ML100050034

R.E. Ginna - Fifth Ten-Year Inservice Inspection Plan and Request for Approval of Alternative (Relief Request) to American Society of Mechanical Engineers Code,Section XI

ML100050034
Person / Time
Site:	Ginna
Issue date:	12/30/2009
From:	Swift P Constellation Energy Group, Ginna
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
WPLNRC-1002245
Download: ML100050034 (124)

Text

Paul Swift Manager, Nuclear Engineering Services CENG a joint vehture of 0%Constelat11on (1-zeDF Oj Enezg R.E. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, New York 14519-9364 585.771.5208 585.771.3392 Fax Paul.swift@cengllc.com December 30, 2009 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

SUBJECT:

Document Control Desk R.E. Ginna Nuclear Power Plant Docket No. 50-244 Fifth Ten-Year Inservice Inspection Plan and Request for Approval of Alternative (Relief Request) to American Society of Mechanical Engineers Code,Section XI (a) Letter from Mr. T. L. Tate (NRC) to Mr. M. K. Nazar (Indiana Michigan Power Company), dated September 28, 2007, "Donald C. Cook Nuclear Plant, Units 1 and 2 -Risk-Informed Safety-Based Inservice Inspection Program for Class 1 and 2 Piping Welds"

REFERENCES:

Pursuant to 10 CFR 50.55a(g)(5)(i), R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC), the licensee for the R.E. Ginna Nuclear Power Plant (Ginna), submits its proposed Fifth Ten-Year Inservice Inspection (ISI) Plan. This ISI plan begins January 1, 2010 and will end on December 31, 2019. A copy of the Ginna ISI plan is provided in Attachment (1).In addition to the ISI plan submittal, one necessary request for relief from the requirements of American Society of Mechanical Engineers (ASME) Code,Section XI is submitted in Attachment (2) for your review and approval.

The relief request is submitted under the provision of 10 CFR 50.55a(a)(3)(i) as an alternative that provides an acceptable level of quality and safety. This is a request for approval for implementation of a risk-informed/safety based inservice inspection program for Class 1 and 2 piping based on ASME Code Case N-716. The relief request is similar to the relief request approved for Donald C. Cook Nuclear Plant in Reference (a).Ginna LLC requests approval of this relief request prior to July 1, 2010 to allow for appropriate planning for the 2011 refueling outage. This relief request is for the duration of the Fifth Ten-Year Interval ISI.Should you have questions regarding this matter, please contact Mr. Thomas L. Harding at (585) 771-5219.-r- _/NiW r_

Document Control Desk December 30, 2009 Page 2 Very truly yours, Paul M. Swift Attachments:

(1)(2)Fifth Ten-Year Inservice Inspection Plan ASME Code,Section XI Relief Request -ISI-01 cc: D. V. Pickett, NRC S. J. Collins, NRC Resident Inspector, NRC ATTACHMENT (1)FIFTH TEN-YEAR INSERVICE INSPECTION PLAN R.E. Ginna Nuclear Power Plant December 30, 2009 OVIE'HVI E WNS"ECTVON EA t Imm, t. JR-Aut, b at.-MD.2.C~ / .~ , Pf on i r;q gal, I fýj k-: Fr"k tF ftj ir; E-PiginEx.rin Priw -5 AV~ br~lit;-see,$7 R.

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 L L Fifth Ten-Year Inspection Interval a joint venture of Rev. 00 Comt eDo Date: December 23, 2009eFInservice Inspection Plan Page 2 of 87 1.0 ASME SECTION XI INSERVICE INSPECTION PLAN 1.1 Purpose This document details the basis and plans for the Fifth Ten-Year Inservice Inspection Interval components, welds, supports, bolting, pump casings, valve bodies and reactor vessel for the R. E. Ginna Power Plant.1.2 General The Ginna Nuclear Power Plant Fifth Interval Inservice Inspection Plan is established in accordance with Title 10 Code of Federal Regulations Part 50 Section 55a (10CFR50.55a).

This plan has been developed to comply with the American Society of Mechanical Engineers (ASME) Section Xl, Rules for Inservice Inspection of Nuclear Power Plant Components and implements the requirements of UFSAR Section 5.2.4.1, Inservice Inspection Program.Ginna received its Construction Permit April 25, 1966. The Nuclear Steam Supply System (NSSS) was furnished by Westinghouse Electric Corporation.

The Architect Engineer for Ginna during construction was Bechtel Corporation.

Nuclear Regulatory Commission (NRC) Regulatory Guide 1.26 and 10CFR50.2 is used to determine the ASME Code classification for Class 1, 2 and 3 Systems, Structures, and Components.

1.3 Inspection Intervals The Operating License for Ginna, Docket No. 50-244, was issued on September 19, 1969. The commercial operation date for Ginna was June 1, 1970. The First Ten-Year inspection interval began for Class 1 on January 1, 1970 and for Class 2 and 3 on May 1, 1973. This first interval for Class 1, 2, and 3 ended December 31, 1979, The Second Ten-Year Inspection Interval concluded on December 31, 1989. The Third Inspection Interval for Ginna ended December 31, 1999. The Fourth Inspection Interval was completed December 31, 2009. The Fifth Inspection Interval is divided into inspection periods consistent with Table IWB-2412-1.

The inspection periods are scheduled as follows: 1st Period: From January 1, 2010 to December 31, 2012 (3 Years)2nd Period: From January 1, 2013 to December 31, 2016 (4 Years)3rd Period: From January 1, 2017 to December 31, 2019 (3 Years)In accordance with ASME Section Xl, IWA-2430, that portion of an inspection interval described as an inspection period may be decreased or extended by as much as 1 year to enable inspections to coincide with a plant refueling outage. However, the adjustments

,EIN R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a o Fifth Ten-Year Inspection Interval Rev. 00ven -eDof Date: December 23, 2009.0& -", Inservice Inspection Plan Page 3 of 87 must not cause successive intervals to be altered by more than 1 year. Outages during the fifth inservice inspection interval are scheduled as follows: Ginna'ula

'ae ~Ki~1 st Period 2 nd Period 3 rd Period Refueling Outages Refueling Outages Refueling Outages 2011 2012 2014 2015 2017 2018 1.4 ASME Section Xl Code of Record for the Fifth Inservice InSpection Interval 10CFR50.55a requires that inservice inspection' of components and system pressure tests conducted during successive 120 month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference, subject to the limitations and modifications of 10CFR 50.55a(b)(2), 12 months prior to the start of the 120 month inspection interval.

On December 31, 2008, the latest edition of the ASME Code Section Xl accepted by the NRC was the 2004 Edition with no Addenda.The ISI Plan for ASME Class 1, 2, and 3 components for the Fifth Inspection Interval has been developed using the ASME Code, Section Xl, 2004 Edition, No Addenda, including the limitations and modifications identified below, hereafter referred to as the Code.1.4.1 10CFR50.55a(b)(2)(xxii), use of the provisions contained in IWA-2220 of the ASME Section XI, 2004 Edition that allow the use of ultrasonic examination as a surface method is prohibited.

1.4.2 10CFR50.55a(b)(2)(xxiv), the use of Appendix VIII and the supplements to Appendix VIII, and Appendix 1 Article 1-3000 of ASME Section Xl 2001 Edition will be used. In addition, paragraphs 10CFR50.55a(b)(2)(xiv, xv and xvi) are required.Note: The limitations of 1.4.2 do not apply when implementing Code Case N-729-1. See Section 5.4 for details.

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval Rev. 00 a joint venture ofRe .O Eneg.sy!'

.'.eDF. Date: December 23, 2009 Inservice Inspection Plan Page 4 of 87 1.4.3 10CFR50.55a(b)(2)(xviii)(C), when qualifying visual examination personnel for VT-3 visual examinations under paragraph IWA-2317 of the ASME Section Xl 2004 Edition, the proficiency of training must be demonstrated by administering an initial qualification examination and administering subsequent re-examinations on a 3-year interval.1.4.4 10CFR50.55a(b)(2)(xviii)(A), level I and II nondestructive examination personnel shall be recertified on a 3 year interval in lieu of the 5 year interval in IWA-2314(a) and IWA-2314(b) of the ASME Section Xl 2004 Edition.1.4.5 1 OCFR50.55a(b)(2)(xix), the provisibns for substitution of alternative examination methods, a combination of methods, or n evly developed techniques in the ASME Section Xl 1997 Addenda of IWA-2240 must be applied. The provisions in IWA-2240 of the ASME Section Xl 2004 Edition are not approved for use.1.4.6 10CFR50.55a(b)(2)(xxi)(A), the provisions of Table IWB-2500-1, Examination Category B-D, Full Penetration Welded Nozzles in Vessels, Items B3.120 and B3.140 in the 1998 Edition of ASME Section XI must be applied. A visual examination with enhanced magnification that has a resolution sensitivity to detect a 1-mil width wire or crack, utilizing the allowable flaw length in Table IWB-3512-1, ASME Section Xl, 2004 Edition with a limiting assumption on the flaw aspect ratio (i.e., a/l=0.5) may be performed instead of an ultrasonic examination.

1.4.7 10CFR50.55a(b)(2)(xxi)(B), the require'ments of Table IWB-2500-1, Examination Category B-G-2, Item B7.80 of the ASME Section XI 1995 Edition is applicable to reused CRD bolting. Ginna does not have CRD pressure retaining bolting.1.5 Ten-Year Inservice Inspection (ISI) Plan Description The Ten-Year Inservice Inspection (ISI) Plan details Ginna's compliance with ASME Code, Articles IWA, IWB, IWC, IWD, and IWF for examination of Class 1, Class 2, and Class 3 welds and supports.

This document defines the Class 1, 2, and 3, components and the Code required examinations for each ASME Section Xl examination category, and the augmented inspection scope.The performance of examinations and tests to be performed during Ginna's fifth interval are contained in the implementation schedule of CENG-GNPP-ISI-005-10.

Requests for Alternatives and Requests for Relief to the NRC and Safety Evaluation Reports are identified in Section 3.0.

OENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval Rev. 00 a joint venture of .e .0 0 raelan'"e-F Date: December 23, 2009 I Inservice Inspection Plan Page 5 of 87 1.6 Additional Programs The following programs are outside the scope of this plan. They are addressed in separate documents." The ASME Section Xl System Pressure Test Program* The ASME Section Xl Containment Inspection Program" The ASME Section Xl Repair / Replacement Program" Snubber Examination and Testing Program" Steam Generator Tube Surveillance Program 2.0 ASME CODE CASE APPLICABILITY This section contains ASME Code Cases applicable to the Fifth Ginna Inservice Inspection Interval.2.1 Adoption of Code Cases All Code Cases adopted for ASME Section Xl activities for use during the Fifth Interval are listed in Tables 2.2, 2.3, and 2.4. The use of Code Cases is in accordance with ASME Section Xl, IWA-2440, 10 CFR 50.55a, and Regulatory Guide 1.147. As permitted by ASME Section Xl and Regulatory Guide 1.147 or 10 CFR 50.55a, ASME Section Xl Code Cases may be adopted and used as described below: 2.1.1 Adoption of Code Cases Listed for Generic Use in Regulatory Guide 1.147 Code Cases that are listed for generic use in the latest revision of Regulatory Guide 1.147 may be included in the ISI program provided any additional provisions specified in the Regulatory Guide are also incorporated.

Table 2.2 identifies those Code Cases approved for generic use.2.1.2 Adoption of Code Cases Not Approved in Regulatory Guide 1.147 Certain Code Cases that have been app-6ved by the ASME Board of Nuclear Codes and Standards may not have been reviewed and approved by the NRC Staff for generic use and listed in Regulatory Guide 1.147. Use of such Code Cases may be requested in the form of a "Request for Alternative" in accordance with 10 CFR 50.55a(a)(3).

Once approved, these Requests for Alternatives will be available for use until such time that the Code Cases are adopted into Regulatory CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint venture of Fifth Ten-Year Inspection Interval Rev. 00-eDF Date: December 23, 2009 Inservice Inspection Plan Page 6 of 87 Guide 1.147, at which time compliance with the provisions contained in the Regulatory Guide is required.Table 2.3 identifies those Code Cases that have been requested through Requests for Alternatives.

These Requests for Alternatives are identified in Section 3 with their correspondence numbers to facilitate retrieval.

2.1.3 Adoption of Code Cases Mandated by 10 CFR 50.55a The NRC may require the licensee to follow an augmented inservice inspection program for systems and components for which the Commission deems that added assurance of structural reliability is necessary.

Many times these"Augmented Requirements" will be contained in Code Cases that ASME has approved.

The NRC may mandate its use and add conditions it believes are necessary via 10 CFR 50.55a. Table 2.4 identifies those Code Cases mandated by 10 CFR50.55a.

2.1.4 Use of Annulled Code Cases As permitted by Regulatory Guide 1.147, Code Cases that have been adopted for use in the current interval that are subsequently annulled by ASME, may be used for the remainder of the interval.2.1.5 Code Case Revisions As permitted by Regulatory Guide 1.147, activities performed to a specific revision of an approved 'code case need not be changed when a subsequent revision of the code case is listed as the approved version in the Regulatory Guide. An exception to this provision would be the inclusion of a limitation or condition on the use of the code case which is necessary to enhance safety.2.1.6 Adoption of Code Cases Issued Subsequent to Filing the Inservice Inspection Plan Code Cases issued by ASME subsequent to filing the Inservice Inspection Plan with the NRC may be incorporated within the provisions of paragraphs 2.1.1 or 2.1.2 by revision to this ISI Inspection Plan. Any subsequent Code Cases shall be added to this inspection plan at the next document revision.

C EN G R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 ajointventu Fifth Ten-Year Inspection Interval Rev. 00 onste eDFDate: December 23, 2009 4%" 1eD Inservice Inspection Plan Page 7 of 87 2.1.7 Non Inservice Inspection Code Cases Code Cases applicable to System Pressure Testing, Containment Inservice Inspection and Repair/Replacement Activities are addressed in their respective programs.2.1.8 Code Cases not approved for use by the NRC Certain Code Cases that have been approved by the ASME Board of Nuclear Codes and Standards may not be approved by the NRC Staff for generic use and listed in Regulatory Guide 1.193, ASME Code Cases Not Approved for Use.However, the NRC may approve their use in specific cases. Code Cases listed in the Regulatory Guide will not be used at Ginna without an approved Request for Alternative in accordance with 10 CFR 50.55a(a)(3).

2.2 ReQulatory Guide 1.147, Revision 15 Approved Code Cases I I I I I.Cod.. Adp rbm- Re at GuideI 14'Code .~~Cse Title-' NRC Limitations Alternative Examination N-460 Coverage for Class I and 2 None Welds Alternative Requirements for N-526 Successive Inspections of Class None I and 2 Vessels Repair/Replacement Activity N-532-4 Documentation Requirements None and Inservice Summary Report I Preparation and Submission CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval a joint venture of Rev. 00 C '-neDFo Date: December 23, 2009.0%. 1 Inservice Inspection Plan Page 8 of 87 27Tb~e22 roe ase Ad~pt~dfrm Regulatory Guide'1 147~Title ~~NRC fiain To achieve consistency with the 10 CFR 50.55a rule change published September 22, 1999 (64 FR 51370), incorporating Appendix VIII,"Performance Demonstration for Ultrasonic Examination Systems," to Section Xl, add the following to the Alternative Methods -specimen requirements:

N-552 Qualification for Nozzle Inside Radius Section from the Outside "At least 50 percent of the flaws in the Surface demonstration test set must be cracks and the maximum misorientation must be demonstrated with cracks. Flaws in nozzles with bore diameters equal to or less than 4 inches may be notches." Add to detection criteria, "The number of false calls must not exceed three." Alternative Additional N-586-1 Examination Requirements for None Classes 1, 2, and 3 Piping, Components, and Supports Alternative Examination Requirements for Steam None N-593 Generator Nozzle to Vessel Welds Ultrasonic Examination of Penetration Nozzles in Vessels, N-613-1 Examination Category B-D, Item None Nos. B3.10 and B3.90, Reactor Nozzle-to-Vessel Welds, Figs.IWB-2500-7(a), (b), and (c)N-624 Successive Inspections None r1~IC ~R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 ,Fifth Ten-Year Inspection Interval a joint venture of Rev. 00 Consteflaio r " eDate: December 23, 2009 Inservice Inspection Plan Page 9 of 87 Table 2.2 -p0de Cases Adopted fror, Regulatory GuOide 1:.147-.'Cae Title ? NRC Limit~ations~

Chemical ranges of the calibration block may vary from the materials specification if (1) it is within the chemical range of the component N-639 Alternative Calibration Block specification to be inspected, and (2)3 Material the phase and grain shape are maintained in the same ranges produced by the thermal process required by the material specification.

In place of a UT examination, licensees may perform a visual examination with enhanced magnification that has a resolution sensitivity to detect a 1--mil width wire or crack, utilizing the allowable flaw length criteria of Table Alternative Requirements for IWB-3512-1 with limiting assumptions N-648-1 Inner Radius Examination of on the flaw aspect ratio. The Class 1 Reactor Vessel Nozzles provisions of Table IWB-2500-1, Examination Category B-D, continue to apply except that, in place of examination volumes, the surfaces to be examined are the external surfaces shown in the figures applicable to this table (the external surface is from point M to point N in the figure).Alternative Examination Requirements of Table IWB-2500-1 and Table IWC-2500-1 for PWR Stainless Steel Residual and Regenerative Heat I Exchangers C ENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 aFifth Ten-Year Inspection Interval a joint venture of -Rev. 00 December 23, 2009 I Inservice Inspection Plan Page 10 of 87 2.3 Code Cases Approved Through Request for Alternatives The following ASME Code Cases are not contained in Regulatory Guide 1.147, Revision 15 and require a Request for Alternative prior to implementation.

See Section 3.0 of this plan for the applicable requests.il N-716 _Alternative i-'Iping ulassitication ana Lxamination I Requirements ISI-01 I 2.4 Code Cases Adopted Via 10 CFR 50.55a The following ASME Code Cases are not contained in Regulatory Guide 1.147, Revision 15, but are mandated in 10 CFR 50.55a as augmented requirements.

N-722 Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 Materials Implemented in the Pressure Test ProQram Conditions specified Alternative Examination Requirements for PWR in paragraphs N-729-1 Reactor Vessel Upper Heads With Nozzles Having (g)(6)(ii)(D)(2)

Pressure-Retaining Partial-Penetration Welds through (6) of 10 CFR 50.55a CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval Rev. 00 a joint venture of/ e .0 Conselltio Date: December 23, 2009.01 Inservice Inspection Plan Page 11 of 87 3.0 RELIEF REQUESTS The ASME Section Xl Code was written to provide requirements for defense in-depth inspections for the nuclear industry.

However,'

not all requirements are applicable or possible to be performed at every plant. Therefore, Constellation has reviewed the Code requirements and determined where those requirements would not be viable at the Ginna Nuclear Plant. 10 CFR 50.55a provides three options to submit these determinations to the staff for review and approval.3.1 Request For Alternatives that Provide an Acceptable Level of Quality and Safety 10 CFR 50.55a(a)(3)(i) allows alternatives to 10 CFR 50.55a(g), when authorized by the NRC, if the proposed alternatives would provide an acceptable level ofquality and safety.In cases where Constellation proposes alternatives to the ASME Section Xl requirements that would provide an acceptable level of quality and safety, a Request for Alternative, as allowed by 1OCFR50.55a(a)(3)(i) will be submitted to the NRC.3.2 Request For Alternatives Required Due to Burden 10 CFR 50.55a(a)(3)(ii) allows alternatives to 10 CFR 50.55a(g), 'when authorized by the NRC, if compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. In cases where Constellation proposes alternatives to ASME SectionXI when compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, a Request for Alternative as allowed by 10CFR50.55a(a)(3)(ii), will be submitted to the NRC.3.3 Relief Request Required due to Impracticality or Limited Examinations 10 CFR 50.55a(g)(5)(iii) allows relief to be submitted in cases where the ASME Section Xl requirements for inservice inspection are considered impractical.

In cases where the ASME Section Xl requirements for inservice inspection are considered impractical, Constellation will' notify the NRC and submit information to support the determination, as required by 10 CFR50.55a(g)(5)(iii).

The submittal of this information will be referred to as a Request for Relief. Per 10 CFR50.55a paragraph (g)(6)(i), the Director of the Office of Nuclear Reactor Regulation will evaluate Requests for Relief per Paragraph (g)(5) and"...may grant such relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility".

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 aE G jnFifth Ten-Year Inspection Interval ajo int venture of.] Rev. 00 Q.'.n"eDFI Date: December 23, 2009 Inservice Inspection Plan Page 12 of 87 In the event that the entire examination volume or surface (as defined in the ASME Code)cannot be examined due to interference by another component or part geometry, then, in accordance with Code Case N-460, a reduction in examination volume or area is acceptable if the reduction is less than 10%. In the event that the reduction in examination volume or area is 10% or greater, a request for relief will be submitted.

NRC Information Notice 98-42 provides additional guidance that all ASME Section XI examinations should meet the examination coverage criteria established in Code Case N-460. Therefore, the guidance included in NRC Information Notice 98-42 will be followed by Constellation when determining whether to prepare a Relief Request or apply the criteria of Code Case N-460 for examinations where less than 100% coverage of any Section Xl examination is obtained.This does not apply to the bare metal visual examination required by Code'Case N-729-1. When performing visual examinations to meet N-729-1, the requirements of N-729-1 shall be used for determining allowance for partial examination of the required surface.3.4 Table of Requests Table 3.4 contains an index of Requests For Alternatives and Requests For Relief written in accordance with 10 CFR50.55a (a)(3) and (g)(5). The applicable Constellation submittal and NRC Safety Evaluation Report (SER) correspondence numbers are also included in Table 3.4 for each Request for Alternative and request for relief.Note that only Requests for Alternatives or Requests for Relief applicable to Inservice Inspection and nondestructive examination requirements for Class 1, 2, and 3 components and component supports are addressed in Table 3.4. Requests for Alternatives or Requests for Relief applicable to System Pressure Testing, Containment Inservice Inspection and Repair/Replacement Activities are addressed in their respective programs.

C ENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval Rev. 0 a joint venture ofRe .0 onsidiatin

• ,l.eDF Date: December 23, 2009 Inservice Inspection Plan Page 13 of 87 1"Ginna, Fifth Interval .,Relief Requests,"Relief ReOjest"i., Relief Request Descrition Mod i 3/4 ' - -I.ISI-01 Request to Use Code Case N-716, Alternative Piping Classification and Examination Requirements Submitted NRC SER TBD N/A CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval a joint ventureof Rev. 00 c'1eFI Date: December 23, 2009 En-o* Inservice Inspection Plan Page 14 of 87 4.0 TEN-YEAR INSERVICE INSPECTION PLAN AND SCHEDULE 4.1 General The Ten-Year Inservice Inspection Plan and Implementing Schedule details Constellation Energy's compliance with ASME Code, Articles IWA, IWB, IWC, IWD, and IWF for examination of Class 1, Class 2, and Class 3 components and their supports.

This document also defines the Class 1, 2, and 3 systems, components and Code classification boundaries, the required examinations.

for each ASME Section XI examination category, and the augmented inspect:ion scope.4.2 Applicable Documents The Fifth Inservice Inspection Plan for ASME Code Class 1, 2, and 3, systems and components (including their supports) was developed after giving due considerations to the following documents and subject to the limitations and modifications listed in 10 CFR 50.55a(b), and to the extent practical within the limitations of design, geometry and materials of construction.

Specific areas within this plan where these documents are used in the preparation of the inspection plan are addressed within each area that is affected.4.2.1 Title 10, Code of Federal Regulations, Part 50.55a, Codes and Standards 4.2.2 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section Xl, 2004 Edition no addenda 4.2.3 ASME Boiler and Pressure Vessel Code,Section V, 2004 Edition,"Nondestructive Examination" 4.2.4 R. G. 1.26 Quality Group Classifications and Standards for Water-Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants, Revision 2, June 1975 4.2.5 Regulatory Guide 1.14, Reactor Coolant Pump Flywheel Integrity.

4.2.6 Regulatory Guide 1.147, In-service Inspection Code Case Applicability, ASME Section Xl, Division 1, Revision 15 4.2.7 EPRI TR-1 12657, Electric Power Research Institute Report for Alternative Requirements of Risk-Informed In-service Inspection Methodology.

4.2.8 Ginna Technical Requirements Manual 4.2.9 Ginna Technical Specifications CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 ajontenture oFifth Ten-Year Inspection Interval Rev. 00 a joint ventUre ofRe ,O Co t. eD Date: December 23, 2009 Inservice Inspection PIan Page 15 of 87 5.0 Augmented Inservice Inspection Requirements Augmented inservice inspection requirements are those examinations that are specified by documents other than the ASME Section XI Code. Typically, these augmented examinations are at the request of the Nuclear Regulatory Commission through such mechanisms as Bulletins, Notices and Regulatory Guides. The augmented examinations addressed by the Ginna ISI Program during the fifth inspection interval are as follows: 5.1 Break Exclusion Region (BER): Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants -Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping (Examination of High Energy Line Break Piping)Source Document:

NUREG-0800, Standard Review Plan Associated Document:

Section 3.6.2, Ginna UFSAR, Rochester Gas and Electric Corporation's Report "Effect of Postulated Pipe Breaks Outside the Containment Building", dated October 29, 1973.Purpose: The purpose of this augmented program is to perform examinations on piping subject to High Energy Line Break (HELB) analysis criteria.Scope: Augmented Inservice Inspection Program for Main Steam and Main Feedwater Piping formally included in the High Energy Program. Table 5.1 Identifies High Energy Piping circumferential butt welds at design break locations and consequential break locations that require examination.

Method: Ultrasonic examinations will be performed per Section XI, IWA-2232.Radiographic examination using B31.1 -1967 or later may be used in lieu of Ultrasonic examinations.

Industry Code or Standards:

ASME Code Section XI, 2004 Edition for Ultrasonics, or ANSI/ASME B31.1, 1967 or later for radiography.

Frequency:

Volumetric examinations will be performed once per 10 year interval per Section 3.6.2, Ginna UFSAR.Acceptance Criteria or Standard:

Flaws detected during examination shall be evaluated by comparing the examination results to the acceptance standards established in ASME Section XI, IWC-3514.

Radiographic examination shall be evaluated using B31.1 -1967 or later.

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 C join vFifth Ten-Year Inspection Interval joint venture of. Rev. 00 0 , 'eDF Date: December 23, 2009 O" Inservice Inspection Plan Page 16 of 87 Requlatory Basis: The regulatory basis for this augmented examination program is NUREG-0800, Standard Review Plan.Table 5.1 Break Exclusion Region (High Energy Program) Weld Examination Population MAIN STEAM -INTERMEDIATE BUILDING -WELDS: Design Basis Break Consequential Break 30A-MS-600-1 30A-MS-600-1 30A-MS-600-1 30A-MS-600-1A 30A-MS-600-1A 30A-MS-600-1A 30A-MS-600-1A 30A-MS-600-1A 30A-MS-600-1A 30A-MS-600-1A 30A-MS-600-1A 30A-MS-600-1A 30A-MS-600-1A J K L D2 D3 E El E2 F G G1 G2 H 30B-MS-600-1 30B-MS-600-1 30B-MS-600-1 30B-MS-600-1 B 30B-MS-600-1 B 30B-MS-600-1 B 30B-MS-600-1 B 30B-MS-600-1 B 30B-MS-600-1 B 30B-MS-600-1 B 30B-MS-600-1 B 30B-MS-600-1 B N 0 P D1 D2 E J K L Li L2 M 36-MS-600-1 36-MS-600-1 30A-MS-600-1A 30A-MS-600-1A 30A-MS-600-lA 30B-MS-600-1 B 30B-MS-600-1 B 30B-MS-600-1 B LI P1 D D1 F1 D H1 JI MAIN STEAM -TURBINE BUILDING -WELDS: Consequential Break Design Basis Break 36-MS-600-1 24A-MS-600-1A 24A-MS-600-1A 24A-MS-600-1A 24B-MS-600-1A 24B-MS-600-1A 24B-MS-600-lA L4 B Cl D1 B C D 24A-MS-600-1A 24A-MS-600-1A 24A-MS-600-1A 24B-MS-600-1A 24B-MS-600-1A 24B-MS-600-1A A B1 D A B1 Cl 36-MS-600-1 L2 CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 C oint vFifth Ten-Year Inspection Interval ajoint venture of Rev. 00 C.O- stebDF ,,1 Date: December 23, 2009 Inservice Inspection Plan Page 17 of 87 FEEDWATER

-TURBINE BUILDING -WELDS: Consequential Break Design Basis Break 20-FW-900-1 20-FW-900-1 20-FW-900-1 20-FW-900-1 8-FW-900-1 8-FW-900-1 8-FW-900-1 8-FW-900-1 J L M M2 B D F H 20-FW-900-1 20-FW-900-1 20-FW-900-1 8-FW-900-1 8-FW-900-1 8-FW-900-1 8-FW-900-1 KI Li M1 A C E G 20-FW-900-1 M3 FEEDWATER LOOP A -WELDS: Consequential Break Desiqn Basis Break 14A-FW-900-1 14A-FW-900-1 14A-FW-900-1 14A-FW-900-1 14A-FW-900-1 14A-FW-900-1 14A-FW-900-1 14A-FW-900-1A 14A-FW-900-1A 14A-FW-900-1A T2 U7 U6 VA Vl V2A W zi Z3 14A-FW-900-1 14A-FW-900-1 14A-FW-900-1 14A-FW-900-1 14A-FW-900-1 14A-FW-900-1 14A-FW-900-1A 14A-FW-900-1A 14A-FW-900-1A 14A-FW-900-1A T3 U5 V VB V2 V2B x z Z2 Z4 20-FW-900-1 14A-FW-900-1 14A-FW-900-1 14A-FW-900-1A M4 Ul u2 AA FEEDWATER LOOP B -WELDS: Consequential Break Design Basis Break 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 F3 F1 G G1 G4 HA H1 H2A J 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 14B-FW-900-1 F5 F2 G2 G3 H HB H2 H2B K 20-FW-900-1 14B-FW-900-1 14B-FW-900-1 B 14B-FW-900-1 B Al F4 N V CEI -R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a join vente of Fifth Ten-Year Inspection Interval a joint venture of Rev. 00__" 'Date: December 23, 2009)Inservice Inspection Plan Page 18 of 87 14B-FW-900-1B L 14B-FW-900-1B M 14B-FW-900-1B N1 14B-FW-900-1B 0 14B-FW-900-1B 01 14B-FW-900-1B P 5.2 RG 1.14 RCP FLYWHEEL:

Reactor Coolant Pump Flywheel Integrity Source Document:

NRC Regulatory Guide 1.14 Associated Document:

UFSAR Section 5.4.1.2.5, WCAP-15666-A Rev 1 Purpose: The reactor coolant pump (RCP) motor flywheels are examined due to a concern about high-energy missiles inside containment that could potentially damage, and cause the simultaneous failure of, multiple trains of redundant safety-related systems.Scope: The scope includes the examination of all RCP flywheels.

Method: Surface and volumetric examinations of all four RCP flywheels shall be conducted in accordance with Ginna UFSAR Section 5.4.1.2.5.

These examinations are to be performed during RCP refurbishment.

Industry Code or Standards:

ASME Code Section Xl, 2004 Edition.Frequency:

WCAP-15666-A Rev 1 was approved by the NRC Staff. This allowed the inspection frequency in Reg. Guide 1.14 of every 10 years to be extended to every 20 years. The inservice inspection plan for the reactor coolant pump (RCP) flywheels consists of either an ultrasonic (UT) examination over the volume from the inner bore of the flywheel to the circle of one-half the outer radius or conduct an ultrasonic (UT) and a surface (MT and/or PT) examination of exposed surfaces defined by the volume of the disassembled flywheels once every 20 years.Acceptance Criteria or Standard:

Any flaws detected during examination shall be forwarded to Constellation Engineering for resolution.

Regulatory Basis: The regulatory basis for this augmented examination program is NRC Regulatory Guide 1.14.5.3 Code Case N-722: Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 Materials Source Document:

10 CFR 50.55a (September 2008)

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval ajoint Venture of Rev. 00'wsteua eD Date: December 23, 2009 Inservice Inspection Plan Page 19 of 87 Associated Document:

ASME Code Case N-722 Purpose: Components in the primary coolant system containing Alloy 600 material and 82/182 dissimilar metal butt welds may be susceptible to degradation caused by primary water stress corrosion cracking (PWSCC). Code Case N-722 was published by ASME to provide for additional leak detection capabilities of locations containing Alloy 600/82/182 materials.

Scope: The scope includes those primary coolant system components containing Alloy 600 material and 82/182 dissimilar metal butt welds. The inspection requirements of ASME Code Case N-722 do not apply to components with pressure retaining welds fabricated with Alloy 600/82/182 materials that have been mitigated by weld overlay or stress improvement.

Method: Bare metal visual examination (VE) with insulation removed. Alternatively, the VE may be performed with insulation in place using remote visual inspection equipment that provides resolution of the component metal surface equivalent to a bare-metal direct VE.The VE may be performed when the system or component is depressurized.

An ultrasonic examination performed from the component inside or outside surface in accordance with the requirements of Table IWB-2500-1 and Appendix VIII shall be acceptable in lieu of the VE.If ultrasonic examinations of butt welds are used to meet the additional actions or inspections discussed under acceptance criteria below, they must be performed using the appropriate supplement of Section XI, Appendix VIII of the ASME Boiler and Pressure Vessel Code.Industry Code or Standards:

ASME Code Case N-722 Frequency:

Examination frequencies are based on the component item number listed in Table 1 of Code Case N-722. The frequency ranges from once each refueling outage to once each inspection interval.Acceptance Criteria or Standard:

ASME Section Xl, IWB-3522.

The following conditions of 50.55a(g)(6)(ii)(E) must be met: If a visual examination determines that leakage is occurring from a specific item listed in Table 1 of ASME Code Case N-722 that is not exempted by the ASME Code, Section Xl, IWB-1220(b)(1), additional actions must be performed to characterize the location, orientation, and length of crack(s) in Alloy 600 nozzle wrought material and location, orientation, and length of crack(s) in Alloy 82/182 butt welds. Alternatively, licensees may R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005

• Fifth Ten-Year Inspection Interval a joint venture of Rev. 00 owe '% Date: December 23, 2009 Inservice Inspection Plan Page 20 of 87 replace the Alloy 600/82/182 materials in all the components under the item number of the leaking component.

If the additional actions required above determ~ine tha"t a flaw is circumferentially oriented and potentially a result of primary water stress corrosion Cracking, licensees shall perform non-visual NDE inspections of components that fall Under that ASME Code Case N-722 item number. The number of components inspected must equal 0r exceed the number of components found to be leaking under that item number. If circumferential cracking is identified in the sample, non-visual NDE must be performed in the remaining components under that item number.Regulatory Basis: 10 CFR 50.55a(g)(6)(ii)(E) 5.4 Code Case N-729-1: Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds Source Document:

10 CFR 50.55a (September 2008) and August 5, 2009 (74 FR 38893).Associated Document:

ASME Code Case N-729-1 Purpose: Components in the primary coolant system containing dissimilar metal butt welds may be susceptible to degradation caused by primary water stress corrosion cracking (PWSCC). Code Case N-729-1 is the ASME Code Case developed to inspect reactor vessel heads to locate PWSCC.Scope: The scope includes PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds. Appendix 1 to the Code Case shall not be used without prior NRC approval.Method: There are two methods required depending on location.1. Bare metal visual examination (VE) of the entire surface of the head, including the intersection of each nozzle with the head, with insulation removed. Alternatively, the VE may be performed with insulation in place using remote visual inspection equipment that provides resolution of the component metal surface equivalent to a bare-metal direct VE.The VE may be performed when the system or component is depressurized.

CENG

• R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval a joint venture of Rev. 0'm!o-ý #-"-eDF Date: December 23, 2009 Inservice Inspection Plan Page 21 of 87 2. In stead of the specified "examination method" requirements for volumetric and surface examinations in Note 6 of Table 1 of Code Case N-729-1, volumetric and/or surface examination of essentially 100% of the required volume or equivalent surfaces of the nozzle tube identified in Figure 2 of Code Case N-729-1 shall be performed.

A volumetric or surface leak path assessment through the J-groove welds shall be performed.

If a surface examination is being substituted for a volumetric examination on a portion of a penetration nozzle that is below the toe of the J-groove weld (point E on Figure 2 of the Code Case), the surface examination shall be of the inside and outside wetted surface of the penetration nozzle not examined volumetrically.

3. Ultrasonic examinations shall be performed using personnel, procedures and equipment that have been qualified by blind demonstration on representative mockups using a methodology that meets the conditions below instead of the qualification requirements of the Code Case. References to Appendix VIII shall be to the 2004 Edition with no addenda.(i) The specimen set shall have an applicable thickness qualification range of +25 percent to -40 percent for nominal depth through-wall thickness.

The specimen set shall include geometric and material conditions that normally require discrimination from primary water stress corrosion cracking (PWSCC) flaws.(ii) The specimen set shall have a minimum of ten (10) flaws which provide an acoustic response similar to PWSCC indications.

All flaws shall be greater than 10 percent of the nominal pipe wall thickness.

A minimum of 20 percent of the total flaws shall initiate from the inside surface and 20 percent from the outside surface.At least 20 percent of the flaws shall be in the depth ranges of 10-30 percent through wall thickness and at least 20 percent within a depth range of 31-50 percent through wall thickness.

At least 20 percent and no more than 60 percent of the flaws shall be oriented axially.(iii) Procedures shall identify the equipment and essential variables and settings used for the qualification, and are consistent with Subarticle VIII-2100 of Section Xl, Appendix VIII. The procedure shall be requalified when an essential variable is changed outside the demonstration range as defined by Subarticle VIII-3130 of Section Xl, Appendix VIII and as allowed by Articles VIII-4100, VIII-4200 and VIII-4300 of Section XI, Appendix VIII. Procedure qualification shall include the equivalent of at least three personnel performance demonstration test sets.Procedure qualification requires at least one successful personnel performance demonstration.(iv) Personnel performance demonstration test acceptance criteria shall meet the personnel performance demonstration detection test acceptance criteria of Table CE G -R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 l ajointventureof .Fifth Ten-Year Inspection Interval Rev. 00 0 ',cO 6",eDF Date: December 23, 2009 Inservice Inspection Plan Page 22 of 87 VIII-S1O0-1 of Section Xl, Appendix VIII, Supplement

10. Examination procedures, equipment, and personnel are qualified for depth sizing and length sizing when the RMS error, as defined by Subarticle ViI1-3120 of Section Xl, Appendix VIII, of the flaw depth measurements, as compared to the true flaw depths, do not exceed 1/8 inch (3 mm), and the root mean square (RMS) error of the flaw length measurements, as compared to the true flaw lengths, do not exceed 3/8 inch (10 mm), respectively.

Industry Code or Standards:

ASME Code Case N-729-1 Frequency:

Examination frequencies are based on the component item number listed in Table 1 of Code Case N-729-1. The visual examination is required every third refueling outage, or five years whichever is less. The volumetric and/or surface examination is required once every inspection interval (10 Years). If flaws attributed to PWSCC are identified, whether acceptable or not for continued service under paragraphs 3130 or 3140 of the Code Case, the reinspection interval must be each refuel outage instead of the reinspection intervals required by Table 1, Note 8 of the Code Case.Acceptance Criteria or Standard:

Code Case N-729-1, Paragraphs 3130 or 3140 as applicable.

Regulatory Basis: 10 CFR 50.55a(g)(6)(ii)(D) 5.5 OWNER ELECTED EXAMINATIONS FOR INTERNAL COMMITMENTS There are no Owner Elected examination commitments for Ginna at this time in the Inservice Inspection Plan.5.6 LICENSE RENEWAL EXAMINATIONS FOR AGING MANAGEMENT COMMITMENTS Ginna's License Renewal Application credits the ASME Section Xl ISI Program for aging management.

The areas that License Renewal takes credit for are: ASME Section Xl Class 1, 2 and 3 Items and Component Supports and ASME Section Xl Class 1 and 2 Bolting.

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint venture of Fifth Ten-Year Inspection Interval Rev. 00 O cEutio-, n ., Date: December 23, 2009 Inservice Inspection Plan Page 23 of 87 6.0 ASME SYSTEMS & EXAMINATION BOUNDARIES 6.1 This section defines those systems that are designated as ASME Class 1, 2, or 3 and provides justification for their inclusion or exclusion within the Fifth Ten-Year Inspection Interval ISI Program. Several of Ginna systems or portions of systems are excluded from system and component volumetric and surface examination as allowed by Articles IWB-1200, IWC-1200, and IWD-1200.

However, these portions of systems are not excluded from the pressure testing requirements of ASME Code, Section Xl, except as allowed by Articles IWA-5000, IWB-5000, IWC-5000, and IWD-5000.

Systems subject to examination and testing are provided in Table 6.1, Ginna ASME Code Class Systems.The ASME Code, Section Xl Code defines the inspection requirements for each of the ASME Code Classes within the fifth inspection interval, which began on January 1i 2010.6.2 Per IWA-1400(a) it is the owner's responsibility to determine the appropriate Code Classes for each component and to identify the system boundaries subject to inspection.

IWA-1400(a), footnote 1, states that classification criteria are specified in 10 CFR50.This reference is to footnote 9 of 10 CFR50.55a which references Regulatory Guide 1.26 and Section 3.2.2 of NUREG-0800.

6.3 The component classifications of the ASME Code (Class 1, 2, or 3) determine the rules and requirements for inspection and define the Section Xl examination boundaries.

Because early vintage nuclear plants were designed and constructed before Section III of the ASME Boiler and Pressure Vessel Code was incorporated into 10 CFR50.55a, the ASME Section Xl Code classifications for ISI may differ from the original design classifications.

Therefore, while the ASME Code classifications determine the rules for repairs and replacements and the component inspection requirements, repairs and replacements are generally performed to meet the specifications of the original design code.6.4 The Code of Federal Regulations provides criteria for the classification of Quality Group A components.

In previous issues of 10 CFR, this criterion was provided in section 50.2(v). Regulatory Guide 1.26, Quality Group Classifications and Standards for Water, Steam, and Radioactive Waste Containing Components of Nuclear Power Plants, provides criteria for the classification of Quality Group B, C, and D components.

Regulatory Guide 1.26 is used for ASME Code, Section Xl component classification at Ginna.

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 join ventFifth Ten-Year Inspection Interval Rev. 00 a joint venture ofRe .0 0 :.# Date: December 23, 2009 Inservice Inspection Plan Page 24 of 87 6.5 The ASME Code Class 1, 2, and 3 systems required to be examined in accordance with ASME Code, Section Xl, are identified on isometric drawings.

These are Plant Controlled drawings.

Section 10, the Component and Isometric Drawings Index, identifies the welds and components subject to examination and tests during the Fifth Inservice Inspection Interval.6.6 A review of ISI boundaries for application of the ASME Code,Section XI, was conducted by Constellation Energy, Inc. to assure proper Class 1, 2, and 3 boundaries prior to updating the Inservice Inspection Plan to the 2004 Edition of ASME Code,Section XI Code for the Fifth Inspection Interval.ATabli 6.1 Man aGin na32 .oASME-CodeCdalass'S'YS teMS __ _ __ _ASME , Syt~ PID#thCde-J~p~ectioin RequireIm-e'nts IWC-2000, 5000 Main Steam 33013-1231 2, 3 IWD-2000, 5000 Not Applicable Main Steam 33013-1232 Non-Code 33013-1236 Not Applicable Feedwater Sheet 1 NnCd IWC-2000, 5000 IWD-2000, 5000 Feedwater 33013-1236 2,3 Sheet 2 IWC-2000, 5000 IWD-2000, 5000 Auxiliary Feedwater 3301 3-1 237 2, 3 IWC-2000, 5000 Standby Auxiliary Feedwater 33013-1238 2, 3 IWD-2000, 5000 CE G _R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 C Iventur Fifth Ten-Year Inspection Interval Rev. 00 a joinit venltUre of.Re .0-rDate: December 23, 2009 Inservice Inspection Plan Page 25 of 87 G~innar"AS M E Code ýClas Systemns _________ANSME% -'.Cede oI- J nsper c~ Reui met System'-, P&lD # hod lsetio qeire nt 33013-1239 IWD-2000, 5000 Diesel Generator-A She3 Sheet 1 33013-1239 IWD-2000, 5000 Diesel Generator-B She3 Sheet 2 IWD-2000, 5000 Auxiliary Coolant Component 33013-1245 3 Cooling Water IWB-2000, 5000 Auxiliary Coolant Component 33013-1246 1, 2, 3 IWC-2000, 5000 Cooling Water Sheet 1 IWD-2000, 5000 IWD-2000, 5000 Auxiliary Coolant Component 33013-1246 3 Cooling Water Sheet 2 IWB-2000, 5000 Auxiliary Coolant Residual 33013-1247 1,2 IWC-2000, 5000 Heat Removal IWC-2000, 5000 Auxiliary Cooling Spent Fuel 33013-1248 2, 3 IWD-2000, 5000 Pool Cooling IWD-2000, 5000 Station Service Cooling Water 33013-1250 3 Sheet 2 33013-1250 IWD-2000, 5000 Station Service Cooling Water Sheet 23 CE G IR. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 joint vnte fFifth Ten-Year Inspection Interval a joint venture of '. .Rev. 00 Date: December 23, 2009 EeInser .vice Inspection Plan Page 26 of 87 GZ-innAM SCde 'Class Syv.stems 21 ASME'.Syste!m R&'D :I SpectCI ionri equ rE'ment"s-Station Service Cooling Water 33013-1250 2,3 IWC-2000, 5000 Sheet 3 IWD-2000, 5000 Reactor Coolant Pressurizer 33013-1258 1, 2 IWB-2000, 5000 IWC-2000, 5000 Reactor Coolant 33013-1260 1, 2 IWB-2000, 5000 IWC-2000, 5000 Containment Spray 33013-1261 2 IWC-2000, 5000 Safety Injection and 33013-1262 2 IWC-2000, 5000 Accumulators Sheet 1 Safety Injection and 33013-1262 1,2 IWB-2000, 5000 Accumulators Sheet 2 IWC-2000, 5000 CLASS 3 NDE EXEMPT RCS Overpressure Protection 33013-1263 3 IWD-1210 Nitrogen Accumulator System CLASS 3 SPT EXEMPT IWD-1210

.lN -R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a jFifth Ten-Year Inspection Interval a joint venture of Rev. 00

ýn=,eD. Date: December 23, 2009 h Inservice Inspection Plan Page 27 of 87 Chemica & VoluCodentrlaIWB-200,t500 N &ID #ASME, Lsestn P3 164 1, , IrsoectiohWCR-20,e 50 nt0 Chemical & Volume Control 33013-1264 1, 2 IWB-2000, 5000 Letdown S IWC-2000, 5000 Chemical & Volume Control 33013-1265 IWB-2000, 5000 Charging Sheet 1 2 IWC-2000, 5000 Chemical Volume Control 33013-1265 2 IWC-2000, 5000 System Sheet 2 Auxiliary Building Chemical 33013-1266 Non-Code Not Applicable Volume Control System 33013-1270 Waste Disposal System Sheet 1 Non-Code Not Applicable CLASS 2 NDE EXEMPT -IWC-1222(a)

CLASS 3 NDE EXEMPT -Waste Disposal-Liquid RC 33013-1272 2, 3 IWD-1210 Drain Tank Sheet 1 23ID11 CLASS 2 SPT EXEMPT -IWA-5110(c)

CLASS 3 SPT EXEMPT -IWD-1210 CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 ajoit vntur ofFifth Ten-Year Inspection IntervalRe.0 a joint venture of.n rc .o Rev. 00 0 C~nstet*n ng ' eDFI Date: December 23, 2009 Inservice Inspection Plan Page 28 of 87 TAblib,1 A Codde Cass Systems _________"Cyte ode' Inhs'pectionRq CLASS 2 NDE EXEMPT -IWC-1222(a)

Waste Disposal-Liquid RC 33013-1272 CLASS 3 NDE EXEMPT-Drain Tank Sheet 2 2,3 IWD-1210 CLASS 2 SPT EXEMPT -IWA-5110(c)

CLASS 3 SPT EXEMPT -IWD-1210 CLASS 2 NDE EXEMPT -IWC-1222(a)

Waste Disposal-Gas 33013-1275 CLASS 3 NDE EXEMPT-Hydrogen Recombiner Sheet 1 2,3 IWD-1210 IWD-5000 CLASS 3 SPT EXEMPT -IWD-1210 CLASS 3 NDE EXEMPT Waste Disposal-Gas 33013-1275 2,3 IWD-1210 Hydrogen Recombiner Sheet 2 CLASS 3 SPT EXEMPT IWD-1210 33013-1277 CLASS 2 NDE EXEMPT -Steam Generator Blowdown Sheet 1 2 IWC-1222(a)

She IWC-5000 CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 o f Fifth Ten-Year Inspection Interval ajoint venture of Rev. 00Comsf ... Date: December 23, 2009 Enew eDF Inservice Inspection Plan Page 29 of 87 G inia 'ASME-Code Cd syste CLASS 1 NDE EXEMPT -IWB-1220(b)

CLASS 2 NDE EXEMPT-IWC-1 222(a)33013-1278 CLASS 3 NDE EXEMPT -Nuclear Sampling Sheet 1 1,2,3 IWD-1210 CLASS 1 -IWB-5000 CLASS 2 SPT EXEMPT -IWA-5110(c)

CLASS 3 SPT EXEMPT -IWD-1210 CLASS 2 NDE EXEMPT -Nuclear Sampling Sheet 2 2 IWC-1222(a)

CLASS 2 -IWC-5000 CLASS 2 NDE EXEMPT -Post Accident Sampling 33013-1279 2 IWC-1222(a)

System CLASS 2 SPT EXEMPT -IWA-51 10(c)Containment HVAC Systems CLASS 2 NDE EXEMPT -Containment Recirculating 33013-1863 2 IWC-1222(c) and Cooling System, Post CLASS 2 SPT EXEMPT -Accident Charcoal Filters IWC-5222(b)

CLASS 2 NDE EXEMPT -IWC-1222(c)

Containment HVAC Systems 33013-1865 CLASS 3 NDE EXEMPT -Containment Purge Supply Sheet 1 2, 3 IWD-1220 (a)Mini-Purge Supply CLASS 2 SPT EXEMPT -IWA-5110(c)

CLASS 3 SPT EXEMPT -IWA-5110(c)

"E G'R R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a I joint venFifth Ten-Year Inspection Interval ajoint venture of Rev. 00 0 conwi 61,e Date: December 23, 2009 E ,eDF Inservice Inspection Plan Page 30 of 87ý_iTabIle 6.1 Code Clas ytm~~Syst~~~rn

~ &D#O e Iseton Requirements CLASS 2 -IWC-2000 CLASS 3 NDE EXEMPT -Containment HVAC Systems 33013-1865 2, 3 IWD-1220 (a)Containment Purge Exhaust Sheet 2 CLASS 2 SPT EXEMPT -IWA-5110(c)

CLASS 3 SPT EXEMPT -IWA-5110(c)

CLASS 2 NDE EXEMPT -Containment HVAC Systems 33013-1866 2 IWC-1222(a)

Penetration Cooling System CLASS 2 SPT EXEMPT -IWA-51 10(c)CLASS 2 NDE EXEMPT -IWC-1222(c)

Auxiliary/Intermediate Bldgs CLASS 3 NDE EXEMPT -HVAC Systems VCT Exhaust, 33013-1870 2, 3 IWD-1220 (a)Charcoal Filter CLASS 2 SPT EXEMPT -IWA-51 10(c) CLASS 3 SPT EXEMPT -IWA-51 10(c)CLASS 2 NDE EXEMPT -Containment Vessel Air & 33013-1882 2 IWC-1222(c)

Proof Test & Breathing Air CLASS 2 SPT EXEMPT -IWA-51 10(c)CLASS 2 NDE EXEMPT -Service Air 33013-1886 2 IWC-1222(a)

Sheet 2 CLASS 2 SPT EXEMPT -IWA-51 10(c)

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint venture of Fifth Ten-Year Inspection Interval Rev. 00 ConW "eDate: December 23, 2009 Inservice Inspection Plan Page 31 of 87 GiC2nna ASMýCode. Class ,Systemns,.,,, System P&ID P: Cod 1nP§ect~on Requiremnents Cls CLASS 2 NIDE EXEMPT IWC-1222(c)

Instrument Air Containment CLASS 3 NDE EXEMPT -Building 33013-1887 2, 3 IWD-1220 (a)CLASS 2 SPT EXEMPT -IWA-5110 (c) CLASS 3 SPT EXEMPT -IWD-1210 CLASS 2 NDE EXEMPT -IWC-1222(c)

Instrument Air Intermediate CLASS 3 NDE EXEMPT -Bldg 33013-1893 2, 3 IWD-1220 (a)CLASS 2 SPT EXEMPT -IWA-5110 (c) CLASS 3 SPT EXEMPT -IWD-1210 CLASS 2 NDE EXEMPT -IWC-1222(c)

CLASS 3 NDE EXEMPT -Primary Water Treatment Sheet 3 2, 3 IWD-1220 (a)CLASS 2 SPT EXEMPT -IWA-5110 (c) CLASS 3 -IWD-5000 Intermediate Bldg. and CLASS 2 NDE EXEMPT -Containment Heating Steam 33013-1915 2 IWC-1222(c) and Condensate CLASS 2 SPT EXEMPT -1n Cn1 IWA-5110 (c)

CENI. R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Sj oFifth Ten-Year Inspection Interval a joint venture of Rev. O00 ConsW~m .,Date: December 23, 2009 O # %eDF Inservice Inspection Plan Page 32 of 87.aible6.1AA

~Gi nn a',,,AS M Code Page32S s'~. ~: AME System P&ID Code Inspection Requirementst Fire Protection Fire Service CLASS 2 NDE EXEMPT -Water Auxiliary Bldg., 33013-1991 2 IWC-1222(c)

Intermediate Bldg. CLASS 2 SPT EXEMPT -Containment Bldg IWA-51 10 (c)Incore Detectors Drive Units CLASS 1 NDE EXEMPT -Skid IWB-1220(b)

Motor Driven and Turbine CLASS 3 NDE EXEMPT -Driven Auxiliary Feedwater 33013-2285 3 IWA-1 320 (e)CLASS 3 SPT EXEMPT -Pumps Lube Oil Skid IWA-1320 (e)

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint ventu re of Fifth Ten-Year Inspection Interval Rev. 00 oe ,,0,eoe Date: December 23, 2009 Inservice Inspection Plan Page 33 of 87 7.0 APPLICATION CRITERIA AND CODE COMPLIANCE 7.1 ASME Section Xl The following provides a summary of the application of ASME Code, Section Xl, to the Ginna Nuclear Power Plant, Ten-Year Inspection Plan for the Fifth Inspection Interval.The application and distribution of examinations for this interval is based upon utilizing Inspection Program B as defined by Articles IWB-2412, IWC-2412, and IWD-2412 of Section XI.The results of this application are summarized by ASME Category and Item number and are contained within Table 7.2. These tables only contain those ASME Item numbers that are relevant to Ginna.7.1.1 EXAMINATION CATEGORY B-A -PRESSURE RETAINING WELDS IN REACTOR VESSEL Reactor vessel examinations will be deferred until the third period as allowed in Table IWB-2500-1.

7.1.2 EXAMINATION CATEGORY B-B -PRESSURE RETAINING WELDS IN VESSELS OTHER THAN REACTOR VESSELS Examination Category B-B requires that primary side of steam generators, pressurizer, and regenerative heat exchangers be examined during the inspection interval.

The examinations may be limited to one of a group of vessels with similar function.Steam Generators There is one Class 1 Examination Category B-B, weld on each steam generator.

Table IWB-2500-1, Category B-B, Note (1) allows the examination to be limited to one vessel among a group of vessels performing a similar function or 50%. There is one tube sheet to head weld on each S/G. One of two tube sheet to head welds is selected for examination, thus meeting the one vessel among a group of vessels performing a similar function or 50% requirement.

Pressurizer Section XI requires examination of both shell to head welds and one foot of one intersecting longitudinal weld per head. The Ginna pressurizer has two longitudinal welds or one per head. Both of the shell to head welds are selected,

( N .R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Ja jointventur Fifth Ten-Year Inspection Interval Rev. 00 c nw eDF Date: December 23, 2009 Inservice Inspection Plan Page 34 of 87 meeting the 100% requirement and one intersecting long seam weld per head is selected, meeting the 100% requirement.

Regenerative Heat Exchangers Code Case N-706 has been adopted at Ginna for use on the, regenerative heat exchangers.

This Case requires' a VT-2 ,Visual Examination every refueling outage for Examination Category B-B 1welds on all three regenerative heat exchangers.

This is done in lieu of the Volumetric examination requirement..

The primary shell-to-tubesheet and head-to-tubesheet welds on all three Ginna regenerative heat exchangers are scheduled for a VT-2 Visual Examination each refueling outage in conjunction with the system pressure test prior to returning to service.7.1.3 EXAMINATION CATEGORY B-D -FULL PENETRATION WELDS OF NOZZLES IN VESSELS Reactor vessel examinations will be deferred until the third period as allowed in Table IWB-2500-1.

Code Case N-706 has been adopted at Ginna for use on the regenerative heat exchangers.

This Case relquires a VT-2 Visual Examination every refueling outage for Examination Category B-D welds on both regenerative heat exchangers.

This is done in lieu of the volumetric examination requirements.

All six primary nozzles welds on both Ginna regenerative heat exchangers are scheduled for a VT-2 Visual Examination each refueling outage in conjunction with the system pressure test prior to returning to service.All of the pressurizer and primary side steam generator nozzle to vessel welds and nozzle inner radii are selected for examination meeting the 100% examination requirement.

7.1.4 EXAMINATION CATEGORY B-F -PRESSURE RETAINING DISSIMILAR METAL WELDS IN VESSEL NOZZLES This category addresses Nozzle-to-Safe End Welds and Piping Welds. Ginna has developed a Code Case N-716 RI-ISI program. All Examination Category B-F Welds have been re-categorized as R-A welds in accordance with Code Case N-716. Code Case N-716 has been submitted to the NRC via Request for Alternative ISI-01. Therefore no examinations will be performed per Examination Category B-F.

R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 C N Fifth Ten-Year Inspection Interval a joint venture of Rev. 00 t eon Date: December 23, 2009 Inservice Inspection Plan Page 35 of 87 7.1.5 EXAMINATION CATEGORY B-G-1 -PRESSURE RETAINING BOLTING, GREATER THAN 2" IN DIAMETER Components with bolting greater than 2" in diameter include the reactor vessel, steam generators, and reactor coolant pumps.The visual examination of the Reactor Vessel Bolting will be deferred to the end of the interval and performed in the third period. For volumetric examination of Pump Studs, one of two sets of reactor coolant pump studs is selected for volumetric examination.

For the visual examination of pump flange surfaces, nuts, bushings and washers, one of the reactor coolant pumps will be examined only if disassembled.

This is the same requirement as for Examination Category B-L-2 and will be met during the repair / replacement process. All other B-G-1 components are scheduled for examination.

7.1.6 EXAMINATION CATEGORY B-G-2 -PRESSURE RETAINING BOLTING, 2" AND LESS IN DIAMETER This category includes manway and valve bolting, in the reactor coolant system and valve bolting in the high pressure safety injection and residual heat removal systems. This bolting will only be examined if associated connections are disassembled.

For components other than piping, bolting will be required only when the component is examined under Examination Category B-B, B-L-2, or B-M-2. This group of bolting is grouped with the components by the "Program Type".All B-G-2 bolting examinations will be scheduled during the repair /replacement process.7.1.7 EXAMINATION CATEGORY B-J -PRESSURE RETAINING WELDS IN PIPING This category addresses piping welds. Ginna has developed a Code Case N-716 RI-ISI program. All Examination Category B-J Welds have been re-categorized as R-A welds in accordance with Code Case N-716. Code Case N-716 has been submitted to the NRC via Request for Alternative ISI-01. Therefore no examinations will be performed per Examination Category B-J.7.1.8 EXAMINATION CATEGORY B-K -WELDED ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND VALVES Examination Category B-K requires examination of Integral Attachments.

For vessel attachments, Note 4 allows for multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels shall be selected for examination.

For single vessels, only one welded attachment shall R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 CE G Fifth Ten-Year Inspection Interval a joint venture ofRev. 0 Co swiat Date: December 23, 2009 Inservice Inspection Plan Page 36 of 87 be selected for examination.

There are two RPVAttachments and one attachment weld on the PZR. One of the two RPV attachment welds, one of the two S/G attachment welds, and the one PZR attachment weld are selected for examination or 66% of all vessel attachment welds.For piping pumps and valves, inspection'of 10% of the total population of integral welded attachments is required.

10% of all piping welded attachments and 10% of all pump welded attachments are selected for examination.

7.1.9 EXAMINATION CATEGORIES B1-L-1 AND B-L-2 -PRESSURE RETAINING WELDS IN PUMP CASINGS, AND PUMP CASINGS This category involves reactor coolant pumps and requires volumetric examination on pump casing welds and visual examination of pump internal pressure retaining surfaces when disassembled.

The reactor, coolant pumps are the only Code Class 1 pumps.B-L-1 requires examination of one purn in each group of pumps performing similar functions in the system. There are two Reactor Coolant Pumps at Ginna each with 3 pump body welds. One of the two Reactor Coolant Pumps is selected for examination of all three body welds meeting the 50% requirement.

B-L-2 requires examination of one pump in each group of pumps performing similar functions in the system. However, this examination is only required when a pump is dissembled for maintenance, repair, or volumetric examination.

No Reactor Coolant Pumps have been selected.

This requirement will be met during the repair / replacement process.7.1.10 EXAMINATION CATEGORIES B-M-1 AND B-M-2 -PRESSURE RETAINING WELDS IN VALVE BODIES, AND VALVE:BODIES This category only involves Reactor Coolant (RC), Residual Heat Removal (RHR), and High Pressure Safety Injection (HPSI) valves. Examinations will be conducted as required in Table IWB-2500-1.

B-M-1 Ginna has no Class 1 valves with body welds. The previous interval listed 4 however, it was determined that these were actually valve to "pup" pieces and not body welds.B-M-2 requires examination of one valve in each group of valves that are of the same size, constructural design, and manufacturing method, and that performs similar functions in the system. However, this examination is only required when a CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval a joint ventbre Of Rev. O00 0 coseoetm Date: December 23, 2009 On-w" ~~DF Inservice Inspection Plan Page 37 of 87 valve is dissembled for maintenance, repair, or volumetric examination.

No valve body internal surfaces have been selected.

This requirement will be met during the repair / replacement process. Table 7.1 below list all valves grouped by system, size, and type and subject to examination under B-M-2.Table71 AVE G ROUING GRP "Lin~e/ tVal'v"e:Nu tr~ibeT ~iie~e bVlvie Rparks >1 6A-RC-2501-A

/ 852A 6 Velan Gate 1 Valve among a 6A-RC-2501-B

/ 852B group of valves 2 6A-RC-2501-A

/ 853A 6 Velan 1Valve among a 6A-RC-2501-B

/ 853B Check group of valves 3 1OA-RCO-2501-A/700 10 Velan Gate 1 Valve among a 1OA-RCO-2501-A

/ 701 group of valves 1OA-RC0-2501-B

/ 720 1OA-RCO-2501-B

/ 721 4 10A-S12-1502-A

/ 842A 10 Darling 1 Valve among a 10A-S12-2501-B

/ 842B Check group of valves 1OA-S12-2501-A

/ 867A 10A-SI2-2501-B

/ 867B 7.1.11 EXAMINATION CATEGORY B-N-1 -INTERIOR OF REACTOR VESSEL These examinations will be conducted each inspection period.7.1.12 EXAMINATION CATEGORY B-N-2 -WELDED CORE SUPPORT STRUCTURES AND INTERIOR ATTACHMENTS TO REACTOR VESSEL Ginna does not have attachment welds within the beltline region. Examination of the attachment welds beyond the beltline region will be deferred until the third period as allowed in Table IWB-2500-1.

7.1.13 EXAMINATION CATEGORY B-N-3 -REMOVABLE CORE SUPPORT STRUCTURES These examinations will be deferred until the third period as allowed in Table IWB-2500-1.

C Nr R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 o--EN ao r. oFifth Ten-Year Inspection Interval a joint venture of. Rev. O00 0o.s!Um ý,(r-eDF Date: December 23, 2009 Inservice Inspection Plan Page 38 of 87 7.1.14 EXAMINATION CATEGORY B-O -PRESSURE RETAINING WELDS IN CONTROL ROD HOUSING AND INSTRUMENT NOZZLE HOUSINGS The ASME Code requires volumetric examination of 10% of peripheral control rod drive housings during the inspection interval.

There are 20 peripheral Control Rod Drive Mechanisms (CRDMs) on the reactor vessel head. To meet the Code requirements, weld examination of two of the CRDMS have been selected.7.1.15 EXAMINATION CATEGORY B-P -ALL PRESSURE RETAINING COMPONENTS The pressure testing program at Ginna is contained in the Pressure Test Program Plan.7.1.16 EXAMINATION CATEGORY B-Q -STEAM GENERATOR TUBING The Steam Generator Tube inspection program at Ginna is governed by Ginna Technical Specification 5.5.8.7.1.17 EXAMINATION CATEGORY C-A -PRESSURE RETAINING WELDS IN PRESSURE VESSELS This category applies to the secondary side of the steam generators and residual heat removal heat exchangers.

The examinations of steam generators are distributed between both generators.

Note (3) states that "In the case of multiple vessels of similar design, size, and service, the required examinations may be limited to one vessel or distributed among the vessels or 50%." The welds on the residual heat removal heat exchangers are inspected per Code Case N-706. The welds are selected for a VT-2 visual examination, one each period in conjunction with the C-H pressure tests.Steam Generators There are two circumferential shell welds, one tubesheet to shell weld and one circumferential head weld on both S/Gs. Table IWC-2500-1, Category C-A, Note (3) allows the examination to be limited to one vessel among a group of vessels performing a similar function or 50%.

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 C oin vFifth Ten-Year Inspection Interval Rev. 00 a joint venture ofRe .0 ofl" ,,, ,-OreDF Date: December 23, 2009* 10% 1 Inservice Inspection Plan Page 39 of 87 Residual Heat Removal Heat Exchangers (RHRHE)Code Case N-706 has been adopted at Ginna for use on the residual heat removal heat exchangers.

This Case requires a VT-2 Visual Examination every refueling outage for Examination Category C-A welds on both regenerative heat exchangers.

This is done in lieu of the volumetric examination requirements.

The one circumferential shell to flange weld and the one circumferential head to shell weld on both Ginna residual heat removal heat exchangers are scheduled for a VT-2 Visual Examination each inspection period in conjunction with the system pressure test.Summary Therefore, the Class 2 vessel shell welds are scheduled as follows. Both RHRHE circumferential shell welds are scheduled each period for 6 examinations and one circumferential shell weld is schedule on each S/G once for a total of eight examinations on Class 2 circumferential shell welds.The Class 2 head vessel welds are scheduled as follows. Both RHRHE circumferential head welds are scheduled each period for 6 examinations and the one circumferential head weld is schedule on one S/G once for a total of seven examinations on Class 2 circumferential head welds.One of the two Class 2 S/G tubesheet to shell welds is selected for examination.

7.1.18 EXAMINATION CATEGORY C-B -PRESSURE RETAINING NOZZLE WELDS IN VESSELS This category applies to secondary side steam generators and residual heat removal heat exchangers.

Note (1) in Table IWC-2500-1, Category C-B, excludes manways and handholes.

Note (3) requires that nozzles selected initially for examination shall be reexamined over the service life of the component.

Note (5)allows that in the case of multiple vessels of similar design, size, and service the required examinations may be limited to one vessel or distributed among the vessels. All the vessel nozzle welds and inner radii on one S/G and both nozzle welds on a RHRHE are selected for examination The Nozzles on the RHRHE are< NPS 12 and therefore, the Nozzle Inner Radius do not require examination in accordance with the examination figure.The examinations of steam generators are distributed between both generators.

Note (4) states that "In the case of multiple vessels of similar design, size, and CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 LFifth Ten-Year Inspection Interval a joint venture of Rev. 00 0 t ,eDF Date: December 23, 2009 Inservice Inspection Plan Page 40 of 87 service, the required examinations may be limited to one vessel or distributed among the vessels or 50%." The welds on the residual heat removal heat exchangers are inspected per Code Case N-706. The welds are selected for a VT-2 visual examination once each period in conjunction with the C-H pressure tests.7.1.19 EXAMINATION CATEGORY C-C -WELDED ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND VALVES Examination Category C-C requires examination of Integral Attachments.

For vessel attachments Note 4 allows for multiple vessels of, similar design, function, and service, only one welded attachment of only one of the multiple vessels shall be selected for examination.

For single vessels, only onewelded attachment shall be selected for examination.

There 'are four attachment welds on each Steam Generator and three attachment welds on each, RHRHE. One of the eight S/G attachment welds and one of the six RHRHE attachment welds are selected for examination for 2/14 or 14% of all vessel attachment welds.For piping pumps and valves, inspection of 10% of the total population of integral welded attachments is required.

10% of all piping welded attachments are selected for examination.

7.1.20 EXAMINATION CATEGORY C-F-I- PRESSURE RETAINING WELDS IN AUSTENITIC STAINLESS STEEL OR HIGH ALLOY PIPING This category addresses Class 2 piping welds. Ginna has developed a Code Case N-716 RISB Risk-Informed Alternate program. All Examination Category C-F-1 welds have been re-categorized as R-A welds in accordance with Code Case N-716. Code Case N-716 has been submitted to the NRC via Request for Alternative ISI-01. Therefore no examinations will be performed per Examination Category C-F-I.7.1.21 EXAMINATION CATEGORY C-F-2 -PRESSURE RETAINING WELDS IN CARBON OR LOW ALLOY STEEL PIPING This category addresses Class 2 piping welds. Ginna has developed a Code Case N-716 RI-ISI program. All Examination Category C-F-2 Welds have been re-categorized as R-A welds in accordance with Code Case N-716. Code Case N-716 has been submitted to the NRC via Request for Alternative ISI-01. Therefore no examinations will be performed per Examination Category C-F-2.

CE G -R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 ajoint ver o .fFifth Ten-Year Inspection Interval a joint venture of Rev. 00 ON; '=w i eDF Date: December 23, 2009 Inservice Inspection Plan Page 41 of 87 7.1.22 EXAMINATION CATEGORY C-G -PRESSURE RETAINING WELDS IN PUMPS AND VALVES Ginna has no Pressure Retaining welds in Class 2 Pumps or Valves 7.1.23 EXAMINATION CATEGORY C-H- ALL PRESSURE RETAINING COMPONENTS The pressure testing program at Ginna iscontained in the Pressure Test Program Plan.7.1.24 EXAMINATION CATEGORY D-A -WELDED ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND VALVES Examination Category D-A requires examinrtion of Integral Attachments.

For vessel attachments Note 3 allows for multiple vessels of similar design, function, and service, only the welded attachment of one of the multiple vessels shall be selected for examination.

For single vessels, only one welded attachment shall be selected for examination.

Ginna has two CCHX and two DG Jacket Water Heat Exchangers that have two attachment welds each.: One integral attachment is selected on one CCHX and one DG Jacket Water Heat Exchanger.

In addition, one integral attachment on the CCW Surge Tank (a single vessel with two integral attachments) is selected for examination.

Three of ten vessel attachment welds are selected for examination or 30% of all vessel attachment welds.For piping pumps and valves, inspection of 10% of the total population of integral welded attachments is required.

10% of all piping welded attachments are selected for examination.

7.1.25 EXAMINATION CATEGORY D-B -ALL PRESSURE RETAINING COMPONENTS The pressure testing program at Ginna is contained in the Pressure Test Program Plan.7.1.26 EXAMINATION CATEGORY F-A -SUPPORTS Examination Category F-A requires 25% of Class 1 Piping Supports, 15% of Class 2 Piping Supports, and 10% of Class 3 Piping Supports to be examined during the inspection interval.

The supports have been separated by type as defined in Note 1 to Examination Category F-A. The type has been added to the Item number to clearly identify each support by type. Twenty-five percent (25%)of the Class 1 supports have been selected and are prorated by type and system.Fifteen percent (15%) of the Class 2 supports have been selected and are

' i I"- R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint venture of Fifth Ten-Year Inspection Interval Rev. 00 Date: December 23, 2009 o .natio -w eD! Inservice Inspection Plan Page 42 of 87 prorated by type and system. Ten percent (10%) of the Class 3 supports have been selected and are prorated by type and system. Ginna has selected more supports than are required by Section Xl. The extra supports are those in which an associated Integral welded attachment is selected for examination.

If those extra supports are removed from the schedule then the proration will be checked to ensure the Code requirement is still met.For multiple components other than piping, within a system of similar design, function, and service, the supports of only one of the multiple components are required to be examined.

For supports other than piping supports the components are scheduled as follows: the one PZR, three of six S/G, three of the six RHRHEs, two of six Regen HX, two of four RHR Pumps, two of four CCWHE, both on the CCW Surge Tank, two of four DG JWCHX, four of the six RPV (two are inaccessible), and five of ten RCPs. Twenty-six of the forty-nine non-piping supports are selected for examination or 53% of all non-piping supports.7.1.27 EXAMINATION CATEGORY R-A The alternative Code Case N-716, RIS-B Program for piping as described in Relief Request ISI-01. The RIS-B Program has been substituted for the current program for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-1 and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety. These welds are selected as provided in Structural Integrity Associates, Inc. (SI) Calculation Package 0800472.302, Ginna N-716 Evaluation.

7.2 Augmented Programs 7.2.1 Code Case N-729-1 -PRESSURE RETAINING PARTIAL PENETRATION WELDS IN VESSELS Code Case N-729-1 has been mandated by 10 CFR 50.55a. This Code Case requires examination of Reactor Vessel Upper Heads. The head was replaced in the fourth interval with a head and nozzles containing PWSCC resistant material.Based on the materials, the reactor vessel head will be visually examined (bare metal) every third refueling outage and the nozzles and partial-penetration welds in the head will be volumetrically examined once this interval within 10 years of the replacement date.

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Ejointve Fifth Ten-Year Inspection Interval Rev. 00 a joint venture atR v.O Ioatin = Date: December 23, 2009Inservice Inspection Plan Page 43 of 87 7.2.2 MRP- 139 Category A Welds Welds of PWSCC-Resistant Material have a MRP-139 requirement to use the owner's existing ASME Code examination program or approved alternative.

The NRC has stated that the RI-ISI Program is not an alternative to the MRP Program.Ginna has 4 welds that are Category A welds. All four of these welds are Code Item Number IR1.20 (Risk Informed elements not subject to a damage mechanism).

These 4 welds are each scheduled once during the 5 th interval.

All four welds are selected for examination under the MRP Category A Augmented Program.This meets the Augmented Program MRP-139 Category A examination requirements.

7.2.3 RG 1.14 Reactor Coolant Pump Flywheel Inspections The Reactor Coolant Pump Motor flywheels will be inspected as required by Ginna UFSAR Section 5.4.1.2.5 and Commitment Change 2003-004.

Inservice inspection of each reactor coolant purrip flywheel shall be performed at least once every twenty years, when disassembled.

The inspection consist of either an ultrasonic (UT) examination over the volume from the inner bore of the flywheel to the circle of one-half the outer radius or conduct an ultrasonic (UT) and a surface (MT and/or PT) examination of exposed surfaces defined by the volume of the disassembled.

The examinations are controlled in the Preventive Maintenance Program.7.2.4 Break Exclusion Re-gion (BER)This augmented program was performed, during previous intervals under the high energy piping augmented program. The purpose of this augmented program is to perform examinations on Main Feedwater and Main Steam piping subject to High Energy Line Break (HELB) analysis criteria.

Ginna has 115 Consequential and Design Basis Break weld locations that have been assigned to the BER Augmented Program. These locations are identified in Table 5.1 of Section 5.1. All BER welds are categorized as Code Item Number R1.20 (Risk Informed elements not subject to a damage mechanism).

Although the risk-informed program would require only a sampling all 115 of these weld locations are scheduled for examination once during the 5 th interval per Section 3.6.2, of the Ginna UFSAR.

R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 CENG. Fifth Ten-Year Inspection Interval a joint venture of Rev. O00 O"A"t" 'In Date: December 23, 2009 10%" Inservice Inspection Plan Page 44 of 87 able 7.2 Ginn, Code Category-ummar" 9ateor Reuie to xamination N UmbeWr, " Item'di .' -xam, ~ ubro be EkiviAeo reng Exm'Numbe De~sriptin Metod o.mponet q ind Numteem No.'. I. jne~i equired ,. In B-A Pressure Retaining Welds in Reactor Vessel Reactor Vessel B-A B13.11 Circumferential Shell Volumetric 4 4 100%0') All Welds 0 0 4 Welds B-A 11.30 Reactor Vessel Shell-to-Volumetric 1 1(2) 100% Weld 1(2) 0 1(2)B-A 1.30Flange Weld Category.Total 5 5 0 0 5 Note 1: Deferral of examination to the end of the interval is permissible per Examination Category B-A.Notes for Cat. B-A Note 2: The shell-to-flange weld shall be examined in the first andthird-periods.

The examination in the first period need only be performed from the flange face from the flange face with the examination in the third period from the shell.B-B Pressure Retaining Welds in VesselsOther ThanReactor Vessels.;--, -.Pressurizer Shell-to-B-B B2.11 Head Welds Volumetric 2 2 100% Bothwelds 1 0 1 Circumferential Pressurizer Shell-to-1 ft (300 mm)B-B B2.12 Volumetric 2 2 100% of one weld 1 0 1 Head Welds Longitudinal Voper head Steam Generators B-B B2.40 (Primary Side) Volumetric 2 1(1) 50% Weld(") 0 1 0 Tubesheet-to-Head Weld Heat Exchangers ( Each B-B B2.60(2) (Primary Side) VT-2 3 18(2) 100% Refueling 6 6 6 ,Tubesheet-to-Head

,-. Outage ,

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint ventue Fifth Ten-Year Inspection Interval Rev. 00 0 'neFWtm ,4f, Date: December 23, 2009"n- Inservice Inspection Plan Page 45 of 87 Caegr D'"... ,.,'escn!ption';., Methodi 'C":"ompentsin

..... ined... 4P~ercentage. .Examined i'nr Examinied; 'Examined-C.... ... .rNDf-d Itvl In ierst.o, in -e6nd intird It. ... .. , ... ... Pe oi ' "Period x. -

Heat Exchangers

• Each B-B B2.80(2) (Primary Side) VT-2 3 18(2) 100% Refueling 6 6 6_Tubesheet-to-Shell ut ge Category Total 12 41 (2) 14 13 14 Note 1: The examination may be limited to one vessel among the group of-vessels performing a similar function. (Ref. Table IWB-2500-1, Examination Category B-B, Note 1)Notes for Cat. B-B Note 2: Code Case N-706 has been adopted requiring a VT-2 visual examination in lieu of the Code-required Volumetric Examination.

For these item numbers the VT.-2 Visual Examination is required each-refueling:outage.

Therefore the number required during the interval is six times the total number of components.

This is also reflected in the category total.B-D Full Penetration Welded Nozzles in-Vessels B-D B3.90 Reactor Vessel Nozzle- Volumetric 6 6 100%(2) Same as 1st 0 0 6 to-Vessel Welds Interval B-D B3.100 Reactor Vessel Nozzle Volumetric 6 6 100%(2) Same as 1st 0 0 6 Inside Radius Section Interval B-D B3.1201 3 1 Pressurizer Nozzle Volumetric 5 5 100% All nozzles 1 1 3 Inside Radius Section or visual Steam Generators B-D B3.140 t 3) (Primary Side) Nozzle Volumetric 100% All nozzles 2 2 0 Inside Radius Section or visual Heat Exchangers Same as 1st B-D(') B3.150 (Primary Side) Nozzle-to-VT-2 6 36(1) 100% Interval 12 12 12 Vessel Welds Itv C EN G R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval a joint venture ofR e .0 Constllafln 4FDate: December 23, 2009~1~Core " eDF Inservice Inspection Plan Page 46 of 87 Taln. G na,,,., .CategorySmm r MM k", r 'Requiredjq, bejxuiiiedl Nubrt uber oNme Ite Eam Nmbr of Examination~

'Number to be~ be eb Category.

C6.-mponents, inPercentage Exaineedain Examid Numbere tMethod Item No..Durng ie,. ,ed Examined Examined, Ittra on eFirst. ;i iSecond inihird-_____ ____ ____ ___ _ ____ __ _ _____ ___ ____ ____ ____ ___ Period., .Pra ~ 'eid Heat Exchangers Same as 1st B-D(1) B3.160 (Primary Side) Nozzle VT-2 6 36(l) 100% Interval 12 12 12 Inside Radius Section Category Total 33 81() 27 27 39 Note 1: Code Case N-706 has been adopted requiring a VT-2 visual examination in lieu of the Code-required Volumetric Examination.

For these item numbers the VT-2 Visual Examination is required each refueling outage. Therefore the number required during the interval is six times the total number of components.

This is-also reflected in the category total.Notes for Cat. B-D Note 2: Deferral of examination-to the end of the interval.

is permissiblez perýExamination-Category B-A (note 5). -Note 3: These Item Numbers are from the 1998 Edition of Section XIas-required by_10r CFR 50.55a(b)(xxi)(A).

If visual examination is performed, the examination shall-be in accordance with 10-CFR50.55a(b)(xxi)(A) as defined in Section 1.4.6 of this inspection plan.B-G-1 Pressure Retaining Bolting, Greater Than .2 in. (50 mm)in Diameter--.---.

Reactor Vessel Closure Same as for 0 0 1 B-G-1 B6.10 Head Nuts Visual, v-i 1 1 100% 1st interval B-G-1 B6.20 Reactor Vessel Closure Volumetric 1 1 100% Same. as for 0 0 1 Studs fst interval B-G-1 B6.40 Reactor Vessel Threads Volumetric 1 1 100% Same as for 0 0 1 in Flange 1st interval Reactor Vessel -Closure Same as for 0 0 1 Washers, BushingsVT-1 1 100% 1st interval C EN G R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 C jointventueoFifth Ten-Year Inspection Interval Rev. 00 ,.¶nsteflaton Date: December 23, 2009 Er-w ,.,eDF Inservice Inspection Plan Page 47 of 87 sj" ,4_2 GinnýV-C eCate4 mar-y-, 7 edt ExaNuinber~to Numbert Nuber to, of'-' Requir 0 ination, Numbrnti&o b 'b ~ .C o ater te. i Exam N be E'xamined in Ex'hmied Exami.ed' Examined Nuamberi Compon ents- in-Di Percentage ) `xam' ined'iý.xa

'Ninbn~of.

",aminatiin B-G-1 B6.180 Pumps Bolts and Studs Volumetric 2 1 50% Same asfor 0 0 1 1st interval Pumps Flange Surface Same as for B-G-1 B6.190 when connection Visual, VT-i 2 0 0%(0) 1ste inat r2) 0 00 disassembled 1_stinterval(2_

Pumps Nuts, Bushings, Visual, VT-i 2 0 0%(O) Same as for B-G-1 B6.200 and Washers 1st interval(2) 0 0 0 Category:Total

-10_ .5 0 0 5 Note 1: Not Required unless disassembled

-.Note 2: For heat exchangers, piping, pumps; and-valves-, examinationsare limited to components selected for examination under Examination Categories B-B, B-J, B-L-2, and B-M-2..(Ref. Table IWB-2500-1, Examination Category.

B-G-1, Note 3)B-G-2 Pressure Retaining Bolting, 2 in. (50 mm) and Less in- Diameter .Same as for B-G-2Pressurizer Bolts, Studs Visual, V-1i 1 0 0%(') 1st interval 0 0 0 and Nuts (2)Steam Generator Bolts, 0Same as for B-G-2 B7.30 Visual, VT-1 8 0 0%(l) 1st interval 0 0 0 Studs; and Nuts (2)(3)Valves Bolts, Studs and 0Same as for B-G-2 B7.70 Visual, VT-1 12 0 0%(l) 1st interval 0 0 0 Nuts (2)(3)(4)Category Total 21 0

• 0 0 O0 L _______________I____

___ ___I_____________

R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 CFifth Ten-Year Inspection Interval a joint venture of Rev 00 0 1% eDFne c Date: December 23, 2009 Inspection Plan Page 48 of 87 Tal"72Ginpa"C'

~aegr~ummr: ; Numer o,: ; -q , _

iNmb~ro~b

Required todubrt Nibrt ubrt ItemE mExam: e o be Examined ExEE'ni Ne& be aegj Nme ,Descnption Meth C omponentsin

-Percentage e i xa da Nufibr M'tod mP!,During, Rn 1tm4o- ntra equirea lntrvl Ain'First<'

Ainfecond' iThr '~-~-'~ _____ -. *Period ~ Reid" ero Note 1: Not required unless disassembled Note 2: Examination is only required once per interval Notes for Cat.B-G-2 Note 3: For heat exchangers, piping, pumps, and valves, examinations are limited to components selected for examination under Examination Categories B-B, B-J, B-L-2, and B-M-2. (Ref. Table IWB-2500-1, Examination Category B-G-2, Note 2)Note 4: Only one valve of each group of valves is required as outlined in B-M-2.B-K Welded Attachments for Vessels, Piping, Pumps, and Valves B-K B10.10 Pressure Vessels Surface 3 2 66%(2) Same as for 0 0 2 Welded Attachments 1st interval-(1)

Piping Welded Surface 14 2 10%-- Same as1 0 B-K B10.20 Attachments 1st interval,_....

B-K B10.30 Pumps Welded Surface 6 1 10% Same as for 0 1 0_Attachments II _II_1st interval I Category Total 23 5 1 2 2 Note 1: For multiple vessels of similar design, function, andservice, only one welded attachment of only one of the multiple vessels Notes for Cat. B-K shall be selected for examination.

For single vessels, only one welded attachment shall be selected for examination.

Note 2: There are two welded attachments on the RPV and 1 on the PZR for atotal of 3. Must examine 1 on each vessel or multiple vessels, therefore 2 are required for 66%B-L-1 I Pressure Retaining Welds in Pump Casings B-L-1 B12.10 Pumps Pump Casing Welds (B-L-1)

REN G R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval ajoint venture of Rev. 00_Consel Date: December 23, 2009 Inservice Inspection Plan Page 49 of 87~ &AW~ ~ ~>K~ab~e .2 Ginna Code Ciatgory,-,umniay U-'Examequireto Number to, NDmber23 0 Number I ' " ...+ .... Re u edto .Item -e,. xam bie Exmnxamieod

-.....'Category NubrDescriptio E-~ Mt~d C mpoets incntg Examifned in, Examined', ,Exeaminhed'

'Examhined'

~.hiNo Duin r, Inten/al:i4is n~c~; nh~______ , eriod .-Period~ *Pe 'JOd Notes for Cat. B-L-1 Note 1: Examination is limited to at least one pump in each group of pumps performing similar functions in the system.Note 2: There are two Reactor Coolant Pumps. Each pump has three casing welds, therefore 3 welds of 6 is required for 50%.B-L-2 Pump Casings Pumps Pump Casing % Same as for B-L-2 B12.20 Internal Surfaces Visual, VT-3 2 0 0%( first interval(')

0 0 0 (B-L-2)B-M-2 Valve Bodies Valve Body Internal %2) Same as forJ B-M-2 B12.50 Surfaces, Exceeding Visual, VT-3 12 0 %() first interval 0 0 I 0______ ~NPS 4 (ON 100) (B-M-2)j (1 jj Category Total 12 0. 0 0 0 0 (Note 1: Examination is limited to at least one valve in each group-of valves that are of the same size, constructural design, B I manufacturing method, and that perform similar functions in the system. Ginna's valves are grouped into four groups in Table Notes for Cat. B-M-2 7.1._ Note 2: Not required unless disassembled C ENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint ventur Fifth Ten-Year Inspection Interval Rev. 00 CDate: December 23, 2009 r.0?,e" eDF Inservice Inspection Plan Page 50 of 87 Cat -Smay I~e ~eory Smar>>ico : Numbierto

Number6to: ,NUmber be .be.. Ube, be ,,Examin ddin-Exdained Examined, Exarini Category TotalI 3 1 3 Notes for Cat. B-N-1 I-Note 1: Ginna has one examination required for each period, thus-3.total examinations.required.

B-N-2 Welded Core Support Structures-and.lnterior Attachments to. Reactor Vessels Reactor Vessel (PWR)B1 3.60 Interior Attachments Visual,V-Beyond Beltline Region R. E. Ginna Nuclear Power Plant CENG-GNPP-ISl-005 C EN G .Fifth Ten-Year Inspection Interval Ia joint venture of -Rev. 00 CansI 'Date: December 23, 2009 Inservice Inspection Plan Page 51 of 87 c1te~ 4 7itmGinna Coide car" __SNumberto Numbert6 Nubert 1"'ý nai e T xainD, ia"tExioe ebed B-0 Pressure Retaining Welds in Control Rod Drive and Instrument Nozzle Housings Reactor Vessel (PWR) Volumetric 10%terperipheral B-0 B14.20 Welds in Control Rod or surface 20 2 10%(1) peRipeaD D_________jrive CRD Housing (from ID) _ ______ ______ ______ housings ____ ____ ____Category Total 20 2 0 0 2 Notes for Cat. B-a Note 1: There are 20 Peripheral CRDMs, .10% of 20 is 2 CRDMs. There is 1 weld per CRDM for a total of 2 welds.C-A Pressure Retaining Welds in Pressure Vessels Each C-A C1.10 Pressure Vessels Shell Volumetric 6 8(2) inspection 3 23 Circumferential Welds interval,')(5)Each C-A Cs20 PrursRe Vessels Headi Control 4 (rv and 3 2 2 Circumferential Welds Vouerc 4() insetional 3(2 Each Pressure Vessels Volumetric 00%-1.0Tubesheet-to-Shell Weld Voruric 2 1.%... insperalo 0 1 Category Total 12 16 6 5 5 C EN G R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a Fifth Ten-Year Inspection Interval Rev. 00 ajoint ventureo atv O C 'on " 'feDF Date: December 23, 2009 Inservice Inspection Plan Page 52 of 87* Table 7~.2 n Cde Categoy ummary , Numbroo be,~oNmetoNmet caeoy tmExm ~ ume f be Eiminationiý.

ýNumiber to &K'e* c. sb-Nmern E DeExip~m mib~Prenae Eaiedi Exmie dExa"E"'eGomponents enxarnne Ceo Iter~o Duig , 6- n..jirit rnteial i~~o1~tn1!

First:i xinSeond a in Tid Nube Meho &rn h id,;.Note 1 The examination may be limited to one vessel among the group of vessels of similar design, size, and function. (Ref Table IWC-2500-1, Examination Category C-A, Note 3)Note 2: There are two circumferential shell welds on both S/Gs & 1 circumferential shell weld on both RHRHEs. One weld on each S/G is scheduled.

Both RHRHE shell welds are scheduled (VT-2) each period. Therefore the number required during the interval on the RHRHE welds is three times the total number of components.

This is also reflected in the category total.Notes for Cat. C-A Note 3: There is one circumferential head weld on both S/Gs and both RHRHEs. Examination on 1 of 2 S/G head welds is scheduled.

Both RHRHE head welds are scheduled (VT-2) each period. Therefore the number required during the interval on the RHRHE welds is three times the total number of components.

This is also reflected in the category total.Note 4: There is one tubesheet-to-shell weld on both S/G, therefore.l-of.2-tubesheet-to-shell welds are required to be examined for 50%Note 5: Code Case N-706 has been adopted requiring a VT-2 visual examination in lieu of the Code-required Volumetric Examination for the RHRHEs. For these welds-a VT-2 Visual Examination:is required-once each period. Therefore the number required during the interval is three times the total number of components.

This-is also reflected in the category total.C-B Pressure Retaining Nozzle Welds in Vessels -Nozzles Without Reinforcing Plate in Vessels > 1/2in. (13mm) Surface and Each-C-B C2.21 Nominal Thickness volumetric 2 1 50%(2) inspection 0 0 1 Nozzle-to-Shell (Nozzle interval(')

to Head or Nozzle to I_____ ___ Nozzle) Weld I I I I I I I C ENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval Rev. 00 a joint venture ofRe .0 O ConsW .'h*","'eDF Date: December 23, 2009 Inservice Inspection Plan Page 53 of 87 Tabe .2GinaCoe~atgoy mraiy ,_________

NumberofRequired to : Eamnaton -" Number to Nuimber to. Number to.Ct>.~~~~tem Exam ~be -Examnined7Nme ob b b.b ,Numbe.. Met Cs ho"mined.in

." .xarbq m Ed L-ned-Exa Examine&,-

Itebfoý,Cm No.etih h t euie .Intrai nFrt eod i hr Nozzles Without Reinforcing Plate in. Each C-B C2.22 Vessels > 1/2 in. (13mm) Volumetric 2150%(2) inspection 00 1 Nominal Thickness interval(e Nozzle Inside Radius Section Nozzles With Reinforcing Plate in Vessels > 1/2in.C-B C2.31 (13mm) Nominal VT-2 12 (3) Once-Per 4 4 Thickness Reinforcing Period Plate Welds to Nozzle and Vessel Nozzles With Reinforcing Plate in Vessels > 1/2in.(13mm) Nominal C-B C2.32 Thickness Nozzle-to-VT-2 4 12" (3) Once Per 4 4 Shell (Nozzle to Head or Period Nozzle to Nozzle) Welds When Inside of Vessel Is______LAccessible

_ E N1 R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a ente Fifth Ten-Year Inspection Interval a joint venlture of Rev. O00 COrtsmiation

":eDF Date: December 23, 2009 Inservice Inspection Plan Page 54 of 87 l7.2 iinnfia CoetategoJ mmry of eq ir dt~ xaNumrbeir to, ýNumnb r to. ~.Number to item .Ex Numberrf`

b...... n Examination -Numbee to be be e e Category-Numbeescrition Method Components in Percentage ) Ekminedinv Examined Examined-Number. te N T- Dufi Reurd-F -- Interval w First' i1n, econd' -_InKThird

---~~,- ..~ -Inter~v al- Ieid._ _ _ _ _ _Nozzle to Shell (Nozzle to Head or Nozzle to C B C2.33 Nozzle) Welds When VT-2 12 36 (3) Once Per 12 12 12 Inside of Vessel Is Period Inaccessible 22 Category Total 24 62 20 20 22 Note 1: The examination may be limited to one vessel among the group of vessels-of similar design, size, and function. (Ref. Table IWC-2500-1, Examination Category C-B, Note 4)-- --Note 2: There are two nozzles and nozzle inner radii on both S/G, therefore 2/4 nozzles-are required to-be examined or 50%Note 3: The RHRHE Nozzles are examined in accordance with Code. Case N-706. Examination.is scheduled each period. Therefore the number required during the interval on the RHRHE welds is three times the total number of components.

This is also reflected in the category total.C-C Welded Attachments for Vessels, Piping, Pumps, and Valves C-0 C3.10 Pressure Vessels Welded Attachments(1)

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 C N joinvetureoFifth Ten-Year Inspection Interval Rev. 00)IiMo #'9eD Date: December 23, 2009-0 Inservice Inspection Plan Page 55 of 87 ile 7. 2 Ginn A'C od CAteor SiirVma I Notes for Cat. C-C Note 1: For multiple vessels of similar design,. function, and-service, .only one welded attachment of only-one of the multiple vessels shall be selected for examination.-

For single vessels, only-one welded attachment shall be selected for examination. (Ref.Table IWC-2500-1, Examination Category C-C, Note 4)Note 2: There are four Welded Attachments on both S/G and three welded attachments on both RHRHE, therefore 2/14 welded attachments are required to be-examined for 14%Note 3: Examination is required whenever component support member deformation is.identified. (Ref. Table IWC-2500-1, Examination Category C-C, Note 6)Note 4: There are 113 Class 2 piping welded- attachments at Ginna. 10% sample population is 11.3 welded attachments.

To satisfy the minimum requirement of 11.3, Ginna has selected 12 Class 2 piping welded attachments.

D-A Welded Attachments for Vessels, Piping, Pumps, and Valves C ENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a Fifth Ten-Year Inspection Interval a joint venture of Rev. O00% *feDF Plan Date: December 23, 2009 Inservice Inspection Plan Page 56 of 87 Tabl 7 ~Gina Cde_.Cat0egory,&ummary Number'to Numberto Nimberto.Item ......".........

....... '" ....... Exao eqi red toU a t N mber b6 e C o uD inthod Componentsin N ..Percentage.

e Examnined TEained' ; xammei,;I, tem 'o. 2 Drig Re qured..i.tervaF .. -. -nFrt-Each identified occurrence D-A D1.10 Pressure Vessels Visual, VT-1 10 3 30% and each 0 2 1 Welded Attachments inspection interval (1)(2)(3)Each identified.-......

0-A 01.20 Piping Welded Visual, VT-I- 92 10 10% occurrence 4 3 3 Attachments V Tand each inspection interval (3)Category Total 102 13.. 4 5 4 Note 1: For multiple vessels of similar design, function, and service, the welded. attachments-of only-one of the multiple vessels shall be selected for examination.

For single vessels, only one welded attachment shallbe selected-for examination. (Ref. Table IWC-2500-1, Examination Category D-A, Note 3.Note 2: Ginna has two CCHX and two DG Jacket Water Heat Exchangers that have two attachment welds each, a single vessel with two integral attachments.

One integral attachment is selected on one CCHX;,one DG Jacket Water Heat Exchanger, and one Notes for Cat. D-A integral attachment on the CCW Surge Tank is selected forexamination.

Three of ten vessel attachment welds are selected for examination or 30% of all vessel attachment welds.Note 3: Examination is required whenever component support member deformation is identified. (Ref. Table IWD-2500-1, Examination Category D-A, Note 4)Note 4: There are 92 Class 3 piping welded attachments at Ginna. 10% sample population is 9.2 welded attachments.

To satisfy the minimum requirement of 9.2, Ginna has selected 10 Class 3 piping welded attachments.

CENG a joint-venture of R. E. Ginna Nuclear Power Plant Fifth Ten-Year Inspection Interval CENG-GNPP-ISI-005 Rev. 00 Date: December 23, 2009 Inservice Inspection Plan Page 57 of 87 i L IElG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Sjen Fifth Ten-Year Inspection Interval Rev. 00 Se '-JeDF PlanDate:

December 23, 2009"-t" Inservice Inspection PlanPage 58 of 87"C.a .'Summary A d ly, T-TAbld, Category I;-,tem : T'Des~criptio~n I 7.I NUMberof Rbeqjired'to Nbt Num er to Eaination INumber~t 1:6e< " T Nymber t: be I s 3 Piping Supports Directional lass 3 Piping Supports lulti-directional lass 3 Piping Supports hermal Movement lass 3 Piping C EN G R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval Rev. 00 ajoint venture ofRe .O 0 cW"s' Date: December 23, 2009 E- Inservice Inspection Plan Page 59 of 87 Table 7xaGnnaCnteo CCategooS umm-y N Numbrto Numberb,to Nubeto"'Inttem w 1 6tachf Requiredn itoeexamiation umber eoate ben Cate Numberti p entsqd t ed Exmed: PercentIgeW Examinedain Examinedg FE-Aminedo t Examined 2 aegry escptin- Meto Copnet in During e ftem N. s, nteo Required Intervalo fnRirst An econd fo CHI thi________ k '/ Pro&.eid K. eid Note 1: The total percentage sample shall be comprised of supports form each system, where the individual sample sizes are proportional to the total number of non-exempt supports of each typerand function within each system. (Ref. Table IWF-2500-1, Examination Category F-A, Note 2)Note 2: Ginna has selected more supports than are required by Section Xe. The extra supports are those in which an associated Integral welded attachment is selected for examination.

Notes for Cat. F-A Note 3: For multiple components other than piping, within a system of similar design, function, and service, the supports of only one of the multiple components are required to be examined. (Ref. Table IWF-2500-1, Examination Category F-A, Note 3)Note 4: One PZR, three of six SIG, three of the six RHRHEs, two of six-Regen HX, two of four RHR Pumps, two of four CCWHE, both on the CCW Surge Tank, two of four DG JWCHX" for of six4 RPV, and five. of ten RCPs. Twenty-six of the forty-nine non piping supports or 53%.______ _____1 Note 5: Two of the RPV supports are inaccessible and therefore exempt in. accordance with IWF-1 230.R-A Risk Informed Piping Welds R-A R1.11 N-716 Elements Subject Volumetric 114 24, 21% Element (1) 8 8 8 to Thermal Fatigue N-716 Elements Subject Volumetric to Intergranular or (> 2" NPS) 2 0 0 0 0 R-A R1.16 Transgranular Stress (s2" NPS) 33% Element ()(2)Corrosion Cracking Visual VT-2 4 2 1 0 1 (IGSCC, TGSCC) (< 2" NPS)R1.20 N-716 Elements not Volumetric 400 29 9 10 10 R-A Subject to a Damage (> 2" NPS) 8% Element ()(2)R1.20S Mechanism Visual, VT-2 235 22 8 8 6 (<_2" NPS)Category TotalI 755 77 26 26 25 II _ _ _ I _ _ _ _ _ _ _ _ _

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 CjointvENGumo Fifth Ten-Year Inspection Interval Rev. 00 a joint venture of 20 t DDate: December 23, 2009 Inspection Plan Page 60 of 87 I1 I I CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 C Fifth Ten-Year Inspection Interval Rev. 00~cOIwtmiI

-,"-eDF Date: December 23, 2009 I Inservice Inspection Plan Page 61 of 87 MTR Ca3tegorySuema Requiredto It~rn Exm I~umueof b EaiEd Exaination Nmert be Interva RA R1.20 Welds that are in the RI- volumetri 1)6scrvolumetricý Category Total -n 4 _ 100_ _ _ _ _ _ri _,,§ora asoD .,_ _ ?cmned __ _ n uirement to be consistent with existing /Droaram. or once every 10vears.I 11 Catei Break Exclusion Region (High Energy Line Break) Piping Welds Volumetric 115 115 100% 1 37 1_35 43 Category TotalI 115 115 37 35 43_____________________________________________________________

L .1. A A A &Notes for BER Notel: Table 5.1 of Section 5.1 identifies High Energy Piping circumferential butt welds at design break locations and consequential break locations that are scheduled to be examined.

R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a jFifth Ten-Year Inspection Interval a joint venture Of __________________________Rev.

00 0 Ihflatio0 Date: December 23, 2009 Inservice Inspection Plan Page 62 of 87 8.0 LISTING OF DEFINITIONS 8.1 Authorized Inspection Agency -an organization that is empowered by an Enforcement Authority to provide inspection personnel and services as required by ASME Section XI.8.2 Authorized Nuclear Inservice Inspector

-a person' who is employed and has been qualified by an Authorized Inspection Agency to verify that examinations, tests, and repairs are performed in accordance with the rules and requirements of ASME Section Xl.8.3 Authorized Nuclear Inservice Inspector Supervisor

-a person who is employed by an Authorized Inspection Agency to supervise Authorized Nuclear Inservice Inspectors and who is qualified as an Authorized Nuclear Inservice Inspector.

8.4 Component

-a vessel, concrete containment, pump, valve, storage tank, piping system, or component support.8.5 Component ID -Unique Plant Identifier for the ScheduleWorks database 8.6 Enforcement Authority

-a regional or local governing:

body, isuch as State or Municipality of the United States or a Province of Canada, empowered to enact and enforce Boiler and Pressure Vessel Code legislation.

8.7 Engineering Evaluation

-an evaluation of indications that exceed allowable acceptance standards to determine if the margins required by the Design Specifications and Construction Code are maintained.

8.8 Evaluation

-the process of determining the significance of examination or test results, including the comparison of examination or test results, with applicable acceptance criteria or previous results.8.9 Examination

-denotes the performance of nondestructive testing and visual observation such as volumetric examinations (radiography, ultrasonic and eddy current), surface examinations (liquid penetrant or magnetic particle), and visual examinations (VT-1, VT-2, and VT-3).8.10 Examination Category -a grouping of items to be examined or tested.8.11 Fabrication

-actions such as forming, machining, assembling, welding, brazing, heat treating, examination, testing, inspection, and certification, but excluding design, required to manufacture components, parts or appurtenances.

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval a joint venture of. Rev. 00 Coel eD Date: December 23, 2009 Inservice Inspection Plan Page 63 of 87 8.12 General Corrosion

-an approximately uniform wastage of a surface of a component, through chemical or electrochemical action, free of deep pits or cracks.8.13 Hold Time -the time after pressurization to test conditions before the visual examinations commence.8.14 Inservice Inspection

-methods and actions for assuring the structural and pressure-retaining integrity of safety-related nuclear power plant components in accordance with the rules of Section Xl of the ASME Code.8.15 Inservice Inspection Program Owner -individual in the Constellation Generation Group's Corporate Engineering Programs Unit responsible for development and oversight of station activities required to implement the Inservice Inspection Program.8.16 Inservice Life -the period of time from the initial use of an item until its retirement from service.8.17 Inservice Test -a special test procedure for obtaining, through measurement or observation, information to determine the operational readiness of a system or component.

8.18 Inspection

-verification of the performance of examinations and tests by an Inspector.

8.19 Inspection Proqram -the plan and schedule for performing examinations or tests as required by Section Xl of the ASME Code.8.20 Inspector-an Authorized Nuclear Inservice Inspector.

8.21 Nondestructive Examination

-an examination by the visual, surface, or volumetric method.8.22 Normal Plant Operating Conditions

-the operating, conditions during reactor startup, operation at power, hot standby, and reactor cooldown to cold shutdown conditions.

Test conditions are excluded.8.23 Normal Plant Operation

-the conditions of startup, hot standby, operation within the normal power range, and cooldown and shutdown of the power plant.8.24 Open Ended -a condition of piping or tubing that permits free discharge to the atmosphere or containment atmosphere.

C ENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint venture of Fifth Ten-Year Inspection Interval Rev. 00 Carmtelmi Date: December 23, 2009 Inservice Inspection Plan Page 64 of 87 8.25 Operating Convenience

-a provision to facilitate plant operation but not required to perform a specific function in shutting down a reactor to cold shutdown condition or in mitigating the consequences of an accident.8.26 Operational Readiness

-the ability of a component or system to perform its intended function when required.8.27 Original Construction Code -the original Codes(s) Edition and Addenda under which the component was constructed and installed.

8.28 Owner -the organization legally responsible for the operation, maintenance, safety, and power generation of the nuclear power plant, (i.e.; Constellation Energy).8.29 Piping System -a functional nuclear power plant system with boundaries defined by a plant Design Specification, and/or System Flow Diagrams.8.30 Regulatory Authority

-a federal government agency, such as the United States Nuclear Regulatory Commission, that is empowered to issue and enforce regulations affecting the design, construction, and operation of nuclear power plants.8.31 Relevant Condition

-a condition observed during a visual examination that requires supplemental examination, corrective measure, repair, replacement, or analytical evaluation.

8.32 Summary Number -Unique ISI designator for the ScheduleWorks database.8.33 System Hydrostatic Test Boundary -the boundary subject to test pressurization during a system hydrostatic test shall be defined by the system boundary (or each portion of the boundary) within which the components have the same minimum required classification and are designed to the same primary pressure rating as governed by the system function and the internal fluid operating conditions, respectively.

8.34 System Leakage Test Boundary -the boundary subject to test pressurization during a system leakage test shall extend to the pressure retaining components within the system boundary pressurized during normal plant operation, or during a system operability test.

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval Rev. 00 a joint venture of Rev. 00 0 on" 'eDF Date: December 23, 2009 Inservice Inspection Plan Page 65 of 87 8.35 System Pressure Test -a test in which the pressure retaining components within each system boundary is subject to system pressure under which visual examination VT-2 is performed to detect leakage. The test may be conducted in conjunction with one or more of the following system tests:* System Leakaqe Test -a system leakage test conducted with the system normal operating conditions, or during a system operability test." System Hydrostatic Test -a system hydrostatic test conducted during a plant shutdown at a pressure above nominal operating pressure or system pressure for which overpressure protection is provided.* System Pneumatic Test -a system pneumatic test conducted in lieu of a hydrostatic pressure test for components within the scope of IWC and IWD.8.36 Test -a procedure to obtain information through measurement or observation to determine the operational readiness of a component or system while under controlled conditions.

8.37 Verify -to determine that a particular action has been performed in accordance with the rules and requirements of Section Xl of the ASME Code either by witnessing the action or by reviewing records.8.38 Visual Examination VT-2 -the VT-2 Visual Examination is conducted in accordance with ASME Section Xl, IWA-2212 to determine the presence of leakage from pressure retaining components.

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 ajoint ventu~reof .Fifth Ten-Year Inspection Interval Rev. 00 ComW eDFn Date: December 23, 2009 Inservice Inspection Plan Page 66 of 87 9.0 RECORDS AND REPORTS 9.1 Records and reports for the Inservice Inspection Plan, Implementation Schedules, outage examination schedules, examination results, procedures, certifications, test, repairs, and replacements are maintained in accordance with the requirements of ASME, Section Xl, Article IWA-6000.9.2 The Owner's Activity Reports (OAR-i) are prepared in accordance with Code Case N-532-4, Repair/Replacement Activity Documentation Requirements and Inservice Summary Report.9.3 Upon completion of all repair / replacement activities, Form NIS-2A shall be prepared and maintained by Ginna.9.4 The Owner's Activity Report, Form OAR-1 shall be processed within 90 calendar days of the completion of each refueling outage. As a minimum it shall contain the following:

• A listing of items with flaws or relevant conditions that exceed the acceptance criteria and the required evaluation to determine acceptability for continued service shall be provided in the format of Table 1 of Code Case N-532-4.* An abstract for repair/replacement activities that were required due to an item containing a flaw or relevant condition that exceed acceptance criteria of Section XI, Division 1 shall be provided in the format of Table 2 of Code Case N-532-4. This information is required even if the discovery of the flaw or relevant condition that necessitated the repair/replacement activity did not result from an examination or test required by Section Xl, Division 1. If the acceptance criteria for a particular item are not specified in Section Xl, Division 1, the provisions of IWA-3100(b) shall be used to determine which repair/replacement activities are required to be included in the abstract." If no items met the criteria of (a) or (b), the term "None" shall be recorded on the applicable table." If there are multiple inspection plans with different intervals, periods, Editions, or Addenda, they shall be identified on Form OAR-1." The completed Form OAR-1 shall be submitted to the regulatory and enforcement authorities having jurisdiction at the plant site.

R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Ca E Nu .Fifth Ten-Year Inspection Interval ajoint venture of Rev. 00 0 nsteumu " 0,eDF Date: December 23, 2009 Inservice Inspection Plan Page 67 of 87 10.0 COMPONENT

& ISOMETRIC DRAWINGS 10.1 The Component and Isometric Drawings Index, Table 10.1, identifies the welds and components subject to examination and tests. The drawings included within this section identify lines classified as ASME Class 1, 2, 3, and High Energy pressure boundary within the scope of the inservice inspection program (including Risk informed and Augmented programs).

10.2 The piping integral attachment welds for examination categories B-K and C-C are identified on the corresponding Component Support drawings.10.3 The Code boundaries shown on these sketches were obtained from the plant controlled P&ID drawings.

C EN G R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint venture of Fifth Ten-Year Inspection Interval Rev. 00 Constellat Date: December 23, 2009 0 Enem, I eDF.0'6 Inservice Inspection Plan Page 68 of 87ýZ inng'.'T, ýsorn 'DJ J dd t: bon0Of n0winag-4i x Z,ýMg W i n--'g- -'N 6-1 em raw ra M(Wh kvý "A on-00 1k-M 33013-1260 Reactor Coolant (RC) P & I D W679,1440 A-1 Reactor Pressure Vessel (RRC01) W679,1441 AAA Reactor Pressure Vessel 33013-1260 Reactor Coolant (RC) P & I D A-2 Reactor Pressure Vessel Nozzle 33013-1260 Reactor Coolant (RC) P & I D Primary Coolant Loop A (general drawing; replaced by 304-014 A-3-1 A-3-1 A, A-3-1 B and A-3-1 C) 304-631 33013-1260 Reactor Coolant (RC) P & I D A-3-1 A Primary Coolant Loop A Hot Leg 304-014 A-3-1 B Primary Coolant Loop A Crossover Leg 33013-1260 Reactor Coolant (RC) P & I D 33013-1260 Reactor Coolant (RC) P & I D A-3-1 C Primary Coolant Loop A Cold Leg 304-631 33013-1258 Reactor Coolant Pressurizer (RC) P & I D 33013-1260 Reactor Coolant (RC) P & I D Primary Coolant Loop B (general drawing; replaced by 304-014 A-3-2 A-3-2A, A-3-213 and A-3-2C) 304-631 33013-1260 Reactor Coolant (RC) P & I D A-3-2A Primary Coolant Loop B Hot Leg 304-014 A-3-213 Primary Coolant Loop B Crossover Leg 33013-1260 Reactor Coolant (RC) P & I D 33013-1260 Reactor Coolant (RC) P & I D A-3-2C Primary Coolant Loop B Cold Leg 304-631 A-4 Pressurizer (TRC01) 33013-1258 Reactor Coolant Pressuri7er (RC) P & in C ENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint venture of Fifth Ten-Year Inspection Interval Rev. 00 0 onsttion 6eDF Date: December 23, 2009 E .eF Inservice Inspection Plan Page 69 of 87-~ M~7'I abler iUX mo enta od Iso tr r]awn lcec~;rR' rncawrawing , :ew-.,.~ADes6~htin

___________6

~ Q6~l~681J251 681 J252 681 J253 33013-1258 Reactor Coolant Pressurizer (RC) P & ID 681 J251 681 J252 A-4A Pressurizer Nozzle Details 681 J253 33013-1231 Main Steam (MS) (Safety Related) P & ID 33013-1236-2 Feedwater (FW) P & ID A-5 Steam Generator (EMS01A & EMS01B) 33013-1260 Reactor Coolant (RC) P & ID A-6 Support Locations n/a A-6A Steam Generator "A" Pad Details n/a A-6B Steam Generator "B" Pad Details n/a A-7 Reactor Coolant Pump (PRC01A & PRC01 B) n/a D-521015 D-521049 A-7A Reactor Coolant Pump "A" (PRC01A) D-521050 D-521015 D-521049 A-7B Reactor Coolant Pump "A" Wall Pads D-521050 D-521015 D-521049 A-7C Reactor Coolant Pump "B" D-521050 CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval Rev. 00 0 comnal 1Date: December 23, 2009 E -4FInservice Inspection Plan Page 70 of 87* -~ ~ ~ ~" * , 2, C sIndexi,-

-D-521015 D-521049 A-7D Reactor Coolant Pump "B" Wall Pads D-521050 A-7E S/G & RCP Comp. Support Attachments n/a A-7F ISO # changed to B-1A n/a A-7G S/G Upper Support -Rigid Struts n/a C-381-357-15 C-381-357-16 C-381-357-17 Chemical & Volume Control Letdown (CVCS) P 33013-1264

& ID Chemical and Volume Control System A-8 Regenerative heat Exchanger 33013-1265-1 Charging (CVCS) P & ID C-381-353-2 33013-1258 Reactor Coolant Pressurizer (RC) P & ID A-9 Pressurizer Spray Line 33013-1260 Reactor Coolant (RC) P & ID C-381-353-1 33013-1258 Reactor Coolant Pressurizer (RC) P & ID A-10-1A Pressurizer Spray Line -A Loop 33013-1260 Reactor Coolant (RC) P & ID C-381-353-1 33013-1258 Reactor Coolant Pressurizer (RC) P & ID A-10-11B Pressurizer Spray Line -B Loop 33013-1260 Reactor Coolant (RC) P & ID A-11 2" Auxiliary Spray Line C-381-357-18 CENG a joint venture of C.ý1t -4,,DF R. E. Ginna Nuclear Power Plant Fifth Ten-Year Inspection Interval CENG-GNPP-ISI-005 Rev. 00 Inservice Inspection Plan Date: December 23, 2009 Page 71 of 87_________________________

~1~

CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval a joint ventureof Rev. O00 0 -emntmn *'eDF *Date: December 23, 2009 Qe ,-DF Inservice Inspection Plan Page 72 of 87 S10 *.E.~inn~~ ~Tta lng11 6on eonrtand ]S!omefri" Wawino i~nde@x~~___________Dsig Ref9scnptlnde, Wing. Dawn N C-381-355-10 A-21 Safety Injection Pump Discharge 33013-1262-2 Safety Injection and Accumulators (SI) P & ID A-22 Letdown to Excess Letdown HTX 33013-1260 Reactor Coolant (RC) P & ID C-381-357-16 C-381-357-17 33013-1260 Reactor Coolant (RC) P & ID Chemical & Volume Control Letdown (CVCS) P A-23 2" Letdown Line to Regen HTX 33013-1264

&ID C-381-357-17 A-23A 2" Letdown Line 33013-1260 Reactor Coolant (RC) P & ID C-381-357-15 Chemical & Volume Control Letdown (CVCS) P A-24 Chemical & Volume Control Letdown 33013-1264

& ID C-381-357-20 33013-1260 Reactor Coolant (RC) P & I D Chemical and Volume Control System A-25 2" Charging Line 33013-1265-1 Charging (CVCS) P & ID C-381-357-20 Chemical and Volume Control System A-25A 2" Charging Line 33013-1265-1 Charging (CVCS) P & ID C-381-357-25 33013-1260 Reactor Coolant (RC) P & ID Chemical and Volume Control System A-26 2" Alternate Charging Line 33013-1265-1 Charging (CVCS) P & ID C-381-357-19 A-27 2" Alternate Charging Line 33013-1260 Reactor Coolant (RC) P & ID CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 o' ven Fifth Ten-Year Inspection Interval ajoint ventureof Rev. O00 Qon-e'eati" eDF Date: December 23, 2009 Inservice Inspection Plan Insrvie Ispctin PanPage 73 of 87_______~-E i ~~e~Chemical and Volume Control System 33013-1265-1 Charging (CVCS) P & ID C-381-357-21 Chemical and Volume Control System A-29 Charging Line 33013-1265-1 Charging (CVCS) P & ID C-381-357-20 C-381-357-21 Chemical and Volume Control System A-30 2" Charging Line 33013-1265-1 Charging (CVCS) P & ID C-381-357-27 Chemical and Volume Control System A-31 RCP "A" CVCS Seal Water Supply 33013-1265-1 Charging (CVCS) P & ID C-381-357-26 Chemical and Volume Control System A-31A RCP "A" CVCS Seal Water Supply 33013-1265-1 Charging (CVCS) P & ID C-381-357-32 Chemical and Volume Control System A-32 RCP "B" CVCS Seal Water Supply 33013-1265-1 Charging (CVCS) P & ID C-381-357-31 Chemical and Volume Control System A-32A RCP "B" CVCS Seal Water Supply 33013-1265-1 Charging (CVCS) P & ID A-33 RPV Closure Head -CRDM Orientation 083NE001 33013-1230 33013-1236-2 Feedwater (FW) P & ID B-1 Steam Generator (EMS01A & EMS01 B) 33013-1260 Reactor Coolant (RC) P & ID C EN G R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint venture of Fifth Ten-Year Inspection Interval Rev. 00 Consteflation-r " 4,*'eOF Date: December 23, 2009.#%F IInservice Inspection Plan Page 74 of 87 B-1A SIG Upper Support -Location Plans (Formerly A-7F) n/a____________________

8-2 ISO # changed to 0-43 n/a B-3 ISO # changed to D-1 n/a B-4 ISO # chan edtoC0-44 n/a 8-5 ISO # changed to 0-45 nla C-381 -357-10 Auxiliary Building Chemical Volume Control B-6 8" Charging Pump Discharge to Filter 33013-1265-2 System (CVCS) P & ID B-6A 24" Charging Pump Discharge Filter 33013-391-E Chemical and Volume Control System B-7 Seal Water Injection filters 1 & 2 3301 3-1265-1 Charging (CVCS) P & ID 0-381-350-1 B-8-lA Main Steam -A Loop 33013-1231 Main Steam (MS) (Safety Related) P & ID-381-350-1 B-8-1 B Main Steam -B Loop 3301 3-1231 Main Steam (MS) (Safety Related) P & ID 0-38 1-350-2 8-9 Main Steam 33013-1231 Main Steam (MS) (Safety Related) P & ID 0-381-350-2

____________________

B-9A Main Steam 33013-1231 Main Steam (MS) (Safety Related) P & ID C-381-350-3 0 B-10 Main Steam 33013-1231 Main Steam (MS) (Safety Related) P & ID C ENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint yenture of Fifth Ten-Year Inspection Interval Rev. 00 0 C tons -eDF Date: December 23, 2009 EnSDF- m0% 1 Inservice Inspection Plan Page 75 of C-381 -35 0-3_____________________

B-10A Main Steam 33013-1231 Main Steam (MS) (Safety Related) P & ID C-381-351-4____________________

B-li Feedwater 3301 3-1 236-2 Feedwater(FVV)

P & ID C-381-351-5____________________

3301 3-1236-2 Feedwater (FW) P & ID B-12 Feedwater "A" Loop 7705E158_____________________

0-381-351-1____________________

3301 3-1 236-2 Feedwater (FW) P & ID B-i13 Feedwater "B" Loop 7705E158_____________________

C-381-351-2____________________

B-i14 Feedwater 3301 3-1 236-2 Feedwater (FW) P & I D 3301 3-1262-1 Safety Injection and Accumulators (SI) P & ID Auxiliary Building Chemical Volume Control 33013-1266 System Boric Acid (CVCS) P & ID B-i5 8" Low Pressure Safety Injection 1 869E53-1 ___________________

3301 3-i1262-1 Safety Injection and Accumulators (SI) P & ID B-16 8" L.P. Safety Injection 1 869E53-1____________________

33013-1262-1 Safety Injection and Accumulators (SI) P & ID 1 869E53-1 ____________________

B-16A 8" L.P. Safety Injection 1 869E53-2___________________

Auxiliary Coolant Residual Heat Removal (AC)33013-1247 P & ID 33013-1262-1 Safety Injection and Accumulators (SI) P & ID B-16B 8" & 4" L.P. Safety Injection 17869E53-CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 ot Fifth Ten-Year Inspection Interval ajoint ventureof Rev. 00 Sot@nerDation Date: December 23, 2009 Inservice Inspection Plan Page 76 of 87 Auxioiar Cooat hi A, -di Auxiliary Coolant Residual Heat Removal (AC)33013-1247 P_&_ID B-17 Low Head Safety Injection C-381-354-3 Auxiliary Coolant Residual Heat Removal (AC)33013-1247 P&ID B-18 6" Low Head Safety Injection C-381 -354-2 __________________

Auxiliary Coolant Residual Heat Removal (AC)33013-1247 P&ID 33013-1261 Containment Spray (Sl) P & ID 33013-1262-2 Safety Injection and Accumulators (SI) P & ID C-381-359-9 B-19 Safety Injection 1869E53-3 Auxiliary Coolant Residual Heat Removal (AC)33013-1247 P&ID B-20 Residual Heat Removal C-381-354-5 Auxiliary Coolant Residual Heat Removal (AC)33013-1247 P&ID B-20A Residual Heat Removal C-381-354-4 Auxiliary Coolant Residual Heat Removal (AC)33013-1247 P&ID B-21 Residual Heat Removal C-381-354-4 Auxiliary Coolant Residual Heat Removal (AC)33013-1247 P&ID B-22 Residual Heat Removal C-381-354-1 Auxiliary Coolant Residual Heat Removal (AC)33013-1247 P &ID B-23 Residual Heat Removal C-381-354 C EN G R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint venture of Fifth Ten-Year Inspection Interval Rev. 00 0 -'t" ':'eDF -Plan Date: December 23, 2009 Inservice Inspection Plan Page 77 of 87-innQ~C-381 -354-7_____________________

Auxiliary Coolant Residual Heat Removal (AC)B-24 Residual Heat Removal 33013-1247 P& ID C-381 -354-8 Auxiliary Coolant Residual Heat Removal (AC)B-25 Residual Heat Removal 33013-1247 P &ID C-381 -354-9 Auxiliary Coolant Residual Heat Removal (AC)B-26 Residual Heat Removal 3301 3-1247 P &ID C-381-354-10 Auxiliary Coolant Residual Heat Removal (AC)B-27 Residual Heat Removal 33013-1247 P&ID 33013-12454 Auxiliary Coolant Residual Heat Removal (AC)B-28 RHR Pump 33013-1247 P&ID C-381-356-14 Auxiliary Coolant Component Cooling Water B-29 Aux Cool Pen 128 to RCP "B" BRG Cool 33013-1246-1 (AC)C-381-356-16 B-30 Aux Cool Pen 127 to RCP "A" BRG Cool 33013-1246 C-381 -350-6 B-31 "B" SiG Blowdown 33013-1277 C-381-357-121 Chemical & Volume Control Letdown (CVCS) P B-32 CVCS Letdown -Pen 112 to RH E 33013-1264

& ID C-381-357-28 B-34 Seal Water "A" RCP to PEN 108 33013-1265 CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 C Eoin vFifth Ten-Year Inspection Interval aont eueo Rev. O0 O mqyll- #-eDF Date: December 23, 2009 Inservice Inspection Plan Page 78 of 87 a ~Table, 1 0..mponenit~and Isometric Drawings IndexK ISI Drvawing.

DeinDaigN-einRfrneDrwn ,WOng______

i~mn -. 'jiptin;~C-381-357-1 Chemical & Volume Control Letdown (CVCS) P B-35 CVCS Pen 112 to Non-Regen HTX 33013-1264

& ID C-381-357-2 Chemical & Volume Control Letdown (CVCS) P B-36 CVCS Non-Regen HTX to Wall Enclosure 33013-1264

& ID C-381-357-2 Chemical & Volume Control Letdown (CVCS) P B-36A CVCS Non-Regen HTX to Wall Enclosure 33013-1264

& ID C-381-355-2 B-37 High Pressure Safety Injection 33013-1262 C-381-355-11 B-38 High Pressure Safety Injection 33013-1262-2 Safety Injection and Accumulators (Sl) P & ID C-381-355 Sht 10 B-39 H.P. Safety Injection 33013-1262 Sht 2 C-381-355-3 B-40 High Pressure Safety Injection 33013-1262-1 Safety Injection and Accumulators (Sl) P & ID C-381-355-3 B-41 High Pressure Safety Injection 33013-1262-1 Safety Injection and Accumulators (SI) P & ID C-381-355-1 B-42 High Pressure Safety Injection 33013-1262-1 Safety Injection and Accumulators (SI) P & ID C-381-355-12 B-43 High Pressure Safety Injection 33013-1262-2 Safety Injection and Accumulators (SI) P & ID C-381-355-13 B-44 High Pressure Safety Injection 33013-1262-2 Safety Injection and Accumulators (SI) P & ID B-45 High Pressure Safety Injection C-381-355-14 C ENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint venture of Fifth Ten-Year Inspection Interval Rev. 00 0 Ec.-,'.Cte" -ýeDF Date: December 23, 2009 ,0D Inservice Inspection Plan Page 79 of 87-07 ' Tal oOno~ 6fint, ndIýAffiOf f~~i"' JIex-,, J~Drawing ~~~~~& DeOsign- Dr~ir~N einRf~~ rwn 33013-1262-2 Safety Injection and Accumulators (Sl) P & ID C-381-359-8 B-46 Containment Spray 33013-1261 Containment Spray (SI) P & ID C-381-359-7 B-47 Containment Spray 33013-1261 Containment Spray (SI) P & ID C-381-359-5 B-48 Containment Spray -Upper Ring 33013-1261 Containment Spray (Sl) P & ID Auxiliary Coolant Residual Heat Removal (AC)B-49 Residual Heat Removal 33013-1247 P&ID Auxiliary Coolant Residual Heat Removal (AC)33013-1247 P&ID B-49A Residual Heat Removal 33013-2085 C-381-358-13 C-381-358-18 Station Service Cooling Water Safety Related B-50 SVC Water to RX Bldg Fan "A" Cooler 33013-1250-3 (SW) P & ID C-381-358-9 C-381-358-21 Station Service Cooling Water Safety Related B-50A SVC Water From RX Bldg Fan "A" Cooler 33013-1250-3 (SW) P & ID C-381-358-13 C-381-358-24 Station Service Cooling Water Safety Related B-51 SVC Water to RX Bldg Fan "B" Cooler 33013-1250-3 (SW) P & ID B-51A SVC Water to RX Bldg Fan "B" Cooler C-381-358-24 CEN _ R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval Rev. 00 0 c-wuzi-,, -'"eDF Date: December 23, 2009 E" Inservice Inspection Plan Page 80 of 87 C ENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 ot Fifth Ten-Year Inspection Interval iont ventureo Rev. 00 0onstl eDate: December 23, 2009.0% 1F Inservice Inspection Plan Page 81 of 87-Ta~n ~lfrf e rawing~Station Service Cooling Water Safety Related 33013-1250-3 (SW) P & ID 33013-1261 Containment Spray (Sl) P & ID B-54 S.I. Pump Disch & Accum Test Line 33013-1262-1 Safety Injection and Accumulators (Sl) P & ID Westinghouse 4807-1 Auxiliary Coolant Residual Heat Removal (AC)B-109 RHR Heat Exchangers "A" & "B" 33013-1247 P & ID Westinghouse 4807-1 Auxiliary Coolant Residual Heat Removal (AC)B-1 09A RHR Heat Exchangers "A" & "B" 33013-1247 P & ID C-381-352-4 C-381-352-5 C-1 Auxiliary Feedwater Pumps Discharge 33013-1237 Auxiliary Feedwater (FW) P & ID C-381-352-2 C-381-352-4 C-IA Auxiliary Feedwater Pump Discharge 33013-1237 Auxiliary Feedwater (FW) P & ID C-381-352-4 C-1B Auxiliary Feedwater Pump Discharge 33013-1237 Auxiliary Feedwater (FW) P & ID C-381-352-3 C-IC Auxiliary Feedwater Pump Discharge 33013-1237 Auxiliary Feedwater (FW) P & I D C-381-252-2 C-1D Auxiliary Feedwater Pump Discharge 33013-1237 Auxiliary Feedwater (FW) P & I D C-381-352-1 C-1E Auxiliary Feedwater Pump Discharge 33013-1237 Auxiliary Feedwater (FW) P & ID C-381-352-1 C-1F Auxiliary Feedwater Pump Discharge 33013-1237 Auxiliary Feedwater (FW) P & ID CENG a joint venture of o -ta-ua 6"-DF R. E. Ginna Nuclear Power Plant Fifth Ten-Year Inspection Interval CENG-GNPP-ISI-005 Rev. 00 Inservice Inspection Plan Date: December 23, 2009 Page 82 of 87 CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval Rev. 00 0 w Date: December 23, 2009 10% Inservice Inspection Plan Page 83 of 87-, 4 ~~R E. G~inna..2 x-..-.-.I~t ~ Dsig Reer~~e~rawing,,.-

~IS~ODin' Dawing Nq No ~~~~ ~~9~~~sgn; 2 -~ ~ ~ -ecpd , Auxiliary Coolant Component Cooling Water 33013-1246-1 (AC)C-381-358-3 C-381-358-5 Station Service Cooling Water Safety Related C-11 Service Water 33013-1250-2 (SW) P & ID C-381-358-1 C-381-358-2 C-381-358-6 Station Service Cooling Water Safety Related 33013-1250-1 (SW) P & ID Station Service Cooling Water Safety Related C-12 Service Water 33013-1250-2 (SW) P & ID C-381-358-37 C-381-358-38 Station Service Cooling Water Safety Related C-13 SW to A & B Edg Water Coolers 33013-1250-1 (SW) P & ID C-381-358-9 Station Service Cooling Water Safety Related C-14 SW from Pens 308, 311, 315 & 323 33013-1250-3 (SW) P & ID C-381-358-40 Station Service Cooling Water Safety Related C-15 SW- Edg Bldg & Turbine Bldg 33013-1250-1 (SW) P & ID C-381-358-14 Station Service Cooling Water Safety Related C-16 SW to N/C Water Chillers 33013-1250-i (SW) P & ID CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 a joint venture of Fifth Ten-Year Inspection Interval Rev. 00 conte ion 'eDF Date: December 23, 2009 oEne0 r Inservice Inspection Plan Page 84 of 87 Iae8 f8 Table 10.1e OACompio net-.nd Itometric Drwni ne-S IS Drwn Deig Reeec rwn Station Service Cooling Water Safety Related 33013-1250-3 (SW) P & ID 33013-1237 Auxiliary Feedwater (FW) P & ID Station Service Cooling Water Safety Related C-16A SW to A/C Water Chillers 33013-1250-3 (SW) P & ID C-381-358-13 Station Service Cooling Water Safety Related C-16B SW to A/C Water Chillers 33013-1250-3 (SW) P & ID C-381-358-34 C-381-358-35 Station Service Cooling Water Safety Related C-1 7 Service Water 3301 3-1250-1 (SW) P & ID C-381-358-10 Station Service Cooling Water Safety Related C-1 8 Service Water 33013-1250-3 (SW) P & ID C-381-358-40 Station Service Cooling Water Safety Related C-19 SW -"B" Edg to Turbine Bldg 33013-1250-1 (SW) P & ID C-381-040 C-20 Standby Auxiliary Feedwater 33013-1238 Standby Auxiliary Feedwater (FW) P & ID C-381-042-1 C-381-042-2 C-381-042-3 C-381-043-1 C-21 Standby Auxiliary Feedwater 33013-1238 Standby Auxiliary Feedwater (FW) P & ID C-22 Standby Auxiliary Feedwater C-381-042-4 CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 Fifth Ten-Year Inspection Interval Rev. 00 0 cteIttion 4eDF Date: December 23, 2009 Inservice Inspection Plan Page 85 of 87 1>%jRc X RjEGn n a Table~ 10. Co pnn and~- Ismeri DrI Inde C-381-043-4 33013-1238 Standby Auxiliary Feedwater (FW) P & ID C-381-042-6 C-22A SAFW Pump Recirc to CD Supply Tk 33013-1238 Standby Auxiliary Feedwater (FW) P & ID C-381-042-5 C-381-043-5 C-23 SAFW Discharge Crossover 33013-1238 Standby Auxiliary Feedwater (FWV) P & ID C-381-041 C-24 Standby Auxiliary Feedwater 33013-1238 Standby Auxiliary Feedwater (FW) P & ID C-381-043-1 C-381-043-2 C-381-043-3 C-25 Standby Auxiliary Feedwater 33013-1238 Standby Auxiliary Feedwater (FW) P & ID C-351-025-1 C-351-025-2 Station Service Cooling Water Safety Related 0-26 Standby Auxiliary Feedwater 33013-1250-2 (SW) P & ID C-381-024-1 C-381-024-2 Station Service Cooling Water Safety-Related C-27 Standby Auxiliary Feedwater 33013-1250-2 (SW) P & ID C-381-024-2 C-381-024-3 C-381-025-3 C-381-025-4 C-28 Standby Auxiliary Feedwater 33013-1238 Standby Auxiliary Feedwater (FW) P & ID CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 C jointventueoFifth Ten-Year Inspection Interval Rev. 00 0 "eDate: December 23, 2009 En-w Inservice Inspection Plan Page 86 of 87 CENG R. E. Ginna Nuclear Power Plant CENG-GNPP-ISI-005 ot ven Fifth Ten-Year Inspection Interval a oin ventureof Rev. 00 0 Co " Date: December 23, 2009 Inservice Inspection Plan Page 87 of 87~~ ~'A~> 1 TblIOACT CmppneriarfiskomtrOic Drawnci ndI~~~ ~~~ Design. DrtwgNo~ '~C-39 "A" Edg Exhaust & Supports 33013-1239 C-40 "B" Edg Exhaust & Supports 33013-1239 D-1260-7 Auxiliary Coolant Component Cooling Water C-41 A & B Component Cooling Water HTX's 33013-1245 (AC) P & ID Auxiliary Coolant Component Cooling Water C-42 Component Cooling Water Surge Tank 33013-1245 (AC) P & ID Auxiliary Coolant Component Cooling Water 33013-1246-2 (AC) P & ID Chemical & Volume Control Letdown (CVCS) P C-43 Non-Regenerative Heat Exchanger (formerly B-2) 33013-1264

& ID Auxiliary Coolant Component Cooling Water 33013-1246-2 (AC) P & ID Auxiliary Building Chemical Volume Control C-44 Seal Water Heat Exchanger (formerly B-4) 33013-1265-2 System (CVCS) P & ID Auxiliary Building Chemical Volume Control C-45 Volume Control Tank (former B-5) 33013-1265-2 System (CVCS) P & ID Chemical & Volume Control Letdown (CVCS) P 33013-1264

&ID Auxiliary Building Chemical Volume Control D-1 Filter (formerly B-3) 33013-1265-2 System (CVCS) P & ID HE-5 & HE-6 Feedwater-High Energy 33013-1236-2 Feedwater System (FW) P& ID HE-7 & HE-7A Main Steam -High Energy 33013-1232 Main Steam (MS) P & ID ATTACHMENT (2)ASME CODE, SECTION XI RELIEF REQUEST -- ISI-01 R.E. Ginna Nuclear Power Plant December 30, 2009 CONSTELLATION NUCLEAR R.E. GINNA NUCLEAR POWER PLANT REQUEST FOR ALTERNATIVE ISI-01 Application of ASME Code Case N-716 RISK-INFORMED

/ SAFETY-BASED INSERVICE INSPECTION PROGRAM PLAN Table of Contents 1. Introduction 1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 1.2 PRA Quality 2. Proposed Alternative to Current Inservice Inspection Programs 2.1 ASME Section XI 2.2 Augmented Programs 3. Risk-Informed

/ Safety-Based ISI Process 3.1 Safety Significance Determination 3.2 Failure Potential Assessment 3.3 Element and NDE Selection 3.3.1 Additional Examinations 3.3.2 Program Relief Requests 3.4 Risk Impact Assessment 3.4.1 Quantitative Analysis 3.4.2 Defense-in-Depth

4. Implementation and Monitoring Program 5. Proposed ISI Program Plan Change 6. References/Documentation ATTACHMENT A -R.E. Ginna PRA Quality Review Page 1 of 33 CONSTELLATION NUCLEAR R.E. GINNA NUCLEAR POWER PLANT REQUEST FOR ALTERNATIVE ISI-01 1. INTRODUCTION R.E. Ginna Nuclear Power Plant (REGNPP) is currently in the fourth inservice inspection (ISI)interval as defined by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Section Xl Code for Inspection Program B. REGNPP plans to implement a risk-informed I safety-based inservice inspection (RIS_B) program beginning with the first inspection period of the fifth ISI interval.

The fifth interval will commence January 1, 2010 for REGNPP.The ASME Section Xl code of record for the fourth ISI interval at REGNPP is the 1995 Edition with 1996 Addenda for Examination Category B-F, B-J, C-F-I, and C-F-2 Class 1 and 2 piping components.

The ASME Section Xl code of record for the fifth ISI interval at REGNPP will be the 2004 Edition for these welds.The objective of this submittal is to request the use of the RISB process for the inservice inspection of Class 1 and 2 piping. The RISB process used in this submittal is based upon ASME Code Case N-716, Alternative Piping Classification and Examination Requirements, Section Xl Division 1, which is founded in large part on the RI-ISI process as described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, Revised Risk-Informed Inservice Inspection Evaluation Procedure.

1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 As a risk-informed application, this submittal meets the intent and principles of Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis," and Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking

/nservice Inspection of Piping. Additional information is provided in Section 3.4.2 relative to defense-in-depth.

1.2 Probabilistic Safety Assessment (PSA) Quality The Ginna PRA has been updated to meet ASME PRA Standard RA-Sb-2005 and Regulatory Guide (RG) 1.200, Revision 1. The updated PRA model meets the Capability Category II supporting requirements (SRs) and combined Category II and Category III SRs where both requirements are equivalent (e.g., SR IF-D5a). Based on this updated PRA, several internal floods and high energy line breaks (HELB) were added to the PRA model to meet the ASME PRA Standard.

There are no internal floods piping segments or HELB piping segments in the high safety significant (HSS) scope with a Core Damage Frequency (CDF) > 1 E-06/yr or a Large Early Release Frequency (LERF) >1 E-07 year.Page 2 of 33 An industry peer review of the updated PRA model was conducted in June 2009. The peer review utilized the process described in Nuclear Energy Institute document NEI-05-04 and the ASME PRA Standard, including consideration of the NRC staff positions provided in Appendix A of RG 1.200, Revision 1. A summary of the findings and the impact of those findings on the PRA model are provided (see Attachment A of this document).

2. PROPOSED ALTERNATIVE TO CURRENT ISI PROGRAMS 2.1 ASME Section XI ASME Section Xl Examination Categories B-F, B-J, C-F-l, and C-F-2 currently contain requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components.

The alternative RISB Program for piping is described in Code Case N-716. The RIS_B Program will be substituted for the current program for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-1 and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety. Other non-related portions of the ASME Section Xl Code will be unaffected.

2.2 Augmented Programs The impact of the RISB application on the various plant augmented inspection programs listed below were considered.

This section documents only those plant augmented inspection programs that address common piping with the RIS_B application scope (e.g., Class 1 and 2 piping)." The original plant augmented inspection program for high-energy line breaks outside containment is unaffected by this application.

• The plant augmented inspection programs previously implemented in response to NRC Bulletins 88-08, Thermal Stresses in Piping Connected to Reactor Coolant Systems, and 88-11, Pressurizer Surge Line Thermal Stratification, are subsumed by the new RIS-B Program since the thermal fatigue concerns addressed by these bulletins were explicitly considered in this application process.* The plant augmented inspection program for flow accelerated corrosion (FAC) per GL 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning, is relied upon to manage this damage mechanism.

The RISB Program will not be used to eliminate any GL 89-08 requirements.

• A plant augmented inspection program is being implemented at REGNPP in response to MRP-139, Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guidelines.

The requirements of MRP-139 will be used for the inspection and management of PWSCC susceptible welds and will supplement the RISB Program selection process. The RIS_B Program will not be used to eliminate any MRP-139 requirements.

Page 3 of 33 REGNPP has conducted an evaluation in accordance with MRP-146, Materials Reliability Program: Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines, and these results have been incorporated into the RISB Program.3. RISK-INFORMED

/ SAFETY-BASED ISI PROCESS The process used to develop the RIS_B Program conformed to the methodology described in Code Case N-716 and consisted of the following steps:* Safety Significance Determination

• Failure Potential Assessment

• Element and NDE Selection" Risk Impact Assessment" Implementation Program" Feedback Loop 3.1 Safety Significance Determination The systems assessed in the RIS_B Program are provided in Table 3.1. The piping and instrumentation diagrams and additional plant information including the existing plant ISI Program were used to define the piping system boundaries.

Per Code Case N-716 requirements, piping welds are assigned safety-significance categories, which are used to determine the treatment requirements.

High safety-significant (HSS) welds are determined in accordance with the requirements below.Low safety-significant (LSS) welds include ail other Class 2, 3, or Non-Class welds.(1) Class 1 portions of the reactor coolant pressure boundary (RCPB), except as provided in 10 CFR 50.55a(c)(2)(i) and (c)(2)(ii);

(2) Applicable portions of the shutdown cooling pressure boundary function.

That is, Class 1 and 2 welds of systems or portions of systems needed to utilize the normal shutdown cooling flow path either: (a) As part of the RCPB from the reactor pressure Vessel (RPV) to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; or (b) Other systems or portions of systems from the RPV to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds;Page 4 of 33 (3) That portion of the Class 2 feedwater system [> 4 inch nominal pipe size (NPS)] of pressurized water reactors (PWRs) from the steam generator to the outer containment isolation valve;(4) Piping within the break exclusion region (> NPS 4) for high-energy piping systems as defined by the Owner. This may include Class 3 or Non-Class piping; and (5) Any piping segment whose contribution to CDF is greater than 1 E-06 (and per NRC feedback on the Grand Gulf and DC Cook RISB pilot applications 1 E-07 for LERF) based upon a plant-specific PSA of pressure boundary failures (e.g., pipe whip, jet impingement, spray, inventory losses). This may include Class 3 or Non-Class piping.3.2 Failure Potential Assessment Failure potential estimates were generated utilizing industry failure history, plant-specific failure history, and other relevant information.

These failure estimates were determined using the guidance provided in EPRI TR-112657 (i.e., the EPRI RI-ISI methodology).

One deviation to the EPRI RI-ISI methodology has been implemented in the failure potential assessment for REGNPP. Table 3-16 of EPRI TR-1 12657 contains criteria for assessing the potential for thermal stratification, cycling, and striping (TASCS). Key attributes for horizontal or slightly sloped piping greater than NPS 1 include: 1. The potential exists for low flow in a pipe section connected to a component allowing mixing of hot and cold fluids; or 2. The potential exists for leakage flow past a valve, including in-leakage, out-leakage and cross-leakage allowing mixing of hot and cold fluids; or 3. The potential exists for convective heating in dead-ended pipe sections connected to a source of hot fluid; or 4. The potential exists for two phase (steam/water) flow; or 5. The potential exists for turbulent penetration into a relatively colder branch pipe connected to header piping containing hot fluid with turbulent flow;AND AT > 50 0 F, AND Richardson Number > 4 (this value predicts the potential buoyancy of a stratified flow)These criteria, based on meeting a high cycle fatigue endurance limit with the actual AT assumed equal to the greatest potential AT for the transient, will identify locations where stratification is likely to occur, but allows for no assessment of severity.

As such, many locations will be identified as subject to TASCS where no significant potential for thermal Page 5 of 33 fatigue exists. The critical attribute missing from the existing methodology that would allow consideration of fatigue severity is a criterion that addresses the potential for fluid cycling. The impact of this additional consideration on the existing TASCS susceptibility criteria is presented below.> Turbulent Penetration TASCS Turbulent penetration typically occurs in lines connected to piping containing hot flowing fluid. In the case of downward sloping lines that then turn horizontal, significant top-to-bottom cyclic ATs can develop in the horizontal sections if the horizontal section is less than about 25 pipe diameters from the reactor coolant piping. Therefore, TASCS is considered for this configuration.

For upward sloping branch lines connected to the hot fluid source that turn horizontal or in horizontal branch lines, natural convective effects combined with effects of turbulence penetration will keep the line filled with hot water. If there is no potential for in-leakage towards the hot fluid source from the outboard end of the line, this will result in a well-mixed fluid condition where significant top-to-bottom ATs will not occur. Therefore, TASCS is not considered for these configurations.

Even in fairly long lines, where some heat loss from the outside of the piping will tend to occur and some fluid stratification may be present, there is no significant potential for cycling as has been observed for the in-leakage case. The effect of TASCS will not be significant under these conditions and can be neglected.

> Low flow TASCS In some situations, the transient startup of a system (e.g., shutdown cooling suction piping) creates the potential for fluid stratification as flow is established.

In cases where no cold fluid source exists, the hot flowing fluid will fairly rapidly displace the cold fluid in stagnant lines, while fluid mixing will occur in the piping further removed from the hot source and stratified conditions will exist only briefly as the line fills with hot fluid. As such, since the situation is transient in nature, it can be assumed that the criteria for thermal transients (TT) will govern.> Valve leakage TASCS Sometimes a very small leakage flow of hot water can occur outward past a valve into a line that is relatively colder, creating a significant temperature difference.

However, since this is generally a "steady-state" phenomenon with no potential for cyclic temperature changes, the effect of TASCS is not significant and can be neglected.

> Convection Heating TASCS Similarly, there sometimes exists the potential for heat transfer across a valve to an isolated section beyond the valve, resulting in fluid stratification due to natural convection.

However, since there is no potential for cyclic temperature changes in this case, the effect of TASCS is not significant and can be neglected.

Page 6 of 33 In summary, these additional considerations for determining the potential for thermal fatigue as a result of the effects of TASCS provide an allowance for considering cycle severity.

The above criteria have previously been submitted by EPRI to the NRC for generic approval [letters dated February 28, 2001 and March 28, 2001, from P.J.O'Regan (EPRI) to Dr. B. Sheron (USNRC), Extension of Risk-Informed Inservice Inspection Methodology].

The methodology used in the REGNPP RIS_B application for assessing TASCS potential conforms to these updated criteria.

Final materials reliability program (MRP) guidance on the subject of TASCS has been incorporated into the REGNPP RISB application in accordance with MRP-146. It should be noted that the NRC has granted approval for RI-ISI relief requests incorporating these TASCS criteria at several facilities, including Comanche Peak (NRC letter dated September 28, 2001)and South Texas Project (NRC letter dated March 5, 2002).Table 3.2 summarizes the failure potential assessment by system for each degradation mechanism that was identified as potentially operative.

3.3 Element and NDE Selection Code Case N-716 and lessons learned from the Grand Gulf and DC Cook RISB pilot applications provide criteria for identifying the number and location of required examinations.

Ten percent of the HSS welds shall be selected for examination as follows: (1) Examinations shall be prorated equally among systems to the extent practical, and each system shall individually meet the following requirements: (a) A minimum of 25% of the population identified as susceptible to each degradation mechanism and degradation mechanism combination shall be selected.(b) If the examinations selected above exceed 10% of the total number of HSS welds, the examinations may be reduced by prorating among each degradation mechanism and degradation mechanism combination, to the extent practical, such that at least 10% of the HSS population is inspected.(c) If the examinations selected above are not at least 10% of the HSS weld population, additional welds shall be selected so that the total number selected for examination is at least 10%.(2) At least 10% of the RCPB welds shall be selected.(3) For the RCPB, at least two-thirds of the examinations shall be located between the inside first isolation valve (IFIV) (i.e., isolation valve closest to the RPV) and the RPV.(4) A minimum of 10% of the welds in that portion of the RCPB that lies outside containment (OC) (e.g., portions of the main feedwater system in BWRs) shall be selected.(5) A minimum of 10% of the welds within the break exclusion region (BER) shall be selected.Page 7 of 33 Currently, there are one hundred fifteen BER program welds at REGNPP. These will be examined in accordance with the existing augmented program, which is unaffected by this new RISB program.In contrast to a number of RI-ISI Program applications where the percentage of Class 1 piping locations selected for examination has fallen substantially below 10%, Code Case N-716 mandates that at least 10% be chosen. A brief summary is provided below, and the results of the selections are presented in Table 3.3. Section 4 of EPRI TR-1 12657 was used as guidance in determining the examination requirements for these locations.

s (1) (2) Class 3 andQ(4 Class 1 Welds.1) Class 2 Welds(2) Class Welds (3) All Piping Welds(4)Total Selected Total Selected Total Selected Total Selected 543 56 958 14 74 7 1575 77 Notes (1) Includes all Category B-F and B-J locations.

All 543 Class 1 piping weld locations are HSS.(2) Includes all Category C-F-1 and C-F-2 locations.

Of the 958 Class 2 piping weld locations, 138 are HSS and the remaining 820 are LSS (3) These 74 Class 3 and non-nuclear safety (0 Class) piping weld locations are HSS as a result of the BER Program (there are four Class 3 in main steam,, 52 NNS in feedwater and 18 NNS in main steam).(4) Regardless of safety significance, Class 1 and 2 in-scope piping components will continue to be pressure tested as required by the ASME Section Xl Program. VT-2 visual examinations are scheduled in accordance with the station's pressure test program that remains unaffected by the RIS_B Program.3.3.1 Additional Examinations If the flaw is from original construction or otherwise acceptable, Code rules do not require any additional inspections.

Any service-induced flaW that exceeds the dimensions of allowable flaws in ASME Code Section X1 IWB/C/D-3500 is noted as unacceptable.

The noted unacceptable service-induced flaw can subsequently be accepted by supplemental examinations, evaluation or repair/replacement.

Supplemental examinations shall be performed in accordance with IWB/C/D-3200.

Analytical evaluation of flaws shall be per the requirements of ASME Code Section Xl, IWB/C/D-3600.

As part of performing evaluation to IWB/C/D-3600, the degradation mechanism that is responsible for the flaw will be determined and accounted for in the evaluation.

If the nature and type of the flaw is service-induced, then similar systems or trains will be examined.

If the unacceptable service-induced flaw is accepted by analytical evaluation then appropriate successive examinations shall be scheduled per IWB/C/D-2420(b) and (c). If the evaluation that was performed under IWB/C/D-3600 determines the flaw is found unacceptable for continued operation, it shall be repaired/replaced in accordance with IWA-4000 and/or applicable ASME Section XI Code Cases. The need for extensive root cause analysis beyond that required for IWB/C/D-3600 evaluation will be dependent on practical considerations (i.e., the practicality of performing additional NDE or removing the flaw for further evaluation during the outage). The NRC is involved in the Page 8 of 33 process at several points. For preemptive weld overlays, a relief request in accordance with 10 CFR 50.55a(a)(3) is usually required for design and installation.

Should a flaw be discovered during an examination, a notification in accordance with 10 CFR 50.72 or 10 CFR 50.73 may be required.

IWB-3134(b) and IWB-3144(b) require the analytical evaluation to be submitted to the NRC.Finally, the evaluation will be documented in the corrective action program and the Owner submittals required by Section Xl.The evaluation will include whether other elements in the segment or additional segments are subject to the same root cause conditions.

Additional examinations will be performed on those elements with the same root cause conditions or degradation mechanisms.

The additional examinations will include HSS elements up to a number equivalent to the number of elements required to be inspected during the current outage. If unacceptable flaws or relevant conditions are again found similar to the initial problem, the remaining elements identified as susceptible will be examined during the current outage. No additional examinations need be performed if there are no additional elements identified as being susceptible to the same root cause conditions.

3.3.2 Program Relief Requests An attempt has been made to select RISB locations for examination such that a minimum of >90% coverage (i.e., Code Case N-460 criteria) is attainable.

However, some limitations will not be known until the examination is performed since some locations may be examined for the first time by the specified techniques.

In instances where locations at the time of the examination fail to meet the >90%coverage requirement, the process outlined in 10 CFR 50.55a will be followed.Consistent with previously approved RI-ISI submittals, REGNPP will calculate coverage and use additional examinations or techniques in the same manner it has for traditional Section Xl examinations.

Experience has shown this process to be weld-specific (e.g., joint configuration).

As such, the effect on risk, if any, will not be known until that time. Relief requests will be submitted per the guidance of 10CFR50.55a(g)(5)(iv) within one (1) year after the end of the interval.No REGNPP relief requests are being withdrawn due to the RIS_B application.

3.4 Risk Impact Assessment The RISB Program development has been conducted in accordance with Regulatory Guide 1.174 and the requirements of Code Case N-716, and the risk from implementation of this program is expected to remain neutral or decrease when compared to that estimated from current requirements.

Page 9 of 33 This evaluation categorized segments as high safety significant or low safety significant in accordance with Code Case N-716, and then determined what inspection changes are proposed for each system. The changes include changing the number and location of inspections and in many cases improving the effectiveness of the inspection to account for the findings of the RISB degradation mechanism assessment.

For example, examinations of locations subject to thermal fatigue will be conducted on an expanded volume and will be focused to enhance the probability of detection (POD)during the inspection process.3.4.1 Quantitative Analysis Code Case N-716 has adopted the EPRI TR-1 12657 process for risk impact analyses whereby limits are imposed to ensure that the change in risk of implementing the RIS_B Program meets the requirements of Regulatory Guides 1.174 and 11178. The EPRI criterion requires that the cumulative change in CDF and LERF be less than 1E-07 and 1E-08 per year per system, respectively.

For LSS welds, CCDP and CLERP values of 1E-4 and 1E-5 are generally conservatively used, unless pipe segments in the plant internal flooding study are found with higher values. With the exception of the main steam system, a review of the internal flood study results confirmed that CCDP and CLERP values are less than the assumed the 1 E-4 and 1 E-5 values used for CCDP and CLERP. A separate CCDP and CLERP are assigned to low safety significant main steam (LSSMS) elements in the table below.With respect to assigning failure potential for LSS piping, the criteria are defined by Table 3 of the N-716 Code Case. That is, those locations identified as susceptible to FAC (or another mechanism and also susceptible to water hammer) are assigned a high failure potential.

Those locations susceptible to thermal fatigue, erosion-cavitation, corrosion or stress corrosion cracking are assigned to a medium failure potential and those locations that are identified as not susceptible to degradation are assigned a low failure potential.

In order to streamline the risk impact assessment, a review was conducted to verify that the LSS piping was not susceptible to FAC or water hammer. This review was conducted similar to that done for a traditional RI-ISI application.

Thus, the High failure potential category is not applicable to LSS piping. In lieu of conducting a formal degradation mechanism evaluation for all LSS piping (e.g.to determine if thermal fatigue is applicable), these locations were conservatively assigned to the Medium failure potential

('"Assume Medium" in Table 3.4) for use in the change-in-risk assessment.

Experience with previous industry RI-ISI applications shows this to be conservative.

REGNPP has conducted a risk impact analysis per the requirements of Section 5 of Code Case N-716 that is consistent with the "Simplified Risk Quantification Method" described in Section 3.7 of EPRI TR-1 12657. The analysis estimates Page 10 of 33 the net change in risk due to the positive and negative influences of adding and removing locations from the inspection program.The conditional core damage probability (CCDP) and conditional large early release probability (CLERP) values used to assess risk impact were estimated based on pipe break location.

Based on these estimated values, a corresponding consequence rank was assigned per the requirements of EPRI TR-1 12657 and upper bound threshold values were used as provided below.Consistent with the EPRI risk-informed methodology, the upper bound for all break locations that fall within the high consequence rank range was based on the highest CCDP value obtained (e.g., Large LOCA for REGNPP) for pipe breaks in the scope of N-716.CCDP and CLERP Values Based on Break Location Break Location Estimated Consequence Upper Bound Designation CCDP CLERP Rank CCDP CLERP LOCA 3.OE-02 3.OE-03 HIGH 3.0E-02 3.OE-03 RCPB pipe breaks that result in a loss of coolant accident -The highest CCDP for Large LOCA (2.4E-2) was used (0.1 margin used for CLERP). Unisolated RCPB piping of all sizes.ILOCA) I 9.OE-05 9.OE-06 MEDIUM 1.OE-04 1.OE-05 Isolable LOCA (1 open valve) -RCPB pipe breaks that result in an isolable LOCA -Calculated based on Large LOCA CCDP of 3E-2 and valve fail to close probability of -3E-3 (0.1 margin used for CLERP). Between 1 st and 2nd isolation valve on charging, letdown and pressurizer relief.PLOCA(1) 3.0E-05 3.0E-06 MEDIUM 1.OE-04 1.0E-05 Potential LOCA (1 closed valve) -RCPB pipe breaks that result in a potential LOCA -Calculated based on Large LOCA CCDP of 3E-2 and valve rupture probability of -IE-3 (0.1 margin used for CLERP). Piping between 1st and 2nd isolation valves on safety injection, alternate charging, drain lines on RC PLOCASD(1) 9.OE-05 9.0E-06 MEDIUM 1.OE-04 1.OE-05 Potential LOCA (1 valve) -RCPB pipe breaks that occur in shutdown cooling piping result in a potential LOCA at power and isolable LOCA during shutdown (between MOVs or CVs). LOCA CCDP and MOV failure on demand is judged to be appropriate for lines inside containment (0.1 margin used for CLERP)PPLOCASD 1) 1 9.0E-06 9.OE-07 MEDIUM 1.OE-04 1.OE-05 Potential LOCA (2 valves) -Class 2 potential LOCA in shutdown cooling piping downstream of 2nd isolation valve (suction MOVs) and upstream of 2nd isolation valve (return lines check valves). LOCA CCDP and failure of 2 MOVs to close on demand (3E-4) is judged to be appropriate for lines inside containment (0.1 margin used for CLERP)IPLOCA) I <lE-06 <lE-07 MEDIUM L.OE-04 1.OE-05 Class 2 isolable-potential LOCA breaks that require two valves in series to fail (one rupture and one fail to close) -based on Large LOCA CCDP of 3E-2 and 2 valve failures 3E-6 (0.1 margin used for CLERP). Vent and drain lines on CVCS letdown.Seal 2.OE-05 2.OE-06 MEDIUM L.OE-04 1.OE-05 RCP seal injection lines require a seal LOCA to occur -based on small LOCA CCDP of 2E-4 and 0.1 probability of seal LOCA (0.1 margin used for CLERP)Iseal <lE-06 <lE-07 MEDIUM L.OE-04 1.OE-05 Isolable RCP seal injection lines require a seal LOCA to occur -based on small LOCA CCDP of 2E-4 and 0.1 probability of seal LOCA and 3E-3 valve failure probability (0.1 margin used for CLERP)Page 11 of 33 CCDP and CLERP Values Based on Break Location Break Location Estimated Consequence Upper Bound Designation CCDP CLERP Rank CCDP CLERP SLB 3.OE-03 3.OE-04 HIGH 3.OE-02 3.OE-03 BER main steam and feedwater breaks (numerous initiating events) with several CCDP and CLERP -used bounding CCDP (0. 1 margin used for CLERP)LSSMS 2.OE-03 2.OE-04 HIGH 3.OE-02 3.OE-03 Class 2 low safety significant main steam breaks (numerous initiating events) with several CCDP and CLERP -used bounding CCDP (0.1 margin used for CLERP)Class 2 LSS 1.OE-04 I .OE-05 MEDIUM 1.OE-04 1.OE-05 Class 2 pipe breaks that occur in the remaining system piping designated as low safety significant

-Estimated based on upper bound for Medium Consequence.

Notes 1. The Ginna PRA does not explicitly model potential and isolable LOCA events, because such events are subsumed by the LOCA initiators in the PRA. That is, the frequency of a LOCA in this limited piping downstream of the first RCPB isolation valve times the probability that the valve fails is a small contributor to the total LOCA frequency.

The N-716 methodology must evaluate these segments individually; thus, it is necessary to estimate their contribution.

This is estimated by taking the LOCA CCDP and multiplying it by the valve failure probability.

2. Although the calculated CCDP and CLERP values for the IPLOCA and Iseal break locations fall in the "Low" consequence rank range, a "Medium" consequence rank is conservatively used for risk impact.The likelihood of pressure boundary failure (PBF) is determined by the presence of different degradation mechanisms and the rank is based on the relative failure probability.

The basic likelihood of PBF for a piping location with no degradation mechanism present is given as x, and is expected to have a value less than 1E-08. Piping locations identified as medium failure potential have a likelihood of 20x 0.These PBF likelihoods are consistent with References 9 and 14 of EPRI TR-1 12657. In addition, the analysis was performed both with and without taking credit for enhanced inspection effectiveness due to an increased POD from application of the RISB approach.Table 3.4 presents summaries of the RISB Program versus ASME Section Xl program requirements on a per system basis, as indicated below.1995 Edition of ASME Code with 1996 Addenda for the selection of Category B-F, B-J, C-F-1 and C-F-2 piping welds for the Fourth Interval ISI Program.As indicated in the following table, this evaluation has demonstrated that unacceptable risk impacts will not occur from implementation of the RIS_B Program, and satisfies the acceptance criteria of Regulatory Guide 1.174 and Code Case N-716. In addition, sensitivity cases with the estimated CCDP and CLERP values in the above table show that the change is risk is similar and meets the acceptance criteria.

Also, for cases where the RISB selections exceeded SXI selections in Table 3.4, they were set equal to SXI to confirm that Page 12 of 33 the use of conservative CCDP and CLERP are not non conservative relative to meeting the acceptance criteria.REGNPP Risk Impact Results With POD Credit Without POD Credit SystemWihPDCei_______

Delta CDF Delta LERF Delta CDF Delta LERF Chemical & Volume Control -1.04E-10

-1.04E-11

-4.OOE-11

-4.OOE-12 Main Feedwater O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 HPSI (SI) 3.01E-10 3.01E-11 3.01E-10 3.01E-11 Main Steam 2.13E-08 2.13E-09 2.13E-08 2.13E-09 Reactor Coolant -6.93E-08

-6.93E-09

-2.25E-08

-2.25E-09 Residual Heat Removal -5.OOE-13

-5.OOE-14

-5.OOE-13

-5.00E-14 Auxiliary Feedwater O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Containment Spray O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Total -4.78E-08

-4.78E-09

-9.39E-10

-9.39E-11 3.4.2 Defense-in-Depth The intent of the inspections mandated by ASME Section Xi for piping welds is to identify conditions such as flaws or indications that may be precursors to leaks or ruptures in a system's pressure boundary.

Currently, the process for selecting inspection locations is based upon structural discontinuity and stress analysis results. As depicted in ASME White Paper 92-01-01 Rev. 1, Evaluation of Inservice Inspection Requirements for Class 1, Category B-J Pressure Retaining Welds, this method has been ineffective in identifying leaks or failures.

EPRI TR-1 12657 and Code Case N-716 provide a more robust selection process founded on actual service experience with nuclear plant piping failure data.This process has two key independent ingredients; that is, a determination of each location's susceptibility to degradation and secondly, an independent assessment of the consequence of the piping failure. These two ingredients assure defense-in-depth is maintained.

First, by evaluating a location's susceptibility to degradation, the likelihood of finding flaws or indications that may be precursors to leak or ruptures is increased.

Secondly, a generic assessment of high-consequence sites has been determined by Code Case N-716 supplemented by plant-specific evaluations thereby requiring a minimum threshold of inspection for important piping whose failure would result in a LOCA or BER break. Finally, Code Case N-716 requires that any piping on a plant-specific basis that has a contribution to CDF of greater than 1 E-06 (or 1 E-07 for LERF) be included in the scope of the application.

REGNPP did not identify any such piping.All locations within the Class 1, 2, and 3 pressure boundaries will continue to be pressure tested in accordance with the Code, regardless of its safety significance.

Page 13 of 33

4. IMPLEMENTATION AND MONITORING PROGRAM Upon approval of the RISB Program, procedures thatcomply with the guidelines described in Code Case N-716 will be prepared to implement and monitor the program. The new program will be implemented at the beginning of the Fifth ISI Interval.

No changes to the Technical Specifications or Updated Final Safety Analysis Report are necessary for program implementation.

The applicable aspects of the ASME Code not affected by this change will be retained, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements.

Existing ASME Section XI program implementing procedures will be retained and modified to address the RIS_B process, as appropriate.

The monitoring and corrective action program will contain the following elements: A. Identify B. Characterize C. (1) Evaluate, determine the cause and extent of the condition identified (2) Evaluate, develop a corrective action plan or plans D. Decide E. Implement F. Monitor G. Trend The RIS_B Program is a living program requiring feedback of new relevant information to ensure the appropriate identification of HSS piping locations.

As a minimum, this review will be conducted on an ASME period basis. In addition, significant changes may require more frequent adjustment as directed by NRC Bulletin or Generic Letter requirements, or by industry and plant-specific feedback.For preservice examinations, REGNPP will follow the rules contained in Section 3.0 of N-716.Welds classified HSS require preservice inspection.

The examination volumes, techniques, and procedures shall be in accordance with Table 1. Welds classified as LSS do not require preservice inspection.

5. PROPOSED ISI PROGRAM PLAN CHANGE A comparison between the new Fifth ISI interval RISB Program and previous Fourth ISI Interval ASME Section Xl program requirements for in-scope piping is provided in Table 5, as indicated below.Page 14 of 33 1995 Edition of ASME Code with 1996 Addenda was used for the selection of Category B-F, B-J, C-F-1 and C-F-2 piping welds for the Fourth Interval ISI Program.REGNPP will implement the new program at the beginning of the Fifth ISI Interval.Page 15 of 33

6. REFERENCES/DOCUMENTATION EPRI Report 1006937, Extension of EPRI Risk Informed ISI Methodology to Break Exclusion Region Programs.EPRI TR-112657, Revised Risk-Informed Inservice Inspection Evaluation Procedure, Rev. B-A.ASME Code Case N-716, Alternative Piping Classification and Examination Requirements, Section X1 Division 1. April 19, 2006.Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis.Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping.Regulatory Guide 1.200, Rev 1 "An Approach For Determining The Technical Adequacy Of Probabilistic Risk Assessment Results For Risk-Informed Activities." USNRC Safety Evaluation for Grand Gulf Nuclear Station Unit 1, Request for Alternative GG-ISI-002-1mplement Risk-Informed ISI based on ASME Code Case N-716, dated September 21, 2007.USNRC Safety Evaluation for DC Cook Nuclear Plant, Units 1 and 2, Risk-Informed Safety-Based ISI program for Class 1 and 2 Piping Welds, dated September 28, 2007 Supporting Onsite Documentation SI Calculation No. 0800470.302, N-716 Evaluation of Ginna, Revision 0.SI Calculation No. 0800470.301, Degradation Mechanism Evaluation for Ginna, Revision 2.Page 16 of 33 AFW BER cc CCDP CDF CLERP CS CVCS DMs ECSCC E-C FAC FW FW1 FW2 FW3 FW4 HPSI HSS IFIV IGSCC ILOCA IPLOCA LERF LOCA LSS MIC MS MS1 MS2 MSIV OC PIT PLOCA PLOCA -OC PLOCASD PPLOCA PPLOCASD POD PWR: FW PWSCC RC RCPB RCPBIFIV RCPBOC RHR RI-ISI RIS_B SDC SI SLB SXI List of Acronyms Auxiliary Feedwater System Break Exclusion Region Crevice Corrosion Condition Core Damage Probability Core Damage Frequency Condition Large Early Release Probability Containment Spray System Chemical & Volume Control System Degradation Mechanisms External Chloride Stress Corrosion Cracking Erosion-Cavitation Flow-Accelerated Corrosion Feedwater System Feedwater piping between steam generator and first isolation valve inside containment Feedwater piping between first isolation valve inside containment and containment penetration Feedwater piping outside containment between containment penetration and isolation valve Feedwater piping outside containment upstream of the outer isolation valve High Pressure Safety Injection System High Safety Significant Inside First Isolation Valve Intergranular Stress Corrosion Cracking Isolable Loss of Coolant Accident Isolable Potential Loss of Coolant Accident Large Early Release Frequency Loss of Coolant Accident Low Safety Significant Microbiologically-Influenced Corrosion Main Steam System Main Steam piping outside containment between containment penetration and MSIV Main Steam piping outside containment downstream of MSIV Main Steam Isolation Valve Outside Containment Pitting Potential Loss of Coolant Accident Potential Loss of Coolant Accident -Outside Containment Potential Loss of Coolant Accident (Shutdown Function)Potential Loss of Coolant Accident (2 Valves)Potential Loss of Coolant Accident (Shutdown Function, 2 Valves)Probability of Detection Pressurized Water Reactor: Feedwater Primary Water Stress Corrosion Cracking Reactor Coolant System Reactor Coolant Pressure Boundary Reactor Coolant Pressure Boundary Inside First Isolation Valve Reactor Coolant Pressure Boundary Outside Containment Residual Heat Removal System Risk-Informed Inservice Inspection Risk-Informed

/ Safety-Based Inservice Inspection Shutdown Cooling Safety Injection System Steam Line Break Section XI Page 17 of 33 TASCS Thermal Stratification, Cycling, and Striping TGSCC Transgranular Stress Corrosion Cracking TT Thermal Transients Vol/Sur Volumetric and Surface Sur Surface Page 18 of 33 Table 3.1 N-716 Safety Significance Determination N-716 Safety Significance Determination Safety SytmDsrpin Weld Significance System Description Count CDF > 1E-6 RCPB SDC PWR: FW BER LERF>1E-7 High Low CVCS -Chemical & Volume 237 1 'Control 24 *" 16 " " 1 FW -Feedwater 52 " " 46 " 1 HPSI -High Pressure Safety 52 V I Injection 357 " 47 1"v MS -Main Steam 62 13 1" *" *RC -Reactor Coolant 216 1 " 25 V, V, RHR -Residual Heat Removal 51 " " 214 _ _AFW -Auxiliary Feedwater 63 1 CS -Containment Spray 100 V" 38 V ., 505 1V

SUMMARY

51 1 V RESULTS 16 11 FOR ALL SYSTEMS 99 V " 46 " I 820 _" TOTALS 1575 Page 19 of 33 Table 3.2 Failure Potential Assessment Summary Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive System(l)

TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC CVCS(2) V FW HPSI(2)MS(2)RC I/ v RHR (2)AFW(2)CS(2)Notes 1. Systems are described in Table 3.1.2. A degradation mechanism assessment was not performed on low safety significant piping segments.

This includes the AFW and CS systems in their entirety, as well as portions of the CVCS, HPSI, MS and RHR systems.Page 20 of 33 Table 3.3 N-716 Element Selections Weld N716 Selection Considerations System Count Selections HSS LSS DMs RCPB RCPB (IFIV) RCPB (OC) BER CVCS 55 IT T 8 CVCS 28 None V V 16 CVCS 154 None / 0 CVCS 24 0 FW 68 None V 12 FW 46 None 0 HPSI 6 IGSCC V 2 HPSI 46 None V 4 HPSI 357 0 MS 47 None V 5 MS 62 0 RC 9 TT,TASCS V V 3 RC 35 TT T V 9 RC 15 TASCS V V 4 RC 153 None V V 6 RC 17 None I 1 RHR 25 None V_ 3 RHR 51 None 4 RHR 214 0 AFW 63 0 CS 100 0 35 TT V V 9 55 IT T 8 15 TASCS V V 4 Summary 9 TT,TASCS V V 3 Results 6 IGSCC V" 2 All 181 None V V 22 Systems 242 None V 8 115 None V 17 97 None 4 820 0 Totals 755 820 77 Note 1. Systems are described in Table 3.1.Page 21 of 33 Table 3.4: Risk Impact Analvsis Results Safety Break Failure Potential Inspections CDF Impact LERF Impact System Significance Location DMs Rank SXI RIS B Delta w/POD w/o POD w/POD w/o POD CVCS High ILOCA TT Medium 0 3 3 -5.40E-11

-3.OOE-11

-5.40E-12

-3.OOE-12 CVCS High PLOCA TT Medium 0 5 5 -9.00E-1 I -5.OOE-1 1 -9.OOE-12

-5.OOE-12 CVCS High ILOCA None Low 0 0 0 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 CVCS High PLOCA None Low 0 0 0 0.00E+00 O.OOE+00 0.OOE+00 0.00E+00 CVCS High IPLOCA None Low 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 CVCS High SEAL None Low 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 CVCS High ISEAL None Low 0 0 0 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 CVCS Low LSS Assume Medium 4 0 -4 4.OOE- I1 4.00E- 11 4.00E-12 4.00E-12 CVCS Total -1.04E-10

-4.OOE-11

-1.04E-11

-4.OOE-12 FW Total High SLB None Low 12 12 0 O.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 HPSI High PLOCA IGSCC Medium 2 0 -2 2.00E-11 2.00E-11 2.OOE-12 2.OOE-12 HPSI High PLOCA None Low 3 1 -2 1.00E-12 1.OOE-12 1.00E-13 1.OOE-13 HPSI Low LSS Assume Medium 28 0 -28 2.80E-10 2.80E-10 2.80E-11 2.80E- 11 HPSI Total 3.01E-10 3.01E-10 3.01E-11 3.01E-11 MS High SLB None Low 7 5 -2 3.OOE-10 3.OOE-10 3.OOE- 1 3.OOE-11 MS Low LSSMS Assume Medium 7 0 -7 2.1OE-08 2.1OE-08 2.1OE-09 2.1OE-09 MS Total 2.13E-08 2.13E-08 2.13E-09 2.13E-09 RC High LOCA TASCS Medium 0 4 4 -2.16E-08

-1.20E-08

-2.16E-09

-1.20E-09 RC High LOCA TT,TASCS Medium 2 3 1 -1.26E-08

-3.OOE-09

-1.26E-09

-3.OOE-10 RC High LOCA TT Medium 5 9 4 -3.96E-08

-1.20E-08

-3.96E-09

-1.20E-09 RC High LOCA None Low 34 4 -30 4.50E-09 4.50E-09 4.50E-10 4.50E-10 RC High ILOCA None Low 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 RC High PLOCA None Low 1 0 -1 5.00E-13 5.OOE-13 5.OOE-14 5.OOE-14 RC Total 0 -6.93E-08

-2.25E-08

-6.93E-09

-2.25E-09 RHR High PLOCASD None Low 6 3 -3 1.50E-12 1.50E-12 1.50E-13 1.50E-13 RHR High PPLOCASD None Low 0 4 4 -2.OOE-12

-2.OOE-12

-2.OOE-13

-2.OOE-13 RHR Low LSS Assume Medium 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 RHR Total -5.OOE-13

-5.OOE-13

-5.OOE-14

-5.OOE-14 AFW Total Low LSS Assume Medium 0 0 0 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 CS Total Low LSS Assume Medium 0 0 0 O.OOE+00 O.OOE+00 0.00E+00 O.OOE+00 Grand Total 111 53 4.78E-08

-9.39E-10

-4.78E-09

-9.39E-11 Notes 1.2.Systems are described in Table 3.1.The "LSS" break location designation in Table 3.4 is used to identify those Code Section 2(a) of N-716 (e.g., not part of the BER scope).Class 2 locations that are not HSS because they do not meet any of the five HSS criteria of Page 22 of 33

3. The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium" or "Low" dependent upon potential susceptibly to the various types of degradation mechanisms.

[Note: LSS locations were conservatively assumed to be a rank of Medium (i.e., "Assume Medium").]

4. Only those ASME Section Xl Code inspection locations that received a volumetric examination in addition to a surface examination are included in the count. Inspection locations previously subjected to a surface examination only were not considered in accordance with Section 3.7.1 of EPRI TR-1 12657.Page 23 of 33 Table 5: Inspection Location Selections Comparison Between ASME XI and RIS_B System Safety Significance Break Failure Potential Code Weld Section XI Code Case N716 High Low Location DMs Rank Category Count Vol Surface RISB Other CVCS IV, ILOCA TT Medium B-J 34 0 13 3 CVCS I PLOCA TT Medium B-J 21 0 11 5 CVCS V ILOCA None Low B-J 29 0 7 0 CVCS *" PLOCA None Low B-J 22 0 14 0 CVCS IPLOCA None Low B-J 4 0 1 0 CVCS " SEAL None Low B-J 28 0 3 0 16 VT-2 CVCS " ISEAL None Low B-J 99 0 11 0 CVCS I LSS N/A Assume Medium B-J 24 4 1 0 FW " SLB None Low C-F-2, Q 114 12 0 12 HPSI "' PLOCA IGSCC Medium B-J 6 2 0 0 2 VT-2 HIPSI " PLOCA None Low B-J 46 3 8 1 3 VT-2 HPSI V LSS N/A Assume Medium B-J 357 28 20 0 MS " SLB None Low C-F-2, Q 47 7 0 5 MS " LSSMS N/A Assume Medium C-F-2 62 7 3 0 RC V LOCA TASCS Medium B-J 15 0 3 4 RC " LOCA TT,TASCS Medium B-J 9 2 0 3 RC -LOCA TT Medium B-F,B-J 35 5 6 9 RC -LOCA None Low B-F,B-J 153 34 16 4 2 VT-2 RC I ILOCA None Low B-J 6 0 2 0 RC "" PLOCA None Low B-J 11 1 1 0 1 VT-2 RHR PLOCASD None Low B-J 25 6 0 3 RHR PPLOCASD None Low C-F-i 51 0 0 4 RHR " LSS N/A Assume Medium C-F-1 214 0 0 0 AFW " LSS N/A Assume Medium C-F-2 63 0 0 0 CS *" LSS N/A Assume Medium C-F-i 100 0 0 0 Notes 1. Systems are described in Table 3.1.2. The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium" or "Low" dependent upon potential susceptibly to the various types of degradation mechanisms.

[Note: LSS locations were conservatively assumed to be a rank of Medium (i.e., "Assume Medium").]

3. The column labeled "Other" is generally used to identify plant augmented inspection program locations credited per Section 4 of Code Case N-716. Code Case N-716 allows the existing plant augmented inspection programrfor IGSCC (Categories B through G) in a BWR to be credited toward the 10% requirement.

This option is not applicable for the REGNPP RISB application.

The "Other" column has been retained in this table solely for uniformity purposes with other RISB application template submittals and to identify the number of RIS_B VT-2 exam selections not credited in the risk impact assessment similar to Section Xl Surface only exams, which are also not credited.Page 24 of 33 ATTACHMENT A REGNPP Probabilistic Risk Assessment Quality Review Introduction Constellation Energy Nuclear Group (CENG) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating CENG nuclear generation sites. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to Probabilistic Risk Assessment (PRA).PRA Maintenance and Update The CENG risk management process ensures that the applicable PRA model remains an accurate reflection of the as-built and as-operated plants. This process is defined in the CENG risk management program, which consists of a governing procedure CNG-CM-2.01, Probabilistic Risk Assessment (PRA) Directive, and subordinate implementation procedures.

CENG procedure CNG-CM-1.01-3 003,"Probabilistic Risk Assessment Configuration Control" delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating CENG nuclear generation sites.CNG-CM- 10 1-3004, "PRA Process for Internal Evaluations", includes the process to meet the overall CENG risk management program, including CNG-CM- 1.01-3003, defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operational experience), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plant, the following activities are routinely performed:

• Design changes and procedure changes are reviewed for their impact on the PRA model.PRA screens are required for all design and procedure changes.9 New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.* Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated based upon percentage changes in PRA failure rates or frequencies.

This has recently been noted as an area for improvement in our configuration control procedure.

This will be revised to require a data update on a periodic frequency as well.In addition to these activities, CENG risk management procedures provide the guidance for particular risk management and PRA quality and maintenance activities.

This guidance includes: " Documentation of the PRA model, PRA products, and bases documents." The approach for controlling electronic storage of Risk Management (RM) products including PRA update information, PRA models, and PRA applications.

• Guidelines for updating the full power, internal events PRA models for CENG nuclear generation sites." Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (1OCFR50.65 (a)(4)).Page 25 of 33 ATTACHMENT A REGNPP Probabilistic Risk Assessment Quality Review PRA Peer Review Revision 6.5a of the PRA model was subjected to an industry Peer Review during the week of June 1, 2009. The Peer Review resulted in 25 findings.

All of the findings that could impact Risk-Informed ISI have been incorporated into PRA Model 7.0. This model is the living evaluation tool that contains the modeling resolution of the current set of open CRMPs (i.e. open items identified, but not completely documented in our RG 1.200 notebooks).

This provides the ability to track the cumulative risk impact of all of the open items. Each of the Peer Review findings as well as any open CRMPs that could impact Risk-Informed ISI are listed below (each of the 25 findings have 2009 Peer Review in the title and list the Supporting Requirement (SR)): CRMP 699: Add AOV 1 12A transferring to the hold-up tank as a VCT diversion This issue was corrected in the model used for Risk-Informed ISI.CRMP 702: Loss of CCW without Letdown Isolation could lead to an Excessive Charging Temperatures This issue was corrected in the model used for Risk-Informed ISI.CRMP 768: Correct the maintenance alignment term for TIOOPSDCA.

This issue was corrected in the model used for Risk-Informed ISI.CRMP 791: 2009 Peer Review SY-Al 1: Add logic for operator actions to align bleed and feed given less than 2 PORVs available.

This finding was corrected in the model used for Risk-Informed ISI.CRMP 797: 2009 Peer Review SY-A 13: Model City Water DG Support for all Systems not just GE-Betz This finding was corrected in the model used for Risk-Informed ISI.CRMP 800: 2009 Peer Review HR-G3: Provide a More Detailed Basis and Operator Interviews for Key Operator Actions All of the key operator actions identified by the Peer Review team were reviewed for technical adequacy and interviewed with operators.

Any changes identified were made and the events re-calculated.

The updated values were incorporated into the model used for Risk-Informed ISI. Re-calculating these HRA events had no significant impact on the PRA results.CRMP 802: 2009 Peer Review HR-I1: Completely populate the HRA Calculator for Each HRA No specific items identified for improvement beyond documentation.

Page 26 of 33 ATTACHMENT A REGNPP Probabilistic Risk Assessment Quality Review CRMP 805: 2009 Peer Review IF-B2: Perform a more systematic consideration of human-caused floods A systematic review of human induced flooding was perform and documented in the PRA model. No significant additional flooding sources were identified.

CRMP 806: 2009 Peer Review IF-C3: Document the Pipe Whip/Jet Impingement and Spray Modeling This is a documentation issue only. Expanded documentation has been incorporated in the PRA model.No modeling changes were required.CRMP 807: 2009 Peer Review IF-C3b: Model and Document the potential barrier failures due to flooding loads This is a documentation issue only. An expanded discussion of potential barrier failures has been included in the PRA model. No model changes were required.CRMP 808: 2009 Peer Review IF-C8: Improve the Documentation on the.Screening of Flooding Scenarios This is a documentation issue only. The basis for the qualiiative screening of flood scenarios has been added to the GRPA. No modeling changes were required.CRMP 809: 2009 Peer Review IF-D5a: Improve Documentation on material condition, water hammer, and maintenance induced flooding.Additional research into plant material condition, water hammer issues, and maintenance induced flooding was performed and included in the documentation.

No issues were discovered which would invalidate the use of generic flood initiation frequencies.

CRMP 810: 2009 Peer Review IF-D6: Improve the Documentation on Maintenance Induced Floods A systematic review of human induced flooding was performed and documented in the PRA model.No significant additional flooding sources were identified.

CRMP 811: 2009 Peer Review IF-D7: Improve the Documentation on Quantitative screening This is a documentation issue only. A discussion of the basis for each quantitative screening has been included in the PRA model.CRMP 812: 2009 Peer Review IF-E5: Interview and Document Flooding HRAs All flooding HRA events were re-examined and an operator interview was performed.

Several HRA events scenarios were modified based on this re-examination and/or operator input. However, all HRA values remained the same or decreased, with the exception of one value which increased by a negligible amount. As such, these changes will not adversely impact Risk-Informed ISI.Page 27 of 33 ATTACHMENT A REGNPP Probabilistic Risk Assessment Quality Review CRMP 813: 2009 Peer Review IF-F1: Document the Flood Notebook per the Fleet Procedures This is a documentation issue only. No impact on Risk-Informed ISI.CRMP 814: 2009 Peer Review IF-F3: Expand the Uncertainty Analysis in the IF Notebook This is a documentation issue only. An expanded discussion of uncertainties has been included in the PRA model.CRMP 816: 2009 Peer Review MU-D1: Database to identify applications requiring change after model update does not exist No impact on Risk-Informed ISI.CRMP 818: 2009 Peer Review DA-C 13: Improve the Modeling on Concurrent Planned Unavailabilities The current modeling is conservative.

Further, no currently allowed redundant maintenance activities are risk significant.

Conservative non-risk significant modeling will not impact Risk-Informed ISI.CRMP 828: 2009 Peer Review SY-A18: Consider Explicitly Modeling Partially Shadowed Unavailabilities to Reduce Risk The conservative modeling of these unavailabilities does not appear in the review of the top 95% of the cutsets. Conservative non-risk significant modeling will not impact Risk-Informed ISI.CRMP 834: 2009 Peer Review LE-C 10: Document why no credit need be taken for scrubbing of releases Ignoring credit for scrubbing of releases is conservative.

Conservatism in the modeling will not adversely impact Risk-Informed ISI.CRMP 835: 2009 Peer Review LE-C2a: Document why post-CDF Operator Actions would Provide no Reduction in LERF Ignoring post-CDF operator actions to prevent LERF is conservative.

Conservatism in the modeling will not adversely impact Risk-Informed iSI.CRMP 837: 2009 Peer Review LE-C9a: No credit was given to equipment survivability or human actions that could be impacted by containment failure Ignoring credit for equipment recovery is conservative.

Conservatism in the modeling will not adversely impact Risk-Informed ISI.CRMP 838: 2009 Peer Review SC-A2: Use CET >1200 F for 30 minutes versus Core Uncovery Use of core uncovery vs. CET temperature is conservative.

Conservatism in the modeling will not adversely impact Risk-Informed ISI.Page 28 of 33 ATTACHMENT A REGNPP Probabilistic Risk Assessment Quality Review CRMP 845: 2009 Peer Review IE-ClO: IE-CIO requires a comparison of plant-specific and generic initiating events and an explanation any differences Documentation issue only. Comparison table exists, but no explanation of differences was provided.No impact on Risk-Informed ISI.CRMP 847: 2009 Peer Review IE-C 13: Documentation does not provide or reference IE uncertainty data distribution No correlated uncertainties exist within initiating events that are related to Risk-Informed ISI that would affect the mean values used in the evaluation.

CRMP 855: 2009 Peer Review OU-B5: Define the Circular Logic Process and Provide more examples Documentation issue only. No impact on Risk-Informed ISI.CRMP 856: Combine Maintenance Unavailabilities within a System unless a Logical Reason Exists for Differences This finding was corrected in the model used for MSPI.Based on the results of the Ginna peer review, the PRA model with the changes made to address the findings discussed above, is of sufficient technical adequacy to support a Risk-Informed ISI application.

Flood Supporting Requirements The peer review was performed against Category II of the internal events standard.

Per EPRI document TR- 1018427 'Nondestructive Evaluation:

Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs', this is adequate for Risk-Informed ISI. However, there are three SR's where, for an ISI submittal under N-716 EPRI methodology, the NRC has generally expected that Category III be met for the SR. These are addressed below: (SR) IF C3: Capability Category III includes jet impingement, pipe whip, and humidity, condensation, and temperature concerns.

Risk Informed ISI requires that all SSC failures induced by a pipe break be considered.

The PRA model explicitly addressed the potential for jet impingement and pipe whip from steam line piping to impact the 4160 VAC electrical busses in the turbine building.

Plant arrangement drawings were reviewed and walkdowns were performed to ensure that no steam or feedwater piping is within 10 pipe diameters of these busses. Pipe breaks more than 10 pipe diameters from the busses need not be considered based on NUREG/CR-2913.

The impact of temperature, humidity, and condensation, on the 4160 VAC busses was also explicitly addressed by increasing failure rate of these busses during high temperature/humidity conditions.

Since a spatial analysis of the potential for pipe whip or jet impingement to impact other equipment in the TB or the IB has not been performed, it was assumed that all other equipment in the TB and 113 is failed by these breaks. This also addresses temperature/humidity/condensation concerns.Page 29 of 33 ATTACHMENT A REGNPP Probabilistic Risk Assessment Quality Review A similar analysis was performed to conservatively model the impact of feedwater line pipe breaks.(SR) IF-C6: This SR permits screening out of flood areas based on, in part, the success of human actions to isolate and terminate the flood. The endorsed RI-ISI methods require determination of the flood scenario with and without human intervention which corresponds to the Capability Category III (i.e. scenarios are not screened out based on human actions).Ginna's Flood Model does allow limited use of screening of flood initiators when considering human actions recoveries.

However, these are very limited in scope and would not realisitically challenge the 1E-06 criteria for core damage frequency as noted in N-716. All have frequencies which, when considering hardware, are well below 1E-06 with the following exception:

Flood Area IBN -Fire Protection flood at up to 2000 gpm.This flood has a frequency of 1.70E-05.

There are over three hours to mitigate and stop the flood prior to failing AFW Pumps. There is indication in the Control Room of the flood due to the Fire Pumps running and the flood is easily stopped by securing the pumps. The combination what would be a very reliable human action and the CCDP with MFW and feed and bleed, and standby AFW available would make the annual CDF impact orders of magnitude less then 1E-06.Thus, the Ginna Flood IE screening method does not screen any floods which could challenge the N-716 criteria.(SR) IF-D5a: This SR addresses the development of flood initiating (pipe rupture) frequencies for use during the scenario development.

Risk Informed-ISI is premised on inspecting locations with the highest risk, driven mostly by failure frequency.

The plant 'specific information collected and used should include experience related to degradation mechanisms that could indicate increased likelihood of pipe failure at particular locations.

Ginna's high energy program (e.g., main steam and feedwater) has not changed and remains an inspection of all high energy welds during the 10 year period. Therefore the Risk-Informed ISI process does not apply to the high energy portions of the plant. During the development of the risk informed inspection locations and risk informed process for the remainder of the class 1, 2, and 3 piping, operating experience from the ISI program and condition reports were reviewed for the last 10 years to determine if the pipe failure frequencies were above the generic frequencies.

Industry OE was also reviewed.

The review failed to note any cases where Ginna would be an outlier and would require use of more conservative frequencies.

PRA Configuration Control The PRA model configuration is procedurally controlled by CNG-CM-1.01-3003, "PRA Configuration Control".

This procedure provides the control and processes for maintaining the PRA model consistent with the as-operated, as-built plant.Design changes and procedure changes are monitored for impact on the PRA. Issues requiring action are entered into the PRA model Configuration Risk Management Program (CRMP) database as a CRMP Issue. These issues are prioritized in accordance with their significance for implementation into Page 30 of 33 ATTACHMENT A REGNPP Probabilistic Risk Assessment Quality Review future PRA updates. Significant CRMP Issues that are the result of errors are entered into the site corrective action program.CNG-CM-1.01-3004, PRA Process for Internal Evaluations provides the guidance for documentation of RG 1.200 compliance.

This includes all internal events (including Flood Modeling) and Fire PRA compliance documentation.

General Conclusion Regarding PRA Capability for Risk-Informed In-Service Inspection The CENG PRA maintenance and update processes and technical capability evaluations described above provide a robust basis for concluding that the PRA is suitable for use in this risk-informed process. In the risk-informed inservice inspection (RI-ISI) program at GINNA, the EPRI Risk Informed ISI methodology is used to define alternative inservice inspection requirements.

Plant-specific PRA-derived risk significance information is used during the RI-ISI plan development to support the consequence assessment, risk ranking, element selection and delta risk evaluation steps.The importance of PRA consequence results, and therefore the scope of PRA technical capability, is tempered by three fundamental components of the EPRI methodology.

First, PRA consequence results are binned into one of three conditional core damage probability (CCDP) and conditional large early release probability (CLERP) ranges before any welds are chosen for RI-ISI inspection as illustrated below. Broad ranges are used to define these bins so that the impact of uncertainty is minimized and only substantial PRA changes would be expected to have an impact on the consequence ranking results.Further, the LSS classification were conservatively binned as High Risk. None of the Medium Risk break locations remotely challenged the High classification (Highest was 1.3E-05).Consequence Results Binning Groups Consequence Category CCDP Range CLERP Range High CCDP > 1E-4 CLERP > 1E-5 Medium 1E-6 < CCDP < 1E-4 1E-7 < CLERP < 1E-5 Low CCDP < 1E-6 CLERP < 1E-7 The risk importance of a weld is therefore not tied directly to a specific PRA result. Instead, it depends only on the range in which the PRA result falls. As a consequence, any PRA modeling uncertainties would be mitigated by the wide binning provided in the methodology.

Additionally, conservatism in the binning process (e.g., as would typically be introduced through PRA attributes meeting ASME PRA Standard Capability Category I versus II) will tend to result in a larger inspection population.

Secondly, the impacts of particular PRA consequence results are further dampened by the joint consideration of the weld failure potential via a non-PRA-dependent damage mechanism assessment.

The results of the consequence assessment and the damage mechanism assessment are combined to determine the risk ranking of each pipe segment (and ultimately each element) according to the EPRI Risk Matrix, Page 31 of 33 ATTACHMENT A REGNPP Probabilistic Risk Assessment Quality Review Thirdly, the EPRI RI-ISI methodology uses an absolute risk ranking approach.

As such, conservatism in either the consequence assessment or the failure potential assessment will result in a larger inspection population rather than masking other important components.

That is, providing more realism into the PRA model (e.g., by meeting higher capability categories) most likely would result in a smaller inspection population.

These three facets of the methodology reduce the importance and influence of PRA on the final list of candidate welds.The limited manner of PRA involvement in the RI-ISI process is also reflected in the risk-informed license application guidance provided in Regulatory Guide 1.174. Section 2.2.6 of Regulatory Guide 1.174 provides the following insight into PRA capability requirements for this type of application: "There are, however, some applications that, because of the nature of the proposed change, have a limited impact on risk, and this is reflected in the impact on the elements of the risk model.An example is risk-informed inservice inspection (RI-ISI).

In this application, risk significance was used as one criterion for selecting pipe segments to be periodically examinedfor cracking.

During the staff review it became clear that a high level of emphasis on PRA technical acceptability was not necessary.

Therefore, the staff review ofplant-specific RI-ISI typically will include only a limited scope review of PRA technical acceptability." Conclusion Regarding PRA Capability for Risk-Informed 1SI The Ginna PRA model is suitable for use in the risk informed in-service inspection application.

This conclusion is based on: " The PRA maintenance and update processes in place,* Successful peer review to category II Page 32 of 33 ATTACHMENT A REGNPP Probabilistic Risk Assessment Quality Review PRA Quality References

1. NEI-00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," Rev.A3.2. Deleted 3. U.S. Nuclear Regulatory Commission, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Regulatory Guide 1.200, Revision 1, January 2007.4. WCAP-16464-NP, "Westinghouse Owner's Group Mitigating Systems Performance Index Cross Comparison," Revision 0, August 2005.5. Deleted.6. American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-Sb-2005, New York, New York, December 2005.7. U.S. Nuclear Regulatory Commission Memorandum to Michael T. Lesar from Farouk Eltawila, "Notice of Clarification to Revision 1 of Regulatory Guide 1.200," for publication as a Federal Register Notice, July 27, 2007.8. Revised Risk-Informed Inservice Inspection Evaluation Procedure, EPRI TR-1 12657, Revision B-A, December 1999.9. U.S. Nuclear Regulatory Commission, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 1, November 2002.10. Nondestructive Evaluation:

Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs, EPRI TR-10 18427, December 2008.Page 33 of 33