NRC 2009-0095, License Renewal Commitment Reactor Vessel Lnternals Program Submittal
| ML092750419 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/02/2009 |
| From: | Meyer L Point Beach |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC 2009-0095 | |
| Download: ML092750419 (53) | |
Text
POINT BEACH October 2,2009 NRC 2009-0095 10 CFR 54 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units Iand 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 License Renewal Commitment Reactor Vessel lnternals Program Submittal
References:
( 1 Nuclear Management Company, LLC, letter to NRC, dated February 25, 2004, Application for Renewed Operating License (ML040580020)
(2) FPL Energy Point Beach, LLC letter to NRC, dated August 29, 2008, Reactor Vessel lnternals Program Regulatory Commitment Change (ML082480192)
In support of the Point Beach Nuclear Plant (PBNP) license renewal application Reference (I),
Nuclear Management Company, LLC, the former license holder for PBNP, committed to submitting a Reactor Vessel lnternals Program. The commitment, as modified in Reference (2),
included a plan to develop an aging management program for the reactor internals based on the guidance contained in an Electric Power Research Institute (EPRI) Material Reliability Program (MRP) technical report. The Reactor Vessel lnternals Program was to be submitted to the NRC for approval one year prior to the commencement of renewed operation.
Enclosure Iprovides PBNP program document AM 3-44, Reactor Vessel lnternals Program.
This program is based upon the guidance contained in the EPRI technical report MRP-227, Pressurized Water Reactor lnternals Inspection and Evaluation Guidelines, Revision 0.
Summary of Regulatory Commitments This letter contains no new Regulatory Commitments. This letter fulfills the following Regulatory Commitment:
o Commitment 39 - FPL Energy Point Beach will submit the Reactor Vessel lnternals Aging Management Program to the NRC for review one year prior to commencement of renewed operation.
NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241
Document Control Desk Page 2 In accordance with 10 CFR 50.91, a copy of this letter is being provided to the designated Wisconsin Official.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on October 2, 2009.
Very truly yours, NextEra Energy Point Beach, LLC
/%.? fl Larry Meyer Site Vice President Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW
ENCLOSURE I NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS I AND 2 AM 3-44 REVISION 0 REACTOR \/ESSEL INTERNALS PROGRAM 50 pages follow
ACTOR VESSEL INTE DOCUMENT TYPE: Adrninistrative REVISION: 0 EFFECTIVE DATE: September 30,2009 APPROVAL AUTHORITY: Department Manager PROCEDURE OWNER (title): Group Head OWNER GROUP: Program Engineering
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30. 2009 REACTOR VESSEL INTERNALS PROGRAM TABLE OF CONTENTS SECTION TITLE PAGE 1.0 PURPOSE ........................................................................................................................... 3
2.0 BACKGROUND
............................................................................................................. 4 3 .0 RESPONSIBILITIES.......................................................................................................... 6 4.0 PROCEDURE ................................................................................................................... 7
5.0 REFERENCES
.................................................................................................................. 27 5.1 Source Documents............................................................................................................. 27 5.2 Reference Documents .......................................................................................................27 5.3 Records .............................................................................................................................. 27 6.0 BASES ............................................................................................................................. 2 8 Page 2 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM 1.0 PURPOSE 1.1 Scope This document describes Point Beach Nuclear Plant (PBNP) requirements for implementation of a Reactor Vessel Internals Program for PBNP Units 1 and 2. The reactor vessel internals consist of two (2) basic assemblies: (1) an upper internals assembly which is removed dunhg each refueling outage to access the reactor core; and (2) a lower internals assembly which can be removed following a complete core unload.
The reactor vessel internals h c t i o n to:
a Provide support, guidance, and protection for the reactor core; 0 Provide a passageway for the distribution of the reactor coolant flow to the reactor core; Provide a passageway for support, guidance, and protection for control elements and in-vessel/core instrumentation; and a Provide gamma and neutron shielding for the reactor vessel.
This document implements a commitment to the NRC to manage the effects of aging for systems, structures, and components (SSC) within the scope of License Renewal (LR) as described in NP 7.7.25, PBNP Renewed License Program. LR Regulatory comnitments 4, 5, 6, 8,29, and 39 require the implementation of a Reactor Vessel Internals Program.
@-I, B-2)
This document demonstrates that the Reactor Vessel Internals Program meets the requirements of NUREG- 1801, Revision 1, "Generic Aging Lessons Learned (GALL)
Report,"Section XI.Ml6, "PWR Vessel Internals." This program was developed using EPRI MRP-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines," and NEI 03-08, "Guideline for the Management of Materials Issues." The Reactor Vessel Internals Program is a living document and will be revised periodically to reflect the latest plant configurations and industry experience.
1.2 Obiective The objectives of the Reactor Vessel Internals Program are to:
1.2.1 Demonstrate that the effects of aging on the Reactor Vessel Internals will be adequately managed for the period of extended operation in accordance with 10 CFR 54.
1.2.2 Summarize the role of existing PBNP Aging Management Programs (AMPS) in the Reactor Vessel Internals Program.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM 1.2.3 Participate in and implement the industry-defined (EPRVMRP and PWROG) pressurized water reactor (PWR) reactor vessel intemals requirements and guidance for managing aging of reactor intemals.
1.2.4 Provide an inspection plan summary for the PBNP reactor vessel internals.
2.0 BACKGROUND
The management of aging degradation effects in reactor vessel intemals is required for nuclear plants considering or entering license renewal, as specified in the NRC Standard Review Plan.
The U.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 guideliiles for managing the effects of aging within PWR reactor internals. In 1997, the Westinghouse Owners Group (WOG) issued WCAP-14577, "License Renewal Evaluation: Aging Management for Reactor Intemals," which was reissued as Revision l-A in 2001 after receiving NRC Staff review and approval. Later, an effort was engaged by the EPRI MRP to address the PWR internals aging management issue for the three currently operating U.S. reactor designs - Westinghouse, Combustion Engineering (CE), and Babcock & Wilcox (B&W).
The MRP first established a framework and strategy for the aging management of PWR intemals components using proven and familiar methods for inspection, monitoring, surveillance, and comunication. Based upon that framework and strategy, and on the accumulated industry research data, the following elements of an Aging Management Program were further developed.
s, Screening criteria were developed, considering chemical composition, neutron fluence exposure, temperature history, and representative stress levels, for determining the relative susceptibility of PWR intemals components to each of eight postulated aging mechanisms.
PWR intemals components were categorized, based on the screening criteria, into categories that ranged fkom:
- Components for which the effects from the postulated aging mechanisms are insignificant,
- Components that are moderately susceptible to the aging effects, and
- Components that are significantly susceptible to the aging effects.
Functionality assessments were performed, based on representative plant designs of PMlX intemals components and assemblies of components using irradiated and aged material properties, 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 examination timing, and the need and timing of subsequent inspections.
Items considered included component accessibility, operating experience, existing evaluations, and prior examination results.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM The industry effort, as coordinated by the EPRI MRP, has finalized initial Inspection and Evaluation Guidelines (I&E Guidelines) for reactor internals and submitted the document to the NRC with a request for a formal Safety Evaluation Report (SER). A supporting document addressing inspection requirements has been issued. The industry guidance is contained within two separate EPRI MRP documents:
a MRP-227, "PWR Internals Inspection and Evaluation Guidelines," (hereafter referred to as "the I&E Guidelines" or simply "MRP-227") provides the industry bacltground, listing of reactor internals components requiring inspection, type of NDE required for each component, timing for initial inspections, and criteria for evaluating inspection results. MRP-227 provides a standardized approach to PWR internals aging management for each unique reactor design (Westinghouse, B&W, and CE). The document was submitted to the NRC for a formal evaluation and review.
MRP-228, "Inspection Standard for Reactor Internals Components," provides guidance on the qualification/demonstration of the NDE techniques and other criteria pertaining to the actual performance of the inspections.
The Pressurized Water Reactor Owners Group (PWROG) has also recently begun efforts to develop "generic acceptance criteria" for the MRP-227 inspections, where feasible, for some of the reactor internals components. Final reports are to be developed and be available for industry use in support of planned license renewal inspection commitments. In some cases, individual plants will develop plant-specific acceptance criteria for some internals components where a generic approach is not practical.
The PBNP reactor vessel internals are integral with the reactor coolant system (RCS) of a Westinghouse two-loop nuclear steam supply system (NSSS). A typical illustration of which is provided in Attachment A, Figure 1.
As described in NUREG-1 839, the PBNP reactor vessel internals are designed to support, align, and guide the core components and to support and guide in-core instrumentation. The reactor vessel internals consist of two basic assemblies - an upper internals assembly that is removed during each refueling operation to obtain access to the reactor core, and a lower internals assembly that can be removed, if desired, following a complete core unload.
The lower internals assembly is supported in the vessel by resting on a ledge in the vessel flange region and is closely guided at the bottom by radial supportlclevis assemblies. The upper internals assembly is clamped at this same ledge by the reactor vessel head. The bottom of the upper internals assembly is closely guided by the core barrel alignment pins of the lower internals assembly.
The lower internals comprise the core barrel, thermal shield, core baffle assembly, lower core plate, intermediate diffuser plate, bottom support plate, and supporting structures. The upper internals package (upper core support structure) is a rigid member composed of the top support plate and deep beam sections, support columns, control rod guide tube assemblies, and the upper core plate. Upon upper internals assembly installation, the last three parts are physically located inside the core barrel.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM PBNP was granted a license for extended operation by the NRC through the issuance of a SER in NUREG-1839. In the SER, the NRC concluded that the PBNP License Renewal Application (LRA) adequately identified the reactor vessel internals systems, structures, and components that are subject to an aging management review (AMR), as required by 10 CFR 54.21(a)(l).
The U.S. industry, as noted tlxough the efforts of the MRP and PMrROG, has further investigated the components and subcomponents that require aging management to support continued reliable function. As designated by the protocols of NEI 03-08, "Guidelines for the Management of Materials Issues," each plant will be required to use MRP-227 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. MRP-227 was issued in December 2008, and plant AMPS must therefore be completed by December 201 1, or sooner, if required by plant-specific License Renewal commitments.
The categorization and analysis used in the development of MRP-227 are not intended to supersede any ASME B&PV Code Section XI requirements. Any components that are classified as core support structures, as defmed in ASME B&PV Code Section XI IWlB 2500, Category B-N-3, have requirements that remain in effect and may only be altered as allowed by 10 CFR 50.55a.
The information contained in this program fully complies with the requirements and guidance of the referenced documents. This program will manage aging effects of the reactor vessel internals so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation.
3.0 RESPONSIBILITIES The Nuclear Chief Operator Officer and Vice President, Nuclear Engineering Support are ultimately responsible for the successful implementation of the Reactor Vessel Internals Program.
The overall responsibility for the development, revision and implementation of the Reactor Vessel Internals Program resides with the PBNP Programs Engineering Department.
Responsibilities of the various interfacing groups are described below.
3.1 Programs Engineering Department 3.1.1 Preparation, maintenance and ownership of the Reactor Vessel Internals Program whiclz implement EPRI MRP requirements and guidance.
3.1.2 Development of refueling outage examination plans.
3.1.3 Development of a recommended strategy for the management of Reactor Vessel Internals materials.
3.1.4 Ensure compliance with regulatory requirements.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM 3.1.5 Serve as the contact for outside technical communications (NEI, INPO, NRC, EPRI, ASME, PWR Owners Group, etc.).
3.1.6 Participate in industry owners groups and industry investigations of aging affects applicable to reactor vessel internals.
3.1.7 Monitor and participate in industry initiatives with regard to bafflelformer and barrel/foimer bolt performance to support aging management for the Unit 1 bolting.
3.1.8 Incorporate applicable results of industry initiatives related to void swelling in the Reactor Vessel Internals Program.
3.1.9 Provide analysis and response to significant industry events.
3.1.10 Conduct periodic self-assessments of the Reactor Vessel Internals Program.
3.2 Design Engineering 3.2.1 Preparation of Design Change Packages (DCP) packages for repairs or modifications that would result in a configuration change to existing Reactor Vessel Internals components.
3.2.2 Disposition of Condition Reports associated with examination results.
4.0 PROCEDURE 4.1 Program Overview This reactor internals program utilizes a combination of prevention, mitigation, and condition monitoring. Where applicable, credit is taken for existing programs such as water chemistry, inspections prescribed by the ASME Section XI Inservice Inspection Program, thimble tube inspections, and past and future mitigation projects such as split pin replacements, combined with augmented inspectioils or evaluations as recommended by MRP-227.
Aging degradation mechanisms that impact internals have been identified and documented in PBNP Aging Management Reviews in support of License Renewal. The overall outcome of the reviews and the additional work performed by the industry, as summarized in MRP-227, is to provide appropriate augmented inspections for reactor internals components to provide early detection of the degradation mechanisms of concern. Therefore, this Reactor Vessel Internals Program is consistent with the existing PBNP AMR methodology and the additional industry work summarized in MRP-227.
All sources are consistent and address conceins about component degradation resulting from the following eight material aging degradation mechanisms identified as affecting reactor internals:
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009
=ACTOR VESSEL INTERNALS PROGRAM Stress Con-osion Cracking Stress corrosion cracking (SCC) refers to local, nonductile cracking of a material due to a combination of tensile stress, environment, and metallurgical properties. The actual mechanism that causes SCC involves a complex interaction of environmental and inetallurgical factors. The aging effect is cracling.
Irradiation-Assisted Stress Corrosion Cracking Irradiation-assisted stress corrosion cracking (IASCC) is a unique form of SCC that occurs only in highly irradiated components. The aging effect is cracking.
Wear Wear 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.
Fatigue Fatigue is defined as the structural deterioration that can occur as the result of repeated stresslstrain cycles caused by fluctuating loads and 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 craclc 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 craclc 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 craclc initiation and growth resistance are governed by a number of material, structural, 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 cracling.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM e Thermal Aging Embrittlernent Thermal aging embrittlement is the exposure of delta ferrite within cast austenitic stainless steel (CASS) and precipitation-hardenable (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 fiacture toughness.
o Irradiation Embrittlement 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 h c t i o n 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.
o Void Swelling and Irradiation Growth Void swelling is defined as a gradual increase in the volume of a component caused by formation of microscopic cavities in the material. These cavities result from the nucleation and growth of clusters 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
(>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 void swelling may result in cracking under stress.
o Thermal and Irradiation-Enhanced Stress Relaxation or Irradiation-Enhanced Creep The loss of preload aging effect can be caused by the aging mechanisms of stress relaxation or creep. Thermal stress relaxation (or primary creep) is defmed 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.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM Creep (or more precisely, secondary creep) is a slow, time- and temperature-dependent, plastic deforrnation of materials that can occur at stress levels below the yield strength (elastic limit). Creep occurs at elevated temperatures where continuous deformation takes place under constant strain. Secondary creep in austenitic stainless steels is associated with temperatures higher than those relevant to PWR intemals even after talcing into account gamma heating. However, irradiation-enhanced creep (or more simply, irradiation creep) or irradiation-enhanced stress relaxation (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 subsequent degradation by fatigue or wear and result in cracking.
The Reactor Vessel Intemals Program is focused on meeting the requirements of the 10 elements of an aging management program as described in NUREG-1801, GALL Report, Revision 1,Section XI.Ml6 for PWR Vessel Intemals. This is demonstrated through application of existing PBNP AMR methodology that credits inspections prescribed by the ASME Section XI Inservice Inspection Program, existing PBNP programs, and additional augmented inspections based on MRP-227 requirements and guidelines. A description of the applicable existing PBNP programs and compliance with the elements of the GALL is contained in the following subsections.
4.2 Existing PBNP Promams PBNP's overall strategy for managing aging in reactor intemals components is supported by the following existing programs:
- Water Chemistry Control Program
- ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program Thimble Tube Inspection 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,they will be managed under the existing programs.
Brief descriptions of the programs are included in the following subsections.
4.2.1 Water Chemistry Control Program The PBNP Water Chemistry Control Program is used to mitigate aging effects on component surfaces that are exposed to water as process fluid. Chemistry programs are used to control water chemistry for impurities that accelerate corrosion and contaminants that may cause cracking due to SCC. This program relies on monitoring and control of water chemistry to keep peak levels of various contaminants below the system-specific limits. The PBNP Water Chemistry Control Program is based on the current revision of EPRI PWR Primary Water Chemistry Guidelines. Later revisions of the guidelines will be used when issued. The limits imposed by the PBNP program meet the intent of the industry standard for addressing water chemistry.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 E A C T O R VESSEL INTERNALS PROGRAM 4.2.2 ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program The PBNP ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection (ISI) Program is implemented to monitor for aging effects such as cracking, loss of preload due to stress relaxation or irradiation creep, loss of material, and reduction of fracture toughness due to thermal embrittlement.
For PBNP, inspections conducted under the reactor internals program will be controlled as a combination of ASME Section XI IS1 exams on core support structures and augmented exams performed under that IS1 Program for the remaining reactor internals components addressed within MRP-227.
4.2.3 Thimble Tube Inspection Program Thimble tubes are long, slender, stainless steel tubes that are seal welded at one end with flux thimble tube plugs, which pass through the vessel penetration, through the lower internals assembly, and finally extend to the top of the fuel assembly. The bottom-mounted instrumentation @MI) colulnn assemblies provide a path for the flux thimbles into the core from the bottom of the vessel and protect the flux thimbles during operation of the reactor.
The thimble tube provides a path for the neutron flux detector into the core and is subject to reactor coolant pressure on the outside and containment pressure on the inside.
The PBNP thimble tube inspection program is an existing plant-specific program that satisfies NRC Bulletin 88-09 requirements that a tube wear inspection procedure be established and maintained for Westinghouse-supplied reactors that use bottom-mounted flux thimble tube instrumentation. The program follows Branch Technical Position RI;SB-1, Aging Management Review - Generic, which is included in Appendix A of NUREG-1800. The program includes eddy current testing requirements for thimble tubes and criteria for determining sample size, inspection frequency, flaw evaluation, and corrective action in accordance with NRC Bulletin 88-09.
4.2.4 License Renewal Programs The license renewal processes conducted at PBNP Units 1 and 2 created a number of programs to ensure that the integrity of structures and components is maintained throughout the periods of extended operation at both sites.
Specific programs concerning Reactor Vessel Internals materials include:
o LR-AMP-015-RVINT, Reactor Vessel Internals Program Basis Document for License Renewal e LR-AMP-001-WCHEM, Water Chemistry Control Program Basis Document for License Renewal Page 11 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM:
o LR-AMP-017-IWBCD, ASME Section XI, Subsections IWB, IWC, and IWD Insewice Inspection Program Basis Document for License Renewal e LR-AMP-006-TTI, "Thimble Tube Inspection Program Basis Document for License Renewal" 4.3 Joint Industry Issues Programs Applicable issues programs and their current examination requirements include:
4.3.1 WCAP-14577, Aging Management for Reactor Internals The Westinghouse Owners Group (WOG, now PVJROG) topical report WCAP-14577 contains a technical evaluation of aging degradation mechanisms and aging effects for Westinghouse RVI components. The WOG sent the report to the NRC staff to demonstrate that WOG member plant owners that subscribed to the WCAP could adequately manage effects of aging on RVI during the period of extended operation, using approved aging management methodologies of the WCAP to develop plant-specific aging management programs.
The aging management review for the PBNP internals was completed in accordance with the requirements of WCAP-14577.
4.3.2 MRP-227, Reactor Internals Inspection and Evaluation Guidelines MRP-227, as discussed in Section 2.0, was developed by a team of industry experts including utility representatives, NSSS vendors, independent consultants, and international cormnittee 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.
- a. MRP-227 RVI Component Categorizations MRP-227 used a screening and ranking process to aid in the identification of required inspections for specific RVI components. MRP-227 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 United States were evaluated in the MRP program; and appropriate inspection, evaluation, and implementation requirements for reactor internals were defined.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTMTIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL JNTERNALS PROGRAM Based on the completed evaluations, the PBNP RVI components are categorized within MRP-227 as "Primary" components, "Expansion" components, "Existing Programs" components, or No Additional Measures" components, as described as follows:
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.
Expansion Those PMrR inteillals that are highly or moderately susceptible to the effects of at least one of the eight aging mechanisms, but for which functionality assessment has shown a degree of tolerance to those effects, were placed in the Expansion group. The schedule for implementation of aging management requirements for Expansion components depends on the findings fiom the examinations of the Primary components at individual plants.
0 Existing Programs 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 managing those effects, were placed in the Existing Programs group.
e No Additional Measures Programs Those 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 mode, effects, and criticality analysis (FMECA) and the functionality assessment. No further action is required by these guidelines for managing the aging of the No Additional Measures components.
- b. NEI 03-08 Guidance Within MRP-227 The industry program requirements of MRP-227 are classified in accordance with the requirements of the NEI 03-08 protocols. The MRP-227 guideline includes Mandatory, Needed, and Good Practice elements as follows:
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 WACTOR VESSEL INTERNALS PROGRAM Mandatory There is one Mandatory element:
Each commercial U S . P WR unit shall develop and docunzent a PWR reactor internals aging nzanagenzent program within 36 nzonthsfollowing issuance of MRP-22 7, Rev. 0.
PBNP Applicability: This document meets this mandatory element.
Needed There are three Needed elements:
Each conznzercial U S . P WR unit shall implement MRP-227, Tables 4-1 through 4-9 and Tables 5-1 through 5-3for the applicable design within 24 ~nonths following issuance of MRP-227-A.
PBNP Applicability: Initial PBNP inspections have been scheduled to coincide with the 10-year interval IS1 program inspections. The applicable Westinghouse tables contained in MRP-227 are Table 4.3 (Primary), Table 4.6 (Expansion), and Table 4.9 (Existing), for PBNP, these tables are Attaclments B, C, and D, respectively.
Examinations specijied in the MRP-227 guidelines shall be conducted in accordance with Inspection Standard MW-228.
PBNP Applicability: Inspection standards developed under MRP-228 will be used for augmented inspection at PBNP as applicable where required by MRP-227 directives.
Examination results that do not nzeet the exanzination acceptance criteria defined in Section 5 of the MW-227 guidelines shall be recorded and entered in the plant corrective action program and dispositioned.
PBNP Applicability: PBNP will comply with this requirement.
Good Practice There is one Good Practice element:
Each conznzercial U S . PWR unit shouldprovide a sumrnary report of all inspections and monitoring, itenzs requiring evaluation, and new repairs to the M W Program Manager within 120 days of the conzpletion of an outage during which PWR internals are examined. The MRP tenzplate should be usedfor the report.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM PBNP Applicability: PBNP will participate in future industry efforts and will adhere to industry directives for reporting, response, and follow-up.
- c. MRP-227 Applicability to PBNP The applicability of MRP-227 to PBNP requires compliance with the followh~gMRP-227 assumptions:
0 Operation of 30 years or less with high-leakage core loading patterns (fiesh fuel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel management strategy for the remaining 30 years of operation.
PBNP Applicability: A low leakage fuel management strategy was implemented starting with fuel cycle 8 for Unit 1 and fuel cycle 6 for Unit 2. The core loading pattern was changed significantly prior to 30 years of operation.
0 Base load operation, i.e., typically operates at fixed power levels and does not usually vary power on a calendar or load demand schedule.
PBNP Applicability: Early in operating life, PBNP followed load as needed depending upon Wisconsin Electric Power Company system requirements. Load following operations diminished in the late 1970's as system requirements changed. Since this time, PBNP has typically operated at a fixed power level. The resulting load follow cycles are well within the reactor vessel internal's design basis transient analysis.
The existence of the load following history will not affect the component categorizations nor recommended inspections since PBNP has been operated within its design and licensing basis.
In addition, other conservatisms exist in that PBNP only operated with a high leakage core for 7 cycles on Unit 1 and 5 cycles on Unit 2, which is significantly less than the 30 years assumed in MRP-227.
Thus, MRP-227 is applicable to PBNP.
0 No design changes beyond those identified in general industry guidance or recommended by the original vendors.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM PBNP Applicability: PBNP has made modifications to the reactor internals. These modifications were all performed with the involvement of Westinghouse, the reactor vessel internals designer.
MRP-227 states that the recomrnendations are applicable to all U.S.
PWR operating plants as of May 2007 for the three designs considered. PBNP has not made any modifications to reactor internals components since May 2007 other than replacement of split pins for Unit 1 with an upgraded material in 2008. The split pin replacement was performed by Westinghouse. The modification will have no impact on the applicability of MRP-227 and is an example of the PBNP proactive approach to managing aging reactor internals.
Based on the above, MRP-227 is applicable for PBNP 4.3.3 Ongoing Industry Programs The U.S. industry, through both the EPRVMRP and the PWROG, continues to sponsor activities related to 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. PBNP will maintain cognizance of industry activities related to PWR internals inspection and aging management; and will address/implement industry guidance, stemming from those activities, as appropriate under NEI 03-08 practices.
4.3.4 Point Beach Reactor Internals Aging Management Program Attributes The ley elements of the previously discussed aging management activities, which are used in the Reactor Vessel Internals Program, are described below.
The results of an evaluation of each key element against NUREG-1801, "Generic Aging Lessons Learned (GALL) Report," Sections XI.Ml6, "PWR Vessel Internals," is also provided below.
4.3.5 GALL Element 1: Scope of Program The PBNP reactor vessel internals consist of two basic assemblies: (1) an upper internals assembly that is removed during each refueling operation to obtain access to the reactor core, and (2) a lower internals assembly that can be removed, if desired, following a complete core unload.
Page 16 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009
=ACTOR VESSEL INTERNALS PROGRAM The results of the industry research provided by MRP-227, summarized in the tables of Attachments B, C and D 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 frequency of inspection, and examination acceptance criteria. The PBNP Reactor Vessel Internals Program scope is based on previously established and approved GALL Report approaches through application of the WCAP-14577 methodologies to determine those components that require aging management. Lilcewise, the additional information provided in the industry MRP-227 is rooted in the GALL methodology and provides a basis for augmented inspections that were required to complete this PBNP Reactor Vessel Internals Program by providing the inspection method, frequency of inspection, and examination acceptance criteria.
This program also credits the One-Time Inspection Program for the management of stress relaxation of the lower intenials hold-down spring.
This element is consistent with the corresponding NUREG-1801 aging management program elements.
4.3.6 GALL Element 2: Preventative Actions The Water Chemistry Control Program is credited for monitoring and control of reactor coolant water chemistry to prevent or mitigate tlie effects of SCC and IASCC. The PBNP Water Chemistry Control Program is based on the EPRI PWR Water Chemistry Guidelines.
The PBNP Unit 1 and 2 split pins were replaced witli a 3 16 stainless steel material in 2008 and 2005, respectively. A portion of the baffle former bolts in the PBNP Unit 2 internals have been replaced with a more crack resistant material. There are no preventive actions to mitigate thermal aging, neutron irradiation embrittlement or void swelling.
This element is consistent with the corresponding NUREG-1 801 aging management program elements.
4.3.7 GALL Element 3: Parameters Monitored or Inspected The program monitors, inspects, and/or tests for the effects of the eight aging degradation inechanisms on the intended fimction of the reactor vessel internals components through condition monitoring activities in accordance with the augmented requirements defined under industry directives as contained in MRP-227 and ASME Section XI.
Page 17 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM The ASME Section XI Program consists of periodic volumetric, surface, and/or visual examination of components for assessment, signs of degradation, and corrective actions. Attachments ByC, and D provide a detailed listing of the components and subcomponents and the parameters monitored, inspected, and/or tested.
This element is consistent with the corresponding NUREG-1801 aging management program elements.
4.3.8 GALL Element 4: Detection of Aging Effects Detection of indications that are required by the ASME Section XI IS1 Program is well established and field-proven through the application of the Section XI IS1 Program Those augmented inspections that are taken from the MRP-227 recommendations will be applied through use of the MRP-228 Inspection Standard.
Inspection can be used to detect physical effects of degradation including craclcing, 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 measurement. Three different visual techniques include VT-3, VT-1, and EVT-1. 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- 1 Visual Examinations 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 IW-B-3520. VT-1 visual examination is intended to identify crack-like surface flaws. Unacceptable conditions for a VT-1 examination are:
- Cracl-like surface flaws on the welds joining the attachment to the vessel wall that exceed the allowable linear flaw standards of IW-B-3510.
o 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. No additional VT-1 inspections over and above those required by ASME Section XI IS1 have been specified in this program.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM EVT-1 Enllanced Visual Examination for the Detection of Surface Breaking Flaws In the augmented inspections detailed in tlze MRP-227 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 technique. The EVT-1 specification augments the VT-1 requirements to provide more rigorous inspection standards for stress corrosion cracking 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-1) with additional requirements (such as camera scanning speed) currently being developed by the industry. Any recommendation for EVT-1 inspection will require additional analysis to establish flaw-tolerance criteria. The industry is currently developing a consensus approach for acceptance criteria methodologies to support plant-specific augmented examinations. PBNP has been an active participant in these initiatives and will follow the industry directive. These acceptance criteria methodologies may be determined either generically or on a plant-specific basis because both loads and component dimensions may vary fiom plant to plant within a typical PWR design.
VT-3 Examination for General Condition Monitoring In the augmented inspections detailed in the MRP-227 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 cracling 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 (such as baffle-edge bolts), where a single failure does not co~npromisethe 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 IWB-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; Page 19 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM Loose, missing, craclced, or fractured parts, bolting, or fasteners; B) 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 thiclcness by more than 5 percent; e Wear of mating surfaces that may lead to loss of function; sr Structural degradation of interior attachments such that the original cross-sectional area is reduced more than 5 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 Testing 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 ultrasoilic 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 assumed 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 through 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.
Page 20 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM Failure of a single bolt does not compromise the function of the entire assembly. Bolting systems in the reactor 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 of 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 demonstrate that the system will maintain reliability and integrity in continuing to perform the intended function of the component.
This analysis is highly plant-specific. Therefore, any recommendation for UT inspection of bolts assumes that the plant owner will work with the designer to establish minimum acceptable bolting patterns prior to the inspection to support continued operation. For Westinghouse-designed plants, minimum acceptable bolting patterns for baffle-former and barrel-former bolts are available through the PWROG. PBNP has been a full participant in the development of the PWROG documents and has full access and use.
This element is consistent with the coi-responding NUREG-1801 aging management program elements.
4.3.9 GALL Element 5: Monitoring and Trending Operating experience with PWR reactor internals has been generally proactive. Flux thimble wear and control rod guide tube split pin cracking issues were identified by the industry and continue to be actively managed.
The extremely low frequency of failure in reactor internals inakes monitoring and trending based on operating kxperience somewhat impractical. The majority of the materials aging degradation models used to develop the MRP-227 Guidelines are based on test data from reactor internals components removed from service. The data is 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. PBNP has in the past and will continue to maintain cognizance of industry activities and shared information related to PWR internals inspection and aging management.
Inspections credited in Attachment D are based on utilizing the PBNP ASME Section XI Inservice Inspection Program and the augmented inspections derived from the industry program contained in Attachments B and C. These inspections, where practical, will be scheduled to be conducted in conjunction with typical 10-year interval IS1 examinations.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM Attachments ByC and D 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, inspection of the "Primary" components provides reasonable assurance for demonstrating component current capacity to perform the intended functions.
Reporting requirements are included as part of the MRP-227 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.
Tlis element is consistent with the col-responding NUREG-1801 aging management program elements.
4.3.10 GALL Element 6: Acceptance Criteria For examinations conducted in accordance with ASME Section XI, indications or relevant conditions of degradation detected will be evaluated in accordance with IWB-3 100, which refers to acceptance standards contained in IWB-3400 and IWB-3500.
Inspection acceptance and expansion criteria are provided in Attachment E.
These criteria will be reviewed periodically as the industry continues to develop and refine the information and will be included in updates to PBNP procedures to enable the examiner to identify examination acceptance criteria considering state-of-the-art information and techniques.
Augmented inspections, as defined by the MRP-227 requirements, 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. An example of an analytical evaluation is using a minimum bolting WCAP approach such as those cornrnonly used to support continued component or assembly functionality. Additional analysis to establish acceptable bolting pattern evaluation criteria for the baffle-former bolt assembly, as contained in various industry documents, is also considered in determining the acceptance of inspection results to support continued component or assembly functionality. The industry, through various cooperative efforts, is working to construct a consensus set of tools in line with accepted and proven methodologies to support this element. Additional analysis to establish Attachment E expansion component evaluation criteria is being performed through the efforts of the PWROG. Status is monitored through direct PBNP cognizance of industry (including PWROG) activities related to PWR internals inspection and aging management.
This element is consistent with the corresponding NUREG-1801 aging management program elements.
Page 22 of 50
POINT BEACH NUCLEAR PLANT Ah4 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM 4.3.1 1 GALL Element 7: Corrective Actions Corrective actions are implemented in accordance with the requirements of 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," and FPL-1, Quality Assurance Topical Report. The PBNP corrective action program includes non-safety related structures, systems, and components."
This element is consistent with the corresponding NUREG-1801 aging management program elements.
4.3.12 GALL Element 8: Confmation Process The confirmation process is part of the corrective action program, which is implemented in accordance with the requirements of 10 CFR 50, Appendix B and FPL-1, Quality Assurance Topical Report. The PBNP corrective action program includes non-safety related structures, systems, and components."
This element is consistent with the corresponding NUREG-1801 aging management program elements.
4.3.13 GALL Element 9: Administrative Controls The Reactor Vessel Internals Program is implemented through the documents discussed in Section 5.2. These implementing documents are subject to administrative controls, including a formal review and approval process, in accordance with the requirements of 10 CFR 50, Appendix B and FPL-1, Quality Assurance Topical Report.
This element is consistent with the corresponding NUREG-1801 aging management program elements.
4.3.14 GALL Element 10: Operating Experience Extensive industry and PBNP operating experience has been reviewed during the development of the Reactor Vessel Internals Aging Management Program.
The experience reviewed includes NRC Information Notices 84-18, "Stress Corrosion Cracking in PWR Systems" and 98-1 1, "Cracking of Reactor Vessel Inteinal Baffle Former Bolts in Foreign Plants." Most of the industry operating experience reviewed has involved cracking of austenitic stainless steel baffle-former bolts, or SCC of high-strength internals bolting. SCC of control rod guide tube split pins has also been reported.
A review of plant-specific operating experience with reactor vessel internals reveals that PBNP has responded to industry operating experience regarding reactor vessel internals degradation. Examples that demonstrate PBNPYs response to industry operating experience with RVI are described below:
Page 23 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTFWTIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNUS PROGRAM PBNP Control Rod Guide Tubes Split Pins The control rod guide tube split pins were replaced at PBNP Units 1 and 2.
The new pins are a material upgrade from Inconel X-750 to 3 16SS in support of managing aging in the component.
Fuel Damage Due to Baffle Gap Water Jetting -Conversion to Upflow Design Both Point Beach units have experienced fuel rod damage due to baffle gap jetting. Both units were placed in service with a c'downflow" design for coolant flow in the former region between the baffle plates and the core barrel.
This design minimizes core bypass flow, but exhibits a fairly large pressure differential across the baffle plates near the top of the core. The baffle plates at the two units are of both a buttjointed bolted configuration and a customized joint configuration.
One fuel pin failure at PBNP Unit 1 in 1975 was mentioned in NRC IE Circular 80-17, ccFuelPin Damage due to Water Jet from Baffle Plate Comer."
This same document describes the peening techniques that were used to reduce the gaps between baffle segments. The fuel rod damage was limited to only one of eight fuel assemblies in core locations adjacent to identified susceptible joints.
Baffle plateljoint peening at PBNP was done in accordance with:
8 Baffle Peening (Unit I), Westinghouse Procedure ccMRS2.3.1 Gen-1,"
Revision 1 8 Baffle Peening (Unit 2), Westinghouse Procedure "MRS 2.3.1 Gen-1,"
Revision 2 While peening the susceptible locations appeared to be an effective solution to the baffle jetting problem, subsequent fuel damage was experienced after several years.
At that time, a more effective solution was implemented by converting the bafflebarrel flow design to the ccupflowy~ configuration. This modification effectively eliminated the pressure differential across the baffle joints which caused the jetting problem. The upflow modifications were performed in the 1986-1987 time frame in accordance with:
8 PBNP Modification 86-058 (Unit 1) - RV lower intemals upflow conversion.
8 PBNP Modification 86-059 (Unit 2) - RV lower intemals upflow conversion.
Page 24 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 E A C T O R VESSEL INTERNALS PROGRAM Installation of Flexureless hserts The original internals configuration at both units incorporated removable inserts at the top of the control rod guide tubes to minimize bypass flow from the core outlet region to the upper head region. These inserts were attached to the upper guide tube housing plate by four flexures, which were manufactured from Inconel X-750. These flexures, over time, exhibited craclcing due to primary water stress corrosion cracking, and a modified "flexureless" design was developed as a replacement.
Replacement of the original inserts at Point Beach has been performed in accordance with:
o PBNP Modification 84-235 (Unit 1) - Remove flexures and install flexureless inserts.
o PBNP Modification 84-236 (Unit 2) - Remove flexures and install flexureless inserts.
Examination and Replacement of Baffle to Former Bolts at PBNP Unit 2 In 1997, the Westinghouse Owners Group (WOG) formed a task group to investigate the issue of reactor vessel internals bolting integrity. A small group of domestic utilities that operate Westinghouse two-loop plants collectively developed a proactive plan for inspection and replacement of bolts that join the baffle to former plates in the reactor lower internals. A n augmented examination via UT was conducted on the baffle-former bolts of PBNP Unit 2. The UT examination identified a number of bolts with indications indicative of crack like flaws. A number of bolts sufficient to guarantee the structural margins of the baffle-former joints were replaced, including all bolts with UT indications. The replacement bolts are fabricated from a more LASCC-resistant material.
PBNP Unit 2 Hold-Down Spring Evaluation The preload from the PBNP Unit 2 core barrel hold-down spring was evaluated to justify extending the use of the current core barrel hold-down spring for the period of extended operation. During the October 2006 outage, Westinghouse measured the height of the core barrel hold-down spring for PBNP Unit 2.
The material for the PBNP Unit 2 hold-down spring is quenched and tempered modified Type 403 martensitic stainless steel per Westinghouse Specification PDS-10725-HA, Based on the new and as-found core barrel hold-down spring height measurements, the lower internals hold down capacity of PBNP Unit 2 was shown to be adequate for an increased design life of 60 years.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATTIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM PWROG Control Rod Guide Cards Inspection Program at PBNP Unit 1 The PWROG is currently conducting upper intelnals control rod guide tube card wear measurements on a sample of guide tubes from selected representative pilot plants to approximate the remaining life of the guide tube guide cards. This is a proactive effort by the U.S. industry to establish criteria for inspection and gather data to support aging management of the component.
PBNP proactively inspected guide cards in Unit 1 during the 2008 outage as one of the representative pilot inspection plants. Inspection outcomes are being evaluated to ensure compliance with MRP-227 specifications.
Industry operating experience is routinely reviewed by PBNP engineers using INPO Operating Experience (OE), the Nuclear Network, and other information sources as directed under the applicable procedure, for the determination of additional actions and lessons learned. These insiglits, as applicable, can be incorporated into the plant systems quarterly health reports and further evaluated for incorporation into plant programs.
A key element of the MRP-227 Guideline is the reporting of age-related degradation of reactor vessel components. PBNP, through its participation in PWROG and EPRI-MRP activities, will continue to benefit from the reporting of inspection infomation and will share its own operating experience with the industry through those groups or INPO, as appropriate.
PBNP will continue to participate in industry groups studying RVI materials degradation issues, such as the EPRI MRP RI-ITG and Westinghouse Owner's Group (WOG), so as to gain the full benefit of the data that will be generated. The WOG also sponsors research related to RVI degradation issues that are specific to Westinghouse-built PWRs. As new infollnation and technology becomes available, the Reactor Vessel Internals Program will be modified to incorporate enhanced inspections of appropriate components as necessary. As such, PBNP committed to submit the Reactor Vessel Internals Program to the NRC for review and approval one year prior to entering the period of extended operation.
A review of NRC Inspection Reports, QA AuditJSurveillance Reports, and Self-Assessments since 1999 revealed no issues or findings that could impact the effectiveness of the Reactor Vessel Internals Program. As additional operating experience is obtained, lessons learned may be used to adjust this program.
This element is consistent with the corresponding NUREG-1801 aging management program elements.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM 4.3.15 Sunmary The Reactor Vessel Internals Program complies with or exceeds the corresponding aging management attribute in NUREG- 1801, "Generic Aging Lessons Learned (GALL) Report,"Section XI.Ml6, "PWR Vessel Internals."
5.0 REFERENCES
5.1 Source Documents 5.1.1 WCAP-15960, "Aging Management Review Report for the Point Beach Reactor Vessel Internals" 5.1.2 LR-AMR-0 127-RC, License Renewal Aging Management Review Report, Reactor Vessel Intemals 5.1.3 LR-AMP-015-RVINT, Reactor Vessel Intemals Program Basis Document for License Renewal 5.1.4 Westinghouse Letter WEP-02-8, "Point Beach Units 1 and 2 Reactor Internals CMTR Summary," dated September 12,2002 5.2 Reference Documents 5.2.1 NEI 03-08, "Guideline for the Management of Materials Issues" 5.2.2 EPRI MRP-227, "Materials Reliability Program: Pressurized Water Reactor Intemals Inspection and Evaluation Guidelines" 5.2.3 ER-AA-105, "Reactor Coolant Systein Materials Degradation Management Program (RCS MDMP)"
5.2.4 NP 7.7.25, PBNP Renewed License Program 5.2.5 NDE-756, Remote Visual Examination of Reactor Pressure Vessel Interior and Components 5.2.6 NUREG-1801, "Generic Aging Lessons Learned (GALL) Report,"
Revision 1, September, 2005 5.3 Records None.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM 6.0 BASES B-1 LR-AMP-015-RVINT, Reactor Vessel Internals Program Basis Document for License Renewal B-2 NUREG-1839, "US NRC Safety Evaluation Report Related to the License Renewal of the Point Beach Nuclear Plant, Unit 1 and 2" Page 28 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHMENT A FIGURE 1 Overview Of Typical Westinghouse Internals ROD TRAVEL HOUSING INSTRUMENTATION PORTS THERMAL SLEEVE LIFTING LUG CLOSURE HEAD ASSEMBLY HOLD-DOWN SPRlNG CONTROL ROD GUIDE TUBE CONTROL ROD DRIVE SHAFT INLET NOZZLE CONTROL ROD CLUSTER (WITHDRAWI ACCESS PORT REACTOR VESSEL LOWER CORE PLATE Page 29 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACrnNT A FIGURE 2 Westinghouse Control Rod Guide Card (PBNP 14 X 14 Fuel Assembly)
Page 30 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHMENT A FIGURE 3 Typical Westinghouse Control Rod Guide Tube Assembly Lower Flags Weld Page 3 1 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE! MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHMENT A FIGURE 4 Major Fabrication Welds In Typical Westinghouse Core Barrel COP mmi to Support Plae Weld Page 32 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHMENT A FIGURE 5 Bolt Locations In Typical Westingl~ouseBaffle-Former-Barrel Structure Page 33 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL, Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHMENT A FIGURE 6 Baffle-Edge Bolt And Baffle-Former Bolt Locations At High Fluence Seams In Bolted Baffle-Former Assembly Page 34 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revisio~i0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACrnNT A FIGURE 7 High Fluelice Sean Locations In Westinghouse Baffle-Former Assembly High Flucncc Scams
' "r Page 35 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHMENT A FIGURE 8 Exaggerated View Of Void Swelling Induced Distortion In Westinghouse Baffle-Former Assembly Potenlid Gaps at Bd&-Formcr Plale Levels Polcntlal Bmlng Along High FIuence Seam Page 36 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 ATTACE-IMENT A FIGURE 9 Vertical Displacement Of Westinghosue Baffle Plates Caused By Void Swelling FIGURE 10 Schematic Cross-Sections Of The Westinghouse Hold-Down Springs TOP SUPPORT PLATE Page 37 of SO
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHMENT A FIGURE 11 Location Of Westinghouse Thennal Slseld Flexures
.ma1 Shield Care Support Page 38 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACrnNT A FIGURE 12 Schematic Indicating Location Of Westinghouse Lower Core Support Structure. Additional Details Shown In Figure 13 FIGURE 13 Westinghouse Lower Core Support Structure And Bottom Mounted Instrumentation Columns. Core Support Column Bolts Fasten The Core Support Columns To The Lower Core Plate.
LOWER CORE PLATE DIFFUSER PLATE CORE SUPPORT PLATEFORGING BOTTOM MOUNTED IWSTRUMENTATION COLUMN Page 39 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNAL8 PROGRAM ATTACHlMENT A FIGURE 14 Typical Westinghouse Core Support Column. Core Support Column Bolts Fasten The Top Of The Support Column To The Lower Core Plate FIGURE 15 Examples Of Westinghouse Bottom Mounted Instrumentation Column Designs Page 40 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHIVIENT A FIGURE 16 Typical Westingl~ouseThermal Shield Flexure Page 41 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACrnNT B WESTINGHOUSE PLANTS PRIMARY COMPONENTS Expansion Link Examination Method/Frequency Item Schedule Effect (Mechanism) Examination Coverage (Note 1) (Note 1)
Control Rod Guide Unit 1 - 2008 Loss of Material (Wear) None Visual (VT-3) examination no later than 2 20% examination of the Tube Assembly refueling outages from the beginning of the number of CRGT assemblies, Guide plates (cards) Unit 2 - 2015 license renewal period, and no earlier than two with all guide cards within refueling outages prior to the start of the license each selected CRGT renewal period. Subsequent examinations are assembly examined.
required on a ten-year interval. See Figure 2 Control Rod Guide Unit 1 - 2013 Cracking (SCC, Fatigue) Bottom-mounted Enhanced visual (EVT-1) examination to 100% of outer (accessible)
Tube Assembly instrumentation determine the presence of crack-like surface CRGT lower flange weld Lower flange welds Unit 2 - 2015 (BMI) column flaws in flange welds no later than 2 refueling surfaces and adjacent base bodies, outages from the beginning of the license metal.
Lower support renewal period and subsequent examination on a See Figure 3 column bodies ten-year interval.
(cast)
Core Barrel Unit I -2013 Cracking (SCC) Remaining core Periodic enhanced visual (EVT-1) examination, 100% of one side of the Assembly barrel welds, no later than 2 refueling outages &om the accessible surfaces of the Upper core barrel Unit 2 - 2015 Lower support beginning of the license renewal period and selected weld and adjacent flange weld column bodies subsequent examination on a ten-year interval. base metal.
(non cast) See Figure 4 Baffle-Former Unitl-2013 Cracking(LASCC,Fatigue)that None Visual (VT-3) examination, with baseline Bolts and locking devices on Assembly results in examination between 20 and 40 EFPY and high fluence seams. 100% of Baffle-edge bolts -
Unit 2 2015 0 Lost or broken locking devices subsequent examinations on a ten-year interval. components accessible from 0 Failed or missing bolts core side.
0 Protrusion of bolt heads See Figure 5 Page 42 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHMENT B
'WESTINGHOUSE PLANTS PRIMARY COMPONENTS Expansion Link Examination Method/Frequency Item Schedule Effect (Mechanism) Examination Coverage (Note 1) (Note 1)
Baffle-Former Unit 1 - 2013 Cracking (IASCC, Fatigue) Lower support Baseline volumetric (UT) examination between 100% of accessible bolts or as Assembly column bolts, 25 and 35 EFPY, with subsequent examination supported by plant-specific Baffle-former bolts. Unit 2 - 2015 Barrel-Former after 10 to 15 additional EFPY to confirm justification. Heads accessible bolts stability of bolting pattern. Re-examination for from the core side. UT high-leakage core designs requires continuing accessibility may be affected examinations on a ten-year interval. by complexity of head and locking device designs.
See Figures 5 and 6 Baffle-Former Unit 1 - 2013 Distortion (Void Swelling), or None Visual (VT-3) examination to check for Core side surface as Assembly Cracking (IASCC) that results in: evidence of distortion, with baseline indicated.
Assembly Unit 2 - 20 15
- Abnormal interaction with fuel examination between 20 and 40 EFPY and See Figures 6,7, 8 and 9 assemblies subsequent examinations on a ten-year interval.
e Gaps along high fluence baffle joint 0 Vertical displacement of baffle plates near high fluencejoint s Broken or damaged edge bolt locking systems along high fluence baffle joint.
Alignment and NIA - Point Distortion (Loss of Load) None NIA NIA Interfacing Beach hold See Figure 10 Components down springs Intemals hold down are 403 SS.
spring (304 SS)
Thermal Shield Unit 1 - 2013 Cracking (Fatigue) or Loss of None Visual (VT-3) no later than 2 refueling outages 100% of thermal shield Assembly Thermal Material (Wear) that results in from the beginning of the license renewal flexures.
shield flexures Unit 2 - 20 15 thermal shield flexures excessive period. Subsequent examinations on a ten-year See Figures 11 and 16 wear, fracture, or complete interval.
separation Notes: 1. Examination acceptance criteria and expansion criteria are in Attaclment E.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHMENT C WESTINGHOUSE PLANTS EXPANSION COMPONENTS Expansion Link Examination Method/Frequency Item Schedule Effect (Mechanism) Examination Coverage (Note 1) (Note 1)
Core Barrel Based on Cracking (IASCC, Fatigue) Baffle-former bolts Volumetric (UT) examination, with initial and 100% of accessible bolts.
Assembly results of subsequent examinations dependent on results Accessibility may be limited Barrel-former bolts Primary Link of baffle-former bolt examinations. by presence of thermal Components shields of neutron pads.
See Figure 5 Lower Support Based on Cracking (IASCC, Fatigue) Baffle-formerbolts Volumetric (UT) examination, with initial and 100% of accessible bolts or as Assembly results of subsequent examinations dependent on results supported by plant-specific Lower support column Primary Link of baffle-former bolt examinations. justification.
bolts Components See Figures 12 and 13 Core Barrel Based on Craclcing (SCC, Fatigue) Upper core barrel Enhanced visual (EVT-1) examination, with 100% of one side of the Assembly results of flange weld initial examination and re-examination accessible surfaces of the Core barrel flange, Primary Link frequency dependent on the examination results selected weld and adjacent Core barrel outlet Components for upper core barrel flange. base metal.
nozzles, Lower core barrel flange weld See Figure 4 Lower Support Based on Cracking (IASCC) Upper core barrel Enhanced visual (EVT-I) examination, with 100% of accessible surfaces.
Assembly results of flange weld initial examination and re-examination Lower support column Primary Link kequency dependent on the examination results See Figures 13 and 14 bodies (non cast) Components for upper core barrel flange weld.
Lower Support Based on Cracking (MSCC) including the Control rod guide Visual (EVT-1) examination 100% of accessible support Assembly results of detection of fractured support tube (CRGT) lower columns.
Lower support column Primary Link columns flanges bodies (cast) Components See Figures 13 and 14 Page 44 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 WACTOR VESSEL INTERNALS PROGRAM ATTACHMENT C MESTINGHOUSE PLANTS EXPANSION COMPONENTS Expansion Link Examination Methodmrequency Item Schedule Effect (Mechanism) Examination Coverage (Note 1) (Note 1)
Bottom Mounted Based on Cracking (Fatigue) including the Control rod guide Visual (VT-3) examination of BMI column 100% of BMI column bodies Instrumentation results of detection of completely fractured tube (CRGT) lower bodies as indicated by difficulty of for which difficulty is System Primary Link column bodies flanges insertiodwithdrawal of flux thimbles. Flux detected during flux thimble Bottom-mounted Components thimble insertiodwithdrawal to be monitored at insertiodwithdrawal.
instrumentation (BMI) each inspection interval.
column bodies See Figures 13 and 15 Notes: 1. Examination acceptance criteria in Attachment E Page 45 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHhENT D WESTINGHOUSE PLANTS EXISTING PROGRAM COMPONENTS Item Schedule Effect (Mechanism) Primary Link Examination Methodmrequency Examination Coverage Core Barrel Unit I - Spring Loss of material (Wear) ASME Code Visual (VT-3) examination to determine All accessible surfaces at Assembly 2010 Section XI general condition for excessive wear. specified frequency.
Core barrel flange Unit 2 - Fall 2009 Upper lnternals Unit 1 - Spring Cracking (SCC, Fatigue) ASME Code Visual (VT-3) examination All accessible surfaces at Assembly 2010 Section XI specified frequency.
Upper support ring or Unit 2 - Fall skirt 2009 Lower Internals Unit 1- Spring Cracking (IASCC, Fatigue) ASME Code Visual (VT-3) examination of the lower core All accessible surfaces at Assembly 2010 Section XI plates to detect evidence of distortion andlor specified frequency.
Lower core plate XI, Unit 2 -Fall loss of bolt integrity.
lower core plate 2009 (XL= 14ft. Core)
(Note 1) p p p p Lower lnternals Unit 1 - Spring Loss of material (Wear) ASME Code Visual (VT-3) examination All accessible surfaces at Assembly 2010 Section XI specified frequency.
Lower core plate XL Unit 2 -Fall lower core plate 2009 (XL= 14ft. Core)
(Note 1 )
Bottom Mounted Unit 1 - Spring Loss of material (Wear) NUREG- 1801 Rev. Surface (ET) examination Eddy current surface Instrumentation 2010 1 examination as defined in System Unit 2 - Fall plant response to DEB 88-09.
Flux thimble tubes 2009 Alignment and Unit 1- Spring Loss of material (Wear) ASME Code Visual (VT-3) examination All accessible surfaces at Interfacing 2010 Section XI specified frequency.
Components Unit 2 - Fall (Note 2)
Clevis insert bolts 2009 Alignment and Unit 1 - Spring Loss of material (Wear) ASME Code Visual (VT-3) examination All accessible surfaces at Interfacing 2010 Section XI specified frequency.
Components Unit 2 - Fall Upper core plate 2009 alignment pins Notes: 1. XL = "Extra Long" referring to Westinghouse plants with 14-foot cores.
- 2. Bolt was screened in because of stress relaxation and associated cracking; however, wear of the clevis/insert is the issue.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHMENT E ACCEPTANCE CRITERIA AND EXPANSION CRITERIA Examination Additional Examination Expansion Item Shown In Acceptance Criteria Expansion Criteria Acceptance Criteria Links(s)
(Note 1)
Control Rod Guide Figure 2 Visual (VT-3) examination None NIA NIA Tube Assembly The specific relevant condition is Guide plates (cards) wear that could lead to loss of control rod alignment and impede control assembly insetion.
Control Rod Guide Figure 3 Enhanced visual (EVT-1) a. Bottom-mounted a. Confirmation of surface-breaking indications a. For BMI column bodies, Tube Assembly examination instrumentation in two or more CRGT lower flange welds, the specific relevant Lower flange welds The specific relevant condition is (BMI) column combined with flux thimble condition for the VT-3 a detectable crack-like surface bodies insetion/withdrawal dEculty, shall require examination is completely indication. b. Lower support visual (VT-3) examination of BMI column kctured column bodies.
column bodies bodies by the completion of the next (cast) refueling outage. b. For cast lower support column bodies, the
- b. Confirmation of surface-breaking indications specific relevant condition in two or more CRGT lower flange welds is a detectable crack-like shall require EVT-1 examination of cast surface indication.
lower support column bodies within three fuel cycles following the initial obsemation.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTFUTIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHMENT E ACCEPTANCE CRITERIA AND EXPANSION CRITERIA Examination Additional Examination Expansion Item Shown In Acceptance Criteria Expansion Criteria Links(s) Acceptance Criteria (Note 1)
Core Barrel Figure 4 Periodic enhanced visual (EVT-1) a. Remaining core a. The confirmed detection and sizing of a a and b. The specific relevant Assembly examination. barrel welds surFace-breaking indication with a length condition is a Upper core barrel greater than two inches in the upper core detectable crack-like The specific relevant condition is barrel flange weld shall require that the surface indication.
flange weld b. Lower support a detectable crack-like surface column bodies EVT-1 examination, and any supplementary indication. (non cast) UT examination, be expanded to include the core barrel-to-support plate weld by the completion of the next refbeling outage. If extensive confirmed indications in the core barrel-to-support plate weld are detected, further expansion of the EVT-1 examination shall include the remaining core barrel assembly welds.
- b. If extensive cracking in the remaining core barrel welds is detected, EVT-1 examination shall be expanded to include the upper six inches of the accessible surfaces of the noncast lower support column bodies within three fuel cycles following the initial observation.
Baffle-Former Figures 5 and Visual (VT-3) examination. None NIA NIA Assembly 6 Baffle-edge bolt The specific relevant conditions are missing or broken locking devices, failed or missing bolts, and protrusion of bolt heads.
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POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 REACTOR VESSEL INTERNALS PROGRAM ATTACHMENT E ACCEPTANCE CRITERIA AND EXPANSION CRITERIA Examination Expansion Additional Examination Item Shown In Acceptance Criteria Expansion Criteria Links(s) Acceptance Criteria (Note 1)
Baffle-Former Figures 5 and 6 Volumetric (UT) examination. a. Lower support a. Confirmation that more than 5% of the a and b. The examination Assembly column bolts baffle-former bolts actually examined on the acceptance criteria Baffle-former bolts The examination acceptance four baffle plates at the largest distance from for the UT of the criteria for the UT of the b. Baffle-former the core (presumed to be the lowest dose lower support column bafne-former bolts shall be bolts locations) contain unacceptable indications bolts and the established as part of the shall require UT examination of the lower baffle-former examination technical support column bolts within the next three bolts shall be justification. fuel cycles. established as part of the examination
- b. Confirmation that more than 5% of the lower technical support column bolts actually examined justification.
contain unacceptable indications shall require UT examination of the baffle-former bolts.
Baffle-Former Figures 5,6,7, Visual (VT-3) examination. None N/A N/A Assembly &and9 Assembly The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, vertical displacement of shroud plates near high fluence joints, and broken or damaged edge bolt locking systems along high fluence baffle plate joints.
Page 49 of 50
POINT BEACH NUCLEAR PLANT AM 3-44 ADMINISTRATIVE MANUAL Revision 0 September 30,2009 ATTACHMENT E ACCEPTANCE CRITERIA AND EXPANSION CRITERIA Examination Expansion Additional Examination Item Shown In Acceptance Criteria Expansion Criteria Linlrs(s) Acceptance Criteria (Note 1)
Alignment and -
N/A Point Direct physical measurement or None N/A N/A Interfacing Beach hold spring height.
Components down springs lntemals hold down are 403 SS.
spring Thermal Shield Figure 11 Visual (VT-3) examination. None NIA N/A Assembly The speciiic relevkt conditions Thermal shield for thermal shield flexures are flexures excessive wear, fracture or complete separation.
Notes: 1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).
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