP and CSB/LSS flexure weld are either bounded by, or very similar to, the equivalent loads for the reference plant.5. Review the RVI design transients to confirm that there are no significant differences between MPS2 and the reference plant with regard to both the types of transient events and the numbers of lifetime occurrences in the original design bases.6. Identify those design transients that have a significant impact on fatigue usage in RVI components.

7. Use the stress calculation methodology defined in the analysis for the reference plant to calculate stresses in the MPS2 CSP and CSB/LSS flexure weld.8. Use the fatigue analysis methodology defined in the analysis for the reference plant (consistent with ASME Code practice for that era) to calculate fatigue cumulative usage factors for the MPS2 CSP and CSB/LSS flexure weld.Consider significant design transients with numbers of cycles from the original design bases, and also the projected 60-year cycle counts for the corresponding transients based on MPS2 operating history.The results of the evaluation identified that the analyzed components had a cumulative usage factor that did not exceed the screening value allowed by MRP-191. Therefore the MPS2 CSP and CSB/LSS flexure welds screened out for fatigue cracking concerns under MRP-227-A.

However, existing ASME Section Xl program requirements remain applicable.

2. The MPS2 CSP and CSB/LSS flexure weld are not newly identified components for aging management review, but were considered in the development of the license renewal application.

The fatigue evaluation described above was performed as a permitted option in the implementation of MRP-227-A in lieu of inspection.

Because MRP-191 conservatively screened in certain components based on a lack of plant specific fatigue susceptibility information and operational history, MRP-227-A permitted the option to re-evaluate the fatigue screening of those components on a plant-specific basis. Since the evaluation is not required to demonstrate compliance with a MPS2 licensing basis requirement, no FSAR description of the evaluation is required to be included in the FSAR.

Serial No 14-286 Docket No. 50-336 Attachment 1, Page 12 of 13 RAI 9 Explain or correct the following inconsistencies between the MPS2 "Primary" and"Expansion" inspection category components identified in Table 2 and 3 of the RVI Program Description, and Tables 4-2 and 4-5 of MRP-227-A:

1. Table 3 lists Core Shroud Assembly (Welded) -Remaining Axial Welds with applicability to plant designs with core shrouds assembled in two vertical sections.MRP-227-A Table 4-5 has Core Shroud Assembly (welded) -Remaining axial welds, ribs and rings, but this component is only applicable to plant designs with core shrouds assembled with full-height shroud plates.2. Core Support Barrel Assembly -Core Barrel Assembly Axial Welds,"Expansion" component in Table 4-5 of MRP-227-A, is not included in Table 3 of the RVI Program Description.

DNC Response 1. The MPS2 AMP submittal Table 2 entry for the expansion examination of the Remaining Axial Welds is correct for welded core shrouds assembled in two vertical sections, which is the applicable design of MPS2. The published version of MRP-227-A Table 4-5 had an error that inadvertently omitted the correct entry for designs like MPS2. EPRI has been notified of the publishing error.2. The MPS2 AMP submittal Table 3 entry for the expansion examination of the Core Support Barrel Assembly Axial Welds was inadvertently omitted from the submitted program for MPS2 but was included in internal plant documents.

It is reproduced below. This entry follows the last entry in Table 3 of Enclosure 1, page 25 of 44.Core Support All plants Cracking Core barrel Enhanced visual (EVT-1) 100% of one side of the Barrel (SCC) assembly girth examination, with initial and accessible weld and Assembly welds subsequent examinations adjacent base metal Core barrel dependent on the results of surfaces for the weld assembly axial core barrel assembly girth with the highest welds weld examinations, calculated operating stress.See Figure 4-15.

Serial No 14-286 Docket No. 50-336 Attachment 1, Page 13 of 13 NRC References

1. 2/25/2013 Summary of Telecon with EPRI and Westinghouse Electric Company, March 15, 2013 (ADAMS Accession No. ML13067A262)

2. MRP-227-A Applicability Guidelines for Combustion Engineering and Westinghouse Pressurized Water Reactor Designs, Enclosure to MRP Letter 2013-025, October 14, 2013 3. "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components," May 19, 2000 (NRC ADAMS Accession No. ML003717179)

4. Summary Tables for CASS Position, March 12, 2014 (ADAMS Accession No.ML14072A012)

5. Email re: Summary Tables for CASS Position, from Joseph Holonich (NRC) to Kyle Amberge (EPRI) dated March 12, 2014 (ADAMS Accession No. ML14071A411)

6. LR Interim Staff Guidance LR-ISG-2011-04:

Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors.

May 28, 2013 (ADAMS Accession No. ML12270A251)

DNC References

1. 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. 1013234.2. "Project No. 704 -BWRVIP Response to NRC Request for Additional Information on BWRVIP-234", BWRVIP Letter 2014-086 dated May 23, 1014 from A. McGehee and D Madison to Joseph Holonich (ADAMS Accession No. ML14174A841).