2CAN091701, Response to Request for Additional Information Reactor Vessel Internals Aging Management Program Plan

Response to Request for Additional Information Reactor Vessel Internals Aging Management Program Plan

ML17251A758
Person / Time
Site:	Arkansas Nuclear
Issue date:	09/06/2017
From:	Pyle S Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2CAN091701, CAC MF8155
Download: ML17251A758 (13)

Text

~Entergy Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4704 Stephenie L. Pyle Manager, Regulatory Assurance Arkansas Nuclear One 2CAN091701 September 6, 2017 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Response to Request for Additional Information Reactor Vessel Internals Aging Management Program Plan Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6

REFERENCES:

1. Entergy letter to NRC, Reactor Vessel Internals Aging Management Program Plan, dated July 18, 2016 (2CAN071603) (ML16202A168)

2. NRC email to Entergy, dated May 9, 2017, Request for Additional Information Regarding Reactor Vessel Internals Aging Management Plan Review (CAC No. MF8155) (2CNA051701) (ML17123A297)

Dear Sir or Madam:

Entergy Operations, Inc. (Entergy) submitted the Arkansas Nuclear One, Unit 2 (AN0-2)

Reactor Vessel Internals (RVI) Aging Management Program (AMP) Plan (Reference 1) to fulfill a commitment made as part of the AN0-2 License Renewal Application. The plan identified the reactor vessel internals components that must be included for aging management review and identified the augmented inspection plan for those components.

During review of Reference 1, the NRC identified the need for additional information to continue the review. The NRC issued a Request for Additional Information (RAI) via Reference 2. This submittal contains the RAI response.

During the development of the responses, it was noted that a line item in Table 5.2 of Reference 1 had been omitted. Table 5-2 is essentially a duplicate of Table 4-5 from MRP-227-A that contains the expansion component items for Combustion Engineering-designed plants. When MRP-227-A was issued, the row in the table that identifies this specific expansion component item was inadvertently omitted. This omission was subsequently identified in MRP Letter 2012-034, but was missed in the development of the AN0-2 RVI AMP plan. .

2CAN091701 Page 2 of 2 The subject missing line item (remaining axial welds in the core shroud assembly) is an expansion RVI component. Examination of this item is contingent on the extent of aging degradation observed in the associated lead primary RVI component (core shroud plate-former plate weld). This inspection has not been performed as of this time.

The affected page from Table 5-2 of the AN0-2 RVI AMP plan is provided in Enclosure 2 of this submittal.

This submittal contains no new regulatory commitments.

If you have any questions or require additional information, please contact Stephenie Pyle at 4 79-858-4 704.

Sincerely, SLP/rwc Enclosures

1. Response to Request for Additional Information - AN0-2 RVI AMP

2. Revised Page of Table 5-2 of RVI AMP cc: Mr. Kriss Kennedy Regional Administrator U. S. Nuclear Regulatory Commission RGN-IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P. o. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Thomas Wengert MS 0-08B1 One White Flint North 11555 Rockville Pike Rockville, MD 20852

Enclosure 1 to 2CAN091701 Response to Request for Additional Information AN0-2 RVI AMP to 2CAN091701 Page 1of8 Response to Request for Additional Information AN0-2 RVI AMP By letter dated July 18, 2016, Entergy Operations, Inc. (Entergy) submitted the Reactor Vessel Internals (RVI) Aging Management Program (AMP) for Arkansas Nuclear One, Unit 2 (AN0-2).

The RVI AMP was submitted to fulfill License Renewal Commitment Nos. 15 and 19, which are described in the AN0-2 Safety Analysis Report (SAR), Sections 18.1.23 and 18.1.24, respectively.

In the course of review, the NRC staff determined that additional information was needed (reference NRC letter dated May 9, 2017, ML17123A297). The questions included in the NRC's request for information (RAI), along with the respective Entergy responses, are included below.

RAl-1 Action Item 1 - Applicability of Failure Modes. Effects. and Critical Analysis and Functionality Analysis Assumptions

Background

Section 4.2.1, "Applicability of FMECA [Failure Modes, Effects, and Critical Analysis] and Functionality Analysis Assumptions," of the NRC's safety evaluation (SE) (ADAMS Accession No. ML11308A770) for Electric Power Research Institute (EPRI) Topical Report, Materials Reliability Program (MRP)-227-A (ADAMS Accession No. ML120170453), "Material Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines,"

describes plant-specific Applicant/Licensee Action Item 1.

Section 5.1 "SE Section 4.2.1, Applicant/Licensee Action Item 1 (Applicability of FMECA and Functionality Analysis Assumptions)," and Section 1.8.4.1, "MRP-227-A Applicability to AN0-2,"

of the Attachment to the AN0-2 RVI AMP submittal, "PWR [Pressurized-Water Reactor Internals Aging Management Program Plan for Arkansas Nuclear One, Unit 2" (ADAMS Accession No. ML16202A167), discuss the applicability to MRP-227-A and the licensee's assessment of its compliance with Action Item 1.

In order to resolve the generic issue of the information needed from licensees to resolve Action Item 1, Westinghouse Commercial Atomic Power (WCAP) report WCAP-17780-P, "Reactor Internals Aging Management MRP-227-A Applicability for Combustion Engineering and Westinghouse Pressurized Water Reactor Designs" (ADAMS Accession No. ML13183A373; not publicly available, proprietary information), and EPRI MRP Letter 2013-025, "MRP-227-A Applicability Template Guideline" (ADAMS Accession No. ML13322A454), were developed. An NRC staff evaluation (ADAMS Accession No. ML14309A484) assessed MRP Letter 2013-025 and the technical basis contained in WCAP-17780-P, and concluded that if a licensee demonstrates that its plant complies with the guidance in MRP Letter 2013-025, there is reasonable assurance that the Inspection and Evaluation (l&E) guidance of MRP-227-A will be applicable to the specific plant.

to 2CAN091701 Page 2 of 8 MRP Letter 2013-025 provides two plant-specific questions to be addressed by licensees to l demonstrate compliance with MRP-227-A, as it relates to Action Item 1. In its application, the licen~ee did not provide the AN0-2 plant-specific information to answer Question 2 in MRP Letter 2013-025.

Request Provide the plant-specific response to Question 2 in MRP Letter 2013-025:

Question 2: [Has AN0-2 ever utilized] atypical fuel design or fuel management that could render the assumptions of MRP-227-A, regarding core loading/core design, non-representative for that plant [including power changes/uprates]?

If so, describe how the differences were reconciled with the assumptions of MRP-227-A or provide a plant-specific aging management program for affected components, as appropriate.

Entergy Response The AN0-2 core designs and fuel management have been compared to the assumptions of MRP-227-A and the screening criteria of MRP 2013-025. In this comparison it has been determined that the AN0-2 has not utilized atypical fuel design or fuel management that could render the assumptions of MRP-227-A, regarding core loading I core design, non-representative.

MRP-227-A fluence estimates assumed a 60-year plant operation, with 30 years of high leakage cores followed by 30 years of low leakage cores. The operating history of AN0-2 is approximately 22 years of high leakage cores, followed by what is currently 16+ years of low leakage core operation. This provides significant margin to the assumed fluence in MRP-227-A.

AN0-2 meets all MRP-227-A assumptions except for the average core power density and distance. This requires further screening per MRP 2013-025.

AN0-2 meets all screening criteria in MRP 2013-025, as summarized below.

MRP 2013-025 Radial Boundary Limitation Plant-specific applicability of MRP-227-A in the radial direction with no further evaluation required is demonstrated by meeting the following limits for CE-designed plants such as AN0-2:

Heat generation Figure of Merit (FoM), F S 68 Wattslcm 3 - Cycle 24 is considered low leakage, and representative of current and planned future core designs at AN0-2. Cycle 24 indicates a maximum FoM value of 66.32 W/cm 3 .

Average core power density< 110 Watts/cm 3 - The current rated power level at AN0-2 results in a core power density of 103.6 W/cm 3 . It should be noted that AN0-2 has previously implemented a 7.5% power uprate.

to 2CAN091701 Page 3 of 8 MRP 2013-025 Upper Axial Boundary Limitation Plant-specific applicability of MRP-227-A in the upper axial direction with no further evaluation required is demonstrated by meeting the fbllowing limits for CE-designed plants such as AN0-2:

Active fuel to fuel alignment plate distance> 12.4 inches. - The distance from the active fuel to the bottom of the fuel alignment plate at AN0-2 is currently greater than 19 inches and has always maintained a distance of greater than 18 inches.

Average core power density< 110 Watts/cm 3 - The current rated power level at AN0-2 results in a core power density of 103.6 W/cm 3

• This is after a 7.5% power uprate in 2002.

MRP 2013-025 Lower Axial Boundary Limitation Plant-specific applicability of MRP-227-A in the lower axial direction with no further evaluation required is demonstrated by meeting the MRP-227-A, Section 2.4 criteria. The assumptions from MRP-227-A, Section 2.4 are as follows:

30 years of operation with high leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low leakage fuel management strategy for the remaining 30 years of operation. - AN0-2 had been considered operating with a low leakage fuel management strategy beginning with Cycle 6 in which several previously burned assemblies were placed on the core periphery and most of the fresh assemblies were placed throughout the interior of the core. However, this operating history characterization is inconsistent with a finding that the first AN0-2 low leakage core did not occur until sometime between Cycle 11 and Cycle 15 (based on the low leakage FoM radial boundary limitation of MRP 2013-025). This finding supports a determination that AN0-2 Cycle 15 and later cycles meet the low leakage FoM radial boundary limitation of MRP 2013-025. Based on the low leakage FoM radial boundary limitation of MRP 2013-025, the operating history of AN0-2 is approximately 22 years of high leakage cores, followed by what is currently 16+ years of low leakage core operation. This provides significant margin to the assumed fluence in MRP-227-A.

Base load operation, i.e., typically operates at fixed power levels and does not usually vary power on a calendar or load demand schedule. - AN0-2 has not operated in a load follow capacity nor are there plans to operate to a load demand schedule in the future.

No design changes beyond those identified in general industry guidance or recommended by the original vendors. - AN0-2 has not implemented any fuel design changes outside of those recommended by the fuel vendor or in general industry guidance.

In conclusion, there have been no significant changes to AN0-2 in the areas of atypical core designs or fuel management that would invalidate the assumptions or results of MRP-227-A.

to 2CAN091701 Page 4 of 8 RAl-2 Action Item 2 - PWR Vessel Internal Components within the Scope of License Renewal

Background

Table 4-5, "Components and Materials for [Combustion Engineering] CE-Designed Plants," of MRP-191 (ADAMS Accession No. ML091910130), "Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191)," provides a generic list of RVI components and the materials of construction for CE-designed plants.

The materials listed below have been used in the fabrication of PWR RVI components. The materials listed below are susceptible to some of the aging degradation mechanisms that are identified in MRP-227-A.

• Nickel base alloys - lnconel 600 and weld metals -Alloy 82 and 182 and Alloy X:-750;

• Stainless steel type 347 material (excluding baffle-former bolts);

• Precipitation hardened stainless steel materials 4 and 15-5;

• Type 431 stainless steel materials.

Plant-specific material and design variations must be considered when applying MRP-191, Table 4-5 to plant-specific cases. Additionally, RVI components that are categorized as "Existing Programs" components need to be inspected for the applicable plant-specific aging effects.

Request Identify any locations that these materials are used in the AN0-2 RVls, including the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME) Code Section XI components. If these materials are used in the RVI components at AN0-2, provide the information regarding their proposed inspections and basis to demonstrate that the proposed inspections are consistent with the intent of MRP-227-A. Provide any relevant operating experience (OE) associated with the aging degradation mechanisms that are identified in MRP-227-A for these components.

Entergy Response The AN0-2 RVI components, including ASME Section XI B-N-3 components, are not constructed of any of the following materials:

• Nickel base alloys - lnconel 600 and weld metals -Alloy 82 and 182 and Alloy X-750;

• Stainless steel type 347 material;

• Precipitation hardened stainless steel materials 4 and 15-5

• Type 431 stainless steel materials.

AN0-2 does not have baffle-former bolts. The AN0-2 baffle plates are welded together.

Enclosure 1 to 2CAN091701 Page 5 of 8 RAl-3 Condition 7 - Updating of MRP-227 Appendix A Background -

Topical Report Condition 7 of the NRC staff's SE on MRP-227-A states, in part:" ... MRP-227, Appendix A shall be updated to include ... the Operating Experience Summary." Appendix A of MRP-227-A discusses OE under various degradation mechanisms for the RPV internal components of all nuclear steam supply system designs, including CE plants.

Section 2.3.10.1, "AN0-2 Operating Experience," and Section 3.7, "AN0-2 Unit Operating Experience," of the AN0-2 RVI AMP submittal discuss the OE relevant to the age-related degradation of the AN0-2 RVI. Section 2.3.10.1 states that extensive industry and AN0-2 OE has been reviewed during the development of the RVI AMP.

Request Provide a summary of any AN0-2 plant-specific OE relevant to age-related degradation of RVI components, including the plant-specific OE contained in Appendix A of MRP-227-A. Describe the actions taken to address this OE.

Entergy Response A summary of the AN0-2 plant specific operating experience (OE) relevant to age-related degradation of RV! components, including plant specific OE contained in Appendix A of MRP-227-A, is provided below:

Holddown Ring Stress Relaxation Field measurements taken during early fuel cycles revealed significant stress relaxation of the AN0-2 Type 403 stainless steel holddown ring. A series of measurements were taken during refueling outages over approximately a 20-year period. The holddown ring deflection eventually stabilized at an acceptable value and no further action was required. This OE item is not included in Appendix A of MRP-227-A.

ICI Thimble Tube Irradiation Induced Growth Greater than expected irradiation induced growt~ of Zircalloy-4 in-core instrumentation (ICI)

• thimble tubes has been discovered at AN0-2 and a number of other CE designed nuclear power plants. AN0-2 replaced the majority of the ICI thimble tubes in the fall 2006 refueling outage. Periodic measurements continue to be performed during refueling outages to monitor further growth and facilitate future replacements of the thimble tubes as needed. This OE item is included in Appendix A of MRP-227-A.

Enclosure 1 to 2CAN091701 Page 6 of 8 RAl-4 Action Item 3 - Evaluation of the Adequacy of Plant-Specific Existing Programs

Background

The MRP-227-A SE, Section 4.2.3, "Evaluation of the Adequacy of Plant-Specific Existing Programs" states, in part, that applicants/licensees of CE plants are required to justify the acceptability of an existing program, or to identify changes to manage the aging of CE thermal shield positioning pins and CE in-core instrumentation {ICI} thimble tubes for the period of extended operation.

Section 5.3 "SE Section 4.2.3, Applicant/Licensee Action Item 3 (Evaluation of the Adequacy of

.Plant-Specific Existing Programs}," of the AN0-2 RVI AMP submittal states that "AN0-2 complies with Applicant/licensee Action 3 through management and replacement of in-core instrumentation thimble tubes as described in [AN0-2 Engineering Request, ER-AN0-2003-0399-003; Rev. 0, "NCP for AN0-2 ICI Thimble Tube Replacement," March 9, 2006]."

Furthermore, Sections 2.3.10.1 and 3. 7 indicate that in CE-designed plants, zirconium-base alloy thimbles exhibited growth due to irradiation and that this thimble growth was a major aging management issue, and the thimbles were subsequently replaced. The licensee stated that AN0-2 has monitored the growth of the Zircaloy section of the thimble tube due to the high level of neutron radiation exposure and replaced ICI thimble tubes.

It is not clear if the licensee 'continues to monitor the replacement thimble tubes that were installed and will continue to manage the effects of age-related degradation for the replacement in-core instrumentation thimble tubes during the period of extended operation.

Request Clarify whether the replacement in-core instrumentation thimble tubes will be managed during the period of extended operation.

• If yes, explain the details regarding the inspection method and the associated acceptance criteria, and the frequency of inspection. Further, justify that these actions are adequate to manage the applicable aging effects.

• If no, justify that inspections during the period of extended operation are not necessary to manage age-related degradation.

Confirm whether the material of the replacement ICI thimble tubes are consistent with the material assessed in MRP-191, Table 4-5. If not, justify that no additional aging effects require management during the period of extended operation, other than loss of material due to wear, for these replacement ICI thimble tubes.

Entergy Response As noted above in the response to RAl-3, AN0-2 continues to perform periodic measurements of its*ICI thimble tubes for the age-related degradation effects of irradiation induced growth.

These measurements are performed and analyzed in accordance with approved plant procedures. The procedures involve the use of a thimble tube length measurement tool that is inserted into the open end of a thimble tube until the tool's bullet nose is firmly seated at.the to 2CAN091701 Page 7 of 8 bottom of the thimble tube. The measured lengths are then analyzed to project the remaining life of the thimble tubes, i.e., the time remaining until the tubes contact the bottom of the fuel assembly lower end fittings. There is no set measurement frequency period; rather it is dependent on the results of the inspections. Inspections are performed as needed to facilitate future thimble tube replacements. This approach to management of ICI thimble tube growth is in accordance with guidance provided by Westinghouse. The AN0-2 ICI thimble tubes are constructed of Zircalloy-4, consistent with the assumed material in MRP-191:

RAl-5 Physical Measurements Acceptance Criteria

Background

MRP-227-A Section 4.2.5, "Application of Physical Measurements as Part of l&E Guidelines for B&W [Babcock & Wilcox], CE, and Westinghouse RVI Components," states, in part, that applicants/licensees shall identify plant-specific acceptance criteria to be applied when performing the physical measurements for distortion in the gap between the top and bottom core shroud segments in CE units with core barrel shrouds assembled in two vertical sections.

The applicant/licensee shall include its proposed acceptance criteria and an explanation of how the proposed acceptance criteria are consistent with the plant's licensing basis and the need to maintain the functionality of the component being inspected under all licensing basis conditions of operation during the period of extended operation, as part of their submittal to apply the approved version of MRP-227.

Section 5.5, "SE Section 4.2.5, Applicant/Licensee Action Item 5 (Application of Physical Measurements as part of l&E Guidelines for B&W, CE, and Westinghouse RVI Components),"

of the Attachment to the AN0-2 RVI AMP submittal states that AN0-2 has a core barrel shroud assembled in two vertical sections and that, prior to performing the VT-1 examination, AN0-2 will develop acceptance criteria that are consistent with the licensing basis to ensure that the core shroud remains capable of performing its required functions.

The licensee's response did not identify the proposed acceptance criteria and an explanation of how the proposed acceptance criteria is consistent with its licensing basis, such that the component maintains its functionality under all licensing basis conditions of operation during the period of extended operation.

Request Provide the acceptance criteria for the VT-1 examination used for the physical measurement of the distortion in the gap between the top and bottom core shroud segments in CE unit, AN0-2, with core barrel shrouds assembled in two vertical sections. Justify that this acceptance criteria is consistent with the licensing basis to ensure that the core shroud remains capable of performing its required functions.

If the acceptance criteria is not yet developed, discuss the method for determining the acceptance criteria and justify this acceptance criteria will be consistent with the licensing basis to ensure that the core shroud remains capable of performing its required functions.

to 2CAN091701 Page 8 of 8 Entergy Response The same approach and process used to determine the acceptance criteria for the gap measurement between the core shroud segments for Calvert Cliffs and St. Lucie was implemented for AN0-2. It should be noted that Calvert Cliffs and St. Lucie were the lead plants for the Pressurized Water Owners' Group (PWROG) effort to address this issue. This approach and process is summarized below.

Differential thermal expansion (due to gamma heating) and irradiation-induced void swelling could cause vertical gaps to form between the interfacing plates of the upper and lower core shroud subassemblies, with associated horizontal (inward) deflections of the plates. The total vertical gap at any location would include a thermal contribution and a void swelling contribution, along with the as-fabricated gap. The total inward deflection would include thermal and void swelling contributions only.

The maximum thermal gap and inward deflection were explicitly calculated. The maximum void swelling gap was assumed based on the ability to readily detect the presence of gaps during physical examinations of the core shroud. The maximum inward deflection due to void swelling was derived by applying the thermal horizontal/thermal deflection ratio to the assumed vertical void swelling gap. The permissible as-fabricated gap was obtained from as-built drawings. The maximum total gap and inward deflection occur at the innermost corners of the interfacing plates, and there are no through-gaps, because the interfacing plates are welded together at their outer peripheries. The potential adverse effects of this maximum total gap and the associated inward deflection were evaluated and determined to be acceptable.

That portion of the total gap due to differential thermal expansion would only be present during power operation. That portion of the total gap due to irradiation-induced void swelling (along with the as-fabricated gap) would be present under all conditions, including plant shutdown, during which physical examinations of the core shroud will be performed.

During plant shutdown, the maximum value for the gap between the interfacing plates of the upper and lower core shroud subassemblies, which is equal to the permissible as-fabricated gap plus the maximum void swelling gap, is 0.125 inch (1/8 of an inch). Again, the maximum gap would occur at the innermost corners.

Based on these results, a maximum, bounding gap between the interfacing plates of the core shroud upper and lower subassemblie9, as could be present during plant shutdown, is set at 1/8 of an inch. This gap may be used as an acceptance criterion for the physical examination of gaps in the AN0-2 core shroud.

The maximum gap during plant shutdown, constituting the acceptance criterion for physical examination of gaps in the core shroud, must be within the range that can be detected by VT-1 visual examination. Per MRP-228 [Subsection 2.3.6.3, Paragraph b.1.]: 'Remote EVT-1 or VT-1 examination processes shall be demonstrated as capable of resolving lowercase characters ...

with character heights no greater than 0.044 in. (1.1 mm) at the maximum examination distance.' To distinguish between different characters of 0.044-inch height, it is reasonable to conclude that features of one-half that size (i.e., 0.022 inch or greater) can be resolved by VT-1 visual examination. 1"herefore, the acceptance criterion for physical examination of gaps in the core shroud must be;:: 0.022 inch.

I

Enclosure 2 of 2CAN091701 Revised Page of Table 5-2 of RVI AMP

Table 5-2. CE Plants Expansion Category Components from Table 4-5 ofMRP-227-A [4] (continued)

Examination Examination Item Appl icab ility Effect (Mechanism) Primary Link Comments Method/Frequency (1) Coverage/Frequency (1)

Core Support Barrel All plants Cracking (SCC) Core barrel Enhan ced visual (EVT-1 ) 100% of one side of the Contingency if Assembly assembly girth examination, with initial and accessible weld and adjacent indications are found Core barrel assembly welds subsequent examinations base metal surfaces for the in EVT-1 exam of axial welds dependent on the results of weld with the highest Core barrel assembly core barrel assembly girth calculated operating stress. girth welds weld examinations.

See Figures 4-15 of MRP-227-A.

Lower Support All plants except Cracking (SCC , Upper (core Visual (EVT-1) examination . 100% of accessible Contingency if Structure those with core Fatigue) including support barrel) surface.( 2 ) indications are found Lower core support shrouds assembled damaged or fractured flange weld Re-inspection every 10 in EVT-1 exam of beams years following initial See Figures 4-16 and 4-31 of Upper (core support with fu ll-height material inspection . MRP-227-A. barrel) flange weld shroud plates Core Shroud Bolted plant Cracking (IASCC, Core shroud Ultrasonic (UT) 100 % (or as supported by N/A Assembly designs Fatigue) bolts examination. plant-specific analysis) of (Bolted) core support column bolts Core support column Aging Management Re-inspection every 10 with neutron fluence bolts (IE) years following initial exposures > 3 dpa. (2 )

inspection .

(Not applicable for See Figures 4-16 and 4-33 of AN0 -2) MRP-227-A Core Shroud Plant designs with Cracking (IASCC) Core shroud Enhanced visual (EVT-1) Axial weld seams other than N/A Assembly core shrouds plate-former examination . the co re shroud re- entrant (Welded) assembled in two plate weld corner welds at the core mid-Re maining axial welds vertical sections Re-inspection every 10 plane.

years following initial inspection . See Figure 4-12 of MRP-227-A Core Shroud Plant designs with Cracking (IASCC) Shroud plates of Enhanced visual (EVT-1 ) Axial weld seams other than N/A Assembly core shrouds welded core examination . the core shroud re-entrant (Welded) assembled with Aging Management shroud corner welds at the core mid-Remaining axial welds , full-height shroud (IE) assemblies Re-inspection every 10 plane, plus ribs and rings.

Ribs and rings plates years following initial inspection . See Figure 4- 13 of (Not applicable for MRP-227-A AN0-2) 5-15 e structural Integrity Associates, Inc.

Report No. 1500227.40 l .RO