ML12277A174
| ML12277A174 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 09/28/2012 |
| From: | Mogren T Constellation Energy Nuclear Group, EDF Group, Ginna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RIS 2011-007 LR-RVI-PROGPLAN, Rev. 3 | |
| Download: ML12277A174 (120) | |
Text
Thomas G. Mogren Manager - Engineering Services R.E. Ginna Nuclear Power Plant, LLC CENG a joint venture of 1503 Lake Road Ontario, New York 14519-9364 O Canstluon #--%D 585.771.5208 585.771.3392 Fax Thomas.Mogren@cengllc.com September 28, 2012 U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 ATTENTION: Document Control Desk
SUBJECT:
RE. Ginna Nuclear Power Plant Docket No. 50-244 License Renewal Aiin2 Manaaement Submit Revised Reactor Vessel Internals Program Document in Accordance with RIS 2011-07
REFERENCES:
(1) Letter from J. E. Pacher, CENG, to the U.S. NRC Document Control Desk,
Subject:
License Renewal Aging Management, Reactor Vessel Internals Program, dated February 27, 2009 (ADAMS Accession No.
(2) U.S. Nuclear Regulatory Commission, "NRC Regulatory Issue Summary 2011-07, License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management," dated July 21, 2011 (ADAMS Accession No. ML111990086).
(3) Letter from T. G. Mogren, CENG, to the U.S. NRC Document Control Desk,
Subject:
License Renewal Aging Management, Withdraw Reactor Vessel Internals Program Document and Commit to Submit Revised Reactor Vessel Internals Program Document in Accordance with RIS 2011-07, dated September 13, 2011 (ADAMS Accession No.
MLI 1263A004)
By Reference (1), R.E. Ginna Nuclear Power Plant (Ginna) submitted for NRC staff review and approval the document titled "Reactor Vessel Internals Program" based on the EPRI Report MRP-227, Revision 0. In accordance with Reference (2), Ginna is classified as a Category A plant. Therefore, by Reference (3), Ginna chose to: (a) withdraw the submittal in Reference (1);
and (b) commit to resubmit the aging management program (AMP)/inspection plan in accordance with the NRC-approved report MRP-227-A no later than October 1, 2012.
Document Control Desk September 28, 2012 Page 2 By this letter, Ginna submits the document titled "Reactor Vessel Internals Program," Rev. 3, which is based on the NRC-approved report MRP-227-A. There are no regulatory commitments in this submittal.
Should you have questions regarding this matter, please contact Thomas Harding at (585) 771-5219 or via e-mail at Thomas.HardingJr@cengllc.com.
Very truly yours, Thomas G. Mogren
Attachment:
(1) Reactor Vessel Internals Program, LR-RVI-PROGPLAN, Rev. 3 cc: W. M. Dean, NRC M.C. Thadani, NRC Resident Inspector, NRC (Ginna)
ATTACHMENT 1 REACTOR VESSEL INTERNALS PROGRAM LR-RVI-PROGPLAN REVISION 3 R.E. Ginna Nuclear Power Plant, LLC September 28, 2011
LICENSE RENEWAL AGING MANAGEMENT PROGRAM BASIS DOCUMENT GINNA STATION REACTOR VESSEL INTERNALS PROGRAM LR-RVI-PROGPLAN Revision 3 Prepared By: Rober+ L^rcelleo Ar46& 1 A.012I Aging Management Program Owner / Date Reviewed By: Z/.A ý1-2 (1/ 2 -
License Renewal*Cordinator / Date Approved By: ______
ý/I'/2.
. Lic+ne Renewal Project Manager
/ Date
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 TABLE OF CONTENTS 1.0 P URPPOSE .................................................................................................................................................. 4 2.0 BA CK GRO UND ........................................................................................................................................ 5 3.0 PROGRAM OWNER ................................................................................................................................ 9
4.0 DESCRIPTION
OF THE GINNA AMP andINDUSTRY PROGRAMS ............................................ 9 5.0 GINNA AGING M4ANA GEMENT PROGRAMA TTRIBUTE EVALUATIONS ............................ 26 5.1 SRP Element 1 - Scope of Program ............................................................................................ 26 5.2 SRP Element 2 - Preventive Actions ...................... ............. ............. 28 5.3 SRP Element 3 - Parameters Monitored, Inspected, and/or Tested ...........................28 5.4 SRP Element 4 - Detection of Aging Effects................................... . .......... 39 5.5 SRP Element 5- Monitoring and Trending ................ .................................... 42 5.6 SRP Element 6 - Acceptance Criteria ..... ................................................. 43 5.7 SRP Element 7 - Corrective Actions ....... ................................ ... 43 5.8 SRP Element 8 - Confirmation Process ........................................... 44 5.9 SRP Element 9- Administrative Controls ................................................................................. 44 5.10 SRP Element 10 - Operating Experience ...... - ............................................ 45 6.0 DEMONSTRATION ......... .................................................... 47 7.0 REQUIRED PROGRAM ENHANCEMENTS/IMPLEMENTATION SCHEDULE..................... 48 8.0
SUMMARY
OFIMPLEMENTING DOCUMENTS/RESPONSIBLE DEPARTMENT ............... 48
9.0 REFERENCES
..................................................................... ............................................... 49 A TTACHMENTA - Summary of Implementing Documents /Enhancements / Testing /
InitialInspections/Actions................................................................................................... 57 ATTACHMENT B - PBD - PEN&INK Non-Intent Change Formfor Attachment A, Table I .......................... 60 A TTACHMENT C - Ginna Station Aging ManagementReview Summary Table LRAM-R VI .......................... 61 A TTA CHMENT D - MRP-227 Revision 0 Westinghouse Tables........................................................................ 67 ATTACHMENT E- MRP-22 7-A Westinghouse Tables....................................................................................... 80 A TTACHMENT F - Scope LR-R VI-PR OGPLAN ................................... ............ ... 95 A TTACHMENT G (Fig.1) - Typical Westinghouse Internals................................................................................. 101 ATTA CHMENT G (Fig.2) - Guide Card................................ ....................................... ..................... 102 A TTACHMENT G (Fig.3) - Control Rod Guide Tube Assembly .................................. 103 A TTACHMENT G (Fig.4) - Core Barrel Welds Inspected at GinnaDuring2011 RFO ...................................... 104 Page 2 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT G (Fig.5) - Bolt Locations in Westinghouse Baffle-FormerAssembly ....................................... 105 ATTACHMENT G (Fig.6) - Section of Baffle Former-Assembly ........................................................................... 106 A TTACHMENT G (Fig. 7) - Void Swelling Induced Distortionin West. Baffle-FormerAssembly .......... 107 ATTACHMENT G (Fig.8) - VerticalDisplacementof West. Baffle Plates Causedby Void Swelling .................. 108 A TTACHMENT G (Fig.9) - Thermal Shield Flexure............................................................................................. 109 A TTACHMENT G (Fig.10) - Detail Thermal Shield Flexure................................................................................ 110 ATTACHMENT G (Fig.11) - Upper InternalsAssembly ........................................................................................ 111 ATTACHMENT G (Fig.12) - Clevis Insert Details.................................................................................................. 112 ATTACHMENT H- Ginna'sResponse to Applicant/LicenseeAction Itemsfrom Revision 1 to the NRC Safety Evaluation of MRP-227 Rev. 0 ..................................................................................... 113 Page 3 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date. 9/18/2012 1.0 PURPOSE The purpose of this Program Basis Document (PBD) is to document the Ginna Station Reactor Vessel Internals Aging Management Program (AMP), and identify those activities of the Reactor Internals AMP that are credited for license renewal. This document provides a description of the program as it relates to the management of aging effects identified in the aging management review process documented in LRAM-RVI (Ref. 9.35). This PBD also captures additional industry guidance for reactor internals augmented inspections that are contained in the EPRI Materials Reliability Program, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227).
This PBD demonstrates thatfthe program adequately manages the effects of aging for reactor internals components and establishes the basis for providing reasonable assurance that the internals components will continue to perform their intended function through the Ginna license renewal period of extended operation. The Ginna License Renewal Program is committed to Revision 0 ofNUREG-1801, "Generic Aging Lessons Learned (GALL) Report."
The Ginna Station Reactor Internals AMP complies with each of the ten elements in Section XI.M. 16 of this NUREG, as shown in Section 5 of this PBD.
Revision 3 of this PBD provides additional inspection requirements and acceptance criteria for reactor internals not previously documented in the LRAM-RVI. As noted in the LRAM-RVI, additional requirements for inspections were to be developed by the industry on a generic applicability basis. These additional requirements are captured by this PBD and are based on the work of the EPRI-MRP and the PWROG.
The development and submittal of Revision 2 of this PBD to the NRC met a license renewal commitment to work with the industry to develop a reactor vessel internals aging management program. Revision 3 of this PBD addresses Regulatory Issue Summary (RIS) 2011-07 (Ref.
9.85). Per RIS 2011-07, Ginna is classified as a Category A plant. As such, Ginna withdrew the formerly submitted Revision 2 of this PBD from the NRC, revised it to include the inspection results performed under MRP-227 Revision 0 during the Ginna 2011 RFO, and included Ginna's fiture commitment to the requirements of MRP-227-A. Justifications for each gap between MRP-227 Revision 0 and MRP-227-A guidelines are also documented in this PBD. Revision 3 of this PBD is being submitted to the NRC prior to October 1, 2012.
Upon the NRC staff review of MRP-227 Revision 0, a NRC Safety Evaluation was released and later revised. Revision 1 of the Final Safety Evaluation is included in MRP-227-A and identifies the NRC staff s concerns related to conditions and limitations on the use of MRP-227. The Topical Report Conditions are addressed and incorporated in MRP-227-A. The Safety Evaluation also addressed the NRC staffs concerns relating to plant-specific action items. Ginna's response to these plant specific Applicant/Licensee Action Items are included in Attachment H of this PBD.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012
2.0 BACKGROUND
2.1 Ginna License Renewal Background By letter dated July 30, 2002, Rochester Gas & Electric Corporation (RG&E, the applicant at that time) submitted the License Renewal Application (LRA) for Ginna Station in accordance with Title 10, Part 54, of the Code of Federal Regulations (10 CFR Part 54).
Through the LRA. RG&E requested that the U.S. Nuclear Regulatory Commission (NRC) renew the operating license for Ginna (license number DPR- 18) for a period of 20 years beyond the current expiration of midnight, September 18, 2009. The safety evaluation report (SER) NUREG- 1786 documented the technical review of the R.E. Ginna Nuclear Power Plant (Ginna) license renewal application (LRA) by the U.S. Nuclear Regulatory Commission Staff.
Section 6 of the SER concludes that based on the evaluation of the application as discussed in the SER, the staff determined that the requirements of 10 CFR 54.29(a) were met by the Ginna application.
The Renewed Facility Operating License No. DPR-18 for R.E. Ginna Nuclear Power Plant was granted as documented in the NRC letter dated May 19, 2004 (Ref. 9.21). Reference 9.21 identifies the technical basis for issuing the renewed license as being set forth in the NUREG-1786, Safety Evaluation Report Related to the License Renewal of the R.E. Ginna Nuclear Power Plant (Ref. 9.34).
2.2 Ginna Reactor Internals Aging Management Review/Industry Program Background Per SER NUREG 1786, Ginna Station committed to establishing and providing an AMP for the Reactor Internals components based on the required ten elements of the GALL Report (Ref.
9.1). This commitment was formally tracked to completion in NL-2007-000064-004.
The initial work performed to support the Ginna station application for license renewal included an aging management review LRAM-RVI (Ref 9.35) utilizing the methodology that was established by the WOG (now PWROG) published in WCAP-14577-1-A (Ref 9.2). The aging management approach utilized by the WCAP was found acceptable by the NRC as documented in their letter dated Feb. 10, 2001(Ref. 9.66). Per the NRC letter, the WCAP approach was found acceptable, with NRC identified limitations of committing to the accepted aging management programs defined in the WCAP, and completing the renewal applicant action items defined in section 4.1 of the WCAP final safety evaluation report. Therefore, application of the WCAP methodology provides reasonable assurance that the effects of aging will be adequately managed so that the intended functions of the reactor internals would be maintained consistent with the current licensing basis during the period of extended operation.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 As presented in the AMR output table of LRAM-RVI (Ref. 9.35) provided herein as Attachment C, a combination of existing programs, and additional work, to be identified by the "Reactor Vessel Internals Inspection Program" was credited for aging management of the Ginna Reactor Internals. The table in Attachment C, includes in the last column, those line items where additional Industry evaluations or inspection technique development were anticipated in order to complete this LR-RVI-PROGPLAN or "Reactor Vessel Internals Inspection Program" as referenced in the Attachment C summary table.
The additional industry work on the Reactor Vessel Internals Inspection Program, referred to in the Ginna LRAM-RVI, culminated in the submittal of "PWR Internals Inspection and Evaluation Guidelines" MRP-227, Rev. 0 (Ref. 9.38) to the NRC for review and approval in January of 2009. The industry program is intended to provide a consistent approach to aging management of PWR reactor internals components across the PWR fleet. The NRC reviewed and approved version was issued in December 2011 as MRP-227-A: "Materials Reliability Program: Pressurized Water Reactor Internals and Evaluation Guidelines" (Ref. 9.71).
2.3 Ginna Reactor Internals Aging Management Program Background In this PBD the Ginna Station Reactor Vessel Internals (RVI) AMP is compared to the 10 attributes ofNUREG-1801, Generic Aging Lessons Learned (GALL) Report, Rev. 0,Section XI.M. 16, "PWR Vessel Internals" (Ref. 9.1). The AMP credits inspections from existing Ginna station programs, industry programs issued under the guidance of NEI-03-08 (Ref. 9.54) as implemented by Ginna Procedures (Ref 9.83 and 9.82), previous industry operating experience, and incorporates recommendations for additional inspections provided by Industry Guidelines contained in MRP-227 Revision 0 and MRP-227-A (Ref. 9.38 and 9.71). Revision 2 of this PBD satisfies a Ginna NRC licensing commitment for life extension. The commitment is number #31, and is contained in the Ginna SER, NUREG -1786 (Ref. 9.37).
Attachment A provides a list of action items that were required to implement this program.
Actions that have been completed have been noted as such in column 7 and have been shaded.
Actions that must be completed prior to support the next interval of augmented inspections have been added to this attachment as part of Revision 3 of this PBD.
It should be noted that Ginna station was among the first plants to submit an aging management program on reactor internals for regulatory review. As such, this PBD demonstrates, that the previously approved approach documented in the NRC letter of Feb. 10, 2001 (Ref. 9.66) is met through comparison of the Ginna program to the required GALL Report attributes while capturing the intent of the additional industry program (MRP-227 Revision 0) discussed above. As discussed in subsequent sections, as required by license renewal commitments, Ginna implemented to the requirements of MRP-227 Revision 0 during the 2011 RFO, as the NRC approved version had not yet been issued. Ginna has revised this PBD to adopt MRP-227-A for future examinations.
Existing Program aspects of this AMP such as the ASME Section XI program and primary Page 6 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 system chemistry monitoring have been and continue to be on-going programs at Ginna Station. These programs are supplemented as needed to form the basis for the Ginna Station Reactor Vessel internals AMP.
Augmented inspections for Reactor Vessel Internals aging degradation mechanisms were scheduled to coincide with the Fourth Interval ASME Section XI, 10 year In-service Inspections (ISI) program requirements during the 2011 RFO1. These augmented inspections incorporated MRP-227 Revision 0. The next implementation of these augmented inspections, which will include MRP-227-A, or a later approved version, are being scheduled to coincide with the Fifth Interval ASME Section XI, 10 year ISI program requirements.
An additional study was also performed and documented in WCAP 16881-P, "R.E. Ginna Reactor Vessel Internals Program Scheduling Evaluations" (Ref. 9.68), for assessing the initial period of extended operation prior to conducting the inspections in the 2011 RFO. This study provides a technical basis focusing on the components determined to be most limiting by the MRP-227 Revision 0 work. The report concluded that there would be no adverse conditions for the reactor internals during the initial years of extended operation. The study looked at expected aging degradation mechanisms taking into account the extended power uprate, neutron fluence and associated molecular degradation, and cumulative fatigue usage factors.
Although the study was performed for evaluating the initial extension of most of the internals inspections to the 2011 RFO, much of the work serves to validate the prior aging management AMR reviews and provide insight to the acceptability for continued operation beyond the 2011 RFO inspections and through the full 20 year extended license period.
2.4 Ginna Station Aging Management Program Intent The reactor internals AMP utilizes a combination of prevention, mitigation and condition monitoring. Where applicable, credit is taken for existing programs such as inspections prescribed by the ASME Section XI In-Service Inspection Program (Ref 9.30 ), other existing plant programs such as water chemistry (Refs. 9.10 9.11, 9.12), thimble tube inspections (Ref.
9.28) and past and future mitigation projects such as baffle bolt inspections and split pin replacement ( Refs. 9.5, 9.32, 9.73, 9.74, 9.75, and Section 4.2 of this document) combined with augmented inspections or evaluations as recommended by MRP-227.
1Ginna's 4 th Interval ended in 2009. As such, the 10 year ISI inspection was to occur during the 2009 RFO. The submittal of Relief Requests 18, 19, 20, and 21 deferred the 10 year ISI inspections to the 2011 RFO (Ref. 9.72).
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 Aging degradation mechanisms that impact internals have been identified in Ginna Aging Management review LRAM-RVI. The over-all outcome of both the LRAM-RVI and the additional work performed by the Industry in MRP-227 is to provide appropriate aging management of reactor internals components and ensure early detection of degradation mechanisms identified below. This AMP is consistent with both the WCAP 14577-1-A methodology and the additional industry work provided by MRP -227. Both documents address the concern for the following aging mechanisms:
- Irradiation --Assisted Stress Corrosion Cracking (IASCC)
- Wear
- Fatigue
- Thermal Aging Embrittlement
- Irradiation Embrittlement
- Void Swelling and Irradiation Growth
- Thermal and Irradiation-Enhanced Stress Relaxation or Irradiation Enhanced Creep Section 5, provides an evaluation of the ten elements from Section XI.M. 16 of the GALL Report (Ref. 9.1) and incorporates all available programs and activities that are credited for managing the aging effects produced by the mechanisms listed above. All components of the Reactor Internals have been considered in the review. Tables indicating those aging effects requiring aging management based on a specific assembly, component, or sub-component of the reactor internals are included in the summary table in Attachment C as taken from the Ginna AMR, LRAM-RVI.
For those components listed in the Program/Activities column of Attachment C, where credit is taken for "Reactor Vessel Internals Inspections Program," the industry sponsored MRP-227 Reactor Internals Inspection Guidelines (Ref 9.38 and 9.71) tables provide direction on the inspection technique and frequency. The tables of MRP-227 Revision 0 are included as Attachment D. The MRP-227 Revision 0 was used as the basis for the 2011 RFO inspection to ensure that the Girma AMP was aligned with the anticipated NRC approved version of the Revision 0 inspections and frequencies of inspections for those components that credit this PROGPLAN and the anticipated industry work in the original Ginna LRAM-RVI. A brief justification is provided for inspection applicability to the Ginna LRAM-RVI required augmented inspections presented in section 5.3.3 where appropriate. The MRP-227 recommendations are based on additional Industry evaluations documented in references 9.40 through 9.43. The tables of MRP-227-A are included as Attachment E.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 2.5 Ginna Reactor Internals Background The Ginna reactor pressure vessel is a 2 Loop Westinghouse Design with internals similar to those shown in Attachment G (Figure 1). The internals, consisting of the upper and lower core support structure, are designed to support, align, and guide the core components, direct the coolant flow to and from the core components, and to support and guide the in-core instrumentation.
The components of the Ginna reactor internals per the Ginna Station UFSAR are divided into three parts consisting of the lower core support structure (including the entire core barrel and thermal shield), the upper core support structure, and the in-core instrumentation support structure.
It should be noted that industry work is typically divided into lower and upper internals only, where in the third Ginna component instrumentation such as the thimble tubes, is typically considered with the lower internals.
Additional discussions on the Ginna Reactor Internals structures is available in the Ginna Station UFSAR (Ref. 9.67) Section 3.9.5.
3.0 PROGRAM OWNER The Nuclear Engineering Services Group is responsible for maintaining and implementing the Reactor Vessel Internals AMP.
4.0 DESCRIPTION
OF THE GINNA AMP and INDUSTRY PROGRAMS The Ginna Reactor Vessel Internals AMP is based on meeting the requirements of the ten elements of an aging management program as described by NUREG 1801, GALL Report section XI.M16 for PWR Vessel Internals. In the Ginna Station AMP, this is demonstrated through application of AMR methodology that credits inspections prescribed by the ASME Code Section XI In-Service Inspection Program, existing Ginna Station programs, and additional augmented inspections based on MRP-227 recommendations.
.4.1 Existing Ginna Station Programs Ginna Station has a number of existing programs that support aging management of reactor internals. Ginna Station has also performed modifications in response to Industry identified concerns that are described below.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 4.1.1 ASME Code Section XI In-service Inspection (ISI) Requirements The ASME Code Section XI ISI program is an existing program required by 10CFR50.55a.
The Fourth 10-Year Interval utilized the 1995 Edition with the 1996 Addenda (Ref. 9.27). The Fifth 10-Year Interval utilizes the 2004 Edition with no Addenda (Ref. 9.76). The accessible areas of the reactor vessel interior, including internals, are examined using nondestructive examination visual techniques. At Ginna Station, this examination is performed under the direction of NDE procedure EP-VT-1 10, "Visual Examination of the Reactor Vessel and Removable Internal Structures" (Ref. 9.9).
ASME Section XI, IWB-2500, Examination Category B-N-1 applies to the reactor vessel interior. ASME Section XI, IWB-2500, Examination Category B-N-2 applies to interior attachments beyond the beltline region. ASME Section XI, IWB-2500, Examination Category B-N-3, applies to the reactor vessel internals core support structures. Examination of these interior attachments in the reactor vessel (Category B-N-2) and the reactor vessel internals (Category B-N-3) is required once every 10-year ISI interval in accordance with the Ginna Station Inservice Inspection Plan (Ref. 9.30). The examination is performed with the aid of remote visual examination tools in accordance with the EP-VT-1 10 procedure.
Relevant conditions for these examination categories are found in ASME Section XI, IWB-3520, Standards for Examination Category B-N-1, Interior of Reactor vessel Examination Category B-N-2, Welded Core Support Structures and Interior Attachments to Reactor Vessels, and examination Category B-N-3, removable core support structures.
The relevant conditions described include but are not limited to cracks, structural distortion, loose parts, missing parts, and foreign material inside the vessel. The VT-3 examination, combined with other nondestructive examination techniques, can be relied upon to detect wear damage on accessible surfaces and any other normal degradation found inside the reactor vessel. The Ginna Station ASME Code Section XI, 4 th 10-Year Interval ISI examinations were completed during the 2011 RFO.
4.1.2 Primary Water Chemistry Program The Ginna Station Water Chemistry Control Program, (Ref. 9.10), as implemented by plant chemistry procedures CH-101, "Strategic Primary Water Chemistry Plan" (Ref. 9. 11) and CH-120, "Primary System Analysis Schedule and Limits" (Ref. 9.12) can also be credited with mitigating IASCC. The Water Chemistry Control Program limits the concentration of oxygen, halogens, and sulfate species in the primary water, and therefore effectively prevents SCC and greatly reduces the probability of IASCC. The limits imposed by the chemistry monitoring program meet the intent of the EPRI "Pressurized Water Reactor Primary Water Chemistry Guidelines" (Ref. 9.13).
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 4.1.3 LR-TTI-PROGPLAN (Thimble Tubes)
Flux 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, then through the lower internals assembly, and finally extend to the top of the fuel assembly. The bottom mounted instruments (BMI) column 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 flux thimble 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 Ginna thimble tube inspection program is an existing plant specific program that follows Branch Technical position RLSB-1, Generic Aging management review which is included in Appendix A of NUREG-1800 (Ref 9.52). The Ginna Thimble Tube inspection program manages cracking due to SCC of BMI guide tubes and cracking of BMI guide tube fillet welds.
This program requires periodic visual inspection of fillet welds. The program also includes eddy current testing requirements for thimble tubes and includes criteria fbr determining sample size, inspection frequency flaw evaluation, and corrective action in accordance with US NRC Bulletin 88-09 (Ref. 9.53), and NUREG 1801 (Ref 9.1). Specific Ginna actions are contained in the LR-TTI-PROGPLAN (Ref. 9.28).
During the 2011 RFO, taking into account plant and industry OE, as well as the information in NRC Bulletin 88-09, Ginna replaced all 36 flux thimbles with tubes that are chrome plated in the area from 12" above the core plate to 12" below the reactor vessel bottom head. The chrome plating of this portion of the thimble tube has been shown to reduce the wear rate by at least a factor of two or three, which significantly increases the life of the flux thimbles. The related 36 seal table fittings were also replaced with a newer type of high pressure fitting. The new type of fitting prevents leakage and provides for easier maintenance. This replacement was performed through ECP-2009-0123 (Ref. 9.98). Applicable OE which drove Ginna to select the chrome plated flux thimbles is referenced in this Engineering Change Package (ECP).
4.2 Other Ginna Station Projects 4.2.1 Control Rod Guide Tube Split Pin Replacement Project Westinghouse Analysis indicated that failures of original Alloy X-750 Control Rod Guide Tube (CRGT) Split Pins was caused by SCC. The failed pins had been solution heat treated at less than 1800 'F, after which they were age-hardened, and then stressed during subsequent manufacture and installation processes. In general, SCC prevention is aided by adherence to strict primary water chemistry limits that effectively prevent SCC and greatly reduce the probability of IASCC. The limits imposed by the chemistry monitoring program are consistent with TR-10 14986 "EPRI PWR Primary Water Chemistry Guidelines" as described in the previous section.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 In response to the industry concern, all of the CRGT OEM split pins were preemptively replaced at Ginna Station during the 1986 RFO.
Replacement CRGT split pins for Ginna were supplied by Framatome. The replacement pins were supplied as equivalent replacements and manufactured from X-750 material that included special design considerations for installation torques, improved radius of curvature and heat treatments, which were intended to prolong the life of the split pin. Specific pin enhancements are discussed in the Ginna LRAM-RVI.
More recent inspections (2007/2008) have been performed at European facilities with similar split pins, known as Generation 4 split pins. Discussions with individuals familiar with the European experience resulted in knowledge of suspected signal anomalies in 5 Generation 4 split pins. Discussions indicate that additional review of the signal anomalies has resulted in a preliminary conclusion that SCC may have initiated in 1 of the 5 split pins investigated.
Based on the preliminary information available from the 2007/2008 experience, the split pin issue was entered into the Ginna Station Corrective Action Process. A contingency plan for split pin failure had been initiated for the Ginna Station CRGT split pins (Attachment A, Table 1, NL-2009-000023; CA-2008-003832; Ref. 9.69). The contingency included a specific Metal Impact Monitoring System (MIMS) signal characteristic for control room recognition based on the potential failed split pin piece, size and weight, followed by prompt plant shut down in the case where the anticipated MIMS signal characteristic is received in the control room.
The decision was made to replace all of Ginna's 66 split pins during the 2011 RFO. The replacement split pin was Westinghouse designed and made of a Cold Worked, Type 316 Stainless Steel material, Grade B8M (Class 2 strain-hardened) from ASME SA-193 specified in ASME Code Case N-60-5. The replacement Westinghouse split pin was analyzed for the long term affects of Irradiated Assisted Stress Corrosion Cracking, Primary Water Stress Corrosion Cracking, Irradiation Stress Relaxation, Irradiation Swelling and Desensification, as well as embrittlement and Toughness for the conditions of the Ginna reactor. Each of these properties was determined to maintain its integrity over a forty year period at a 100% capacity factor (Ref. 9.75 and 9.90). Ginna intends to verify the integrity of the replaced split pins after twenty years of service life (Ginna Action Item NL-2012-000094-005).
4.2.2 Baffle-Former Bolt Inspection/Replacement Project Baffle-former bolts are potentially subject to IASCC, irradiation embrittlement, stress relaxation, and void swelling. Because cracking has typically occurred at the juncture of the bolt head and shank, and since this area is inaccessible for visual examination, VT-3 examination alone is not an effective method for managing aging effects that affect baffle-former bolts. VT-3 examination would detect missing bolt heads. Ultrasonic testing (UT) can be used to determine if cracks exist in bolts. Alternatively, selected bolts can be preemptively replaced with bolts fabricated from a more crack-resistant material, thus precluding or postponing future examinations. Because there is significant mechanical redundancy in the Page 12 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 number of baffle-former bolts used in the baffle-former assembly, only a fraction of the total number of bolts would require replacement or an alternate pattern determined to guarantee integrity of the baffle-former joint until the end of extended plant life.
During the 1999 RYO, a preemptive baffle-former bolt replacement was completed at Ginna Station as part of a WOG project known as "B" cubed (Refs. 9.32 & 9.33 ). Ginna Station Action Report (Ref. 9.6) summarizes that all accessible baffle-former bolts were inspected by UT and that a total of 56 replacement bolts were installed during the 1999 refueling outage.
The replaced bolts were reported to have recordable defect-like indications. The bolts of concern were also part of a pre-qualified minimum bolt pattern for two-loop nuclear plants that was generated by the Westinghouse Owners Group and documented in WCAP-15036 (Ref.
9.7).
A sample of intact bolts with indications (as well as two bolt head segments) was sent to the Westinghouse hot cell for metallurgical analysis (Ref. 9.25). Destructive analysis of the bolts with flaw-like UT indications revealed no evidence of head/shank cracking. Fractographic analysis of one bolt head revealed fracture features typical of intergranular cracking which would be expected of IASCC. TEM studies showed evidence of voids near the threaded end of one bolt, but not at the head end. The void volume (0.004% maximum observed in the 347 SS bolt material) was small and preliminary extrapolation to the end of extended life using a simple square law suggested that void swelling should not be a concern. Detailed results of the hot cell tests and analyses are summarized in the AMR report for the reactor vessel internals and in Reference 9.25.
The more recent work provided by MRP-227 calls for examination of baffle-former bolts based on additional evaluations and determinations that the 1999 inspections may have been performed early in plant life prior to the bolts reaching their maximum expected degradation peaks. MRP-227 Revision 0 specified the baffle-former bolts as a Primary component. As referenced in Attachment D, Table 1.2, of this PBD, the examination coverage is 100% of accessible bolts or as supported by plant-specific justification.
Prior to the 2011 RFO, Ginna Engineering was not confident in the techniques available to volumetrically examine Ginna's baffle-former bolts. This was due to the fact that UT results were subject to interpretation When it was previously performed in 1999, because of the internal hex head design of Ginna's original baffle-former bolts. Industry Operating Experience and site benchmarking (Ref. 9.77 and 9.78) also supported this conclusion. As a result, Ginna made the decision to replace a specified number of bolts as laid out in WCAP-17354 (Ref. 9.79). Because replacement was not specifically stated in MRP-227 Revision 0 as a means to ensure the integrity of the baffle-former Bolt's, NL-2010-000049-001 was generated to document a deviation in accordance with Ginna interface procedure IP-IIT-8, which implements the reactor coolant system materials degradation management program as part of the Nuclear Energy Institute's (NEI) 03-08 program. The document was provided to the Page 13 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 MRP-227 Industry Protocol Group and determined by the group to be a plant specific justification, as described in Table 4-3 of MRP-227 Revision 0, instead of a deviation.
The intent of the justification was to replace a specified pattern of 182 original baffle-former bolts, per WCAP- 17354, with new baffle-former bolts made of a more resistant material to Irradiated Assisted Stress Corrosion Cracking (IASCC). As a result of significant difficulties encountered in the: field during the replacement campaign, which included a maximum achieved production rate of less than 5 baffle-former bolts per day and required 3 locations to be left empty in the baffle-former structure (Ref 9.74 and 9.80), Ginna modified the plant specific justification to meet the MRP-227 Revision 0 requirements (Ref. 9.73). In addition to volumetric examination of the 56 replacement baffle-former bolts that were installed in 1999, this modified justification consisted of replacement of the 25 original baffle-former bolts, examination of 24 of these removed bolts via UT examinations from the threaded end 2, and inspection of 99 additional original baffle-former Bolts via UT examinations at the socket end.
The vendor was able to make volumetric examination of the baffle-former bolts possible, by redesigning the UT probe that was originally used during the 1999 RFO. During the replacement campaign, the vendor took measurements of the Ginna internal hex head designed baffle-former bolt. It'was realized that the measurements taken in the field were not as depicted on plant drawings from the time of original manufacturing of the bolt. It was discovered that the depth of the internal hex head was in fact deeper than originally documented. With this knowledge, the vendor moved the transducer further down the probe and was able to obtain the proper signal (Ref. 9.81).
ECP-12-000028 was created in order to document the updated justification for meeting the MRP-227 Revision 0 requirements, with respect to the inspection of the baffle-former bolts (Ref. 9.73). This plant specific justification provides assurance of the structural and hydraulic stability of the baffle-former assembly until the next baffle-former bolt inspection. The results provide confirmation that examination of the components within the expansion criteria, which include the lower support column bolts and the barrel-former bolts, was not necessary.
The Replacement of 25 of the original baffle-former bolts with new bolts made of a more resistant material that is less susceptible to Irradiated Assisted Stress Corrosion Cracking (IASCC) ensures that these bolts will maintain their design function. Volumetric Examination of the threaded end of 24 of the original baffle-former bolts that were removed provides evidence that the original bolt material has not been affected by IASCC. Volumetric Examination of the 56 replacement baffle-former bolts installed in 1999 verifies that the aging effects of the 11 year old bolts are being managed. Volumetric examination of 99 original baffle-former bolts, with 98 of the 99 bolts having satisfactory results, also provides assurance that the original bolt material has not been affected by IASCC. In addition to these volumetric 2 28 Baffle Former Bolts were removed. Due to difficulty of removal, 4 of these bolts had to be removed with contingency EDM burns. These contingent removals damaged the bolt to the point where UT from the threaded end was not possible.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 examinations, as part of the ten-year In-Service Inspection exams, a visual inspection of all baffle-edge bolts, baffle-former bolts, and seams of the baffle plates was performed. All results were satisfactory, with no anomalies found. All edge-bolts and baffle-former bolts, along with the tack welds of their respective lock washers, were intact and no issues were observed on the seams of the baffle plates.
Of the sample set of baffle-former bolts examined, which was of statistical significance and representative of the entire population (Ref. 9.91), there was only one bolt which had flaw-like indications, located at position 4-5-1-6 (Quadrant-Plate-Column-Former). This is a less than 1% failure rate in the sample set of 179 bolts inspected and a less than 1% failure rate in the sample set of original baffle-former bolts in the age range of 25 to 35 Effective Full Power Years3.
The specific combination of replacement and inspection has been analyzed for structural acceptability through the current interval, until the next MRP-227 required inspection date (Ref 9.91 and 9.92). The replacement baffle-former bolts have also been evaluated to meet the replacement-specifications stated by the OEM (Ref. 9.93). The techniques to volumetrically examine both the original bolts that have been removed (Ref. 9.88) and the original bolts that were examined in place (Ref. 9.81) have been demonstrated and technically justified. Further, the impact of having three locations in the baffle plates with no bolts installed has been analyzed by two vendors, each using a different methodology, and determined to not be significant to the safety of the plant (Ref. 9.94, 9.95, 9.96, 9.97, and 9.99).
MRP-227-A also specifies the baffle-former bolts as a "Primary" component. As referenced in Attachment E, Table 1.1 of this PBD, the examination coverage is 100% of accessible bolts, via a volumetric (UT) examination. If the volumetric examination of the baffle-former bolts is not satisfactory at the time that this exam is implemented at Ginna, then 100% of the accessible barrel-former bolts and a 100% of the accessible lower support column bolts will be inspected volumetrically, per the expansion link. A plant specific justification is also allowed in lieu of performing a UT examination on 100% of the accessible lower support column bolts.
4.3 Industry Programs 4.3.1 WCAP-14577-1-A As described previously in Section 2, the WOG topical report WCAP-14577-1-A, (Ref. 9.2) 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 3 MRP-227 Revision 0 specified an Examination Method/Frequency of a baseline volumetric examination between 25 and 35 EFPY.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 effects of aging on Reactor Vessel Internals 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 Ginna Station internals, documented in LRAM-RVI, including required applicant action items, was completed in accordance with the requirements of the WCAP-14577-1-A (Ref. 9.2). Attachment C tables present the results of the Ginna Station evaluation and the resulting programs, as well as additional inspections that are required for those components determined to be most susceptible to degradation mechanisms that are credited for license renewal. Those components listed in the Attachment C "Program/Activity" column that credit the "Reactor Vessel Internals Inspection Program," are addressed by the additional evaluations and analysis as described in the next section.
4.3.2 MRP-227 The MRP-227 Reactor Internals Inspections and Evaluation Guidelines were developed by a team of industry and NSSS vendors and International committee representatives who reviewed available data and industry experience on materials aging. The objective of the group was to develop a systematic approach for identifying and prioritizing inspection and evaluation requirements for reactor internals.
The key sequential steps in the process included the following:
The development of screening criteria, with susceptibility levels for the eight postulated aging mechanisms relevant to reactor internals and their effects; An initial component screening and categorization,. using the susceptibility levels and FMECA (failure modes, effects, and criticality assessment) to identify the relative ranking of the components;
- Functionality assessment of degradation for components and assemblies of components; Aging management strategy development combining the results of functionality assessment with component accessibility, operating experience, existing evaluations, and prior examination results to determine the appropriate aging management methodology, baseline examination timing, and the need for and the timing of subsequent inspections.
Through this process, the reactor internals for all three NSSS vendors 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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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 The MRP-227 utilized the screening and ranking process to aid in the identification of required inspections for Primary and Expansion Components and credits Exiting component inspections when they were deemed adequate.
The basic description of each classification is 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 these 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 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 at individual plants.
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 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 FMECA and the functionality assessment. No further action is required by these guidelines for managing the aging of the No Additional Measures components.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 The categorization and analysis processes used by the MRP-227 approach is not intended to supersede any ASME B&PV Code Section XI (Ref .9.27 and 9.76) requirements. Any components that are classified as core support structures, as defined in ASME B&PV Code Section XI IWB 2500, Category B-N-3 have requirements that remain in effect and may only be altered as allowed by 10 CFR 50.55a.
4.3.2.1 MRP-227 Revision 0 Requirements Revision 0 of MRP-227 was issued to the industry in December 2008. The Industry program outputs of MRP-227 Revision 0 are classified in accordance with the requirements of the NEI-03-08 Guidelines (Ref. 9.54) and have been applied at Ginna Station through application of CNG-AM-4.01 (Ref 9.51) and IP-IIT-8 (Ref 9.47). For the MRP-227 Revision 0 guideline there are Mandatory, Needed and Good Practice elements as follows:
There is one Mandatory Element:
Mandatory: Each commercial U.S. PWR unit shall develop and document a PWR reactor internalsaging managementprogram (AMP) within thirty-six months following issuance of MRP-227-Rev. 0.
Appendix A describes each of the attributes that comprise an acceptable AMP.
MRP-227-Rev. 0 is the first published version of these guidelines.
There are three Needed elements:
Needed: Each commercial U.S. PWR unit shall implement Tables 4-1 through 4-9 and Tables 5-1 through 5-3for the applicable design within twenty-four monthsfollowing issuance of MRP-22 7-A.
Implementation of these guidelines is to take effect 24 months following issuance of MRP-227-A. MRP-227-A is the version that will have incorporated the changes proposed by the MRP in response to U.S. Nuclear Regulatory Commission (NRC) Requests for Additional Information, recommendations in the NRC Safety Evaluation and other necessary revisions identified since the previous publication of the report.
Earlier implementation may be required by plant-specific regulatory comnmitments (for example, license renewal approvals). Plants implementing these guidelines prior to the issuance of the "NRC-approved" version would thus implement the requirements in accordance with the current published version of these guidelines.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 Needed: Examinationsspecified in these guidelinesshall be conducted in accordancewith the Inspection StandardMRP-228.
Needed: Examination results that do not meet the examination acceptance criteriadefined in Section 5 of these guidelinesshall be recordedand entered in the plant corrective action programand dispositioned.
There is one Good. Practice element:
Good Practice: Each commercial US. PWR unit shouldprovide a summary report of all inspections and monitoring, items requiringevaluation, and new repairsto the MRP Program Managerwithin 120 days of the completion of an outage during which PWR internals are examined The MRP template should be usedfor the report.
This summary of the results will be compiled into an overall industry report which will track industry progress, aid in evaluation of significant issues, identification of fleet trends and determination of any needed revisions to these guidelines. The industry report will be updated biennially for the benefit of the fleet, the regulator, the PWROG and other industry stakeholders. This biennial report will serve to assist in review of operating experience, and required monitoring and trending for aging management programs established by the industry.
In order to ensure completeness and consistency of reporting, the MRP will provide a template listing the requested information.
4.3.2.2 MRP-227-A Requirements MRP-227-A was issued to the industry in December 2011. This revision incorporates the changes proposed by the NRC in the Safety Evaluation as well as those proposed by the EPRI Materials Reliability Program in response to NRC Requests for Information (RAIs). The Industry program outputs of MRP-227-A are classified in accordance with the requirements of the NEI-03-08 Guidelines (Ref. 9.84) and will be applied at Ginna Station through application of CNG-AM-4.01 (Ref 9.82) and IP-IIT-8 (Ref 9.83). For the MRP-227-A guideline there are Mandatory, Needed and Good Practice elements as follows:
There is one Mandatory Element:
Mandatory: Each commercial U.S. P WR unit shall develop anddocument a programfor management of aging of reactor internalcomponents within thirty-six months following issuance of MRP-22 7-Rev. 0 (that is, no later than December 31, 2011).
MRP-227-Rev. 0 is the first published version of these guidelines.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 There are five "Needed" Elements:
Needed: Each commercial US. PWR unit shall implement Tables 4-1 through 4-9 and Tables 5-1 through 5-3for the applicable design within twenty-four monthsfollowing issuance of MRP-22 7-A.
Implementation of these guidelines is to take effect 24 months following issuance of MRP-227-A (that is, no later than December 31, 2013). Implementation means performance of Inspections of applicable components within the time frame specified in the guidance provided in the applicable tables. MRP-227-A is the current version that has incorporated the changes proposed by the MRP in response to U.S. Nuclear Regulatory Commission (NRC) Requests for Additional Information, recommendations in the NRC Safety Evaluation and other necessary Revisions identified since the previous publication of the report (MRP-227 Rev. 0).
Earlier implementation may be required by plant-specific regulatory commitments (for example, license renewal approvals). Plants implementing these guidelines prior to the issuance of the "NRC-approved" version would thus implement the requirements in accordance with the current published version of these guidelines.
Consistent with the requirements of NEI 03-08, if the guidance contained in Table 4-1 through 4-9 and/or Tables 5-1 through 5-3 cannot, need not, or will not be implemented as written, a technical justification must be prepared that clearly states what requirement cannot, need not, or will not be met and why; what alternative action is being taken to satisfy the objective or intent of the guidance; and, why the alternative action is acceptable. Examples of alternatives that may be justifiable are: elevation of an Expansion component to Primary; substitution of an equivalent or more rigorous examination than is required by the tables; or destructive testing in lieu of nondestructive examination, such as the case where one or more of the primary components is being replaced. Since the Expansion components are also "needed" requirements, the technical justification for not fully implementing a Primary component examination or not implementing it in a manner consistent with its intent, would be expected to include disposition of the associated Expansion components.
When submittal of a deviation from work products or elements is required, the justification shall be reviewed and approved in accordance with the applicable plant procedures with the additional responsibility for deviation from a 'Needed' element that an internal independent review is performed and that concurrence is "obtained from the responsible utility executive.
Further, as stipulated in the Implementation Protocol (Appendix B) of NEI 03-08, a utility is required to notify the Issue Program (e.g., the MRP) and the NRC.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 Needed: Examinationsspecified in these guidelinesshall be conducted in accordancewith the Inspection StandardMRP-228.
Needed: Examination results that do not meet the examination acceptance criteriadefined in Section 5 of these guidelines shall be recordedand entered in the plant corrective action programand dispositioned.
Needed: Each commercial US. PWR unit shall provide a summary reportof all inspections and monitoring, items requiringevaluation, and new repairsto the MRP ProgramManager within 120 days of the completion of an outage during which PWR internals within the scope of MRP-227 are examined This summary of the results will be compiled into an overall industry report which will track industry progress, aid in evaluation of significant issues, identification of fleet trends and determination of any needed revisions to these guidelines. The industry report will be updated biennially for the benefit of the fleet, the regulator, the PWROG and other industry stakeholders. This biennial report will serve to assist in review of operating experience, and required monitoring and trending for aging management programs established by the industry.
In order to ensure completeness and consistency of reporting, the MRP will provide a template listing the requested information.
Needed: If an engineeringevaluation is used to dispositionan examination result that does not meet the examination acceptance criteriain Section 5, this engineeringevaluation shall be conducted in accordancewith a NRC-approvedevaluation methodology.
4.3.2.3 MRP-227 Implementation at Ginna Station MRP-227 Revision 0 was issued in December 2008. The one mandatory requirement was that each commercial U.S. PWR unit shall develop and document a PWR reactor internals aging management program within thirty-six months following its issuance. Gi:rma fulfilled this requirement with the issuance of Revision 2 of this PBD in February 2009. This was subsequently sent to the NRC. Upon the issuance of MRP-227-A, which was December 2011 (six months following Ginna's implementation of the inspections specified in MRP-227 Revision 0), Ginna withdrew its submittal of this PBD from the NRC. Per NRC Regulatory Issue Summary 2011-07 (Ref. 9.85), this was not necessary for Ginna, as it was classified under Category A. Plants classified as Category A are plants with renewed licenses that have already submitted an AMP/inspection plan based on MRP-227 Revision 0 to comply with an existing commitment. The NRC Expectation for these plants are that licensees may withdraw their current AMP/inspection plan submittals and provide new/revised commitments to resubmit AMPs/inspection plans in accordance with MRP-227-A no later than October 1, Page 21 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 2012. Future submittals should address all information required by MRP-227-A. Ginna withdrew their PBD from the NRC with the intent of including the results from the implementation of MRP-227 Revision 0 as well as committing to the implementation of MRP-227-A in future inspections and provide a justification for each deviation from the MRP-227-A guidelines; then re-submitting it to the NRC.
The first needed requirement of MRP-227 Revision 0 was that each commercial U.S. PWR unit shall implement Tables 4-1 through 4-9 and Tables 5-1 through 5-3 for the applicable design within twenty-four months following issuance of MRP-227-A. The supporting details of this requirement state that, "earlier implementation may be required by plant-specific regulatory commitments and that plants implementing these guidelines prior to the issuance of the "NRC-approved" version would implement the requirements in accordance with the current published version [i.e. MRP-227 Revision 0] of these guidelines." Due to License Renewal Commitment # 31, Ginna committed to Revision 0 of MRP-227 and implemented its inspection requirements during the Spring 2011 RFO. It therefore used the aforementioned supporting detail and did not have to wait until the issuance of the -A revision to implement the guidelines. The MRP-227 Revision 0 inspections were completed during the 2011 Spring RFO in order to coincide with the 10 Year ISI Program Inspections. The applicable Westinghouse tables contained in MRP-227 Revision 0 that were implemented are Table 4-3 (Primary), Table 4-6 (Expansion), Table 4-9 (Existing), and Table 5-3 (Acceptance and Expansion Criteria).
These tables are attached to this PBD, for reference, in the Attachment D Tables 1.1, 1.2, 1.3, and 1.4 respectively. Upon the issuance of MRP-227-A, in December 2011, several changes were made to the applicable Westinghouse Tables 4-3 (Primary), 4-6 (Expansion), 4-9 (Existing), and 5-3 (Acceptance and Expansion Criteria) from Revision 0. These tables are attached to this PBD, for reference, in Attachment E Tables 1.1, 1.2, 1.3, and 1.4 respectively.
The supporting detail which stated, "Earlier implementation may be required by plant-specific regulatory commitments (for example, license renewal approvals). Plants implementing these guidelines prior to the issuance of the "NRC-approved" version would thus implement the requirements in accordance with the current published version of these guidelines," in MRP-227 Revision 0 was also stated in MRP-227-A, for the respective needed requirement.
However, looking forward to the next MRP-227 inspections which Ginna must perform, Ginna commits to implementing the guidelines of MRP-227-A. If an inspection cannot, need not, or will not be implemented as written, a technical justification will be prepared in accordance with the applicable guidelines. Action items have been created in order to track the implementation ofIMRP-227-A and are included in Attachment A of this document.
The second needed requirement of MRP-227 Revision 0 is that examinations specified in these guidelines shall be conducted in accordance with the Inspection Standard MRP-228. This same needed requirement exists in MRP-227-A. Ginna has used this Inspection Standard in its implementation of MRP-227 Revision 0 during the 2011 RFO where required by outputs. This Inspection Standard will also be used during implementation of MRP-227-A.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 The third needed requirement of MRP-227 Revision 0 is that examination results that do not meet the examination acceptance criteria defined in Section 5 of these guidelines shall be recorded and entered in the plant Corrective Action Program and dispositioned. As described in Section 5.3.3 of this PBD, all inspection results were found to be satisfactory. Therefore, Ginna did not need to apply CNG-CA-1.01-1000 (Ref 9.18). This requirement exists in MRP-227-A. Ginna's Corrective Action Program will be used, if needed, during implementation of MRP-227-A.
The one good practice requirement of MRP-227 Revision 0 stated that each commercial U.S.
PWR unit should provide a summary report of all inspections and monitoring, items requiring evaluation, and new repairs to the MRP Program Manager within 120 days of the completion of an outage during which PWR internals are examined. The MRP template should be used for the report. Ginna has fulfilled this requirement following the 2011 RFO. This good practice requirement was changed to a needed requirement in MRP-227-A. Ginna will fulfill this requirement upon completion of its implementation of MRP-227-A.
MRP-227-A had one additional needed requirement that MRP-227 Revision 0 did not. As stated, "if an engineering evaluation is used to disposition an examination result that does not meet the examination acceptance criteria in Section 5, this engineering evaluation shall be conducted in accordance with a NRC-approved evaluation methodology." This needed requirement will be followed during implementation of MRP-227-A.
It should be noted that Appendix A of MRP-227 Revision 0 also includes a description of the attributes that comprise an acceptable AMP. These attributes are similar to the previously discussed attributes of the GALL Report and used for the Aging Management Review for Ginna station, LRAM-RVI as discussed previously in Sections 2.3 and 4.0. Evaluation of the Ginna License Renewal Reactor Vessel Internals Program Basis Document (LR-RVI-PROGPLAN) against GALL attribute elements is provided in Section 5 of this PBD.
As part of the Ginra License Renewal Project, Ginna Station agreed to participate in Industry activities associated with the development of the standard Industry Guideline for Inspection and Evaluation of Reactor Internals described above. These efforts have defined the required inspections and examination techniques for those components that credit this LR-RVI-PROGPLAN and the additional industry work. Those components correspond to the ones that are listed in the last column of the Attachment C tables that credit the "Reactor Internal Inspections Program" as being the basis for the recommended inspections. The results of the industry recommended inspections, published in the MRP-227 Revision 0 guidelines, serve as the basis for identifying any augmented inspections that are required to complete this PROGPLAN.
At the time that the MRP-227 Revision 0 guideline was submitted to the NRC to produce a Page 23 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 Safety Evaluation Review (SER) review and approval, discussions with the NRC indicated that the Ginna program could not be based solely on the MRP-227 Revision 0 work. Therefore, as discussed in previous section of this LR-RVI-PROGPLAN, the Ginna program has been rooted in meeting the GALL Report attributes and the previous approved work of the PWROG WCAP 14577-1-A. However, this LR-RVI-PROGPLAN also captures the results of additional evaluations, inspection recommendations, and the SER of MRP-227, as summarized in the tables of Attachments D and E.
4.3.2.4 MRP-227 Applicability to Ginna Station The applicability of MRP-227 to Ginna Station requires compliance with the following analysis assumptions:
30 years or less of operation with high leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel management strategy for the remaining 30 years of operation.
Ginna Station fuel management program changed from a high to a low leakage core loading pattern prior to 30 years of operation.
0 Base load operation, i.e., typically operates at fixed power levels and does not usually vary power on a calendar or load demand schedule.
Ginna operates as a base load unit.
- No design changes beyond those identified in general industry guidance or recommended by the original vendors.
MRP-227 states that the recommendations are applicable to all U.S. PWR operating plants as of May 2007 for the three designs considered. Ginna Station has not made any modifications to internals components in the time since May 2007, that was beyond general industry guidance or recommended by the original vendor.
Conclusion:
Based on the above review, the MRP-227 work is representative for Ginna Station.
4.3.3 Ongoing Industry Programs Both the PWROG and the EPRI/MRP Industry Issue Program continue to sponsor activities related to Reactor Vessel Internals aging management. CENG and Ginna Station participation in this industry program work will continue.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 4.4 Summary of the Ginna AMP and Industry Programs It should be noted that both the WCAP-14577-1 -A and the MRP-227 approach to aging management are based on the GALL approach to aging management strategies. That is, to determine which reactor internals passive components are most susceptible to the aging mechanisms of concern; and then determine the proper inspection or mitigating program that provides reasonable assurance that the component will continue to perform its intended function through the period of extended operation.
The WCAP based approach was used at Ginna for the initial basis of the license renewal application and subsequent NRC SER. Applying the WCAP approach produced a Ginna specific AMR which produced the AMR summary table provided in Attachment C. The table credits the "Reactor Vessel Internals Inspection Program" for certain components as shown in the last column of the table. The Reactor Vessel Internals Inspection Program refers to this LR-RVI-PROGPLAN and is based on additional Industry MRP requirements that are summarized by the Tables in Attachments D and E.
The additional evaluations and analysis completed by the MRP industry group, provided needed clarification to the level of inspection quality needed to determine the proper examination method and frequencies. The tables provided by the MRP-227 Attachments D and E provide the level of examination required for each of the components evaluated. MRP-227 Revision 0 inspections were completed at Ginna during the 2011 RFO. Ginna made a license renewal commitment to follow this industry guideline. At the time of the 2011 RFO inspection, MRP-227-A had not been issued. Ginna intends to follow the requirements in MRP-227-A for subsequent examinations. Justification for each difference from MRP-227-A, based on having completed the exams to MRP-227 Revision 0, is provided in this PBD.
WCAP-14577 and MRP-227 approaches have yielded tables that credit existing programs or prescribe additional inspections that provide the input basis for revision of the reactor internals inspection implementation procedure EP-VT- 110.
This procedure is used for the In-service Inspection of the reactor internals in accordance with the ASME Section XI Code. Action item NL-2009-000021-001 documents that this procedure has been updated based on the augmented inspection required by this PROGPLAN.
It is the Ginna Station position that use of the AMR produced by the WCAP methodology, combined with the additional augmented inspections required by the Industry tables provided in Attachments D and E, provides reasonable assurance that the reactor internals passive components will continue to perform their intended functions through the period of extended operation.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 5.0 GINNA AGING MANAGEMENT PROGRAM ATTRIBUTE EVALUATIONS The attributes of the Ginna Station Reactor Internals AMP and their compliance with NUREG-1801 (Generic Aging Lessons Learned (GALL) Report),Section XI.M. 16, "PWR Vessel Internals" (Ref. 9.1) are described in this section.
5.1 SRP Element 1 -:Scope of Program Gall Report Element Definition:
The program is focused on managing the effects of crack initiation and growth due to stress corrosion cracking (SCC) or irradiation assisted stress corrosion cracking (IASCC), and loss of fracture toughness due to neutron irradiation embrittlement or void swelling. The program contains preventive measures to mitigate SCC or IASCC; ISI to monitor the effects of cracking on the intended function of the components; and repair and/or replacement as needed to maintain the ability to perform the intended function. Loss of fracture toughness is of consequence only if cracks exist. Cracking is expected to initiate at the surface and is detectable by augmented inspection.
The program provides guidelines to assure safety function integrity of the subject safety related reactor pressure vessel internal components, both non-bolted and bolted components. The program consists of the following elements: (a) identify the most susceptible or limiting items, (b) develop appropriate inspection techniques to permit detection and characterizing of the feature (cracks) of interest and demonstrate the effectiveness of the proposed technique, and (c) implement the inspection during the license renewal term. For example, appropriate inspection techniques may include enhancing visual VT-1 examinations for non-bolted components and demonstrated acceptable inspection methods for bolted components.
5.1.1 Ginna Program Scope The Ginna Station RVIs consist of two basic assemblies:
- Upper internals assembly that is removed during each refueling operation to obtain access to the reactor core. The top of this assembly is clamped to a ledge below the vessel-head mating surface by the reactor vessel head. The core barrel fuel alignment pins of the lower internals assembly guide the bottom of the upper internals assembly.
- Lower internals assembly that can be removed, if desired, following a complete core unload. This assembly is clamped at the same ledge below the vessel-head mating surface and closely guided at the bottom by radial/clevis assemblies.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 Additional Reactor Vessel Internals details are provided in Section 3.9.5 and Section 4.2.1 of the UFSAR.
5.1.2 Subcomponents Subject to an Aging Management Review The subcomponents of the Reactor Vessel Internals that require aging management review are indicated in the previously submitted, Table 2.3.1.3 of the R.E. Ginna Nuclear Power Plant Application for Renewal Operating License, (Ref 9.36). This table is included herein for convenience as Attachment F. The table lists the subcomponents of the Reactor Vessel Internals that require aging management review along with each subcomponent's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3 of the Ginna License Renewal Application.
The Ginna Station Reactor Internals Aging Management Review was conducted and documented in the Ginna Station Aging Management Review Report, LRAM-RVI (Ref. 9.35).
The table, summarizing the results of that review, is reproduced herein as Attachment C, "Ginna Station Aging Management Review Summary Table." The table identifies those components that are most susceptible or limiting for Ginna, based on the initial LRAM-RVI results. The last column in the table lists the Program/Activity that is credited to address the component and aging effect during the period of extended operation.
It should be noted that the program/activity credited in the last column referred to as "Reactor Vessel Internals Inspection Program," is this LR-RVI-PROGPLAN, which is based on additional industry MRP activities discussed in sections 4.3.2 and 4.4 above. The additional studies provided by MRP-227 Revision 0 and MRP-227-A, summarized herein in the tables of Attachments D and E respectively, provide the basis for the required augmented inspections, inspection techniques to permit detection and characterizing of the feature (cracks) of interest, prescribed frequency of inspection, and examination acceptance criteria.
The Ginna program scope is based on previously established and approved GALL Report approaches through application of the WCAP 14577-1-A methodologies to determine those components that require aging management. Likewise, the additional evaluation summary results provided by the Industry MRP-227 results of Attachments D and E are rooted in the GALL methodology and provide a basis for augmented inspections that were required to complete this Gimia Station Program by providing the inspection method, frequency of inspection, and examination acceptance criteria.
5.1.3
Conclusion:
This element complies with the corresponding aging management attribute in NUREG- 1801,Section XI.M16.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 5.2 SRP Element 2 - Preventive Actions GALL Report Element Definition:
The requirements of ASME Section XI, Subsection IWB, provide guidance on detection, but do not provide guidance on methods to mitigate cracking. Maintaining high water purity reduces susceptibility to cracking due to SCC. Reactor coolant water chemistry is monitored and maintained in accordance with the EPRI guidelines in TR-105714. The program description and evaluation and technical basis of monitoring and maintaining reactor water chemistry are presented in Chapter XI.M2, "Water Chemistry."
5.2.1 Ginna Preventative Action The Ginna's reactor internals AMP includes the following existing program that complies with the requirements of this SRP Element. A description and applicability to the Ginna Station reactor internals AMP is provided in the following subsection.
5.2.2 Primary Water Chemistry Control Program The Ginna Primary Water Chemistry Control Program (Ref 9.10), as implemented by plant chemistry procedures CH-101, "Strategic Primary Water Chemistry Plan" (Ref. 9. 11) and CH-120 "Primary System Analysis Schedule and Limits" (Ref. 9.12) reduces the susceptibility of RVI components to SCC and IASCC. The chemistry monitoring program places limits on oxygen, halogens, and sulfate species that effectively prevent SCC for austenitic stainless steel components and greatly reduce the probability of IASCC. The limits imposed by the chemistry monitoring program are consistent with the EPRI "PWR Primary Water Chemistry Guidelines" (Ref. 9.13).
5.2.3
Conclusion:
This element complies with the corresponding aging management attribute in NUREG-1801,Section XI.M16.
5.3 SRP Element 3 - Parameters Monitored, Inspected, and/or Tested GALL Report Element Definition:
The program monitors the effects of cracking on the intended function of the component by detection and sizing of cracks by augmentation of the In-Service Inspection requirements in accordance with the requirements of the ASME Code,Section XI, Table IWB 2500-1.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 5.3.1 Ginna Parameters Monitored, Inspected, and/or Tested The Ginna program monitors for the end effect of the identified degradation mechanisms of cracking, loss of bolted joint integrity and wear through application of existing ASME Code Section XI, ISI Program, augmented inspections, and other existing plant and industry programs.
5.3.2 The Ginna Station In-service Inspection Program This existing program (Ref. 9.30 and 9.45) is credited for inspection of those reactor internals assemblies requiring aging management as identified by the Ginna Station AMR and listed in Attachment C. Industry programs also credited use of the ASME Code inspections where inspections are adequate to detect internals degradation mechanisms such as those listed in the Existing Programs of Attachment D, Table 1.3, and Attachment E, Table 1.3.
The requirements of ASME Section XI, Subsection IWB, prescribe using VT-3 methods for examination of Categories B-N-1, B-N-2 and B-N-3. These examinations have been performed during each 10-year ISI interval, i.e. 1979, 1989 and 1999 when the upper and lower internals assemblies were removed from the vessel 4 . Inspections and examinations are documented in procedure EP-VT- 110 (Ref. 9.9).
The tables provided in Attachments C and D identify which components credit the use of the ASME Section XI inspections for license renewal.
5.3.3 Augmented Examinations Per the Ginna Station AMR, LRAM-RVI, the components requiring aging management are listed in summary table Attachment C. In Attachment C, credit was taken for the existing Section XI ISI inspection Program, Water Chemistry Program, and this Reactor Vessel Internals Inspection Program, as noted in the last column identified as "Program/Activity."
Augmented inspections, frequencies, and inspection techniques were to be determined based on the additional evaluations provided by the Industry in order to supplement the ASME Section XI Program inspections. These evaluations were to provide needed clarification as to the level of inspection quality needed; and to determine the proper examination method and frequencies.
4 Ginna's 4'h Interval ended in 2009. As such, the 10 year IS!inspection was to occur during the 2009 RFO. The submittal of Relief Requests 18, 19, 20, and 21 deferred the 10 year ISl inspections to the 2011 RFO (Ref. 9.72).
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 The summary tables provided by MRP-227, Attachments D and E, provide the level of examination, method, and frequency of inspection required for each of the components evaluated by the industry, which as previously discussed in Section 4.4.
Therefore, augmented inspections credited in Attachment C that are listed under the Reactor Vessel Internals Inspection Program, each being specified in section 5.3.3.1, are credited as part of the additional industry evaluations summarized in Attachments D and E. As discussed in Section 4.3.2, Ginna performed the inspections listed in Attachment D during the 2011 RFO, due to License Renewal Commitments, and has committed to performing the inspections listed in Attachment E for future outages at the specified frequency.
In the case where Attachments D and E may list additional Primary Components that require inspections that do not appear in the Attachment C list, the additional components of Attachments D and E are included for augmented inspection as listed in Section 5.3.3.2.
Therefore, the Ginna program not only meets the original requirements of0the LRAM-RVI methodology previously approved by the Ginna Station SER, but also captures the MRP-227 guidelines; both in following Revision 0 during the 2011 RFO, due to License Renewal commitments, and in following the NRC reviewed and approved revision in future, applicable, outages.
5.3.3.1 Augmented Inspections required by LRAM-RVI Attachment 10.1 These components are addressed in the order that they are found in the Table of Attachment C.
5.3.3.1.1 Lower Core Plate and Fuel Pins The lower core plate (LCP) is inspected as part of the ISI program to VT-3 as shown in Figure 19 of procedure EP-VT-1 10. It should be noted that the evaluations of the lower core plate performed per MRP-227 Revision 0 and MRP-227-A have shown that no additional measures for the 12 foot core design is required for the LCP used at Ginna Station5 . Therefore, the existing ISI program was credited to detect aging in the lower core plate during the 2011 RFO when MRP-227 Revision 0 was implemented. The ISI program will also be credited to detect aging in the lower core plate when MRP-227-A is implemented at Ginna. No augmented inspections are required.
5 Reference Attachment D, Table 1.3, Existing Programs, Lower Internals Assembly with regard to MRP-227 Revision 0 and Attachment E, Table 1.3, Existing Programs, Lower Internals Assembly with regard to MRP-227-A Page 30 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 5.3.3.1.2 Lower Support Columns The MRP tables divide this component into two subassemblies, lower support column bolts and lower support column bodies.
5.3.3.1.2.1 Lower Support Column Bolts The lower support column bolts are identified as expansion components in the MRP-227 Revision 0 table of Attachment D Table 1.2. They are tied to the examination results of the baffle-former bolts. Since the plant specific justification for the baffle-former bolts was fulfilled during the ?2011 RFO, this expansion criterion was not entered.
The lower support column bolts are also identified as expansion components in the MRP-227-A table of Attachment E Table 1.2. If the volumetric examination of the baffle-former bolts is not satisfactory at the time that this exam is implemented at Ginna, then 100% of the accessible lower support column bolts will be volumetrically examined, unless a plant specific justification is implemented.
5.3.3.1.2.2 Lower Support Column Bodies The lower support column bodies at Ginna are made of ASTM Type 304 stainless steel, hot rolled, annealed, and pickled. They are identified as expansion components in the MRP-227 Revision 0 table of Attachment D Table 1.2. They are tied to the examination results of the upper core barrel flange weld of the Core Barrel Assembly. Since the examination results for the upper core barrel flange weld of the Core Barrel Assembly were satisfactory, this expansion criterion was not entered.
The lower support column bodies are also identified as expansion components in the MRP-227-A table of Attachment E Table 1.2. If the enhanced visual examination (EVT-1) of the upper core barrel flange weld is not satisfactory at the time that this exam is implemented at Ginna, then 100% of the accessible surfaces of the lower support column bodies will be examined via EVT-1, per the expansion link.
5.3.3.1.3 Core Barrell and Flange The core barrel assembly has been evaluated by the industry program to be a "Primary" component in MPP-227 Revision 0, as shown in Attachment D, Table 1.1. The additional evaluations show that the upper core barrel flange weld is the primary component of interest.
The requirement was to inspect 100% of one side of the accessible surfaces of the selected weld and adjacent base metal, via an EVT-1 examination to determine the presence of crack-like surface flaws. Ginna initially inspected the inner diameter upper core barrel flange weld, Page 31 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 but upon some difficulty of achieving the criteria of an EVT- 1 exam, due to difficulties of the camera focusing on the weld since it was flush to the core barrel, Ginna examined the outer diameter upper core barrel flange weld. This exam was completed successfully, meeting the criteria of an EVT-1 exam, with no anomalies found. While attempting to inspect the inner diameter of the weld, Ginna was able to take credit for an EVT-1 examination for part of the weld length on the inner diameter and a complete VT-3 examination on the inner diameter.
The core barrel assembly has also been evaluated by the industry program to be a "Primary" component in MRP-227-A, as shown in Attachment E, Table 1.1. The upper core barrel flan e weld, upper and lower core barrel cylinder girth welds, and the lower core barrel flange weld are the primary components of interest. For ease of reference, these welds are shown on Attachment G, Figure 4.
With regard to the upper core barrel flange weld, the requirement is to inspect 100% of one side of the accessible surfaces of the selected weld and adjacent base metal, via an EVT- 1 examination to determine the presence of crack-like surface flaws. If the EVT-1 examination of the upper core barrel flange weld is not satisfactory at the time that this exam is implemented at Ginna, then 100% of one side of the accessible surfaces of the core barrel outlet nozzle welds and adjacent base metal will be examined via EVT-1, per the expansion link. As stated above, this requirement was met during the 2011 RFO, with no expansion link required.
With regard to the upper and lower core barrel cylinder girth welds, the requirement is to inspect 100% of one side of the accessible surfaces of the selected weld and adjacent base metal, via an EVT-1 examination. If the EVT-1 examination of the upper and lower core barrel cylinder girth welds is not satisfactory at the time that this exam is implemented at Ginna, then 100% of one side of the accessible surfaces of the selected weld and base metal of the upper and lower core barrel cylinder axial welds will be examined via EVT- 1, per the expansion link. Both the inner diameter and the outer diameter of the upper core barrel cylinder girth weld was visually inspected, per VT-3 requirements, during the 2011 RFO per the ISI program. This weld was determined to be satisfactory per the VT-3 inspection on both the inner diameter and the outer diameter.
6 Ginna has a core barrel-to-support plate weld. For consistency with terminology in MRP-227-A, it will be referred to as the lower core barrel flange weld in this PBD.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 With regard to the lower core barrel flange weld, the requirement is to inspect 100% of one side of the accessible surfaces of the selected weld and adjacent base metal, via an EVT-1 examination. The lower core barrel flange weld was visually inspected, per VT-3 requirements, during the 2011 RFO per the ISI program. This weld was determined to be satisfactory per the VT-3 inspection. As part of this inspection, roughly one foot of the lower axial weld (the weld between the lower core barrel cylinder girth weld and the lower core barrel flange weld) was inspected to VT-3 criteria and determined to be satisfactory.
During the 2011 RFO, Ginna did not inspect the lower core barrel cylinder girth weld. Also, Ginna did not inspect the upper core barrel cylinder girth weld or the lower core barrel flange weld to the EVT-1 criteria, as specified in MRP-227-A, due to the fact that Ginna performed exams according to MRP-227 Revision 0. It should be noted, however, that as a whole, Ginna did inspect 75% of the core barrel circumferential welds with no sign of degradation. Further, Ginna did inspect the core barrel outlet nozzle welds to VT-3 criteria as part of the ISI exams and did not see wear. 100% of the CRGT flange welds were also inspected to EVT-1 criteria and no wear was noted. Although these welds are under a different form of stress than the lower core barrel cylinder girth weld and were examined to a less rigorous examination than the EVT-1 requirements specified for the lower core barrel cylinder girth weld, they are made of stainless steel, just as the lower core barrel cylinder girth weld, and are in relatively high fluence areas.
The core barrel is not particularly susceptible to fatigue, as indicated by the low fatigue usage factors, when compared to other components of the core that were examined and determined to be satisfactory. Specifically, the upper core barrel girth weld has a fatigue usage factor that is fifteen times less than the lower core plate and nine times less than the core barrel outlet nozzles. Further, the lower core barrel girth weld has a fatigue usage factor that is ten times less than that of the lower core plate and six times less than that of the core barrel outlet nozzles (Ref. 9.68). As stated in the above paragraph, the core barrel outlet nozzle welds were inspected as part of the ISI exams and determined to be satisfactory. The lower core plate was also inspected during the 2011 RFO as part of the ISI exams, to VT-3 inspection standards, and was also determined to be satisfactory.
Therefore it is safe to conclude that although the lower core barrel cylinder girth weld was not inspected and the upper core barrel cylinder girth weld and the lower core barrel flange weld were not inspected to EVT-1 criteria, they are in an acceptable condition to continue to maintain their function in supporting the core. There is reasonable assurance that they will continue to perform their intended function. It is acceptable to not inspect the upper core barrel cylinder girth weld, the lower core barrel cylinder girth weld, and the lower core barrel flange weld to the EVT-1 criteria until the next MRP-227 examination interval.
Since the integrity of the upper and lower core barrel cylinder girth welds has been evaluated to be satisfactory until they are next inspected, the expansion link to inspect the upper and lower core barrel cylinder axial welds does not have to be entered at this time.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 5.3.3.1.4 Thermal Shield The thermal shield has been evaluated by the industry program to be a "Primary" component in MRP-227 Revision 0, as shown in Attachment D, Table 1.1. The additional evaluations show that the thermal shield flexures are the primary components of interest. The requirement was to inspect 100% of the thermal shield flexures, via a VT-3 examination. This exam was completed successfully, with satisfactory results.
The thermal shield has been evaluated by the industry program to be a "Primary" component in MRP-227-A, as shown in Attachment E, Table 1.1. The thermal shield flexures are the primary components of interest. The requirement is to inspect 100% of the thermal shield flexures, via a VT-3 examination. This examination will be implemented at Ginna.
5.3.3.1.5 Bolting The last component group of Attachment C summary table is "bolting." It includes the lower support column bolts, baffle-former bolts, and barrel-former bolts.
5.3.3.1.5.1 Lower Support Column Bolts The Lower support column bolts have been addressed in Section 5.3.3.1.2.1.
5.3.3.1.5.2 Baffle-Former Bolts The baffle-former bolts have been addressed in Section 4.2.2.
5.3.3.1.5.3 Barrel-Former Bolts The barrel-former bolts are identified as expansion components in the MRP-227 Revision 0 table of Attachment D Table 1.2. They are tied to the examination results of the baffle-former bolts. Since the plant specific justification for the baffle-former bolts was fulfilled and the acceptance criteria was met during the 2011 RFO, this expansion criterion was not entered.
The barrel-former bolts are also identified as expansion components in the MRP-227-A table of Attachment E Table 1.2. If the volumetric examination of the baffle-former bolts is not satisfactory at the time that this exam is implemented at Ginna, then 100% of the accessible barrel-former bolts will be volumetrically examined, per the expansion link.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 5.3.3.2 Augmented Inspections required by MRP-227 This section includes augmented inspections for those intemals components that have been classified as "Primary Components" by the industry list provided in Attachments D and E.
This section adds additional industry augmented inspections required by the "Primary Component" list of Attachment D, Table 1.1 and Attachment E. Table 1.1. By adding the augmented inspections for the Primary Components, Ginna station will not only meet the original Ginna AMR intent; but, has also meet the intent of the industry program described by MRP-227, Revision 0, and will also meet the intent of the industry program described by MRP-227-A.
5.3.3.2.1 Control Rod Guide Tube Assembly (Guide Cards)
The Control Rod Guide Tube Guide Cards have been evaluated by the industry program to be a "Primary" component in MRP-227 Revision 0, as shown in Attachment D, Table 1.1. The requirement was to inspect 20% of the number of CRGT assemblies, with all guide cards within each selected CRGT examined, via a VT-3 examination. Due to the fact that Ginna elected to replace all of its split pins during the 2011 RFO, as described in section 4.2.1, all of the 29 active control rod guide tube assemblies were accessible and therefore all guide cards within these assemblies were accessible. Ginna took advantage of the opportunity and inspected all of its 261 guide cards (9 guide card per CRGT). Westinghouse provided WCAP-17480 (Ref. 9.86) to evaluate the amount of wear in the guide cards. It was evaluated that the wear on the guide cards is low and determined that significant wear will not be achieved for another 35 - 41 Effective Full Power Years for the most limiting guide card.
The Control Rod Guide Tube Guide Cards have been evaluated by the industry program to be a "Primary" component in MRP-227 -A, as shown in Attachment E, Table 1.1. The requirement is to inspect 20% of the number of CRGT assemblies, with all guide cards within each selected CRGT examined, via a VT-3 examination. Since WCAP-17480 has demonstrated that Ginna's guide card wear is minimal, and will not show significant wear until well beyond Ginna's first period of extended operation, relief from this exam will be sought prior to the next MRP-227 inspection. If not granted, Ginna will exam the required amount of guide cards per MRP-227 Rev A.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 5.3.3.2.2 Control Rod Guide Tube Assembly (Lower Flange Welds)
The Control Rod Guide Tube lower flange welds have been evaluated by the industry program to be a "Primary" component in MRP-227 Revision 0, as shown in Attachment D, Table 1.1.
The requirement was to inspect 100% of the outer accessible CRGT lower flange weld surfaces and adjacent base metal, via an EVT-1 exam to determine the presence of crack-like surface flaws. Ginna performed this inspection during the 2011 RFO on each of the 29 active CRGT's. All lower flange welds were determined to be satisfactory.
The Control Rod Guide Tube lower flange welds have been evaluated by the industry program to be a "Primary" component in MRP-227-A, as shown in Attachment E, Table 1.1. The requirement is to inspect 100% of the outer accessible CRGT lower flange weld surfaces and adjacent base metal on the individual periphery CRGT assemblies, via an EVT-1 exam to determine the presence of crack-like surface flaws. Ginna is committed to performing this inspection when it implements MRP-227-A.
5.3.3.2.3 Baffle-Former Assembly (Baffle-Edge Bolts)
The baffle-edge bolts have been evaluated by the industry program to be a "Primary" component in MRP-227 Revision 0, as shown in Attachment D, Table 1.1. The requirement was to inspect 100% of the accessible components from the core side of the bolts and locking devices on the high fluence seems, via a VT-3 examination. Ginna performed this inspection during the 2011 KFO on all of its baffle-edge bolts. All baffle-edge bolts and their associated washers (locking devices) were present, with no anomalies found.
The baffle-edge bolts have been evaluated by the industry program to be a "Primary" component in MRP-227-A, as shown in Attachment E, Table 1.1. The requirement is to inspect 100% of the accessible components from the core side of the bolts and locking devices on the high fluence seems, via a VT-3 examination. Ginna is committed to performing this inspection when it implements MRP-227-A.
5.3.3.2.4 Baffle-Former Assembly (Assembly)
The baffle former assembly has been evaluated by the industry program to be a "Primary" component in MRP-227 Revision 0, as shown in Attachment D, Table 1.1. The requirement was to perform a visual examination (VT-3) to check for evidence of distortion on the core side surface. Ginna performed this inspection during the 2011 RFO. No anomalies were found.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 The baffle former assembly has been evaluated by the industry program to be a "Primary" component in MRP-227-A, as shown in Attachment E, Table 1.1. The requirement is to perform a visual examination (VT-3) to check for evidence of distortion on the core side surface to include the baffle plates, baffle-edge bolts, and indirect effects of void swelling in the former plates. Ginna is committed to performing this inspection when it implements MRP-227-A.
5.3.3.2.5 Alignment and Interfacing Components (Internals Hold Down Spring)
The internals hold down spring has been evaluated by the industry program to be a "Primary" component in MRP-227 Revision 0, as shown in Attachment D, Table 1.1. The requirement was to perform a direct measurement of the spring height at several points around the circumference of the spring, with a statistically adequate number of measurements at each point to minimize uncertainty. This examination was only applicable to plants with Type 304 stainless steel hold down spring. Since Ginna has a Type 410 stainless steel hold down spring, it was exempt from this examination and therefore did not perform it.
The internals hold down spring has been evaluated by the industry program to be a "Primary" component in MRP-227-A, as shown in Attachment E, Table 1.1. The requirement is to perform a direct measurement of the spring height at several points around the circumference of the spring, with a statistically adequate number of measurements at each point to minimize uncertainty. This examination is only applicable to plants with Type 304 stainless steel hold down spring. Since Ginna has a Type 410 stainless steel hold down spring, it is exempt from this examination and therefore does not intend on performing it when it performs the examinations of MRP-227-A.
5.3.4 Industry Issues Program Outputs 5.3.4.1 Thimble Tubes The Ginna Thimble Tube Inspection Program is an existing plant specific program that follows Branch Technical position RLSB-1, "Aging Management Review - Generic" which is included in Appendix A of NUREG-1800 (Ref 9.52). The Ginna Thimble Tube Inspection Program manages cracking due to SCC of BMI guide tubes and cracking of BMI guide tube fillet welds. This program requires periodic visual inspection of fillet welds.
The program also includes eddy current testing requirements for thimble tubes and includes criteria for determining sample size, inspection frequency flaw evaluation and corrective action in accordance with US NRC Bulletin 88-09 (Ref. 9.53), and NUREG 1801 (Ref 9.1). Specific Ginna actions are summarized in reference 9.28. Replacement of Ginna's Thimble Tubes during the 2011 RFO is discussed in section 4.1.3 of this PBD.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 5.3.4.2 Industry Reactor Internals Guidelines-Discussion Per the Ginna Station license renewal SER (Ref. 9.37), Ginna Station committed to participate in industry activities associated with developing a standard industry program to provide inspection and evaluation guidelines for reactor internals inspections. The industry program results are summarized as inspection and evaluation guidelines as documented in Ref 9.38.
Augmented inspections as prescribed by the industry efforts are included in Attachment D, Tables 1.1, 1.2, 1.3, and 1.4.
The tables in Attachment D represent a consensus view of industry experts that call for additional inspections as noted. The tables are based on additional evaluations, analyses, and expert opinion reviews that were performed by the industry in order to determine which components of the reactor internals require augmented inspections. This approach demonstrated that degradation mechanisms would not inhibit the intended functions of the internals components as they continue to operate through the period of extended operation.
These evaluations form the basis for the completion of Ginna LRAM-RVI table Attachment C components that indicate additional evaluations by the industry were required to identify the necessary inspections, frequencies, and methods.
The industry program compiled available research data; additional evaluation techniques, such as an IASCC predictive model; and expert opinion evaluations in order to determine what reactor internals components required additional inspection. The industry outputs are contained in References 9.40, 9.41, 9.42 and 9.43. Based on these outputs, a screening and ranking process was developed and applied to all removable reactor internals components in order to identify those components that required additional inspections above and beyond the inspections required by the ASME Section XI Code requirements.
As previously mentioned in Section 4.3.2, inspections as identified in the tables of Attachment D are from MRP-227 Revision 0. They were intended to provide reasonable assurance that the existing licensing basis functions would not be impacted by the degradation mechanisms identified in the tables. The augmented examinations are based on a hierarchy of components that leads to the tables provided in Attachment D. The tables rank the components as Primary, Expansion, and Exiting where the "Primary" component list is predicted to contain the most susceptible component to the particular degradation mechanism. These components are expected to display the identified degradation mechanism at first on-set. Upon NRC review of MRP-227 Revision 0, MRP-227-A was issued.
In the MRP process, each of the vessel internals subcomponents has been categorized using the screening and categorization process described in MRP-191 (Ref. 9.43).
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 Augmented inspections by the MRP-227 work have been used in Section 5.3.3 to complete this LR-RVI-PROGPLAN which is credited in the LRAM-RVI Table Attachmaent C. These augmented inspections have been taken from the tables of Attachment D and were completed during the 2011 RFO, in order to coincide with the Ginna Station fourth interval 10 year ISI examinations. Existing Ginna Station NDE procedure EP-VT-1 10 (Ref. 9.9) was modified following the 20019 RFO through action item NL-2009-000021-001, as shown in Attachment A, Table 1. Modifications to the procedure were delayed until after the 2009 RFO as portions of the upper internals were examined during the Fall 2009 RFO. The 2009 RFO components were part of the normal "period" examination required by the ASME Section XI ISI Program (Ref 9.87).
5.3.4.3
Conclusion:
The Ginna Station AMP for reactor internals monitors for the effects of aging mechanisms by inspecting for cracking, wear and other relevant conditions in reactor internals components.
These inspections performed in accordance with the requirements of the ASME Code,Section XI, Table IWB 2500-1, existing programs and the guidance of the MRP-227 program recommendations provide reasonable assurance that the intended functions of the components will continue to be met through the period of extended operation.
This element complies with the corresponding aging management attribute in NUREG- 1801,Section XI.M16.
5.4 SRP Element 4 - Detection of Aging Effects GALL Report Element Definition:
The extent and schedule of the inspection and test techniques prescribed by the aging management program are designed to maintain structural integrity and ensure that aging effects will be discovered and repaired before the loss of intended function. Inspection can reveal crack initiation and growth. Vessel internal components are inspected in accordance with the requirements of ASME Section XI, Subsection IWB, examination category B-N-3 for all accessible surfaces of reactor core support structure that can be removed from the vessel. The ASME Section XI inspection specifies visual VT-3 examination to deternmine the general mechanical and structural condition of the component supports by (a) verifying parameters, such as clearances, settings, and physical displacements, and (b) detecting discontinuities and imperfections, such as loss of integrity at bolted or welded connections, loose or missing parts, debris, corrosion, 'wear, or erosion.
However, visual VT-3 examination is to be augmented to detect tight or fine cracks. Also, historically the VT-3 examinations have not identified bolt cracking because cracking occurs at the juncture of the bolt head and shank, which is not accessible for visual inspection.
Creviced and other inaccessible regions are difficult to inspect visually. This AMP recommends more stringent inspections such as enhanced visual VT-1 examinations or Page 39 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ultrasonic methods of volumetric inspection, for certain selected components and locations.
The inspection technique is capable of detecting the critical flaw size with adequate margin.
The critical flaw size is determined based on the service loading condition and service degraded material properties. For non-bolted components, augmented ISI may include enhancement of the visual VT-1 examination of Section XI IWA-2210. A description of such an enhanced visual VT-1 examination should include the ability to achieve a 0.0005-in.
resolution, with the conditions (e.g., lighting and surface cleanliness) of the inservice examination bounded by those used to demonstrate the resolution of the inspection technique.
For bolted components, augmented ISI is to include other demonstrated acceptable inspection methods to detect cracks between the bolt head and the shank. Alternatively, the applicant may perform a component-specific evaluation, including a mechanical loading assessment to determine the maximum tensile loading on the component during ASME Code Level A, B, C, and D conditions. If the loading is compressive or low enough (<5 ksi) to preclude fracture, then supplemental inspection of the component is not required. Failure to meet this criterion requires continued use of the augmented inspection methods.
5.4.1 Ginna Detection of Aging Effects 5.4.1.1 Visual examinations The extent of the Ginna inspections that were performed during the 2011 RFO, and that will be performed in future RFO's, in accordance with the requirements of ASME Section XI requirements and augmented inspections prescribed by the industry program are listed in EP-VT- 110 and Attachments D and E. Visual examinations are typically used to detect cracking and wear. The type of visual examination used will be, as a minimum, the VT-3 examination specified by ASME Section XI, Subsection IWB, Category B-N-3.
Augmented examinations for cracking may consist of VT-1, or Enhanced VT-1 (EVT-1) examination which is defined as a visual examination with a finer resolution than the standard VT-1 requirements of ASME Section XI, or volumetric examination when warranted.
Augmented inspections of Attachments C, D, and E identify the effects of aging, the examination parameters to be monitored, and the frequency and condition monitoring programs needed to maintain the continued functionality of the SSC.
Industry inspection standards for augmented inspections are contained in MRP-228 (Ref. 9.56).
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 Other examination methodologies selected for use in the industry guidelines are visual VT-I and EVT- 1 examinations. These examinations were selected where a greater degree of detection capability than VT-3 examination is needed to manage the aging effect. Unlike the detection of general degradation conditions by VT-3 examination, VT-I and EVT-1 examinations are conducted to detect discontinuities and imperfections on the surface of components, including conditions such as cracks, wear, corrosion, or erosion. Specifically, VT-1 is used for the detection of surface discontinuities such as gaps, while EVT-1 is used for the detection of surface breaking flaws.
When specified, a VT-I examination is conducted in accordance with the requirements of the ASME Code (Ref. 9.57). EVT-1 examination will be conducted in accordance with any additional requirements for EVT-1 examination (such as camera scanning speed) as specified in the MRP 228 Inspection Standard (Ref 9.56).
The current ASME Code (Ref. 9.57) requirements for VT-1 examination became more rigorous than the previous ASME Code versions. Many previous VT- 1 examinations were only required to discern a 1/32" black line on a gray background. These limitations led the NRC and industry to adopt modified visual examinations for use in detecting flaws discovered in boiling water reactor (BWR) internals. The most recent research conducted by the EPRI Non-Destructive Examination (NDE) Center established the VT-I character heights specified in Reference 9.57 as equally or better able to detect the degradation effects than the modified visual examination requirements developed previously.
5.4.1.2 Volumetric Examinations Ultrasonic examination was selected for volumetric examinations for cases where visual or surface examination is unable to detect the effect of the age related degradation mechanism such as for baffle-former bolts inspections. Ultrasonic testing (UT) or other equivalent NDE technique may be used to detect cracking. UT of the baffle-former bolts is used to detect cracking caused by IASCC under the baffle-former bolt head with go/no go criteria applied for detection.
Indications recorded during the volumetric examinations of the baffle-former bolts were evaluated using the procedures similar to those found in Ref. 9.7 which were developed for the 1999 inspection of baffle-former bolts. Technical Justifications for volumetric exams of baffle-former bolts implemented during the 2011 RFO are found in references 9.81, 9.88, and 9.89.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 5.4.1.3
Conclusion:
Detection of indications that are required by the ASME Section XI, ISI Program are well established and field-proven through the application of the Section XI Program. Those augmented inspections that are taken from the MRP-227 recommendations have been applied through use of the MRP 228 Inspection Standard (Ref. 9.56).
This element is consistent with the corresponding aging management program attribute in NUREG-1801,Section XI.M16.
5.5 SRP Element 5 - Monitoring and Trending GALL Report Element Definition:
Inspection schedules in accordance with IWB-2400, assessment of susceptible or limiting components or locations, and reliable examination methods provide timely detection of cracks.
The scope of examination expansion and re-inspection beyond the baseline inspection are required if flaws are detected.
5.5.1 Ginna Program -- Monitoring and Trending Inspections credited in LRAM-RVI Attachment C, are based on utilizing the Ginna 10 year ISI program, and the augmented inspections described in Section 5.3.3 derived from the industry program contained in Attachments D and E. These inspections were conducted in conjunction with the 4th Interval, 10 Year ISI examinations and will be conducted in conjunction with the 5 th Interval, 10 Year ISI examinations.
Attachments D and E, Tables 1.1, 1.2, 1.3, and 1.4 identify the aging effects and timely inspection/mitigation actions for Primary components. As discussed in Section 2.3 and Reference 9.68, arn assessment of the MRP-227 "Primary" components provides reasonable assurance for demonstrating the "Primary" components current capacity to perform their intended functions.
Reporting requirements are included as part of the MRP-227 guidelines. Inspection result reporting will enable the industry to monitor reactor internals degradation on an on-going 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 pre-emptive action through update of the MRP guidelines.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 5.5.2
Conclusion:
This element of the Ginna Station program complies with the corresponding aging management attribute in NUREG- 1801,Section XI.M16.
5.6 SRP Element 6 - Acceptance Criteria GALL Report Element Definition:
Any indication or relevant condition of degradation is evaluated in accordance with IWB-3 100 by comparing ISI results with the acceptance standards of IWB-3400 and IWB-3500.
5.6.1 Ginna Program Acceptance Criteria Those recordable indications that are the result of inspections required by the exiting Ginna Station ISI program scope are evaluated in accordance with the applicable requirements of the ASME Code through the existing CNG-CA-1.01-1000 Corrective Action Program.
With regards to the MRP-227 Revision 0 exams that have been conducted, inspection acceptance and expansion criteria were provided in Attachment D Table 1.4. These criteria were also included in EP-VT- 110, which enabled the examiner to identify examination acceptance criteria. As stated in Section 5.3.3, all examinations were satisfactory and therefore the expansion criteria did not have to be entered.
For future examinations, with regards to MRP-227-A, inspection and acceptance criteria are provided in Attachment E, Table 1.4. Augmented inspections discussed in Section 5.3.3, that result in recordable relevant conditions, will be entered into the plant Corrective Action Program and addressed by appropriate actions that may include enhanced inspection, repair, replacement, mitigation actions or analytical evaluations using a "Flaw Evaluation" handbook-type approach such as those documented in WCAP 17016 NP (Ref. 9.61) or Reference 9.62 to support continued component or assembly functionality.
5.6.2
Conclusion:
This element complies with the corresponding aging management attribute in NUREG- 1801,Section XI.M16.
5.7 SRP Element 7 - Corrective Actions GALL Report Element Definition:
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 Repair and replacement procedures are equivalent to those requirements in ASME Section XI.
Repair is in conformance with IWB-4000 and replacement occurs according to IWB-7000. As discussed in the appendix to this report, the staff finds the requirements of 10 CFR Part 50, Appendix B, acceptable in addressing corrective actions.
5.7.1 Ginna Program The NRC has addressed this element in NUREG- 1801, Appendix, Quality Assurance for Aging Management Programs.
R.E. Ginna has an established 10 CFR Part 50, Appendix B, Program that addresses the elements of corrective actions, confirmation process, and administrative controls. The R.E.
Ginna Program includes non-safety related structures, systems, and components. (CNG-CA-1.01-1000, Ref. 9.18) 5.7.2
Conclusion:
This element complies with the corresponding aging management attribute in NUREG-1801,Section XI.M 16.
5.8 SRP Element 8 - Confirmation Process GALL Report Element Definition:
Site quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B. As discussed in the appendix to this report, the staff finds the requirements of 10 CFR Part 50, Appendix B, acceptable in addressing the confirmation process and administrative controls.
5.8.1 Ginna Program See SRP Element 7- "Corrective Actions" in Section 5.7 above.
5.8.2
Conclusion:
This element complies with the corresponding aging management attribute in NUREG-1801,Section XI.M16.
5.9 SRP Element 9 - Administrative Controls GALL Report Element Definition Page 44 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 See Item 8, above.
5.9.1 Ginna Program See SRP Element 7- "Corrective Actions" above.
5.9.2
Conclusion:
This element complies with the corresponding aging management attribute in NUREG- 1801,Section XI.M 16.
5.10 SRP Element 10 - Operating Experience Gall Report Element Definition:
Because the ASME Code is a consensus document that has been widely used over a long period of time, it has been shown to be generally effective in managing aging effects in Class 1, 2, or 3 components and their integral attachments in light-water cooled power plants.
In PWRs, stainless steel components have generally not been found to be affected by SCC because of low dissolved oxygen levels and control of primary water chemistry. However, the potential for SCC exists due to inadvertent introduction of contaminants into the primary coolant system from unacceptable levels of contaminants in the boric acid; introduction through the free surface of the spent fuel pool, which can be a natural collector of airborne contaminants (NRC IN 84-18); introduction of relatively high levels of oxygen during shutdown; or from aggressive chemistries that may develop in creviced regions. Cracking has occurred in Stainless Steel baffle-former bolts in a number of foreign plants (NRC IN 98-11) and has now been observed in plants in the United States.
5.10.1 Ginna Program Extensive industry and Ginna Station operating experience has been reviewed during the development of the Reactor Vessel Internals Program. The experience reviewed includes NRC Information Notice 98-11, "Cracking Of Reactor Vessel Internal Baffle Former Bolts In Foreign Plants," and NRC Information Notice 84-18, "Stress Corrosion Cracking in PWR Systems." 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 guide tube split pins has also been reported.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 As described previously, a review of plant-specific experience with reactor vessel internals reveals that Ginna Station has responded proactively to industry issues with respect to reactor internals degradation. Examples that demonstrate this proactive response are the preemptive replacement of control rod guide tube split pins, both in 1986 and 2011, and augmented examination and preemptive replacement of baffle-former bolts in 1999 and 2011. These were previously discussed in sections 4.2.1 and 4.2.2.
Early plant operating experience related to hot functional testing and reactor internals is documented in Ref. 9.65. Inspections performed as part of the 10 year ISI program would be expected to discover overall general internals structure degradation.
Ginna Station has also participated in the EPRI Reactor Vessel Internals Issue Task Group (EPRI RI-ITG, Focus Group, and Core Writers Group), and now participates in the MRP Assessment ITG, which has now replaced the Core Writers Group as the group responsible for the Reactor Internals Guidelines.
The Assessment ITG continues to monitor and engage in ongoing research efforts on aging of reactor vessel internals as they represent themselves in the US and Worldwide research efforts and will continue to provide guidance to utilities on corrective actions for aging effects, as conditions warrant. The Pressurized Water Reactor Owners Group (PWROG) has established a materials sub-committee on reactor internals primarily to establish expansion component evaluation criteria, as discussed above, and to provide an interface with the MRP working groups to monitor activities related to reactor vessel internals.
In addition, industry operating experience is routinely reviewed by Ginna Station System engineers and program owners using INPO OE and the barrier analysis concept for the determination of additional actions and lessons learned that can be incorporated into the plant systems quarterly health reports and for consideration for incorporation into plant programs.
Specifically, with regards to the Reactor Vessel Internals, Ginna has taken note of the recent OE that D.C. Cooke and Surry has issued regarding baffle-former bolt failures (OE34581 and OE33420, respectively). Previous OE from domestic and foreign plants regarding baffle-former bolt failures is what led to Ginna preemptively replacing its Stainless Steel Type 347 bolts with Stainless Steel Type 316 bolts. Likewise, OE from other plants is what led Ginna to preemptively replacing its split pins in 1986 and 2011.
A key element of the MRP-227 Guideline is the reporting of age related degradation of reactor vessel components. Ginna Station, through its participation in PWROG and EPRI-MRP activities, will continue to benefit from the reporting of inspection information and will share its own operating experience with the industry through those groups or INPO, as appropriate.
Most notably, Ginna has issued OE34231 to the industry, describing the results of its MRP-227 inspections. It has also reported the results of its MRP-227 inspections to the MRP Program Manager within 120 days following completion of the 2011 RFO, which was the outage during which the PWR internals were examined, as required by MRP-227.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 5.10.2
Conclusion:
This element complies with the corresponding aging management attribute in NUREG- 1801,Section XI.M16.
6.0 DEMONSTRATION The examinations required by ASME Section XI for the Ginna Station reactor vessel internals have been performed during each ten-year interval since plant operations commenced. The examination results are well documented and no unacceptable conditions have been reported.
Ginna Station has been proactive in responding to industry experience regarding reactor vessel internals degradation. Alloy X-750 guide tube split pins were replaced at Ginna Station in 1986 in response to SCC failures of these pins in other Westinghouse plants. The replacement pins for Ginna we:re fabricated from the same nickel-based alloy, but with modified geometry and heat treatment to increase resistance to SCC. These split pins were again preemptively replaced in 2011 with a more resistant, Cold Worked Type 316 material.
All accessible Type 347 stainless steel baffle-former bolts in the Ginna Station reactor vessel internals were inspected by UT during the 1999 refueling outage (Ref. 9.6). The UT examination identified a number of bolts with flaw-like indications. A number of bolts sufficient to guarantee the structural adequacy of the baffle-former assembly were replaced.
The replacement bolts were fabricated from Type 316 stainless steel, which is expected to be more resistant to IASCC than Type 347. Destructive metallurgical analysis of a sample of intact bolts containing flaw-like UT indications suggested false positive UT indications.
Further details of the test results are documented in the AMR report (Ref. 9.35) for the Ginna reactor vessel intermals and in the "Ginna Hot Cell Testing Report" (Ref. 9.25). Further baffle-former bolt work was done during the 2011 RFO due to MRP-227 requirements.
The Water Chemistry Control Program at Ginna Station has been effective in maintaining oxygen, halogens, and sulfate at levels sufficiently low to prevent SCC of the reactor vessel internals. LR-H2OC-PROGPLAN (Ref. 9.10), "Water Chemistry Control Program Basis Document," contains details on the effectiveness of primary water chemistry control.
Review of QA audit reports, NRC inspection reports, and INPO evaluations since 1999 indicate no unacceptable issues related to reactor vessel internals inspections.
As described in previous sections of this PROGPLAN, augmented inspections, derived from the evaluations that support the industry MRP-227 Guidelines, have been utilized in this PROGPLAN to build on existing plant programs. This approach is expected to encourage detection of a degradation mechanism at its first appearance. This approach provides reasonable assurance that the internals components will continue to perform their intended function through the period of extended operation.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 This PROGPLAN also fulfills the approved license renewal methodology requirement to identify the most susceptible components and to inspect those components with an indication detection level commensurate with the expected degradation mechanism indication.
These examinations were performed at Ginna Station during the 2011 RFO. The augmented inspections discussed in Section 5.3.3 were integrated into the inspection procedures used to perform the ASME Code Section XI, Fourth 10-Year Interval ISI examinations procedure EP-VT-110 (Ref. 9.9).
As discussed in Section 4.3.3, the industry MRP-227 Guidelines also provide for updates as experience is gained through inspection results. This feedback loop will enable updates based on actual inspection experience. Ginna has submitted its results from the MRP-227 Revision 0 exams it performed during the 2011 RFO to the MRP group.
As documented in quarterly System and Program Health Reports, Operating Experience (OE) reports are continuously reviewed by system engineers and program engineers for applicable issues that indicate operating procedures or programs require updates based on new OE.
The augmented inspections described in Section 5.3.3, combined with the ASME Section XI, ISI program inspections, existing Ginna Station programs, and use of OE reports provide reasonable assurance that the reactor internals will continue to perform their intended functions through the period of extended operation.
7.0 REQUIRED PROGRAM ENHANCEMENTS / IMPLEMENTATION SCHEDULE The information that was included in this section is now consolidated in Attachment A, Table 1.
8.0
SUMMARY
OF IMPLEMENTING DOCUMENTS / RESPONSIBLE DEPARTMENT The information that was included in this section is now consolidated in Attachment A, Table 1.
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GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012
9.0 REFERENCES
9.1 Generic Aging Lessons Learned (GALL) Report, NUREG-1801, Revision 0, Volumes 1 & 2, U. S. Nuclear Regulatory Commission, July 2001.
9.2 WCAP-14577 Rev. l-A: "License Renewal Evaluation: Aging Management For Reactor Internals," Westinghouse Electric Company, March 2001 9.3 NRC Information Notice 84-18, "Stress Corrosion Cracking in PWR Systems,"
March 7, 1984 9.4 "Void Swelling of Austenitic Internals and its Possible Implications for PWRs,"
presentation by F.A. Garner at the EPRI RPV Internals/JOBB ITG Meeting, July 20-21, 1999, Milwaukee, Wisconsin 9.5 NRC Information Notice 98-11 :"Cracking Of Reactor Vessel Internal Baffle Former Bolts In Foreign Plants," March 25, 1998 9.6 Action Report 99-0469, "Reactor Vessel Baffle Bolts with UT Indications."
FCMS Document Number: 1999-0469.
9.7 WCAP-15036, Rev. 1, "Determination of Acceptable Baffle-Barrel Bolting Distribution Under Fault Conditions" 9.8 10 CFR 54. "Requirements for Renewal of Operating Licenses for Nuclear Power Plants" 9.9 EP-VT-1 10, "Visual Examination of the Reactor Vessel and Removable Internal Structures," Revision 3. April 25, 2011 9.10 LR-H20C-PROGPLAN, "Water Chemistry Control Program," Revision 2.
March 6, 2009.
9.11 CH-101, "Strategic Primary Water Chemistry Plan," Revision 3. April 20, 2012 9.12 CH-120, "Primary System Analysis Schedule and Limits," Revision 8. May 11, 2012 9.13 "Pressurized Water Reactor Primary Water Chemistry Guidelines," Revision 6.
EPRI Product ID 1014986. December 2007.
9.14 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants" Page 49 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 9.15 ANSI N 18.7-1976 "Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants" 9.16 Updated Final Safety Analysis Report (UFSAR), Chapter 17, "Quality Assurance Program for Station Operation" 9.17 ND-QAP, "Quality Assurance Program," Revision 01100, dated August 13, 2008 9.18 CNG-CA- 1.01-1000, "Corrective Action Program," Revision 00600, dated Dec.
7, 2011 9.19 A-1603.2, "Work Order Initiation," Revision 18, dated March 21, 2006 9.20 IP-OPS-4, "Post-Maintenance Operability Testing," Revision 00100, dated August 13, 2008 9.21 NRC Letter dated May 19, 2004, Issuance of Renewed Facility Operating License No DPR- 18 For R.E. Ginna Nuclear Power Plant 9.22 CNG-PR- 101-1011, "Control of Station Specific Procedure Change Process" Revision 00300, dated November 4, 2008 9.23 IP-PRO-4, "Procedure Adherence Requirements," Revision 01500, dated 11-9 2007 9.24 CBG-PR-3.01-1000, "Records Management," Revision 00101, dated August 18, 2008 9.25 Conermanm, J., Shogun, R. P., Junker, W., Wilson, I.L.W., and Spellward, P.,
Westinghouse Letter Report STD-MR-01-0012, Rev. 0, "Ginna Hot Cell Testing Report," Nov. 26, 2001 9.26 LR-PSPM-PROGPLAN, "Periodic Surveillance and Preventive Maintenance Program" 9.27 American Society of Mechanical Engineers Boiler and Pressure 'Vessel Code,Section XI, Rules for In-service Inspection of Nuclear Power Plant Components, 1995 Edition, 1996 Addenda, American Society of Mechanical Engineers, New York, NY 9.28 LR-TTI-PROGPLAN, "Thimble Tube Inspection Program," Revision 4, May 2, 2009 Page 50 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 9.29 WCAP-15437, "Study of the Sensitivity of Acceptable Baffle-Barrel-Former Bolting Patterns to Fuel Type for Two-Loop Domestic Plants," September 2000 9.30 CENG-GNPP-ISI-005, Revision 2, "Ginna Station Fifth Ten-Year Inservice Inspection Plan." December 12, 2011.
9.31 Intentional Blank 9.32 WCAP-1483 1, "Baffle-Former Bolt Program for Westinghouse Owners Group Phase 5: Operability Investigation II," December 1997 9.33 WCAP-15029, "Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distribution Under Faulted Load Conditions,"
November 1998 9.34 NUREG 1786, Safety evaluation Report Related to the License Renewal of the R.E. Ginna Nuclear Power Plant, May 2004.
9.35 Ginna Nuclear Power Plant, License Renewal Project, Aging Management Review Report, LRAM-RVI, Revision 0, April 9, 2002.
9.36 Application for Renewed Operating License, R. E. Ginna Nuclear Power Plant, Section 2.3.1.3, Reactor Vessel Internals, August 2002.
9.37 Safety Evaluation Report Related to the License Renewal of R. E. Ginna Nuclear Power Plant (NUREG-1786), May 2004.
9.38 Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0), EPRI Product ID 1016596, December 2008 9.39 Appendix D: Materials Guidelines: Implementation Protocol, in "Guidelines for the Management of Materials Issues," NEI 03-08, Nuclear Energy Institute, Washington, DC.
9.40 Materials Reliability Program: Framework and Strategies for Managing Aging Effects in Reactor Internals (MRP-134), EPRI, Palo Alto, CA: June 2005.
1008203.
9.41 Materials Reliability Program: Development of Material Constitutive Model for Irradiated Austenitic Stainless Steels (MRP-135), EPRI, Palo Alto, CA: 1011127.
Page 51 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 9.42 Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175), EPRI, Palo Alto, CA:
1012081.
9.43 Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191), EPRI, Palo Alto, CA: 2006. 1013234.
9.44 ASME Boiler and Pressure Vessel Code,Section XI, "Rules for In-service Inspection of Nuclear Plant Components," 2004 Edition.
9.45 LR-IWBD-PROGPLAN, "ASME Section XI, Subsection IWB, IWC, & IWD Inservice Inspection Program," Revision 5. April 17, 2009 9.46 Intentional Blank 9.47 IP-IIT-8, "RCS Materials Degradation Management Program," Revision 1. July 23, 2007.
9.48 Materials Reliability Program: Pressurized Water Reactor Issue Management Tables (MRP-205), EPRI, Palo Alto, CA: 2006. 1014446.
9.49 Intentional Blank 9.50 Intentional Blank 9.51 CNG-AM-4.01, "RCS Material Degradation Management Program," Revision 1.
July 16, 2007.
9.52 NUREG -1800, "Standard Review Plan for Review of License Renewal Application for Nuclear Power Plants," July 2001.
9.53 U.S. NRC Bulletin 88-09, "Thimble Tube Thinning in Westinghouse Reactors" March 28,1988.
9.54 NEI-03-08, Guideline for the Management of Materials Issues, Revision 1. April 2007 9.55 RG&E Letter to NRC R.E Mecredy to Allen R. Johnson dated Ap. 8, 1993, Response to NRC Bulletin NO 88-09 Thimble Tube thinning in Westinghouse reactors" RGO- 14088.
Page 52 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 9.56 Materials Reliability Program: Inspection Standard for Reactor Internals (MRP-228). EPRI, Palo Alto, CA: 2009. 1016609.
9.57 ASME Boiler & Pressure Vessel Code,Section XI, Division 1, "Rules for In-service Inspection of Nuclear Power Plant Components," American Society of Mechanical Engineers, New York, NY, 2001 Edition, Plus 2003 Addenda, or later.
9.58 Evaluation of Remote Visual Examination Methods. EPRI, Palo Alto, CA: 2006.
1013537.
9.59 Letter to Reactor Internals Focus Group from MRP,
Subject:
Minutes of the Expert Panel Meetings on Expansion Criteria for Reactor Internals I&E Guidelines, MRP 2008-036(via email), June 12, 2008.
9.60 EP-3-S-0719, "Preparation of Aging Management Program Basis Document Guideline," Revision 0.
9.61 WCAP- 17016-NP, "Handbook of Flaw acceptance Criteria for selected R.E.
Ginna Reactor Vessel Internals MRP-227 Primary Components" 9.62 WCAP-17096-NP, "Reactor Internals Acceptance Criteria Methodology and Data Requirements" 9.63 CENG Letter of May 10, 2008 "Fourth Ten Year Interval In-service Inspection Program, Submittal of Relief Request 18,19, 20 and 21" 9.64 WCAP -15036 Rev 1, Determination of Acceptable Baffle-Barrel Bolting For Two Loop Westinghouse Domestic Plants 9.65 WCAP - 7408, Robert E. Ginna Unit No. 1, Internals Assurance Program 9.66 NRC Letter, "Acceptance for Referencing of Generic License renewal Program Topical report entitled "License renewal Evaluation: Aging Management for Reactor Internals" WCAP-14577, Revision 1, Oct. 2000" dated Feb 10, 2001 9.67 Ginna Station, Updated Final Safety Analysis Report (UFSAR) 9.68 WCAP 16881-P - "R.E. Ginna Reactor Vessel Internals Program Scheduling Evaluations." November 2008 9.69 Westinghouse Letter, LTR-RIDA-09-28, R.E. Ginna, Aging Management Inspection Evaluations for Guide Tube Support Pins. June 1, 2009 Page 53 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 9.70 EP-3-S-0719, "Preparation of Aging Management Program Basis Document Guideline, Revision 0, dated February 28, 2007 9.71 Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. A), EPRI Product ID 1022863, December 2011 9.72 Ginna Nuclear Power Plant - Relief Requests 18 -21, Fourth Ten-Year Interval Inservice Inspection Program Submittal of Relief Request Numbers 18, 19, 20, and 21, May 10, 2008 (ML081400722) 9.73 ECP-12-000028 Revision 0, "MRP-227 Plant Specific Evaluation fbr Baffle Former Bolts." February 28, 2012 9.74 ECP-10-000422 Revision 1, "Baffle Former Bolt (BFB) Design Change Modification due to 2011 RFO Scope Change." June 10, 2011 9.75 ECP- 11-000173 Revision 0, "ESR- 11-0091 ESR (000) - Reactor Vessel Lower Guide Tube Support Pin (Split Pin) Replacement." June 6, 2011 9.76 American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Rules for In-service Inspection of Nuclear Power Plant Components, 2004 Edition (no Addenda), American Society of Mechanical Engineers, New York, NY 9.77 BM-2009-000078: Tier III Benchmark at Beznau, Switzerland.
9.78 BM-2010-000115: Tier III Benchmark at Surry Power Station.
9.79 WCAP- 173'54-P, Revision 0, "Determination of Acceptable Baffle-Barrel Bolting for R.E. Ginna. January 2011.
9.80 CR-2011-003963, "ECP-10-00422 - At three locations where existing baffle bolts were removed Areva was unable to install new replacement bolts." Date discovered: 5/19/2011.
9.81 AREVA Technical Justification 51-9162318 Revision 0, "Technical Justification for Angle Beam Hex Probe." December 20, 2011.
9.82 CNG-AM-4.01 Revision 2, "RCS Material Degradation Management Program."
January 14, :2011.
Page 54 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 9.83 IP-IIT-8, "RCS Materials Degradation Management Program," Revision 3. July 16, 2010.
9.84 NEI-03-08, Guideline for the Management of Materials Issues, Revision 2.
January 2010 9.85 NRC Regulatory Issue Summary 2011-07 License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management. July 21, 2011 (ML111990086) 9.86 WCAP-17480, Revision 0, "R.E. Ginna - 14x14 Upper Internals Guide Tube -
Guide Card Wear Evaluation." January 2012.
9.87 Fourth Interval Inservice Inspection (ISI) Program. June 30, 2009.
9.88 AREVA Technical Justification 51-9162304 Revision 0, "Technical Justification for Baffle Bolt Threaded End Volumetric Examination at R.E. Ginna Nuclear Power Plant." June 16, 2011.
9.89 AREVA Technical Justification 51-9153507 Revision 0, "Technical Justification for Replacement Baffle Bolt Volumetric Examination at R.E. Ginna Nuclear Power Plant Unit 1." February 17, 2011.
9.90 WCAP-17367-P, Revision 0, "Constellation Energy R.E. Ginna Replacement CW 316 SS Guide Tube Support Pin Design Equivalency Report."' March 2011.
9.91 AREVA Calculation 32-9161762, Revision 0, "Monte Carlo Simulation of Baffle-to-Former Bolt Failure Patterns for R.E. Ginna." May 2011.
9.92 AREVA Calculation 51-9161677, Revision 0, "Weibull Assessment of Ginna Baffle Bolt Failure Rates." May 2011.
9.93 AREVA Calculation 32-5001743, Revision 8, "W Replacement Baffle Bolt Analysis." April 2011.
9.94 Westinghouse LTR-RIDA-11-129, Revision 0, "Thermal-Hydraulic Evaluations for Abandoned Baffle-Former Bolt Hole Locations at R.E. Ginna." May 17, 2011.
9.95 Westinghouse LTR-RIDA-11-132, Revision 0, "Analysis of Hypothetical Distributions of Failed and Unfailed Baffle-Former Bolts at R.E. Ginna." May 27, 2011.
9.96 AREVA Calculation 32-9161854, Revision 1, "Structural Justification for Abandoned Bolts." May 2011.
Page 55 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 9.97 AREVA Calculation 32-9161446, Revision 1, Thermal Hydraulic Analysis for R.E.
Ginna Nuclear Power Plant Operation with Missing Baffle-to Former Plate Bolts."
May 2011.
9.98 ECP-2009-0123 Revision 0, "Replace Incore Thimble Tubes and Seal Table High Pressure Fittings." August 25, 2010.
9.99 Westinghouse LTR-RIDA-12-149, Revision 0, R.E. Ginna Momentum Flux and Core Bypass Flow Technical Evaluation and Low-Cycle Fatigue Clarification."
September 2012.
Page 56 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT A Table 1 Summary Of Implementing Documents/ Enhancements/ Testinid Initial Inspections/ Actions 1 2 3 4 5 6 7 Discussed in PBD AMR Implementing Document Document Effective Description of Activity / Enhancement AIT Element Activity Number-Title Revision # Step #(s) Number No. and Title Number Section 4.4- N/A EP-VT-I 10 0 TBD Revise EP-VT-1 10 for the 2011 RFO to include NL-2009-Summary augmented inspections required by the LR-RVI- 000021-PROGPLAN Section 5.3.3 001
"_(Complete)
Section 5.2.2 2 Pri-Sched (CH-120) 02900 Program Basis Capture to reflect Pri-Sched Chemistry program is Complete Primary Water required to be in place by the RVI program.
Chemistry Control Program Meets intent of Pressurized Water Reactor Primary Water Chemistry Guidelines Rev. 6. 1014986 December 2007.
N/A Action FATIGUE -PRO N/A N/A Evaluate need for any additional FATIGUE pro NL-2009-Item 2 &7 monitoring that is required. i.e. cycle counting or 000022-tabulation of RX internals fatigue evaluations. 001 Section 7 (Complete)
LRAM-RVI Page 57 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT A Table 1 Summary Of Implementing Documents/ Enhancements/ Testing/ Initial Inspections/ Actions 1 2 3 4 5 6 7 Discussed in PBD AMR Implementing Document Document Effective Description of Activity / Enhancement AIT Element Activity Number-Title Revision # Step #(s) Number No. and Title Number Section 4.2.1 N/A CR-2008-009906 0Complete Westinghouse Letter LTR-RIDA-09-28. Aging NL-2009-Control Rod Management Inspection Evaluations for Guide Tube Split 000023-Guide Tube Split CA-2008-003 832 Pins 001I& 002 pin Replacement (Complete)
Project ODMI For Split Pins 5.6 Acceptance N/A ansion Comnponent 0 N/A Develop Expansion Component Evaluation Criteria with NL-2009-Criteria Evaluation (PWROG) Criteria PWROG in order to disposition recordable indications. 000024-
'001 (Complete) 5.3.3.1.1 N/A Clarify Item 8.2.13 Of the EP- 0000 8.2.13 Clarifv checklist to include Fig-19. Implied in text NL-2009-Lower Core Plate VT- 110 checklist to include discussion but not included in list of figures referenced 000020-figure 19. 001
__ 9.(Complete)
_gure__
4.2.2 Baffle Applicant Establish Acceptable Bolt 0 N/A Develop 60/80 year option for acceptable BFB NL-2009-Former Bolt Action Patterns for 2011 BFB, replacement patterns. 000019-Inspection Item 10 Options. 001 Replacement (Complete)
Project Page 58 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 12 _3 4 5 6 -7 Discussed in PBD AMR ImplementiqngDocument Document Effective Desc 'ipti of Activity/Enhancement AlT Elemn Activity. Numnbr-itle. RevisionH., ~Step #(s) -- Niumbe
-Noand-Title- Number -
N/A N/A LRAM-RVI 0 N/A Update AMR to be consistent with MRP-227 when NL-2009-approved. vv002-001 N/A N/A LR-RVI-PROGPLAN 3 N/A Plan for second MRP-227 exams during the 5T Interval 10 NL-2012-year ISI inspections. 000094-001 N/A N/A LR-RVI-PROGPLAN 3 N/A Determine the need for the inspection of the CRGT Guide NL-2012-Cards prior to the start of the MRP-227-A exams, based 000094-off analysis in WCAP-17480. 003 N/A N/A LR-RVI-PROGPLAN 3 N/A If it is determined that CRGT Guide Card inspections do NL-2012-not need to be completed during the second MRP-227 000094-exams, per NL-2012-000094-003, then seek the 004 appropriate relief from the industry.
N/A N/A LR-RVI-PROGPLAN 3 N/A Inspect CRGT Split Pins following twenty years of in- NL-2012-service life to ensure that assumptions made in WCAP- 000094-17367, that Westinghouse CW 316 Split Pin will maintina 005 integrity for forty years of operation at 100% capacity factor, is maintained.
N/A N/A LV-RVI-PROGPLAN 3 N/A Perform MRP-227-A Inspections prior to 2021. NL-2012-000094-1_ 1 1 1 1 007 Page 59 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT B PBD - PEN&INK Non-Intent Change Form for Attachment A. Table 1 This process is only applicable to the information in Attachment A, Table 1, columns: "4- Document Revision #" and "5- Effective Step #(s) ". See section 6.2.2, Page 60 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT C Ginna Station Aging Management Review Summary Table LRAM-RVI COMPONENT INTENDED MATERIAL ENVIRONMENT AGING EFFECTS PROGRAM/ACTIVITY GROUP FUNCTION REQUIRING MANAGEMENT 1 Lower Core Core Support Flow Stainless Steel Primary Water Cracking due to IASCC
- ASME Section XI Inservice Plate and Fuel Distribution Reduction in Fracture Inspection Program Examination Pins Toughness due to Category B-N-3 Irradiation Embrittlement 9 Reactor Vessel Internals Inspection Program (MRP-227-A Table 4-9)
Loss of Material due to
- ASME Section XI Inservice Wear (Fuel Pins only) Inspection Program Examination Category B-N-3 Cracking due to SCC
- Water Chemistry Control Program Lower Support Core Support Flow Forged Stainless Primary Water Cracking due to SCC
- Water Chemistry Control Program Forging Distribution Steel Lower Support Core Support Stainless Steel Primary Water Cracking due to IASCC
- ASME Section XI Inservice Reduction in Fracture Inspection Program Examination Columns Toughness due to Category B-N-3 Irradiation Embrittlement 9 Reactor Vessel Internals Inspection Program (MRP-227-A Table 4-6)
Cracking due to SCC
- Water Chemistry Control Program Page 61 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT C Ginna Station Aging Management Review Summary Table LRAM-RVI COMPONENT INTENDED MATERIAL ENVIRONMENT AGING EFFECTS PROGRAM/ACTIVITY GROUP FUNCTION REQUIRING MANAGEMENT' Core Barrel and Core Support Stainless Steel Primary Water Cracking due to IASCC
- ASME Section XI Inservice Flange Flow Distribution (Lower core barrel only) Inspection Program Reduction in Fracture Examination Category B-N-3 Toughness due to 0 Reactor Vessel Internals Inspection Irradiation Embrittlement Program (MRP-227-A Tables 4-3 & 4-9)
(Lower core barrel only)
Loss of Material due to
- ASME Section XI Inservice Inspection Program Examination Wear (Flange only) Category B-N-3
- Reactor Vessel Internals Inspection Program (MRP-227-A Tables 4-3 & 4-9)
Cracking due to SCC
- Water Chemistry Control Program Radial Keys and Core Support Stainless Steel Primary Water Loss of Material due to
- ASME Section XI Inservice Clevis Inserts (Radial Keys) Wear Inspection Program Examination Alloy 600 Category B-N-3 (Clevis Inserts) Cracking due to SCC
- Water Chemistry Control Program Page 62 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT C Ginna Station Aging Management Review Summary Table LRAM-RVI COMPONENT INTENDED MATERIAL ENVIRONMENT AGING EFFECTS PROGRAMIACTIVITY GROUP FUNCTION REQUIRING MANAGEMENT' Baffle and Former Core Support Stainless Steel Primary Water Cracking due to IASCC
- ASME Section XI Inservice Assembly Flow Reduction in Fracture Inspection Program Examination Distribution Toughness due to Category B-N-3 Irradiation Embrittlement Cracking due to SCC
- Water Chemistry Control Program Core Barrel Outlet Flow Stainless Steel Primary Water Cracking due to SCC
- Water Chemistry Control Program Distribution Nozzle Secondary Core Core Support Stainless Steel Primary Water Cracking due to SCC
- Water Chemistry Control Program Support Flow Distribution Diffuser Plates Flow Stainless Steel Primary Water Cracking due to SCC e Water Chemistry Control Program Distribution Page 63 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT C Ginna Station Aging Management Review Summary Table LRAM-RVI COMPONENT INTENDED MATERIAL ENVIRONMENT AGING EFFECTS PROGRAMIACTIVITY GROUP FUNCTION REQUIRING MANAGEMENT 1 Upper Support Plate Guide and Stainless Steel Primary Water Cracking due to SCC e Water Chemistry Control Program Assembly Support RCCAs Hold-down Spring Hold-down Stainless Steel Primary Water Loss of Preload due to
- ASME Section XI Inservice Spring Stress Relaxation Inspection Program Loss of Material due to Examination Category B-N-3 Wear Cracking due to SCC
- Water Chemistry Control Program Head/Vessel Core Support Stainless Steel Primary Water Cracking due to SCC e Water Chemistry Control Program Alignment Pins Thermal Shield and Shield Vessel Stainless Steel Primary Water Cracking due to IASCC e ASME Section XI Inservice Neutron Panels (Ginna Inspection Program Neutron Panels - N/A) Reduction in Fracture Examination Category B-N-3 Toughness due to
- Reactor Vessel Internals Inspection Irradiation Embrittlement Program (MRP-227-A Table 4-3)
Cracking due to SCC
- Water Chemistry Control Program Page 64 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT C Ginna Station Aging Management Review Summary Table LRAM-RVI COMPONENT INTENDED MATERIAL ENVIRONMENT AGING EFFECTS PROGRAM/ACTIVITY GROUP FUNCTION REQUIRING MANAGEMENT 1 BMI Columns and Guide and Stainless Steel Primary Water Cracking due to SCC
- Water Chemistry Control Program Support Flux Thimble Tubes Loss of Material due to
- Thimble Tube Inspection Program Instrumentation Wear (Thimbles Only)
Head Cooling Spray Flow Stainless Steel Primary Water Cracking due to SCC
- Water Chemistry Control Program Nozzles Distribution Upper Guide and Stainless Steel Primary Water Cracking due to SCC 9 Water Chemistry Control Program Instrumentation Support Column, Conduit and Thermocouples Supports Bolting (Upper Core Support Stainless Steel Primary Water Loss of Mechanical o ASME Section XI In-service Support Column, Alloy X-750 Closure Inspection Program Guide Tube, Clevis (Clevis Integrity due to Stress Examination Category B-N-3 Insert) Insert Bolts) Relaxation Cracking due to SCC
- Water Chemistry Control Program Page 65 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT C Ginna Station Aginy-Management Review Summary Table LRAM-RVI COMPONENT INTENDED MATERIAL ENVIRONMENT AGING EFFECTS PROGRAM/ACTIVITY GROUP FUNCTION REQUIRING MANAGEMENT' Bolting (Lower Core Support Stainless Steel Primary Water Cracking due to SCC,
- ASME Section XI Inservice Support Column, IASCC and Loss of Inspection Program Baffle/Former, Mechanical Examination Category B-N-3 Barrel/Former) Closure Integrity due to IASCC, Reduction in
- Reactor Vessel Internals Inspection Fracture Toughness, Program (MRP-227-A Tables 4-3 & 4-6)
Irradiation Creep and Stress Relaxation
- Water Chemistry Control Program Page 66 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT D - Table 1.1 Westinghouse Plants Primary Components (MRP-227 Revision 0 Table 4-3)
Item Applicability Effect Expansion Examination Method/Frequency Examination (Mechanism) Link Coverage Control Rod Guide Fig. 3 Loss of Material None Visual (VT-3) examination no later than 2 20% examination of the Tibe AQemhlv- Ginide (Wear) rpeelina g utnage from the hbginnina of mnmhbr nofCRT plates (cards) the license renewal period, and no earlier assemblies, with all guide than two refueling outages prior to the cards within each selected start of the license renewal period. CRGT assembly Subsequent examinations are required on a examined.
ten-year interval.
Control Rod Guide Fig. 3 Cracking (SCC, Bottom- Enhanced visual (EVT-1) examination to 100% of outer (accessible)
Tube Assembly- Fatigue) mounted determine the presence of crack-like CRGT lower flange weld Lower flange welds All instrumentation surface flaws in flange welds no later than surfaces and adjacent base (BMI) column 2 refueling outages from the beginning of metal.
bodies, Lower the license renewal period and subsequent support column examination on a ten-year interval.
bodies (cast)
Core Barrel Fig. 4 Cracking (SCC) Remaining core Periodic enhanced visual (EVT-1) 100% of one side of the Assembly- Upper core barrel welds, examination, no later than 2 refueling accessible surfaces of the barrel flange weld Lower support outages from the beginning of the license selected weld and column bodies renewal period and subsequent adjacent base metal.
(non cast) examination on a ten-year interval.
Page 67 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT D - Table 1.1 WiQtinohnnep Plante Primarv (amnnnpntgtn (M[RP_..7V7 E TIhMP AA*%
Item Applicability Effect Expansion Examination Method/Frequency Examination (Mechanism) Link Coverage Baffle-Former Fig. 7 Cracking (IASCC, None Visual (VT-3 examination, with baseline Bolts and locking devices Assembly Baffle-edge Fatigue) that results examination between 20 and 40 EFPY and on high fluence seams.
bolts in lost or broken- " ....--- examinations on a ten-year 100% of components locking devices, interval. accessible from core side.
failed or missing bolts or protrusion of bolt heads Baffle-Former Fig. 5 Cracking (IASCC, Lower support Baseline volumetric (UT) examination 100% of accessible bolts Assembly Baffle- Fatigue) column bolts, between 25 and 35 EFPY, with subsequent or as supported by plant-former bolts. Barrel-Former examination after 10 to 15 additional EFPY specific justification.
bolts. Heads accessible from the to confirm stability of bolting pattern. Re- Heas accessib ility core side. UT accessibility examination for high-leakage core designs may be affected by requires continuing examinations on a ten- complexity of head and year interval. locking device designs.
Page 68 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012
- . .-
ATTACHMENT D - Table 1.1 Westinhouse Plants Primary Comr onents (MRP-227 Revision 0 Table 4-3)
Item Applicability Effect Expansion Examination Method/Frequency Examination (Mechanism) Link Coverage Baffle-Former Fig. 6 Distortion (Void None Visual (VT-3) examination to check for Core side surface as Assembly Swelling), or evidence of distortion, with baseline indicated.
Assembly Cracking (IASCC) examination between 20 and 40 EFPY and that results in: subsequent examinations on a ten-year
-abnorm al i nt erval .
interaction with fuel interval.
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.
Page 69 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 S -
ATTACHMENT D - Table 1.1 Wdpetinorhnuug Plants Primnur fnmann fe IM1P _77'7 Upwini~n ft ah~ip A-.7 Item Applicability Effect Expansion Examination Method/Frequency Examination (Mechanism) Link Coverage Alignment and N/A Ginna Distortion (Loss of None Direct measurement of spring height within Measurements should be Interfacing Only plants with Load) Note: This three cycles of the beginning of the license taken at several points Components Internals 304 stainless steel mechanism was not renewal period.Iftst set of around the circumference hold down spring hold down springs strictly identified in measurements is not sufficient to determine of the spring, with a the original list of measurements i s e de t statistically adequate Ginna is 410 SS. age related life, spring height measurements must be number of measurements degradation taken during the next two outages, in order at each point to minimize mechanisms [7]. to extrapolate the expected spring height to uncertainty. Replacement 60 years. of 304 springs by 403 springs is required when the spring stiffness is determined to relax beyond design tolerance.
Thermal Shield Fig. 9 & 10 Cracking (Fatigue) None Visual (VT-3) no later than 2 refueling 100% of thermal shield Assembly Thermal or Loss of Material outages from the beginning of the license flexures.
shield flexures (Wear) that results renewal period. Subsequent examinations in thermal shield flexures excessive on a ten-year interval.
wear, fracture, or complete separation Page 70 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012
________________________ . -
ATTACHMENT D - Table 1.2 Westinghouse Plants Expansion Components (MRP-227 Revision 0 Table 4-6)
Item Applicability Effect Expansion Examination Method/Frequency Examination (Mechanism) Link Coverage Core Barrel Assembly Fig. 4 Cracking (IASCC, Baffle-former Volumetric (UT) examination, with initial 100% of accessible bolts.
Barrel-former bolts Fatigue) bolts and subsequent examinations dependent on Accessibility may be bf-former r.s.lts ofALMll_____Dit _ bolt emiaio limited bv presence of IIIthermal shields.
Lower Support Fig. I Cracking (IASCC, Baffle-former Volumetric (UT) examination, with initial 100% of accessible bolts Assembly Fatigue) bolts and subsequent examinations dependent on or as supported by plant-Lower support column results of baffle-former bolt examinations, specific justification.
bolts Core Barrel Assembly Fig. 4 Cracking (IASCC, Upper core Enhanced visual (EVT-1) examination, 100% of one side of the Core barrel flange, core Fatigue) barrel flange with initial examination and re- accessible surfaces of the barrel outlet nozzles, weld examination frequency dependent on the selected weld and adjacent Lower core barrel examination results for upper core barrel base metal.
Lower Support Ginna= A276 Cracking (IASCC) Upper core Enhanced visual (EVT-1) examination with 100% of accessible Assembly type 304 barrel flange initial examination and re-examination support columns.
Lower support column Fig. 1 weld frequency dependent upon the examination bodies (non cast) results for upper core barrel flange weld.
Lower Support N/A Ginna Cracking (IASCC) Control rod Enhanced visual (EVT- 1) examination. 100% of accessible Assembly Lower Including the guide tube support columns.
Support Column bodies detection of (CRGT) lower (cast) fractured support flanges columns Bottom Mounted Fig.1 Cracking (Fatigue) Control rod Visual (VT-3) examination of BMI column 100% of BMI column Instrumentation including the guide tube bodies as indicated by difficulty of bodies for which difficulty System detection of (CRGT) lower insertion/withdrawal of flux thimbles. Flux is detected during flux Bottom-mounted completely flanges thimble insertion/withdrawal to be thimble instrumentation (BMI) fractured column insertion/withdrawal.
column bodies bodies monitored at each inspection interval.
Page 71 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT D - Table 1.3 Westinghouse Plants Existing Programs Components (MRP-227 Revision 0 Table 4-9)
Item Applicability Effect Expansion Examination Method/Frequency Examination (Mechanism) Link Coverage Core Barrel Assembly Figure 4 Loss of material ASME Code Visual (VT-3) examination to determine All accessible surfaces at Core barrel flange (Wear) Section XY l n for. excessive wear. specied*,,,
fre*q-uen-cy Upper Internals Figure 1 Cracking (SCC, ASME Code Visual (VT-3) examination All accessible surfaces at Assembly Fatigue) Section XI specified frequency.
Upper support ring or skirt Lower Internals N/A Ginna is Cracking (SCC, ASME Code Visual (VT-3) examination of the lower All accessible surfaces at Assembly -12' Core Fatigue) Section XI core plates to detect evidence of distortion specified frequency.
Lower core plate XL and/or loss of bolt integrity.
lower core plate (XL= 14ft. Core)
Lower Internals N/A Ginna is Loss of material ASME Code Visual (VT-3) examination All accessible surfaces at Assembly -12' Core (Wear) Section XI specified frequency.
Lower core plate XL lower core plate (XL= 14ft. Core)
Bottom Mounted Fig I Loss of material NUREG-1801 Surface (ET) examination Eddy current surface Instrumentation (Wear) Rev. 1 examination as defined in System plant response to 1EB 88-Flux thimble tubes 09.
Alignment and Fig. 12 Loss of material ASME Code Visual (VT-3) examination All accessible surfaces at Interfacing (Wear) Section XI specified frequency.
Components Clevis insert bolts Alignment and Fig. 11 Loss of material ASME Code Visual (VT-3) examination All accessible surfaces at Interfacing (Wear) Section XI specified frequency.
Components Upper core plate alignment pins Page 72 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT D - Table 1.4 Westinghouse Plants Examination Acceptance and Expansion Criteria (MRP-227 Revision 0 Table 5-3)
I -~
~V Item Applicability Examination Expansion Expansion Criteria Additional Acceptance Links(s) Examination Criteria Acceptance Criteria
_________________(Note 1)
Control Rod Guide Fig 2 & 3. Visual (VT-3) None N/A N/A Tube Assembly Examination Guide plates The specific (cards) relevant condition is wear that could lead to loss of control rod alignment and impede control assembly insertion.
Note 1: The examination acceptance criterion for visual examination is the absence of the specified relevance condition(s).
Page 73 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT D - Table 1.4 Westinghouse Plants Examination Acceptance and Expansion Criteria (MRP-227 Revision 0 Table 5-3)
Item Applicability Examination Expansion Expansion Criteria Additional Acceptance Links(s) Examination Criteria Acceptance Criteria (Note 1)
Control Rod Guide Fig .3 Enhanced visual a. Bottom- a. Confirmation of surface breaking a. For BMI column Tube Assembly (EVT-1) mounted indications in two or more CRGT lower bodies, the specific Lower Flange welds examination instrumentation flange welds, combined with flux thimble relevant condition for the The specific (BMI) column insertion/withdrawal difficulty, shall VT-3 examination is relevant condition is bodies inertio tal diffict sa completely fractured a detectable crack- b. Lower require visual (VT-3) examination of BMI column bodies.
like surface support column column bodies by the completion of the indication, bodies (CAST) next refueling outage. b. For cast lower support column bodies, the
- b. Confirmation of surface breaking specific relevant condition indications in two or more CRGT lower is a detectable crack-like flange welds shall require EVT- I surface indication.
examination of cast lower support column bodies within three fuel cycles following the initial observation.
Page 74 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT D - Table 1.4 Westinghouse Plants Examination Acceptance and Expansion Criteria (MRP-227 Revision 0 Table 5-3)
Item Applicability Examination Expansion Expansion Criteria Additional Acceptance Links(s) Examination Criteria Acceptance Criteria (Note 1)
Core Barrel Assembly Fig. 4 Periodic enhanced a. Remaining The confirmed detection and sizing of a a and b. the specific Upper core barrel visual (EVT-1) core barrel surface-breaking indication with a length relevant condition is a flange weld examination, welds greater than two inches in the upper core detectable crack-like surface indication.
The specific b. lower barrel flange weld shall require that the relevant condition is support column EVT- 1 examination, and any a detectable crack- bodies (non- supplementary UT examination, be like surface cast) expanded to include the core barrel-to-indication. support plate weld by the completion of the next refueling 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 oftheaccessible surfaces of the non cast lower support column bodies within three fuel cycles following the initial observation.
Page 75 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT D - Table 1.4 Westinghouse Plants Examination Acceptance and Expansion Criteria (MRP-227 Revision 0 Table 5-3)
Item Applicability Examination Expansion Expansion Criteria Additional Acceptance Links(s) Examination Criteria Acceptance Criteria (Note 1)
Baffle-Former Fig 5& 6 Visual (VT-3) None N/A N/A Assembly examination.
Baffle-edge bolt The specific relevant conditions are missing or broken locking devices, failed or missing bolts, and protrusion of bolt heads.
Page 76 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT D - Table 1.4 Westinghouse Plants Examination Acceptance and Expansion Criteria (MRP-227 Revision 0 Table 5-3)
Item Applicability Examination Expansion Expansion Criteria Additional Acceptance Links(s) Examination Criteria Acceptance Criteria (Note 1)
Baffle-Former Fig 5&6 Volumetric (UT) a. Lower a. Confirmation that more than 5% of the a and b. The examination Assembly examination. support column baffle-former bolts actually examined on acceptance criteria for the Baffle-former bolts bolts the four baffle plates at the largest distance UT of the lower support The examination column bolts and the The examination ~~from the core (presumed to be the lowest clm ot n h acceptance criteria b. Barrel- dose cons) mednto beptable barrel-former bolts shall for the UT of the former bolts dose locations) contain unacceptable be established as part of baffle-former bolts indications shall require UT examination of the examination technical shall be established the lower support column bolts within the justification.
as part of the next three fuel cycles.
examination technical b. confirmation that more than 5% of the justification. lower support column bolts actually examined contain unacceptable indications shall require UT examination of the barrel-former bolts.
Page 77 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT D - Table 1.4 Westinghouse Plants Examination Acceptance and Expansion Criteria (MRP-227 Revision 0 Table 5-3)
Item Applicability Examination Expansion Expansion Criteria Additional Acceptance Links(s) Examination Criteria Acceptance Criteria (Note 1)
Baffle-Former Visual (VT-3) None N/A N/A Assembly Fig. 5 & 6 examination.
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 78 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT D - Table 1.4 Westinghouse Plants Examination Acceptance and Expansion Criteria (MRP-227 Revision 0 Table 5-3L Item Applicability Examination Expansion Expansion Criteria Additional Acceptance Links(s) Examination Criteria Acceptance Criteria (Note 1)
Alignment and All plants with Direct physical None N/A N/A Interfacing 304 stainless steel measurement or Components hold down springs spring height.
Internals hold down spring NOTE: The examination N/A - Ginna acceptance criterion Spring is 410 SS for this measurement is that the remaining compressible height of the spring shall provide hold-down forces within the plant-specific design tolerance.
N/A Ghina Thermal Shield Fig. 9 Visual (VT-3) None N/A N/A Assembly examination.
Thermal shield flexures The specific relevant conditions for thermal shield flexures are excessive wear, fracture or complete separation.
Page 79 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.1 Westinghouse Plants Primary Components (MRP-227-A Table 4-3)
Examination(NoteMethod/Frequency Examination ADDlicabilitv Itiviechanism) E (Note 1)Link Expansion !) Coverage Control Rod Guide All plants Loss of Material None Visual (VT-3) examination no later than 20% examination of the Tube Assembly (Wear) 2 refueling outages from the beginning number of CRGT Guide plates (cards) of the license renewal period, and no assemblies, with all earlier than two refueling outages prior guide cards within each to the start of the license renewal selected CRGT period. Subsequent examinations are assembly examined.
required on a ten-year interval.
See Figure 4-20 Control Rod Guide All plants Cracking (SCC, Bottom-mounted Enhanced visual (EVT-1) examination 100% of outer Tube Assembly Fatigue) instrumentation to determine the presence of crack-like (accessible) CRGT Lower flange welds Aging (BMI) column surface flaws in flange welds no later lower flange weld Management (IE bodies, than 2 refueling outages from the surfaces and adjacent and TE) Lower support beginning of the license renewal period base metal on the column bodies and subsequent examination on a ten- individual periphery (cast) year interval. CRGT assemblies.
Upper core plate (Note 2)
Lower support forging/casting See Figure 4-21.
Page 80 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.1 Westinghouse Plants Primary Components (MRP-227-A Table 4-3)
Effect Expansion Examination Method/Frequency Examination
_ve (I4 Iý t 1.
IMechanism) Link (Note 1)e 1) Coverage Core Barrel Assembly All plants Cracking (SCC) Lower support Periodic enhanced visual (EVT-1) 100% of one side of the Upper core barrel flange column bodies examination, no later than 2 refueling accessible surfaces of weld (non cast) outages from the beginning of the the selected weld and Core barrel license renewal period and subsequent adjacent base metal outlet nozzle examination on a ten-year interval. (Note 4).
welds See Figure 4-22.
Core Barrel Assembly All plants Cracking (SCC, Upper and Periodic enhanced visual (EVT-1) 100% of one side of the Upper and lower core IASCC, Fatigue) lower core examination, no later than 2 refueling accessible surfaces of barrel cylinder girth welds barrel cylinder outages from the beginning of the the selected weld and axial welds license renewal period and subsequent adjacent base metal examination on a ten-year interval. (Note 4).
See Figure 4-22 Core Barrel Assembly All plants Cracking (SCC, None Periodic enhanced visual (EVT-1) 100% of one side of Lower core barrel flange Fatigue) examination, no later than 2 refueling the accessible weld (Note 5) outages from the beginning of the surfaces of the license renewal period and subsequent selected weld and examination on a ten-year interval, adjacent base metal, (Note 4).
Page 81 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.1 Westinghouse Plants Primary Components (MRP-227-A Table 4-3)
Effect Expansion Examination Method/Frequency Examination Item Applicability (Mechanism) Link (Note 1) (Note 1) Coverage Baffle-Former Assembly All plants with Cracking (IASCC, None Visual (VT-3) examination, with Bolts and locking Baffle-edge bolts baffle-edge Fatigue) that baseline examination between 20 and devices on high bolts results in 40 EFPY and subsequent fluence seams. 100%
- Lost or broken examinations on a ten-year interval, of components locking devices accessible from core
- Failed or side (Note 3).
missing bolts o Protrusion of See Figure 4-23.
bolt heads Aging Management (IE and ISR)
(Note 6)
Baffle-Former Assembly All plants Cracking (IASCC, Lower support Baseline volumetric (UT) examination 100% of accessible Baffle-former bolts Fatigue) column bolts, between 25 and 35 EFPY, with bolts (Note 3). Heads Aging Barrel-former subsequent examination on a ten-year accessible from the Management (IE bolts interval, core side. UT and ISR) accessibility may be (Note 6) affected by complexity of head and locking device designs.
See Figures 4-23 and 4-24.
Page 82 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.1 Westinghouse Plants Primary Components (MRP-227-A Table 4-3)
Expansion Examination Method/Frequency Examination Item Applicability Effect (Mechanism) Link (Note 1) (Note 1) Coverage Baffle-Former All plants Distortion (Void Swelling), None Visual (VT-3) examination to check Core side surface as Assembly or Cracking (IASCC) for evidence of distortion, with indicated.
Assembly that results in baseline examination between 20 (Includes: Baffle e Abnormal and 40 EFPY and subsequent See Figures 4-24, 4-25, plates, baffle interaction with fuel examinations on a ten-year 4-26 and 4-27.
edge bolts and assemblies interval.
indirect effects 9 Gaps along high fluence of void swelling in baffle joint former
- Vertical displacement of plates) baffle plates near high fluence joint o Broken or damaged edge bolt locking systems along high fluence baffle joint Alignment and All plants with Distortion (Loss of Load) None Direct measurement of spring Measurements should Interfacing 304 stainless Note: This mechanism height within three cycles of the be taken at several Components steel hold down was not strictly identified in beginning of the license renewal points around the Internals hold springs the original list of age- period. If the first set of circumference of the down spring related degradation measurements is not sufficient to spring, with a mechanisms [7]. determine life, spring height statistically adequate measurements must be taken number of during the next two outages, in measurements at each order to extrapolate the expected point to minimize spring height to 60 years. uncertainty.
See Figure 4-28.
Page 83 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.1 Westinghouse Plants Primary Components (MRP-227-A Table 4-3)
Expansion Examination Method/Frequency Examination Item Applicability Effect (Mechanism) Link (Note 1) (Note 1) Coverage Thermal Shield All plants with Cracking None Visual (VT-3) no later than 2 100% of thermal shield Assembly thermal shields (Fatigue) refueling outages from the flexures.
Thermal shield or Loss of beginning of the license renewal flexures Material (Wear) period. Subsequent examinations See Figures 4-29 and that results in on a ten-year interval. 4-36.
thermal shield flexures excessive wear, fracture, or complete separation Notes to Table 4-3:
- 1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3.
- 2. Aminimum of 75% of the total identified sample population must be examined.
- 3. Aminimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined for inspection credit.
- 4. A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined from either the inner or outer diameter for inspection credit.
- 5. The lower core barrel flange weld may be alternatively designated as the core barrel-to-support plate weld in some Westinghouse plant designs.
- 6. Void swelling effects on this component is managed through management of void swelling on the entire baffle-former assembly.
Page 84 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.2 Westinghouse Plants Expansion Components (MRP-227-A Table 4-6) sPrimary Link Examination Method/Frequency Examination Item Applicability Effect (Mechanism) (Note 1) (Note 1) Coverage Upper Internals All plants Cracking CRGT lower Enhanced visual (EV'I-1) 100% of accessible Assembly (Fatigue, Wear) flange weld examination. Re-inspection every surfaces (Note 2).
Upper core plate 10 years following initial inspection.
Lower Internals All plants Cracking CRGT lower Enhanced visual (EVT- 100% of accessible Assembly Aging flange weld 1) examination. Re-inspection surfaces (Note 2).
Lower support Management (TE in every 10 years following initial forging or Casting) inspection. See Figure 4-33.
castings Core Barrel All plants Cracking Baffle-former Volumetric (UT) examination. 100% of accessible Assembly (IASCC, Fatigue) bolts Re-inspection every 10 years bolts. Accessibility may Barrel-former bolts Aging Management (IE, following initial inspection, be limited by presence Void Swelling and ISR) of thermal shields or neutron pads (Note 2).
See Figure 4-23.
Lower Support All plants Cracking Baffle-former Volumetric (UT) examination. 100% of accessible Assembly (IASCC, Fatigue) bolts Re-inspection every 10 years bolts or as supported by Lower support Aging following initial inspection. plant-specific column bolts Management (IE and
,ISR) j' s+ificati"o (Note 2).
See Figures 4-32 and 4-33.
Page 85 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.2 Westinghouse Plants Expansion Components (MRP-227-A Table 4-6)
Item Applicability IT Effect (Mechanism) I Primary Link IExamination(Non Method/Frequency
- 1) Examination
((Note 1) Note 1) Coverage Core Barrel All plants Cracking (SCC, Fatigue) upper core Enhanced visual (EVT-i) I00% of one side of the Assembly Aging Management (IE of barrel flange examination. Re-inspection every accessible surfaces of Core barrel outlet lower sections) weld 10 years following initial the selected weld and nozzle inspection. adjacent base metal welds (Note 2).
See Figure 4-22.
Core Barrel All plants Cracking (SCC,IASCC) Upper and Enhanced visual (EVT-1) 100% of one side of the Assembly Aging Management (IE) lower core examination. Re-inspection every accessible surfaces of Upper and lower barrel cylinder 10 years following initial the selected weld and core barrel girth welds inspection. adjacent base metal cylinder axial (Note 2).
welds See Figure 4-22.
Lower Support All plants Cracking Upper core Enhanced visual (EVT-1) 100% of accessible Assembly (IASCC) barrel flange examination. Re-inspection every surfaces (Note 2).
Lower support Aging Management (IE) weld 10 years following initial column bodies inspection. See Figure 4-34.
(non cast)
Lower Support All plants Cracking (IASCC) Control rod Visual (EVT-1) examination. Re- 100% of accessible Assembly including the detection of guide tube inspection every 10 years following support columns (Note Lower support fractured support columns (CRGT) lower initial inspection. 2).
column bodies Aging Management (IE) flanges (cast) See Figure 4-34.
Page 86 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.2 Westinghouse Plants Expansion Components (MRP-227-A Table 4-6)
P Examination Method/Frequency Item Applicability Effect (Mechanism) (Note 1) (Note 1) Coverage Bottom Mounted All plants Cracking (Fatigue) Control rod Visual (VT-3) examination of BMI 100% of BMi column Instrumentation including the detection of guide tube column bodies as indicated by bodies for which System completely fractured (CRGT) lower difficulty of insertion/withdrawal of difficulty is detected Bottom-mounted column bodies flanges flux thimbles. Re-inspection every during flux thimble instrumentation Aging Management (IE) 10 years following initial insertion/withdrawal.
(BMI) column inspection. Flux thimble bodies insertion/withdrawal to be See Figure 4-35.
monitored at each inspection interval.
Notes to Table 4-6:
- 1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3.
- 2. A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).
Page 87 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.3 W nh l ',Io,+n .t1 DDii"? A T".lF A- OI Item Applicability Effect (Mechanism) Reference Examination Method Examination Coverage Core Barrel All plants Loss of material (Wear) ASME Code Visual (VT-3) examination to All accessible surfaces at Assembly Section XI determine general condition for specified frequency.
Core barrel flange excessive wear.
Upper Internals All plants Cracking (SCC, Fatigue) ASME Code Visual (VT-3) examination. All accessible surfaces at Assembly Section XI specified frequency.
Upper support ring or skirt Lower Internals All plants Cracking (IASCC, Fatigue) ASME Code Visual (VT-3) examination of the lower All accessible surfaces at Assembly Aging Management (IE) Section Xl core plates to detect evidence of specified frequency.
Lower core plate distortion and/or loss of bolt integrity.
XL lower core plate (Note 1)
Lower Internals All plants Loss of material (Wear) ASME Code Visual (VT-3) examination. All accessible surfaces at Assembly Section XI specified frequency.
Lower core plate XL lower core plate (Note 1)
Bottom Mounted All plants Loss of material (Wear) NUREG-1801 Surface (ET) examination. Eddy current surface Instrumentation Rev. 1 examination as defined in System plant response to IEB 88-Flux thimble tubes 09.
Alignment and All plants Loss of material (Wear) ASME Code Visual (VT-3) examination. All accessible surfaces at Interfacing (Note 2) Section XI specified frequency.
Components Clevis insert bolts Alignment and All plants Loss of material (Wear) ASME Code Visual (VT-3) examination. All accessible surfaces at Interfacing Section XI specified frequency.
Components Upper core plate alignment pins Notes to Table 4-9:
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.
Page 88 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.4 Westinghouse Plants Examination Acceptance and Expansion Criteria (MRP-227-A Table 5-3)
Applicability Examination Acceptance Examnatin Exansin Expansion Expansion Criteria I Additional Amition Item ItmApiaiiy Criteria (Note 1) Link(s) Examination Acceptance Criteria Control Rod All plants V,.... *k-rj CaminatII IIone N/A N/A Guide The specific relevant Tube Assembly condition is wear that Guide plates could lead to loss of (cards) control rod alignment and impede control assembly insertion.
Control Rod All plants Enhanced visual (EVT- a.Bottom a. Confirmation of surface breaking a. For BMI column Guide 1) examination. The mounted indications in two or more CRGT bodies, the specific Tube Assembly specific relevant condition instrumentation lower flange welds, combined with relevant condition for Lower flange is a detectable crack-like (BMI) column flux thimble insertion/withdrawal the VT-3 examination is welds surface indication, bodies difficulty, shall require visual (VT-3) completely fractured
- b. Lower examination of BMI column bodies column bodies.
support by the completion of the next b. For cast lower column bodies refueling outage. support column bodies, (cast), upper b. Confirmation of surface breaking upper core plate and core indications in two or more CRGT lower support plate and lower lower flange welds shall require forging/castings, support forging EVT-1 examination of cast lower the specific relevant or support column bodies, upper condition is a casting core plate and lower support detectable forging/castings within three crack-like surface fuel cycles following the initial indication.
observation.
Page 89 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.4 Westinghouse Plants Examination Acceptance and Expansion Criteria (MRP-227-A Table 5-3)
Item Applicability Examination Acceptance Expansion I Expansion Criteria Additional Examination Criteria (Note 1) Link(s) Acceptance Criteria Core Barrel All lants- Periodic enhanced visual a. Core barrel a. The confirmned detection and a and b. The specific Assembly (EVT-1) examination. The outlet nozzle sizing of a surface-breaking relevant condition for Upper core barrel specific relevant condition welds indication with a length greater the expansion core flange weld is a detectable crack-like b. Lower than two inches in the upper core barrel outlet nozzle surface indication, support column barrel flange weld shall require that weld and lower support bodies (non the EVT-1 examination be column body cast) expanded to include the core barrel examination is a outlet nozzle welds by the detectable crack-like completion of the next refueling surface indication.
outage.
- b. If extensive cracking in the core barrel outlet nozzle welds is detected, EVT-1 examination shall be expanded to include the upper six inches of the accessible surfaces of the non-cast lower support column bodies within three fuel cycles following the initial observation.
Core Barrel All plants Periodic enhanced visual None None None Assembly (EVT-1) examination.
Lower core barrel flange weld (Note The specific relevant
- 2) condition is a detectable crack-like surface indication.
Page 90 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.4 Westinghouse Plants Examination Acceptance and Expansion Criteria (MRP-227-A Table 5-3)
Item Applicability Examination Acceptance Expansion Expansion Criteria I Additional Amition A Criteria (Note 1) Link(s) Examination Acceptance Criteria Coe-,1us Mil JICILb Periodic enhanced Upper core The confirmred detection and sizing The specific relevant Assembly visual (EVT-1) barrel cylinder of a surface-breaking indication condition for the Upper core barrel examination, axial welds with a length greater than two expansion upper core cylinder girth inches in the upper core barrel barrel cylinder axial welds The specific relevant cylinder girth welds shall require weld examination is a condition is a that the EVT-1 examination be detectable crack-like detectable crack-like expanded to include the upper core surface indication.
surface indication, barrel cylinder axial welds by the completion of the next refueling outage.
Core Barrel All plants Periodic enhanced Lower core The confirmed detection and sizing The specific relevant Assembly visual (EVT-1) barrel cylinder of a surface-breaking indication condition for the Lower core barrel examination, axial welds with a length greater than two expansion lower core cylinder girth inches in the lower core barrel barrel cylinder axial welds The specific relevant cylinder girth welds shall require weld examination is a condition is a that the EVT-1 examination be detectable crack-like detectable crack-like expanded to include the lower core surface indication.
surface indication, barrel cylinder axial welds by the completion of the next refueling outage.
Page 91 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.4 Westinghouse Plants Examination Acceptance and Expansion Criteria (MRP-227-A Table 5-3)
Examination Acceptance Expansion Expansion Criteria Additional Item Applicability Crtra(oe1 Criteria (Note 1) I iks Link(s) xmnto Acceptance Criteria Baffle-Former Ali plants Visual (VT-3) examination. None N/A NIA Assembly with baffle- The specific relevant Baffle-edge bolts edge bolts conditions are missing or broken locking devices, failed or missing bolts, and protrusion of bolt heads.
Baffle-Former All plants Volumetric (UT) a. Lower a. Confirmation that more than 5% a and b. The Assembly examination. The support of the baffle-former bolts actually examination Baffle-former bolts examination acceptance column bolts examined on the four baffle plates acceptance criteria for criteria for the UT of the b. Barrel-former at the largest distance from the the UT of the lower baffle-former bolts shall be bolts core (presumed to be the lowest support column bolts established as part of dose locations) contain and the barrel former the examination technical unacceptable indications shall bolts shall be justification. require UT examination of the established as part of lower support column bolts within the examination the next three fuel cycles, technical justification.
- 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 92 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.4 Westinghouse Plants Examination Acceptance and Expansion Criteria (MRP-227-A Table 5-3)
Examination Acceptance I Expansion Expansion Criteria T Additional Amition Item Applicability Criteria (Note 1) Link(s) Examination Acceptance Criteria Baffle-Former All plants Visual (vT-3) examination. None N/A N/A Assembly The specific relevant Assembly 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 All plants Direct physical None N/A N/A Interfacing with 304 measurement of spring Components stainless height. The examination Internals hold steel hold acceptance criterion for down spring down springs this measurement is that the remaining compressible height of the spring shall provide hold-down forces within the plant-specific design tolerance.
Page 93 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT E - Table 1.4 Westinghouse Plants Examination Acceptance and Expansion Criteria (MRP-227-A Table 5-3)
Examination Acceptance Expansion Expansion Criteria Additional Item Applicability Criteria (Note 1) Link(s) Examination Acceptance Criteria Thermal Shield All plants Visuai (VT-3) examination. None NiA N/A Assembly with thermal The specific relevant Thermal shield shields conditions for thermal flexures shield flexures are excessive wear, fracture, or complete separation.
Notes to Table 5-3:
- 1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).
- 2. The lower core barrel flange weld may alternatively be designated as the core barrel-to-support plate weld in some Westinghouse plant designs.
Page 94 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT F Scope LR-RVI-PROGPLAN Subcomponent Passive Function Aging Management Reference
'r" le 1.1* 1 T ;,,., -K,,M ,-,h . fql/
Table 3.2-1 Line Number (31)
LOWER CORE PLATE CORE SUPPORT Table 3.2-1 Line Number (33)
AND FUEL PINS FLOW DISTRIBUTION Table 3.2-2 Line Number (10)
These apply to both passive functions.
Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (31)
CORE SUPPORT Table 3.2-1 Line Number (33)
LOWER SUPPORT Table 3.2-2 Line Number (7)
FORGING FLOW DISTRIBUTION Table 3.2-2 Line Number (8)
Table 3.2-2 Line Number (9)
These apply to both passive functions.
LOWER SUPPORT Table 3.2-1 Line Number (8)
CORE SUPPORT Table 3.2-1 Line Number (31)
COLUMNS Table 3.2-1 Line Number (33)
Table 3.2-2 Line Number (7)
Table 3.2-1 Line Number (8)
CORE BARREL AND CORE SUPPORT Table 3.2-1 Line Number (31)
FLANGE FLOW DISTRIBUTION Table 3.2-1 Line Number (33)
Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (10)
-Page 95 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT F Scope LR-RVT-PR0GPI.AN Subcomponent Passive Function Aging Management Reference These aIpply , ,... passive functions.
Table 3.2-1 Line Number (8)
RADIAL KEYS AND Table 3.2-1 Line Number (28)
CLEVIS INSERTS CORE SUPPORT Table 3.2-1 Line Number (33)
Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (9)
Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (31)
BAFFLE AND FORMER CORE SUPPORT Table 3.2-1 Line Number (33)
ASSEMBLY FLOW DISTRIBUTION Table 3.2-2 Line Number (7)
These apply to both passive functions.
Table 3.2-1 Line Number (8)
CORE BARREL Table 3.2-1 Line Number (31)
FLOW DISTRIBUTION Table 3.2-1 Line Number (33)
OUTLET NOZZLE Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (8)
Table 3.2-2 Line Number (9)
Page 96 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMdENT F - SCODe LR-RVI-PROGPLAN Subcomponent Passive Function Aging Management Reference SECONDARY CORE CORE SUPPORT Table 3.2-2 Line Number (11) This SUPPORT~tl rFLOW DITRIBUr YTIONK app 1 is bt pass-ver- cion~LIfs.
DIFFUSER PLATES FLOW DISTRIBUTION Table 3.2-2 Line Number (11)
This applies to both passive functions.
UPPER SUPPORT GUIDE AND SUPPORT Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (33)
PLATE ASSEMBLY RCCA'S Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (9)
Table 3.2-1 Line Number (8)
UPPER CORE PLATE Table 3.2-1 Line Number (33)
CORE SUPPORT Table 3.2-2 Line Number (7)
AND FUEL Table 3.2-2 Line Number (9)
ALIGNMENT PINS Table 3.2-2 Line Number (10)
These apply to both passive functions.
UPPER SUPPORT GUIDE AND SUPPORT Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (33)
COLUMNS RCCA'S Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (9)
RCCA GUIDE TUBES GUIDE AND SUPPORT Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (33)
AND FLOW DOWNCOMERS RCCAS Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (9)
Page 97 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT F - Scope LR-RVI-PROGPLAN Subcomponent Passive Function Aging Management Reference GUIDE TUBE GUIDE AND SUPPORT Table 3.2-1 Line Number (8) 1 a v J.Y.-z. JLII*IN MIMIel /I)
SUPPORT PINS RCCAS Table 3.2-2 Line Number (10)
Table 3.2-2 Line Number (11)
Table 3.2-1 Line Number (8)
UPPER CORE PLATE GUIDE AND SUPPORT Table 3.2-1 Line Number (28)
ALIGNMENT PINS RCCAS Table 3.2-1 Line Number (33)
Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (9)
Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (30)
Table 3.2-1 Line Number (33)
HOLD-DOWN SPRING CORE SUPPORT Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (9)
Table 3.2-2 Line Number (10)
Table 3.2-2 Line Number (12)
HEAD/VESSEL CORE SUPPORT Table 3.2-2 Line Number (11)
ALIGNMENT PINS Page 98 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT F - Scope LR-RVI-PROGPLAN BMI COLUMNS AND GUIDE AND SUPPORT Table 3.2-1 Line Number (28)
FLUX THIMBLES INSTRUMENTATION Table 3.2-2 Line Number (11)
HEAD COOLING FLOW DISTRIBUTION Table 3.2-2 Line Number (11)
SPRAY NOZZLES UPPER INSTRUMENTATION GUIDE AND SUPPORT Table 3.2-2 Line Number (11)
COLUMN, CONDUIT THERMOCOUPLES AND SUPPORTS Table 3.2-1 Line Number (8)
BOLTING (UPPER Table 3.2-1 Line Number (30)
SUPPORT COLUMN, Table 3.2-1 Line Number (31)
CORE SUPPORT Table 3.2-1 Line Number (33)
GUIDE TUBE, CLEVIS Table 3.2-1 Line Number (36)
Table 3.2-2 Line Number (7)
INSERT) Table 3.2-2 Line Number (9)
Table 3.2-2 Line Number (12)
Page 99 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT F - Scope.
LR-RVI-PROGPLAN Subcomponent Passive Function Aging Management Reference Table 3.2-1 Line Number (5)
T Table 3.2-1 Line Number (o)
Table 3.2-1 Line Number (12)
SUPPORT COLUMN, Table 3.2-1 Line Number (13)
BAFFLE/FORMER, Table 3.2-1 Line Number (31)
Table 3.2-1 Line Number (33)
BARREL/FORMER) Table 3.2-1 Line Number (36)
Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (12)
Page 100 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT G Fig. 1 Typical Westinghouse Internals Page 101 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT G Figure 2 Guide Card Wear Area Page 102 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT G Figure 3 Control Rod Guide Tube Assembly Lower Flange Weld (Upper)
Lower Flange Weld (Lower)
Page 103 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT G Figure 4 Core Barrel Welds Inspected at Ginna During 2011 RFO Axial Weld (1400):
Upper Core Barrel Flange Weld:
VT-3 Inspection in 2011 RFO per ISI Exam 100% EVT-1 Inspection on OD in 2011 RFO per MRP-227 Rev. 0 Exam Partial EVT-1 Inspection on ID in 2011 RFO 100% VT-3 Inspection on ID in 2011 RFO
% Core Barrel Outlet Nozzle (Typ. 2 Places):
VT-3 Inspection in 2011 RFO per ISI Exam Upper Core Barrel Cylinder Girth Weld:
VT-3 Inspection on OD and ID in 2011 RFO per ISI Exam Lower Barrel Axial Weld (280*):
Not Inspected in 2011 RFO Lower Core Barrel Cylinder Girth Weld:
Not Inspected in 2011 RFO Lower Barrel Axial Weld (80*):
Not Inspected in 2011 RFO Lower Core Barrel Flange Weld:
VT-3 Inspection in 2011 RFO per ISI Exam Page 104 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT G Figure 5 Bolt Locations in a Typical Westin2house Baffle Former Barrel Structure COMER EDGE DM=
BAFFLE 70FGDIE BOLT Page 105 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT G Fieure 6 Section of Baffle Former Assembly Page 106 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT G Figure 7 Void Swelling Induced Distorion in Westing-house Baffle-Former Assembly Page 107 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT G Figure 8 Vertical Displacement of Westinehouse Baffle Plates Caused by Void Swellin!
Page 108 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT G Figure 9 - Thermal Shield Flexure Page 109 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT G Fisaure 10 detail thermal shields flexure Page 110 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT G Figure 11 Upper Internals Assembly (flPICAL)
T~fEB*OCO1PLt COLIM4 (TYW iCAL)
-SUPPORT TUJBE (Ttplci'z UPPER CO~REF~
UPPER FUEL PIN (xrzlcAL)
Page 111 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT G Figure 12 Clevis Insert Details 4Z.4~Q4
. 4. 4 * .: .
AL.741/2 ii 44 1..
44' 444444
-r 4<
44 4
~-1.4 4444 4 .4 ~44 4< 44 4 44 4 4 4 4 4 - 4~44 4 44444 4444 4 4. 44.
REACTOR VESSEL 4 .444 .4 4 .7 CLEIVS A 44. 74 44 4~44 4 4 4.4 ~j4 4-4~. 4 44 .444 4
~44 444 4 44 44 444 44 44 - 4
.4 I 44
.4-1 474\
44 4 -. 4 444 4 4 444444 4 .44 444
.4444 44
.4 4 4 4 44 Page 112 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT H-Ginna's Response to Applicant/Licensee Action Items from Revision 1 to the NRC Safety Evaluation of MRP-227 Rev. 0 Applicant/Licensee Action Item 1:
Applicant/Licensee Action Item 1 required that each applicant/licensee refer to the assumptions regarding plant design and operating history made in the FMECA of MRP-227-A and functionality analysis for reactors of their design which support MRP-227 and describe the process used for determining plant-specific differences in the design of their RVI components or plant operating conditions, which result in different component inspection categories.
Section 4.3.2.4 in this PBD lists the assumptions that MRP-227-A used in creating the inspection guidelines for each plant. Each assumption was verified for applicability to Ginna Station.
In addition to verifying the assumptions made to create MRP-227-A, Ginna Station has also reviewed Table 3-3, "final disposition of Westinghouse internals," of MRP-227-A, which is the result of the Failure Modes, Effects, and Criticality Analyses (FMECA) for Westinghouse internals. There are no RVI components flagged in the prior Ginna plant specific review under the LRAM-RVI (Ref. 9.35) that are not addressed in the FMECA which supports MRP-227-A. As stated previously in this PBD, Ginna has committed to examining each of the designated primary components, unless otherwise specified, as well as each of the expansion components, if necessary. Those components listed as "Existing" in Table 3-3 have been reviewed and verified to be credited as part of an existing program at Ginna Station. The C-tubes and sheaths of the Control Rod Guide Tube Assembly, as well as various BMI components are listed as "No Additional Measures" in MRP-227-A Table 3-3.
Per foot note 2 in Table 3-3 of MRP-227-A, the C-tubes and sheaths were placed in the "No Additional Measures" group, because decisions on remediation of wear and degradation of these components are to be based on the results of the Control Rod Guide Tube Guide Card inspections. As stated in section 5.3.3.2.1 of this PBD, the condition of the Ginna guide cards was evaluated as having low wear. It was determined that significant wear will not be achieved for another 35 - 41 Effective Full Power Years for the most limiting guide card. For this reason, it is concluded that no additional measures need to be taken for the C-tubes and sheaths.
As described in section 4.1.3 of this PBD, the Flux Thimble Tubes were replaced during the 2011 RFO. Furthermore, Ginna has a program dedicated to the BMI's that ensures their integrity is maintained.
Page 113 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT H -
Ginna's Response to Applicant/Licensee Action Items from Revision 1 to the NRC Safety Evaluation of MRP-227 Rev. 0 Although Table 3-3 groups the Guide Tube Support Pins (Split Pins) into the "Existing" category, section 4.4.3 of MRP-227-A states that the guidance for the split pins is limited to plant specific recommendations and the owner should review their specific design and follow supplier recommendations. As discussed in sections 4.2.1 and 5.10.1 of this PBD, Ginna has replaced all of its split pins during the 2011 RFO with improved material that is more resistant to SCC.
CENG and Ginna will continue to monitor RVI operational experience and future MRP changes for additional plant-specific RVI AMP guidance.
Applicant/Licensee Action Item 2:
Applicant/Licensee Action Item 2 required that each applicant/licensee identify which RVI components are within the scope of LR for its facility. Specifically, a review of the information in Table 4-4 of MRP- 191, "Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design," was required for Westinghouse plants to identify whether these tables contain all of the RVI components that are within the scope of LR for their facilities in accordance with 10 CFR 54.4.
A review of Table 4-4 of MRP- 191 against the RVI components that are in the LR scope for Ginna has been performed. It has been determined that all components listed in Ginna's LR scope are listed in Table 4-4 of MRP-191.
Applicant/Licensee Action Item 3:
Applicant/Licensee Action Item 3 required that each applicant/licensee perform a plant specific analysis to justify the acceptability of the existing guide tube support pin (split pin) program at Westinghouse plants.
As previously discussed in sections 4.2.1 and 6.0 of this PBD, Ginna replaced its split pins during the 2011 RFO. Therefore, this action has been fulfilled through Ginna having a proactive approach in replacing potentially age related degraded components. The design life of the replaced split pins is 40 years. Therefore, no further inspection is required for the period of extended operation.
Page 114 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT H -
Ginna's Response to Applicant/Licensee Action Items from Revision 1 to the NRC Safety Evaluation of MRP-227 Rev. 0 Applicant/Licensee Action Item 4:
Applicant/Licensee Action Item 4 required that each B&W applicant/licensee confirm that the core support structure upper flange weld was stress relieved during the original fabrication of the Reactor Pressure Vessel.
Ginna has a Westinghouse designed Reactor Pressure Vessel. Therefore this Action Item does not apply to Ginna.
Applicant/Licensee Action Item 5:
Applicant/Licensee Action Item 5 required that each applicant/licensee identify plant-specific acceptance criteria to be applied when performing the physical measurements required by the NRC-approved version of MRP-227 for loss of compressibility for Westinghouse hold down springs.
The applicability to inspect the internals hold down spring in Table 4-3 of MRP-227-A is for hold down springs of material 304 Stainless Steel. Ginna's hold down spring is made of 410 Stainless Steel. Therefore, this Action Item does not apply to Ginna.
Applicant/Licensee Action Item 6:
Applicant/Licensee Action Item 6 required that each B&W applicant/licensee justify the acceptability of specified inaccessible components for continued operation through the period of extended operation.
Ginna has a Westinghouse designed Reactor Pressure Vessel. Therefore this Action Item does not apply to Ginna.
Applicant/Licensee Action Item 7:
Applicant/Licensee Action Item 7 required that each applicant/licensee of Westinghouse reactors are required to develop plant-specific analyses to be applied for their facilities to demonstrate that Westinghouse lower support column bodies will maintain their Page 115 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT H-Ginna's Response to Applicant/Licensee Action Items from Revision 1 to the NRC Safety Evaluation of MRP-227 Rev. 0 functionality during the period of extended operation or for additional RVI components that may be fabricated from CASS, martensitic stainless steel or precipitation hardened stainless steel materials.
The Ginna lower support column bodies are fabricated from ASTM A276 Type 304 Stainless Steel that was hot rolled, annealed, and pickled. Since this material is not CASS, martensitic stainless steel, nor precipitation hardened stainless steel it does not require further plant specific analysis.
Additional RVI components that are within the scope of LR, as reviewed in accordance with applicant/licensee Action Item 2, are not CASS and are either inspected per MRP-227-A examinations or are credited through other programs.
Applicant/Licensee Action Item 8:
Applicant/Licensee Action Item 8 required that each applicant/licensee make a submittal for NRC review and approval to credit their implementations of MRP-227, as amended by this SE, as an AMP for the RVI components at their facility.
This AMP for the Ginna reactor internals demonstrates that the program adequately manages the effects of aging for reactor internals components and establishes the basis for providing reasonable assurance that the internals components will continue to perform their intended function through the Ginna license renewal period of extended operation.
The Ginna License Renewal Program is committed to Revision 0 ofNUREG-1801, "Generic Aging Lessons Learned (GALL) Report." The Ginna Station Reactor Internals AMP complies with each of the ten elements in Section XI.M. 16 of this NUREG, as shown in Section :5 of this Program Basis Document (PBD).
The development and submittal of Revision 2 of this PBD to the NRC met a license renewal commitment to work with the industry to develop a reactor vessel internals aging management program. Revision 3 of this PBD addresses Regulatory Issue Summary (RIS) 2011-07. Per RIS 2011-07, Ginna is classified as a Category A plant. As such, Ginna retracted the formerly submitted Revision 2 of this PBD from the NRC, revised it to include the inspection results performed under MRP-227 Revision 0 during the Ginna 2011 RFO, and included Ginna's future commitment to the requirements of MRP-227-A.
Justifications for each gap between MRP-227 Revision 0 and MRP-227-A guidelines are also documented in this PBD. Revision 3 of this PBD is being submitted to the NRC Page 116 of 117
GINNA STATION LR-RVI-PROGPLAN LICENSE RENEWAL PROJECT Revision 3 REACTOR VESSEL INTERNALS PROGRAM Date: 9/18/2012 ATTACHMENT H -
Ginna's Response to Applicant/Licensee Action Items from Revision 1 to the NRC Safety Evaluation of MRP-227 Rev. 0 prior to October 1., 2012 and satisfies this Action Item. The Ginna Station UFSAR will be revised to reflect the submittal to the NRC (NL-2012-000094-006).
Page 117 of 117