ML12276A006
| ML12276A006 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 08/31/2012 |
| From: | Paesano P Westinghouse, Westinghouse |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| WCAP-17347-NP, Rev 1 | |
| Download: ML12276A006 (101) | |
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0 WCAP-17347-NP August 201 2 Revision 1 0
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S PWR Vessel Internals Program 0 Plan for Aging Management of S
0 Reactor Internals at 0 Fort Calhoun Station S
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WCAP-17347-NP 0 Revision 1 0
S PWR Vessel Internals Program Plan for Aging Management 0 of Reactor Internals at Fort Calhoun Station 0
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0 Bethany A. Paden*
0 Reactor Internals Aging Management Karli N. Szweda*
0 Reactor Internals Aging Management 0
Richard A. Basel*
Reactor Internals Aging Management 0
August 2012 0
0 Approved: Patricia C. Paesano*, Manager Reactor Internals Aging Management 0
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- Electronically approved records are authenticated in the electronic document management system.
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Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Twp, PA 16066
© 2012 Westinghouse Electric Company LLC All Rights Reserved
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii TABLE OF CONTENTS LIST O F TA BL ES ........................................................................................................................................ v 0
SL IST O F FIGURE S ..................................................................................................................................... vi LIST O F A C RON Y M S ............................................................................................................................... vii A C KN OW LED GEM EN TS ......................................................................................................................... ix P U RP O SEE ..................................................................................................................................... 1-1 0
2 BA C KG R O U N D .......................................................................................................................... 2-1 3 PROGRAM OWNER ............................................................................................................. 3-1 4 DESCRIPTION OF THE FT. CALHOUN REACTOR INTERNALS AGING MANAGEMENT PROGRAMS AND INDUSTRY PROGRAMS ............................................. 4-1 4.1 Existing Ft. Calhoun Programs ........................................................................................ 4-4 4.1.1 ASME Section XI IWB, IWC, IWD, IWF Inservice Inspection Program ...... 4-4 4.1.2 Reactor Vessel Internals Inspection Program .................................................. 4-4 4.1.3 C hem istry Program .......................................................................................... 4-5 4.2 Supporting Ft. Calhoun Programs and Aging Management Supportive Plant Enhancem ents .................................................................................................................. 4-5 4.2.1 Reactor Internals Aging Management Review Process ................................... 4-5 4.3 Industry Programs ............................................................................................................ 4-6 4.3.1 CE NPSD- 1216, Aging Management for Reactor Internals ............................ 4-6 4.3.2 MRP-227-A, Reactor Internals Inspection and Evaluation Guidelines ........... 4-6 4.3.3 Ongoing Industry Programs ............................................................................. 4-9 4 .4 Summ ary ........................................................................................................................ 4-10 5 FT. CALHOUN REACTOR INTERNALS AGING MANAGEMENT PROGRAM A TT RIBU T ES .............................................................................................................................. 5-1 5.1 GALL Revision 2 Element 1: Scope of Program ........................................................... 5-1 5.2 GALL Revision 2 Element 2: Preventive Actions .......................................................... 5-3 5.3 GALL Revision 2 Element 3: Parameters Monitored or Inspected ................................ 5-4 5.4 GALL Revision 2 Element 4: Detection of Aging Effects ............................................. 5-5 5.5 GALL Revision 2 Element 5: Monitoring and Trending ................................................ 5-9 5.6 GALL Revision 2 Element 6: Acceptance Criteria ...................................................... 5-10 5.7 GALL Revision 2 Element 7: Corrective Actions ........................................................ 5-12 5.8 GALL Revision 2 Element 8: Confirmation Process .................................................... 5-13 5.9 GALL Revision 2 Element 9: Administrative Controls ................................................ 5-13 5.10 GALL Revision 2 Element 10: Operating Experience ................................................. 5-14 6 D EM ON STR AT ION .................................................................................................................... 6-1 0
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0 6.1 Demonstration of Topical Report Conditions Compliance to SE on MRP-227, 0 R ev ision 0 ........................................................................................... ............................ 6-2 6.2 Demonstration of Applicant/Licensee Action item Compliance to SE on MRP-227, 0 R evision 0 ........................................................................................................................ 6-3 0 6.2.1 SE Applicant/Licensee Action Item 1: Applicability of FMECA and 0 Functionality Analysis Assum ptions ............................................................... 6-3 6.2.2 SE Applicant/Licensee Action Item 2: PWR Vessel Intemals Components 0
within the Scope of License Renewal ............................................................. 6-5 0 6.2.3 SE Applicant/Licensee Action Item 3: Evaluation of the Adequacy of 0 Plant-Specific Existing Programs ............................. 6-6 6.2.4 SE Applicant/Licensee Action Item 4: B&W Core Support Structure Upper 0
Flange Stress R elief ......................................................................................... 6-6 0 6.2.5 SE Applicant/Licensee Action Item 5: Application of Physical 0 Measurements as part of I&E Guidelines for B&W, CE, and W estinghouse RVI Components ..................................................................... 6-7 0
6.2.6 SE Applicant/Licensee Action Item 6: Evaluation of Inaccessible B&W 0 C omponents ................................................................................................... 6-7 0 6.2.7 SE Applicant/Licensee Action Item 7: Plant-Specific Evaluation of CASS M aterials .......................................................................................................... 6-8 0
6.2.8 SE Applicant/Licensee Action Item 8: Submittal of Information for Staff 0 R eview and Approval .................................................................................... 6-10 0 7 PROJECTED PROGRAM ENHANCEMENT AND IMPLEMENTATION SCHEDULE ........ 7-1 0
0 8 IM PLEM EN TIN G DOCU M EN TS .............................................................................................. 8-1 0 9 R EFER ENC E S ............................................................................................................................. 9-1 0
0 APPENDIX A ILLUSTRATIONS .............................................................................................. A-1 0 APPENDIX B FT. CALHOUN NUCLEAR PLANT LICENSE RENEWAL AGING 0
MANAGEMENT REVIEW
SUMMARY
TABLES .......................................... B-1 0 0
APPENDIX C MRP-227-A AUGMENTED INSPECTIONS ................................................... C-1 0 0
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Table 6-1 Topical Report Conditions Compliance to SE on MRP-227, Revision 0 ........................ 6-2 0 Table 6-2 Summary of FCS CASS Components and their Susceptibility to TE ............................. 6-9 Table 6-3 FCS Reactor Internals: Martensitic SS and PH-SS Components .................................... 6-9 0 Table 7-1 Aging Management Program Enhancement and Inspection Implementation Summary .7-1 Table B-1 LRA Aging Management Evaluation Summary - Ft. Calhoun LRA Table 2.3.1.1-1 Reactor Vessel Internals Component Types Subject to Aging Management Review .... B- 1 0 Table C-I MRP-227-A Primary Inspection and Monitoring Recommendations for CE-Designed Intern als .......................................................................................................................... C- 1 0
Table C-2 MRP-227-A Expansion Inspection and Monitoring Recommendations for CE-Designed 0 Intern als .......................................................................................................................... C -6 Table C-3 MRP-227-A Existing Inspection and Aging Management Programs Credited in 0 Recommendations for CE-Designed Internals ................................................................ C-8 0 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for CE-0 Designed Internals .......................................................................................................... C -9 S
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LIST OF FIGURES 0
S Figure A-I Typical CE Internals ......................................................................................................... A -1 0 Figure A-2 Core Support Barrel Assem bly .............................................................................................. A-2 0 Figure A-3 Illustration of Typical Bolting in Core Plates ........................................................................ A-3 S Figure A-4 Typical Bolted Core Shroud Assembly ................................................................................. A-4 Figure A-5 High-Fluence Seam Locations ........................................................................................ A-5 Figure A-6 Exaggerated View of Void Swelling Induced Distortion in Assembly ..............
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Figure A-7 Typical Core Support Barrel Structure .................................................................................. A-7 Figure A-8 Core Support Barrel, Core Shroud Assembly, and Lower Support Structure ....................... A-8 Figure A-9 Schematic View of Lower Support Structure Assembly ....................................................... A-9 0 Figure A- 10 Lower Support Structure Assembly .................................................................................... A-9 0 Figure A-Il (a) CE Schematic Illustration of a Portion of the Fuel Alignment Plate, and (b) CE S Radial-View Schematic Illustration of the Guide Tubes ............................................ A-10 0 Figure A- 12 (a) Fuel Bundle Alignment Plate and (b) Upper Guide Structure ................................ A-Il 0 Figure A- 13 CE Schematic Illustration of the Control Element Assembly (CEA) Shroud Assembly.. A- 12 0 Figure A-14 Control Rod Shroud Assembly ...................................... A-13 0
Figure A- 15 Isometric View of the Lower Support Structure in the CE Core Shrouds with S
Full-Height Shroud Plates Units .................................................................................. A- 14 Figure A- 16 Bolting in a Typical Westinghouse Baffle-Former Structure ....................................... A-15 0 Figure A- 17 CE Core Support Columns ........................................................................................... A- 16 0
S Figure A- 18 CE Lower Core Support Structure - Cross-Section ..................................................... A- 17 0
Figure A- 19 Potential crack locations for CE welded core shroud assembled in stacked sections ....... A-18 0 Figure A-20 Locations of potential separation between core shroud sections caused by swelling S induced warping of thick flange plates in CE welded core shroud assembled in stacked sections ................................................................................................................ A- 19 S
Figure A-21 CE welded core shroud with full height panels .................................................................
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Figure A-22 CE lower support structures for welded core shrouds: separate core barrel and lower support structure assembly with lower flange and core support plate .............................. A-21 S
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LIST OF ACRONYMS AMP Aging Management Program Plan AMR Aging Management Review ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel B&W Babcock & Wilcox BWR boiling water reactor CASS cast austenitic stainless steel CE Combustion Engineering CEA control element assembly CEDM control element drive mechanism CEOG Combustion Engineering Owners Group CFR Code of Federal Regulations CLB current licensing basis dpa displacements per atom EFPY effective full-power year EPRI Electric Power Research Institute EVT enhanced visual testing (a visual NDE method that includes EVT- 1)
FCS Fort Calhoun Station FMECA failure modes, effects, and criticality analysis GALL Generic Aging Lessons Learned I&E Inspection and Evaluation IASCC irradiation-assisted stress corrosion cracking ICI in-core instrumentation ID inside diameter IE irradiation embrittlement IGA intergranular attack IGSCC intergranular stress corrosion cracking INPO Institute of Nuclear Power Operations ISI inservice inspection ISR irradiation-enhanced stress relaxation LOCA loss-of-coolant accident LRA License Renewal Application LRAAI License Renewal Applicant Action Item MRP Materials Reliability Program NDE nondestructive examination NEI Nuclear Energy Institute NOS Nuclear Oversight Section NRC United States Nuclear Regulatory Commission NSSS nuclear steam supply system OD outer diameter OE Operating Experience OEM Original Equipment Manufacturer OER Operating Experience Report OPPD Omaha Public Power District 0
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0 LIST OF ACRONYMS (cont.) 0 PH precipitation-hardening 0 PWR pressurized water reactor 0 PWROG Pressurized Water Reactor Owners Group 0 PWSCC primary water stress corrosion cracking QA quality assurance S RCS reactor coolant system 0 RIS Regulatory Issue Summary 0 RO refueling outage RV reactor vessel 0 RVI reactor vessel internals 0 SCC stress corrosion cracking 0 SER Safety Evaluation Report SRP Standard Review Plan 0
SS stainless steel 0 TE thermal embrittlement 0 TLAA time-limited aging analysis TR technical report 0
UGS upper guide structure 0 UT ultrasonic testing (a volumetric NDE method) 0 VT visual testing (a visual NDE method that includes VT- 1 and VT-3) 0 0
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0 ACKNOWLEDGEMENTS 0 The authors would like to thank the members of the Omaha Public Power District (OPPD) Fort Calhoun 0 Aging Management Program Team led by Bob Lisowyj and Kristen Maassen, and our associates at 0 Westinghouse, including Dr. Randy Lott, for their efforts in supporting development of this WCAP.
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- 1 PURPOSE 0
- The purpose of this report is to document the Omaha Public Power District Fort Calhoun Station (hereafter "Ft. Calhoun") Reactor Vessel Internals (RVI) Aging Management Program Plan (AMP). The 0purpose of the AMP is to manage the effects of aging on RVI through the license renewal period, which begins at the expiration of the original license on August 9, 2013. This document provides a description of the program as it relates to the management of aging effects identified in various regulatory and updated industry-generated documents in addition to the program described in the Ft. Calhoun document
- EA-FC 134 [1 ] in support of license renewal program evaluations. This AMP is supported by existing 0Ft. Calhoun documents and procedures and, as required by industry experience or directives in the future, will be updated or supported by additional documents to provide clear and concise direction for the effective management of aging degradation in reactor internals components. These actions provide assurance that operations at Ft. Calhoun will continue to be conducted in accordance with the current
- licensing basis (CLB) for the RVI by fulfilling License Renewal commitments [2], following American
- Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section XI Inservice Inspection (ISI) programs [3, 4, and 8] and meeting industry requirements [5]. Based on the programs sponsored by United States utilities through the Electric Power Research Institute (EPRI) managed
- Materials Reliability Program (MRP) and the Pressurized Water Reactor Owners Group (PWROG), this AMP fully captures the intent of the additional industry guidance for reactor internals augmented inspections.
0The main objectives for the Ft. Calhoun RVI AMP are to:
S 0 Demonstrate that the effects of aging on the RVI will be adequately managed for the period of extended operation in accordance with 10 CFR 54 [6].
0* Define and implement the industry-defined (EPRI/MRP and PWROG) pressurized water reactor 0(PWR) RVI requirements and guidance for managing aging of reactor internals.
0 0 Provide an inspection plan summary for the Ft. Calhoun reactor internals.
0Ft. Calhoun License Renewal Commitments 16, 17, and 18 [2] for the "Reactor Vessel Intemals Inspection Program" commit Ft. Calhoun as follows:
0 Commitment 16:
- Visual inspections of the core shrouds at Palisadesand FCS in 1995 and 1993, respectively, revealedno panel separationandno missing bolts. Ten-year inservice inspections were performed at FCS in 1992 and will be performedagain in 2003 andprior to the period of 5extended operation. The results of these inspections, the Palisadesin-service inspection results, and the results of industryprograms will be monitored to determine if additionalaction, such as 5ultrasonic 0 inspection, is necessary.
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The EPRI MRP is developing an actionplan to address potentialSCC of reactor vessel internals.
OPPD is participatingin this programand will take action, as necessary, in response to any recommendationsandfindings comingfrom the evaluation.
Commitment 17:
0 OPPDhas incorporatedan augmented inspection of the thermalshield bolting or pins within the Reactor Vessel InternalsInspection Program.OPPDcontinues to monitor thermal shield vibrationsas a task within the Reactor Vessel InternalsInspection Program (B.2.8).
Commitment 18: 0 0
The following enhancements will be made to the Reactor Vessel Internals Inspection Program:A fluence and stress analysis will be performed to identify criticallocations. A fracture mechanics analysisfor criticallocations will be performed to determineflaw acceptance criteriaand resolution requiredto detectflaws. Appropriateinspection techniques will be implemented based on analyses.
(Forthe R Vlflow skirt) Thefracture mechanics analysis committed to in Section B. 2.8 of the LRA will be performed [2].
Augmented inspections, based on required program enhancements resulting from industry programs, will become part of the Ft. Calhoun Inservice Inspection Program [8]. Corrective actions for augmented inspections will be developed and will use repair and replacement procedures equivalent to those requirements in ASME B&PV Code,Section XI, or they will use processes determined to be equivalent to or more rigorous than currently defined procedures as determined independently by OPPD or in cooperation with the industry.
This AMP for the Ft. Calhoun reactor internals demonstrates that the program adequately manages the effects of aging for reactor internals components and establishes the basis for providing reasonable assurance that the intemals components will continue to perform their intended function through the Ft. Calhoun license renewal period of extended operation. Since MRP-227-A was released, the NRC published a Regulatory Issue Summary (RIS) that implemented new guidelines as to when plants must submit an inspection plan, which is October 1, 2012, for FCS. Revision I of this AMP incorporates the changes from MRP-227-A and the RIS.
The development and implementation of this program meets the guidelines provided in the RIS [26].
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- 2 BACKGROUND S
- The management of aging degradation effects in reactor internals is required for nuclear plants considering or entering license renewal, as specified in the NRC Standard Review Plan (SRP) [10]. The 0U.S. nuclear power industry has been actively engaged in recent years in a significant effort to support the industry goal of responding to these requirements. Various programs have been underway within the
- industry over the past decade to develop guidelines for managing the effects of aging within PWR reactor internals. Later, an effort was engaged by the EPRI MRP to address the PWR internals aging management 0issue for the following three currently operating U.S. reactor designs: Westinghouse, Combustion
The MRP first established a framework and strategy for the aging management of PWR internals components using proven and familiar methods for inspection, monitoring, surveillance, and 0communication. Based on that framework and strategy and on the accumulated industry research data, the
- following elements of an AMP were further developed [11, 12, and 13]:
0Screening criteria were developed, considering chemical composition, neutron fluence exposure, temperature history, and representative stress levels, for determining the relative susceptibility of PWR internals components to each of eight postulated aging mechanisms (further discussed in Section 4 of this program).
- PWR internals components were categorized, based on the screening criteria, as follows:
- Components for which the effects from the postulated aging mechanisms are insignificant
- Components that are moderately susceptible to the aging effects
- - Components that are significantly susceptible to the aging effects 0
Functionality assessments were performed based on representative plant designs of PWR internals components and assemblies of components, using irradiated and aged material 0properties, 0 to determine the effects of the degradation mechanisms on component functionality.
Aging management strategies were developed combining the results of the functionality assessment with several contributing factors to determine the appropriate aging management methodology, baseline 0examination timing, and the need and timing of subsequent inspections. The considered factors include component accessibility, operating experience (OE), existing evaluations, and prior examination results.
The industry guidance is contained within two separate EPRI MRP documents:
- MRP-227-A [5], "PWR Internals Inspection and Evaluation Guidelines" (hereafter referred to as "the I&E Guidelines" or simply "MRP-227-A") provides industry background for the guidelines, lists of reactor internals components requiring inspection, and the timing for initial inspections of 0those components. For each component, the guidelines require a specific type of nondestructive 5examination (NDE) and give criteria for evaluating inspection results. MRP-227-A provides a standardized approach to PWR internals aging management for each unique reactor design S(Westinghouse, B&W, and CE).
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2-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 MRP-228 [13], "Inspection Standard for PWR Internals," provides guidance on the qualification/demonstration of the required NDE techniques and other criteria pertaining to the actual performance of the inspections.
The PWROG has also developed "Reactor Internals Acceptance Criteria Methodology and Data Requirements" for the MRP-227, Revision 0 inspections, where feasible [14]. This document has been submitted to the NRC for review, and will be updated to incorporate changes from MRP-227-A [5]. Final reports have been developed and are available for industry use in support of planned license renewal inspection commitments. Individual plants will develop plant-specific acceptance criteria for internals components if a generic approach is not practical or available.
As described in MRP-232 [15], the primary functions of the Ft. Calhoun internals are to provide support, guidance, and protection to the reactor core, provide a passageway for the distribution of the reactor coolant flow to the reactor core, provide support, guidance, and protection for in-core instrumentation (ICI), and provide gamma and neutron shielding for the reactor vessel. The RVI consist of the following three main assemblies: an upper internals assembly (also known as an "upper guide structure" (UGS) at Ft. Calhoun) [7] that is removed during refueling as a single component to provide access to the fuel assemblies, the core support barrel, and the lower support structure. In addition, there are three other RVI assemblies in CE-designed plants: the control element assembly (CEA) shroud assembly, the core shroud assembly, and the in-core instrumentation support system. Because the design at Ft. Calhoun is different from other CE-designed plants, some of the components have a different naming scheme. As mentioned, one of these is the "upper guide structure," which is similar to the "upper internals assembly" in other CE plants. These components have some differences in design, but they serve similar functions, and the MRP-227-A augmented inspection requirements are still applicable. FCS is one of two CE-designed plants that have bolted core shrouds. A brief summary of the design characteristics for these internals specific to Ft. Calhoun is provided in the following subsection. Note that the component names at Ft.
Calhoun are used rather than the generic names for the other CE-designed plants. The general arrangement of the CE-designed PWR internals components as well as the specific configuration of Ft.
Calhoun is shown in Figure A- 1.
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- Upper Guide Structure 0
The UGS is located above the reactor core within the core support barrel assembly. It is removed during refueling as a single component in order to provide access to the fuel assemblies. The UGS is an integral 0assembly that includes the fuel alignment plate, the CEA shroud assemblies, the in-core instrumentation support systems, and the hold-down ring. The functions of the UGS are to provide alignment and support to the fuel assemblies, to maintain CEA shroud spacing, to prevent movement of the fuel assemblies in the case of a severe accident condition, and to protect the CEAs from crossflow effects in the upper 0plenum.
0 The UGS support plate rests on the core support barrel upper flange [5 and 7].
- Core Support Barrel Assembly 0The core support barrel assembly consists of the core support barrel cylinders, nozzles, flange, alignment
- keys, snubbers, and the thermal shield [5 and 7].
The core support barrel is a cylinder which contains the core and other internals components. Its function 0is to resist static loads from the fuel assemblies and other internals components and dynamic loads from Snormal operating hydraulic flow, seismic events, and loss-of-coolant accident (LOCA) events. The core support barrel supports the lower support structure, and the core support plate rests on top of the lower support structure. The core support barrel upper flange is a thick ring that suspends the core support barrel 0from 0 a ledge on the reactor vessel.
The thermal shield is a cylindrical structure that reduces the neutron flux and radiation heating in the reactor vessel wall. The thermal shield rests upon eight equally spaced lugs at the outer periphery of the core support barrel. The lower end of the thermal shield is positioned radially with thermal shield bolts,
- which hold the thermal shield tight against the core support barrel [5 and 7].
0 Lower Support Structure 0The lower support structure consists of the core support plate and the core support columns and beams.
0The core support plate positions and supports the reactor core and directs the reactor coolant flow into each fuel assembly. The weight of the core is transmitted to the core support barrel through the lower support structure via the core support columns. Fuel alignment holes in the core support plate engage 0lower fuel assembly alignment pins to provide guidance for and limit lateral movement of the individual 0fuel assemblies. Ft. Calhoun uses bolts to attach the core support plate to the core support columns [5 and
- 7].
- Core Shroud Assembly 0
0The core shroud assembly is located within the core support barrel and directly below the UGS.
Ft. Calhoun has a bolted core shroud assembly where the core shroud plates are fastened to the centering 0plates with structural bolts. The centering plates, and thus the assembly, are attached to the core support 0barrel by structural bolts as well. The core shroud assembly provides a boundary between the reactor coolant bypass flow on the inside of the core support barrel and the reactor coolant flow through the fuel assemblies. It also limits the amount of coolant bypass flow and reduces the lateral motion of the fuel
- assemblies [5 and 7].
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2-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Control Element Assembly Shroud Assemblies The CEA shroud assemblies consist of the CEA shrouds, the CEA shroud bolts, and the extension shaft guide tubes. The CEA shrouds protect the CEAs from cross-flow effects in the upper plenum. The lower end of the shrouds is bolted to the fuel alignment plate. The control element drive mechanisms are positioned on the reactor vessel closure head and are coupled to the control elements by the control element drive extension shafts. The control element shroud assemblies were considered in the development of this AMP. The control element drive mechanisms and control elements were not included in the list of internals components [5 and 7].
In-Core Instrumentation Support System The ICI support system consists of in-core instrumentation guide tubes and components that provide support to the in-core instrumentation. For plants with top-entry, in-core instrumentation assemblies, such as Ft. Calhoun, the in-core instrumentation is inserted through the reactor vessel head and into a guide tube via a nozzle. The guide tubes interface with the fuel alignment plate to align with holes in the fuel assemblies. ICI elements are inserted from the guide tubes directly into the fuel assemblies [5 and 7].
Ft. Calhoun License Extension Ft. Calhoun was granted a license for extended operation by the NRC through the issuance of a SER in NUREG- 1782 [2]. In the SER, the NRC concluded that the Ft. Calhoun License Renewal Application (LRA) [9] adequately identified the RVI systems, structures, and components that are subject to an Aging Management Review (AMR), as required by 10 CFR 54.21 (a)(1) [6]. A list of the Ft. Calhoun RVI components and subcomponents that are subject to AMP requirements, and have already been reviewed 0 by the NRC as part of the LRA, is included in Table B-1.
In accordance with 10 CFR Part 54 [6], frequently referred to as the License Renewal Rule, Ft. Calhoun has developed a procedure to direct the performance of AMRs of mechanical structures and components
[1]. The U.S. industry, through the efforts of the MRP and PWROG, has further investigated the components and subcomponents that require aging management to support continued reliable function.
MRP-227- A contains a NEI 03-08 [16] "Mandatory" requirement that each plant will be required to use MRP-227-A and MRP-228 to develop and implement an AMP for reactor internals no later than three years after the initial industry issuance of MRP-227, Revision 0 in December 2008. Therefore, plant AMPs must be completed by December 2011 or sooner, as required by plant-specific license renewal commitments. According to [26], FCS must submit their AMP for review and approval by the NRC in accordance with MRP-227-A [5] no later than October 1, 2012.
0 The information contained in this AMP fully complies with the requirements and guidance of the referenced documents. The AMP will manage aging effects of the RVI so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation.
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0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3 PROGRAM OWNER S The Reactor Vessel Internals Program [ 1] is part of the Ft. Calhoun aging management programs. The 0 successful implementation and comprehensive long-term management of the Ft. Calhoun RVI AMP will require interaction with multiple industry organizations including, but not limited to, the ASME, MRP, NRC, and PWROG. The responsibilities of the individual Ft. Calhoun groups are provided in applicable plant procedures. The Engineering Programs department has overall responsibility for maintaining and 0 implementing the RVI inspection program and the RVI AMP [I]. OPPD will maintain cognizance of industry activities related to PWR internals inspection and aging management and will address and 0 implement industry guidance stemming from those activities, as appropriate under NEI 03-08 practices
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- 4 DESCRIPTION OF THE FT. CALHOUN REACTOR INTERNALS
- AGING MANAGEMENT PROGRAMS AND INDUSTRY
- PROGRAMS The U.S. nuclear industry, through the combined efforts of utilities, vendors, and independent consultants, has defined a generic guideline to assist utilities in developing reactor internals plant-specific aging management programs based on inspection and evaluation. As noted in Section 3, the PWR Vessel Internals Program is a part of the Ft. Calhoun aging management programs. The intent of this 0 program is to ensure the long-term integrity and safe operation of the reactor internals components.
Ft. Calhoun has developed this AMP in conformance with the 10 Generic Aging Lessons Learned
This reactor internals AMP utilizes a combination of prevention, mitigation, and condition monitoring.
0Where applicable, credit is taken for existing programs such as chemistry [18] and inspections prescribed
- by the ASME Section XI Inservice Inspection Program [3] and [4]. These existing programs combined with the LRA commitment for augmented inspections or evaluations as recommended by MRP-227-A reflect on OPPD's proactive approach for aging management.
Aging degradation mechanisms that impact internals have been identified and documented in a Ft. Calhoun AMR [1] prepared using the corporate procedural guidance document as described in [ 1] in support of the License Renewal effort. The overall outcome of the reviews and the additional work 0performed by the industry, as summarized in MRP-227-A, are to provide appropriate augmented
- inspections for reactor internals components to provide early detection of the degradation mechanisms of concern. Therefore, this AMP is consistent with the existing Ft. Calhoun AMR methodology and the additional industry work summarized in MRP-227-A. All sources are consistent and address concerns 0about component degradation resulting from the following eight material aging degradation mechanisms identified as affecting reactor internals:
0Stress corrosion cracking (SCC) refers to local, non-ductile cracking of a material due to a combination of tensile stress, corrosive environment, and susceptible material. The actual mechanism that causes SCC involves a complex interaction of environmental and metallurgical 0factors.
0 The aging effect is cracking.
Irradiation-Assisted Stress Corrosion Cracking 0* Irradiation-assisted stress corrosion cracking (IASCC) is a unique form of SCC that occurs only
- in highly irradiated components. The aging effect is cracking.
0 Wear 0Wear is caused by the relative motion between adjacent surfaces, with the extent determined by the relative properties of the adjacent materials and their surface condition. The aging effect is loss of material.
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4-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Fatioe Fatigue is defined as the structural deterioration that can occur as the result of repeated stress/strain cycles caused by fluctuating loads or temperatures. After repeated cyclic loading of sufficient magnitude, microstructural damage can accumulate, leading to macroscopic crack initiation at the most highly affected locations. Subsequent mechanical or thermal cyclic loading can lead to growth of the initiated crack. Corrosion fatigue is included in the degradation description.
Low-cycle fatigue is defined as cyclic loads that cause significant plastic strain in the highly stressed regions, where the number of applied cycles is increased to the point where the crack eventually initiates. When the cyclic loads are such that significant plastic deformation does not occur in the highly stressed regions, but the loads are of such increased frequency that a fatigue crack eventually initiates, the damage accumulated is said to have been caused by high-cycle fatigue. The aging effects of low-cycle fatigue and high-cycle fatigue are additive.
Fatigue crack initiation and growth resistance are governed by a number of material, structural, S and environmental factors such as stress range, loading frequency, surface condition, and presence of deleterious chemical species. Cracks typically initiate at local geometric stress concentrations such as notches, surface defects, and structural discontinuities. The aging effect is cracking.
0 Thermal Aging Embrittlement Thermal aging embrittlement is the exposure of delta ferrite within cast austenitic stainless steel (CASS) and precipitation-hardened (PH) stainless steel to high inservice temperatures, which can result in an increase in tensile strength, a decrease in ductility, and a loss-of-fracture toughness.
Some degree of thermal aging embrittlement can also occur at normal operating temperatures for CASS and PH stainless steel internals. CASS components have a duplex microstructure and are particularly susceptible to this mechanism. While the initial aging effect is loss of ductility and toughness, unstable crack extension is the eventual aging effect if a crack is present and the local applied stress intensity exceeds the reduced fracture toughness.
Irradiation Embrittlement o
Irradiation embrittlement is also referred to as neutron embrittlement. When exposed to high-energy neutrons, the mechanical properties of stainless steel and nickel-based alloys can be changed. Such changes in mechanical properties include increasing yield strength, increasing ultimate strength, decreasing ductility, and a loss-of-fracture toughness. The irradiation embrittlement aging mechanism is a function of both temperature and neutron fluence. While the initial aging effect is loss of ductility and toughness, unstable crack extension is the eventual aging effect if a crack is present and the local applied stress intensity exceeds the reduced fracture toughness.
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- Void Swelling and Irradiation Growth Void swelling is defined as a gradual increase in the volume of a component caused by formation 0of microscopic cavities in the material. These cavities result from the nucleation and growth of 0clusters of irradiation-produced vacancies. Helium produced by nuclear transmutations can have a significant impact on the nucleation and growth of cavities in the material. Void swelling may produce dimensional changes that exceed the tolerances on a component. Strain gradients produced by differential swelling in the system may produce significant stresses. Severe swelling 0(> 5 percent by volume) has been correlated with extremely low fracture toughness values. Also included in this mechanism is irradiation growth of anisotropic materials, which is known to cause significant dimensional changes within in-core instrumentation tubes that are fabricated from zirconium alloys. While the initial aging effect is dimensional change and distortion, severe 0void swelling may result in cracking under stress.
Thermal and Irradiation-Enhanced Stress Relaxation or Irradiation-Enhanced Creep 0The loss of preload aging effect can be caused by the aging mechanisms of stress relaxation or creep. Thermal stress relaxation (or primary creep) is defined as the unloading of preloaded components due to long-term exposure to elevated temperatures, as seen in PWR internals. Stress relaxation occurs under conditions of constant strain where part of the elastic strain is replaced with plastic strain. Available data show that thermal stress relaxation appears to reach saturation in a short time (< 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />) at PWR internals temperatures.
Creep (or more precisely, secondary creep) is a slow, time and temperature-dependent, plastic deformation of materials that can occur at stress levels below the yield strength (elastic limit).
0Creep occurs at elevated temperatures where continuous deformation takes place under constant stress. Secondary creep in austenitic stainless steels is associated with temperatures higher than those relevant to PWR internals even after taking into account gamma heating. However, 0irradiation-enhanced creep (or more simply, irradiation creep) or irradiation-enhanced stress 0relaxation (ISR) is an athermal process that depends on the neutron fluence and stress, and it can also be affected by void swelling should it occur. The aging effect is a loss of mechanical closure integrity (or preload) that can lead to unanticipated loading that, in turn, may eventually cause 0subsequent 0 degradation by fatigue or wear and result in cracking.
0The Ft. Calhoun RVI AMP is focused on meeting the requirements of the 10 elements of an AMP as described in NUREG-1801, GALL Report Section XI.M I6A for PWR Vessel Internals [25]. In the Ft.
- Calhoun RVI AMP, this is demonstrated through application of existing Ft. Calhoun AMR methodology 0that credits inspections prescribed by the ASME Code Section XI Inservice Inspection Program, existing 5Ft. Calhoun programs, and additional augmented inspections based on MRP-227-A recommendations. A description of the applicable existing Ft. Calhoun programs and compliance with the elements of the 0GALL 0 Report is contained in the following subsections.
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4-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 4.1 EXISTING FT. CALHOUN PROGRAMS Ft. Calhoun's overall strategy for managing aging in reactor internals components is supported by the following existing programs:
- ASME Section XI IWB, IWC, IWD, IWF Inservice Inspection Program [3] and [4]
- Reactor Vessel Internals Inspection Program
- Chemistry Program These are established programs that support the aging management of RCS components in addition to the RVI components. Although affiliated with and supporting the RVI AMP, these programs will continue to be managed under the existing structure. Brief descriptions of the programs are included in the following subsections.
4.1.1 ASME Section XI IWB, IWC, IWD, IWF Inservice Inspection Program The ASME Section XI IWB, IWC, IWD, IWF Inservice Inspection Program [3] and [4] is an existing program that facilitates inspections to identify and correct degradation in Class 1, 2, and 3 piping, components, their supports, and integral attachments. The program includes periodic visual, surface and/or volumetric examinations and leakage tests of all Class 1, 2, and 3 pressure-retaining components, their supports and integral attachments, including welds, pump casings, valve bodies, pressure-retaining bolting, piping/component supports, and reactor head closure studs. These are identified in ASME Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components" [ 19], or commitments requiring augmented inservice inspections, and are within the scope of license renewal. This program is in accordance with 10 CFR 50.55a.
S The evaluation of this program against the 10 attributes in the GALL Report for Programs XI.M 1, XI.M3, and XI.S3 [25] in support of the Ft. Calhoun LRA remains applicable.
4.1.2 Reactor Vessel Internals Inspection Program The Reactor Vessel Internals Inspection Program is an existing program that manages the aging effects for reactor vessel internals. The program provides for: (a) Inservice Inspection (ISI) in accordance with ASME Section XI requirements, including examinations performed during the 10-year ISI examination; (b) Participation in industry initiatives to evaluate the significance of void swelling; (c) Monitoring and control of reactor coolant water chemistry in accordance with the EPRI guidelines (see Chemistry Program) to mitigate SCC or IASCC; and (d) Participation in industry initiatives that will generate additional data on aging mechanisms relevant to RVI and develop appropriate inspection techniques to permit detection and characterization of features of interest.
Ft. Calhoun is unique in the CE-designed fleet in its continued application of a thermal shield.
Commitment 17 that requires Ft. Calhoun to monitor and address the aging of the thermal shield is satisfied by the Ft. Calhoun Reactor Vessel Internals Inspection Program. The effects of material degradation have been monitored.
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- The evaluation of this program against the 10 attributes in the GALL Report for Program XI.M 16 A [25]
- in support of the Ft. Calhoun LRA remains applicable.
04.1.3 Chemistry Program The Chemistry Program is an existing program that is credited for managing aging effects by controlling the environment to which internal surfaces of systems and components are exposed. Such effects include 0 the following:
Loss of material due to general, pitting, and crevice corrosion 0
Cracking due to SCC Steam generator tube degradation caused by denting, intergranular attack (IGA), and outer
- diameter (OD) SCC 0The aging effects are minimized by controlling the chemical species that cause the underlying 0mechanisms that produce them. The program provides assurance that an elevated level of contaminants and, where applicable, oxygen does not exist in the systems and components covered by the program, thus minimizing the occurrences of aging effects, and maintaining each component's ability to perform
- The evaluation of this program against the 10 attributes in the GALL Report for Program XI.M2 [25] in support of the Ft. Calhoun LRA remains applicable.
- 4.2 SUPPORTING FT. CALHOUN PROGRAMS AND AGING MANAGEMENT
- SUPPORTIVE PLANT ENHANCEMENTS 4.2.1 Reactor Internals Aging Management Review Process 0
5A comprehensive review of aging management of reactor internals was performed according to the requirements of the License Renewal Rule [6]. EA-FC 134 [1] documents the results of the AMR performed in support of Ft. Calhoun license renewal for reactor internals. The Ft. Calhoun LRA was
- approved by the NRC in NUREG- 1782 [2]. RVI components specifically noted as requiring aging management, as identified in the NUREG, are summarized in this AMP.
5The 0 referenced documents supported the LRA as follows:
- 1. Identified applicable aging effects requiring management
- 2. Associated AMPs to manage those aging effects described in Section 4.0 0
- 3. Identified enhancements or modifications to existing programs, new AMPs, or any other actions required to support the conclusions reached in the calculation 0
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4-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 AMRs were performed for each Ft. Calhoun system that contained long-lived, passive components requiring AMR. The results are of the AMRs are fully incorporated in the Ft. Calhoun RVI AMR [I].
4.3 INDUSTRY PROGRAMS 4.3.1 CE NPSD-1216, Aging Management for Reactor Internals The Combustion Engineering Owners Group (CEOG) topical report CE NPSD- 1216 [11] contains a technical evaluation of aging degradation mechanisms and aging effects for CE RVI components. The CEOG report provided guidance for CEOG member plant owners to manage effects of aging on RVI during the period of extended operation, using approved aging management methodologies to develop plant-specific AMPs.
The AMR for the Ft. Calhoun internals, documented in [1], was completed in a manner consistent with the approach of CE NPSD-1216 [11 ]. Both the Ft. Calhoun-specific AMR document and the generic CE document were completed in accordance with 10 CFR 54.
4.3.2 MRP-227-A, Reactor Internals Inspection and Evaluation Guidelines MRP-227-A, as discussed in Section 2, was developed by a team of industry experts including utility representatives, NSSS vendors, independent consultants, and international committee representatives who reviewed available data and industry experience on materials aging. The objective of the group was to develop a consistent, systematic approach for identifying and prioritizing inspection and evaluation requirements for reactor internals. The following subsections briefly describe the industry process.
4.3.2.1 MRP-227-A, RVI Component Categorizations MRP-227-A used a screening and ranking process to aid in the identification of required inspections for specific RVI components. MRP-227-A credited existing component inspections, when they were deemed adequate, as a result of detailed expert panel assessments conducted in conjunction with the development of the industry document. Through the elements of the process, the reactor internals for all currently licensed and operating PWR designs in the U.S. were evaluated in the MRP program and appropriate inspection, evaluation, and implementation requirements for reactor internals were defined.
0 Based on the completed evaluations, the RVI components are categorized within MRP-227-A as "Primary" components, "Expansion" components, "Existing Programs" components, or "No Additional Measures" components, described as follows:
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- Primary Those PWR internals that are highly susceptible to the effects of at least one of the eight aging mechanisms were placed in the Primary group. The aging management requirements that are needed to ensure functionality of Primary components are described in the I&E guidelines. The Primary group also includes components that have shown a degree of tolerance to a specific aging degradation effect, but for which no highly susceptible component exists or for which no highly susceptible component is accessible.
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Those PWR internals that are highly or moderately susceptible to the effects of at least one of the 0eight aging mechanisms, but for which functionality assessment has shown a degree of tolerance 0to those effects, were placed in the Expansion group. The schedule for implementation of aging management requirements for Expansion components depends on the findings from the examinations of the Primary components at individual plants.
0Existing Programs 0
Those PWR internals that are susceptible to the effects of at least one of the eight aging mechanisms and for which generic and plant-specific existing AMP elements are capable of 0managing those effects, were placed in the Existing Programs group.
- No Additional Measures 0Those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria were placed in the No Additional Measures group. Additional components were placed in the No Additional Measures group as a result of a failure modes, effects, and criticality analysis (FMECA) and the functionality assessment. No further action is required by these guidelines for Smanaging the aging of the No Additional Measures components.
The categorization and analysis used in the development of MRP-227-A are not intended to supersede any ASME B&PV Code Section XI requirements. Any components that are classified as core support
- that remain in effect and may only be altered as allowed by 10 CFR 50.55a.
0 4.3.2.2 NEI 03-08 Guidance within MRP-227-A The industry program requirements of MRP-227-A are classified in accordance with the requirements of
0 Mandatory There is one Mandatory element:
- 1. Each commercial U.S. PWR unit shall develop and document a programfor management of aging of reactorinternalcomponents within thirty-six months following issuanceof MRP-227-Rev. 0 (that is, no laterthan December 31, 2011).
- Ft. Calhoun Applicability: MRP-227, Revision 0 was officially issued by the industry in
- December 2008. An AMP must therefore be developed by December 2011. OPPD developed this AMP for Ft. Calhoun to meet its license renewal commitment that pre-dates the implementation date contained in MRP-227, Revision 0.
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4-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 According to the NRC Regulatory Issue Summary (RIS) [26], FCS must submit their AMP for review and approval by the NRC in accordance with MRP-227-A [5] no later than October 1, 2012.
' Needed There are four Needed elements:
- 1. Each commercial U.S. PWR unit shall implement MRP-227-A, Tables 4-1 through 4-9 and Tables 5-1 through 5-3for the applicabledesign within twenty-four months following issuance of MRP-227-A.
Ft. Calhoun Applicability: MRP-227-A augmented inspections will be incorporated in the Ft.
Calhoun ISI for the license renewal period. The applicable CE tables contained in MRP-227-A are Table 4-2 (Primary), Table 4-5 (Expansion), Table 4-8 (Existing), and Table 5-2 (Examination Acceptance and Expansion Criteria) and are attached herein as Appendix C Tables C- 1, C-2, C-3, and C-4, respectively.
There are no plant specific licensing commitments that require FCS to implement an AMP according MRP-227, Revision 0 prior to December 31, 2011.
- 2. Examinationsspecified in the MRP-227-A guidelines shall be conducted in accordancewith Inspection Standard,MRP-228 [13].
Ft. Calhoun Applicability: Inspection standards will be in accordance with the requirements of MRP-228 [13]. These inspection standards will be used for augmented inspection at Ft. Calhoun as required by MRP-227-A directives.
- 3. Examination results that do not meet the examinationacceptance criteriadefined in Section 5 of the MRP-227-A guidelines shall be recordedand entered in the plant corrective actionprogram anddispositioned.
Ft. Calhoun Applicability: Ft. Calhoun will comply with this requirement.
- 4. Each commercial U.S. PWR unit shall provide a summary reportof all inspections and monitoring,items requiringevaluation, and new repairsto the MRP Program Managerwithin 120 days of the completion ofan outage during which PWR internalswithin the scope of MRP-227 are examined.
S Ft. Calhoun Applicability: As discussed in Section 4, OPPD will participate in future industry efforts and will adhere to industry directives for reporting, response, and follow-up.
4.3.2.3 GALL AMP Development Guidance It should be noted that Section XI.Ml 6A of NUREG-1801, Revision 2 [25], includes a description of the attributes that make up an acceptable AMP. These attributes are similar to the previously discussed 0
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-9 attributes of the GALL Report and are consistent with the Ft. Calhoun AMR process. Evaluation of the
- Ft. Calhoun RVI AMP against GALL Report attribute elements is provided in Section 5 of this program plan. Ft. Calhoun is committed to meeting the requirements of Revision 0 of the GALL, but will be meeting the intent of Revision 2 of the GALL.
As part of License Renewal, Ft. Calhoun agreed to participate in industry activities associated with the development of the standard Industry Guideline for Inspection and Evaluation of Reactor Internals. The industry efforts have defined the required inspections and examination techniques for those components 0critical to aging management of RVI. The results of the industry-recommended inspections, as published in MRP-227-A, serve as the basis for identifying any augmented inspections that are required to complete
- the Ft. Calhoun RVI AMP.
- 4.3.2.4 MRP-227-A Applicability to Ft. Calhoun
- The applicability of MRP-227-A to Ft. Calhoun requires compliance with the following MRP-227-A assumptions:
30 years of operation with high-leakage core loadingpatterns (freshfuel assemblies loaded in peripherallocations)followed by implementation ofa low-leakagefuel management strategyfor the remaining 30 years of operation.
- Ft. Calhoun Applicability: Ft. Calhoun fuel management program changed from a high to a low-leakage core loading pattern prior to 30 years of operation.
0 Base load operation,i.e., typically operates atfixedpower levels anddoes not usually vary power
- on a calendaror load demand schedule.
0 Ft. Calhoun Applicability: Ft. Calhoun operates as a base load unit.
0No design changes beyond those identified in general industryguidance or recommended by the originalvendors.
- Ft. Calhoun Applicability: MRP-227-A states that the recommendations are applicable to all U.S.
SPWR operating plants as of May 2007 for the three designs considered. Ft. Calhoun has not made any modifications to reactor internals components since May 2007.
Based on the Ft. Calhoun applicability, as stated, the MRP-227-A work is representative for Ft. Calhoun.
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- 4.3.3 Ongoing Industry Programs The U.S. industry, through both the EPRI/MRP and the PWROG, continues to sponsor activities related 0to RVI aging management, including planned development of a standard NRC submittal template, development of a plant-specific implementation program template for currently licensed U.S. PWR plants, and development of acceptance criteria and inspection disposition processes. OPPD will maintain cognizance of industry activities related to PWR internals inspection and aging management and will S
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4.4
SUMMARY
It should be noted that the OPPD, Ft. Calhoun, MRP, and PWROG approaches to aging management are based on the GALL Report approach to aging management strategies. This approach includes a determination of which reactor internals passive components are most susceptible to the aging mechanisms of concern followed by determination of the proper inspection or mitigating program to provide reasonable assurance that the component will continue to perform its intended function through the period of extended operation. The GALL Report-based approach was used at Ft. Calhoun for the initial basis of the LRA that resulted in the NRC SER in NUREG-1782 [2].
The approach used to develop Ft. Calhoun AMPs is fully compliant with regulatory directives and 0 approved documents. The additional evaluations and analyses completed by the MRP industry group have 0 provided clarification on the level of inspection quality needed to determine the proper examination 0 method and frequencies. The tables provided in MRP-227-A and included as Appendix C provide the level of examination required for each of the components evaluated. 0 0
It is the Ft. Calhoun position that use of the AMR produced by the LRA methodology, combined with any 0 additional augmented inspections required by the MRP-227-A industry tables provided in Appendix C, provides reasonable assurance that the reactor internals passive components will continue to perform their 0 intended functions through the period of extended operation. 0 0
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- PROGRAM ATTRIBUTES 0
The Ft. Calhoun RVI AMP is credited for aging management of RVI components for the following eight 0aging 0 degradation mechanisms and their associated effects:
- Irradiation-assisted stress corrosion cracking 0 Wear (loss of material)
- Fatigue (cracking) 0 Thermal aging embrittlement (reduction in fracture toughness)
- Irradiation embrittlement (reduction in fracture toughness) 9 Void swelling and irradiation growth (distortion) 0 Thermal and irradiation-enhanced stress relaxation or irradiation-enhanced creep (loss of preload, or loss of mechanical closure integrity)
- Report), Section XI.M 16A, "PWR Vessel Internals" [25] are described in this section. The GALL Report identifies 10 attributes for successful component aging management. The framework for assessing the effectiveness of the projected program is established by the use of the 10 elements of the GALL Report.
- Ft. Calhoun fully utilized the GALL Report process contained in NUREG- 1801 [17] in performing the AMR of the reactor internals in the license renewal process. Ft. Calhoun has committed to: (1) participate in industry initiatives that will generate additional data on aging mechanisms relevant to reactor internals; 0(2) revise the Reactor Vessel Internals Program as applicable to incorporate industry recommendations for augmented inspections and techniques resulting from the industry initiatives; and, (3) submit a revised Reactor Vessel Internals Program to the NRC for review and approval.
This AMP is consistent with the NUREG- 1801 process and includes consideration of the augmented inspections identified in MRP-227-A. It fully meets the requirements of the Ft. Calhoun commitments and GALL, Revision 2. Specific details of the Ft. Calhoun Reactor Internals AMP are summarized in the following subsections.
- 5.1 GALL REVISION 2 ELEMENT 1: SCOPE OF PROGRAM 0
GALL Report AMP Element Description 0"The scope of the program includes all R VI components at the Fort CalhounStation, which is
- built to a CE NSSS design. The scope of the programapplies the methodology andguidance in the most recently NRC-endorsed version ofMRP-227, which provides augmented inspection and flaw evaluation methodologyfor assuring thefunctional integrity of safety-relatedinternals in 0commercial operating U.S. PWR nuclearpower plants designed by B& W, CE, and Westinghouse.
The scope of components consideredfor inspection under MRP-227 guidance includes core support structures (typically denoted as Examination CategoryB-N-3 by the ASME Code,Section XI), those R VI components that serve an intended license renewalsafety function pursuant to
- criteriain 10 CFR 54.4(a)(1), and other R VI components whose failure couldprevent satisfactory 0
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0 5-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 accomplishmentof any of thefunctions identified in 10 CFR 54.4(a)(1)(i), (ii), or (iii). The scope of the program does not include consumable items, such asfuel assemblies, reactivity control assemblies, and nuclear instrumentation,because these components are not typically within the scope of the components that are requiredto be subject to an aging management review (AMR),
as defined by the criteriaset in 10 CFR 54.21(a)(1). The scope of the programalso does not include welded attachments to the internalsurface of the reactorvessel because these components are consideredto be ASME Code Class 1 appurtenancesto the reactor vessel and are adequately managed in accordancewith an applicant'sAMP that correspondsto GALL AMP XI.M1, "ASME Code,Section XI Inservice Inspection, Subsections IWB, IWC, and IWD. 0 The scope of the program includes the response bases to applicable license renewal applicant action items (LRAAls) on the MiRP-227 methodology, and any additionalprograms,actions, or activities that are discussed in these LRAAI responses and creditedfor aging management of the applicant'sRVI components. The LRAAIs are identified in the staff's safety evaluation on MRP-227 and include applicable action items on meeting those assumptions thatformed the basis of the MRP 's augmentedinspection andflaw evaluation methodology (as discussed in Section 2.4 of MRP-22 7), and NSSS vendor-specific or plant-specific LRAAIs as well. The responses to the LRAAIs on MRP-22 7 areprovided in Appendix C of the LRA.
S The guidance ofAMRP-227 specifies applicabilitylimitations to base-loadedplantsand thefuel loading management assumptions upon which thefunctionality analyses were based. These limitations and assumptions requirea determinationof applicability by the applicantfor each reactor and are covered in Section 2.4 of MRP-227" [25].
Ft. Calhoun Program Scope 0
The Ft. Calhoun RVI consist of three basic assemblies: (1) an upper guide structure, (2) a core support barrel assembly, and (3) a lower support structure. Additional RVI details are provided in Section 2 of this AMP and in applicable sections of the Ft. Calhoun LRA [9].
0 The Ft. Calhoun RVI subcomponents that required an AMR are indicated in Table 3.1-1, 3.1-2, and 3.1-3 in the Ft. Calhoun LRA [9]. The portion of these tables associated with the internals is included as part of the tables in Appendix B. The tables in the LRA list the subcomponents of the RVI that required an AMR along with each subcomponent passive function(s) and reference(s) to the corresponding AMR table(s) in the Ft. Calhoun LRA.
The Ft. Calhoun Reactor Internals AMR was conducted and documented in the Ft. Calhoun AMR [1].0 The table summarizing the results of that review is also included in the tables of Appendix B. The tables 0 identify those aging effects that require management for those components requiring AMR. A column in the tables lists the program/activity that is credited to address the component and aging effect during the period of extended operation. Ft. Calhoun submitted the information to the NRC for review [9] and received approval in NUREG-1782 [2].
0 The results of the industry research provided by MRP-227-A, summarized in the tables of Appendix C, provide the basis for the required augmented inspections, inspection techniques to permit detection and characterizing of the aging effects (cracks, loss of material, loss of preload, etc.) of interest, prescribed 0
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0 0 SWESTINGHOUSE NON-PROPRIETARY CLASS 3 5-3 frequency of inspection, and examination acceptance criteria. The information provided in MRP-227-A is
- rooted in the GALL methodology. The basic assumptions of MRP-227-A, Section 2.4 are met by FCS and are addressed in subsection 4.3.2.4 of this AMP. The Topical Report Conditions and Applicant/Licensee Action Items provided by the NRC in the SE on MRP-227, Revision 0 [5] are met by
- FCS and demonstration of compliance is addressed in Section 6.1 of this AMP for the Topical Report
- Conditions and in Section 6.2 for the Applicant/Licensee Action Items. The Ft. Calhoun RVI AMP scope is additionally based on previously established and approved GALL Report approaches through application of supporting methodologies to determine those components that require aging management.
0 Conclusion This element complies with the corresponding aging management attribute in NUREG- 180 1,
- Section XI.M16A [25] and Commitments 16, 17, and 18 in the Ft. Calhoun SER.
- 5.2 GALL REVISION 2 ELEMENT 2: PREVENTIVE ACTIONS
- GALL Report AMP Element Description The guidance in MRP-22 7 relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive aging mechanisms (e.g., loss of materialinduced by 0general, pitting corrosion,crevice corrosion,or stress corrosion crackingor any of itsforms
[SCC, PWSCC, or IASCC]). Reactor coolant water chemistry is monitoredand maintainedin accordancewith the Water Chemistry Program. The programdescription, evaluation, and technicalbasis of water chemistry arepresentedin GALL AMP XI.M2, "Water Chemistry" [25].
- Ft. Calhoun Preventive Action The Ft. Calhoun reactor internals AMP includes the Primary Water Chemistry Program [18] as an existing program that complies with the requirements of this element. A description and applicability to the Ft. Calhoun reactor internals AMP is provided in the following subsection.
0 Primary Water Chemistry Program 0To mitigate aging effects on component surfaces that are exposed to water as process fluid, chemistry programs are used to control water chemistry for impurities (e.g., dissolved oxygen, chloride, fluoride, and sulfate) that accelerate corrosion. This program relies on monitoring and control of water chemistry to keep peak levels of various contaminants below the system-specific limits. The Ft. Calhoun Chemistry
- Guidelines.
This program is consistent with the corresponding program described in the GALL Report [ 17].
0 0The limits of known detrimental contaminants imposed by the chemistry monitoring program are consistent with the EPRI PWR Primary Water Chemistry Guidelines [18].
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This element complies with the corresponding aging management attribute in NUREG-1801,Section XI.MI6A [25] and Commitment 16 in the Ft. Calhoun SER.
5.3 GALL REVISION 2 ELEMENT 3: PARAMETERS MONITORED OR INSPECTED GALL Report AMP Element Description "The programmanages thefollowing age-relateddegradationeffects and mechanisms that are applicablein general to the R VI components at the facility: (a) cracking induced by SCC, PWSCC, IASCC, orfatigue/cyclical loading; (b) loss ofmaterial induced by wear, (c) loss of fracture toughness induced by either thermal agingor neutron irradiationembrittlement; (d) changes in dimension due to void swelling and irradiationgrowth, distortion,or deflection; and (e)loss ofpreload caused by thermal and irradiation-enhancedstress relaxation or creep. For the management ofcracking, the program monitors the evidence of surface breakinglinear discontinuitiesif a visual inspection technique is used as the non-destruction examination (NDE) method, orfor relevantflaw presentationsignals if a volumetric UT method is used as the NDE method For the management of loss of material,the program monitorsfor gross or abnormal surface conditions that may be indicative of loss of materialoccurring in the components. For 0 the management of loss ofpreload,the program monitorsfor gross surface conditions that may be indicative of loosening in applicablebolted,fastened, keyed, orpinned connections. The program does not directly monitorfor loss offracture toughness that is induced by thermal aging or neutron irradiationembrittlement, or by void swelling and irradiationgrowth; instead,the impact of loss offracture toughness on component integrity is indirectlymanaged by using visual or volumetric examination techniques to monitorfor cracking in the components and by applying applicable reducedfracture toughness propertiesin theflaw evaluations if cracking is detected in the components and is extensive enough to warranta supplementalflaw growth orflaw tolerance evaluation under MRP-22 7 guidance or ASME Code,Section XI requirements. The program uses physical measurements to monitorfor any dimensionalchanges due to void swelling, irradiationgrowth, distortion,or deflection.
Specifically, the program implements the parametersmonitored/inspectedcriteriafor CE designed Primary Components in Table 4-2 of MRP-22 7. Additionally, the program implements the parametersmonitored/inspectedcriteriafor CE designedExpansion Components in Table 4-5 of MRP-227. The parametersmonitored/inspectedfor Existing Program Componentsfollow the basesfor referencedExisting programs,such as the requirementsfor ASME Code Class R VI 0 components in ASME Code,Section XI, Table IWB-2500-1, Examination CategoriesB-N-3, as implemented through the applicant'sASME Code,Section XI program,or the recommended programfor inspecting Westinghouse-designedflux thimble tubes in GALL AMP XIM3 7, "Flux Thimble Tube Inspection. " No inspections, exceptfor those specified in ASME Code, Section XW, 0 are requiredfor components that are identified as requiring "No Additional Measure," in accordancewith the analyses reportedin MRP-227" [25].
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The Ft. Calhoun AMP monitors, inspects, and/or tests for the effects of the eight aging degradation mechanisms on the intended function of the Ft. Calhoun PWR internals components through inspection and condition monitoring activities in accordance with the augmented requirements defined under
- industry directives as contained in MRP-227-A and in accordance with ASME Section XI.
This AMP implements the requirements for the Primary Component inspections from Table 4-2 of MRP-
- 227-A (included in Appendix C of this AMP as Table C-1), the Expansion Component inspections from
- Component inspections from Table 4-8 of MRP-227-A (included in Appendix C of this AMP as Table C-3). These tables contain requirements for monitoring and inspecting the RVI through the period of extended operation to address the effects of the eight aging degradation mechanisms.
For license renewal, the ASME Section XI Program consists of periodic volumetric, surface, and/or visual examination of components for assessment, signs of degradation, and corrective actions. The 0requirements of MRP-227-A only augment and do not replace or modify the requirements of ASME Section XI. This program is consistent with the corresponding program described in the GALL Report
- [17].
0Appendices B and C of this AMP provide a detailed listing of the components and subcomponents and the parameters monitored, inspected, and/or tested.
S Conclusion 0This element complies with or exceeds the corresponding aging management attribute in NUREG- 1801,Section XI.M16A [25] and Commitment 16 in the Ft. Calhoun SER.
- 5.4 GALL REVISION 2 ELEMENT 4: DETECTION OF AGING EFFECTS
- GALL Report AMP Element Description
- The detection of aging effects is covered in two places: (a) the guidance in Section 4 of MRP-22 7 0 provides an introductorydiscussion andjustificationof the examination methods selectedfor detecting the agingeffects of interest; and (b) standardsfor examination methods, procedures, andpersonnel areprovided in a companion document, MRP-228. In all cases, well-established methods were selected. These methods include volumetric UT examination methods for detecting 0 flaws in bolting,physical measurementsfor detecting changes in dimension, and various visual (VT-3, VT-1, andEVT-1) examinationsfor detecting effects rangingfrom general conditions to detection and sizing of surface-breakingdiscontinuities. Surface examinations may also be used as an alternative to visual examinationsfor detection and sizing of surface-breaking 0discontinuities.
0 Cracking caused by SCC, IASCC, andfatigue is monitored/inspectedby either VT-I or EVT-1 examination (for internalsother than bolting) or by volumetric UT examination (bolting). The 0 VT-3 visual methods may be appliedfor the detection of cracking only when theflaw tolerance of 0
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5-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 the component or affected assembly, as evaluatedfor reducedfracturetoughnessproperties,is known and has been shown to be tolerantof easily detected largeflaws, even under reduced fracture toughness conditions. In addition, VT-3 examinations are used to monitor/inspectfor loss ofmaterialinduced by wear andfor generalaging conditions, such as gross distortion caused by void swelling and irradiationgrowth or by gross effects of loss ofpreloadcaused by thermal and irradiation-enhancedstress relaxation and creep.
In addition, the program adopts the recommended guidance in MRP-227 for defining the Expansion criteriathat needed to be appliedto inspections of Primary Components and Existing Requirement Components andfor expanding the examinations to include additionalExpansion Components. As a result, inspectionsperformed on the R VI components areperformed consistent with the inspectionfrequency and samplingbasesfor Primary Components, Existing Requirement Components, and Expansion Components in MRP-227, which have been demonstrated to be in conformance with the inspectioncriteria,sampling basis criteria,and sample Expansion criteriain Section A. 1.2.3.4 of NRC Branch PositionRLSB-1.
Specifically, the program implements the parametersmonitored/inspectedcriteriaand basesfor inspecting the relevantparameterconditionsfor CE designedPrimary Components in Table 4-2 of MRP-227 andfor CE designedexpansion components in Table 4-5 ofMRP-227.
The program is supplemented by thefollowing plant-specificPrimary Component and Expansion Component inspectionsfor the program (as applicable): for FCS, no additionalPrimaryor Expansion components are relevant to the scope of agingmanagementfor the RVI.
In addition, in some cases (as defined in MRP-22 7), physical measurements are used as supplemental techniques to managefor the gross effects of wear, loss ofpreload due to stress relaxation, orfor changes in dimension due to void swelling, deflection or distortion. The physical measurements methods appliedin accordancewith this program include no specific component at FCS. Therefore, FCS will not be requiredto perform physical measurements of 0 any components per MRP-227-A [25]. 0 Ft. Calhoun Detection of Aging Effects Detection of indications that are required by the ASME Code Section XI ISI Program is well established and field-proven through the application of the Section XI ISI Program. Those augmented -inspections that are taken from the MRP-227-A recommendations will be applied through use of the MRP-228 Inspection Standard. This AMP implements the augmented inspection requirements of Table 4-2, Table 4-5, and Table 4-8 from MRP-227-A for the Primary, Expansion, and Existing Components, respectively. These 0 are included in Appendix C of this AMP for reference. These tables include the inspection frequency and sampling basis. For the Expansion Components of MRP-227-A, this AMP implements the expansion requirements of Table 5-2 of MRP-227-A (included in Appendix C of this AMP as Table C-4).
Inspection can be used to detect physical effects of degradation including cracking, fracture, wear, and distortion. The choice of an inspection technique depends on the nature and extent of the expected damage. The recommendations supporting aging management for the reactor internals, as contained in this report, are built around three basic inspection techniques: (1) visual, (2) ultrasonic, and (3) physical S
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-7 measurement. The visual techniques include VT-3, VT-1, and EVT- 1 (enhanced visual test). The
- assumptions and process used to select the appropriate inspection technique are described in the following subsections. Inspection standards developed by the industry for the application of these techniques for augmented reactor internals inspections are documented in MRP-228.
- VT-I Visual Examinations The acceptance criteria for visual examinations conducted under categories B-N-2 (welded core support 0structures and interior attachments to reactor vessels) and B-N-3 (removable core support structures) are
- defined in IWB-3520 [19]. VT-i visual examination is intended to identify crack-like surface flaws.
0Unacceptable conditions for a VT-I examination are:
- Crack-like surface flaws on the welds joining the attachment to the vessel wall that exceed the
- allowable linear flaw standards of IWB-3510 0
- Structural degradation of attachment welds such that the original cross-sectional area is reduced by more than 10 percent
- These requirements are defined to ensure the integrity of attachment welds on the ferritic pressure vessel.
Although the IWB-3520 criteria do not directly apply to austenitic stainless steel internals, the clear intent is to ensure that the structure will meet minimum flaw tolerance fracture requirements. In the MRP-227-0A recommendations, VT-I examinations have been identified for components requiring close visual examinations with some estimate of the scale of deformation or wear. In MRP-227-A, note that VT- I has only been selected to detect distortion as evidenced by small gaps between the upper-to-lower mating surfaces of CE-welded core shrouds assembled in two vertical sections. Therefore, no additional VT-I 0inspections over and above those required by ASME Section XI ISI have been specified.
EVT- 1 Enhanced Visual Examination for the Detection of Surface Breaking Flaws 0In the augmented inspections detailed in MRP-227-A for reactor internals, the EVT- 1 enhanced visual examination has been identified for inspection of components where surface-breaking flaws are a potential concern. Any visual inspection for cracking requires a reasonable expectation that the flaw length and crack mouth opening displacement meet the resolution requirements of the observation 0technique. The EVT- 1 specification augments the VT-I requirements to provide more rigorous inspection standards for SCC and has been demonstrated for similar inspections in boiling water reactor (BWR)
- internals. Enhanced visual examination (i.e., EVT- 1) is also conducted in accordance with the requirements described for visual examination (i.e., VT-I) with additional requirements (such as camera scanning speed) currently being developed by the industry. Any recommendation for EVT- 1 inspection 0will require additional analysis to establish flaw-tolerance criteria, which must take into account potential embrittlement due to thermal aging or neutron irradiation. The industry, through the PWROG, has developed an approach for acceptance criteria methodologies to support plant-specific augmented examinations. This work is summarized in WCAP- 17096-NP, "Reactor Internals Acceptance Criteria 0Methodology and Data Requirements" [14]. The acceptance criteria developed using these methodologies may be created on either a generic or plant-specific basis because both loads and component dimensions
- may vary from plant-to-plant within a typical PWR design.
0 VT-3 Examination for General Condition Monitoring 0
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In the augmented inspections detailed in MRP-227-A for reactor internals, the VT-3 visual examination has been identified for inspection of components where general condition monitoring is required. The VT-3 examination is intended to identify individual components with significant levels of existing degradation. As the VT-3 examination is not intended to detect the early stages of component cracking or other incipient degradation effects, it should not be used when failure of an individual component could threaten either plant safety or operational stability. The VT-3 examination may be appropriate for inspecting highly redundant components, where a single failure does not compromise the function or integrity of the critical assembly.
The acceptance criteria for visual examinations conducted under categories B-N-2 (welded core support structures and interior attachments to reactor vessels) and B-N-3 (removable core support structures) are defined in 1WB-3520. These criteria are designed to provide general guidelines. The unacceptable conditions for a VT-3 examination are:
Structural distortion or displacement of parts to the extent that component function may be impaired
- Loose, missing, cracked, or fractured parts, bolting, or fasteners 0 Foreign materials or accumulation of corrosion products that could interfere with control rod motion or could result in blockage of coolant flow through fuel
- Corrosion or erosion that reduces the nominal section thickness by more than five percent
- Wear of mating surfaces that may lead to loss of function S
- Structural degradation of interior attachments such that the original cross-sectional area is reduced more than five percent The VT-3 examination is intended for use in situations where the degradation is readily observable. It is meant to provide an indication of condition, and quantitative acceptance criteria are not generally required. In any particular recommendation for VT-3 visual examination, it should be possible to identify the specific conditions of concern. For instance, the unacceptable conditions for wear indicate wear that might lead to loss of function. Guidelines for wear in a critical-alignment component may be very different from the guidelines for wear in a large structural component.
Ultrasonic Testing0 0
Volumetric examinations in the form of ultrasonic testing (UT) techniques can be used to identify and determine the length and depth of a crack in a component. Although access to the surface of the component is required to apply the ultrasonic signals, the flaw may exist in the bulk of the material. In this proposed strategy, UT inspections have been recommended exclusively for detection of flaws in bolts. For the bolt inspections, any bolt with a detected flaw should be assumned to have failed. The size of the flaw in the bolt is not critical because crack growth rates are generally high, and it is assumed that the observed flaw will result in failure prior to the next inspection opportunity. It has generally been observed 0
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0 0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-9 0through examination performance demonstrations that UT can reliably (90 percent or greater reliability) detect flaws that reduce the cross-sectional area of a bolt by 35 percent [ 13].
0Failure of a single bolt does not compromise the function of the entire assembly. Bolting systems in the 0reactor internals are highly redundant. For any system of bolts, it is possible to demonstrate multiple minimum acceptable bolting patterns. The evaluation program must demonstrate that the remaining bolts meet the requirements for a minimum bolting pattern for continued operation. The evaluation procedures must also demonstrate that the pattern of remaining bolts contains sufficient margin such that continuation 0of the bolt failure rate will not result in failure of the system to meet the requirements for minimum acceptable bolting pattern before the next inspection.
Establishment of the minimum acceptable bolting pattern for any system of bolts requires analysis to 0demonstrate that the system will maintain reliability and integrity in continuing to perform the intended function of the component. This analysis is highly plant-specific, and it is recommended that, prior to UT inspection of bolts, a minimum acceptable bolting pattern be established to support continued operation.
0Physical Measurement Examination 0
Continued functionality can be confirmed by physical measurements to evaluate the impact caused by various degradation mechanisms such as wear or loss of functionality as a result of loss of preload or 0material deformation. For FCS, direct physical measurements are not required for any components in the RVI.
Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG-l1801,
- Section XI.M16A [25] and Commitment 16 in the Ft. Calhoun SER.
0*5.5 GALL REVISION 2 ELEMENT 5: MONITORING AND TRENDING
- GALL Report AMP Element Description 0 The methods for monitoring,recording, evaluating,and trending the data that resultfrom the S program'sinspections are given in Section 6 ofMRP-227 and its subsections. The evaluation methods include recommendationsforflaw depth sizing andfor crack growth determinationsas well forperforming applicable limit load, linear elastic and elastic-plasticfracture analyses of relevantflaw indications. The examinations and re-examinationsrequiredby the MRP-22 7
- guidance, together with the requirements specifiedin MRP-228for inspection methodologies, 0inspection procedures, and inspectionpersonnel,provide timely detection, reporting,and corrective actions with respect to the effects of the age-relateddegradationmechanisms within the scope of the program. The extent of the examinations, beginning with the sample of 0susceptible PWR internalscomponent locations identified as Primary Component locations, with S the potentialfor inclusion of Expansion Component locations if the effects aregreaterthan anticipated,plus the continuation of the Existing Programsactivities,such as the ASME Code, Section XM, Examination CategoryB-N-3 examinationsfor core support structures,provides a
- high degree ofconfidence in the total program [25].
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5-10 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Ft. Calhoun Monitoring and Trending Operating experience with PWR reactor internals has been generally proactive. The extremely low frequency of failure in reactor internals makes monitoring and trending based on operating experience somewhat impractical. The majority of the materials aging degradation models used to develop the MRP-227-A guidelines are based on test data from reactor internals components removed from service. The data are used to identify trends in materials degradation and forecast potential component degradation.
The industry continues to share both material test data and operating experience through the auspices of the MRP and PWROG. OPPD has in the past and will continue to maintain cognizance of industry activities and shared information related to PWR internals inspection and aging management as demonstrated in the Corrective Action Program [19], [27], [28]3 [29], and [30].
Inspections credited in Appendix B are based on utilizing the Ft. Calhoun 10-year ISI program and the augmented inspections derived from MRP-227-A and repeated here in Appendix C. The MRP-227-A inspections only augment and do not replace the existing ASME Section XI ISI requirements. These inspections, where practical, are scheduled to be conducted in conjunction with typical 10-year ISI examinations.
S Appendix C, Tables C-1, C-2, and C-3 identify the augmented Primary and Expansion inspection and monitoring recommendations, and the Existing programs credited for inspection and aging management.
As discussed in MRP-227-A, inspection of the "Primary" components provides reasonable assurance for demonstrating component current capacity to perform the intended functions. Table C-4 in Appendix C identifies the MRP-227-A expansion criteria from the Primary components. If these expansion criteria are met for a component, the associated Expansion component is to be inspected to manage the aging degradation.
0 Reporting requirements are included as part of the MRP-227-A guidelines. Consistent reporting of inspection results across all PWR designs will enable the industry to monitor reactor internals degradation on an ongoing industry basis as the period of extended operation moves forward. Reporting of examination results will allow the industry to monitor and trend results and take appropriate preemptive action through update of the MRP guidelines.
Conclusion 0
This element complies with or exceeds the corresponding aging management attribute in NUREG- 1801,Section XI.M16A [25] and Commitment 16 in the Ft. Calhoun SER.
5.6 GALL REVISION 2 ELEMENT 6: ACCEPTANCE CRITERIA 0
GALL Report AMP Element Description "Section 5 of MRP-227 provides specific examination acceptance criteriafor the Primary and Expansion Component examinations. For components addressedby examinations referenced to ASME Code,Section XI, the IWB-3500 acceptance criteria apply. For other components covered by Existing Programs, the examination acceptance criteria are described within the Existing Program reference document.
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-11 WThe guidance in MRP-22 7 contains three types of examination acceptancecriteria:
- For visual examination (and surface examination as an alternative to visual examination), the examination acceptance criterion is the absence of any of the specific, 0descriptive relevant conditions; in addition, there are requirements to record and disposition surface breaking indications that are detected and sized for length by VT-1I/EVT-1 examinations, 0 For volumetric examination, the examination acceptance criterion is the capabilityfor reliable detection of indications in bolting, as demonstratedin the examination Technical Justification; in addition, there are requirementsfor system-level assessment of bolted or pinned assemblies with unacceptable volumetric (UT) examination indications that 0 exceed specifiedlimits, and
- For physical measurements, the examination acceptance criterion for the acceptable tolerance in the measured differential height from the top of the plenum rib pads to the 0vessel seatingsurface in B& Wplants are given in Table 5-1 of MRP-22 7. The acceptance criterionfor physical measurementsperformed on the height limits of the Westinghouse-designed hold-down springs are not applicable, as Fort Calhoun is a CE-designed plant." [25].
Ft. Calhoun Acceptance Criteria Those recordable indications that are the result of inspections required by the existing Ft. Calhoun ISI 0program scope are evaluated in accordance with the applicable requirements of the ASME Code through
- the existing Corrective Action Program [19].
0 Inspection acceptance and expansion criteria are provided in Appendix C, Table C-4. These criteria will 0be reviewed periodically as the industry continues to develop and refine the information and will be 0included in updates to Ft. Calhoun procedures to enable the examiner to identify examination acceptance criteria considering state-of-the-art information and techniques.
0Augmented inspections, as defined by the MRP-227-A requirements included in this AMP as Appendix 0C, Table C- 1, Table C-2, and Table C-3, that result in recordable relevant conditions will be entered into the plant Corrective Action Program and addressed by appropriate actions that may include enhanced inspection, repair, replacement, mitigation actions, or analytical evaluations.
0The industry, through various cooperative efforts, is working to construct a consensus set of tools in line 0with accepted and proven methodologies to support this element One of these tools is the PWROG document WCAP- 17096-NP, "Reactor Intermals Acceptance Criteria Methodology and Data Requirements" [ 14], which details acceptance criteria methodology for the MRP-227, Revision 0 Primary 0and Expansion components. WCAP- 17096-NP is currently under revision to incorporate the changes
- made in MRP-227-A. Status is monitored through direct OPPD participation in industry (including 0 PWROG) activities related to PWR intemals inspection and aging management.
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5-12 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG- 1801,Section XI.M I6A [25] and Commitment 16 in the Ft. Calhoun SER.
5.7 GALL REVISION 2 ELEMENT 7: CORRECTIVE ACTIONS GALL Report AMP Element Description "Corrective actions following the detection of unacceptable conditions are fundamentally provided for in each plant's corrective action program. Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the plant corrective action program, which may require repair, replacement, or analytical evaluationfor continuedservice until the next inspection. The disposition will ensure that design basisfunctions of the reactor internals components will continue to be fulflled for all licensing basis loads and events. Examples of methodologies that can be used to analytically disposition unacceptable conditions are found in the ASME Code,Section XI or in Section 6 of MRP-22 7. Section 6 of MRP-227 describes the options that are available for disposition of detected conditions that exceed the examination acceptance criteriaof Section 5 of the report. These include engineering evaluation methods, as well as supplementary examinations to further characterize the detected condition, or the alternative of component repair and replacement procedures. The latter are subject to the requirements of the ASME Code, Section Xl. The implementation of the guidance in MRP-227, plus the implementation of any ASME Code requirements, provides an acceptable level of aging management of safety-related components addressed in accordance with the corrective actions of 10 CFR Part50, Appendix B or its equivalent, as applicable.
Other alternativecorrective action bases may be used to disposition relevant conditions if they have been previously approved or endorsed by the NRC. Examples ofpreviously NRC-endorsed alternative corrective actions bases include those corrective actions basesfor Westinghouse-design RVI components that are defined in Tables 4-1, 4-2, 4-3, 4-4, 4-5, 4-6, 4-7 and 4-8 of Westinghouse Report No. WCAP-14577-Rev. 1-A, orfor B& W-designed R VI components in B& W Report No. BA W-2248. Westinghouse Report No. WCAP-14577-Rev. 1-A was endorsedfor use in an NRC SE to the Westinghouse Owners Group, datedFebruary10, 2001. B&WReport No..
BA W-2248 was endorsedfor use in an SE to FramatomeTechnologies on behalfof the B& W Owners Group, dated December 9, 1999. Alternative corrective action bases not approved or endorsed by the NRC will be submittedfor NRC approvalprior to their implementation" [25].
Ft. Calhoun Corrective Actions 0
Corrective actions are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B and the Ft. Calhoun quality assurance program [19]. Controls are established to assure that conditions adverse to quality are identified and documented and that appropriate remedial action is taken. For significant conditions adverse to quality, necessary corrective action is promptly determined and recorded. Corrective action includes determining the cause and extent of the condition, and taking appropriate action to preclude similar problems in the future. The controls also assure that corrective action is implemented in a timely manner.
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-13 0Corrective actions are implemented through the initiation of an Action Request in accordance with plant procedures [19]. Equipment deficiencies may be initially documented by a work order, but the corrective action process specifies that an Action Request also be initiated if required. This approach ensures that identified problems are corrected in a timely manner.
0 Conclusion This element complies with the corresponding aging management attribute in NUREG- 1801,
- Section XI.MI6A [25] and Commitment 16 in the Ft. Calhoun SER.
- 5.8 GALL REVISION 2 ELEMENT 8: CONFIRMATION PROCESS
- GALL Report AMP Element Description "Site quality assuranceprocedures,review and approvalprocesses, andadministrativecontrols are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B, or their
- equivalent, as applicable. It is expected that the implementation of the guidancein MRP-227 will provide an acceptable level of qualityfor inspection,flaw evaluation, andother elements of
- aging management of the PWR internalsthat are addressedin accordance with thel.O CFR Part 50, Appendix B, or their equivalent (as applicable),confirmationprocess, and administrative 0*controls" [25].
Ft. Calhoun Confirmation Process
- Ft. Calhoun has an established 10 CFR Part 50, Appendix B Program [19] that addresses the elements of Scorrective actions, confirmation process, and administrative controls. The Ft. Calhoun Program includes non-safety-related structures, systems, and components. Quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of
- 10 CFR 50, Appendix B.
Conclusion 0This element complies with or exceeds the corresponding aging management attribute in NUREG- 1801,
- Section XI.Ml6A [25] and Commitment 16 in the Ft. Calhoun SER.
0 5.9 GALL REVISION 2 ELEMENT 9: ADMINISTRATIVE CONTROLS
- GALL Report AMP Element Description 0
"The administrativecontrolsfor such programs, including their implementingproceduresand review and approvalprocesses, are under existing site 10 CFR 50 Appendix B Quality Assurance 0Programs, or their equivalent, as applicable.Such a program is thus expected to be established with a sufficient level of documentation and administrativecontrols to ensure effective long-term
- implementation" [25].
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5-14 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Ft. Calhoun Administrative Controls Ft. Calhoun has an established 10 CFR Part 50, Appendix B program [ 19] that addresses the elements of corrective actions, confirmation process, and administrative controls. The Ft. Calhoun program includes non-safety-related structures, systems, and components. Quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B.
Conclusion 0
This element complies with or exceeds the corresponding aging management attribute in NUREG- 1801,Section XI.M16A [25] and Commitment 16 in the Ft. Calhoun SER.
5.10 GALL REVISION 2 ELEMENT 10: OPERATING EXPERIENCE GALL Report AMP Element Description "Relativelyfew incidents ofPWR internals agingdegradationhave been reportedin operating U.S. commercialPWR plants.A summary of observationsto date is provided in Appendix A of MRP-227-A. The applicant is expected to review subsequent operatingexperiencefor impact on its program or to participatein industry initiativesthatperform thisfunction.
0 The applicationof the MRP-227 guidance will establish a considerable amount of operating experience over the next few years. Section 7 of MRP-22 7 describes the reportingrequirements for these applications,and the planfor evaluating the accumulatedadditionaloperating experience" [25].
Ft. Calhoun Operating Experience Extensive industry and Ft. Calhoun operating experience has been reviewed during the development of the RVI AMP. The experience reviewed includes NRC Information Notices 84-18, "Stress Corrosion Cracking in Pressurized Water Reactor Systems" [21] and 98-11, "Cracking of Reactor Vessel Internal Baffle Former Bolts in Foreign Plants" [22]. To date, no cracking has been discovered in bolting for CE-designed reactor vessel internals. In 1998, the CEOG performed an assessment of the cracking of the baffle former bolts reported in foreign PWRs, including the potential impact of the cracking on domestic CE plants. The CEOG report, NPSD-1098, "Evaluation of the Applicability of Baffle Bolt Cracking to Ft.
Calhoun and Palisades Internals Bolts" [23], states that the most likely mechanism for the cracking of cold-worked 316 stainless steel baffle former bolts in foreign plants is IASCC. The prime factors 0 contributing to IASCC susceptibility (as experienced by the bolts in foreign plants) are not applicable to the bolts at Ft. Calhoun Station [8].
Early plant operating experience related to hot functional testing and reactor internals is documented in plant historical records. Inspections performed as part of the 10-year ISI program have been conducted as designated by existing commitments and would be expected to discover overall general internals structure degradation. To date, very little degradation has been observed industry-wide.
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-15 Industry operating experience is routinely reviewed by OPPD engineers using Institute of Nuclear Power Operations (INPO) Operating Experience (OE), the Nuclear Network, and other information sources as 0 directed under the applicable procedure, for the determination of additional actions and lessons learned.
0 These insights, as applicable, can be incorporated in the plant systems quarterly health reports and further 0 evaluated for incorporation in plant programs.
A review of industry and plant-specific experience with reactor vessel internals reveals that the 0 U.S. industry, including Ft. Calhoun, has responded proactively to industry issues relative to reactor 0 internals degradation. A key element of the MRP-227-A guidelines is the reporting of age-related 0 degradation of reactor vessel components. OPPD, through its participation in PWROG and EPRI-MRP activities, will continue to benefit from the reporting of inspection information and will share its own 0 operating experience with the industry through the reporting requirements of Section 7 of MRP-227-A.
The collected information from MRP-227-A augmented inspections will benefit the industry in its 0 continued response to RVI aging degradation.
0 Conclusion 0
0 This element complies with or exceeds the corresponding aging management attribute in NUREG- 1801,Section XI.M16A [25] and Commitments 16, 17 and 18 in the Ft. Calhoun SER.
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Ft. Calhoun has demonstrated a long-term commitment to aging management of reactor internals. This AMP is based on an established history of programs to identify and monitor potential aging degradation 0in the reactor internals. Programs and activities undertaken in the course of fulfilling that commitment
- include:
The examinations required by ASME Section XI for the Ft. Calhoun RVI have been performed as 0required 0 since plant operations commenced.
As documented in Ft. Calhoun operational procedures, Operating Experience Reports (OERs) are continuously reviewed by Ft. Calhoun personnel for applicable issues that indicate a need for updated operating procedures or programs.
Reviews of Nuclear Oversight Section (NOS) audit reports, NRC inspection reports, and INPO evaluations indicate no unacceptable issues related to reactor vessel internals inspections.
0The Chemistry Program at Ft. Calhoun has been effective in maintaining the levels of oxygen, halides, and sulfate sufficiently low to prevent SCC of the reactor vessel internals.
- OPPD has actively participated in past and ongoing EPRI and PWROG RVI activities. OPPD will continue to maintain cognizance of industry activities related to PWR internals inspection and aging management and will address/implement the industry guidance stemming from those activities as appropriate under NEI 03-08 practices.
This AMP fulfills the approved license renewal methodology requirement to identify the most susceptible components and to inspect those components using inspection techniques with the capability to detect the expected degradation mechanism indication or indications. Augmented inspections, derived from the information contained in the industry I&E Guidelines (MRP-227-A) [5], have been utilized in this AMP Sto build on existing plant programs. This approach is expected to encourage detection of a degradation 5mechanism at its first appearance, which is consistent with the ASME B&PV Code approach to inspections. This approach provides reasonable assurance that the internals components will continue to perform their intended function through the period of extended operation.
Typically, MRP-227-A augmented reactor internals examinations applicable to Ft. Calhoun must be performed no later than two refueling outages from the beginning of the license renewal period. A schedule for performing reactor vessel internals inspections at Ft. Calhoun is provided in Table 7-1. The 0augmented inspections needed for compliance with MRP-227-A will be integrated into the Reactor Vessel Internals Inspection Program. Integration of the required inspections will be tracked to completion, and according to the good practice element in MRP-227-A, Ft. Calhoun will continue to participate in future industry efforts on reactor internals and will adhere to industry directives for reporting, response, and follow-up as experience is gained through inspection results. This feedback loop will enable updates 0based on actual inspection experience.
The augmented inspections described in this document, as summarized in Appendix C, combined with the SASME Code Section XI ISI program inspections, existing Ft. Calhoun programs, and use of Operating S
S WCAP-17347-NP August 2012 Revision 1 S
0
6-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Experience Reports (OERs), provide reasonable assurance that the reactor internals at Ft. Calhoun will 0 continue to perform their intended functions through the period of extended operation. 0 Table 6-1 lists the seven topical report conditions and Section 6.2 lists the eight applicant/licensee action items that came out of the NRC review of MRP-227, Revision 0, as listed in [26], as well as their compliance within this AMP.
6.1 DEMONSTRATION OF TOPICAL REPORT CONDITIONS COMPLIANCE TO 0 SE ON MRP-227, REVISION 0 0
0 Table 6-1 Topical Report Conditions Compliance to SE on MRP-227, Revision 0 0
Topical Condition Applicable/Not Applicable Compliance in AMP 0
- 1. High consequence components in the Applicable The lower core support beams, core support 0 "No Additional Measures" Inspection Category barrel assembly upper cylinder, and upper core barrel flange are added to Table C-2 as 0
"Expansion Components" linked to the "Primary 0 Component," the upper core support barrel flange weld.
0
- 2. Inspection of components subject to Applicable The core support barrel assembly lower cylinder 0
irradiation-assisted stress corrosion girth welds are moved from Table C-2, 0 cracking "Expansion Components" to Table C-1, "Primary Components." 0
- 3. Inspection of high consequence Applicable The core support column welds are moved from 0 components subject to multiple degradation mechanisms Table C-2, "Expansion Components" to 0 Table C-1, "Primary Components."
0
- 4. Imposition of minimum examination Applicable Notes 3 through 5 were added to Table C-I, as coverage criteria for "Expansion" well as Note 2 to Table C-2 to reflect this 0 inspection category components condition. 0
- 5. Examination frequencies for baffle-former bolts and core shroud bolts Applicable In Table C- I for the core shroud bolts, the inspection frequency was changed from 10 to 15 0
additional effective full-power years (EFPY) to 0 10 additional years.
0
- 6. Periodicity of the re-examination of Applicable "Re-inspection every 10 years following initial "Expansion" inspection category inspection" was added to every component under 0
components the Examination Method/Frequency column in 0 Table C-2.
0
- 7. Updating of MRP-227, Revision 0, Applicable Section 5 is updated to reflect XI.M I6A from Appendix A GALL Revision 2. 0 0
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0 ItWESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3
- 6.2 DEMONSTRATION OF APPLICANT/LICENSEE ACTION ITEM
- t COMPLIANCE TO SE ON MRP-227, REVISION 0 0 6.2.1 SE Applicant/Licensee Action Item 1: Applicability of FMECA and Functionality
- t Analysis Assumptions As addressedin Section 3.2.5.1 of this SE, each applicant/licenseeis responsiblefor assessing its plant's design andoperatinghistory and demonstratingthat the approved version ofMRP-227 is 0t applicable to thefacility. Each applicant/licenseeshall refer, in particular,to the assumptions
- t regardingplant design and operatinghistory made in the FMECA andfunctionality analysesfor
- reactorsof their design (i.e., Westinghouse, CE, or B& W) which support MRP-227 and describe the process usedfor determiningplant-specific differences in the design of their R VI components 0 or plant operatingconditions, which result in different component inspectioncategories. The 0t applicant/licenseeshall submit this evaluationfor NRC review andapprovalas partof its application to implement the approvedversion of AMP-227. This is Applicant/Licensee Action Item 1 [5].
- t FCS Compliance The process used to provide reasonable assurance that FCS is represented by the generic industry program 0t assumptions with regard to neutron fluence, temperature, stress values, and materials used in the
- t development of MRP-227-A is:
- 1. Identification of typical CE-designed pressurized water reactor (PWR) RVI components (Table 4-
- 5 of MRP-191).
0 a. Confirmation that no additional items were identified by this comparison (primarily
- t supports A/LAI 2).
- b. Confirmation that the materials from Table 4-5 of MRP- 191 are consistent with FCS RVI component materials.
- c. Confirmation that the design and fabrication of FCS RVI components are the same as, or equivalent to, the typical CE-designed PWR RVI components.
- 4. Confirmation that the FCS operating history is consistent with the assumptions in MRP-227-A regarding core loading patterns.
- 5. Confirmation that the FCS RVI materials operated at temperatures within the original design basis 0t parameters.
0 6. Determination of stress values based on design basis documents.
- 7. Confirmation that any changes to the FCS RVI components do not impact the application of the
- t MRP-227-A generic aging management strategy.
- WCAP-17347-NP August 2012 Revision 1 0t 0t
6-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 The FCS RVI components are represented by the design and operating history assumptions regarding neutron fluence, temperature, materials, and stress values in the MRP- 191 generic FMECA and in the MRP-232 functionality analysis based on the following:
- 1. FCS operating history is consistent with the assumptions in MRP-227-A with regard to neutron fluence.
- a. The FMECA and functionality analysis for MRP-227-A are based on the assumption of 30 years of operation with high-leakage core loading patterns followed by 30 years of low-leakage core fuel management strategy. FCS switched to use of a low-leakage core design in fuel cycle 8 (04/02/83) at 10 years of operation [31]. In fuel cycle 10 (01/16/86), FCS used extreme low radial leakage fuel management, but reverted to low radial leakage fuel management for fuel cycles 11 through 13 [31]. Since fuel cycle 14 (05/01/92), FCS has used extreme low radial leakage fuel management [31 ] for all subsequent fuel cycles as stated in WCAP- 17347-P. Therefore, with the switch to low radial leakage fuel in fuel cycle 10 (0 1/16/86) [31], FCS meets the fluence and fuel management assumptions in MRP_ 191 and the requirements for MRP-227-A application.
- b. FCS has operated under base load conditions over the life of the plant, as stated in subsection 4.3.2.4. Therefore, FCS satisfies the assumptions in MRP documents regarding operational parameters affecting fluence.
- 2. The FCS RVI operate between Thor and T 0old [31 ], which are, approximately, not less than 543°F for Tcold or higher than 592°F for Thor. The design temperature for the vessel is 650'F [31 ]. FCS 0 operating history is within original design basis parameters and therefore consistent with the assumptions used to develop the MRP-227-A aging management strategy with regard to temperature operational parameters.
- 3. FCS RVI components and materials are covered by the list of generic CE-designed PWR RVI components (MRP-191, Table 4-5), as summarized in [31 ].
- a. No additional components are identified for FCS by this comparison [2].
- b. FCS RVI component materials are consistent with, or nearly equivalent to, those materials identified in Table 4-5 of MRP-191 for CE-designed plants. Where differences exist, there 0 is no impact on the FCS RVI program or the component is already credited as being managed under an alternate FCS aging management program.
- c. Design and fabrication of FCS RVI components are the same as, or equivalent to, the typical CE-designed PWR RVI components.
- 4. Modifications to the FCS RVI made over the lifetime of the plant are those specifically directed by the original equipment manufacturer (OEM') as discussed in subsection 4.3.2.4. The design has been maintained over the lifetime of the plant as specified by the OEM, operational parameters with regard to fluence and temperature are compliant with MRP-227-A requirements, and the components and materials are equivalent to those considered in MRP- 191. Therefore, the FCS RVI are represented by the assumptions in MRP-191, MRP-227-A, and MRP-232, confirming the applicability of the generic FMECA.
For the purposes of this WCAP, OEM is defined as CE and Westinghouse.
WCAP- 17347-NP August 2012 Revision 1
0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-5 Conclusion 0
FCS complies with Applicant/Licensee Action Item (A/LAI 1) of the NRC SE regarding MRP-227, Revision 0. Therefore, the requirement is met for application of MRP-227-A as a strategy 0for managing age-related material degradation in the RVI components.
0 6.2.2 SE Applicant/Licensee Action Item 2: PWR Vessel Internals Components within 0the Scope of License Renewal As discussed in Section 3.2.5.2 of this SE, consistent with the requirements addressedin 10 CFR 054.4, each applicant/licenseeis responsiblefor identifying which RVI components are within the scope of LR for itsfacility. Applicants/licenseesshall review the information in Tables 4-1 and 4-2 in MRP-189, Revision 1, and Tables 4-4 and 4-5 in MRP-191 and identify whether these tables contain all of the R VI components that are within the scope of LR for theirfacilities in accordancewith 10 CFR 54.4. If the tables do not identify all the RVI components that are within the scope of LR for itsfacility, the applicantor licensee shall identify the missing component(s) 0 andpropose any necessary modifications to the program defined in MRP-22 7, as modified by this SE, when submitting its plant-specificAMP. The AMP shallprovide assurancethat the effects of aging on the missing component(s) will be managedfor the periodof extended operation. This 0*issue is Applicant/Licensee Action Item 2 [5].
- FCS Compliance 0This action item requires comparison of the FCS RVI components that are within the scope of license 0renewal for FCS to those components contained in Table 4-5 of MRP- 191. A detailed tabulation of the FCS RVI components was completed. This tabulation compared favorably to the typical CE-designed PWR RVI components in MRP- 191. All required components in the FCS program [2] are consistent with 0those contained in MRP-191. Several components had a different material than specified in MRP-191, 0but these differences have no effect on the recommended MRP aging strategy or are already managed by an alternate FCS program; therefore, no modifications to the program details in MRP-227-A need to be proposed. This supports the requirement that the AMP shall provide assurance that the effects of aging on the FCS RVI components within the scope of license renewal, but not included in the generic CE-
- designed PWR RVI components from Table 4-5 of MRP- 191, will be managed for the period of extended operation.
- inspection sampling approach for aging management of the RVI specified in MRP-227-A are applicable to FCS with no modifications.
Conclusion 0
- FCS complies with Applicant /Licensee Action Item 2 of the NRC SE on MRP-227, Revision 0 and therefore meets the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor intemals components.
0 0
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WCAP-17347-NP August 2012 Revision 1 0
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6-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6.2.3 SE Applicant/Licensee Action Item 3: Evaluation of the Adequacy of Plant-Specific Existing Programs As addressedin Section 3.2.5.3 in this SE, applicants/licenseesof CE and Westinghouse are requiredto perform plant-specificanalysis either tojustify the acceptabilityofan applicant's/licensee's existing programs,or to identify changes to the programsthat should be implemented to manage the aging of these componentsfor the periodof extended operation. The results of this plant-specific analyses and a description of the plant-specificprogramsbeing relied on to manage agingof these components shall be submitted as part of the applicant's/licensee'sAMP application.The CE and Westinghouse components identifiedfor this type ofplant-specific evaluation include: CE thermal shieldpositioningpins and CE in-core instrumentationthimble tubes (Section 4.3.2 in MRP-22 7), and Westinghouse guide tube support pins (splitpins) (Section 4.3.3 in MIRP-227). This is Applicant/Licensee Action Item 3 [5].
0 FCS Compliance FCS is compliant with the requirements in Table 4-8 of MRP-227-A as applicable to FCS, as shown in Appendix C, Table C-3. This is detailed in the plant-specific program documents for ASME Section XI
[3, 4], the plant-specific Thermal Shield Integrity Program [33], and the plant-specific in-core instrumentation thimble tube program [31 and 34].
Conclusion 0
FCS complies with Licensee/Action Item 3 of the NRC SE on MRP-227, Revision 0 and therefore meets the requirement for application of MRP-227-A as a strategy for managing age- related material degradation in reactor internals components.
S 6.2.4 SE Applicant/Licensee Action Item 4: B&W Core Support Structure Upper Flange Stress Relief 0
As discussed in Section 3.2.5.4 of this SE, the B& W applicants/licenseesshall confirm that the core support structure upperflange weld was stress relievedduring the originalfabricationof the Reactor Pressure Vessel in order to confirm the applicabilityof MRP-22 7, as approvedby the NRC, to theirfacility. If the upperflange weld has not been stress relieved, then this component shall be inspected as a "Primary"inspection category component. If necessary, the examination methods andfrequency for non-stress relievedB& W core support structure upperflange welds shall be consistent with the recommendations in MRP-227, as approvedby the NRC,for the Westinghouse and CE upper core support barrelwelds. The examination coveragefor this B& W flange weld shall conform to the staff's imposed criteria as describedin Sections 3.3.1 and4.3.1 of this SE. The applicant's/licensee'sresolution of this plant-specific action item shall be submitted to the NRCfor review and approval. This is Applicant/Licensee Action Item 4 [5].
0 0
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0 WCAP- 17347-NP August 2012 Revision 1 0
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0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-7
- FCS Compliance 0
- This applicant/licensee action item is not applicable to FCS.
Conclusion
6.2.5 SE Applicant/Licensee Action Item 5: Application of Physical Measurements as
- part of I&E Guidelines for B&W, CE, and Westinghouse RVI Components As addressedin Section 3.3.5 in this SE, applicants/licenseesshall identify plant-specific
- acceptance criteriato be appliedwhen performing the physical measurements requiredby the
- NRC-approved version ofMRP-227for loss of compressibilityfor Westinghouse hold down springs, andfor distortion in the gap between the top and bottom core shroud segments in CE units with core barrelshrouds assembled in two verticalsections. The applicant/licenseeshall include its proposedacceptance criteriaand an explanation of how the proposedacceptance criteriaare consistent with the plants' licensing basis and the need to maintainthe functionality of the component being inspectedunder all licensingbasis conditions of operationduring the period of extended operationas part of their submittal to apply the approvedversion of MRP-227. This is Applicant/Licensee Action Item 5 [5].
0*
- FCS Compliance 0FCS 0 has full height, bolted core shroud plates; therefore, the physical measurement exam is not needed.
Conclusion Item 5 of the NRC SE on MRP-227, Revision 0 is not applicable to FCS.
0*Licensee/Action 6.2.6 SE Applicant/Licensee Action Item 6: Evaluation of Inaccessible B&W Components
- As addressedin Section 3.3.6 in this SE, MRP-227 does not propose to inspect thefollowing 0inaccessible components: the B& W core barrelcylinders (including vertical and circumferential seam welds), B& Wformer plates, B& W external baffle-to-baffle bolts and theirlocking devices, B& W core barrel-to-formerbolts and their locking devices, and B& W core barrelassembly 0internal baffle-to-baffle bolts. The MRP also identified that although the B& W core barrel assembly internalbaffle-to-baffle bolts are accessible, the bolts are non-inspectableusing currently availableexamination techniques.
0 Applicants/licenseesshalljustify the acceptability of these componentsfor continued operation through the period of extended operation by performing an evaluation, or by proposinga scheduled replacement of the components. As part of their applicationto implement the approved version of MRP-227, applicants/licenseesshallprovide theirjustificationfor the continued 0 operabilityof each of the inaccessiblecomponents and, if necessary,provide theirplanfor the 0
0 WCAP-17347-NP August 2012 Revision 1 0
0
6-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 replacement of the componentsfor NRC review and approval. This is Applicant/Licensee Action Item 6 [5].
FCS Compliance This applicant/licensee action item is not applicable to FCS.
Conclusion Licensee/Action Item 6 of the NRC SE on MRP-227, Revision 0 is not applicable to FCS.
6.2.7 SE Applicant/Licensee Action Item 7: Plant-Specific Evaluation of CASS Materials As discussed in Section 3.3.7 of this SE, the applicants/licenseesofB& W, CE, and Westinghouse reactors are required to develop plant-specific analyses to be appliedfor theirfacilities to demonstrate that B& W IMI guide tube assembly spiders and CRGT spacer castings, CE lower support columns, and Westinghouse lower support column bodies will maintain their functionality during the period of extended operationorfor additionalR VI components that may befabricatedfromCASS, martensiticstainless steel orprecipitationhardenedstainless steel materials. These analyses shall also consider the possible loss offracture toughness in these components due to thermaland irradiationembrittlement, and may also need to consider limitations on accessibilityfor inspection and the resolution/sensitivityofthe inspection techniques. The requirementmay not apply to components that were previously evaluatedas not requiringaging management during development ofMRP-227. That is, the requirementwould apply to components fabricatedfromsusceptible materialsfor which an individuallicensee has 0 determined agingmanagement is required,for example during their review performed in accordance with Applicant/LicenseeAction Item 2. The plant-specific analysis shall be consistent with the plant's licensing basis and the need to maintain thefunctionality ofthe components being evaluated under all licensing basis conditionsof operation. The applicant/licenseeshall 0 include the plant-specificanalysis as part of their submittal to apply the approvedversion of 0 MvRP-227. This is Applicant/Licensee Action Item 7 [5].
FCS Compliance 0 Applicant/Licensee Action Item 7 from the NRC's final SE on MRP-227, Revision 0 states that for assessment of CASS materials, the licensees or applicant for License Renewal may apply the criteria in the NRC letter of May 19, 2000, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Stainless Steel components" (NRC ADAMS Accession No. ML003717179) as the basis for 0 determining whether the CASS materials are susceptible to the thermal aging mechanism. If the application of the applicable screening criteria [32] for the component material demonstrates that the components are not susceptible to either thermal embrittlement or irradiation embrittlement, or the synergistic effects of thermal embrittlement (TE) and irradiation embrittlement (IE) combined, then no 0 other evaluation would be necessary.
The FCS RVI CASS components and the assessment of their susceptibility to TE are summarized in S Table 6-2. Based on the criteria of [32], the FCS CASS lower core support columns are not susceptible to 0 WCAP- 17347-NP August 2012 Revision 1
0 0
0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-9 0 TE. Conclusive confirmation of material composition under TE susceptibility thresholds was not demonstrated for the CASS CEA components; thus, it is conservatively assumed that these components are potentially susceptible to TE. Under MRP-191 the CASS CEA components were screened-in for TE, 0 SCC, and irradiation embrittlement, and dispositioned for these mechanisms under the FMECA. Thus, 0 the conservative assumption that the CASS CEA components are susceptible to TE has already been addressed in the development of the MRP-227-A inspection requirements.
0 Table 6-2 Summary of FCS CASS Components and their Susceptibility to TE 0 Susceptibility to TE Calculated Ferrite 0 CASS Component Molybdenum Content Casting Content (based on the NRC criteria [32])
0 Core/Lower S ort/Colun Low, 0.5 max Static Support Columns _<20% Not susceptible to TE 0
0 Potentially susceptible to 0 CEA Low, 0.5 max Static Not available TE 0
0 The FCS martensitic SS and PH-SS RVI components are summarized in Table 6-3 [31].
0 Table 6-3 FCS Reactor Internals: Martensitic SS and PH-SS Components S
Component Name Material Upper Internals Assembly, Martensitic SS 0 Hold-down Ring 403 SS Modified PH-SS (austenitic)
S CEA, Shroud Assembly, CEA Shroud Bolts
(
A286 A286 SS 0 Core Support Barrel Assembly, Core Barrel Snubber PH-SS (austenitic)
Lug Bolts A286 SS 0
0 0
0 Conclusion 0 The results of this CASS evaluation do not conflict with the MRP-227-A strategy for aging management 0 of RVI. It is concluded that continued application of the strategy of MRP-227-A will meet the requirement for managing age-related degradation of the FCS CASS RVI components.
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0 6-10 WESTINGHOUSE NON-PROPRIETARY CLASS 3 S 6.2.8 SE Applicant/Licensee Action Item 8: Submittal of Information for Staff Review 0
and Approval 0 0
As addressedin Section 3.5.1 in this SE, applicants/licenseesshall make a submittalfor NRC review andapprovalto credit their implementation ofMRP-22 7, as amendedby this SE, as an 6
AMP for the R VI components at theirfacility. This submittalshall include the information 0 identified in Section 3.5.1 of this SE. This is Applicant/LicenseeAction Item 8 [5]. 0 FCS Compliance 0
0 FCS will submit their AMP based on [5] for NRC review and approval. 0 0
Conclusion 0
FCS complies with Licensee/Action Item 8 of the NRC SE on MRP-227, Revision 0 and therefore meets 0 the requirement for application of MRP-227-A as a strategy for managing age-related material 0 degradation in reactor intemals components.
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0 0
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0000000000000000000000000000000000000000090o WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-1 7 PROJECTED PROGRAM ENHANCEMENT AND IMPLEMENTATION SCHEDULE The requirements of MRP-227-A are based on an 18-month refueling cycle and consider both EFPYs and cumulative operation. The information contained in Table 7-1 is based on this information and includes a description of the past inspections, as well as the latest scope of inspections pertaining to the reactor internals AMP. Should a change occur in plant operational practices or should operating experience result in changes to the projections, appropriate updates will be made to affected plant documentation in accordance with approved procedures.
Table 7-1 Aging Management Program Enhancement and Inspection Implementation Summary Refueling Project Estimated Inspection Method and Outage Month/Year EFPYt 1 AMP-Related Scope Criteria Comments RO-26 Spring 2011 24.0 Not applicable Not applicable Fuel out. Flood prevented completion of this outage.
RO-27 Fall 2014 25.5 ASME Code Section XI Not applicable Fuel out and core barrel out. Original license expires on August 9, 2013.
RO-28 Spring 2016 27.0 Initial MRP-227- A augmented MRP-227-A inspections in Fuel out and core barrel out.
inspection for core shroud accordance with MRP-228 assembly, upper (core support specifications barrel) flange weld, lower cylinder girth welds, core support column welds, lower flange weld, core support plate, fuel alignment plate, and instrument guide tubes completed during or before this outage.
RO-29 Fall 2017 28.5 Not applicable Fuel out.
RO-30 Spring 2019 30.0 Not applicable Not applicable Fuel out.
RO-31 Fall 2020 31.5 Not applicable Not applicable Fuel out.
RO-32 Spring 2022 33.0 Not applicable Not applicable Fuel out.
RO-33 Fall 2023 34.5 ASME Code Section XI MRP-227-A inspections in Fuel out.
Initial MRP-227-A augmented accordance with MRP-228 inspections for core shroud bolts specifications (FCS: panel-to-former bolts)
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7-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 7-1 Aging Management Program Enhancement and Inspection Implementation Summary (cont.)
Refueling Project Estimated Inspection Method and Outage Month/Year EFPY() AMP-Related Scope Criteria Comments RO-34 Spring 2025 36.0 Subsequent MRP-227-A augmented MRP-227-A inspections in Fuel out and core barrel out.
inspections for core shroud accordance with MRP-228 assembly, upper (core support specifications barrel) flange weld, lower cylinder girth welds, core support column welds, lower flange weld, core support plate, fuel alignment plate, and instrument guide tubes completed during or before this outage.
RO-35 Fall 2026 37.5 Not applicable Not applicable Fuel out.
RO-36 Spring 2028 39.0 Not applicable Not applicable Fuel out.
RO-37 Fall 2029 40.5 Not applicable Not applicable Fuel out.
RO-38 Spring 2031 42.0 Not applicable Not applicable Fuel out.
RO-39 Fall 2032 43.5 Subsequent MRP-227-A augmented MRP-227-A inspections in Fuel out.
inspections for core shroud bolts accordance with MRP-228 (FCS: panel-to-former bolts) specifications RO-40 Spring 2034 45.0 Not applicable Not applicable License expires August 9, 2033.
Note:
- 1. EFPY values estimated based on an assumption of 100 percent reliability.
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0 0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-1 8 IMPLEMENTING DOCUMENTS 0 As noted within this Ft. Calhoun AMP document, the PWR Vessel Internals Program is a part of the Ft.
Calhoun Nuclear Plant aging management program. The Ft. Calhoun RVI AMP also references the 0 Reactor Vessel Internals Inspection Program, Chemistry Program and the ASME Code Section XI 0 Inservice Inspection, Subsections IWB, IWC, IWD, and IWF Program. MRP-227-A augmented examinations (Appendix C), recommended as a result of industry programs, will be included in the 0 existing Reactor Vessel Internals Inspection Program [8].
0 Ft. Calhoun documents associated with the existing Ft. Calhoun programs and considered to be implementing documents of the PWR Vessel Internals Program are:
0
- Reactor Vessel Internals Inspection Program [I]
0
- Chemistry Program Implementation [18]
- Inservice Inspection Programs [3] and [4]
0 The PWR Vessel Internals AMP relies on the Chemistry Program for maintaining high water purity to 0 reduce susceptibility to cracking due to SCC. The Chemistry Program was evaluated and found to be consistent with the GALL Report [2] and [18]. Additional procedures may be updated or created as OE 0 for augmented examinations is accumulated.
S Based on this information, the updated AMP for Ft. Calhoun RVI provides reasonable assurance that the aging effects will be managed such that the components within the scope of license renewal will continue 0 to perform their intended functions consistent with the CLB for the period of extended operation.
0 0
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0 SWESTINGHOUSE NON-PROPRIETARY CLASS 3 9-1 9 REFERENCES
- 1. Fort Calhoun Station Engineering Analysis, EA-FC 134, Rev. 1, "Reactor Vessel Internals Scoping, Screening, and Aging Management Review for License Renewal," October 16, 2002.
- 2. U.S. Nuclear Regulatory Commission, NUREG- 1782, "Safety Evaluation Report Related to the License Renewal of the Fort Calhoun Station," Docket 50-285, October 2003.
- 3. Fort Calhoun Station Document, PED-QP-33, Rev. 8, "Inservice Inspection and Inservice Test
- Program," October 11, 2011.
- 4. Fort Calhoun Station Document, PED-SEI-27, Rev. 9, "Inservice Inspection and Test Program,"
- October 11, 2007.
- 5. MaterialsReliability Program:PressurizedWater Reactor InternalsInspection and Evaluation
- 6. U.S. Nuclear Regulatory Commission, Code of Federal Regulations, 10 CFR Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants."
- 7. Fort Calhoun Station System Training Manual, STM-39, Vol. 39, Rev. 16, "Reactor Vessel & Core Construction System."
- 8. Fort Calhoun Station Engineering Analysis, EA-FC 137, Rev. 0, "FCS Inservice Inspection Program," February 10, 2005.
- 9. Fort Calhoun Station, LIC-02-0042, "Fort Calhoun Station Unit 1 Revised Application for Renewed
- Operating License," April 5, 2002.
- 10. U.S. Nuclear Regulatory Commission, NUREG- 1800 (Initial Issue), "Standard Review Plan for
- Review of License Renewal Applications for Nuclear Power Plants," April 2001.
- 11. Combustion Engineering Owners Group Report, CE NPSD- 1216, Rev. 0, "Generic Aging Management Review Report for the Reactor Vessel Internals," March 2001.
- 12. Materials Reliability Program:Screening, Categorization,andRanking ofReactorInternals
- Components for Westinghouse and Combustion EngineeringPWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.
- 13. MaterialsReliability Program:Inspection Standardfor PWR Internals (MRP-228). EPRI, Palo Alto, CA: 2009. 1016609.
- 14. Westinghouse Report, WCAP- 17096-NP, Rev. 2, "Reactor Internals Acceptance Criteria 0Methodology and Data Requirements," December 2009.
0 15. MaterialsReliability Program:Aging Management Strategiesfor Westinghouse and Combustion
- 16. Nuclear Energy Institute, NEI 03-08, "Guidelines for the Management of Materials Issues," March
- 2009.
- 17. U.S. Nuclear Regulatory Commission, NUREG-1801, "Generic Aging Lessons Learned (GALL)
- Report," July 2001.
- 18. Fort Calhoun Station Engineering Analysis, EA-FC-00-082, Rev. 0, "Chemistry Program Evaluation for License Renewal," January 28, 2005.
0 WCAP-17347-NP August 2012 Revision 1 0
0
9-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3
- 19. ASME Boiler and Pressure Vessel Code Section XI, 2003 Edition.
- 20. Fort Calhoun Station, FCSG-24, Rev. 25, "Corrective Action Program Guideline," September 9, 2010.
- 21. U.S. Nuclear Regulatory Commission Information Notice, IN 84-18, "Stress Corrosion Cracking in Pressurized Water Reactor Systems," March 7, 1984.
- 22. U.S. Nuclear Regulatory Commission Information Notice, IN 98-11, "Cracking of Reactor Vessel Internal Baffle Former Bolts in Foreign Plants," March 25, 1998.
- 23. Combustion Engineering Owners Group, CE NPSD-1098, Rev. 0, "Evaluation of the Applicability of Baffle Bolt Cracking to Ft. Calhoun and Palisades Internals Bolts," April 1998.
- 24. Westinghouse Report, WCAP- 16131, Rev. 0, "Reactor Internals Material Aging Fort Calhoun Station," November 2003.
- 25. U.S. Nuclear Regulatory Commission, NUREG-1801, "Generic Aging Lessons Learned (GALL) 0 Report," Revision 2, December 2010.
- 26. U.S. Nuclear Regulatory Commission, ML111990086, "NRC Regulatory Issue Summary 2011-07 License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management," July 21, 2011.
- 27. Fort Calhoun Station, FCSG-24-1, Rev. 0, "Condition Report Initiation," April 30, 2012.
- 28. Fort Calhoun Station, FCSG-24-4, Rev. Oa, "Condition Report and Cause Evaluation," April 30, 2012.
- 29. Fort Calhoun Station, FCSG-24-6, Rev. 0, "Corrective Action Implementation and Condition Report Closure," April 30, 2012.
- 30. Fort Calhoun Station, FCSG-24-10, Rev. 0, "Corrective Action Program Trending," April 30, 2012.
- 31. Westinghouse Letter, CFTC-12-72, Rev. 0, "Updating AMP Plan for Reactor Vessel Internals to MRP-227-A - Draft Report Comment Resolution," August 30, 2012.
- 32. U.S. Nuclear Regulatory Commission Letter, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components," May 19, 2000, (NRC ADAMS Accession No. ML003717179).
- 33. Westinghouse Report, WCAP-17471-P, Rev. 0, "OPPD Thermal Shield Integrity Program Task 2 Report," November 2011.
- 34. Westinghouse Design and Analysis Report, DAR-ME-05-15, Rev. 00, "Recommendations for Replacement ICI Thimble Measurement," September 30, 2005.
0 0
0 0
0 0
0 0
WCAP- 17347-NP August 2012 Revision I 0
0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-I APPENDIX A ILLUSTRATIONS r In-Core Instrumentation 0 Support Plate 0 I - CEDM Nozzle 0 Instrumentation Nozzle 0
Alignment Pin 0
Upper Guide 0 Structure 0
0 24" ID 0 Inlet Nozzle 0
_ Core Support 0 Barrel 0
0 Core Shroud Figure A-1 Typical CE Internals WCAP- 17347-NP August 2012 Revision 1
S 0
0 A-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 0 0
0 0
UPPER-- 0 CORE SUPPORT 0 BARREL 0
S 0
S 0
S 0
0 0
0 S
S Figure A-2 Core Support Barrel Assembly (Note: This figure is from [24].)
WCAP-17347-NP August 2012 Revision 1
0 0
0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-3 0
0 0
0 0
0 0
0 0
0 0o 0
0 0
0 0
0 0
0 0
0 0 a)t 0 CD S "a 0
0 Figure A-3 Illustration of Typical Bolting in Core Plates (Note: Ft. Calhoun has a similar configuration in the core shroud where the baffle-former bolts are equivalent to core shroud bolts without edge bolts.)
WCAP-17347-NP August 2012 Revision I
A-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 0
Core Shroud Plate S
0 0
0
/2 Centering Plate 0 0
0 S
0 0
0 0
0 0
0 0
S 0
0 0
Figure A-4 Typical Bolted Core Shroud Assembly 0 0
0 WCAP-17347-NP August 2012 Revision I
0 0
A-5 0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 High Fluence Seams I Figure A-5 High-Fluence Seam Locations WCAP- 17347-NP August 2012 Revision 1
0 A-6 A-6 WESTINGHOUSE NON-PROPRIETARY CLASS WESTINGHOUSE NON-PROPRIETARY CLASS 3 3 0 0
F- 0 0
0 0
0 0
0 Potential gaps at 0 panel-to-former S plate levels 0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 Figure A-6 Exaggerated View of Void Swelling Induced Distortion in Assembly WCAP- 17347-NP August 2012 Revision 1
WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-7 0 Flange Weld 0
Axial Weld 0
0 I Upper Core Barrel to 0 Lower Core Barrel Circumferential Weld 0
0 0
0 0 Lower Barrel Axial Weld 0
0 0
0 0
0 %% Lower Barrel 0 Circumferential Weld 0
0 Lower Barrel Axial Weld Core Barrel to Support Plate Weld Figure A-7 Typical Core Support Barrel Structure WCAP- 17347-NP August 2012 Revision 1
0 A-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 0 0
S
'KxFlange Weld 0
Upper Core 0
Barrel Flange 0
0 0
Core Barrel 0 0
0 0
0 0
Core Shroud Assembly 0 0
0 0
Core Support Plate Lower Support S Structure 0
0 0
Figure A-8 Core Support Barrel, Core Shroud Assembly, and Lower Support Structure WCAP- 17347-NP August 2012 Revision 1
-P A-WI3S GnqGSE CLAsSS3 VON 'OPRJEAR Assembly ofLowerSupport Structure Figure A-9 Schematic View 0
Core Core Shroud Barrel Assembly 0
0 Core Support Core Plate ColUpport Columns support 0 B~eams Figure A-1O Lower Support Structure Assembly 0
Revision L~ffarre1I~ 4 A-9
A-10 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-I 0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 0
0 0
0 0
0 0
S 0
0 (a) 0 0
0 0
0 D
0 CEA Guide Tubes
. Plate Extensions (b)
Figure A-1I (a) CE Schematic Illustration of a Portion of the Fuel Alignment Plate, and (b) CE Radial-View Schematic Illustration of the Guide Tubes WCAP-17347-NP August 2012 Revision 1
0 0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-II 0
0 S
0 0
S 0
0 (a) 0 Control Rod 0 Shroud Grid Assembly \
0 S
0 Control Rod 0 Shroud 0
Fuel 0 Bundle Alignment\
Plate (b)
Figure A-12 (a) Fuel Bundle Alignment Plate and (b) Upper Guide Structure WCAP- 17347-NP August 2012 Revision I
A-12 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-12 WESTINGHOUSE NON-PROPRIETARY CLASS 3 0
-C-%-
0 0
(D I
.1 -
0 IIT II I0 JZA 0 0
I I 0 A iCý3 0 0
0 0
0 0
0 i ii I
I L I I S
.1~
to. Mi I
I I
n KRE 1
I l i !
d Ad~I 0
0 Shroud Instrumnent Tubes Figure A-13 CE Schematic Illustration of the Control Element Assembly (CEA) Shroud Assembly WCAP-17347-NP August 2012 Revision 1
0 0
0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-13 0 -- ~ W . -
0 0
0 0
0 Control 0 Rod Shroud 0
0 0
0 0
S S Figure A-14 Control Rod Shroud Assembly 0
0 0
0 WCAP-17347-NP August 2012 Revision 1
0 0
A-14 WESTINGHOUSE NON-PROPRIETARY CLASS 3 0
0 0
0 S
0 0
0 0
0 0
S 0
0 0
U 0
0 0
Illustrates the deep beam grid structure (number 3), as well as the fuel alignment pins (numbers I and 2) 0 0
Figure A-15 Isometric View of the Lower Support Structure in the CE Core Shrouds with Full-Height Shroud Plates Units WCAP-17347-NP August 2012 Revision 1
S 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-15 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
S 0
0 0
S COANEM EDGE BRACIC'r S BAFFLE TO FORMER SOLT 0
0 0
Figure A-16 Bolting in a Typical Westinghouse Baffle-Former Structure (Note: Ft. Calhoun has a similar configuration in the core shroud where the baffle-former bolts are equivalent to panel-to-former bolts and there are no edge bolts.)
WCAP- 17347-NP August 2012 Revision I
A-16 0
WESTINGHOUSE NON-PROPRIETARY CLASS WESTINGHOUSE NON-PROPRIETARY CLASS 33 A-I 6 0 0
0 0
0 0
0 0
0 0
0 0
0 0
I . 0 0
0 a) Tb) c) 0 a) Early support column design 0 b) "Winged" support column design used in plants with second-generation core support assemblies c) Later support column design used in plants with second-generation core support assemblies 0 0
Figure A-17 CE Core Support Columns 0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 WCAP- 17347-NP August 2012 Revision 1
WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-17 0 Core Support Plate Support Column 0
0 0
0 0
0 Support Beam Flange Support Beam 0
0 Figure A-18 CE Lower Core Support Structure - Cross-Section 0 (Note: This figure is from [24]).
0 0
0 0
0 0
S S
0 WCAP-17347-NP August 2012 Revision 1
A-18 WESTINGHOUSE NON-PROPRIETARY CLASS 3 S
A- 18 WESTINGHOUSE NON-PROPRIETARY CLASS 3 S
0 S
S Weld lotons poln y affeted by swefliug inhizfa1 dalsum=
S 0
0 0
0 0
0 0
0 0
0 0
S 0
Figure A-19 Potential crack locations for CE welded core shroud assembled in stacked sections (This component is not applicable to FCS)
WCAP-17347-NP August 2012 Revision 1
0 0
0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-19 0
0 0
0 S
0 0
0 0
0 0
0 0
0 0
0 0
S 0
0 0
Figure A-20 Locations of potential separation between core shroud sections caused by swelling induced warping of thick flange plates in CE welded core shroud assembled in stacked sections (This component is not applicable to FCS)
WCAP- 17347-NP August 2012 Revision I
S A-20 WESTINGHOUSE NON-PROPRIETARY CLASS 3 0 Guide Lug 0 0
Top Plate 0
0 Ring 0
Brace 0 S
0 Rib Bottom Plate Figure A-21 CE welded core shroud with full height panels (This component is not applicable to FCS) 0 0
0 0
0 0
0 0
WCAP- 17347-NP August 2012 0 Revision 1 0
0
0 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-21 0
0 0
S 0
0 S
0 0
0 0
0 0
0 0
0 0
0 Figure A-22 CE lower support structures for welded core shrouds: separate core barrel and lower 0 support structure assembly with lower flange and core support plate 0
0 0
WCAP- 17347-NP August 2012 Revision 1
0 0 WESUNGHOUSE NON-PROPRIETARY CLASS 3 B-1 0
0 APPENDIX B S FT. CALHOUN NUCLEAR PLANT LICENSE RENEWAL AGING MANAGEMENT REVIEW
SUMMARY
TABLES The content and numerical identifiers in Table B-1 of this Appendix are extracted from Tables 2.3.1.1-1, 3.1-1, 3.1-2, and 3.1-3 in the Ft. Calhoun LRA [9].
Table B-1 LRA Aging Management Evaluation Summary - Ft. Calhoun LRA Table 2.3.1.1-1 Reactor Vessel Internals Component Types Subject to Aging Management Review Ft. Calhoun LRA FCS LRA Tables 3.1-1, 3.1-2, and 3.1-3 Tables 3.1-1, 3.1-2, 0 Component Type and 3.1-3 Aging Aging Management Row Number Effect/Mechanism Program 0
3.1.1.01 Cumulative fatigue TLAA, evaluated in mu atgue accordance with 10 damage CFR 54.21 (c) 0 3.1.1.08 Changes in dimension 0 due to void swelling Loss of fracture 0 CEA Shroud Bolts 3.1.1.32 toughness due to neutron irradiation embrittlement, PWR vessel internals; water chemistry and void swelling 0
3.1.1.34 Crack due initiation to SCCand 0 growth IASCC and PWR vessel internals; water chemistry 3.1.1.37 Loss of preload due to Inservice inspection; stress relaxation loose part monitoring TLAA, evaluated in 3.1.1.01 Cumulative fatigue accordance with 10 damage CFR 54.21(c) 3.1.1.08 Changes in dimension due to void swelling Loss of fracture CSB Snubber Bolts 3.1.1.32 toughness due to neutron PWR vessel internals; irradiation embrittlement, water chemistry and void swelling 3.1.1.34 Crack due initiation growth to SCCand and PWR vessel water internals; chemistry IASCC 3.1.1.37 Loss of preload due to Inservice inspection; stress relaxation loose part monitoring WCAP- 17347-NP August 2012 Revision 1
13-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table B-I LRA Aging Management Evaluation Summary - Ft. Calhoun LRA Table 2.3.1.1-1 (cont.) Reactor Vessel Internals Component Types Subject to Aging Management Review Ft. Calhoun LRA FCS LRA Tables 3.1-1, 3.1-2, and 3.1-3 Tables 3.1-1, 3.1-2, Component Type and 3.1-3 Aging Aging Management 0 Row Number EffectlMechanism Program 0
0 3.1.1.08 Changes in dimension Plant specific 0 due to void swelling 0
Loss of fracture 3.1.1.32 toughness due to neutron PWR vessel internals; 0
irradiation embrittlement, water chemistry 0 and void swelling Crack initiation and 0 3.1.1.34 growth due to SCC and PWR vessel internals; 0 IASCC water chemistry 0 Thermal Shield Bolts and 3.1.1.37 Loss of preload due to Inservice inspection; 0 Core Shroud Bolts stress relaxation loose part monitoring 3.1.3.01 Cracking Chemistry program 0
0 3.1.3.02 Reactor vessel Cracking internals inspection 0
program 0
3.1.3.03 Loss of preload Inservice inspection 0
program 0 3.1.3.04 Reduction of fracture Reactor vessel 0 toughness integrity program 0
3.1.1.01 Cumulative fatigue TLAA, evaluated in mu atgue accordance with 10 0 damage CFR 54.21 (c) 0 3.1.1.08 Changes in dimension Plant specific 0 due to void swelling CEA Shroud Spanner Nuts, Loss of fracture 0
and ICI Support Bolting 3.1.1.32 toughness due to neutron PWR vessel internals; 0 irradiation embrittlement, water chemistry 0 and void swelling Crack initiation and 0 3.1.1.34 growth due to SCC and PWR vessel internals; 0 IASCC water chemistry 0
0 0
0 0
0 0
WCAP- 17347-NP August 2012 0 Revision 1 0
0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-3 Table B-I LRA Aging Management Evaluation Summary - Ft. Calhoun LRA Table 2.3.1.1-1 (cont.) Reactor Vessel Internals Component Types Subject to Aging Management Review Ft. Calhoun LRA FCS LRA Tables 3.1-1, 3.1-2, and 3.1-3 Tables 3.1-1, 3.1-2, Component Type and 3.1-3 Aging Aging Management 0 Row Number Effect/Mechanism Program S
3.1.1.01 Cumulative fatigue TLAA, evaluated in muat gue accordance with 10 damage CFR 54.21 (c) 3.1.1.08 Changes in dimension Plant specific due to void swelling 0 CSB Bolts and Lower Internals Assembly Bolts 3.1.1.32 Loss of fracture toughness due to neutron PWR vessel internals; 0 irradiation embrittlement, water chemistry and void swelling 0 3.1.1.34 Crack initiation and PWR vessel internals; growth due to SCC and water chemistry 0 IASCC TLAA, evaluated in 0 3.1.1.01 Cumulative fatigue accordance with 10 damage CFR 54.21 (c) 0 0 3.1.1.08 Changes in dimension Plant specific due to void swelling 0 CEA Shrouds, Base, Tube, Loss of fracture and Transition Piece 3.1.1.26 toughness due to thermal Thermal aging and 0 aging, neutron irradiation neutron irradiation 0 embrittlement, and void swelling embrittlement 0 Crack initiation and 3.1.1.34 growth due to SCC and PWR vessel internals; 0 IASCC water chemistry S 3.1.1.08 Changes in dimension Plant specific 0 due to void swelling Loss of fracture 0 CSB, Core Support Ring 3.1.1.32 toughness due to neutron irradiation embrittlement, PWR vessel internals; water chemistry and void swelling 3.1.1.34 Crack initiation and PWR vessel internals;
.1.1.34 growth due to SCC and water chemistry IASCC WCAP-1 7347-NP August 2012 Revision I
B-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table B-1 LRA Aging Management Evaluation Summary - Ft. Calhoun LRA Table 2.3.1.1-1 (cont.) Reactor Vessel Internals Component Types Subject to Aging Management Review Ft. Calhoun LRA FCS LRA Tables 3.1-1, 3.1-2, and 3.1-3 Tables 3.1-1, 3.1-2, Component Type and 3.1-3 Row Number Aging Effect/Mechanism Aging Management Program S
0 3.1.1.08 Changes in dimension Plant specific S due to void swelling 0
3.1.1.29 Loss of material due to wear Inservice inspection 0 Loss of fracture S
CSB Alignment Key and CSB Upper Flange 3.1.1.32 toughness due to neutron PWR vessel internals; 0 irradiation embrittlement, water chemistry and void swelling 0
Crack initiation and 0 3.1.1.34 growth due to SCC and PWR vessel internals; 0
IASCC water chemistry S
TLAA, evaluated in 3.1.1.01 Cumulative fatigue accordance with 10 CFR 54.21 (c)
S damage S
3.1.1.08 Changes in dimension Plant specific S due to void swelling S
Loss of fracture CSB Nozzle 3.1.1.32 toughness due to neutron PWR vessel internals; S irradiation embrittlement, and void swelling water chemistry S
S 3.1.1.34 Crack initiation and growth due to SCC and PWR vessel internals; water chemistry S IASCC S
3.1.1.08 Changes in dimension due to void swelling S
Loss of fracture 0
CSB - Spacer, Locking 3.1.1.32 toughness due to neutron PWR vessel internals; S Collar, Dowel Pin, and Locking Bar irradiation embrittlement, and void swelling water chemistry S
Crack initiation and S 3.1.1.34 growth due to SCC and PWR vessel internals; S
IASCC water chemistry S
S S
S S
S S
WCAP- 17347-NP August 2012 S Revision 1 0
0 0
0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-5 0
Table B-I LRA Aging Management Evaluation Summary - Ft. Calhoun LRA Table 2.3.1.1-1 0 (cont.) Reactor Vessel Internals Component Types Subject to Aging Management Review 0 Ft. Calhoun LRA FCS LRA Tables 3.1-1, 3.1-2, and 3.1-3 0 Tables 3.1-1, 3.1-2, Component Type and 3.1-3 Aging Aging Management 0 Row Number Effect/Mechanism Program 0
0 3.1.1.08 Changes in dimension Plant specific due to void swelling 0
3.1.1.29 Loss of material due to Inservice inspection 0 wear 0 Loss of fracture CSB Snubber Spacer Block 3.1.1.32 toughness due to neutron PWR vessel internals; 0 irradiation embrittlement, water chemistry 0 and void swelling 0 3.1.1.34 Crack initiation and growth due to SCC and PWR vessel internals; IASCC water chemistry 0
3.1.1.01 Cumulative fatigue TLAA, evaluated in 0 mu atgue accordance with 10 damage CFR 54.21 (c) 0 0 3.1.1.08 Changes in dimension due to void swelling Plant specific 0
Core Shroud Loss of fracture 0 3.1.1.32 toughness due to neutron PWR vessel internals; irradiation embrittlement, water chemistry 0 and void swelling 0 3.1.1.34 Crack initiation and PWR vessel interals; growth due to SCC and water chemistry IASCC S
0 0
0 0
0 S
S 0
WCAP-17347-NP August 2012 Revision 1
0 0
0 B-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 0 0
Table B-1 LRA Aging Management Evaluation Summary - Ft. Calhoun LRA Table 2.3.1.1-1 (cont.) Reactor Vessel Internals Component Types Subject to Aging Management Review 0 Ft. Calhoun LRA FCS LRA Tables 3.1-1, 3.1-2, and 3.1-3 0 Tables 3.1-1, 3.1-2, 0 Component Type and 3.1-3 Aging Aging Management Row Number Effect/Mechanism Program 0
0 Changes in Dimensions Void swelling as a result 0
of helium bubble 0 nucleation and growth Reactor Vessel 3.1.2.08 from nuclear 0
transmutation reactions Inte ral 0 of nickel or boron in the Program austenitic stainless steel 0 or nickel-based alloy material.
0 Cracking 0
-Primary Water Stress Corrosion Cracking 0
-irradiation-assisted 0 3.1.2.09 stress corrosion cracking Alloy 600 Program in the presence of oxygen 0
Flow Skirt concentrations > 5 ppb, halogen concentrations >
0 150 ppb, and fluence 0 levels > 5 E20 n/cm2 Fatigue 0
3.1.2.10 Due to repeated Fatigue Monitoring Program 0
stress/strain cycles caused by fluctuating 0 loads and temperatures 0 Reduction of Fracture Toughness 0 11 Due to changes in the Reactor Vessel properties of the stainless Internals Inspection 0 steel and nickel-base Program 0 alloys used in reactor internals 0 3.1.1.08 Changes in dimension Plant specific 0 due to void swelling 0
Fuel Assembly Alignment 3.1.1.29 Loss of material due to Inservice inspection Plate wear 0
Crack initiation and 3.1.1.34 growth due to SCC and PWR vessel internals; 0 IASCC water chemistry 0
0 0
0 0
0 WCAP-17347-NP 0
August 2012 Revision 1 0 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-7 Table B-1 LRA Aging Management Evaluation Summary - Ft. Calhoun LRA Table 2.3.1.1-1 (cont.) Reactor Vessel Internals Component Types Subject to Aging Management Review Ft. Calhoun LRA FCS LRA Tables 3.1-1, 3.1-2, and 3.1-3 0 Tables 3.1-1, 3.1-2, Component Type and 3.1-3 Aging Aging Management 0 Row Number Effect/Mechanism Program 0 3.1.1.01 Cumulative fatigue TLAA, evaluated in accordance with 10 0 damage CFR 54.21 (c)
S ICI Guide Tube & Supports 3.1.1.08 Changes in dimension Plant specific due to void swelling Crack initiation and 3.1.1.34 growth due to SCC and PWR vessel internals; IASCC water chemistry 3.1.1.08 Changes in dimension Plant specific due to void swelling ICI Support Plate & Gusset 0 Crack initiation and growth due to SCC and PWR vessel internals; water chemistry
________________ ___________IASCC 0 3.1.1.08 Changes in dimension Plant specific due to void swelling 0 Instrument Tube &
Supports 3.1.1.34 Crack initiation and PWR vessel internals; growth due to SCC and water chemistry S IASCC 3.1.1.08 Changes in dimension Plant specific due to void swelling 0 Lower Internals Assembly - 3.1.1.32 Loss of fracture toughness due to neutron PWR vessel internals; Manhole Cover Plate & irradiation embrittlement, water chemistry Bottom Plate and void swelling 3.1.1.34 Crack initiation and PWR vessel internals; growth due to SCC and water chemistry 0 _________________ ___________IASCC TLAA, evaluated in 3.1.1.01 Cumulative fatigue accordance with 10 damage CFR 54.21 (c) 0 3.1.1.08 Changes in dimension Plant specific due to void swelling Lower Internals Assembly - Loss of fracture Core Support Columns 3.1.1.26 toughness due to thermal Thermal aging and aging, neutron irradiation neutron irradiation embrittlement, and void embrittlement swelling Crack initiation and 3.1.1.34 growth due to SCC and PWR vessel internals; IASCC water chemistry WCAP- 17347-NP August 2012 Revision I
B-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table B-I LRA Aging Management Evaluation Summary - Ft. Calhoun LRA Table 2.3.1.1-1 (cont.) Reactor Vessel Internals Component Types Subject to Aging Management Review Ft. Calhoun LRA FCS LRA Tables 3.1-1, 3.1-2, and 3.1-3 Tables 3.1-1, 3.1-2, Component Type and 3.1-3 Aging Aging Management Row Number Effect/Mechanism Program S 3.1.1.01 Cumulative fatigue TLAA, evaluated in 0
C muat ge damage accordance with 10 CFR 54.21 (c)
S 3.1.1.08 Changes in dimension Plant specific 0
Lower Internals Assembly - due to void swelling 0 Core Support Plate and Loss of fracture S Support Beams and Flanges 3.1.1.32 toughness due to neutron PWR vessel internals; irradiation embrittlement, water chemistry S and void swelling Crack initiation S
to SCCand 3.1.1.34 growth due and PWR vessel internals; S
IASCC water chemistry 3.1.1.08 Changes in dimension Plant specific S
due to void swelling Plantspecific S Lower Internals Assembly - 3.1.1.32 Loss of fracture toughness due to neutron PWR vessel internals; S
Anchor Block and Dowel irradiation embrittlement, water chemistry S Pins and void swelling Crack initiation and S
3.1.1.34 growth due to SCC and PWR vessel internals; S
IASCC water chemistry TLAA, evaluated in S
3.1.1.01 Cumulative fatigue accordance with 10 damage CFR 54.21 (c)
S 3.1.3.02 Reactor Vessel S Cracking Internals Inspection Program S
Thermal Shield 3.1.3.04 Reduction of Fracture Reactor Vessel S Toughness Integrity Program S 3.1.3.05 Reactor Vessel Changes in Dimensions Internals Inspection S Program S 3.1.3.14 Fatigue TLAA S
3.1.1.01 Cumulative fatigue TLAA, evaluated in accordance with 10 S
damage CFR 54.21 (c) S Thermal Shield - Positioning 3.1.3.02 Cracking Reactor Vessel Internals Inspection S
Program S
Pin & Shim 3.1.3.03 Loss of Preload L
Inservice Inspection oProgram S
S 3.1.3.04 Reduction of Fracture Reactor Vessel Toughness Integrity Program S S
S WCAP- 17347-NP August 2012 Revision 1 S S
S
WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-9 Table B-1 LRA Aging Management Evaluation Summary - Ft. Calhoun LRA Table 2.3.1.1-1 (cont.) Reactor Vessel Internals Component Types Subject to Aging Management Review Ft. Calhoun LRA FCS LRA Tables 3.1-1, 3.1-2, and 3.1-3 S Tables 3.1-1, 3.1-2, Component Type and 3.1-3 Aging Aging Management Row Number Effect/Mechanism Program 0
0 3.1.1.08 Changes in dimension Plant specific S UGS - Ring Shim, Tab, & due to void swelling Plate 3.1.1.34 Crack initiation and PWR vessel internals; growth due to SCC and wr vesstrn IASCC water chemistry 3.1.1.08 Changes in dimension Plant specific UGS - Dowel Pin & Locking due to void swelling 0 Strip 3.1.1.34 Crack initiation and PWR vessel internals; growth due to SCC and water chemistry IASCC waterchemistry S 3.1.1.08 Changes in dimension Plant specific due to void swelling Plantspecific UGS - Guide PinUi nw GS - G uid 3.1.1.29 P ear due to Loss of material Inservice inspection 0
0 3.1.1.34 Crack initiation and growth due to SCC and PWR vessel internals; water chemistry 0 _________________ ____________IASCC________
3.1.1.08 Changes in dimension Plant specific due to void swelling Plantspecific 3.1.1.29 Loss of material due to lii ti 0 UGS - Alignment Lug wear 0 3.1.1.34 Crack initiation and PWR vessel internals; growth due to SCC and water chemistry S ________________ ___________IASCC S 3.1.1.08 Changes in dimension Plant specific S UGS - Alignment Lug Screw due to void swelling and Nut 3.1.1.34 Crack initiation and PWR vessel internals; growth due to SCC and wr vesstry 0 IASCC water chemistry 0 3.1.1.08 Changes in dimension Plant specific due to void swelling 0 UGS - Key Slot Tab Crack initiation and S 3.1.1.34 growth due to SCC and PWR vessel internals; IASCC water chemistry 3.1.1.08 Changes in dimension Plant specific due to void swelling 3.1.1.29 Loss of material UGS - Hold-down Ring wear due to Inservice inspection 3.1.1.34 Crack initiation and PWR vessel internals; growth due to SCC and water chemistry
_________________ ____________IASCC________
WCAP- 17347-NP August 2012 Revision 1
0 0
0 B-10 WESTINGHOUSE NON-PROPRIETARY CLASS 3 0 0
Table B-1 LRA Aging Management Evaluation Summary - Ft. Calhoun LRA Table 2.3.1.1-1 (cont.) Reactor Vessel Internals Component Types Subject to Aging Management Review 0 Ft. Calhoun LRA FCS LRA Tables 3.1-1, 3.1-2, and 3.1-3 0 Tables 3.1-1, 3.1-2, 0 Component Type and 3.1-3 Aging Aging Management Row Number Effect/Mechanism Program 0
0 3.1.1.08 Changes in dimension Plant specific 0 due to void swelling 0
3.1.1.29 Loss of material due to Inservice inspection UGS - Support Plate &
Sleeves wear 0 Crack initiation and 0 3.1.1.34 growth due to SCC and PWR vessel internals;
__ 1 IASCC water chemistry 0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 WCAP- 17347-NP August 2012 0 Revision 1 0
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-1 APPENDIX C MRP-227-A AUGMENTED INSPECTIONS Table C-I MRP-227-A Primary Inspection and Monitoring Recommendations for CE-Designed Internals Examination Effect' Expansion Link Method/Frequency Item Applicability (Mechanism) (Note 1) (Note 1) Examination Coverage Core Shroud Bolted plant Cracking (IASCC, Core support column Baseline volumetric (UT) 100% of accessible bolts (see Assembly (Bolted) designs Fatigue) bolts, barrel-shroud examination between 25 and 35 Note 3). Heads are accessible Core shroud bolts Aging bolts EFPY, with subsequent from the core side. UT Management (IE (FCS: former-to-core examination on a ten-year accessibility may be affected (FCS: panel-to- head and and ISR) barrel bolts) interval, by complexity of former bolts) (Note 2) locking device designs.
See Figures A-3.
Core Shroud Plant designs Cracking (IASCC) Remaining axial welds Enhanced visual (EVT-1) Axial and horizontal weld Assembly with core Acing examination no later than 2 seams at the core shroud re-(Welded) shrouds Aging refueling outages from the entrant comers as visible from Core shroud plate- assembled in Management (IE) beginning of the license the core side of the shroud, former plate weld two vertical (Note 2) renewal period and subsequent within six inches of central sections examination on a ten-year flange and horizontal (Not applicable to interval. stiffeners.
FCS) See Figures A-19 and A-20.
Core Shroud Plant designs Cracking (IASCC) Remaining axial welds, Enhanced visual (EVT-1) Axial weld seams at the core Assembly with core Ribs and rings examination no later than 2 shroud re-entrant comers, at (Welded) shrouds Aging refueling outages from the the core mid-plane (+/- three Shroud plates assembled Management (IE) beginning of the license feet in height) as visible from with full- (Note 2) renewal period and subsequent the core side of the shroud.
(Not applicable to height shroud examination on a ten-year See Figure A-21.
FCS) plates I interval. II WCAP-17347-NP August 2012 Revision I
C-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table C-i MRP-227-A Primary Inspection and Monitoring Recommendations for CE-Designed Internals (cont.)
Examination Effect Expansion Link Method/Frequency item Applicability (Mechanism) (Note 1) (Note 1) Examination Coverage Core Shroud Bolted plant Distortion (Void None Visual (VT-3) examination no Core-side surfaces as Assembly (Bolted) designs Swelling), later than 2 refueling outages indicated.
Assembly including: from the beginning of the See Figures A-4, A-5 and A-6.
- Abnormal license renewal period.
interaction with Subsequent examinations on a fuel assemblies ten-year interval.
- Gaps along high fluence shroud plate joints eVertical displacement of shroud plates near high fluence joint Aging Management (IE)
Core Shroud Plant designs Distortion (Void None Visual (VT-1) examination no If a gap exists, make three to Assembly with core Swelling), as later than 2 refueling outages five measurements of gap (Welded) shrouds evidenced by from the beginning of the opening from the core side at Assembly assembled in separation between license renewal period, the core shroud re-entrant two vertical the upper and Subsequent examinations on a comers. Then, evaluate the (Not applicable to sections lower core shroud ten-year interval. swelling on a plant specific FCS) segments basis to determine frequency Aging and method for additional Management (IE) examinations.
See Figures A-19 and A-20.
August 2012 WCAP- 17347-NP WCAP- 17347-NP August 2012 Revision I
WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-3 Table C-I MRP-227-A Primary Inspection and Monitoring Recommendations for CE-Designed Internals (cont.)
Examination Effect Expansion Link Method/Frequency Item Applicability (Mechanism) (Note 1) (Note 1) Examination Coverage Core Support All plants Cracking (SCC) Lower core support Enhanced visual (EVT- 1) 100% of the accessible Barrel Assembly beams examination no later than surfaces of the upper flange Upper (core support Core support barrel 2 refueling outages from the weld (Note 4).
barrel) flange weld assembly upper beginning of the license See Figures A-7 and A-8.
cylinder renewal period. Subsequent Upper core barrel examinations interval. on a ten-year flange Core Support All plants Cracking (SCC, Lower cylinder axial Enhanced visual (EVT- 1) 100% of the accessible Barrel Assembly IASCC) welds examination no later than 2 surfaces of the lower cylinder Lower cylinder girth Aging refueling outages from the welds (Note 4).
welds Management (IE) beginning of the license See Figures A-7 and A-8.
renewal period. Subsequent examination on a ten-year interval.
Lower Support All plants Cracking (SCC, None Visual (VT-3) examination no 100% of the accessible Structure IASCC, Fatigue later than 2 refueling outages surfaces of the core support Core support column including damaged from the beginning of the column welds (Note 5).
welds or fractured license renewal period. See Figures A-22 and A-I 7.
material) Subsequent examinations on a Aging ten-year interval.
WCAP-17347-NP August 2012 Revision 1
C-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table C-I MRP-227-A Primary Inspection and Monitoring Recommendations for CE-Designed Internals (cont.)
Examination Effect Expansion Link Method/Frequency Item Applicability (Mechanism) (Note 1) (Note 1) Examination Coverage Core Support All plants Cracking (Fatigue) None If fatigue life cannot be Examination coverage to be Barrel Assembly demonstrated by time-limited defined by evaluation to Lower flange weld aging analysis (TLAA), determine the potential enhanced visual (EVT-1) location and extent of fatigue examination, no later than cracking.
2 refueling outages from the See Figures A-7, A-8, and A-beginning of the license 22.
renewal period. Subsequent examination on a ten-year interval.
Lower Support All plants with Cracking (Fatigue) None If fatigue life cannot be Examination coverage to be Structure a core support Aging demonstrated by time-limited defined by evaluation to Core support plate plate Management (IE) aging analysis (TLAA), determine the potential enhanced visual (EVT-1) location and extent of fatigue examination, no later than 2 cracking.
refueling outages from the See Figures A-2, A-9, A-10, beginning of the license A-18 and A-22.
renewal period. Subsequent examination on a ten-year interval.
Upper Internals All plants with Cracking (Fatigue) None If fatigue life cannot be Examination coverage to be Assembly core shrouds demonstrated by time-limited defined by evaluation to Fuel alignment plate assembled aging analysis (TLAA), determine the potential with full- enhanced visual (EVT-1) location and extent of fatigue height shroud examination, no later than cracking.
plates 2 refueling outages from the See Figure A-Il and A-12.
beginning of the license renewal period. Subsequent examination on a ten-year interval.
WCAP- 17347-NP August 2012 Revision 1
WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-5 Table C-i MRP-227-A Primary Inspection and Monitoring Recommendations for CE-Designed Internals (cont.)
Examination Effect Expansion Link Method/Frequency Item Applicability (Mechanism) (Note 1) (Note 1) Examination Coverage Control Element All plants with Cracking (SCC, Remaining instrument Visual (VT-3) examination, no 100% of tubes in peripheral Assembly instrument Fatigue) that guide tubes within the later than 2 refueling outages CEA shroud assemblies (i.e.,
Instrument guide guide tubes in results in missing CEA shroud assemblies from the beginning of the those adjacent to the perimeter tubes the CEA supports or license renewal period, of the fuel alignment plate).
shroud separation at the Subsequent examination on a See Figures A- 13 and A- 14.
assembly welded joint ten-year interval.
between the tubes Plant-specific component and supports integrity assessments may be required if degradation is detected and remedial action is needed.
Lower Support All plants with Cracking (Fatigue) None Enhanced visual (EVT-1) Examine beam-to-beam welds, Structure core shrouds that results in a examination, no later than in the axial elevation from the Deep beams assembled detectable surface- 2 refueling outages from the beam top surface to four with full- breaking indication beginning of the license inches below.
(Not applicable to height shroud in the welds or renewal period. Subsequent See Figure A-15.
FCS) plates beams examination on a ten-year Aging interval, if adequacy of Management (IE) remaining fatigue life cannot be demonstrated.
Notes:
- 1. Examination acceptance criteria and expansion criteria for the CE components are in Table C-4.
- 2. Void swelling effects on this component is managed through management of void swelling on the entire core shroud assembly.
- 3. A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table C4, must be examined for inspection credit.
- 4. A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Table C4, must be examined from either the inner or outer diameter for inspection credit.
- 5. A minimum of 75% of the total population of core support column welds.
WCAP- 17347-NP August 2012 Revision 1
C-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table C-2 MRP-227-A Expansion Inspection and Monitoring Recommendations for CE-Designed Internals Examination Effect Primary Link Method/Frequency Item Applicability (Mechanism) (Note 1) (Note 1) Examination Coverage Core Shroud Bolted plant Cracking (IASCC, Core shroud bolts Volumetric (UT) examination. 100% (or as supported by plant-Assembly (Bolted) designs Fatigue) (FCS: panel-to- Re-inspection every 10 years specific justification; Note 2) of Barrel-shroud bolts Aging Management former bolts) following initial inspection, barrel-shroud and guide lug (FCS: former-to-core (IE and ISR) insert bolts with neutron barrel bolts) fluence exposures > 3 displacements per atom (dpa).
See Figure A-16.
(Guide lugs not applicable to FCS)
Core Support Barrel All plants Cracking (SCC, Upper (core Enhanced visual (EVT-1) 100% of accessible welds and Assembly Fatigue) support barrel) examination, adjacent base metal. (Note 2).
Lower core barrel flange weld Re-inspection every 10 years See Figures A-7 and A-8.
flange following initial inspection.
Core Support Barrel All plants Cracking (SCC) Upper (core Enhanced visual (EVT-1) 100% of accessible surfaces of Assembly Aging Management support barrel) examination, the welds and adjacent base Upper cylinder (IE) flange weld Re-inspection every 10 years metal (Note 2).
(including welds) following initial inspection. See Figure A-7.
Core Support Barrel All plants Cracking (SCC) Upper (core Enhanced visual (EVT-1) 100% of accessible bottom Assembly support barrel) examination. surface of the flange (Note 2).
Upper core barrel flange weld Re-inspection every 10 years See Figure A-7.
flange following initial inspection.
Core Support Barrel All plants Cracking (SCC) Core barrel Enhanced visual (EVT-1) 100% of one side of the Assembly assembly girth examination, with initial and accessible weld and adjacent Core barrel assembly welds subsequent examinations base metal surfaces for the weld axial welds dependent on the results of core with the highest calculated barrel assembly girth weld operating stress.
examinations. See Figure A-7.
WCAP- 17347-NP August 2012 Revision I
WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-7 Table C-2 MRP-227-A Expansion Inspection and Monitoring Recommendations for CE-Designed Internals Examination Effect Primary Link Method/Frequency Item Applicability (Mechanism) (Note 1) (Note 1) Examination Coverage Lower Support All plants Cracking (SCC, Upper (core Visual (EVT-1) examination. 100% of accessible surfaces Structure except those Fatigue) including support barrel) Re-inspection every 10 years (Note 2).
Lower core support with core damaged or flange weld following initial inspection. See Figures A-22 and A-i17.
beams shrouds fractured material assembled with full-height shroud plates Core Shroud Bolted plant Cracking (IASCC, Core shroud bolts Ultrasonic (UT) examination. 100% (or as supported by plant-Assembly (Bolted) designs Fatigue) Re-inspection every 10 years specific analysis) of core Core support column Aging Management following initial inspection, support column bolts with bolts (IE) neutron fluence exposures > 3 dpa (Note 2).
See Figure A-22.
Core Shroud Plant designs Cracking (IASCC) Shroud plates of Enhanced visual (EVT-1) Axial weld seams other than the Assembly (Welded) with core Aging Management welded core examination, core shroud re-entrant comer Remaining axial shrouds (IE) shroud assemblies Re-inspection every 10 years welds at the core mid-plane, welds, Ribs and rings assembled with following initial inspection, plus ribs and rings.
full-height See Figure A-2 1.
shroud plates Control Element All plants with Cracking (SCC, Peripheral Visual (VT-3) examination. 100% of tubes in CEA shroud Assembly instrument Fatigue) that results instrument guide Re-inspection every 10 years assemblies (Note 2).
Remaining instrument guide tubes in in missing supports tubes within the following initial inspection. See Figures A-13 and A-14.
guide tubes the CEA shroud or separation at the CEA shroud assembly welded joint assemblies between the tubes and supports.
Notes:
- 1. Examination acceptance criteria and expansion criteria for the CE components are in Table C-4.
- 2. A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).
WCAP-17347-NP August 2012 Revision I
C-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table C-3 MRP-227-A Existing Inspection and Aging Management Programs Credited in Recommendations for CE-Designed Internals Effect Item Applicability (Mechanism) Reference Examination Method Examination Coverage Core Shroud All plants Loss of material ASME Code Section Visual (VT-3) examination, First 10-year ISI after 40 years Assembly (Wear) XI general condition examination of operation, and at each Guide lugs Aging for detection of excessive or subsequent inspection interval.
Guide lug inserts and Management (ISR) asymmetrical wear. Accessible surfaces at specified bolts frequency.
(Not applicable to FCS)
Lower Support All plants Cracking (SCC, ASME Code Section Visual (VT-3) examination to Accessible surfaces at specified Structure with core IASCC, Fatigue) Xl detect severed fuel alignment frequency.
Fuel alignment pins shrouds Aging pins, missing locking tabs, or assembled Management (1E excessive wear on the fuel (Not applicable to with full- and ISR) alignment pin nose or flange.
FCS) height shroud plates Lower Support All plants Loss of material ASME Code Section Visual (VT-3) examination. Accessible surfaces at specified Structure with core (Wear) XI frequency.
Fuel alignment pins shrouds Aging assembled in Management (IE (Not applicable to two vertical and ISR)
FCS) sections Core Barrel All plants Loss of material ASME Code Section Visual (VT-3) examination. Area of the upper flange Assembly (Wear) XI potentially susceptible to wear.
Upper flange August 2012 WCAP- 17347-NP August 2012 Revision I
WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-9 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for CE-Designed Internals Examination Acceptance Criteria Additional Examination Item Applicability (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Core Shroud Bolted plant Volumetric (UT) a. Core support a. Confirmation that >5% of a and b. The examination Assembly (Bolted) designs examination, column bolts the Core shroud bolts (FCS: acceptance criteria for the UT Core shroud bolts The examination b. Barrel-shroud panel-to-former bolts) in the of the core support column (FCS: panel-to-former acceptance criteria bolts (FCS: former- four plates at the largest bolts and barrel-shroud bolts bolts) for the UT of the to-core barrel bolts) distance from the core contain (FCS: former-to-core barrel core shroud bolts unacceptable indications shall bolts) shall be established as (FCS: panel-to- require UT examination of the part of the examination former bolts) shall lower support column bolts technical justification.
be established as barrel within the next 3 part of the refueling cycles.
examination b. Confirmation that >5% of technical the core support column bolts justification. contain unacceptable indication shall require UT examination of the barrel-shroud bolts (FCS:
former-to-core barrel bolts) within the next 3 refueling cycles.
Core Shroud Plant designs Visual (EVT-1) Remaining axial Confirmation that a surface- The specific relevant Assembly with core examination, welds breaking indication > 2 inches condition is a detectable crack-(Welded) shrouds in length has been detected and like surface indication.
Core shroud plate- assembled in The specific sized in the core shroud plate-former plate weld two vertical relevant former plate weld at the core sections condition is a shroud re-entrant comers (as (Not applicable to detectable visible from the core side of the FCS) crack-like surface shroud), within 6 inches of the indication, central flange and horizontal stiffeners, shall require EVT-I examination of all remaining axial welds by the completion of the next refueling outage.
WCAP- 17347-NP August 2012 Revision I
C-IO WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for CE-Designed Internals (cont.)
Examination Acceptance Criteria Additional Examination Item Applicability (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Core Shroud Assembly Plant designs Visual (EVT-1) a. Remaining axial a. Confirmation that a surface- The specific relevant (Welded) with core examination, welds breaking indication > 2 inches condition is a detectable Shroud plates shrouds b. Ribs and rings in length has been detected and crack-like surface indication.
assembled The specific sized in the axial weld seams at (Not applicable to FCS) with full- relevant condition is the core shroud re-entrant height a detectable crack- comers at the core mid-plane shroud plates like surface shall require EVT-1 or UT indication, examination of all remaining axial welds by the completion of the next refueling outage.
- b. If extensive cracking is detected in the remaining axial welds, an EVT- 1 examination shall be required of all accessible rib and ring welds by the completion of the next 1 refueling outage.
August 2012 WCAP- 17347-NP WCAP- August 2012 Revision I
WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-11 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for CE-Designed Internals (cont.)
Examination Acceptance Criteria Additional Examination Item Applicability (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Core Shroud Assembly Bolted plant Visual (VT-3) None N/A N/A (Bolted) designs examination.
Assembly The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, and vertical displacement of shroud plates near high fluence joints.
Core Shroud Assembly Plant designs Visual (VT-1) None N/A N/A (Welded) with core examination.
Assembly shrouds assembled in The specific (Not applicable to FCS) two vertical relevant condition is sections evidence of physical separation between the upper and lower core shroud sections.
WCAP- 17347-NP August 2012 Revision 1
C-12 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for CE-Designed Internals (cont.)
Examination Acceptance Criteria Additional Examination Item Applicability (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Core Support Barrel All plants Visual (EVT- 1) Lower core support Confirmation that a surface- The specific relevant Assembly examination, beams breaking indication >2 inches condition is a detectable Upper (core support The specific Upper core barrel in length has been detected and crack-like surface indication.
barrel) flange weld relevant condition is cylinder (including sized in the upper flange weld a detectable crack- welds) shall require that an EVT- I like surface Uexamination of the lower core indication. Upper core barrel support beams, upper core
,flange barrel cylinder, and upper core barrel flange be performed by the completion of the next refueling outage.
Core Support Barrel All plants Visual (EVT-1) Lower cylinder axial Confirmation that a surface- The specific relevant Assembly examination, welds breaking indication >2 inches condition for the expansion Lower cylinder girth The specific in length has been detected and lower cylinder axial welds is welds relevant condition is sized in the lower cylinder girth a detectable crack-like a detectable crack- weld shall require that an EVT- surface indication.
like surface I examination of all accessible indication, lower cylinder axial welds by the completion of the next refueling outage.
Lower Support All plants Visual (VT-3) None None Structure examination.
Core support column The specific welds relevant condition is missing or separated welds.
WCAP-17347-NP August 2012 Revision I
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-13 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for CE-Designed Internals (cont.)
Examination Acceptance Criteria Additional Examination Item Applicability (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Core Support Barrel All plants Visual (EVT-1) None N/A N/A Assembly examination.
Lower flange weld The specific relevant condition is a detectable crack-like indication.
Lower Support All plants Visual (EVT-1) None N/A N/A Structure with a core examination.
Core support plate support plate The specific relevant condition is a detectable crack-like surface indication Upper Internals All plants Visual (EVT-I) None N/A N/A Assembly with core examination.
Fuel alignment plate shrouds The specific (Not applicable to FCS) assembled with fulla-eetal relevant condition cak is with ull- a detectable crack-height shroud like surface plates indication.
WCAP- 17347-NP August 2012 Revision 1
C-14 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for CE-Designed [nternals (cont.)
Examination Acceptance Criteria Additional Examination Item Applicability (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Control Element All plants Visual (VT-3) Remaining Confirmed evidence of missing The specific relevant Assembly with examination, instrument tubes supports or separation at the conditions are missing Instrument guide tubes instruments The specific within the CEA welded joint between the tubes supports and separation at the tubes in the relevant conditions shroud assemblies and supports shall require the welded joint between the CEA shroud are missing supports visual (VT-3) examination to tubes and the supports.
assembly and separation at the be expanded to the remaining welded joint instrument tubes within the between the tubes CEA shroud assemblies by and the supports. completion of the next refueling outage.
Lower Support All plants Visual (EVT-1) None N/A N/A Structure with core examination.
Deep beams shrouds The specific assembled relevant condition is (Not applicable to FCS) with full- a detectable crack-height shroud like indication.
plates Note:
I. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).
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