ML11280A121

Indian Point, Units 2 & 3 - License Renewal Application - Completion of Commitment #30 Re Reactor Vessel Internals Inspection Plan

ML11280A121
Person / Time
Site:	Indian Point
Issue date:	09/28/2011
From:	Dacimo F Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-11-107
Download: ML11280A121 (61)

Text

Entergy Enter-qy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249 Tel (914) 788-2055 Fred Dacimo Vice President Operations License Renewal NL-1 1-107 September 28, 2011 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

REFERENCE:

License Renewal Application

-Completion of Commitment

1. 30 Regarding the Reactor Vessel Internals inspection Plan Indian Point Nuclear Generating Unit Nos. 2 and 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64 1. Entergy Letter dated April 23, 2007, Fred Dacimo to Document Control Desk, "License Renewal Application" (NL-07-039)

2. Entergy Letter dated July 14, 2010, Fred Dacimo to Document Control Desk, "Amendment 9 to License Renewal Application (LAR) -Reactor Vessel Internals Program" (NL-10-063)

3. EPRI, Materials Reliability Program (MRP), Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227)4. NRC, "Final safety Evaluation of EPRI Report, Materials Reliability Program Report 1016596, Revision 0, Pressurized Water reactor (PWR) Internals Inspection and Evaluation Guidelines" dated June 22, 2011

Dear Sir or Madam:

Entergy Nuclear Operations, Inc. applied for renewal of the Indian Point Nuclear Generating Unit Nos. 2 and 3 operating licenses by the reference 1 letter which included a list of regulatory commitments.

The commitment list contained commitment

1. 30 for submitting an inspection plan for reactor vessel internals.

Reference 2 provided the Indian Point Nuclear Generating Unit Nos. 2 and 3 Reactor Vessel Internals Program.This letter contains the inspection plan satisfying the completion of commitment

1. 30 to the License Renewal Application regarding the Aging Management Programs for Reactor Vessel Internals.

The Indian Point Energy Center (IPEC) Reactor Vessel Internals Inspection Plan was developed in accordance with the results of industry programs applicable to the reactor vessel internals and addresses the action items and conditions stated in the NRC Final Safety Evaluation of MRP-227 (Reference 4).

NL-1 1-107 Docket Nos. 50-247 and 50-286 Page 2 of 2 There are no new commitments identified in this submittal.

If you have any questions, or require additional information, please contact Mr. Robert Walpole at 914-734-6710.

Si ly, FD/cbr

Attachment:

Indian Point Energy Center Reactor Vessel Internals Inspection Plan cc: Mr. William Dean, Regional Administrator, NRC Region I Mr. J. Boska, Senior Project Manager, NRC, NRR, DORL Mr. David Wrona, NRC Branch Chief, Engineering Review Branch I Mr. Sherwin E. Turk, NRC Office of General Counsel, Special Counsel NRC Resident Inspectors Office, Indian Point Mr. Paul Eddy, NYS Dept. of Public Service Mr. Francis J. Murray, Jr., President and CEO, NYSERDA ATTACHMENT TO NL-1 1-107 Indian Point Energy Center Reactor Vessel Internals Inspection Plan ENTERGY NUCLEAR OPERATIONS, INC INDIAN POINT NUCLEAR GENERATING UNITS 2 AND 3 DOCKET NOS. 50-247 & 50-286 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan 1 INTRODUCTION

1.1 Aging

Management Program Inspection Plan The EPRI MRP guidelines define a supplemental inspection program for managing aging effects on the reactor vessel internals and were used to develop this inspection plan for IPEC Units 2 and 3. The EPRI MRP Reactor Internals Focus Group developed the MRP-227 Guidelines to support the demonstration of continued functionality, with requirements for inspections to detect the effects of aging along with requirements for the evaluation of detected aging effects, if any.The development of MRP-227 combined the results of component functionality assessments with component accessibility, operating experience, existing evaluations and prior examination results to determine the appropriate aging management methods, initial examination timing and the need and timing of subsequent inspections and identified the components and locations for supplemental examination.

In accordance with MRP-227, this inspection plan includes:* Identification of items for inspection,* Specification of the type of examination appropriate for each degradation mechanism,* Specification of the required level of examination qualification,* Schedule of initial inspection and frequency of subsequent inspections," Criteria for sampling and coverage,* Criteria for expansion of scope if unanticipated indications are found,* Inspection acceptance criteria,* Methods for evaluating examination results not meeting the acceptance criteria," Updating the program based on industry-wide results, and* Contingency measures to repair, replace or mitigate.Page 1 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I hidian Point Energy Center Reactor Vessel Internals Inspection Plan 2 BACKGROUND OF IPEC REACTOR VESSEL INTERNALS DESIGN This section provides a summary of the design characteristics for the IPEC Westinghouse PWR internals.

2.1 Westinghouse

Internals Design Characteristics A schematic view of a typical set of Westinghouse-designed PWR internals is Figure 2-1. More detailed views of selected internals components are Figures 2-2 through 2-16 at the end of this section. These figures are typical and are not an exact representation of the IPEC internals.

To help in the categorization of IPEC internals design characteristics as discussed in MRP-227 Section 3.1.3, the following information is provided.

IPEC Units 2 and 3 are Westinghouse four loop plants with a downflow baffle-barrel region flow design, and a top hat design upper support plate. Unit 2 had an original thermal output of 2758 MWth and Unit 3 had an original thermal output of 3025 MWth.Page 2 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan CONTROL ROD DRIVE MECHANISM UPPER SUPPORT PLATE ROD TRAVEL HOUSING INSTRUMENTATION PORTS THERMAL SLEEVE LIFTING LUG INTERNALS SUPPORT LEDGE CORE BARREL SUPPORT COLUMN UPPER CORE PLATE OUTLET NOZZLE BAFFLE RADIAL SUPPORT BAFFLE CORE SUPPORT COLUMNS INSTRUMENTATION THIMBLE GUIDES RADIAL SUPPORT CORE SUPPORT -CLOSURE HEAD ASSEMBLY HOLD-DOWN SPRING-CONTROL ROD GUIDE TUBE CONTROL ROD DRIVE SHAFT INLET NOZZLE-CONTROL ROD CLUSTER (WmrHDRAWI

-- ACCESS PORT-REACTOR VESSEL CORE PLATE Figure 2-1 Overview of typical Westinghouse internals Page 3 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Westinghouse 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 following a complete core off-load.The reactor core is positioned and supported by the upper internals and lower internals assemblies.

The individual fuel assemblies are positioned by fuel alignment pins in the upper core plate and the lower core plate. These pins control the orientation of the core with respect to the upper and lower internals assemblies.

The lower internals are aligned with the upper internals by the upper core plate alignment pins and secondarily by the head/vessel alignment pins. The lower internals are aligned to the vessel by the lower radial support/clevis assemblies and by the head/vessel alignment pins. Thus, the core is aligned with the vessel by a number of interfacing components.

The lower internals assembly is supported in the vessel by clamping to a ledge below the vessel-head mating surface and is closely guided at the bottom by radial support/clevis 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.Upper Internals Assembly The major sub-assemblies that constitute the upper internals assembly are the: (1) upper core plate (UCP); (2) upper support column assemblies; (3) control rod guide tube assemblies; and (4)upper support plate (USP).During reactor operation, the upper internals assembly is preloaded against the fuel assembly springs and the internals hold down spring by the reactor vessel head pressing down on the outside edge of the USP. The USP acts as the divider between the upper plenum and the reactor vessel head and as a relatively stiff base for the rest of the upper internals.

The upper support columns and the control rod guide tubes are attached to the USP. The UCP, in turn, is attached to the upper support columns. The USP design at IPEC is designated as a top hat design.The UCP is perforated to permit coolant to pass from the core below into the upper plenum between the USP and the UCP. The coolant then exits through the outlet nozzles in the core barrel. The UCP positions and laterally supports the core by fuel alignment pins extending below the plate. The UCP contacts and preloads the fuel assembly springs and thus maintains contact of the fuel assemblies with the lower coreeplate (LCP) during reactor operation.

The upper support columns vertically position the UCP and are designed to take the uplifting hydraulic flow loads and fuel spring loads on the UCP. The control rod guide tubes are bolted to the USP and pinned at the UCP so they can be easily removed if replacement is desired. The control rod guide tubes are designed to guide the control rods in and out of the fuel assemblies to control power generation.

Guide tube cards are located within each control rod guide tube Page 4 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan to guide the absorber rods. The control rod guide tubes are also slotted in their lower sections to allow coolant exiting the core to flow into the upper plenum.The upper instrumentation columns are bolted to the USP. These columns support the thermocouple guide tubes that lead the thermocouples from the reactor head through the upper plenum to just above the UCP.The UCP alignment pins locate the UCP laterally with respect to the lower internals assembly.The pins must laterally support the UCP so that the plate is free to expand radially and move axially during differential thermal expansion between the upper internals and the core barrel. The UCP alignment pins are the interfacing components between the UCP and the core barrel.Lower Internals Assembly The fuel assemblies are supported inside the lower internals assembly on top of the LCP. The functions of the LCP are to position and support the core and provide a metered control of reactor coolant flow into each fuel assembly.

The LCP is elevated above the lower support casting by support columns and bolted to a ring support attached to the inside diameter of the core barrel. The support columns transmit vertical fuel assembly loads from the LCP to the much thicker lower support casting. The function of the lower support casting is to provide support for the core. The lower support casting is welded to and supported by the core barrel, which transmits vertical loads to the vessel through the core barrel flange.The primary function of the core barrel is to support the core. A large number of components are attached to the core barrel, including the baffle/former assembly, the core barrel outlet nozzles, the thermal shields, the alignment pins that engage the UCP, the lower support casting, and the LCP. The lower radial support/clevis assemblies restrain large transverse motions of the core barrel but at the same time allow unrestricted radial and axial thermal expansion.

The baffle and former assembly consists of vertical plates called baffles and horizontal support plates called formers. The baffle plates are bolted to the formers by the baffle/former bolts, and the formers are attached to the core barrel inside diameter by the barrel/former bolts. Baffle plates are secured to each other at selected corners by edge bolts. In addition, at IPEC, corner brackets are installed behind and bolted to the baffle plates. The baffle/former assembly forms the interface between the core and the core barrel. The baffles provide a barrier between the core and the former region so that a high concentration of flow in the core region can be maintained.

A secondary benefit is to reduce the neutron flux on the vessel.The function of the core barrel outlet nozzles is to direct the reactor coolant, after it leaves the core, radially outward through the reactor vessel outlet nozzles. The core barrel outlet nozzles are located in the upper portion of the core barrel directly below the flange and are attached to the core barrel, each in line with a vessel outlet nozzle.Page 5 NL-1 1-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Additional neutron shielding of the reactor vessel is provided in the active core region by thermal shields attached to the outside of the core barrel.A flux thimble is a long, slender stainless steel tube that passes from an external seal table, through a bottom mounted nozzle penetration, through the lower internals assembly, and finally extends to the top of a fuel assembly.

The flux thimble provides a path for a neutron flux detector into the core and is subjected to reactor coolant pressure and temperature on the outside surface and to atmospheric conditions on the inside. The flux thimble path from the seal table to the bottom mounted nozzles is defined by flux thimble guide tubes, which are part of the primary pressure boundary and not part of the internals.

The bottom-mounted instrumentation (BMI)columns provide a path for the flux thimbles from the bottom of the vessel into the core. The BMI columns align the flux thimble paths with instrumentation thimbles in the fuel assembly.In the upper internals assembly, the upper support plate, the upper support columns, and the upper core plate are considered core support structures.

In the lower internals assembly, the lower core plate, the lower support casting, the lower support columns, the core barrel including the core barrel flange, the radial support/clevis assemblies, the baffle plates, and the former plates are classified as core support structures.

Z Wear Area Figure 2-2 Typical Westinghouse control rod guide card (17x17 fuel assembly)Page 6 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan p 1 I D II II Lower Flange Weld Figure 2-3 Typical Westinghouse control rod guide tube assembly Page 7 I NL- 11-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Flange Weld Upper Core Barrel to Lower Core Barrel Circumferential Weld Lower Barrel Circumferential Weld Core Barrel to Support Plate Weld Axial Weld Lower Barrel Axial Weld Lower Barrel Axial Weld Figure 2-4 Major fabrication welds in typical Westinghouse core barrel Page 8 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan CORME EDGE BRAOOMT BAFFLE TO FORME BOLT Figure 2-5 Bolt locations in typical Westinghouse baffle-former-barrel structure.

Page 9 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan (D (U 0 Figure 2-6 Baffle-edge bolt and baffle-former bolt locations at high fluence seams in bolted baffle-former assembly Page 10 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan T High Fluence Seams Figure 2-7 High fluence seam locations in Westinghouse baffle-former assembly Page 11 NL- 11-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Potential Gaps at Baffle-Former Plate Levels Figure 2-8 Exaggerated view of void swelling induced distortion in Westinghouse baffle-former assembly.Page 12 NL-1 1-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Figure 2-9 Vertical displacement of Westinghouse baffle plates caused by void swelling.Page 13 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan CORE BARREL Figure 2-10 Schematic cross-sections of the Westinghouse hold-down springs Page 14 NL-1 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Core Barrel Thermal Shield Thermal Shield Flexure Core Support Figure 2-11 Location of Westinghouse thermal shield flexures Page 15 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Lower Core Plate Lower Core Support Structure Core Support Plate (Forging)Figure 2-12 Schematic indicating location of Westinghouse lower core support structure.

Additional details shown in Figure 2-13 LOWER CORE PLATE DIFFUSER PLATECORE SUPPORT PLATE/FORGING BOTTOM MOUNTED INSTRUMENTATION COLUMN Figure 2-13 Westinghouse lower core support structure and bottom mounted instrumentation columns. Core support column bolts fasten the core support columns to the lower core plate Page 16 NL-1 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan B Figure 2-14 Typical Westinghouse core support column. Core support column bolts fasten the top of the support column to the lower core plate E- 1711 M t Figure 2-15 Examples of Westinghouse bottom mounted instrumentation column designs Page 17 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan/Figure 2-16 Typical Westinghouse thermal shield flexure Page 18 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan 3 INSPECTION PLAN

SUMMARY

Management of component aging effects includes actions to prevent or control aging effects, review of operating experience to better understand the potential for aging effects to occur, inspections to detect the onset of aging effects in susceptible components, protocols for evaluation and remediation of the effects of aging, and procedures to ensure component aging effects are managed in a coordinated program.3.1 Component Inspection and Evaluation Overview This discussion summarizes the guidance of the MRP Inspection

& Evaluation (I&E) guidelines necessary to understand implementation but does not duplicate the full discussion of the technical bases. MRP-227 and its supporting documents provide further information on the technical bases of the program.MRP-227 establishes four groups of reactor internals components with respect to inspections:

Primary, Expansion, Existing Programs and No Additional Measures, as summarized below." 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 which 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 PWR internals that are highly or moderately susceptible to the effects of at least one of the eight aging mechanisms, but for which a 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 will depend on the findings from the examinations of the Primary components." 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.No Additional Measures:

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. Items categorized as Category A in MRP-191 are those for which aging effects are below the screening criteria, so that aging degradation significance is minimal. Primary, expansion, and Page 19 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel hIternals Inspection Plan existing examinations verify that the chemical control program has been effective at controlling stress corrosion cracking and loss of material due to corrosion for Category A components.

Additional components were placed in the No Additional Measures group as a result of Failure Modes, Effects and Criticality Analysis (FMECA) and the functionality assessment.

No further action is required for managing the aging of the No Additional Measures components.

However, any core support structures subject to ASME Section XI Examination Category B-N-3 requirements continue to be subject to those ASME Code requirements throughout the period of extended operation.

The inspections required for Primary and Expansion components were selected from visual, surface and volumetric examination methods that are applicable and appropriate for the expected degradation effect (e.g. cracking caused by particular mechanisms, loss of material caused by wear). The inspection methods include: Visual examinations (VT-3, VT-1, EVT-I), surface examinations, volumetric examinations (specifically UT) and physical measurements.

MRP-227 provides detailed justification for the components selected for inspection and the specific examination methods selected for each. The MRP-228 report, PWR Internals Inspection Standards, provides detailed examination standards and any inspection technical justification or inspection personnel training requirements.

3.2 Inspection

and Evaluation Requirements for Primary Components The inspection requirements for Primary Components at IPEC Units 2 and 3 from MRP-227 are provided in Table 5-2.3.3 Inspection and Evaluation Requirements for Expansion Components The inspection requirements for Expansion Components at IPEC Units 2 and 3 from MRP-227 are provided in Table 5-3.3.4 Inspections of Existing Program Components The list of Existing Program Components at IPEC Units 2 and 3 from MRP-227 are provided in Table 5-4. This includes components in the Section XI ISI Program for IPEC Units 2 and 3 designated as B-N-2 and B-N-3 locations.

The Reactor Vessel Component Inspection Plan conducted as part of the ISI program for IPEC Units 2 and 3 is provided in Table 5-6. The components are inspected as part of the ISI Program.The ISI Program inspections are implemented in accordance with ASME Section XI schedule requirements.

Page 20 NL-1 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan 3.5 Examination Systems Equipment, techniques, procedures and personnel used to perform examinations required under this program will be consistent with the requirements of MRP-228. Indications detected during these examinations will be characterized and reported in accordance with the requirements of MRP-228.3.6 Information Supplied in Response to the NRC Safety Evaluation of MRP-227 As part of the NRC Final Safety Evaluation of MRP-227, a number of action items and conditions were specified by the staff. Table 5-8 documents the IPEC response to the NRC Final Safety Evaluation of MRP-227. Wherever possible, these items have been addressed in the appropriate sections of this document.

All NRC action items and conditions not addressed elsewhere in this document are discussed in this section.SER Section 4.2.1. Applicant/Licensee Action Item 1 IPEC has assessed its plant design and operating history and has determined that MRP-227 is applicable to the facility.

The assumptions regarding plant design and operating history made in MRP-191 are appropriate for IPEC and there are no differences in component inspection categories at IPEC. IPEC Unit 2 (IP2) had the first 8 years of operation with a high leakage core loading pattern. IPEC Unit 3 (IP3) had the first 10 years of operation with a high leakage core loading pattern. The FMECA and functionality analyses were based on the assumption of 30 years of operation with high leakage core loading patterns; therefore, IPEC is bounded by the assumptions in MRP-191. IPEC has always operated as a base-load plant which operates at fixed power levels and does not vary power on a calendar or load demand schedule.SER Section 4.2.2. Applicant/Licensee Action Item 2 IPEC reviewed the information in Table 4-4 of MRP-191 and determined that this table contains all of the RVI components that are within the scope of license renewal. This is shown in Table 5-7.SER Section 4.2.3. Applicant/Licensee Action Item 3 At IP2, the original X750 guide tube support pins (split pins) were replaced in 1995 with an improved X750 Revision B material made from more selective material with more continuous carbide coverage grain boundaries and tighter quality controls, to provide greater resistance to stress corrosion cracking.

At IP3 the original X750 guide tube support pins (split pins) were replaced in 2009 (after 33 years in service) with cold-worked 316 stainless steel. The cold-worked 316 stainless steel is a significant improvement over the X750. At IPEC the effects of Page 21 NL- 11-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan aging on these components will be managed in the period of extended operation based on industry experience and plant specific evaluations.

SER Section 4.2.4, Applicant/Licensee Action Item 4 This action item does not apply to Westinghouse designed units.SER Section 4.2.5, Applicant/Licensee Action Item 5 The IPEC plant specific acceptance criteria for hold down springs and an explanation of how the proposed acceptance criteria are consistent with the IPEC licensing basis and the need to maintain the functionality of the hold down springs under all licensing basis conditions will be developed prior to the first required physical measurement.

In accordance with SER Section 4.2.5, IPEC will submit this information to the NRC as part of the submittal to apply the approved version of MRP-227.SER Section 4.2.6, Applicant/Licensee Action Item 6 This action item does not apply to Westinghouse designed units.SER Section 4.2.7. Applicant/Licensee Action Item 7 The IPEC plant specific analyses to demonstrate the lower support column bodies will maintain their functionality during the period of extended operation will consider the possible loss of fracture toughness in these components due to thermal and irradiation embrittlement.

The analyses will be consistent with the IPEC licensing basis and the need to maintain the functionality of the lower support column bodies under all licensing basis conditions of operation.

In accordance with SER Section 4.2.7, IPEC will submit this information to the NRC as part of the submittal to apply the approved version of MRP-227.SER Section 4.2.8. Applicant/Licensee Action Item 8 This document includes an inspection plan which addresses the identified plant-specific action items contained in the NRC Final Safety Evaluation for MRP-227. IPEC is not requesting any deviations from the guidance provided in MRP-227, as approved by the NRC.Page 22 NL-11-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan 4 EXAMINATION ACCEPTANCE AND EXPANSION CRITERIA 4.1 Examination Acceptance Criteria 4.1.1 Visual (VT-3) Examination Visual (VT-3) examination is an appropriate NDE method for the detection of general degradation conditions in many of the susceptible components.

The ASME Code Section XI, Examination Category B-N-3, provides a set of relevant conditions for the visual (VT-3) examination of removable core support structures in Section IWB. These are: 1. structural distortion or displacement of parts to the extent that component function may be impaired;2. loose, missing, cracked, or fractured parts, bolting, or fasteners;

3. corrosion or erosion that reduces the nominal section thickness by more than 5%;4. wear of mating surfaces that may lead to loss of function; and 5. structural degradation of interior attachments such that the original cross-sectional area is reduced more than 5%.For components in the Existing Programs group, these general relevant conditions are sufficient.

However, for components where visual (VT-3) is specified in the Primary or the Expansion group, more specific descriptions of the relevant conditions are provided in Table 5-5 for the benefit of the examiners.

One or more of these specific relevant condition descriptions may be applicable to the Primary and Expansion components listed in Tables 5-2 and 5-3.The examination acceptance criteria for components requiring visual (VT-3) examination is thus the absence of any of the relevant condition(s) specified in Table 5-5.The disposition can include a supplementary examination to further characterize the relevant condition, an engineering evaluation to show that the component is capable of continued operation with a known relevant condition, or repair/replacement to remediate the relevant condition.

Page 23 NL-11-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan 4.1.2 Visual (VT-1) Examination Visual (VT-1) examination is defined in the ASME Code Section XI as an examination"conducted to detect discontinuities and imperfections on the surface of components, including such conditions as cracks, wear, corrosion, or erosion." The acceptance criterion for any visual (VT-1) examinations is the absence of any relevant conditions defined by the ASME Code, as supplemented by more specific plant inservice inspection requirements.

4.1.3 Enhanced

Visual (EVT-1) Examination Enhanced visual (EVT-l) examination has the same requirements as the ASME Code Section XI visual (VT-1) examination, with additional requirements given in the Inspection Standard, MRP-228. These enhancements are intended to improve the detection and characterization of discontinuities taking into account the remote visual aspect of reactor internals examinations.

As a result, EVT- 1 examinations are capable of detecting small surface-breaking cracks and sizing surface crack length when used in conjunction with sizing aids (e.g. landmarks, ruler, and tape measure).

EVT- 1 examination is the appropriate NDE method for detection of cracking in plates or their welded joints. Thus the relevant condition applied for EVT- 1 examination is the same as for cracking in Section XI which is crack-like surface-breaking indications.

Therefore, until such time as engineering studies provide a basis by which a quantitative amount of degradation can be shown acceptable for the specific component, any observed relevant condition must be dispositioned.

In the interim, the examination acceptance criterion is the absence of any detectable surface-breaking indication.

4.1.4 Surface

Examination Surface ET (eddy current testing) examination is specified as an alternative or as a supplement to visual examinations.

No specific acceptance criteria for surface (ET) examination of PWR internals locations are provided in the ASME Code Section XI. Since surface ET is employed as a signal-based examination, a technical justification per the Inspection Standard, MRP-228 provides the basis for detection and length sizing of surface-breaking or near-surface cracks. The signal-based relevant indication for surface (ET) is thus the same as the relevant condition for enhanced visual (EVT- 1) examination.

The acceptance criteria for enhanced visual (EVT- 1)examinations in 4.1.3 (and accompanying entries in Table 5-5) are therefore applied when this method is used as an alternative or supplement to visual examination.

4.1.5 Volumetric

Examination The intent of volumetric examinations specified for bolts and pins is to detect planar defects. No flaw sizing measurements are recorded or assumed in the acceptance or rejection of individual Page 24 NL-11-107 Docket Nos. 50-247 and 50-286 Attachment 1 hidian Point Energy Center Reactor Vessel Internals Inspection Plan bolts or pins. Individual bolts or pins are accepted based on the absence of relevant indications established as part of the examination technical justification.

When a relevant indication is detected in the cross-sectional area of the bolt or pin, it is assumed to be non-functional and the indication is recorded.

A bolt or pin that passes the criterion of the examination is considered functional.

Because of this pass/fail acceptance of individual bolts or pins, the examination acceptance criterion for volumetric (UT) examination of bolts and pins is based on a reliable detection of indications as established by the individual technical justification for the proposed examination.

This is in keeping with current industry practice.

For example, planar flaws on the order of 30%of the cross-sectional area have been determined reliably detectable in previous bolt NDE technical justifications for baffle-former bolting.Bolted and pinned assemblies are evaluated for acceptance based on a plant specific evaluation.

4.2 Physical

Measurements Examination Acceptance Criteria Continued functionality can be confirmed by physical measurements where, for example, loss of material caused by wear, loss of pre-load of clamping force caused by various degradation mechanisms, or distortion/deflection caused by void swelling may occur. For Westinghouse designs, tolerances are available on a design or plant-specific basis. Specific acceptance criteria will be developed as required, and thus are not provided generically in this plan.4.3 Expansion Criteria The criterion for expanding the scope of examination from the Primary components to their linked Expansion components is contained in Table 5-5 for IPEC.Page 25 NL-11-107 Docket Nos. 50-247 and 50-286 Attachment 1 hidian Point Energy Center Reactor Vessel Internals Inspection Plan 5 TABLES Table 5-1 Indian Point 2 & 3 Component Cross Reference Table 5-2 Primary Components at IPEC Units 2 and 3 Table 5-3 Expansion Components at IPEC Units 2 and 3 Table 5-4 Existing Program Components at IPEC Units 2 and 3 Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Table 5-6 Reactor Vessel Component ISI Program Inspection Plan for IPEC Units 2 and 3 Table 5-7 List of IPEC Reactor Vessel Interior Components and Materials Based on MRP-191 -Table 4-4 Table 5-8 IPEC Response to the NRC Final Safety Evaluation of MRP-227 Page 26 NL-11-107 Docket Nos. 50-247 and 50-286 Attachment 1 hidian Point Energy Center Reactor Vessel hiternals hIspection Plan Table 5-1 Indian Point 2 & 3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227 1 Core Baffle/Former Assembly -Lower Internals Baffle-Former Assembly -Bolts Assembly -Baffle and Baffle-Edge Bolts (Table Former Assembly 4-3 and 5-3)Baffle-Edge Bolts Baffle-Former Assembly -Baffle-Former Bolts (Table Baffle-Former Bolts 4-3 and 5-3)2 Core Baffle/Former Assembly -Lower Internals Baffle-Former Assembly -Plates Assembly -Baffle and Assembly (Table 4-3 and Former Assembly 5-3)Baffle Plates Former Plates 3 Core Barrel Assembly -Bolts and Lower Internals Core Barrel Assembly -Screws Assembly -Baffle and Barrel-Former Bolts (Table Former Assembly 4-6)Barrel-Former Bolts 4 Core Barrel Assembly -Axial Lower Internals Thermal Shield Assembly Flexure Plates (Thermal Shield Assembly -Neutron -Thermal Shield Flexures Flexures)

Panels/Thermal Shield (Table 4-3 and 5-3)Thermal Shield Flexures 5 Core Barrel Assembly -Flange Lower Internals Core Barrel Assembly -Assembly -Core Core Barrel Flange (Table Barrel 4-6 and 4-9)Core Barrel Flange Page 27 NL-11-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-1 Indian Point 2 & 3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227 6 Core Barrel Assembly -Ring Lower Internals None Assembly -Core Core Barrel Assembly -Shell Barrel Core Barrel Assembly -Thermal Upper Core Barrel Shield Lower Core Barrel Lower Internals Assembly -Neutron panels/thermal shield Thermal shield 7 Core Barrel Assembly -Lower None Core Barrel Assembly -Core Barrel Flange Weld Lower Core Barrel Flange Lower Internals Weld (Table 4-6)Core Barrel Assembly -Upper Assembly -Core Core Barrel Flange Weld Barrel Core Barrel Assembly -Upper Core Barrel Flange Core Barrel Flange Weld (Table 4-3 and 5-3)8 Core Barrel Assembly -Outlet Lower Internals Core Barrel Assembly -Nozzles Assembly -Core Core Barrel Outlet Nozzles Barrel (Table 4-6)Core Barrel Outlet Nozzles 9 Lower Internals Assembly -Interfacing Alignment and Interfacing Clevis Insert Bolt Components

-Components

-Clevis Insert Interfacing Bolts (Table 4-9)Components Clevis Insert Bolts Page 28 NL-11-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-1 Indian Point 2 & 3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227 10 Lower Internals Assembly -Interfacing None Clevis Insert Components

-Interfacing Components Clevis Inserts 11 Lower Internals Assembly -Lower Internals None Intermediate Diffuser Plate Assembly -Diffuser Plate Diffuser Plate 12 Lower Internals Assembly -Fuel Lower Internals None Alignment Pin Assembly -Lower Core Plate and Fuel Alignment Pins Fuel Alignment Pins 13 Lower Internals Assembly -Lower Internals Lower Internals Assembly Lower Core Plate Assembly -Lower -Lower Core Plate (Table Core Plate and Fuel 4-9, 2 places)Alignment Pins Lower Core Plate Page 29 NL-11-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-1 Indian Point 2 & 3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227 14 Lower Internals Assembly -Lower Internals None Assembly -Lower" Lower Core Support Castings Support Casting or Forging" Column Cap None Lower Support" Lower Core Support Column Casting Bodies None Lower Internals Lower Support Assembly -Assembly -Lower Lower Support Column Support Column Bodies (Cast) (Table 4-6)Assembly Lower Support Column Bodies 15 Lower Internals Assembly -Lower Internals Lower Support Assembly -Lower Core Support Plate Assembly -Lower Lower Support Column Column Bolt Support Column Bolts (Table 4-6)Assembly Lower Support Column Bolts 16 Lower Internals Assembly -Lower Internals None Lower Core Support Plate Assembly -Lower Column Sleeves Support Column Assembly Lower Support Column Sleeves 17 Lower Internals Assembly -Lower Internals None Radial Key Assembly -Radial Support Keys Radial Support Keys Page 30 NL-11-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-1 Indian Point 2 & 3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227 18 Lower Internals Assembly -Lower Internals None Secondary Core Support Assembly -Secondary Core Support (SCS)Assembly SCS Base Plate 19 RCCA Guide Tube Assembly -Upper Internals None Bolt Assembly -Control Rod Guide Tube Assemblies and Flow Downcomers Bolts 20 RCCA Guide Tube Assembly -Upper Internals Control Rod Guide Tube Guide Tube (including Lower Assembly -Control Assembly -Lower Flange Flange Welds) Rod Guide Tube Welds (Table 4-3 and 5-3)Assemblies and Flow Downcomers Flanges -lower 21 RCCA Guide Tube Assembly -Upper Internals Control Rod Guide Tube Guide Plates Assembly -Control Assembly -Guide Plates Rod Guide Tube (Cards) (Table 4-3 and 5-3)Assemblies and Flow Downcomers Guide Plates/Cards 22 RCCA Guide Tube Assembly -Upper Internals None Support Pin Assembly -Control Rod Guide Tube Assemblies and Flow Downcomers Guide Tube Support Pins Page 31 NL-1 1-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-1 Indian Point 2 & 3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227 23 Core Plate Alignment Pin Interfacing Alignment and Interfacing Components

-Components

-Upper Core Interfacing Plate Alignment Pins Components (Table 4-9)Upper Core Plate Alignment Pins 24 Head/Vessel Alignment Pin Interfacing None Components

-Interfacing Components Head and Vessel Alignment Pins 25 Hold-down Spring Interfacing Alignment and Interfacing Components

-Components

-Internals Interfacing Hold Down Spring (Table Components 4-3 and 5-3)Internals Hold Down Spring 26 Mixing Devices Upper Internals None Assembly -Mixing-Support Column Orifice Base Devices-Support Column Mixer Mixing devices 27 Support Column Upper Internals None Assembly -Upper Support Column Assemblies Column Bodies Page 32 NL-1 1-107 Docket Nos. 50-247 and 50-286 Attachment 1 Indian Point Energy Center Reactor Vessel hiternals hIspection Plan Table 5-1 Indian Point 2 & 3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227 28 Upper Core Plate, Fuel Alignment Upper Internals None Pin Assembly -Upper Core Plate and Fuel Alignment Pins Fuel Alignment Pins 29 Upper Support Plate, Support Upper Internals Upper Internals Assembly Assembly (Including Ring) Assembly -Upper -Upper Support Ring or Support Plate Skirt (Table 4-9)Assembly Upper Support Plate 30 Upper Support Column Bolt Upper Internals None Assembly -Upper Support Column Assemblies Bolts 31 Bottom Mounted Instrumentation Lower Internals Bottom Mounted Column Assembly -Bottom- Instrumentation System -Mounted Bottom Mounted Instrumentation (BMI) Instrumentation (BMI)Column Assemblies Column Bodies (Table 4-6)BMI Column Bodies 32 Flux Thimble Guide Tube Lower Internals Bottom Mounted Assembly -Flux Instrumentation System -Thimbles (Tubes) Flux Thimble Tubes (Table 4-9)Flux Thimbles (Tubes)Page 33 NL-11-107 Docket Nos. 50-247 and 50-286 Attachment 1 hidian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-1 Indian Point 2 & 3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227 33 Thermocouple Conduit Upper Internals None Assembly -Upper Instrumentation Conduit and Support Conduits Page 34 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability (Mechanism)

Expansion Link Method/Frequency Examination Coverage Control Rod Guide Tube IPEC Units 2 and Loss of Material None Visual (VT-3) examination no 20% examination of the Assembly 3 (Wear) later than 2 refueling outages number of CRGT Guide plates (cards) from the beginning of the assemblies, with all guide license renewal period, cards within each selected Subsequent examinations are CRGT assembly examined.required on a ten-year interval.See Figure 2-2 Control Rod Guide Tube IPEC Units 2 and Cracking (SCC, Bottom-mounted Enhanced visual (EVT-1) 100% of outer (accessible)

Assembly 3 Fatigue) instrumentation examination to determine the CRGT lower flange weld Lower flange welds (BMI) column presence of crack-like surface surfaces and adjacent base bodies, flaws in flange welds no later metal.Lower support than 2 refueling outages from column bodies the beginning of the license See Figure 2-3 (cast) renewal period and subsequent examination on a ten-year interval.Core Barrel Assembly IPEC Units 2 and Cracking (SCC) None Enhanced visual (EVT- I) 100% of one side of the Upper core barrel flange weld 3 examination, no later than 2 accessible surfaces of the refueling outages from the selected weld and adjacent beginning of the license renewal base metal.period and subsequent examination on a ten-year See Figure 2-4 interval.Page 35 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability (Mechanism)

Expansion Link Method/Frequency Examination Coverage Baffle-Former Assembly IPEC Units 2 and Cracking (IASCC, None Visual (VT-3) examination, with Bolts and locking devices Baffle-edge bolts 3 Fatigue) that results baseline examination between on high fluence seams.in 20 and 40 EFPY and subsequent 100% of components" Lost or broken examinations on a ten-year accessible from core side.locking devices interval.

75% of a component's total" Failed or missing (accessible

+ inaccessible) bolts inspection area or volume" Protrusion of bolt will be examined or, when heads addressing a set of like components (e.g., bolting), that the inspection examine a minimum sample size of 75 percent of the total population of like components.

For the inspection of a set of like components, it is understood that essentially 100% of the volume/area of each accessible like component will be examined.I See Figure 2-5 Page 36 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I hidian Point Energy Center Reactor Vessel hiternals hIspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 A ibit Effect Examination Examination Coverage Item Applicability (Mechanism)

Expansion Link Method/Frequency ExaminationCoverage Baffle-Former Assembly IPEC Units 2 and Cracking (IASCC, Lower support Baseline volumetric (UT) 100% of accessible bolts or Baffle-former bolts 3 Fatigue) column bolts, examination between 25 and 35 as supported by plant-Barrel-former bolts EFPY, with subsequent specific justification.

Heads examination after 10 years to accessible from the core confirm stability of bolting side. UT accessibility may pattern, be affected by complexity of head and locking device designs. 75% of a component's total (accessible

+ inaccessible) inspection area or volume will be examined or, when addressing a set of like components (e.g., bolting), that the inspection examine a minimum sample size of 75 percent of the total population of like components.

For the inspection of a set of like components, it is understood that essentially 100% of the volume/area of each accessible like component will be examined.I See Figures 2-5 and 2-6.Page 37 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability (Mechanism)

Expansion Link Method/Frequency Examination Coverage Baffle-Former Assembly IPEC Units 2 and Distortion (Void None Visual (VT-3) examination to Core side surface as Assembly 3 Swelling), or check for evidence of distortion, indicated.

Cracking (IASCC) with baseline examination that results in between 20 and 40 EFPY and See Figures 2-6, 2-7, 2-8* Abnormal subsequent examinations on a and 2-9.interaction with ten-year interval.fuel assemblies

-Gaps along high fluence baffle joint-Vertical displacement of baffle plates near high fluence joint-Broken or damaged edge bolt locking systems-along high fluence baffle joint Alignment and Interfacing IPEC Units 2 and Distortion (Loss of None Direct measurement of spring Measurements should be Components 3 Load) height within three cycles of the taken at several points Internals hold down spring beginning of the license renewal around the circumference of Note: This period. If the first set of the spring, with a mechanism was not measurements is not sufficient to statistically adequate strictly identified in determine life, spring height number of measurements at the original list of measurements must be taken each point to minimize age-related during the next two outages, in uncertainty.

Replacement degradation order to extrapolate the expected of 304 springs by 403 mechanisms, spring height to 60 years. springs is required when the spring stiffness is determined to relax beyond design tolerance.

See Figure 2-10 Page 38 NL-1 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability (Mechanism)

Expansion Link Method/Frequency Examination Coverage Thermal Shield Assembly IPEC Units 2 and Cracking (Fatigue)

None Visual (VT-3) no later than 2 100% of thermal shield Thermal shield flexures 3 or Loss of refueling outages from the flexures Materials (Wear) beginning of the license renewal that results in period. Subsequent See Figures 2-I1 and 2-16 thermal shield examinations on a ten year flexures excessive interval.wear, fracture or complete separation Core Barrel Assembly IPEC Units 2 and Cracking (IASCC, None Enhanced visual (EVT-l) 100% of one side of the Upper and lower core barrel 3 Neutron examination, no later than 2 accessible surfaces of the welds Embrittlement) refueling outages from the selected weld and adjacent beginning of the license renewal base metal.period and subsequent examination on a ten-year See Figure 2-4 interval.Core Barrel Assembly IPEC Units 2 and Cracking (IASCC, None Enhanced visual (EVT-1) 100% of one side of the Lower core barrel flange weld 3 Neutron examination, no later than 2 accessible surfaces of the Embrittlement) refueling outages from the selected weld and adjacent (At IPEC this weld is the lower beginning of the license renewal base metal.core barrel to lower support period and subsequent casting weld. IPEC does not examination on a ten-year See Figure 2-4 (Core Barrel have a lower core barrel flange) interval.

to Support Plate Weld)Page 39 NL-l 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel hiternals hIspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Effect EaiainMto Item Applicability (Mechanism)

Primary Link Examination Method Examination Coverage Core Barrel Assembly IPEC Units 2 Cracking (IASCC, Baffle-former Volumetric (UT) examination, 100% of accessible bolts.Barrel-former bolts and 3 Fatigue) bolts with initial examinations The inspection shall dependent on results of baffle- examine a minimum former bolt examinations.

Re- sample size of 75% of the examinations at 10 year total population of bolts.intervals once degradation is Accessibility may be identified in the primary limited by presence of component.

thermal shields. 75% of a component's total (accessible

+ inaccessible) inspection area or volume will be examined or, when addressing a set of like components (e.g., bolting), that the inspection examine a minimum sample size of 75 percent of the total population of like components.

For the inspection of a set of like components, it is understood that essentially 100% of the volume/area of each accessible like component will be examined.See Figure 2-5 Page 40 NL-l 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Item Applicability (EMechanism)

Primary Link Examination Method Examination Coverage Lower Support Assembly IPEC Units 2 Cracking (IASCC, Baffle-former Volumetric (UT) examination, 100% of accessible bolts Lower support column bolts and 3 Fatigue) bolts with initial examinations or as supported by plant-dependent on results of baffle- specific justification.

The former bolt examinations.

Re- inspection shall examine a examinations at 10 year minimum sample size of intervals once degradation is 75 percent of the total identified in the primary population of bolts. 75%component.

of a component's total (accessible

+ inaccessible) inspection area or volume will be examined or, when addressing a set of like components (e.g., bolting), that the inspection examine a minimum sample size of 75 percent of the total population of like components.

For the inspection of a set of like components, it is understood that essentially 100% of the volume/area of each accessible like component will be examined.See Figures 2-12 and 2-13 Page 41 NL- I1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Effect Coverage Meho Item Applicability (Mechanism)

Primary Link Examination Method Examination Coverage Core Barrel Assembly IPEC Units 2 Cracking (SCC, Upper core barrel Enhanced visual (EVT-1) 100% of one side of the Core barrel flange, and 3 Fatigue) flange weld examination, with initial accessible surfaces of the examination frequency selected weld and adjacent Core barrel outlet nozzles dependent on the examination base metal. 75% of a results for upper core barrel component's total flange. Re-examinations at 10 (accessible

+ inaccessible) year intervals once degradation inspection area or volume is identified in the primary will be examined or, when component.

addressing a set of like components (e.g., bolting), that the inspection examine a minimum sample size of 75 percent of the total population of like components.

For the inspection of a set of like components, it is understood that essentially 100% of the volume/area of each accessible like component will be examined.See Figure 2-4 Page 42 NL-1 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Effect EaiainMto Item Applicability (Mechanism)

Primary Link Examination Method Examination Coverage Lower Support Assembly IPEC lower support column Lower support column bodies bodies are cast.(non cast) They are captured in the next Item of this table.Lower Support Assembly IPEC Units 2 Cracking Control rod guide Visual (EVT-1) examination.

100% of accessible Lower support column bodies and 3 (IASCC) tube (CRGT) Re-examinations at 10 year support columns. 75% of a including the lower flanges intervals once degradation is component's total (cast) detection of identified in the primary (accessible

+ inaccessible) fractured support component.

inspection area or volume columns will be examined or, when addressing a set of like components (e.g., bolting), that the inspection examine a minimum sample size of 75 percent of the total population of like components.

For the inspection of a set of like components, it is understood that essentially 100% of the volume/area of each accessible like component will be examined.See Figure 2-14 Page 43 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Item Applicability Effect Primary Link Examination Method Examination Coverage Bottom Mounted IPEC Units 2 Cracking Control rod guide Visual (VT-3) examination of 100% of BMI column Instrumentation System and 3 (Fatigue) tube (CRGT) BMI column bodies as bodies for which difficulty Bottom-mounted including the lower flanges indicated by difficulty of is detected during flux detection of insertion/withdrawal of flux thimble instrumentation (BMI) column completely thimbles.

Flux thimble insertion/withdrawal.

fractured column insertion/withdrawal to be bodies monitored at each inspection interval.

Re-examinations at See Figure 2-15 10 year intervals once degradation is identified in the primary component.

Upper Internals Assembly IPEC Units 2 Cracking (SCC, Control rod guide Enhanced visual (EVT-1) 100% of accessible upper Upper core plate and 3 Fatigue) tube (CRGT) examination, with initial core plate. 75% of a lower flange weld examination frequency component's total dependent on the examination (accessible

+ inaccessible) results for CRGT lower flange inspection area or volume weld. Re-examinations at 10 will be examined or, when year intervals once degradation addressing a set of like is identified in the primary components (e.g., bolting), component.

that the inspection examine a minimum sample size of 75 percent of the total population of like components.

For the inspection of a set of like components, it is understood that essentially 100% of the volume/area of each accessible like component will be examined.

See Figure 2-1 Page 44 NL-1 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Item Effect Examination MethodCoverage ImApplicability (Mechanism)

Primary Link ExaminationCMethod Lower Support Assembly IPEC Units 2 Cracking (SCC, Control rod guide Enhanced visual (EVT-1) 100% of accessible lower Lower support casting and 3 Fatigue) tube (CRGT) examination, with initial support casting. 75% of a lower flange weld examination frequency component's total dependent on the examination (accessible

+ inaccessible) results for CRGT lower flange inspection area or volume weld. Re-examinations at 10 will be examined or, when year intervals once degradation addressing a set of like is identified in the primary components (e.g., bolting), component.

that the inspection examine a minimum sample size of 75 percent of the total population of like components.

For the inspection of a set of like components, it is understood that essentially 100% of the volume/area of each accessible like component will be examined.See Figure 2-1 (Core Support)Page 45 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-4 Existing Program Components at IPEC Units 2 and 3 Effect Item Applicability (Mechanism)

Primary Link Examination Method Examination Coverage Core Barrel Assembly IPEC Units 2 Loss of material ASME Code Visual (VT-3) examination All accessible surfaces at Core barrel flange and 3 (Wear)Section XI to determine general ASME Section XI specified condition for excessive frequency.

wear.Upper Internals Assembly N/A N/A N/A N/A N/A Upper support ring or skirt (This item is N/A because IPEC has a tophat design)Lower Internals Assembly IPEC Units 2 Cracking (IASCC, ASME Code Visual (VT-3) examination All accessible surfaces at Lower core plate and 3 Fatigue)Section XI of the lower core plates to ASME Section XI specified detect evidence of frequency.

distortion and/or loss of bolt integrity.

Lower Internals Assembly IPEC Units 2 Loss of material ASME Code Visual (VT-3) All accessible surfaces at Lower core plate and 3 (Wear)Section XI examination.

ASME Section XI specified frequency.

Bottom Mounted IPEC Units 2 Loss of material N/A Surface (ET) examination.

N/A Instrumentation System and 3 (Wear)Flux thimble tubes Alignment and Interfacing IPEC Units 2 Loss of material ASME Code Visual (VT-3) All accessible surfaces at Components and 3 (Wear)Section XI examination.

ASME Section XI specified Clevis insert bolts frequency.

Alignment and Interfacing IPEC Units 2 Loss of material ASME Code Visual (VT-3) All accessible surfaces at Components and 3 (Wear)Section XI examination.

ASME Section XI specified Upper core plate alignment pins frequency.

Page 46 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Expansion Link(s) Expansion Criteria Additional Examination Item Applicability Criteria (Note 1) E Acceptance Criteria Control Rod Guide Tube IPEC Units 2 Visual (VT-3) None N/A N/A Assembly and 3 examination.

Guide plates (cards)The specific relevant condition is wear that could lead to loss of control rod alignment and impede control assembly insertion.

Control Rod Guide Tube IPEC Units 2 Enhanced visual (EVT-!) a. Bottom-mounted

a. Confirmation of surface- a. For BMI column bodies, Assembly and 3 examination, instrumentation (BMI) breaking indications in two or the specific relevant Lower flange welds column bodies more CRGT lower flange welds, condition for the VT-3 combined with flux thimble examination is completely The specific relevant insertion/withdrawal difficulty, fractured column bodies.condition is a detectable

b. Lower support shall require visual (VT-3)crack-like surface column bodies (cast) examination of BMI column indication.

bodies by the completion of the b. For cast lower support next refueling outage. column bodies, the specific relevant condition is a detectable crack-like

b. Confirmation of surface- surface indication.

breaking indications in two or more CRGT lower flange welds shall require EVT- I examination of cast lower support column bodies within three fuel cycles following the initial observation.

Page 47 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Intenials Inspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Additional Examination Item Applicability Criteria (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Core Barrel Assembly IPEC Units 2 Enhanced visual (EVT-1) None N/A N/A and 3 examination.

Upper core barrel flange weld Upper and lower core barrel The specific relevant welds condition is a detectable crack-like surface Lower core bare] flange indication.

weld (At IPEC this weld is the lower core barrel to lower support casting weld.IPEC does not have a lower core barrel flange)Core barrel flange Core barrel outlet nozzles Baffle-Former Assembly Baffle-edge bolts IPEC Units 2 and 3 Visual (VT-3)examination.

None N/A N/A The specific relevant conditions are missing or broken locking devices, failed or missing bolts, and protrusion of bolt heads.* L L Page 48 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Expansion Link(s) Expansion Criteria Additional Examination Item Applicability Criteria (Note 1) Acceptance Criteria Baffle-Former Assembly IPEC Units 2 Volumetric (UT) a. Lower support a. Confirmation that more than a and b. The examination and 3 examination, column bolts 5% of the baffle-former bolts acceptance criteria for the Baffle-former bolts actually examined on the four UT of the lower support baffle plates at the largest distance column bolts and the The examination

b. Barrel-former bolts from the core (presumed to be the barrel-former bolts shall be acceptance criteria for the lowest dose locations) contain established as part of the UT of the baffle-former unacceptable indications shall examination technical bolts shall be established as require UT examination of the justification.

part of the examination lower support column bolts within technical justification.

the next three fuel cycles.b. Confirmation that more than 5% of the lower support column bolts actually examined contain unacceptable indications shall require UT examination of the barrel-former bolts.Page 49 NL-l 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Item Applicability Examination Acceptance Expansion Link(s) Expansion Criteria Additional Examination Criteria (Note 1) E Acceptance Criteria Baffle-Former Assembly IPEC Units 2 Visual (VT-3) None N/A N/A Assembly and 3 examination.

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.Alignment and Interfacing IPEC Units 2 Direct physical None N/A N/A Components and 3 measurement of spring Internals hold down spring height.The examination acceptance criterion for this measurement is that the remaining compressible height of the spring shall provide hold.down forces within the plant-specific design tolerance.

Page 50 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Expansion Link(s) Expansion Criteria Additional Examination Item Applicability Criteria (Note 1) Acceptance Criteria Thermal Shield Assembly IPEC Units 2 Visual (VT-3) None N/A N/A and 3 examination.

Thermal shield flexures The specific relevant conditions for thermal shield flexures are excessive wear, fracture, or complete separation.

Upper Internals Assembly IPEC Units 2 Enhanced visual (EVT-1) None N/A N/A Upper core plate and 3 examination.

The specific relevant condition is a detectable crack-like surface indication.

Lower Support Assembly IPEC Units 2 Enhanced visual (EVT-I) None N/A N/A Lower support casting and 3 examination.

The specific relevant condition is a detectable crack-like surface indication.

Notes: I. The examination acceptance criterion for visual examination is the absence of the specified relevant condition Page 51 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I hidian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-6 Reactor Vessel Component ISI Program Inspection Plan for IPEC Units 2 and 3 Code Examination Component Extent of Exam Category Method Reactor Vessel Interior B-N-2 VT-I or VT-3 Components and areas as accessible Radial Support Keys Reactor Vessel Interior B-N-2 VT-I or VT-3 Components and areas as accessible Bottom Head Instrumentation Nozzles Reactor Vessel Interior B-N-2 VT- I or VT-3 Components and areas as accessible Outlet and Inlet Nozzle mating surfaces and inside of nozzles Reactor Vessel Interior B-N-2 VT-I or VT-3 Components and areas as accessible Upper internal to vessel mating surface with keys and access slots Reactor Vessel Interior B-N-2 VT-I or VT-3 Components and areas as accessible Vessel flange surface Lower Internals

-Exterior B-N-3 VT-3 Components and areas as accessible Core barrel surface Lower Internals

-Exterior B-N-3 VT-3 Components and areas as accessible Thermal Shield Lower Internals

-Exterior B-N-3 VT-3 Components and areas as accessible Irradiation specimen tubes and guides Lower Internals

-Exterior B-N-3 VT-3 Components and areas as accessible Flexures Page 52 NL-1 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-6 Reactor Vessel Component ISI Program Inspection Plan for IPEC Units 2 and 3 Code Examination Component Extent of Exam Category Method Lower Internals

-Exterior B-N-3 VT-3 Components and areas as accessible Fasteners and locking devices Lower Internals

-Exterior B-N-3 VT-3 Components and areas as accessible Outlet nozzle at 22 deg Lower Internals

-Exterior B-N-3 VT-3 Components and areas as accessible Outlet nozzle at 158 deg Lower Internals

-Exterior B-N-3 VT-3 Components and areas as accessible Outlet nozzle at 202 deg Lower Internals

-Exterior B-N-3 VT-3 Components and areas as accessible Outlet nozzle at 338 deg Lower Internals

-Exterior Bottom B-N-3 VT-3 Components and areas as accessible Lower core support plate Lower Internals

-Exterior Bottom B-N-3 VT-3 Components and areas as accessible Flow distribution plate Lower Internals

-Exterior Bottom B-N-3 VT-3 Components and areas as accessible Lower support casting Lower Internals

-Exterior Bottom B-N-3 VT-3 Components and areas as accessible Core support column Lower Internals

-Exterior Bottom B-N-3 VT-3 Components and areas as accessible Secondary core support Page 53 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-6 Reactor Vessel Component ISI Program Inspection Plan for IPEC Units 2 and 3 Code Examination Component Category Method Extent of Exam Lower Internals

-Exterior Bottom B-N-3 VT-3 Components and areas as accessible Instrumentation guides Lower Internals

-Exterior Bottom B-N-3 VT-3 Components and areas as accessible Radial support keys Lower Internals

-Interior Bottom B-N-3 VT-3 Components and areas as accessible Outlet nozzle at 22 deg Lower Internals

-Interior Bottom B-N-3 VT-3 Components and areas as accessible Outlet nozzle at 158 deg Lower Internals

-Interior Bottom B-N-3 VT-3 Components and areas as accessible Outlet nozzle at 202 deg Lower Internals

-Interior Bottom B-N-3 VT-3 Components and areas as accessible Outlet nozzle at 338 deg Lower Internals

-Interior Bottom B-N-3 VT-3 Components and areas as accessible Core barrel alignment pin Lower Internals

-Interior Bottom B-N-3 VT-3 Components and areas as accessible Lower core plate Lower Internals

-Interior Bottom B-N-3 VT-3 Components and areas as accessible Fuel alignment pins I I Page 54 NL-l 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-7 List of IPEC Reactor Vessel Interior Components and Materials Based on MRP-191 -Table 4-4 UPPER INTERNALS ASSEMBLY Sub Assembly Component Material Category from MRP-191 Table 7-2 Anti-rotation studs and nuts Stainless steel A Bolts Stainless steel A C-tubes Stainless steel C Enclosure pins Stainless steel A Upper guide tube enclosures Stainless steel A Flanges intermediate Stainless steel A Flanges lower Stainless steel A Flexureless inserts Stainless steel A Guide plates/cards Stainless steel C Control rod guide Guide tube support pins (split pins) A X-750 (IP2 only) C tube assemblies and Guide tube support pins (split pins) Stainless steel (IP3 only) A flow downcomers Housing plates Stainless steel A Inserts Stainless steel A Lock bars Stainless steel A Sheaths Stainless steel C Support pin cover plate Stainless steel A Support pin cover plate cap screws Stainless steel A Support pin cover plate locking caps Stainless steel A and tie straps Support pin nuts Alloy X-750 A Support pin nuts Stainless steel A Water flow slot ligaments Stainless steel A Mixing Devices Mixing devices CASS A Upper core plate and Fuel alignment pins Stainless steel A fuel alignment pins Upper core plate Stainless steel A Bolting Stainless steel A Bracketsclampsterminal blocks, and conduit straps Stainless steel A Upper Conduit seal assembly-body, Stainless steel A instrumentation tubesheets conduit and supports Conduit seal assembly-tubes Stainless steel A Conduits Stainless steel A Flange base Stainless steel A Locking caps Stainless steel A Support tubes Stainless steel A UHI flow column bases CASS A Upper plenum UHI flow columns Stainless steel A Page 55 NL-l 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-7 List of IPEC Reactor Vessel Interior Components and Materials Based on MRP-191 -Table 4-4 UPPER INTERNALS ASSEMBLY Sub Assembly Component Material Category from MRP-191 Table 7-2 Adapters Stainless steel A Bolts Stainless steel A Column bases CASS A Upper support Column bodies Stainless steel A column assemblies Extension tubes Stainless steel A Flanges Stainless steel A Lock keys Stainless steel A Nuts Stainless steel A Bolts Stainless steel A Deep beam ribs Stainless steel A Deep beam stiffeners Stainless steel A Flange Stainless steel A Upper support plate Inverted top hat flange Stainless steel A assembly Inverted top hat upper support plate Stainless steel A Lock keys Stainless steel A Ribs Stainless steel A Upper suport plate Stainless steel A Upper support ring or skirt Stainless steel B LOWER INTERNALS ASSEMBLY Sub Assembly Component Material Category from MRP-191 Table 7-2 Baffle bolting locking bar Stainless steel A Baffle edge bolts Stainless steel C Baffle and former Baffle plates Stainless steel B assembly Baffle former bolts Stainless steel C Barrel former bolts Stainless steel C Former plates Stainless steel B BMI column bodies Stainless steel B BMI column bolts Stainless steel A Bottom mounted BMI column collars Stainless steel B instrumentation BMI column cruciforms CASS B (BMI) column BMI column extension bars Stainless steel A.assemblies BMI column extension tubes Stainless steel B BMI column lock caps Stainless steel A BMI column nuts Stainless steel A Core barrel flange Stainless steel B Core barrel Core barrel outlet nozzles Stainless steel B Upper core barrel Stainless steel C Lower core barrel Stainless steel C Diffuser plate Diffuser plate Stainless steel A Page 56 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-7 List of IPEC Reactor Vessel Interior Components and Materials Based on MRP-191 -Table 4-4 LOWER INTERNALS ASSEMBLY Sub Assembly Component Material Category from MRP-191 Table 7-2 Flux thimble tube plugs -IPEC does not use tube plugs. instead tubes are B Flux thimbles (tubes) capped (IP2 has 9 tubes capped, IP3 has 0 tubes capped)Flux thimbles (tubes) Stainless steel C Irradiation specimen guide Stainless steel A Irradiation specimen Irradiation specimen guide bolts Stainless steel A guides Irradiation specimen lock caps Stainless steel A Specimen plugs Stainless steel A Lower core plate Fuel alignment pins Stainless steel A (LCP) and fuel LCP fuel alignment pin bolts Stainless steel A alignment pins LCP fuel alignment pin lock caps Stainless steel A Lower core plate Stainless steel C Lower support column bodies CASS B Lower support Lower support column bolts Stainless steel B column assemblies Lower support column nuts Stainless steel A Lower support column sleeves Stainless steel A Lower support casting or forging Lower support casting CASS A Thermal shield bolts Stainless steel A Neutron Thermal shield dowels Stainless steel A panels/thermal shield Thermal shield flexures Stainless steel B Thermal shield Stainless steel A Radial support key bolts Stainless steel A Radial support keys Radial support key lock keys Stainless steel A Radial support keys Stainless steel A SCS base plate Stainless steel A Secondary core SCS bolts Stainless steel A Suponrt (cor SCS energy absorber Stainless steel A sembly SCS guide posts Stainless steel A SCS housing Stainless steel A SCS lock keys Stainless steel A Clevis insert bolts A X-750 B Clevis insert lock keys Stainless steel A Clevis inserts Alloy 600 A Interfacing Head and vessel allignment pin bolts Stainless steel A Components Head and vessel alignment pin lock Stainless steel A caps Head and vessel allignment pins Stainless steel A Internals hold down spring 304 Stainless steel B Upper core plate alignment pins Stainless steel B Page 57 NL-I 1-107 Docket Nos. 50-247 and 50-286 Attachment I Indian Point Energy Center Reactor Vessel internals Inspection Plan Table 5-8 IPEC Response to the NRC Final Safety Evaluation of MRP-227 MRP-227 SER Item IPEC Response SER Section 4.1.1, Topical Report In accordance with SER Section 4. 1. 1, the upper core plate and the lower Condition I Moving components to support casting have been added to the IPEC "Expansion" inspection category"Expansion" category from "No and are contained in Table 5-3. The components are linked to the "Primary" additional measures" category.

component CRGT lower flange weld. The examination method is consistent with the examinations performed on the CRGT lower flange weld.SER Section 4.1.2, Topical Report In accordance with SER Section 4.1.2, the upper and lower core barrel welds Condition 2 Inspection of and lower core barrel to lower support casting weld have been added to the components subject to irradiation-IPEC "Primary" inspection category and are contained in Table 5-2. The assisted stress corrosion cracking examination method is consistent with the MRP recommendations for these components, the examination coverage conforms to the criteria described in Section 3.3.1 of the NRC SE, and the re-examination frequency is on a 10-year interval consistent with other "Primary" inspection category components.

SER Section 4.1.3, Topical Report No action required.

This item does not apply to components in Westinghouse Condition 3 Inspection of high designed reactors.consequence components subject to multiple degradation mechanisms SER Section 4.1.4, Topical Report In accordance with SER Section 4.1.4, IPEC will meet the minimum inspection Condition 4 Minimum examination coverage specified in the SER. The appropriate wording has been added to coverage criteria for "expansion" Table 5-3 examination coverage.inspection category components SER Section 4.1.5, Topical Report In accordance with SER Section 4.1.5, the examination frequency for baffle-Condition 5 Examination former bolts specifies a 10-year inspection frequency following the baseline frequencies for baffle-former bolts inspection in Table 5-2.SER Section 4. 1.6, Topical Report In accordance with SER Section 4.1.6, Table 5-3 requires a I 0-year re-Condition 6 Periodicity of the re- examination interval for all Expansion inspection category components once examination of "expansion" degradation is identified in the associated Primary inspection category inspection category components component and examination of the expansion category component commences.

SER Section 4.1.7. Topical Report This condition applies to update of the industry guidelines.

No plant-specific ConditionseUpating of industry action required.guideline SER Section 4.2.41, The evaluation of design and operating history demonstrating that MRP-227 is Applicant/Licensee Action Item 4 applicable to IPEC is contained in Section 3.6.SER Section 4.2.2, The IPEC review of components within the scope of license renewal against the Applicant/Licensee Action Item 2 information contained in MRP-191 Table 4-4 is discussed in Section 3.6.SER Section 4.2.3, The IPEC discussion regarding guide tube support pins (split pins) is contained Applicant/Licensee Action Item 3 in Section 3.6.SER Section 4.2.4, No action required.

This item does not apply to Westinghouse designed units.Applicant/Licensee Action Item 4 SER Section 4.2.5. The IPEC discussion regarding hold down springs is contained in Section 3.6.Applicant/Licensee Action Item 5 SER Section 4.2.6, No action required.

This item does not apply to Westinghouse designed units.Applicant/Licensee Action Item 6 SER Section 4.2.7. The IPEC discussion regarding lower support column bodies is contained in Applicant/Licensee Action Item 7 Section 3.6.SER Section 4.2.8, The submittal of information for staff review and approval is discussed in Applicant/Licensee Action Item 8 Section 3.6.Page 58