ML15334A244

Official Exhibit - ENT000641-00-BD01 - Final License Renewal Interim Staff Guidance LR-ISG-2011-04, Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors (May 28, 2013)

ML15334A244
Person / Time
Site:	Indian Point
Issue date:	08/10/2015
From:	Entergy Nuclear Operations
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 28134, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15334A244 (159)

Text

FINAL LICENSE RENEWAL INTERIM STAFF GUIDANCE LR-ISG-2011-04 UPDATED AGING MANAGEMENT CRITERIA FOR REACTOR VESSEL INTERNAL CO MPONENTS FOR PRESSURIZED WATER REACTORS INTRODUCTION This license renewal interim staff guidance (LR-ISG) updates the U.S. Nuclear Regulatory Commission (NRC's) guidance in NUREG-1800, Revision 2, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants," (SRP-LR) and NUREG-1801, Revision 2, "Generic Aging Lessons Learned Report" (GALL Report). This LR-ISG is primarily based on the issuance of Revision 1 to the Final Safety Evaluation (SE) of Electric Power Research Institute (EPRI) Report, Materials Reliability Program (MRP) Report 1016596 (MRP-227), Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," by letter dated December 16, 2011 (SE, Revision 1, on MRP-227) (ADAMS Accession No. ML11308A770). After the issuance of the staff's SE, Revision 1, on MRP-227, EPRI Technical Report No. 1022863 (MRP-227-A) was published in January 2012. MRP-227-A

is the NRC-endorsed version of MRP-227, which incorporates the NRC staff's SE, Revision 1, on MRP-227. Specifically, this LR-ISG revises the recommendations in the GALL Report and the NRC staff's acceptance criteria and review procedures in the SRP-LR to ensure consistency with MRP-227-A. This LR-ISG also provides a framework to ensure that PWR license renewal applicants will adequately address age-related degradation and aging management of reactor vessel internal (RVI) components during the term of the renewed license.

DISCUSSION Current Regulatory Framework

Pursuant to Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants," Section 21(a)(3), of Title 10 of the Code of Federal Regulations (10 CFR 54.21(a)(3)), a license renewal applicant is required to perform an integrated plant assessment (IPA) that demonstrates that the effects of aging on structures and components subject to an aging management review (AMR) are adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis (CLB) for the period of extended operation.

The NRC's guidance in SRP-LR Section 3.0.1 defines the AMR as the identification of the structure and component materials, environments, aging effects, and aging management programs (AMPs) credited for managing the aging effects. In turn, SRP-LR Section A.1.2.3 defines an acceptable AMP as consisting of 10 elements. In addition, 10 CFR 54.21(d) requires the license renewal application (LRA) to contain a final safety analysis report (FSAR) supplement that includes a summary description of the programs and activities for managing the effects of aging and the evaluation of time-limited aging analyses (TLAAs) for the period of extended operation.

GALL Report AMP X I.M 16A, "PWR Vessel Inter nals," provides recomme ndations for an AMP to manage the effects of ag ing for PWR RVI compo nents. In ad dition, the GALL Report provides component-specific AMR items for PWR RVI c o mponents in the following tables:

• Table IV.B2 for Westing house-desig ned RVI components

• Table IV.B3 for CE-designed RVI components

• Table IV.B4 for B&W-designed RVI components SRP-LR Ta ble 3.1-1 pr ovides the specific commodity grou p-based AMR items for PWR RVI components. SRP-LR Sections 3.1.

2.2.1, 3.1.2.

2.3, 3.1.2.2.

9, 3.1.2.2.10 , 3.1.2.2.12, 3.1.2.2.13, and 3.1.2.2.

14 provide the aging management review results for which f u rther evaluation is recommend ed by the GALL Report for PWR RVI components. Finally, S R P-LR Tabl e 3.0-1 provides an example of t he type of informati on to be included in the FSAR Supplement for an AMP for PW R RVI comp onents. Basis for Issuing Interim Guidance On January 12, 2009, EPRI submitted Technica l Report No. 1016596, "Materials Reliability Program: Pressurized Water Reactor Internals I n spection a nd Evaluation Guidelines (MRP-227, Revision 0)," for NRC st aff review and approval. On June 22, 2011, the NRC staff issued its SE on MRP-22 7, Revision 0, which con t ained specific topical re port conditio n items (TRCIs) on the use of MRP-227 and Applicant/Licen see Action It ems (A/LAIs) that must be addressed by those applicants o r licensee s utilizing this topical repor

t. The staff i ssued a revi sion of it s SE on the report methodology (i.e., SE, Revision 1, on MRP-227) by letter dated December 16, 2011.

MRP-227-A, the NRC-en dorsed version of MRP-227, was later published in January 2012 and provides guidance for a PWR licensee or licen se renewal applicant to u s e in the de velopment and implementation of a n AMP for RVI components. MRP-2 27-A also in corporates A/LAIs that are to be ad dressed if th is report is referenced to satisfy the r equirements of 10 CFR 54.21(a)(3) for demonstrating that th e effects of aging on the RVI compo nents, within the scope o f MRP-227, will be adeq uately mana ged. The st aff recomme nds that a P W R license renewal applicant provide its responses to these A/LAIs in Appendix C of the LRA. The use of MRP-2 27-A by a PWR license renewal applicant is n o t a substitu te for performing a plant-specific IPA to identify those struct ures and co mponents subject to an aging management revi ew, in accor dance with 10 CFR 54.21(a)(1).

Regulatory Issue Summary (RIS) 20 11-07, "Lice n se Renewal Submittal Information For Pressurized Water Reactor Internals Aging Management," dated July 21, 2011, was issued, in part, to facilitate a predictable and co nsistent met hod for reviewing the aging management of RVI components for commercial PWR LRAs. An "inspection p l a n" is one aspect of the A/LAIs of the staff's SE, Revision 1, fo r MRP-227.

This "inspe ction plan" p r ovides information about the RVI components to be inspe c ted and a d e scription of how they wi ll be managed for age-related degradation.

Details of a n "inspectio n plan" for t hose PWR plant licen se es that have not submitted but plan to su bmit an LRA in the future will be incor porated into the LRA as part of the 10-element aging management program and aging management revie w line items. Thus, consiste nt with RIS 2011-07, these fu ture license renewal applicants n eed not submit a separate document that contains an "inspect i on plan" in r e sponse to t he A/LAIs of the staff's SE, Revision 1, for MRP-227. Prior to the completion of its review and issuan ce of the SE on MRP-22 7, the staff issued SRP-LR, Re vision 2, an d GALL Rep o rt, Revision 2, in Dece mber 2010. Since SRP-LR, Revision 2, and GALL Report, Revision 2, were based on MRP-227, Revi sion 0, the r e levant portions of t he SRP-LR, Revision 2, and GALL Report, Revision 2, are no w being updated with this LR-ISG to reconcile any differences with MRP-227-A.

ACTION This LR-ISG updates the GALL Repo rt, Revision 2, and SRP-LR, Revision 2, to ensur e consiste ncy with MRP-2 27-A for the aging management of age-related d egradation f o r PWR RVI components during the t e rm of a renewed operating licen se. Appendix A, "Revisions to the GALL Report and SRP-L R ," to this L R-ISG shows these cha nges. The majority of these changes result in the in corporation o f MRP-227-A within the SRP-LR, Re vision 2, an d the GALL Report, Revi sion 2. To b e tter show these chang es, a mark-up is shown in Appendix B, "Mark-Up of Changes to the GALL Report and SRP-LR," to this LR-ISG.

On March 2 0 , 2012, at Volume 77, page 16270, of the Fede ral Register (77 FR 16270), the NRC reques ted public comments on draft LR-ISG-2011-04. Subsequently, as noticed on April 19, 2012, at Volume 77, page 23513, of the Federal Register (77 FR 23513), the NRC issued an e d itorial corre ction to the original notice to specifically identify the ADAMS Accession Nos. for additional documents asso cia t ed with draft LR-ISG-20 11-04. The staff received comments on draft LR-ISG-20 11-04 by letters from EPRI and the Pressurized Water Reactor Owners Group Mate rials Subco mmittee (ADAMS Accession No. ML12146A267) and from the Nuclear Energy Institute (ADAMS Accession No. ML12144A147). The staff considered all comments, and its ev aluation of t hese comments is cont ained in App endix C, "Staff Response to Public Comment s on Draft License Rene wal Interim Staff Guidance 2011-04," of this LR-ISG. The guidance descr ibed in this f i nal LR-ISG supersedes the affected sections of the SRP-LR, Revision 2, and the GALL Report, Revision 2, and is appro v ed for use by the NRC staff and sta k eholders. NEWLY IDE N TIFIED SY STEMS, ST RUCT URES , AND C O MPON ENTS UNDE R 10 CFR 54.37(b) Any structures and components ide n tified in th is LR-ISG as requiring aging management, which were not previously identified in earlier versi ons of the SRP-LR, Revision 2, or GALL Report, Revision 2, are consider ed by the staff to be ne wly-identified structures and components under 10 CFR 54.37(b).

BAC KFITTING A ND IS SUE FIN A LITY This LR-ISG contains gu idance on o ne acceptab le approach for managing the associated aging effects durin g the PEO f o r components within th e scope of license rene wal. The st aff's discussion o n compliance with the requirements of the Backfit Rule, 10 CFR 50.109 is presented below. Compliance with the Backfit Rule and Issue Fina lity Issuance of this LR-ISG does not co nstitute backfitting a s d e fined in 10 CFR 50.109(a)(1), and the NRC staff did not prepare a backfit analysis for issuing this LR-ISG. T here are several rationales fo r this con c lu sion, depen ding on the status of the nuclear po wer plant licensee.

Licensees currently in th e license re newal process

- The ba ckfitting pro v isions in 10 CFR 50.109 do not prote c t an applicant, as backfitting policy consideratio ns are not a pplicable t o an applicant. T herefore, issuance of this LR-ISG do es not const i tute backf itting as define d in 10 CFR 50.109(a)(1). There current ly are no combined lice n ses (i.e., 1 0 CFR Part

52) license renewal applicants; ther efore, the ch anges and n e w positions presented in the LR-ISG ma y be made without considera t ion of the issue finality provisions in 10 CFR Part 52, "Licenses, Certification s , and Approvals for Nuclear Power Plants."

Licensees who already h o ld a renewed license

- This guidan ce is nonb in ding and the LR-ISG would not require curren t holders of r enewed lice n ses to take any action (i.e., progra mmatic or plant hardware changes for managing the associated aging e ffects for co mponents within the scope of this LR-ISG).

Current holders of renewed license s should treat this guidan ce as operating experience an d take actio n s as appro p riate to ensure that applicable agin g managemen t programs are, and will remain, effective.

If, in the future, the NRC decides to take additional a c tion and im pose require ments for managing the associat ed aging effect s for components within the scope of this LR-ISG, the n the NRC would follow the requirements of the Backfit Rule. Current operating licen se or com b in ed license h o lders who have not yet applied for r enewed license s - T he backfitt i n g provisions in 10 CFR 50.109 do n o t protect an y future applicant, as backfitting p o licy consid erations are not applicab le to a future applicant.

Therefore, issuance of this LR-ISG does not co nstitute backfitting a s d e fined in 10 CFR 50.109(a)(1). The issue fina lity provisions o f 10 CFR Pa rt 52 do not extend to the aging management ma tters covered by 10 CFR Part 54, as evidenced by the requirement in 10 CFR 52.107, "Application f o r Renewal,"

stating that application s for renewal of a combined license must be in accordance with 10 CFR Part

54. APPE NDIC E S Appendix A provides the staff's revisions to the S R P-LR, Re vi sion 2, and t he GALL Re port, Revision 2, for managing aging in P W R RVI co mponents and include s the following section s:

• Section 1 -

Revised version of the GALL Report

• Section 2 -

Revised version of the S R P-LR Appendix B provides a mark-up of the SRP-LR, Revision 2, and GALL Report, Revision 2, to better show the changes made as a result of LR-I SG-2011-04 and include s the following sections:

• Section 1 -

Mark-up of changes to th e GALL Rep o rt

• Section 2 -

Mark-up of changes to th e SRP-LR Appendix C provides the staff's ba se s for resolv ing comment s that were received on the draft LR-ISG-201 1-04. REFER E NC ES 1. U.S. Code o f Federal Regulations , "D omestic Licensing of P r oduction an d Utilization Facilit ies," P a rt 50, Chapter I, Title 10 , "Energy."

2. U.S. Code o f Federal Regulations , "L icenses, Certification s , a nd Approvals for Nuclea r Power Plants," Part 52, Chapter I, Title 10, "Energy."

3. U.S. Code o f Federal Regulations , "R equiremen t s for Renewal of Operating Licen s es for Nuclear Power Plants," Part 54, Chapter I, Title 10, "Energy.

" 4. U.S. Nuclear Regulatory Commissio n , "Generic Aging Lesso ns Learned (GALL) Rep o rt," NUREG-18 01, Revision 2, December 2010, ADAMS Accession No. ML103490041.

5. U.S. Nuclear Regulatory Commissio n , "Standard Review Pla n for Review of License Renewal Ap plications for Nuclear Power Plants," NUREG-18 00, Revision 2, December 2010, ADAMS Accession No. ML1 03490036.

6. U.S. Nuclear Regulatory Commissio n , Final Saf e ty Evaluati on of EPRI Report, Mat e rials Reliability Program Repo rt 1016596 (MRP-227), Revision 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, June 22, 20 11, ADAMS Accession No. ML111600498.

7. U.S. Nuclear Regulatory Commissio n , Revision 1 to the Fina l Safety Eval uation of Electric Power Research Institute (E PRI) Report, Materials R eliabili ty Pro g ram (MRP) Report 1016596 (MRP-2 27), Revisio n 0, Pressur i zed Water Reactor Internals Inspe c tion and Evaluation Guidelin es , December 16, 2011, ADAMS Ac cession No.

ML11308A7 70. 8. Electric Power Research Institute, E P RI Technical Report No. 1016596, Materials Reliability Program: Pre ssurized Wa ter Reactor Internals Inspection and Evaluation Guidelines (M RP-227 Revision 0), December 2008, ADAMS Accession No. ML090160204 (Cover le tter from EPRI MR P) an d ADAMS Accession No. ML090160206 (Final Report).

9. Electric Power Research Institute, E P RI Technical Report No. 1022863, Materials Reliability Program: Pre ssurized Wa ter Reactor Internals Inspection and Evaluation Guidelines (M RP-227-A), December 2011, ADAMS Accession No. ML1 2017A193 (Transmittal letter from the EPRI-MRP) and ADAMS Accession Nos. ML12017A194, ML12017A1 96, ML12017A197, ML1 2017A191, ML12017A1 92, ML12017A195 and ML12017A1 99 (Final Report).

10. U.S. Nuclear Regulatory Commissio n , "Updated Aging Management Crit eria for Reactor Vessel Inter nal Compon ents of Pressurized Wat e r Reactors," Federal Register, Vol. 7 7 , No. 54, March 20, 2012, pp. 16270-16271.

11. U.S. Nuclear Regulatory Commissio n , "Updated Aging Management Crit eria for Reactor Vessel Inter nal Compon ents of Pressurized Wat e r Reactors," Federal Register, Vol. 7 7 , No. 76, April 19, 2012, p

p. 23513. 12. T. Wells and E. Fernandez, Electri c Power Rese arch Institut e Materials Reliability Program an d the Pressurized Wate r Reactor Owners Grou p Materials Subcommitt ee, letter to Document Control Desk, U.S. Nuclear Regulatory Commission, May 21, 201 2, ADAMS Accession No.

ML12146A2 67. 13. M. Richter, Nuclear Energy Institute, letter to Cin d y K. Blade y, U.S. Nucl ear Regulatory Commission , May 21, 20 12, ADAMS Accession No. ML12144A147.

14. U.S. Nuclear Regulatory Commissio n , Nuclear Regulatory Commission Regulatory Issue Summary 2 011-07, Lice nse Renewal Sub m ittal Inform ation For Pressurized Water Reactor Internals Aging Manage m e n t , July 21, 2 011, ADAMS Accession No. ML1119 90086. 15. U.S. Nuclear Regulatory Commissio

n. 2008. Memorandum from Dale E. Klein, Chairman, t o Hubert T. Bell, Office of the Inspe c tor General, "Response to Recommen dation 8 of 9/6/07 Audit Report on NRC's License Renewal Program."

(April 1, 200 8). ADAMS Accession No. ML080870286.

A-1 Appendi x A REVISI O N S TO THE G A LL REPO RT AND SRP-LR A-2 Appendix A, Section 1 - Revised version of the GALL Re port (1) Revised ver s ion of GALL Report AMP XI.

M 16 A XI.M16A PWR VESSEL INTERNALS Program Description This progra m relies on implementati on of the Electric Power Research I n stitute (EPRI) Technical Report No. 1022863, "Materials Reliability Program:

Pressurized Wat e r Reactor (PWR) Internals Inspection a nd Evaluation Guidelines," (MRP-227-A) and EPRI Technical Report No. 1016609, "Materials Reliability Program: Inspection St anda rd for PWR Internals," (M RP-228) t o manage the aging eff e cts on the pressurized water reactor (PWR) rea c tor vessel internal (RVI) components. The reco mmended a c tivities in M R P-227-A a nd additiona l plant-spe cific activitie s not defined in MRP-227-A are implemented in accordance with Nuclear Energy Institute (NEI) 03-08, "Guideline for the Managemen t of Materials Issue s." T he staff appr oved the augmented inspection and evaluation (I&E) criteria fo r PWR RVI compon ents in NRC Safety E v aluation (SE), Re vision 1, on MRP-227 by letter dated December 16, 2011.

This progra m is used to manage the effects of a ge-related d egradation mechanisms that are applicable in general to t he PWR RVI components at the facility. These aging effects include: (a) cracking, including st ress corros i on crackin g (SCC), primary water stress corros i o n cracking (PWSCC), irradiation-assisted stress corrosion cr acking (IAS CC), and cracking due t o fatigue/cyclic loading; (b) loss of mat e rial induce d by wear; (c) loss o f fra c ture toughn ess due to either thermal aging or n eutron irradiation embri ttlement; (d) changes in dimensions d ue to void swelling or d i stortion; an d (e) loss of preload due to thermal and irradiatio n-enhanced stress relaxation or creep.

The program applies th e guidance in MRP-227-A for inspect i ng, evaluating, and, if a pplicable, disposit ionin g non-confo r ming RVI c o mponents at the facility. These examinations provide reasonable assurance t hat the effects of age-rel a ted degradation mechanisms will b e managed during the p e riod of extended operation. The program includes expanding periodic examination s and other inspection s, if the extent of the degra dation identified exceeds the expected levels.

MRP-227-A guidance for selectin g RVI components for inclu s ion in the inspection sample is based on a f our-step ran k ing proce s s. Through this proce s s, the RVIs for all three P W R designs wer e assigned t o one of the following fou r groups: "Primary," "Expansion," "E xisting Programs,"

and "No Additional Measures." Defini tions of ea ch group ar e provided in "Generic Aging Lesso ns Learned Report" (GALL Report), Revision 2, Chapter IX

.B. The result of this four-ste p sample se lection pro c ess is a set of "Primary" internals co mponent locations for each of the three plant d e signs that are inspecte d because t hey are exp e cted to show the le ading indica tions of the degradation effects, with another set of "Expansion" internals component locations tha t are specif ied to expand the sample should the indication s be more severe than anticipated.

The degradation effects in a third se t of internals location s ar e deemed to be adequat ely managed by "Existing Programs," such as American Society of Mechanical Engineers A-3 (ASME) Co de,Section X I , 11 Exa m ination Category B-N-3, e x a m inations of core suppor t structures.

A fourth set of internals locations are deemed to require "No Additional Measures."

Evaluation and Technical Basis

1. Scope of Program:

The scope o f the program includes all RVI components based on the plant's app licable nucle ar steam supply system design. The scope of th e program applies the methodology and guidance in MRP-227-A, which provides an augmented inspect i on and flaw evaluation methodology for assurin g the functio nal integrity of safety-related internals in commercial operating U.S. PWR nuclear power plants design ed by Babcock & Wilcox (B&W), Combustion Engineerin g (CE), and Westinghou se. The sco pe of components considered for inspection in MRP-227-A includes co re support st ructures, tho s e RVI components that serve an intended license renewal safety function pursua n t to criteria in 10 CFR 54.4(a)(1), and other RVI components whose failu re could pre v ent satisfacto ry accomplishment of any of the functions identified i n 10 CFR 54.4(a)(1)(i), (ii), or (iii). In a ddition, ASME Code,Section X I includes inspection r equirements for PWR removable core support str u ctures in T able IWB-2500-1, Exa m ination Category B-N-3, which are in additi on to any inspection s th at are implemented in accordance with MRP-2 27-A. The scope o f the program does not include con s umable items, such a s f uel assemblies, reactivity control assemblies, and n u clear instru mentation. The scope o f the program also does not include welded attachments to the internal surface of the reactor vessel becau se these components are conside r ed to be ASME Code Class 1 appurt enances to t he reactor vessel and are managed in accorda n ce with an applicant's AMP that co rresponds to GALL AMP X I.M 1, "ASME Cod e ,Section XI Inservice I n spection, S ubsection s I W B, IWC, and IWD."

2. Preventive Actions:

MRP-227-A relies on PWR water chemistry control to pre v ent or mitigate aging effects th at can be in duced by corrosive aging mechanisms (e.g., loss of material induced by general, pitting corrosion , crevice cor r osion, or str e ss corrosio n cracking o r any of its forms [SCC, PWSCC, or IASCC]). Reacto r coola n t water chemistry is mo nitored and maintained in accordance with the Water Chemistry Progra m , as describe d in GALL AMP X I.M 2, "Wat er Chemistry."

3. Parameters M onitored/Inspected:

The program manag es the follow i ng age-relat ed degradation effects and mechanisms that are ap plicable in g eneral to RVI components at the facility: (a) cracking ind u ced by SCC, PWSCC, IASCC, or fatigue/cyclic loading; (b) loss of material induced by wear; (c) loss o f fracture to ughness ind u ced by either thermal aging or neutron irra diation embrittlement; (d) changes in dimensions due to void swelling or d i stortion; and (e) loss of preload d ue to thermal and irradiation-enhan ced stress r e laxation or creep.

For the management of cracking, th e program moni tors for evidence of surface-bre a king linear discontinu i ties if a visual inspection t e chnique is used as the non-destruct i ve exa m ination (NDE) method, or for relevant flaw presentation signals if a volumetric ultrasonic testing (UT) method is used as the NDE metho d. For the managemen t of loss of material, the program monitors for gross or abn ormal surface condition s that may be indicative of loss of mat e rial occurr ing in the components. For the management of loss of pr eload, the pr ogram monit o rs for gross surface conditions t hat may be i ndicative of loosening in applicable b o lted, fasten ed, keyed, or pinned connection

s. The program does not directly monitor for loss of fracture t oughness th at is induced by thermal aging or neutron irradiation e m brittlement. Instead, th e impact of loss of

11 Refer to the GALL Re port, Chapter I, for ap plica b i lit y of various e d itio ns o f the ASME Code

,Section XI.

A-4 fracture tou ghness on component integrity is i ndirectly mana ged by: (1) using visual or volumetric examination techniques t o monitor for cracking in the components, and (2) applying applicable r educed fract u re toughness propertie s in the flaw evaluations, in case s where cracking is d e tected in th e components and is ex t ensive enough to necessitate a sup p lemental flaw growth or flaw tolerance evaluation.

The pr ogram uses physical measurements to monitor for any dime nsional ch a nges due to void swellin g or distortio

n. Specifically, the program implem ents the parameters monitored/inspecte d criteria co nsistent with the applicable tab l e s in Section 4, "Aging Management Requiremen t s," in MRP-227-A.

4. Detection of Aging Effects:

The inspect i on methods are defined an d establishe d in Section 4 of MRP-227-A.

Standards for impleme n ting the inspection methods are def ined and establishe d in MRP-228.

In all ca ses, well-established inspe c tion methods are sele ct ed. These methods include volume tric UT examination me thods for det ecting flaws in bolting an d various visual (VT-3, VT-1, and EVT-1) exa m inations fo r detecting e ffects rangin g from general conditions t o detection and sizing o f surface-br eaking disco n tinuities. Surface examinations may also be used as an alternative to visual examinations fo r detection a nd sizing of surface-breaking discontinuitie

s. Cracking ca used by SCC, IASCC, and fatigue is monitored/inspected by either VT-1 or EVT-1 examination (for internals other than bolting) or b y volu metric UT exa m ination (boltin g). VT-3 visual methods may be applied for t he detection of cracking in non-redu ndant RVI components only when the flaw toler ance of the component, as evaluated for reduce d fracture to ughness properties, is known an d the component has be en shown to be tolerant of easily det ected large flaws, even under reduced fracture t oughness conditions. V T-3 visual methods are a cceptable for the detection of cracking in redun dant RVI components (e.g., redunda nt bolts or pins used to secure a fastened RVI assembly).

In addition, VT-3 exami nations are used to monito r/inspect f o r loss of m a terial induced by wear and for general aging conditions, such as gross distortio n caused by void swelling and irradiation gr owth or by g r oss effect s of loss of pr eload cause d by thermal and irradiation-enhanced st ress relaxation and cree

p. The program adopts the guidance in MRP-227-A for definin g the "Expa n sion Criteria" that need to be applie d to the insp ection find in gs of "Primary" compone nts and for e x panding the examination s to include additional "E xpansion" components. RVI compo nent inspect i ons are performed consistent wit h the inspection frequen cy and sampling base s for "Primary" components, "Existing Programs" components, and "Expansion" components in MRP-227-A.

In some cases (as defin ed in MRP-2 27-A), phy sical measurements are used as supp lemental techniques t o manage for the gross e ffects of we ar, loss of pr eload due to stress relaxation, or for changes in dimensio ns due to void swelling o r distortion.

Inspection coverages for "Primary" a nd "Expansion" RVI components are implemente d consistent with Sections 3.3.1 and 3.3.2 of the NRC SE, Revision 1, on MRP-227. 5. M onitori ng and Trending:

The methods for monitoring, recording, e v aluating, and trending the data that result from the program's inspection s are given in Section 6 of MRP-227-A and its subsections.

Flaw evaluation methods, inclu d ing recommend ations for fla w depth sizing and for crack growt h determinations as well as for perfor m ing applicable limit loa d , linear ela s tic and elastic-p l astic fracture an alyses of relevant flaw indications, ar e defined in MRP-227-A.

The A-5 examination and re-examinations th at are implemented in accordance with MRP-2 27-A, together with the criteria specifie d in MRP-228 fo r inspectio n methodologies, inspe c tio n procedures, and inspect i on personne l, provide for timely detection, reporting, and implementat ion of corrective actions f o r the aging effects and mechanisms managed by the program.

The program applies a pplicable fra c ture toughn ess properties, inclu d ing reductions f o r thermal aging or neu tron embrittlement, in the flaw evaluations of the components in cases w here cracking is d e tected in a RVI compon ent and is e x tensive enough to warrant a supple m ental flaw growth or flaw tolerance evaluation.

For singly-represented components, the program incl udes crit eria to evaluate the agin g effects in the inaccessible por tions of the components and the resu lting impact on the inten ded function(s) o f the components. For r edundant co mponents (such as redu ndant bolts, screws, pins, keys, o r fasteners, some of which are acce ssible to inspection and some of which are not accessible t o inspect i on), the program includes criteria to evaluate the aging effects in the population o f components that are in accessible t o the applicable inspe c tion techniqu e and the resulting im pact on the intended fun c tion(s) of th e assembly containing t he components.

6. Acceptance Criteria

Section 5 of MRP-227-A, which includes Tab le 5-1 for B&W-designed RVIs, Table 5-2 for CE-designed RVIs, and Table 5-3 for Westinghouse-d e signed RVIs, provides the specific examination and flaw eval uation accept ance criteria for the "Primary" and "Expansion" RVI compo nent examin ation methods. For RVI components addressed by examination s performed in accordan ce with the ASME Code ,Section X I, the accepta n ce criteria in IWB-3500 are applica b le. For RVI components covered by other "Exis t ing Programs," the acceptance criteria are d e scribed wit h in the applicable refere nce document. As applicable, the program establishes a c ceptance crit eria for any physical measurement monitoring methods that are credited for aging management of particular R V I components.

7. Correctiv e Actions:

Any detected condition s that do not satisfy the e x amination acceptance criteria are r equired to b e disposition ed through t he plant corr ective action program, which may require repair, replacement, or analytical evaluation for contin ued service until the next inspection. The disposit ion will ensu r e that desig n basis fun c tions of the r eactor intern als components will continu e to be fulfil l ed for all li censing basis loads and e v ents. The implementat ion of the gu idance in M R P-227-A, p l us the implementation of any ASME Code requirements, provides an acceptab le level of a g ing management of safety-related components addressed in accordance with the corrective act i ons of 10 C F R Part 50, Appendix B or its equivalent, as applica b le. Other alternative correct ive actions b a ses may be us ed to disposition relevant conditions if they have been previously approved or endorsed by the NRC. Alt e rnative corrective actions not approved or endorsed by the NRC will be submitted for NRC approval prior to their implementat ion. 8. Confirmation Process:

Site quality assurance procedure s , review and approval processes, and administrative controls are im plemented in accordance with the recommendatio ns of NEI 03-08 and the requirements of 10 CFR Part 50, Appendix B, or their equivalent, as applicable.

The implementation of t he guidance in MRP-227-A, in conjun ction with NEI 03-08 and other guidance documents, reports, or methodologies r e ferenced in this AMP, provides an acceptable level of quality and an acceptable ba sis for confir ming the quality of inspe c tions, f l aw evaluations, and corrective actions.

A-6 9. Administrative Controls:

The administrative controls fo r these types of progra m s, includin g their implementing proce dures and r e view and approval processes, are implemented in accordance with the recommended i ndustry guidelines and criteria in NEI 03-08, and are under existing site 10 CFR 50 Appendix B, Quality Assurance Programs, or their equivalent, as applicable.

The evaluation in Section 3.5 of the NRC's SE, Revision 1, on MRP-22 7 provides the basis for endorsing NEI 03-08. This include s endorsement of the criteria in NEI 03-08 for notifying the NRC of any deviation from the I&E methodology in MRP-22 7-A and justifying the deviation no later than 4 5 days after it s approval by a license e executive.

10. Operati ng Experience:

The review and assessment of relevant operating experience for it s impacts on t he program, including implement ing procedures, are governed by NEI 03-08 and Appendix A of MRP-227-A. Consist ent with MRP-227-A, the reporting of inspect i o n results an d operating experience is t r eated as a "Needed" ca tegory item under the implementation of NEI 03-08. The program is informed and enhan ced when nece ssary thr ough the systematic an d ongoing review of both plant-spe cific and ind u stry operating experience, as discu ssed in App endix B of the GALL Report, which is documen ted in LR-ISG-2011-05.

References 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants , Office of the Federal Register, Nation al Archives and Records Administration, 2011.

10 CFR Part 50.55a, Codes and Sta ndards , Office of the Fe deral Regist er, National Archives and Records Administration, 2011.

ASME Boiler & Pressure Vessel Code,Section V, Nondestructive Exam ination , 2004 Edition, American Society of Mechanical En gineers, Ne w York, NY.

ASME Boiler & Pressure Vessel Code,Section XI, Rules for I n service I n spection of N u clear Power Plant Co m ponent s , The ASME Boiler and Pressure Vessel Code, 2004 edition as approved in 10 CFR 50.55a, The American Society of Mechanical Engin eers, New York, NY. EPRI Technical Report No. 1016596, Materials Reliability Program

Pressurized Wa ter Reactor Internals Inspection and Evaluation Guidelines (M RP-227, Revision 0), Electric Power Research In stitute, Palo Alto, CA: 2008.

EPRI Technical Report No.1022863, Materials Reliability Program

Pressurized Wa ter Reactor Internals Inspection and Evaluation Guidelines (M RP-227-A), December 2011, ADAMS Accession No. ML12017A193 (Transmittal letter f r om the EPRI-MRP) an d ADAMS Accession Nos. ML12017A194, ML 12017A196, ML12017A197, ML12017A191, ML 12017A192, ML12017A1 95 and ML12017A199, (Final Report).

EPRI 10166 09, Materials Reliability Progra m: In spection Sta ndard for PWR Internals (MRP-228), Electric Power Re search Inst it ute, Palo Alto, CA, July 2009 (Non-publicly available ADAMS Accession No.

ML092120574). The non-proprietary version of the report may be accesse d by members of the public a t ADAMS Ac cession No.

ML092750569.

A-7 NRC Interim Staff Guidance LR-ISG-2011-05, Ongoing Revie w Of Operat ing Experien c e , March 16, 2012, (ADAMS Accession No. ML12044A215).

Nuclear Energy Institute (NEI) Report No. 03-08, Revision 2, Guideline fo r the Manage m ent of Materials Issues , ADAMS Accession No. ML101 050334). NRC Safet y Evaluation from Robert A. Nelson (NRC) to Nei l Wilmshurst (EPRI), Re vision 1 to the Final Sa fety Evaluation of Electric Power Re search Inst it ute (EPRI)

Report, Mat e rials Reliability Progra m (MRP) Report 1016596 (MRP-227), Revision 0, Pressurized Wat e r Reactor Internals Inspe c tion and Eva l uation Guid elines , Dece mber 16, 2011, ADAMS Accession No. ML11308A770.

A-8 (2) Revised ver s ion of GALL Report Chapter IV.B2 B2. REACT O R VE SSE L INTERNALS (PWR) -

WESTINGHOUSE Sy stems, Structures, and Components This section addresses t he Westingh ouse pressu rized-water reactor (PWR) vessel internals, which consist of components in the upper intern als assembly, the control rod guide t ube assembly, the core barr e l assembly, the baffle/fo rmer assembly, the lower internals a ssembly, lower support assembly, thermal shield assembly, bottom mounted instr u mentation system, and alignment and interfacin g components.

Common mi scellaneou s material/environment combinations where aging effects are not expected to degrade the ability of th e structure o r component to perform its intended f unction for the period of extended operation are included in IV.E. Sy stem Interfaces The systems that interfa c e with the r eactor vessel internals include the r eactor pressure vessel (IV.A2).

A-9 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-300 IV.B2-33 (R-108) Alignment and interfacing components: internals hold down spring Stainless steel Reactor coolant and neutron flux Loss of preload due to thermal and irradiation enhanced stress relaxation; changes in dimensions due to void swelling or distortion; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-301 IV.B2-40 (R-112) Alignment and interfacing components: upper core plate alignment pins Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B2.RP-299 IV.B2-34 (R-115) Alignment and interfacing components: upper core plate alignment pins Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-271 IV.B2-10 (R-125) Baffle-to-former assembly: baffle-to-former bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-10 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-272 IV.B2-6 (R-128) Baffle-to-former assembly: baffle-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-270

IV.B2-1 (R-124) Baffle-to-former assembly: baffle and former plates Stainless steel Reactor coolant and neutron flux Changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-270a IV.B2-1 (R-124) Baffle-to-former assembly: baffle and former plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B2.RP-275 IV.B2-6 (R-128) Baffle-to-former assembly: baffle-edge bolts (all plants with baffle-edge bolts) Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-11 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-354 Baffle-to-former assembly: baffle-edge bolts (all plants with baffle-edge bolts) Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-273 IV.B2-10 (R-125) Baffle-to-former assembly: barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry (for SCC mechanisms only)

No IV.B2.RP-274 IV.B2-6 (R-128) Baffle-to-former assembly: barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No A-12 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-284 IV.B2-12 (R-143) Bottom mounted instrument system: flux thimble tubes Stainless steel (with or without chrome plating) Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" or Chapter XI.M37,"Flux Thimble Tube Inspection" No IV.B2.RP-293 IV.B2-24 (R-138) Bottom-mounted instrumentation system: bottom-mounted instrumentation (BMI) column bodies Stainless steel Reactor coolant and neutron flux Cracking due to fatigueChapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-292 Bottom-mounted instrumentation system: bottom-mounted instrument (BMI) column bodies Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-296

Control rod guide tube (CRGT) assemblies: CRGT guide plates (cards)

Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-298 IV.B2-28 (R-118) Control rod guide tube (CRGT) assemblies: CRGT lower flange welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-297 Control rod guide tube (CRGT) assemblies: CRGT lower flange welds Stainless steel (including CASS)Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging and neutron irradiation embrittlement and for CASS, due to thermal aging embrittlement Chapter XI.M16A, "PWR Vessel Internals" No A-13 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-355 Control rod guide tube (CRGT) assemblies: guide tube support pins (split pins)

Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-356 Control rod guide tube (CRGT) assemblies: guide tube support pins (split pins)

Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-387 Core barrel assembly: upper core barrel and lower core barrel circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-387a Core barrel assembly: upper core barrel and lower core barrel vertical (axial) welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-388 Core barrel assembly: upper core barrel and lower core barrel circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-388a Core barrel assembly: upper core barrel and lower core barrel vertical (axial) welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No A-14 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-345 Core barrel assembly: core barrel flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-278 IV.B2-8 (R-120) Core barrel assembly: core barrel outlet nozzle welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-278a Core barrel assembly: core barrel outlet nozzle welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-280 IV.B2-8 (R-120) Core barrel assembly: lower core barrel flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-276 IV.B2-8 (R-120) Core barrel assembly: upper core barrel flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking and irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B2.RP-285 IV.B2-14 (R-137) Lower internals assembly: clevis insert bolts or screws Nickel alloy Reactor coolant and neutron flux Loss of material due to wear; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No A-15 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-399 Lower internals assembly: clevis insert bolts or screws Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to primary water stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-289 IV.B2-20 (R-130) Lower internals assembly: lower core plate and extra-long (XL) lower core plate Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-288 IV.B2-18 (R-132) Lower internals assembly: lower core plate and extra-long (XL) lower core plate Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-291 IV.B2-24 (R-138) Lower support assembly: lower support column bodies (cast)

Cast austenitic stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B2.RP-290 IV.B2-21 (R-140) Lower support assembly: lower support column bodies (cast)

Cast austenitic stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging and neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No

A-16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-291a Lower support assembly: lower support forging or casting Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-290a Lower support assembly: lower support forging or casting Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement (and thermal aging embrittlement for CASS, PH SS, and martensitic SS)

Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-294 IV.B2-24 (R-138) Lower support assembly: lower support column bodies (non-cast)

Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B2.RP-295 Lower support assembly: lower support column bodies (non-cast)

Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-286 IV.B2-16 (R-133) Lower support assembly: lower support column bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-17 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-287 IV.B2-17 (R-135) Lower support assembly: lower support column bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-303 IV.B2-31 (R-53) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Cumulative fatigue damage due to fatigueFatigue is a time-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes - TLAA IV.B2.RP-24 IV.B2-32 (RP-24) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to pitting and crevice corrosion Chapter XI.M2, "Water Chemistry" No A-18 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-382 IV.B2-26 (R-142) Reactor vessel internals: ASME Section XI, Examination Category B-N-3 core support structure components (not already identified as "Existing Programs" components in MRP-227-A)

Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue, stress corrosion cracking, or irradiation-assisted stress corrosion cracking; loss of material due to wear Chapter XI.M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" or Chapter XI.M16A, "PWR Vessel Internals," by invoking applicable 10 CFR 50.55a and ASME Section XI inservice inspection requirements No IV.B2.RP-302 Thermal shield assembly: thermal shield flexures Stainless steel Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-302a Thermal shield assembly: thermal shield flexures Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-265 Reactor internal "No Additional Measures" components Stainless steel; nickel alloy Reactor coolant and neutron flux No additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-291b Upper Internals Assembly; upper core plate Stainless steel Reactor coolant and neutron flux Cracking due to fatigueChapter XI.M16A, "PWR Vessel Internals" No A-19 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-290b Upper Internals Assembly; upper core plate Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-346 Upper Internals Assembly: upper support ring or skirt Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No A-20 (3) Revised ver s ion of GALL Report Chapter IV.B3 B3. REACT O R VE SSE L INTERNALS (PWR) -

COMBUSTI ON E N GI NEERING Sy stems, Structures, and Components This section addresses t he Combustion Enginee ring (CE) pressurized-w a ter reactor (PWR) vessel inter nals, which consist of co mponents in the upper in ternals asse mbly, the control element assembly (CEA), the core support barrel assembly, the core shro ud assembly, and the lower support structure assembly, and encore in strumentation (ICI) components.

Common mi scellaneou s material/environment combinations where aging effects are not expected to degrade the ability of th e structure o r component to perform its intended f unction for the period of extended operation are included in IV.E. Sy stem Interfaces The systems that interfa c e with the r eactor vessel internals include the r eactor pressure vessel (IV.A2).

A-21 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-312 IV.B3-2 (R-149) Control Element Assembly (CEA): instrument guide tubes in peripheral CEA assemblies Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No IV.B3.RP-313 Control Element Assembly (CEA):

remaining instrument guide tubes in CEA assemblies Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-320 IV.B3-9 (R-162) Core shroud assemblies (all plants): guide lugs; guide lug inserts and bolts Stainless steel Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-319 IV.B3-9 (R-162) Core shroud assemblies (all plants): guide lugs; guide lug inserts and bolts Stainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No A-22 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-358 Core shroud assemblies (for bolted core shroud assemblies): assembly components, including shroud plates and former plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-318

IV.B3-8 (R-163) Core shroud assemblies (for bolted core shroud assemblies): assembly components, including shroud plates and former plates Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-316

IV.B3-9 (R-162) Core shroud assemblies (for bolted core shroud assemblies): barrel-shroud bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-23 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-317

IV.B3-7 (R-165) Core shroud assemblies (for bolted core shroud assemblies): barrel-shroud bolts Stainless steel Reactor coolant and neutron flux Loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-314 IV.B3-9 (R-162) Core shroud assemblies (for bolted core shroud assemblies): core shroud bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-315 IV.B3-7 (R-165) Core shroud assemblies (for bolted core shroud assemblies): core shroud bolts Stainless steel Reactor coolant and neutron flux Loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No

A-24 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-359 Core shroud assembly (designs assembled in two vertical sections): core shroud plate-to-former plate welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-322

Core shroud assembly (designs assembled in two vertical sections): core shroud plate-to-former plate welds Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-326

Core shroud assembly (designs assembled in two vertical sections): assembly components, including monitoring of the gap opening at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Changes in dimensions due to void swelling or distortion; loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No A-25 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-326a Core shroud assembly (designs assembled in two vertical sections): assembly components, including monitoring of the gap opening at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-323

Core shroud assembly (designs assembled in two vertical sections): remaining axial welds Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-359a Core shroud assembly (designs assembled in two vertical sections): remaining axial welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No

A-26 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-324 Core shroud assembly (designs assembled with full-height shroud plates): shroud plate axial weld seams at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-360 Core shroud assembly (designs assembled with full-height shroud plates): shroud plate axial weld seams at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-325

Core shroud assembly (designs assembled with full-height shroud plates): remaining axial welds, ribs, and rings Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry""

No

A-27 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-361 Core shroud assembly (designs assembled with full-height shroud plates): remaining axial welds, ribs, and rings Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-362 Core support barrel assembly: lower cylinder circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-362a Core support barrel assembly: lower cylinder circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-362b Core support barrel assembly: lower cylinder vertical (axial) welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No

A-28 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-362c Core support barrel assembly: lower cylinder vertical (axial) welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-329 IV.B3-15 (R-155) Core support barrel assembly: upper cylinder (base metal and welds) and upper core barrel flange (flange base metal)

Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-333 Core support barrel assembly: lower flange Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-328

IV.B3-15 (R-155) Core support barrel assembly: lower core barrel flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-332 IV.B3-17 (R-156) Core support barrel assembly: upper core barrel flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-327 IV.B3-15 (R-155) Core support barrel assembly: upper core barrel flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No A-29 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-357 Incore instruments (ICI): ICI thimble tubes - lower Zircaloy-4 Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-336 IV.B3-22 (R-170) Lower support structure (designs assembled in two vertical sections): fuel alignment pinsStainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-334 IV.B3-23 (R-167) Lower support structure (designs assembled in two vertical sections or with full-height shroud plates): fuel alignment pinsStainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-30 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-334a IV.B3-22 (R-170) Lower support structure (designs assembled in two vertical sections or with full-height shroud plates): fuel alignment pinsStainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-364 Lower support structure (all plants): core support column welds Stainless steel (including CASS) Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement and for column welds made from CASS, thermal embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-363 Lower support structure (all plants): core support column welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-330 IV.B3-23 (R-167) Lower support structure: core support column bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-31 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-331 Lower support structure: core support column bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-335 IV.B3-23 (R-167) Lower support structure (designs except those assembled with full-height shroud plates): lower core support beams Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No IV.B3.RP-365 Lower support structure (designs with a core support plate): core support plate Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-343 Lower support structure (designs with a core support plate): core support plate Stainless steel Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No

A-32 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-342 Lower support structure (designs with core shrouds assembled with full height shroud plates): deep beams Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-366 Lower support structure (designs with core shrouds assembled with full height shroud plates): deep beams Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-339 IV.B3-24 (R-53) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Cumulative fatigue damage due to fatigue Fatigue is a time-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes, TLAA IV.B3.RP-306 Reactor internal "No Additional Measures" components Stainless steel; nickel alloy Reactor coolant and neutron flux No additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists Chapter XI.M16A, "PWR Vessel Internals" No A-33 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-24 IV.B3-25 (RP-24) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to pitting and crevice corrosion Chapter XI.M2, "Water Chemistry" No IV.B3.RP-382 IV.B3-22 (R-170) Reactor vessel internals: ASME Section XI, Examination Category B-N-3 core support structure components (not already identified as "Existing Programs" components in MRP-227-A)

Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue, stress corrosion cracking, or irradiation-assisted stress corrosion cracking; loss of material due to wear Chapter XI.M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" or Chapter XI.M16A, "PWR Vessel Internals," by invoking applicable 10 CFR 50.55a and ASME Section XI inservice inspection requirements No IV.B3.RP-338 Upper internals assembly (designs with core shrouds assembled with full height shroud plates): fuel alignment plate Stainless steel Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-400 Core Support Barrel Assembly: thermal shield positioning pins Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No A-34 (4) Revised ver s ion of GALL Report Chapter IV.B4 B4. REA C T O R VE SSE L INTERN A L S (PWR) -

BAB COCK AND WILC OX Sy stems, Structures, and Components This section addresses t he Babcock and Wilcox (B&W) pressurized-water reactor (PWR) vessel internals, w h ich con s ist of components in the plenum cover assembly, the upper grid assembly, the control r od guide tub e (CRGT) a ssembly, the core supp ort shield a s sembly, the core barrel assembly, the lower grid assembly, incore monitoring instru mentation (IMI) guide tube assembly, and the flow distributor a ssembly.

Common mi scellaneou s material/environment combinations where aging effects are not expected to degrade the ability of th e structure o r component to perform its intended f unction for the period of extended operation are included in IV.E. Sy stem Interfaces The systems that interfa c e with the r eactor vessel internals include the r eactor pressure vessel (IV.A2).

A-35 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-242 IV.B4-4 (R-183) Control rod guide tube (CRGT) assembly: CRGT spacer castings Cast austenitic stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-242a Control rod guide tube (CRGT) assembly: CRGT spacer castings Stainless steel (including CASS) Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No IV.B4.RP-245 IV.B4-13 (R-194) Core barrel assembly (applicable to Crystal River Unit 3 or Davis Besse only):

surveillance specimen holder tube (SSHT) studs/nuts or bolts Nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-245a Core barrel assembly (applicable to Crystal River Unit 3 or Davis Besse only):

surveillance specimen holder tube (SSHT) stud or bolt locking devices Nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No

A-36 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-245b Core barrel assembly (applicable to CR-3 or DB only): surveillance specimen holder tube (SSHT) stud or bolt locking devices Nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-247 IV.B4-13 (R-194) Core barrel assembly: lower core barrel (LCB) bolts Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-247a Core barrel assembly: lower core barrel (LCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "P WR Vessel Internals" No IV.B4.RP-247b Core barrel assembly: lower core barrel (LCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "P WR Vessel Internals" No IV.B4.RP-249 IV.B4-12 (R-196) Core barrel assembly: baffle plates Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-249a Core barrel assembly: baffle plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No A-37 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-241 IV.B4-7 (R-125) Core barrel assembly: baffle-to-former bolts and screws Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, fatigue, and overload Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-241a Core barrel assembly: locking devices (including locking welds) of baffle-to-former bolts and internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, fatigue, and overload Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-240 IV.B4-1 (R-128) Core barrel assembly: baffle-to-former bolts and screws Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No A-38 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-240a Core barrel assembly: locking devices (including locking welds) of baffle-to-former bolts and internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-250 IV.B4-12 (R-196) Core barrel assembly: core barrel cylinder (including vertical and circumferential seam welds); former plates Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-250a Core barrel assembly: core barrel cylinder (including vertical and circumferential seam welds); former plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-375 Core barrel assembly:

internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking, fatigue, or overload Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No A-39 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-375a Core barrel assembly:

internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear Chapter XI.M16A, "P WR Vessel Internals" No IV.B4.RP-244 IV.B4-7 (R-125) Core barrel assembly; external baffle-to-baffle bolts and core barrel-to-former bolts; Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking, fatigue, and overload Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-244a Core barrel assembly: locking devices (including welds) of external baffle-to-baffle bolts and core barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking, or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-40 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-243 IV.B4-1 (R-128) Core barrel assembly: external baffle-to-baffle bolts and core barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-243a Core barrel assembly: locking devices (including welds) of external baffle-to-baffle bolts and core barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of material due to wear Chapter XI.M16A, "P WR Vessel Internals" No IV.B4.RP-248 IV.B4-12 (R-196) Core support shield (CSS) assembly: upper core barrel (UCB) bolts Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-248a Core support shield (CSS) assembly: upper core barrel (UCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No A-41 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-248b Core support shield (CSS) assembly: upper core barrel (UCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-252 IV.B4-16 (R-188) Core support shield (CSS) assembly: CSS vent valve top and bottom retaining rings (valve body components) Stainless steel, including CASS and PH steels Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-252a IV.B4-16 (R-188) Core support shield (CSS) assembly: CSS vent valve top and bottom retaining rings; vent valve locking devices (valve body components) Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-251 IV.B4-15 (R-190) Core support shield (CSS) assembly: CSS top flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of preload (wear) Chapter XI.M16A, "PWR Vessel Internals"No IV.B4.RP-251a IV.B4-15 (R-190) Plenum cover assembly: plenum cover weldment rib pads and plenum cover support flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of preload (wear) Chapter XI.M16A, "PWR Vessel Internals" No A-42 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-256 IV.B4-25 (R-210) Flow distributor assembly: flow distributor bolts Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-256a Flow distributor assembly: flow distributor bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals"No IV.B4.RP-256b Flow distributor assembly: flow distributor bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to distortion or void swelling or distortion Chapter XI.M16A, "P WR Vessel Internals" No IV.B4.RP-259 IV.B4-31 (R-205) Incore Monitoring Instrument (IMI) guide tube assembly: IMI guide tube spider-to-lower grid rib section welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-259a Incore Monitoring Instrument (IMI) guide tube assembly: IMI guide tube spider-to-lower grid rib sections welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No.

A-43 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-258 IV.B4-4 (R-183) Incore Monitoring Instrument (IMI) guide tube assembly: IMI guide tube spiders (castings)

Cast austenitic stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging and neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals"No IV.B4.RP-258a Incore Monitoring Instrumentation (IMI) guide tube assembly: IMI guide tube spiders Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-254 IV.B4-25 (R-210) Lower grid assembly: alloy X-750 lower grid shock pad bolts (Three Mile Island Unit 1, only) Nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-254a Lower grid assembly: alloy X-750 lower grid shock pad bolt locking devices (Three Mile Island Unit 1, only) Nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-254b Lower grid assembly:

alloy X-750 lower grid shock pad bolt locking devices (Three Mile Island Unit 1, only) Nickel Alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals"

No A-44 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-246 IV.B4-12 (R-196) Lower grid assembly: upper thermal shield (UTS) bolts and lower thermal shield (LTS) bolts Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-246a Lower grid assembly: upper thermal shield (UTS) bolt locking devices and lower thermal shield (LTS) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "P WR Vessel Internals" No IV.B4.RP-246b Lower grid assembly: upper thermal shield (UTS) bolt locking devices and lower thermal shield (LTS) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-260 IV.B4-31 (R-205) Lower grid fuel assembly: (a) pads; (b) pad-to-rib section welds; (c) alloy X-750 dowels, cap screws and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No

A-45 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-260a Lower grid fuel assembly: (a) pads; (b) pad-to-rib section welds; (c) alloy X-750 dowels, cap screws and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-262 IV.B4-32 (R-203) Lower grid assembly: alloy X-750 dowel-to-lower fuel assembly support pad locking welds Nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-261 IV.B4-32 (R-203) Lower grid assembly: alloy X-750 dowel-to-guide block welds Nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.R-53 IV.B4-37 (R-53) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Cumulative fatigue damage due to fatigue Fatigue is a time-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes, TLAA IV.B4.RP-24 IV.B4-38 (RP-24) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to pitting and crevice corrosion Chapter XI.M2, "Water Chemistry"

No A-46 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-376 Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Reduction in ductility and fracture toughness due to neutron irradiation Ductility - Reduction in Fracture Toughness is a TLAA (BAW-2248A) to be evaluated for the period of extended operation. See the SRP, Section 4.7, "Other Plant-Specific TLAAs," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes, TLAA IV.B4.RP-382 IV.B4-42 (R-179) Reactor vessel internals: ASME Section XI, Examination Category B-N-3 core support structure components Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue, stress corrosion cracking, or irradiation-assisted stress corrosion cracking; loss of material due to wear Chapter XI.M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" or Chapter XI.M16A, "PWR Vessel Internals," by invoking applicable 10 CFR 50.55a and ASME Section XI inservice inspection requirements No IV.B4.RP-352 Upper grid assembly: alloy X-750 dowel-to-upper fuel assembly support pad welds (all plants except Davis-Besse) Nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No A-47 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-236 Reactor internal "No Additional Measures" components Stainless steel; nickel alloy Reactor coolant and neutron flux No additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-400 Core support shield assembly: upper (top) flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-401 Core support shield assembly: upper (top) flange weld Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals"No A-48 (5) Revised ver s ion of GALL Report Chapter IX.C and IX.G IX.C Selected Definitions & Use of Terms for Des c ribing and Standardizing MATE RIAL S Stainless st eel Products gr ouped under the term "stainless stee l" include austenitic, f e rritic, martensitic, pre c ip itation-harde ned (PH), or duplex stainless steel (Cr content >11%). These stainless ste e ls may be fabricated using a wrought or cast process. Th ese materials are susce p tible to a variety of aging effect s and mechanisms, including loss o f material due to pittin g and crevice corrosion, and crackin g due to stress c o rrosion crac kin g. In som e cases, when an aging effect is app licable to all of the various stainle ss steel categories, it can be assumed that the term "stainless steel" in the "Material" column of an AMR line-item in the GALL Report encompasses all stainless stee l types. Cast austenitic stainless steel (CASS) is quite susceptible to loss of fracture t oughness d ue to thermal and neutr on irradiatio n embrittleme n t. In addition, MRP-22 7-A indicate s that PH stainless ste e ls or martensitic stainle ss steels may be susceptible t o loss of fra c ture toughn ess by a thermal aging mechanism. Therefore, when loss of fracture tou ghness due to thermal and neutron irradiation embrittlement is an applicable a g ing effect a nd mechanism for a component in the GALL Report, the CASS, PH sta i nless steel, or martensitic stainless ste e l designat ion is spe c ifically identified in an AMR line

-item. Steel with st ainless steel cladding a l so may be co nsidered stainless ste e l when the aging effect is asso ciate d with the stainless ste e l surface of the material, rather than the composite volume of the material.

Exa m ples of stainless st eel designat ions that co mprise this category include A-286, SA193-Gr.

B8, SA193-Gr. B8M, Gr. 660 (A-286), SA193-6, SA193-Gr. B8 or B-8M, SA453, Type 416, Type 403, 410, 420 , and 431 martensitic st ainless steels, Type 15-5, 17-4, and 13-8-Mo PH stainle ss steels, and SA-193, Grade B8 a nd B8M bolting materials.

Exa m ples of wrought austenitic stain l ess materia l s that comprise this category include Type 304, 304NG, 304L, 308, 308L, 309, 309L, 316 and 347.

Exa m ples of CASS that comprise this categ o ry include CF3, CF3M, CF8 and CF8M. [Ref. 6, 7, 30]

A-49 IX.G References

30. Welding Handbook, Seventh Edition, Volume 4, Metals and Their Welda b ility, American Welding So ciety, 1984, p.76-145.

A-50 Appendix A, Section 2 - Revised version of the SRP-LR (1) Revised ver s ion of SRP-LR Table 3.0-1 Ta ble 3.0-1 FSA R Supple m ent for A g ing M a na ge me nt of A p plic a b le Sy s t e m s G A LL Chapter G A LL Progra m De sc ription of Progra m Imple m e n ta tion Sc he dule A p plicable GA L L Re port a nd S R P-LR Ch ap ter Ref e r e n ces X I.M16A PWR Vessel Internals The program relie s on impl ementation of the inspe c t i on and eval u a tion guidelin es in EPRI Tech nical Rep o rt No. 102 286 3 (MRP-22 7-A) and EPRI Tech nical Re port No. 10 16 609 (MRP-22 8) to manage the aging effects on the reacto r vessel internal comp one nts. This progra m is used to manag e (a) crackin g , inclu d ing stress corro s io n cra cki ng, prim ary water st re s s co rr osi on cr ac kin g , irra diat ion-a s sist e d st r e s s cor r o s io n c r ack i n g , a nd c r ack i n g d ue to fatigue/cycli c al loading; (b) loss of material in du ced by wear; (c) loss of fractu re toug hne ss d ue to either thermal a g ing , neutron irra diation embrittleme n t, or void swell i ng; (d) dimen s ion a l chang es d ue to void swelling o r di stortion; an d (e) loss of prelo ad du e to thermal a n d irradi ation-enha nced stress relaxatio n or cree p. Program sho u ld be impleme n ted prio r to perio d of extended operation GALL IV / SRP 3.1 (2) Revised ver s ion of SRP-LR Section 3.1.2, "Acce ptance Crite r ia" 3.1.2.2.9 Re m o ved as a result of LR-ISG-201 1-04 3.1.2.2.10 Re m o ved as a result of LR-ISG-201 1-04 3.1.2.2.12 Re m o ved as a result of LR-ISG-201 1-04 3.1.2.2.13 Re m o ved as a result of LR-ISG-201 1-04 3.1.2.2.14 Re m o ved as a result of LR-ISG-201 1-04 (3) Revised ver s ion of SRP-LR Section 3.1.3, "Review Procedures

" 3.1.3.2.9 Re m o ved as a result of LR-ISG-201 1-04 3.1.3.2.10 Re m o ved as a result of LR-ISG-201 1-04 3.1.3.2.12 Re m o ved as a result of LR-ISG-201 1-04 3.1.3.2.13 Re m o ved as a result of LR-ISG-201 1-04 3.1.3.2.14 Re m o ved as a result of LR-ISG-201 1-04 A-51 (5) Revised ver s ion of SRP-LR Table 3.1-1 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 3 BW R/ PWR Stainl ess steel or nicke l allo y r eactor ve ssel inter nal compo nents e x pose d to reactor coo l ant and ne utron flux Cumul a tive fati gue d a ma ge due to fatig u e F a tigue is a T L AA eval uate d for the per iod of ext end ed o per ation (Se e SRP, Section 4.3 "Metal F a tigue," for ac ceptab le methods to co mpl y w i th 10 CF R 54.2 1 (c)(1) Yes, T L AA (See subsecti on 3.1.2.2.1) IV.B1.R-53 IV.B2.RP-303 IV.B3.RP-339 IV.B4.R-53 IV.B1-14 (R-53) IV.B2-31 (R-53) IV.B3-24 (R-53) IV.B4-37 (R-53) 15 PW R Stainl ess steel Babcock &

Wilcox (including CASS, martensitic SS, and PH SS) and n i ckel all o y react o r vessel i n terna l compo nents expos ed to rea c tor coola n t and n eutro n flu x Red u ction of ductilit y an d fracture toug hn ess due to neutro n irrad i at ion embrittlem ent, and for CASS, marten sitic SS, and PH SS due to thermal agi ng embrittlem ent Ductilit y - Re du ction in fracture toug hn ess is a T L AA to be evalu a ted for the peri od of e x te nde d oper ation, See the SRP, Section 4.7, "Other Plant-Specific T L AAs," for accepta b l e methods of meetin g the re quir e ments of 10 CF R 54.2 1 (c). Yes, T L AA (See subsecti on 3.1.2.2.3.3)

IV.B4.RP-376 N/A 28 PW R Stainl ess steel Comb ustion Engi neer in g "Existi ng Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Loss of materi al du e to w e ar; cracki ng due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-400 N/A 32 PW R Stainl ess steel, nickel a llo y, or CASS react o r vessel intern als, core supp ort structure (not a l rea d y referenc ed as ASME Section XI Exa m inati on Categ o r y B-N-3 core supp ort structure compo nents i n MRP-22 7-A), exp o sed to reactor cool ant an d ne utron flu x Crackin g , or lo ss of material due to w e ar Chapter X I.M1, "ASME Section XI Inse rvice Inspection, Subsections IW B, IW C, and IW D" or Chapter X I.M16A, "PWR Vessel Internals," invoking app lica b le 10 CF R 50.55 a and ASME Sec t ion XI inservic e ins p e c tion requ ireme n ts for these compo nents No IV.B2.RP-382 IV.B3.RP-382 IV.B4.RP-382 IV.B2-26 (R-142) IV.B3-22 (R-170) IV.B4-42 (R-179)

A-52 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 51a PW R Stainl ess steel or nicke l allo y B abcock & W ilcox reactor inter nal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B4.RP-241 IV.B4.RP-241a IV.B4.RP-242a IV.B4.RP-247 IV.B4.RP-247a IV.B4.RP-248 IV.B4.RP-248a IV.B4.RP-249a IV.B4.RP-252a IV.B4.RP-256 IV.B4.RP-256a IV.B4.RP-258a IV.B4.RP-259a IV.B4.RP-261 IV.B4.RP-400 IV.B4-7 (R

-125) N/A N/A IV.B4-13 (R-194) N/A IV.B4-25 (R-210) N/A N/A N/A IV.B4-25 (R-210) N/A N/A N/A IV.B4-32 (R-203) N/A 51b PW R Stainl ess steel or nicke l allo y B abcock & W ilcox reactor inter nal "Exp ansi on" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, fatigu e, or overl oad Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B4.RP-244 IV.B4.RP-244a IV.B4.RP-245 IV.B4.RP-245a IV.B4.RP-246 IV.B4.RP-246a IV.B4.RP-254 IV.B4.RP-254a IV.B4.RP-260a IV.B4.RP-262 IV.B4.RP-352 IV.B4.RP-250a IV.B4.RP-375 IV.B4-7 (R

-125) N/A IV.B4-13 (R-194) N/A IV.B4-12 (R-196) N/A IV.B4-25 (R-210) N/A N/A IV.B4-32 (R-203) N/A N/A N/A 52a PW R Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g reactor internal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-312 IV.B3.RP-314 IV.B3.RP-322 IV.B3.RP-324 IV.B3.RP-326a IV.B3.RP-327 IV.B3.RP-328 IV.B3.RP-342 IV.B3.RP-358 IV.B3.RP-362a IV.B3.RP-363 IV.B3.RP-338 IV.B3.RP-343 IV.B3-2 (R

-149) IV.B3-9 (R

-162) N/A N/A N/A IV.B3-15 (R-155) IV.B3-15 (R-155) N/A N/A N/A N/A N/A N/A A-53 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 52b PW R Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g re actor intern al "Exp ans ion" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-313 IV.B3.RP-316 IV.B3.RP-323 IV.B3.RP-325 IV.B3.RP-329 IV.B3.RP-330 IV.B3.RP-333 IV.B3.RP-335 IV.B3.RP-362c NA IV.B3-9 (R

-162) N/A N/A IV.B3-12 (R-155) IV.B3-23 (R-167) N/A IV.B3-23 (R-167) N/A 52c PW R Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g re actor intern al "Ex i sting Progr a ms" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-320 IV.B3.RP-334 IV.B3-9 (R

-162) IV.B3-23 (R-167) 53a PW R Stainl ess steel or nicke l allo y W e sti ngh ouse reactor inter nal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B2.RP-270a IV.B2.RP-271 IV.B2.RP-275 IV.B2.RP-276 IV.B2.RP-280 IV.B2.RP-298 IV.B2.RP-302 IV.B2.RP-387 N/A IV.B2-10 (R-125) IV.B2-6 (R

-128) IV.B2-8 (R

-120) IV.B2-8 (R

-120) IV.B2-28 (R-118) N/A N/A 53b PW R Stainl ess steel W e stingh ouse reactor intern al "E xpa n s ion" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B2.RP-273 IV.B2.RP-278 IV.B2.RP-286 IV.B2.RP-291 IV.B2.RP-291a IV.B2.RP-291b IV.B2.RP-293 IV.B2.RP-294 IV.B2.RP-387a IV.B2-10 (R-125) IV.B2-8 (R

-120) IV.B2-16 (R-133) IV.B2-24 (R-138) N/A N/A IV.B2-24 (R-138) IV.B2-24 (R-138) N/A 53c PW R Stainl ess steel or nicke l allo y W e sti ngh ouse reactor inter nal "Existin g Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B2.RP-289 IV.B2.RP-301 IV.B2.RP-345 IV.B2.RP-346 IV.B2.RP-399 IV.B2.RP-355 IV.B2-20 (R-130) IV.B2-40 (R-112) N/A N/A N/A N/A A-54 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 54 PW R Stainl ess steel bottom mounte d instru ment sy stem flux thimble tubes (w ith or w i t h o u t ch ro me pl a t i ng) e x po sed to reactor cool ant and neutro n flu x (W esting ho use "Ex i sting Progr a ms" compo nents) Loss of materi al du e to we a r Chapter X I.M16A, "PWR Vessel Internals," or Chapter X I.M37, "Flux T h imble T ube Inspecti on" No IV.B2.RP-284 IV.B2-13 (R-145) 55a PW R Stainl ess steel or nicke l allo y B abcock and W ilc o x reactor inter nal "No Additi ona l Mea s ures" compo nents e x pose d to reactor coo l ant and ne utron flux N o a d d i ti on al ag i n g mana geme n t for reactor intern al "No Ad ditio nal Measur es" co mpon ents unl ess requ ire d b y ASME Section XI, Examin ation Categ o r y B-N-3 or relev ant oper ating e x pe rienc e e x ists Chapter X I.M16A, "PWR Vessel Internals" No IV.B4.RP-236 NA 55b PW R Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g re actor intern al "No Add i tion al Measur es" compo nents e x pose d to reactor coo l ant and ne utron flux N o a d d i ti on al ag i n g mana geme n t for reactor intern al "No Ad ditio nal Measur es" co mpon ents unl ess requ ire d b y ASME Section XI, Examin ation Categ o r y B-N-3 or relev ant oper ating e x pe rienc e e x ists Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-306 NA 55c PW R Stainl ess steel or nicke l allo y W e sti ngh ouse reactor inter nal "No Additi ona l Mea s ures" compo nents e x pose d to reactor coo l ant and ne utron flux N o a d d i ti on al ag i n g mana geme n t for reactor intern al "No Ad ditio nal Measur es" co mpon ents unl ess requ ire d b y ASME Section XI, Examin ation Categ o r y B-N-3 or relev ant oper ating e x pe rienc e e x ists Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-265 NA 56a PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y C o mb usti on Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-315 IV.B3.RP-318 IV.B3.RP-359 IV.B3.RP-360 IV.B3-7 (R

-165) IV.B3-8 (R

-163) N/A N/A A-55 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item Engi neer in g re actor intern al "Primar y" com ponents expos ed to rea c tor coola n t and n eutro n flu x due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar IV.B3.RP-362 IV.B3.RP-364 IV.B3.RP-366 IV.B3.RP-365 IV.B3.RP-326 N/A N/A N/A N/A N/A 56b PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS)

Comb ustion E ngi neer in g "Exp ans ion" re actor intern al compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-317 IV.B3.RP-331 IV.B3.RP-359a IV.B3.RP-361 IV.B3.RP-362b IV.B3-7 (R

-165) N/A N/A N/A N/A 56c PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y C o mb usti on Engi neer in g re actor intern al "Ex i sting Progr a ms" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-319 IV.B3.RP-332 IV.B3.RP-334a IV.B3.RP-336 IV.B3.RP-357 IV.B3-9 (R

-162) IV.B3-17 (R-156) N/A IV.B3-22 (R-170) N/A 58a PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y B abcock & W ilcox reactor inter nal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss Chapter X I.M16A, "PWR Vessel Internals" No IV.B4.RP-240 IV.B4.RP-240a IV.B4.RP-242 IV.B4.RP-247b IV.B4.RP-248b IV.B4.RP-249 IV.B4.RP-251 IV.B4.RP-251a IV.B4.RP-252 IV.B4-1 (R

-128) N/A IV.B4-4 (R

-183) N/A N/A IV.B4-12 (R-196) IV.B4-15 (R-190) N/A IV.B4-16 (R-188)

A-56 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item of prelo ad d ue to w e ar; or loss of materia l due to w e ar IV.B4.RP-254b IV.B4.RP-256b IV.B4.RP-258 IV.B4.RP-259 IV.B4.RP-401 N/A N/A IV.B4-4 (R

-183) IV.B4-31 (R-205) N/A 58b PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y B abcock & W ilcox reactor inter nal "Exp ansi on" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B4.RP-245b IV.B4.RP-246b IV.B4.RP-254b IV.B4.RP-260 IV.B4.RP-243 IV.B4.RP-243a IV.B4.RP-250 IV.B4.RP-375a N/A N/A N/A IV.B4-31 (R-205) IV.B4-1 (R

-128) N/A IV.B4-12 (R-196) N/A 59a PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y W e sti ngh ouse re actor internal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-270 IV.B2.RP-272 IV.B2.RP-296 IV.B2.RP-297 IV.B2.RP-302a IV.B2.RP-354 IV.B2.RP-388 IV.B2.RP-300 IV.B2-1 (R

-124) IV.B2-6 (R

-128) N/A N/A N/A N/A N/A N/A 59b PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS)

W e stingh ouse reactor intern al "E xpa n s ion" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-274 IV.B2.RP-278a IV.B2.RP-287 IV.B2.RP-290 IV.B2.RP-290a IV.B2.RP-290b IV.B2.RP-292 IV.B2.RP-295 IV.B2.RP-388a IV.B2-6 (R

-128) N/A IV.B2-17 (R-135) IV.B2-21 (R-140) N/A N/A IV.B2-21 (R-140) IV.B2-22 (R-141) N/A A-57 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item loss of materia l due to w e ar 59c PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y W e sti ngh ouse re actor intern al "E xisti ng Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-285 IV.B2.RP-288 IV.B2.RP-299 IV.B2.RP-356 IV.B2-14 (R-137) IV.B2-18 (R-132) IV.B2-34 (R-115) N/A B-1 Appendix B MARK-UP OF CHANGES TO THE GALL REPORT AND SRP-LR

B-2 Appendix B, Section 1 - Mark-up of Changes to the GAL L Report In the mark-up, strikethr ough text indicates a de letion and u nderline text indicate s a n insertion.

Double strikethrough text indicate s t he original location of th e moved te xt and a doub le underline text indicate s the final lo cation of the moved te xt. (1) Mark-up of changes to GALL Report AMP XI.M16A XI.M16A PWR VESSEL INTERNALS Program Description This progra m relies on implementati on of the Electric Power Research I n stitute (EPRI)

Technical R eport No.

1016596 1022863, "Materials Reliability Program: Pressurized Water Reactor (PWR) Internals Inspe c tion and Evaluation Guidelines," (MRP-227-A) a nd EPRI Report Technica l No. 1016609 , "Materials Reliability Program: Inspection Stan dard for PWR Internals," (MRP-228) to manage the a g ing effects on the pressurized water reactor (PWR) reactor vessel int e rn al (RVI) components.

Th e recommen ded activitie s in MRP-227-A and addition al plant-spe cific a c tivitie s not define d in MRP-22 7-A are implemente d in accorda n ce with Nuclear Energ y Institute (NEI) 03-08, "Guideline f o r the Managemen t of Materials Issue s." T he staff appr oved the augmented inspection and evaluation (I&E) criteria for PWR RVI components in NRC Sa fety Evaluati on (SE), Revision 1, on MRP-22 7 by letter dated December 16, 2011.

This progra m is used to manage the effects of a ge-related d egradation mechanisms that are applicable in general to t he PWR RVI components at the facility.

These aging effects include : (a) various forms of cracking, including stress c o rrosion cra cking (SCC), which also encompasses primary water stress corrosion cracking (PW S CC), irradiation-assist ed stress corrosion cr acking (IAS CC), or and cracking d u e to fatigue/

cyclic al load ing; (b) loss of material induced by wear; (c) loss of fracture toughness d ue to either t hermal aging or neutron irradiation embrittleme n t; (d) chan ges in dime nsion s due t o void swelling or distortion

and (e) loss of preload due to thermal and irradiatio n-enhanced stress relax a tion or cree

p. The program applies th e guidance in MRP-227-A for inspect i ng, evaluating, and, if a pplicable, disposit ionin g non-confo r ming RVI c o mponents at the facility.

The program conforms to the definition of a sampling

-based condition monitoring program, as defined by the Bran ch Technical P o sition RSL B-1, with periodic examinations and other inspections of high ly-affected internals lo cations. These examinations provide reasonable assurance t hat the effects of age

--related deg radation mechanisms will be managed during th e period of e x tended operation.

The program includes expanding periodic ex aminations a nd other inspections

, if t he extent of the degradation effects identified exceeds the e x pected levels.

The MRP-2 2 7-A guidance for sele ct ing RVI components for inclusion in t he inspect i o n sample is based on a f our-step ran k ing proce s s. Through this proce s s, the reactor internals RVIs for all three PWR designs wer e assigned t o one of the following fou r groups: "Primary , ," "Expansion , ," "Existing Programs

, ," and "No Additional Measures components

." Definitions of e a ch group are provided in "Generic Aging Lesso ns Learned Report" (GALL Report), Revision 2, Chapter IX.B. The result of this four-ste p sample se lection pro c ess is a set of "Primary" I i nternals C c omponent locations for each of the three plant d e signs that are inspecte d because t hey are exp e cted to show the le ading indica tions of the degradation effects, with another set of "Expansion" B-3 I i nternals C c omponent locations that are specifie d to expand the sample should the indications be more severe than anticipated.

The degradation effects in a third se t of internals location s ar e deemed to be adequat ely managed by "Existing Programs," such as American Society of Mechanical Engineers (ASME) Code,Section X I , 11 Exa m ination Cate gory B-N-3 examination s of core su pport structu r es. A fourth set of internals lo cations are d eemed to require "Nn o a A dditional m M easures." As a result, the pr ogram typic a lly identifie s 5 to 15% o f the RVI locations as Primary Comp onent locations for inspections, with another 7 to 10% of the RVI locations to be inspected a s Expansion Component s, as warran t ed by the evaluation of the inspect i on results. A nother 5 to 15% of the internals lo cations are covered by Existing Programs, with the remainder requiring no additiona l measures.

This process thu s use s appropriate component functionality criteria, age-related degradation suscepti b ilit y criteria, an d failure con s equence cr iteria to iden tify the components that will be i n spected un der the program in a ma nner that co nforms to the sampling criteria fo r sampling

-based condition monitoring programs in Section A.1.2.3.4 of NRC Branch Position RLSB

-1. Consequently, the sample se lection pro c ess is a deq uate to assu re that the intended fun c tion(s) of th e PWR reactor internal components are maintained during t h e period of extended operation.

The program's use of visual examination methods in MRP

-2 27 for detect i on of relevant conditions (and the absence of rele vant conditions as a visual examina t ion accepta n ce criterion

) is consistent with the ASME Code,Section X I rules for visual examination. However, the program's adoption of th e MRP-227 guidance for visual examinations goe s beyond the ASME Code,Section X I visual examination criteria because additio nal guidance is incorp orated into MRP-227 to clarify how the particular visual examination methods will be u s ed to detect relevant conditions a nd describe s in more detail how the visual techniques relate t o the specif ic RVI components and how to detect their applicable a g e-related d egradation e ffects. The technical basis for detecting relevant conditions using volumetric ult r asonic te sting (UT) inspection t e chniques can be found in MRP-228, where the review of exi s ting bolt i ng UT examination technica l ju stificat ions h a s demonstrated the indication detection capabili t y of at least two vendors, and where vend or technica l ju stification is a requirement prior to any additional b o lting examinations. Spe c ifica lly, the capability of program's UT volumetric methods to detect loss of integrity of PWR internals bolt s , pins, an d fa steners, su ch as baffle

-fo rme r bolting in B

&W and Westinghouse units, has b een well de monstrated by operating experience.

In addition, t he program's adoption o f the MRP-2 27 guidance and process incorporat es the UT criteria in M R P-228, whi c h calls for t he technica l justification s that are ne eded for volumetric examination method demonstrations, required b y the ASME Code,Section V.

The program also inclu des future in dustry operating experience as in cor porated in p e riodic revisions to MRP-227. T he progra m thus provides reasonable assurance for the long

-term integrity and safe operation of reacto r internals in all commercial operatin g U.S. PWR nuclear power plants.

Age-related degradation in the reacto r internals is managed through an int egrated program.

S pecific feat ures of the integrated program are listed in the f o llowing ten program elements.

Degradation due to chan ges in material propertie s (e.g., loss of fracture t oughness) was considered in the determination of inspection re commendati ons and is manag ed by the requirement to use appr opriately degraded properties in the evaluation of identified d e fects. The integrated p r ogram is implemented by the applicant through an inspect i o n plan that is submitted to the NRC f o r review an d approval with the applicat ion f o r license rene wal. Evaluation and Technical Basis

11 Refer to the GALL Re po rt, Chapte r I, for applicability of various e d i t ions of the ASME Code,Section XI.

B-4 1. Scope of Program:

The scope o f the program includes all RVI components at t he [as an ad m i nistrati ve action ite m for the AMP, the app licant to f ill i n the nam e of the applicant' s nuclea r facilit y, inclu d ing applica b le units], w h ich [is/a re] built to a [

ap plicant to f ill in Westingh ouse, CE, or B&W, as applicable

] based on th e plant's ap plicable nu clear steam supply system NSSS design. The scope of th e program applies the m e thodology and guidance in MRP-22 7-Athe mos t recently NR C-endo rsed version of MRP

-22 7 , which pro v ides an augmented inspection and flaw evaluation methodology for assuring the functional integrity of safety-related intern als in commercial operating U.S. PWR nuclear power plants design ed by Babcock & Wilcox (B&W , ), Combustion Engineerin g (CE ,), and Westinghou se. The sco pe of components considered for inspection u nder in MRP-227 guidan ce include s-A include co re support st ructures (typically denoted as Exa m ination Category B

-N-3 by the ASME Code ,Section X I), , those RVI components that serve an intended license rene wal safety function pursu ant to criteria in 10 CFR 54.4(a)(1), and other RVI comp onents whose failure co uld prevent satisfactory accomplish m ent of any of the functions identif ie d in 10 CFR 54.4(a)(1)(i), (ii), or (iii).

In addition, ASME Code ,Section X I includes inspection requ irements for PWR remo vable core su pport structures in Table IWB-2500-1, Exa m ination Category B-N-3, which are in addition t o any inspection s t hat are implemented in accordance with MRP-2 27-A. The scope o f the program does not include con s umable items, such a s f uel assemblies, reactivity control assemblies, and n u clear instru mentation , because th ese components are not typically within the scop e of the components that are require d to be subject to an aging managemen t review (AMR), as defin ed by the criteria set in 1 0 CFR 54.21(a)(1).. The scope of the program also does n o t include w e lded attach ments to the internal surf ace of the re actor vessel beca u se these components are consider ed to be ASME Code Class 1 appurt enances to the reactor vessel and ar e adequately managed i n accordance with an applicant's AMP that corresponds to GALL AMP X I.M 1, "ASME Code ,Section X I Inservice In spection, Su bsection s IWB, IWC, and IWD."

The scope o f the program includes t he response bases to ap plicable lice n se renewal applicant a c tion items (L RAAIs) on the MRP

-22 7 methodology, and any additional pr ograms, actions, or a c tivities that are discussed in these LRAAI responses and credited for a g ing managemen t of the applicant's RVI components. The LRAAIs are identifie d in the staff

's safety evaluation on MRP

-227 and include applicable act i on items on meeting those assum p ti on s that formed the basis of the MRP' s augmented inspect i on and flaw evaluation methodology (as discussed in Section 2.4 of MRP

-227), and NSSS ven dor-specific or plant-specific LRAAIs as well.

The responses to the LRAAIs on MRP

-227 are provided in Appendix C of the LRA.

The guidance in MRP

-22 7 specifies a pplicability limitations to base-loaded plants and t he fuel loading managemen t assumptio n s upon which the functionality analyses were ba sed. These limita t ions and assumptions require a det ermination of appl icabilit y by the app licant for each reactor and are covered in Section 2.4 of MRP

-227. 2. Preventive Actions:

The guidan ce in MRP-2 2 7-A relies on PWR wat e r chemistry control to prevent or mitigate aging effects th at can be in duced by corrosive aging mechanisms (e.g., loss of material induced by general, pittin g corrosion, crevice corr osion, or str e ss corro si on crackin g or any of its forms [SCC, PWSCC, or IASCC]).

Reactor coola n t water chemistry is mo nitored and maintained in accor dance with t he Water Chemistry Program , as described. Th e program B-5 description , evaluation, and technical basis of water chemist r y are presented in GALL AMP X I.M 2, "Wat er Chemistry."

3. Parameters M onitored/Inspected:

The program manag es the follow i ng age-relat ed degradation effects and mechanisms that are ap plicable in g eneral to the RVI compo nents at the facility:

(a) cracking induced by SCC, PW SCC, IASCC, or fatigue/cyclic al load ing; (b) loss of material induced by wear; (c) loss of fractur e toughness induced by either thermal aging or neutron irra diation embrittlement; (d) changes in dimension s due to void swelling and irradiation growth, or distortion , or d e flection; an d (e) loss of preload cau s ed due tob y thermal an d irradiation

--enhanced st ress relaxation or creep.

For the management of cracking, th e program moni tors for evidence of surface brea king linear discontinu i ties if a visual inspecti on t e chnique is used as the non-destruct i on ve exami nation (NDE) meth od, or for relevant flaw presentation signals if a volumetric ult r asonic te sting (UT) method is used as the NDE metho d. For the managemen t of loss of material, the program monitors for gross or abn ormal surface condition s that may be indicative of loss of mat e rial occurring in the components.

For th e management of loss o f preload, th e program monitors for gross surfa c e condition s that may be indicative of loosening in applicable bolted, fast ened, keyed, or pinned conne ctions. The program does not direct ly monitor for loss of fra c ture toughness t hat is indu ced by thermal agi ng or n eutron irradiation embrittlement , or by void swelling and irradiation g r owth; inste a d. Instead

, the impact of loss of fra c ture toughn ess on component integrity is in directly managed by

(1) using visu al or volume tric examination techniques t o monitor for cracking in the components

, and by (2) applying applicable r educed fracture tou ghness prop erties in the flaw evaluations if , in cas e s where cracking is det ected in the components and is extensive enough to warrant necessitate a supplemental flaw growth or flaw toleran c e evaluatio n under the MRP-227 g u idance or ASME Code ,Section X I requirements.. The pro g ram uses physical measurements to monitor for any dimen s ional changes du e to void swelling or, irra diation growth, distortion , or deflectio n.. Specifically, the program impl ements the parameters monitored/inspecte d criteria for

[as an ad m i nistrative action ite m for the AMP, applicant is to se lect one of the f o llowing to finish the sentence, as applicable to its NSSS vendor for it s internals: "

f or B&W designed Primary Co m ponent s in Table 4

-1 of MRP-2 2 7"; "for CE designed Primary Co m p onents in Ta ble 4-2 of MRP-227"; and "for W e stinghouse designed P r im ary Co mpon ents in T able 4-3 of MRP-227"]. Additionally, the program impl ements the par ameters monitored/inspected cr it eria for [as an adm i n istrative action item for the AMP, ap plicant is to select one o f the followin g to finish the se ntence, as a pplicable t o its NSS S vendor for its internals: "fo r B&W designed Expansion Co m ponent s in Table 4

-4 of MRP-2 2 7"; "for CE designed Expansion Co mponents in Table 4

-5 of MRP-227"; and "for W e stinghouse designed E x pansion Co mponents in Table 4-6 of MRP-227"]. The p a rameters monit ored/inspected for Existing Program Co mp onents follow the b a ses for refe renced Existing Programs, such a s t he requirements for ASME Code Class RVI components in ASME Co de,Section X I, Table IW B-2500-1, Examination Categories B

-N-3, as implemented through t he applicant's ASME Code ,Section X I program, or the reco mmended p r ogram for inspecting W e stinghouse

-designed flux thimble tubes in GALL AMP X I.M 37, "Flux Thimble T ube Inspection." No insp ections, except for those specified in ASME Code ,Section X I, are re quire d for components that ar e identified as requiring "No Additional Measures," in accordance with the analyses rep o rted in MRP

-227. Specifically, the program implem ents the parameters monitored/inspecte d criteria co nsistent with the applicable tab l e s in Section 4, "Aging Management Requiremen t s," in MRP-227-A.

B-6 4. Detection of Aging Effects:

The detection of aging effe cts is covered in two places: (a) the guidance in Section 4 of MRP-227 provides an in troductory discussion an d justificatio n of the exa mination The inspect i on methods selected for detecting th e aging effe cts of intere st; and (b) standards fo r examinatio n are defined and established in Sect ion 4 of MRP-227-A. St andards for impleme n ting the inspection methods , proce dures, are d e fined and pe rsonnel are provided established in a companion document, MRP-228. In all ca ses, well-established inspection methods are were select ed. These methods include volume tric UT examination methods for detecting fla w s in bolting , physical measurements for detecting changes in dimension, and various visual (VT-3, VT-1, and EVT-1) exa m inations fo r detecting e ffects ranging fro m general conditions to detection an d sizing of surface-brea king discontinuities.

Surface examinations may also be used as an alte rnative to visual exami nations for d e tection and sizing o f surface-br eaking disco n tinuities.

Cracking ca used by SCC, IASCC, and fatigue is monitored/inspected by either VT-1 or EVT-1 examination (for internals other than bolting) or b y volu metric UT exa m ination (boltin g). The VT--3 visual methods may b e applied for the detectio n of crackin g in non-red undant RVI components only when the flaw toler ance of the component or affected a ssembly , as evaluated for reduced fracture tou ghness prop erties, is kn own and the component has been sh own to be tolerant of e a sily detecte d large flaw s, even under reduced f r acture toug hness cond itions. VT-3 visual methods are acceptable for t he detection of cracking in redundant RVI compo nents (e.g., redundant bolts or pin s used to secure a fasten ed RVI assembly).

In addition, VT-3 exami nations are used to monito r/inspect f o r loss of m a terial induced by wear and for general aging conditions, such as gross distortio n caused by void swelling and irradiation gr owth or by g r oss effect s of loss of pr eload cause d by thermal and irradiation-enhanced st ress relaxation and cree

p. In addition, t h e The program adopts the recomme nded guidan ce in MRP-2 2 7-A for defining the "Expansion criteria Criter ia" that need to be applie d to the insp ection find in g s of "Primary Component s and Existing Requirement Co mp onents " co mponents and for expanding the examination s to include additional "E x pansion Components. As a result," components. RVI component inspection s performed on the RVI components are performed consist ent with the inspection fr equency and sampling bases for "Pr i m a r y Co mponents, " components, "

Existing Requiremen t Componen ts Programs" components , and "Expansion Components

" co mponents in MRP-227-A, which have been demonstrated to be in conf orma nce with the inspection criteria, sampling basis cr iteria, and sample Expansion criteria in Section A.1.2.3.4 of NRC Br anch Position RLSB

-1. Specifically, the program impl ements the parameters monitored/inspecte d criteria an d bases for inspect i n g the relevant parameter condition s for [as an ad m i nistrative action item for the AMP, applicant is to se lect one of th e following t o finish the sentence, a s applicab le to its NSSS vend or for its internals: "B&W designed P r im ary Co mponents in Table 4

-1 of MRP-22 7"; "CE designed Primary Co m p onents in Ta ble 4-2 of MRP-227;" or "Westingho use designed Primary Co m p onents in Ta ble 4-3 of MRP-227"] and for [

as an ad m i nistrative action item for the AMP, applicant is to se lect one of th e following t o finish the sentence, as applicable t o its NSSS vendor for it s internals: "

f or B&W designed Expansion Com p onents in Table 4-4 of MRP-227;" "for CE designed expan sion com ponents in Tab le 4-5 of MRP-227;" and "for Westinghouse designed Expansion Co mponents in Table 4

-6 o f MRP-227"]. The program is supple m ented by t he following plant-specific Primary Component and Expansion Component inspection s f o r the program (as applicable): [

As a relevant license renewal applicant action item , the ap plicant is to list (using criteria in MRP

-22 7) each B-7 additional R V I com pone nt that need s to be insp ected as an additional p l ant-specific Pri m ary Co m ponent for the applicant' s progra m and each additional RVI co m pon ent that nee ds to be inspected a s an additio nal plant-sp ecific Exp a n s ion Com p o nent for t he applicant' s progra m. For each pla n t specif ic co m ponent added as an additional pr im ary or Ex pansion Co mponent, the list shoul d include th e applicabl e aging effect s that will be m onitored for, the inspe c tion method or methods used for m onit o ring, and th e sam p le size and frequencies for th e exam ination s]. In addition, in som e cas e s (as defin ed in M R P-227-A), physical measurements are used as supplemental technique s to manage for the gross effects o f wear, loss o f preload du e to stress relaxation, or for change s in dimension s due to void swelling

, deflection o r distortion.

The physical measurements methods applied in acco rdance with this program include [

Ap plicant to input physical m easure methods identified by th e MRP in re sponse to N RC RAI No. 11 in the NR C's Requ est for Additional Inform a t ion to Mr. Christen B. L a rson, EPRI MRP on To pical Report MRP

-227 dated Nove m ber 1 2 , 2009]. Inspection coverages for "Primary" a nd "Expansion" RVI components are implemente d consistent with Sections 3.3.1 and 3.3.2 of the NRC SE, Revision 1, on MRP-227. 5. M onitori ng and Trending:

The methods for monitoring, recording, e v aluating, and trending the data that result from the program's inspection s are given in Section 6 of MRP-227-A and its subsections.

The Flaw evaluation methods inclu d e , includin g recommen dations for flaw depth sizing and f o r crack gro w th determi nations as w e ll as for per forming applicable limit load, linear elastic a nd elastic-p l astic fracture an alyses of relevant flaw indications

., a r e defined in MRP-227-A. The examination s examination and re-examinations required by the that are implemented in accordan ce with MRP-227 guidance-A , together with the requirements criteria spe c ified in MRP-228 fo r inspectio n methodologies, inspe c tio n procedure s , and inspe c tion person nel, provide time ly detection, reporting, a nd corrective actions wit h respect to the effects o f the age-r elated degr adation mechanisms within the scop e of the pro g ram. The extent of the examination s , beginning with the sa mple of susceptible PW R internals component locations identified a s Primary Co mponent locations, with t he potential for inclusion of Expansi on Component locations if t he effects ar e greater than anticipat ed, plus the continuation of the Existing Programs activities, su ch as the ASM E Code,Section X I, Exa m ination Cate gory B-N-3 examination s for core su pport structu r es, provides a high deg ree of confid ence in the t o tal for timely detection, reportin g , and implementation of corrective a c tions for th e aging effe cts and mechanisms managed by the program.

The program applies a pplicable fra c ture toughn ess properties, inclu d ing reductions f o r thermal aging or neu tron embrittlement, in the flaw evaluations of the components in cases w here cracking is d e tected in a RVI compon ent and is e x tensive enough to warrant a supple m ental flaw growth or flaw tolerance evaluation.

For singly-represented components, the program incl udes crit eria to evaluate the agin g effects in the inaccessible por tions of the components and the resu lting impact on the inten ded function(s) o f the components. For r edundant co mponents (such as redu ndant bolts, screws, pins, keys, o r fasteners, some of which are acce ssible to inspection and some of which are not accessible t o inspect i on), the program includes criteria to evaluate the aging effects in the population o f components that are in accessible t o the applicable inspe c tion techniqu e and the resulting im pact on the intended fun c tion(s) of th e assembly containing t he components.

B-8 6. Acceptance Criteria

Section 5 of MRP-227-A, which includes Tab le 5-1 for B&W-designed RVIs, Table 5-2 for CE-designed RVIs, and Table 5-3 for Westinghouse-d e signed RVIs, provides the specif ic examination and flaw evaluation acceptance crit eria for the "Primary" an d "Expansion Component examination

s. For " RVI component examination methods. For RVI comp onents addressed b y exa m inations referenced to performed in acco rdance with the ASME Code,Section X I, the IWB-3500 acceptance criteria ap ply.in IWB-3500 are applicable.

For other RVI components covered by other "Existing Programs

, ," the examination acceptance cr iteria are described within the Existing Program applicable reference d o cument. The guidance in MRP

-22 7 contains t h ree types of examination As applica b le, the prog ram establishe s acceptance criteria

• For visual examination (and surface examination as an altern ative to visual examination), the examina t ion accepta n ce criterion is the abse n ce of any of the specif ic, descriptive relevant conditions; in a ddition, ther e are requir e ments to record and disposition surf ace breaking ind i cations th at are detecte d and sized for length by VT

-1/EVT-1 examinations;

• For volumetric examination, the examinati on acceptance cr i t erion is the capability fo r reliable dete c tion of ind i cations in bo lting, as demonstrated in the examina t ion Technical Justifi c ation; in addition, there are requirements for system

-level assessment of bolted or pinned asse mblies with unacceptable volumetric (UT) exa m ination indications that exceed specified limits; and For physical measureme n ts, the ex amination acceptance cr iterion for th e acceptable tolerance in the measured differe ntial height f r om the top of the plenu m rib pads to the vessel seating surface in B

&W plants a r e given in Table 5

-1 of MRP-227. T he acceptan ce criterion f o r physical measureme n ts performed on the he ight lim its of the Westing house-desig ned hold-do wn springs are [The incorporation o f this senten ce is a licen s e renewal applicant a c tion item for Westinghou se PWR applicants only - insert the applicable sentence incorporating t he specif ied any physical measureme n t criteria on ly if the app li cant's facility is based o n a Westing house NSSS design: the Westinghouse applicant is to incorporate t he applicable language and then specify the fit up lim its on the hold d o wn springs, as established on a pla nt-specific b a sis for the design of th e hold-down springs at t he applicant

's Westingh ouse-design ed facilit y].monitoring methods that are credite d for aging managemen t of particula r RVI compo nents. 7. Correctiv e Actions:

Corrective actions fo llowing th e dete c tion of una cceptable conditions are fundamentally provided for in ea ch plant's co rrective action program.

Any detected conditions t hat do not satisfy the examination acceptance criteria are re quired to be disposit ione d through th e plant corre ctive action program, wh ich may req u ire repair, replacement, or analytical evaluatio n for c ontinu ed service u n til the next inspection. The disposit ion will ensu r e that desig n basis fun c tions of the r eactor intern als compon ents will continue to be fulfilled f o r all li censi ng basis loa d s and events.

Example s of methodologies that can be used to analytically dispositio n unaccepta b le conditio n s are found in the ASME Code,Section X I o r in Section 6 of MRP-2 27. Section 6 of MRP-2 27 describe s the option s that are avail able for disposit ion of detected conditions t hat exceed the examinat ion accepta n ce criteria of Section 5 of the report. These include enginee ring evaluation methods, as well as supplementary exa m ina t ions to furth e r characterize the detected conditio n , or t he alte rnative of component repair and re placement procedures.

The latter ar e subject to the require ments of the ASME Code ,Section X I. The implementation of t he guidance in MRP-227 The implementation of the guida nce in MRP-227-A , plus the impleme n tation of an y ASME Co de requirements, provides an acceptable level of aging management of safety-related components add ressed in B-9 accordance with the corr ective action s of 10 CFR Part 50, Appendix B or its equivalen t, as applicable.

Other alternative correct ive action act i ons base s may be use d to disposition relevant conditions if they have been previously approved or endorsed by the NRC.

Exa m ples of previously NRC

-endorsed alt e rnative corrective action s bases include those corrective act i ons base s f o r Westinghou s e-design R V I components that are defined in T ables 4-1, 4-2, 4-3, 4-4, 4-5, 4-6, 4-7 and 4-8 of Westinghou se Report No. WCAP

-14 577-Rev. 1-A, or for B&W

-designed RVI components in B&W Report No. BAW

-2248. Westinghouse Report No. WCAP

-1457 7-Rev. 1-A was endo rsed for use in an NRC SE to the West inghouse Owners Grou p, dated February 10, 2001. B&W Report No. BAW

-2248 was endorsed for use in an SE to Framatome T e chnologie s on behalf of the B&W Owners Grou p, dated December 9, 1999.

Alternative correct ive action bases act i on s not approved or endorsed by the NRC will be submitted for NRC appr oval prior to their implementation.

8. Confirmation Process:

Site quality assurance procedure s , review and approval processes, and administrative controls are im plemented in accordance with the recommendatio ns of NEI 03-08 and the r equirements of 10 CFR Part 50, Appendix B, or their equivalent, as applicable.

It is expected that the The implementati on of the guidance in MRP-227 will provide-A, in conjunct i o n with NEI 03-08 and othe r guidance d o cuments, reports, or methodologie s referenced in this AMP, pro v ides an accep t able level o f quality and an accepta b le basis for inspection confirming the quality of inspect i ons , flaw evaluation, and oth e r elements of aging management of th e PWR internals tha t are addressed in accor dance with t he 10 CFR Part 50, Appendix B, or their equivalent (as applica b le), confirmation process, and administrative cont rols evaluations, and corrective actions. 9. Administrative Controls:

The administrative controls fo r such these types of programs, including t h eir implementing proced ures and review and approval processes, are imp l emented in accordan ce with the r e commende d industry guidelines a n d criteria in NEI 03-08, and are under existing site 10 C F R 50 Appendix B

, Qual ity Assurance Programs, or their equivalent, as applicable.

Such The evaluation in Section 3.5 of the NRC' s SE, Revision1, on MRP-227 provides the basis for en dorsing NEI 03-08. This includes e n dorsement of the criteria in NEI-03-08 for notifying the NRC of a n y deviation from the I&E methodology in MRP-227-A and justifying the deviation no later than 45 days after its approval by a program is thus expected to be establish ed with a su fficient level of documentation and ad mini strative controls to e n sure effective lon g-term imple m entation licensee executive

. 10. Operati ng Experience:

Relatively few incidents of PWR internals aging degra dation have been reported in operating U.S. commercial PWR plants. A summary of observations t o date is provided in Appendix A of MRP-227-A. The applicant is expected to review subsequ ent and assessment of relevant operating experience for impact its impacts on it s th e program or to participate in industry in itiatives that perform this function.

The applicat ion of the MRP

-227 guidance will est ablish a con s iderable a m ount of , including implementing procedure s , are governed by NEI 03-08 and Appendix A of MRP-227-A.

Consistent with MRP-2 27-A, the reporting of in spection re sults and ope rating experience over the next few years. Section 7 of MRP

-227 describ es the repor ting require ments for these application s , and the pla n for evaluating the accu mulated additional opera t ing experience is treated as a "Needed" category item under the implementation of NEI 03-08.

B-10 The program is informed and enhan ced when nece ssary thr ough the systematic an d ongoing review of both plant-spe cific and ind u stry operating experience, as discu ssed in App endix B of the GALL Report, which is documen ted in LR-ISG-2011-05.

References 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants , Office of the Federal Register, Nation al Archives and Records Administration, 2009 20 11. 10 CFR Part 50.55a, Codes and Sta ndards , Office of the Fe deral Regist er, National Archives and Records Administration, 2009.2 011. ASME Boiler & Pressure Vessel Code,Section V, Nondestructive Exam ination , 2004 Edition, American Society of Mechanical En gineers, Ne w York, NY.

ASME Boiler & Pressure Vessel Code,Section XI, Rules for I n service I n spection of N u clear Power Plant Co m ponent s , The ASME Boiler and Pressure Vessel Code, 2004 edition as approved in 10 CFR 50.55a, The American Society of Mechanical Engin eers, New York, NY. B&W Report No. BAW-2 248, De m o n s tration of th e Manage ment of Aging Effects for t he Reactor Vessel Inter nals , Framatome Technologies (now AREVA Technologies), Lynchburg VA, July 1997. (NRC Microfiche Accession Number A0076 , Microfiche Pages 001

- 108). EPRI 10149 86, PWR Primary Water Chem istry Guidelines , V o lume 1, Revision 6, Ele c tric Power Rese arch Institut e, Palo Alto, CA, December 2007. (Non

-publicly available ADAMS Accession Number ML0 81140278). The non

-proprietary vers ion of the re port may ac cessed by me mbers of th e public at ADAMS Accessio n Number ML 081230449 EPRI 10165 96, Materials Reliability Progra m: Pressurized Water Reactor Internals I n spection and Evaluation Guidelin es (MRP-22 7 , -Rev ision. 0), Electric Power Rese arch Institut e, Palo Alto, CA:

2008. EPRI Technical Report No. 1022863, Materials Reliability Program

Pressurized Wa ter Reactor Internals Inspection and Evaluation Guidelines (M RP-227-A), December 2011, ADAMS Accession No. ML12017A193 (Transmittal letter f r om the EPRI-MRP) an d ADAMS Accession Nos. ML12017A194, ML 12017A196, ML12017A197, ML12017A191, ML 12017A192, ML12017A1 95 and ML12017A199, (Final Report).

EPRI Technical Report No. 1016609, Materials Reliability Program: Inspection Stand ard for PWR Internals (MRP-22 8), Electric Power Rese arch Institut e, Palo Alto, CA, July 2009

. (Non-publicly available ADAMS Accession N o.u m ber ML09 2120574). The non-proprietary version of the report may accessed by me mbers of the public at ADAMS Accessio n N o.umber ML092750569.

NRC RAI N o. 11 in the NRC's Requ est for Additional Inform a t ion to the Mr. Christen B. Larson, EPRI MRP on Topical Report MRP-227 dated Nove mber 1 2 , 2009. NRC Safet y Evaluation from C. I. Gri m es [NRC] t o R. A, Newton [Chairman, Westinghouse Owners Gro up], Accepta n ce for Referencing of Generi c Licen s e Renewal Progra m To pical Report Entitled "License Renewal Evaluation: Aging Management for Re actor Intern als," WCAP-1457 7, Revision 1, February 10, 2001. (ADAMS Accession Number ML010430375).

B-11 NRC Safet y Evaluation from C. I. Gri m es [NRC] t o W. R. Gra y [Framato me Technologies], Acceptance for Referencing of Generic Licen s e Renewal Progra m Topical Report Entitled "De m onstrat ion of the Managem ent of Aging Effects for the Reactor Vessel Internals," February 10, 2001. (ADAMS Accession Number ML993490288

). NURE G-18 00, Revision 2, Standard Review Pla n for Review of License Renewal Ap plication s for Nuclear Power Plants, Appendix A.1, "Aging Managemen t Review - Generic (Branch Technical P o sition RLS B-1)," U.S.

Nuclear Regulatory Commission, Wa shington, D C , 2010. Westinghou se Non-Proprietary Class 3 Report No. WCAP-14577-Rev. 1-A, License Renewal Evaluation:

Aging Manage m ent for Reactor Internals , Westinghouse Ele c tric Company, Pittsburgh, PA [March 2 001]. Report was submitted to the NRC Docume nt Control Desk in a letter dated April 9, 200

1. (ADAMS Accession Number ML0 11080790).

NRC Interim Staff Guidance LR-ISG-2011-05, Ongoing Revie w Of Operat ing Experien c e , March 16, 2012, (ADAMS Accession No. ML12044A215).

Nuclear Energy Institute (NEI) Report No. 03-08, Revision 2, Guideline fo r the Manage m ent of Materials Issues , ADAMS Accession No. ML101 050334). NRC Safet y Evaluation from Robert A. Nelson (NRC) to Nei l Wilmshurst (EPRI), Re vision 1 to the Final Sa fety Evaluation of Electric Power Re search Inst it ute (EPRI)

Report, Mat e rials Reliability Progra m (MRP) Report 1016596 (MRP-227), Revision 0, Pressurized Wat e r Reactor Internals Inspe c tion and Eva l uation Guid elines , Dece mber 16, 2011, ADAMS Accession No. ML11308A770.

B-12 (2) Mark-up of changes to GALL Report Chapter IV.B2 B2. REACT OR VESSEL I N TERNAL S (P WR) - WE STINGHOU SE Sy stems, Structures, and Components This section addresses t he Westingh ouse pressu rized -water reactor (PWR) vessel internals and consist s of , which consist of components in the upper intern als assembly, the control rod guide tube assem b lies assembly , the core barrel asse m b ly , the ba ffle/former assembly, the lower internal assembly, and t h e internals assembly, lower support assembly, thermal shield assembly, bottom mou n ted instrumentation sup port structur es. Based o n Regulatory Guide 1.2 6 , "Quality Group Classification s an d Standards for Water

-, Steam-, and Radioactive

-Waste-Containing Component s of Nuclear Power Plants," all struc t u res and co mponents that comprise the reactor vessel are g o verned by Group A or B Quality St andards.system, and alignment and interfacing components.

Common mi scellaneou s material/environment combinations where aging effects are not expected to degrade the ability of th e structure o r component to perform its intended f unction for the period of extended operation are included in IV.E. Sy stem Interfaces The systems that interfa c e with the r eactor vessel internals include the r eactor pressure vessel (IV.A2). Inspection Plan An applicant will submit an inspect i on plan for reactor internals to the NRC for review and approval with the application for lice nse renewal in accordan ce with Chapter X I.M 16 A , "PWR Vessel Inter nals."

B-13 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-300 IV.B2-33 (R-108) Alignment and interfacing components:

internals hold down spring Stainless steel Reactor coolant and neutron flux Loss of preload due to thermal and irradiation enhanced stress relaxation; changes in dimensions due to void swelling or distortion; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals"

Primary components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-301 IV.B2-40 (R-112) Alignment and interfacing components:

upper core plate alignment pins Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking 'Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-299 IV.B2-34 (R-115) Alignment and interfacing components:

upper core plate alignment pins Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-271 IV.B2-10 (R-125) Baffle-to-former assembly: accessible baffle-to-former bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Expansion components see AMR Items IV.B2.RP-273 and IV.B2.RP

-286)SCC mechanisms only)

No B-14 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-272 IV.B2-6 (R-128) Baffle-to-former assembly: accessible baffle-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; change changes in dimensions due to void swelling or distortion

loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) (for Expansion components see AMR Items IV.B2.RP

-274 and IV.B2.RP-287) No IV.B2.RP-270 IV.B2-1 (R-124) Baffle-to-former assembly: baffle and former plates Stainless steel Reactor coolant and neutron flux Changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-270a IV.B2-1 (R-124) Baffle-to-former assembly: baffle and former plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B2.RP-275 IV.B2-6 (R-128) Baffle-to-former assembly: baffle-edge bolts (all plants with baffle-edge bolts)

Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Primary components (identified in the "Structure and Components" column) no Expansion components No B-15 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-354 Baffle-to-former assembly: baffle-edge

bolts (all plants with baffle-edge bolts)

Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion

loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-273 IV.B2-10 (R-125) Baffle-to-former assembly

barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry (for Primary components see AMR Item IV.B2.RP

-271)SCC mechanisms only)

No IV.B2.RP-274 IV.B2-6 (R-128) Baffle-to-former assembly: barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation

embrittlement; changes in dimensions due to void swelling or distortion

loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-272) No B-16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-284 IV.B2-12 (R-143) Bottom mounted instrument system: flux

thimble tubes Stainless steel (with or without

chrome plating)

Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column)

No expansion components; and or Chapter XI.M37 , " ,"Flux Thimble Tube Inspection

"" No IV.B2.RP-293 IV.B2-24 (R-138) Bottom-mounted instrumentation system:

bottom-mounted instrumentation (BMI) column bodies Stainless steel Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-298) No IV.B2.RP-292 IV.B2-21(R-140) Bottom-mounted instrumentation system:

bottom-mounted instrumentationinstrume nt (BMI) column bodies Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals"

Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-297) No IV.B2.RP-296 Control rod guide tube (CRGT) assemblies: CRGT guide plates (cards) Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Primary Components (identified in the "Structure and Components" column) (for Expansion components see AMR Line Item IV.B2.RP-386) No B-17 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-298 IV.B2-28 (R-118) Control rod guide tube (CRGT) assemblies:

CRGT lower flange welds (accessible)

Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Expansion components see AMR Items IV.B2.RP-291 and IV.B2.RP-293)SCC mechanisms only)

No IV.B2.RP-297 Control rod guide tube (CRGT) assemblies: CRGT lower flange welds (accessible)

Stainless steel (including CASS)

Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging and neutron irradiation

embrittlement and for CASS, due to thermal aging embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) (for Expansion components see AMR Items IV.B2.RP

-290 and IV.B2.RP-292) No IV.B2.RP-386 Control rod guide tube (CRGT) assemblies: C

-tubes and sheaths Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) are only the components associated with a primary component that exceeded the acceptance limit.

(for Primary components see AMR Item IV.B2.RP

-296) No IV.B2.RP-355 IV.B2.RP-355 Control rod guide tube (CRGT) assemblies: guide tube support pins (split pins)

NickelStainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking and or fatigue A plant-specific aging management program is to be evaluatedChapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) Yes, plant

-specific No B-18 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-356 Control rod guide tube (CRGT) assemblies: guide tube support pins (split pins)

NickelStainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear A plant-specific aging management program is to be evaluatedChapter XI.M16A, "PWR Vessel Internals" Yes, plant

-specific No IV.B2.RP-387 Core barrel assembly:

upper core barrel axialand lower core barrel circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking , and or irradiation-assisted stress

-corrosion cracking or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B2.RP-276SCC mechanisms only) No IV.B2.RP-387a Core barrel assembly: upper core barrel and lower core barrel vertical (axial) welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-388 Core barrel assembly:

upper core barrel axialand lower core barrel circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-276) No IV.B2.RP-282 388a IV.B2-8(R-120) Core barrel assembly:

upper core barrel flangeand lower core barrel vertical (axial) welds Stainless steel Reactor coolant and neutron flux Cracking Loss of fracture toughness due to stress corrosion cracking and fatigueneutron irradiation embrittlement Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-276) No B-19 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-345 Core barrel assembly: core barrel flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-278 IV.B2-8 (R-120) Core barrel assembly: core barrel outlet nozzle welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking and fatigue Cracking due to stress corrosion cracking or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion component (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B2.RP-276)SCC mechanisms only) No IV.B2.RP-280 278a IV.B2-8(R-120) Core barrel assembly: lower core barrel flange weldoutlet nozzle welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking and irradiation

-assisted stress corrosion cracking Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion component (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-276) No IV.B2.RP-281 280 IV.B2-98 (R-122 120) Core barrel assembly: lower core barrel flange weld Stainless steel Reactor coolant and neutron flux Loss of fracture toughnessCracking due to neutron irradiation embrittlement stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" Expansion Components (identified in the "Structure and Components" column) Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B2.RP

-276)SCC mechanisms only)

No B-20 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-276 IV.B2-8 (R-120) Core barrel assembly: upper core barrel flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking and irradiation-assisted stress corrosion cracking Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Primary components (identified in the "Structure and Components" column) (for Expansion components see AMR Items IV.B2.RP

-278, IV.B2.RP-280, IV.B2.RP

-282, IV.B2.RP-294, IV.B2.RP

-295,IV.B2.RP

-281, IV.B2.RP

-387, and IV.B2.RP

-388) No IV.B2.RP-285 IV.B2-14 (R-137) Lower internals assembly: clevis insert

bolts or screws Nickel alloy Reactor coolant and neutron flux Loss of material due to weardue to wear; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals"

Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-399 Lower internals assembly: clevis insert bolts or screws Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to primary water stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-289 IV.B2-20 (R-130) Lower internals assembly: lower core plate and extra-long (XL) lower core plate Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion cracking , and or fatigue 'Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Existing Program components (identified in the "Structure and Components" column) no Expansion components No B-21 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-288 IV.B2-18 (R-132) Lower internals assembly: lower core plate and extra-long (XL) lower core plate Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals"

Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-291 IV.B2-24 (R-138) Lower support assembly: lower support column bodies (cast)

Cast austenitic stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion cracking Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-298) No IV.B2.RP-290 IV.B2-21 (R-140) Lower support assembly: lower support column bodies (cast)

Cast austenitic stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging and neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-297) No IV.B2.RP-291a Lower support assembly: lower support forging or casting Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-290a Lower support assembly: lower support forging or casting Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement (and thermal aging embrittlement for CASS, PH SS, and martensitic SS) Chapter XI.M16A, "PWR Vessel Internals" No B-22 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-294 IV.B2-24 (R-138) Lower support assembly: lower support

column bodies (non-cast) Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion crackingChapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-276) No IV.B2.RP-295 IV.B2-22(R-141) Lower support assembly: lower support column bodies (non-cast) Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion Components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-276) No IV.B2.RP-286 IV.B2-16 (R-133) Lower support assembly: lower support column bolts Stainless steel

nickel alloy Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" col umn) and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B2.RP-271)SCC mechanisms only) No IV.B2.RP-287 IV.B2-17 (R-135) Lower support assembly: lower support column bolts Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation

embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" Expansion component (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-272) No B-23 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-303 IV.B2-31 (R-53) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Cumulative fatigue damage due to fatigue Fatigue is a time-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes , - TLAA IV.B2.RP-24 IV.B2-32 (RP-24) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to pitting and crevice corrosion Chapter XI.M2, "Water Chemistry" No IV.B2.RP-268 382 IV.B2-26 (R-142) Reactor vessel internal internals: ASME Section XI, Examination Category B-N-3 core support structure components (inaccessible locations)not already identified as "Existing Programs" components in MRP-227-A)

Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue, stress corrosion cracking, and or irradiation

- assisted stress corrosion cracking

loss of material due to wear Chapter XI.M2, "Water Chemistry,"

M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" or Chapter XI.M16A, "PWR Vessel Internals

" ," by invoking applicable 10 CFR 50.55a and ASME Section XI inservice inspection requirements Yes, if accessible Primary, Expansion or Existing program components indicate aging effects that nee d management No IV.B2.RP-302 Thermal shield assembly: thermal shield flexures Stainless steel Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No B-24 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-269 302a Reactor vessel internal components (inaccessible locations)Thermal shield assembly: thermal shield flexures Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; change in dimension due to void swelling; loss of preload due to thermal and irradiation enhanced stress relaxation; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Yes, if accessible Primary, Expansion or Existing program components indicate aging effects that need manag ement NoIV.B2.RP-265 Reactor internal "No Additional Measures" componentsReactor vessel internal components with no additional measures Stainless steel; nickel alloy Reactor coolant and neutron flux No additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience existsCracking due to stress corrosion cracking, and irradiation

-assisted stress corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Note: Components with no additional measures are not uniquely identified in GALL tables

- Components with no additional measures are defined in Section 3.3.1 of MRP

-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" No IV.B2.RP-267 291b Reactor vessel internal components with no additional measures Upper Internals Assembly; upper core plate Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of fracture toughness Cracking due to neutron irradiation embrittlement; change in dimension due to void swelling; loss of preload due to thermal and irradiation enhanced stress relaxationfatigue; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Note

Components with no additional measures are not uniquely identified in GALL tables

- Components with no additional measures are defined in Section 3.3.1 of MRP

-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" No IV.B2.RP-382 IV.B2-26(R-142) Reactor vessel internals: core support structure Stainless steel; nickel alloy; cast austenitic stainless steel Reactor coolant and neutron flux Cracking, or Loss of material due to wear Chapter XI.M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" No B-25 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-302 290b Upper Internals Assembly; upper core plateThermal shield assembly: thermal shield flexures Stainless steel Reactor coolant and neutron flux Cracking due to fatigue; loss Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals"

Primary components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-346 Upper i Internals a Assembly: upper support ring or skirt Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking and or fatigue 'Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) Existing Program components (identified in the "Structure and Components" column) no Expansion components No B-26 (3) Mark-up of changes to GALL Report Chapter IV.B3 B3. REACT OR VESSEL I N TERNAL S (P WR) - C O MBU S TION ENGI NEER ING Sy stems, Structures, and Components This section addresses t he Combustion Enginee ring (CE) pressurized

-water reactor (PWR) vessel inter nals and consists of , which con s ist of components in the up per internals assembly, the control e l ement asse mbly (CEA) shrouds,), the core sup port barrel a ssem b ly , the core shroud asse mbly, and th e lower internal assembly. Based on Regulatory Guide 1.26, "Quality Group Class ifications and Standards for Water

-, Steam-, and Radioa ctive-Waste-Containing Components of Nuclear Power Pl ants," all structures and components that compri se the reactor vessel are governed by Group A or B Quality Standards support structure assembly, and encore instr u mentation (ICI) components

. Common mi scellaneou s material/environment combinations where aging effects are not expected to degrade the ability of th e structure o r component to perform its intended f unction for the period of extended operation are included in IV.E. Sy stem Interfaces The systems that interfa c e with the r eactor vessel internals include the r eactor pressure vessel (IV.A2). Inspection P l an An applicant will submit an inspect i on plan for reactor internals to the NRC for review and approval with the applica tion for lice nse renewal in accordan ce with Chapter X I.M 16 A, "PWR Vessel Inter nals."

B-27 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-312 IV.B3-2 (R-149) Control Element Assembly (CEA):

shroud assemblies:

instrument guide tubes in peripheral CEA assemblies Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Expansion components see AMR Item IV.B3.RP-313)SCC mechanisms only)

No IV.B3.RP-313 Control Element Assembly (CEA):

shroud assemblies: remaining instrument guide tubes in CEA

assemblies Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, ""PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

" and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B3.RP-312)SCC mechanisms only)

No IV.B3.RP-320 IV.B3-9 (R-162) Core shroud assemblies (all plants): guide lugs and; guide lug insertinserts and bolts Stainless steel Reactor coolant and neutron flux Cracking due to fatigue

'Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B3.RP-319 IV.B3-9 (R-162) Core shroud assemblies (all plants): guide lugs and; guide lug insertinserts and bolts Stainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of preload due to thermal and irradiation

enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals

"" Existing Program components (identified in the "Structure and Components" column) no Expansion components No B-28 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-358 Core shroud assemblies (for bolted core shroud assemblies): (a)assembly components, including shroud plates and (b) former plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column)

(for Primary component see AMR Item IV.B3.RP-314) No IV.B3.RP-318 IV.B3-8 (R-163) Core shroud assemblies (for bolted core shroud assemblies): (a)assembly components, including shroud plates and (b) former plates Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement;

changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion components No IV.B3.RP-316 IV.B3-9 (R-162) Core shroud assemblies (for bolted core shroud assemblies): barrel-

shroud bolts with neutron exposures greater than 3 dpa Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, ""PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

" and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B3.RP-314)SCC mechanisms only)

No IV.B3.RP-317 IV.B3-7 (R-165) Core shroud assemblies (for bolted core shroud assemblies): barrel-shroud bolts with neutron exposures greater than 3 dpa Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of fracture toughness due to neutron

irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B3.RP-315) No B-29 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-314 IV.B3-9 (R-162) Core shroud assemblies (for bolted core shroud assemblies): core shroud bolts (accessible)

Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Expansion components see AMR Items IV.B3.RP-316, IV.B3.RP

-330, and IV.B3.RP

-358)SCC mechanisms only)

No IV.B3.RP-315 IV.B3-7(R-165) Core shroud assemblies (for bolted core shroud assemblies): core shroud bolts (accessible)

Stainless steel Reactor coolant and neutron flux Loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of fracture toughness due to neutron

irradiation embrittlement; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals

," Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Items IV.B3.RP-317, and IV.B3.RP

-331)" No IV.B3.RP-359 Core shroud assemblies (welded):

(assembly (designs assembled in two vertical sections): core shroud plates and (b) plate-to-former platesplate welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals

," Primary components (identified in the "Structure and Components" column) no Expansion components

" No B-30 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-322 Core shroud assembly (for welded core shrouds designs assembled in two vertical sections): Core shroud plate

-former plate weld (a) The axial and horizontal weld seams at the core shroud re-entrant corners as visible from the core side of the shroud, within six inches of the central flange and horizontal stiffeners, and (b) the horizontal stiffen ers in core shroud plate-to-former plate weld welds Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Item IV.B3.RP-323)and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-326 Core shroud assembly (for welded core shrouds designs assembled in two vertical sections):

gap betweenassembly components, including monitoring of the upper and lower plates gap opening at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Changes in dimensions due to void swelling or distortion; loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion components No B-31 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-323326a Core shroud assembly (for welded core shrouds designs assembled in two vertical sections): remaining axial welds in assembly components, including monitoring of the gap opening at the core shroud plate-to-former platere-entrant corners Stainless steel Reactor coolant and neutron flux Cracking due to irradiation

-assisted stress -corrosion cracking or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B3.RP-322)SCC mechanisms only)

No IV.B3.RP-324323 Core shroud assembly (for welded core shrouds with full-height shroud plates): axial weld seams at the core shroud re-entrant corners, at the core mid-plane (+3 feet in height) as visible from the core side of the shroud Core shroud assembly (designs assembled in two vertical sections): remaining axial welds Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Item IV.B3.RP-325)and Chapter XI.M2, "Water Chemistry" No B-32 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-360 359a Core shroud assembly (for welded core shrouds with full-height shroud plates): axial weld seams at the core shroud re-entrant corners, at the core mid-plane (+3 feet in height) as visible from the core side of the shroud Core shroud assembly (designs assembled in two vertical sections): remaining axial welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement

change s in dimension s due to void swelling or distortion Chapter XI.M16A, ""PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Item IV.B3.RP-361)" No IV.B3.RP-325324 Core shroud assembly (for welded core shrouds designs assembled with full-height shroud plates): remainingshroud plate axial welds, ribs, and ringsweld seams at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B3.RP-324) No B-33 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-361360 Core shroud assembly (for welded core shrouds designs assembled with full-height shroud plates): remainingshroud plate axial welds, ribs, and ringsweld seams at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B3.RP-360) No IV.B3.RP-362325 Core support barrelshroud assembly: lower cylinder (designs assembled with full-height shroud plates):

remaining axial welds, ribs, and rings Stainless steel Reactor coolant and neutron flux Loss of fracture toughnessCracking due to neutron irradiation embrittlement-assisted stress corrosion cracking Chapter XI.M16A, ""PWR Vessel Internals" Expansion components (identified in the "Structure

" and Components" column)Chapter XI.M2, "Water Chemistry""

(for Primary components see AMR Item IV.B3.RP-327) No IV.B3.RP-329361 IV.B3-15(R-155) Core support barrelshroud assembly: lower cylinder (designs assembled with full-height shroud plates):

remaining axial welds , ribs, and remaining core barrel assembly welds rings Stainless steel Reactor coolant and neutron flux Cracking Loss of fracture toughness due to stress corrosion cracking neutron irradiation embrittlement Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "" PWR Vessel Internals

"" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B3.RP-327) No IV.B3.RP-333362 Core support barrel assembly: lower flange weld, if fatigue life cannot be demonstrated by TLAAcylinder circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Cracking Loss of fracture toughness due to fatigueneutron irradiation embrittlement Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "" PWR Vessel Internals

"" Primary components (identified in the "Structure and Components" column) no Expansion components Yes, evaluate to determine the potential locations and extent of fatigue cracking No B-34 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-389 362a Core support barrel assembly: lower flange weld (if fatigue analysis exists)cylinder circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Cumulative fatigue damage due to fatigueCracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Yes, TLAANo IV.B3.RP-328362b IV.B3-15(R-155) Core support barrel assembly: surfaces of the lower core barrel flange weld (accessible surfaces)cylinder vertical (axial) welds Stainless steel Reactor coolant and neutron flux Cracking Loss of fracture toughness due to stress corrosion cracking and fatigueneutron irradiation embrittlement Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "" PWR Vessel Internals

" Primary components (identified in the "Structure and Components" column) no Expansion components" No IV.B3.RP-332 362c IV.B3-17(R-156) Core support barrel assembly: upper core barrel flangelower cylinder vertical (axial) welds Stainless steel Reactor coolant and neutron flux Loss of material due to wearCracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column) no Expansion componentsChapter XI.M2, "Water Chemistry" No IV.B3.RP-327329 IV.B3-15(R-155) Core support barrel assembly: upper cylinder (base metal and welds) and upper core support barrel flange weld (accessible surfaces)(flange base metal) Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Items IV.B3.RP-329, IV.B3.RP

-335, IV.B3.RP

-362, IV.B3.RP-363, IV.B3.RP

-364)and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-357333 Incore instrumentation (ICI):

ICI thimble tubes

- lowerCore support barrel assembly: lower flange Zircaloy-4 Stainless steel Reactor coolant and neutron flux Loss of materialCracking due to wearstress corrosion cracking or fatigue A plant-specific aging management program is to be evaluatedChapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Yes, plant

-specificNo B-35 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-336328 IV.B3-22 15(R-170)155) Lower support structure: A286 fuel alignment pins (all plants with core shroud assembled in two vertical sections)Core support barrel assembly: lower core barrel flange weld Stainless steel Reactor coolant and neutron flux Cracking Loss of material due to wear; loss of fracture toughness due to neutr on irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxationcorrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column) no Expansion componentsChapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No IV.B3.RP-334332 IV.B3-23 17(R-167)156) Lower support structure: A286 fuel alignment pins (all plants with core shroud assembled with full-height shroud plates)Core support barrel assembly: upper core barrel flange Stainless steel Reactor coolant and neutron flux Cracking Loss of material due to irradiation

-assisted stress corrosion cracking and fatiguewear 'Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B3.RP-364327 IV.B3-15(R-155) LowerCore support structure:barrel assembly: upper core support columnbarrel flange weld Cast austenitic stainlessStainless steel Reactor coolant and neutron flux Loss of fracture toughnessCracking due to neutron irradiation and thermal embrittlement stress corrosion cracking Chapter XI.M16A, ""PWR Vessel Internals" Expansion components (identified in the "Structure

" and Components" column)

(for Primary components see AMR Item IV.B3RP-327)Chapter XI.M2, "Water Chemistry" No IV.B3.RP-363 357 Lower support structure: core support columnIncoreinstruments (ICI): ICI thimble tubes - lower Stainless steel Zircaloy-4 Reactor coolant and neutron flux Loss of fracture toughnessmaterial due to neutron irradiation embrittlementwear Chapter XI.M16A, ""PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B3RP-327)" No B-36 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-330336 IV.B3-23 22(R-167)170) Lower support structure: core support column bolts (designs assembled in two vertical sections): fuel alignment pins Stainless steel Reactor coolant and neutron flux Loss of materialCracking due to wear; loss of fracture toughness due to neutron irradiation

-assisted embrittlement; loss of preload due to thermal and irradiation enhanced stress corrosion cracking and fatigue relaxation or creep Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item 'IV.B3.RP-314) No IV.B3.RP-331334 IV.B3-23(R-167) Lower support structure: core support column bolts (designs assembled in two vertical sections or with full-height shroud plates):

fuel alignment pins Stainless steel Reactor coolant and neutron flux Loss of fracture toughnessCracking due to neutron stress corrosion cracking, irradiation embrittlement-assisted stress corrosion cracking, or fatigue Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item 'IV.B3.RP-315)SCC mechanisms only)

No IV.B3.RP-335 334a IV.B3-23 22(R-167)170) Lower support structure: core support column welds, applicable to all plants except those (designs assembled in two vertical sections or with full-height shroud plates): fuel alignment pins Stainless steel Reactor coolant and neutron flux Cracking Loss of material due to stress corrosion cracking,wear; loss of fracture toughness due to neutron irradiation

-assisted stress corrosion cracking, embrittlement; loss of preload due to thermal and fatigue irradiation enhanced stress relaxation or creep Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B3.RP-327) No B-37 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-365364 Lower support structure: (all plants):

core support platecolumn welds Stainless steel (including CASS)Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement and for column welds made from CASS, thermal embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary component (identified in the "Structure and Components" column) no Expansion components No IV.B3.RP-343363 Lower support structure: (all plants):

core support plate (applicable to plants with a core support plate), if fatigue life cannot be demonstrated by TLAAcolumn welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, or fatigue Chapter XI.M2, "Water Chemistry", and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Primary components (identified in the "Structure and Components" column) no Expansion components Yes, evaluate t o determine the potential locations and extent of fatigue cracking No IV.B3.RP-390330 IV.B3-23(R-167) Lower support structure: core

support plate (applicable to plants with a core support plate), if fatigue analysis exists column bolts Stainless steel Reactor coolant and neutron flux Cumulative Cracking due to irradiation-assisted stress corrosion cracking or fatigue damage due to fatigue Fatigue is a time

-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Yes, TLAA No IV.B3.RP-342331 Lower support structure:

deep beams (applicable assemblies with full height shroud plates)core support column bolts Stainless steel Reactor coolant and neutron flux Cracking Loss of fracture toughness due to stress corrosion cracking, neutron irradiation

-assisted stress corrosion cracking, and fatigue embrittlement Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion components No B-38 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-366335 IV.B3-23(R-167) Lower support structure: deep beams (applicable assemblies (designs except those assembled with full

-height shroud plates)): lower core support beams Stainless steel Reactor coolant and neutron flux Loss of fracture toughnessCracking due to neutron irradiation embrittlement stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion componentsChapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No IV.B3.RP-365

Lower support structure (designs with a core support plate): core support plate Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary component (identified in the "Structure and Components" column) no Expansion components No IV.B3.RP-24343 IV.B3-25(RP-24) Reactor vessel internal componentsLower support structure (designs with a core support plate): core support plate Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of materialCracking due to pitting and crevice corrosion fatigue Chapter XI.M2, "Water Chemistry" M16A, "PWR Vessel Internals" No IV.B3.RP-309342 Reactor vessel internal components (inaccessible locations)Lower support structure (designs with core shrouds assembled with full height shroud plates)

deep beams Stainless steel

nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking, and irradiation-assisted stress

-corrosion cracking , or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Yes, if accessible Primary, Expansion or Existing program components indicate aging effects that need management No B-39 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-311366 Reactor vessel internal components (inaccessible locations)Lower support structure (designs with core shrouds assembled with full height shroud plates): deep beams Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement

change in dimension due to void swelling

loss of preload due to thermal and irradiation enhanced stress relaxation; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Yes, if accessible Primary, Expansion or Existing program components indicate aging effects that need management No IV.B3.RP-339 IV.B3-24(R-53) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Cumulative fatigue damage due to fatigue Fatigue is a time-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes, TLAA IV.B3.RP-306 Reactor internal "No Additional Measures" componentsReactor vessel internal components with no additional measures Stainless steel; nickel alloy Reactor coolant and neutron flux No additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience existsCracking due to stress corrosion cracking, and irradiation

-assisted stress corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Note: Components with no additional measures are not uniquely identified in GALL tables - Components with no additional measures are defined in Section 3.3.1 of MRP-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluat ion Guidelines" No B-40 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-30724 IV.B3-25(RP-24)

Reactor vessel internal components with no additional measures Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of fracture toughnessmaterial due to neutron irradiation embrittlement; change in dimension due to void swelling; loss of preload due to thermal and irradiation enhanced stress relaxation; loss of material due to wear pitting and crevice corrosion Chapter XI.M16A, "PWR Vessel Internals" Note: Components with no additional measures are not uniquely identified in GALL tables - Components with no additional measures are defined in Section 3.3.1 of MRP-227, "Materials Reliability Program: Pressurized M2, "Water Reactor Inte rnals Inspection and Evaluation Guidelines"Chemistry" No IV.B3.RP-382 IV.B3-22(R-170) Reactor vessel internals:

ASME Section XI, Examination Category B-N-3 core support structure components (not already identified as "Existing Programs" components in MRP-227-A)

Stainless steel; nickel alloy

cast austenitic stainless steel Reactor coolant and neutron flux Cracking due to fatigue, stress corrosion cracking , or irradiation-assisted stress corrosion cracking; Loss loss of material due to wear Chapter XI.M1, "ASME Section XI Inservice Inspection, Subsecti ons IWB, IWC, and IWD" or Chapter XI.M16A, "PWR Vessel Internals," by invoking applicable 10 CFR 50.55a and ASME Section XI inservice inspection requirements No IV.B3.RP-338 Upper internals assembly

fuel alignment plate (applicable to plants (designs with core shrouds assembled with full height

shroud plates), if fatigue life cannot be demonstrated by TLAA): fuel alignment plate Stainless steel Reactor coolant and neutron flux Cracking due to fatigue

'Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion components Yes, evaluate to determine the potential locations and extent of fatigue cracking No B-41 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-391 400 Upper internals assembly: fuel alignment plate (applicable to plants with core shrouds assembled with full height shroud plates), if fatigue analysis exists Core Support Barrel Assembly: thermal shield positioning pins Stainless steel Reactor coolant and neutron flux Cumulative fatigue damage due to fatigueCracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue; loss of material due to wear Fatigue is a time

-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Yes, TLAANo B-42 (4) Mark-up of changes to GALL Report Chapter IV.B4 B4. REACT OR VESSEL I N TERNAL S (P WR) - BA BCOCK A N D WILCOX Sy stems, Structures, and Components This section addresses t he Babcock and Wilcox (B&W) pressurized

-water reactor (PWR) vessel internals and consists , which consist of components in the plenum cover and plenum cylind e r assembly , the u pper grid assembly, the control rod g u ide tube (CRGT) assembly, the core support shield a sse mbly, the core barrel assembly, the lower grid assembly, and the flow distributor assembly. Based on Regulatory Guide 1.26, "Quality Group Classifications an d Standards for Water

-, Steam-, and Radioactive

-Waste-Containing Components of Nuclear Power Plants," all structures and co mponents that comprise the reactor vessel are governed by Group A or B Quality Standards.

incor e monitoring instrumentation (IMI) guide tube a ssembly, and the flow distributor assembly.

Common mi scellaneou s material/environment combinations where aging effects are not expected to degrade the ability of th e structure o r component to perform its intended f unction for the period of extended operation are included in IV.E. Sy stem Interfaces The systems that interfa c e with the r eactor vessel internals include the r eactor pressure vessel (IV.A2). Inspection Plan An applica n t will sub m it an inspection plan for reactor i n ternals to the NRC for review and approv al with the applicati on for license renewal in accordanc e with Chapter XI.M16 A , "P WR Vessel Internals."

B-43 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-242 IV.B4-4 (R-183) Control rod guide tube (CRGT) assembly:

accessible surfaces at four screw locations (every 90 degrees) for CRGT spacer castings Cast austenitic stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Items IV.B4.RP-253 and IV.B4.RP

-258) No IV.B4.RP-242a Control rod guide tube (CRGT) assembly: CRGT spacer castings Stainless steel (including CASS)

Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-245 IV.B4-13 (R-194) Core barrel assembly: (a) upper thermal shield bolts; (b) (applicable to Crystal River Unit 3 or Davis Besse only):

surveillance specimen holder tube bolts (Davis

-Besse, only); (c) surveillance specimen tube holder (SSHT) studs , and /nuts (Crystal River Unit 3, only)or bolts Stainless steel; nickelNickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Items IV.B4.RP-247 and IV.B4.RP

-248) No IV.B4.RP-245a Core barrel assembly (applicable to Crystal River Unit 3 or Davis Besse only): surveillance specimen holder tube (SSHT) stud or bolt locking devices Nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No B-44 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-245b Core barrel assembly (applicable to CR-3 or DB only): surveillance specimen holder tube (SSHT) stud or bolt locking devices Nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-247 IV.B4-13 (R-194) Core barrel assembly: accessible lower core barrel (LCB) bolts and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Items IV.B4.RP

-245, IV.B4.RP

-246, IV.B4.RP-254, and IV.B4.RP

-256) No IV.B4.RP-247a Core barrel assembly: lower core barrel (LCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-247b Core barrel assembly: lower core barrel (LCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-249 IV.B4-12 (R-196) Core barrel assembly:

baffle plate accessible surfaces within one inch around each baffle plate flow and bolt hole plates Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Item IV.B4.RP-250) No IV.B4.RP-249a Core barrel assembly:

baffle plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No B-45 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-241 IV.B4-7 (R-125) Core barrel assembly: baffle-to-former bolts and screws Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, fatigue, and overload Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-241 241a IV.B4-7(R-125) Core barrel assembly: baffle/former assembly: (a) accessible baffle

-to-former bolts and screws; (b) accessible locking devices (including locking welds) of baffle-to-former bolts and internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking, irradiation-assisted stress -corrosion cracking , fatigue, and overload Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary Components (identified in the "Structure and Components" colu mn) and Chapter XI.M2, "Water Chemistry" (for Expansion components see AMR Items IV.B4.RP-244 and IV.B4.RP

-375)SCC mechanisms only)

No IV.B4.RP-240 IV.B4-1 (R-128) Core barrel assembly:

baffle/former assembly: (a) accessible baffle-to-former bolts and screws

(b) accessible locking devices (including welds) of baffle-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals."" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Item IV.B4.RP-243.) No IV.B4.RP-240a Core barrel assembly:

locking devices (including locking welds) of baffle-to-former bolts and internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No B-46 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-250 IV.B4-12 (R-196) Core barrel assembly: core barrel cylinder (including vertical and circumferential seam welds); former plates Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B4.RP-249) No IV.B4.RP-250a Core barrel assembly: core barrel cylinder (including vertical and circumferential seam welds); former plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-375 Core barrel assembly:

internal baffle-to-baffle

bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking, fatigue, or overload Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B4.RP-241SCC mechanisms only) No IV.B4.RP-375a Core barrel assembly:

internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-244 IV.B4-7 (R-125) Core barrel assembly; external baffle-to-baffle bolts and core barrel-to-former bolts; Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking, fatigue, and overload Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No B-47 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-244 244a IV.B4-7(R-125) Core barrel assembly; (a) external baffle

-to-baffle bolts; (b) core barrel

-to-former bolts; (c)

locking devices (including welds) of external baffle-to-baffle bolts and core barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking , or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, ""PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

" and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B4.RP-241)SCC mechanisms only)

No IV.B4.RP-243 IV.B4-1 (R-128) Core barrel assembly; (a) external baffle

-to-baffle bolts; (b) core barrel

-to-former bolts; (c) locking devices (including welds) of: external baffle-to-baffle bolts and core barrel-to-former bolts

(d) internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B4.RP-240) No IV.B4.RP-243a Core barrel assembly:

locking devices (including welds) of external baffle-to-baffle bolts and core barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-248 IV.B4-12 (R-196) Core support shield (CSS) assembly:

accessible upper core barrel (UCB) bolts and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Items IV.B4.

RP-245, IV.B4.RP

-246, IV.B4.RP-254, IV.B4.RP

-247, and IV.B4.RP-256) No B-48 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-248a Core support shield (CSS) assembly: upper core barrel (UCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-248b Core support shield (CSS) assembly: upper core barrel (UCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-253 252 IV.B4-2116 (R-191)188) Core support shield (CSS) assembly: (a) CSS cast outlet nozzles (Oconee Unit 3 and Davis-Besse, only); (b)

CSS vent valve discs top and bottom retaining rings (valve body components)

Cast austenitic stainlessStainless steel , including CASS and PH steels Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging

embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Item IV.B4.RP-242) No IV.B4.RP-252 252a IV.B4-16 (R-188) Core support shield (CSS) assembly: (a) CSS vent valve disc shaft or hinge pin (b)

CSS vent valve top retaining ring (c) CSS vent valve and bottom retaining ringrings; vent valve locking devices (valve body components)Stainless steel Reactor coolant and neutron flux Loss of fracture toughness Cracking due to thermal aging embrittlement stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No Expansion components No IV.B4.RP-251 IV.B4-15 (R-190) Core support shield (CSS) assembly: CSS top flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of preload (wear) Chapter XI.M16A, "PWR Vessel Internals" No B-49 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-251 251a IV.B4-15 (R-190) Core support shield (CSS) assembly: CSS top flange; plenum Plenum cover assembly: plenum cover weldment rib pads

and plenum cover support flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear

loss of preload (wear)

Chapter XI.M16A, "PWR Vessel Internals" Primary component (identified in the "Structure and Components" column)

No Expansion components No IV.B4.RP-256 IV.B4-25 (R-210) Flow distributor assembly: flow distributor bolts and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals

," Expansion components (identified in the "Structure and Components" column)" and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Items IV.B4.RP-247 and IV.B4.RP

-248) No IV.B4.RP-256a Flow distributor assembly: flow distributor bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-256b Flow distributor assembly: flow distributor bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to distortion or void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-259 IV.B4-31 (R-205) Incore Monitoring InstrumentationInstrument(IMI) guide tube assembly: accessible top surfaces of IMI guide tube spider-to-lower grid rib sectionssection welds Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of fracture toughness

due to thermal aging, neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see Item IV.B4.RP-260) No IV.B4.RP-259a Incore Monitoring Instrument (IMI) guide tube assembly: IMI guide tube spider-to-lower grid rib sections welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No.

B-50 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-258 IV.B4-4 (R-183) Incore Monitoring InstrumentationInstrument(IMI) guide tube assembly: accessible top surfaces of IMI Incore guide tube spider spiders (castings ) Cast austenitic stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging

, and neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see Item IV.B4.RP-242) No IV.B4.RP-258a Incore Monitoring Instrumentation (IMI) guide tube assembly: IMI guide tube spiders Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-254 IV.B4-25 (R-210) Lower grid assembly: alloy X-750 lower grid shock pad bolts and locking devices (T hree M ile I sland Unit -1, only) Nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals

," Expansion components (identified in the "Structure and Components" column) " and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Items IV.B4.RP-247 and IV.B4.R P-248) No IV.B4.RP-254a Lower grid assembly: alloy X-750 lower grid shock pad bolt locking devices (Three Mile Island Unit 1, only) Nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-254b Lower grid assembly: alloy X-750 lower grid shock pad bolt locking devices (Three Mile Island Unit 1, only) Nickel Alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-246 IV.B4-12 (R-196) Lower grid assembly: upper thermal shield (UTS) bolts and lower thermal shield (LTS) bolts Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking

'Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Items IV.B4.RP-247 and IV.B4.RP

-248) No B-51 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-246a Lower grid assembly: upper thermal shield (UTS) bolt locking devices and lower thermal shield (LTS) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-246b Lower grid assembly: upper thermal shield (UTS) bolt locking devices and lower thermal shield (LTS) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-260 IV.B4-31 (R-205) Lower grid fuel assembly: (a) accessible pads; (b) accessible pad-to-rib section welds; (c) accessible alloy X-750 dowels, cap screws and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B4.RP-259) No IV.B4.RP-260a Lower grid fuel assembly: (a) pads; (b) pad-to-rib section welds; (c) alloy X-750 dowels, cap screws and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-262 IV.B4-32 (R-203) Lower grid assembly: accessible alloy X-750 dowel-to-lower fuel assembly support pad locking welds Nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B4.RP-261) No IV.B4.RP-261 IV.B4-32 (R-203) Lower grid assembly: alloy X-750 dowel-to-guide block welds Nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Items IV.B4.RP

-262 and IV.B4.RP

-352)and Chapter XI.M2, "Water Chemistry" No B-52 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.R-53 IV.B4-37 (R-53) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Cumulative fatigue damage due to fatigue Fatigue is a time-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes, TLAA IV.B4.RP-24 IV.B4-38 (RP-24) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to pitting and crevice corrosion Chapter XI.M2, "Water Chemistry" No IV.B4.RP-376 Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Reduction in ductility and fracture

toughness due to neutron irradiation Ductility - Reduction in Fracture Toughness is a TLAA (BAW-2248A) to be evaluated for the period of extended

operation.

See the SRP, Section 4.7, "Other Plant-Specific TLAAs," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes, TLAA IV.B4.RP-238 382 IV.B4-42 (R-179) Reactor vessel internal internals: ASME Section XI, Examination Category B-N-3 core support structure components (inaccessible locations)

Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue, stress corrosion cracking, and or irradiation

- assisted stress corrosion cracking

loss of material due to wear Chapter XI.M2, "Water Chemistry,"

M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" or Chapter XI.M16A, "PWR Vessel Internals"," by invoking applicable 10 CFR 50.55a and ASME Section XI inservice inspection requirements Yes, if accessible Primary, Expansion or Existing program components indicate aging effects that need management No B-53 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-239 352 Upper grid assembly: alloy X-750 dowel-to-upper fuel assembly support pad welds (all plants except Davis-Besse)Reactor vessel internal components (inaccessible locations)

Stainless steel; nickelNickel alloy Reactor coolant and neutron flux Loss of fracture toughnessCracking due to neutron irradiation embrittlement; change in dimension due to void swelling; loss of preload due to thermal and irradiation enhanced stress relaxation; loss of material due to wear-corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Yes, if accessible Primary, Expansion or Existing program components indicate aging effects that need manag ement No IV.B4.RP-236 Reactor internal "No Additional Measures" componentsReactor vessel internal components with no additional measures Stainless steel; nickel alloy Reactor coolant and neutron flux No additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience existsCracking due to stress corrosion cracking , and irradiation

-assisted stress corrosion cracking Chapter XI.M2, "Water Chemistry" and Chapter XI.M16A, "PWR Vessel Internals" Note: Components with no additional measures are not uniquely identifies in GALL tables

- Components with no additional measures are defined in Section 3.3.1 of MRP

-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" No IV.B4.RP-400 Core support shield assembly: upper (top) flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress-corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No B-54 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-237401 Reactor vessel internal components with no additional measures Core support shield assembly: upper (top) flange weld Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement

change in dimension due to void swelling; loss of preload due to thermal and irradiation enhanced stress relaxation; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Note

Components with no additional measures are not uniquely identified in GALL tables

- Components with no additional measures are defined in Section 3.3.1 of MRP

-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" No IV.B4.RP-382 IV.B4-42(R-179) Reactor vessel internals: core support structure Stainless steel; nickel alloy; cast austenitic stainless steel Reactor coolant and neutron flux Cracking, or Loss of material due to wear Chapter XI.M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" No IV.B4.RP-352 Upper grid assembly: alloy X-750 dowel-to-upper fuel assembly support pad welds (all plants except Davis

-Besse) Nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B4.RP-261) No B-55 (5) Mark-up of changes to GALL Report Chapter IX

.C and IX.G IX.C Selected Definitions & Use of Terms for Des c ribing and Standardizing MATE RIAL S Stainless st eel Products gr ouped under the term "stainless stee l" include wrought or forged auste nitic, ferrit i c, martensitic, precipitation

-hardened (PH), or duplex stainless steel (Cr content >11

%). These stainless stee ls may be fabricated using a wro ught or cast process. Th ese materials are susceptible t o a variety o f aging effects and mechanisms, including loss of material due to pittin g and crevice corrosion, a nd crackin g due to stress corrosion cracking.

In some cases, when the recommen ded AMP an aging effect is app licable to all of the same for PH variou s stainless ste e l or cast cat egories, it can be assu med that the term "stainless steel" in the "Material" colum n of an AMR line-item in the GALL Report encompasse s all stainless ste e l types. Cast austenit i c stainless ste e l (CASS) as f o r stainless steel, PH stainless steel or CASS are included as a part of the stainle s s steel classif i cation. However, C ASS is quite suscept ible to loss of fra c ture toughness d ue to thermal and neutr on irradiatio n embrittleme n t. Therefore, wh en this aging effect is being considered, CASS In ad dition, MRP-227-A indica tes that PH stainless steels or m a rtensitic st ainless steels may be susceptible t o loss of fra c ture toughn ess by a thermal aging mechanism. Therefore, when loss of fracture tou ghness due to thermal and neutron irradiation embrittlement is an applicable a g ing effect a nd mechanism for a component in the GALL Report, the CASS, PH sta i nless steel, or martensitic stainless ste e l designat ion is spe c ifi c ally identified de signated in an AMR line-item.

Steel with st ainless steel cladding a l so may be co nsidered stainless ste e l when the aging effect is asso ciate d with the stainless ste e l surface of the material, rather than the composite volume of the material.

Exa m ples of stainless st eel designat ions that co mprise this category include A-286, SA193-Gr.

B8, SA193-Gr. B8M, Gr. 660 (A-286), SA193-6, SA193-Gr. B8 or B-8M, SA453, Type 416, Type 403, 410, 420 , and Types 431 martensitic stainless ste e ls, Type 15-5, 17-4, and 13-8-Mo PH stainless steels, and SA-193, Gra de B8 and B8M bolting materials.

Exa m ples of wrought austenitic stain l ess materia l s that comprise this category include Type 304, 304NG, 304L, 308, 308L, 3 09, 309L, 31 6 , and 347 , 403, and 41 6.. Exa m ples of CASS designations that comprise th is category

B-56 include CF-3, -8, -3M, CF3, CF3M, CF8 and -8M.CF8 M. [Ref. 6, 7

], 30] IX.G References

30. Welding Handbook, Seventh Edition, Volume 4, Metals and Their Welda b ility, American Welding So ciety, 1984, p.76-145.

B-57 Appendix B, Section 2 - Mark-up of Changes to the SRP-LR In the mark-up, red or green strikethr ough text indicates a de letion and blue underline text indicates an insertion.

Green text i ndicates a move, where a double strikethrough indicates th e original lo ca tion of the te xt and a double underlin e indicate s t he final lo ca tion of the moved text.

(1) Mark-up of changes to S R P-LR Tabl e 3.0-1 Ta ble 3.0-1 FSA R Supple m ent for A g ing M a na ge me nt of A p plic a b le Sy s t e m s G A LL Chapter G A LL Progra m De sc ription of Progra m Imple m e n ta tion Sc he dule A p plicable GA L L Re port a nd S R P-LR Chapter Refer e nce s X I.M16A PWR Vessel Internals The program relie s on impl ementation of the inspe c t i on and eval u a tion guidelin es in EPRI Tech nical Rep o rt No. 101 659 6 1022 863 (MRP-227-A) and EPRI Te chni cal Repo rt No. 1016 609 (MRP-228) to ma nage the aging effe cts on the rea c to r vessel internal com p onent s. This prog ram i s use d to mana ge (a) var i ous for m s of cra c king, in cl uding st re ss cor r o s io n cra c kingS C C , primary wate r stre ss cor r o s io n cr a cki ngP WS C C , irradiatio n-as sist e d st re s s co rro sio n c r ac kin g (IASCC), or an d crackin g d ue to fatigue/cycli c al loading; (b) loss of material in du ced by wear; (c) loss of fractu re toug hne ss d ue to either thermal aging or , neutro n irradiatio n embrittleme n t , or void swell i ng; (d) dimen s ion a l chang es a nd p o tential loss of fractu re toughn ess due to void swelling an d irra diation g r o w th or distortio n; an d (e) lo ss of preloa d due to thermal an d irra diation-e nhan ce d stre ss relaxat i on or cre ep. Program sho u ld be implem ent ed prio r to perio d of extended operation GALL IV / SRP 3.1 (2) Mark-up of changes to S R P-LR Secti on 3.1.2, "Acceptance C r iteria" 3.1.2.2.9 Re m o ved as a result of LR-ISG-201 1-04 Cracking due to Stress Corro sion Cracking and Irradiation

-Assisted Stress Co rrosion Cra cking Cracking du e to SCC an d irradiation

-assisted str e ss corros i o n cracking (I ASCC) could occur in inaccessible location s fo r stainless st eel and nickel

-alloy Primary and Exp ansion PWR reactor v e ssel inter nal components. If agin g effects are identified in accessible locations, th e GALL Report recommends furt her evaluation of the aging effects in inaccessible location s on a plant

-specific basis to ensure t hat this agin g effect is a dequately managed. Acc eptance crit eria are described in Branch Technical Position RLSB

-1 (Appendix A.

1 of this SRP

-LR).

B-58 3.1.2.2.10 Re m o ved as a result of LR-ISG-201 1-04 Loss of Fracture Toughness due to Neutron Irradiation Embrittlement; Change in Dimension due to Void Sw ell i n g; Loss of Preload due to Stress Relaxation; or Loss of Material du e to Wear Loss of fract u re toughness due to ne utron irradia t ion embrittlement, change in dimension due to void swellin g, loss of pr eload due to stress rel a xation, or loss of materia l due to wear could occur in inacce ssible locatio n s for stainle s s steel and nickel-a lloy Primary and Expansion PWR reactor vessel inter nal components. If agin g effects are identified in accessible locations, th e GALL Report recommends furt her evaluation of the aging effects in inaccessible location s on a plant

-specific basis to ensure t hat this agin g effe ct is a dequately managed. Acceptance crit eria are described in Branch Technical Position RLSB

-1 (Appendix A.

1 of this SRP

-LR). 3.1.2.2.12 Re m o ved as a result of LR-ISG-201 1-04 Cracking due to Fatigue EPRI 10165 96, Materials Reliability Progra m: Pressurized W a ter Reactor Internals In spection and Evaluation Guidelin es (MRP-22 7-Rev. 0) id entifies cracking due to f a tigue as an aging effect that can occur for the lo wer flange weld in the core support barrel asse mbly, fuel alignment plate in the upper internals a s sembly, and core suppor t plate lower support stru cture in PW R internals designed by Combustion Engineering. The GALL Report recommends that inspect i on for cracking in t h is compon ent be perfo rmed if acce ptable fatigu e life cannot be demonstrated by TLAA through the perio d of extended operation as defined in 10 CFR 5 4.3. 3.1.2.2.13 Re m o ved as a result of LR-ISG-201 1-04 Cracking due to Stress Corro sion Cracking and Fatigue Cracking du e to stress corrosion cra cking and fa tigue could occur in nickel alloy cont rol rod guide tube assemblies, guid e tube support pins expose d to reacto r coolant, an d neutron flu

x. The GALL Report, AMR Ite m IV.B2.RP-35 5, recomme nds further evaluation of a plant

-spe c ific AMP to ensure this aging effe ct is adequa tely manage

d. Acceptan ce criteria ar e described in Branch Technical P o sition RLS B-1 (Appendix A.1 of this SRP

-LR). 3.1.2.2.14 Re m o ved as a result of LR-ISG-201 1-04 Loss of Material du e to Wear Loss of material due to wear could occur in nickel alloy cont rol rod guide tube assemblies, guid e tube support pins and in Zircaloy-4 in core instrumentation low e r thimble tubes exposed to reactor coolant, and neutron flux. The GALL Report, AMR Items IV.

B 2.RP-356 and IV.B3.RP-357, recommend s further evaluation of a plant-specific AMP to en sure this ag ing effect is adequately managed. Acceptance criteria are d e scribed in Branch Technical Position RLSB

-1 (Appendix A.

1 of this SRP

-LR). (3) Mark-up of changes to S R P-LR Secti on 3.1.3, "Review Procedures

" 3.1.3.2.9 Re m o ved as a result of LR-ISG-201 1-04 Cracking due to Stress Corro sion Cracking and Irradiation

-Assisted Stress Co rrosion Cra cking The GALL Report recommends further evaluation of crackin g due to SCC and IASCC for inaccessible location s fo r P rimary an d E xpansion PWR reactor vessel int e rnal components if aging effect s are identif ied for these components in acce ssible location

s. The reviewe r reviews the applican t's proposed program on a case

-by-case basis to ensure that an adequate program B-59 will be in pla c e for the management of these agi ng effects consi stent wit h the action item documented in the staff

's safety evaluation for MRP

-227, Re vision 0.. 3.1.3.2.10 Re m o ved as a result of LR-ISG-201 1-04 Loss of Fracture Toughness due to Neutron Irradiation Embrittlement; Change in Dimension due to Void Sw ell i n g; Loss of Preload due to Stress Rel axation; or Loss of Material due to Wear The GALL Report recommends further evaluation of loss of f r acture toug hness due t o neutron irradiation e m brittlement, change in dimension due to void swelling, loss of preload d ue to stress relaxation, or loss of mat e rial due to wear for inaccessible lo cat ions for P rimary and E xpansion PWR reactor vessel inte rnal components , if agin g effects are identified fo r these components in access ible l o cations. Th e reviewer reviews the applicant's pr oposed prog ram on a case

-by-case basis to en sure that an adequate pr ogram will be in place fo r the management of these aging effects con s istent with th e action ite m document ed in the sta ff's safety evaluation for MRP-227, Revision 0

. 3.1.3.2.12 Re m o ved as a result of LR-ISG-201 1-04 Cracking due to Fatigue The GALL Report recommends further evaluation of crackin g due to fatigue in the lo wer flange weld in the core support barrel asse mbly , fuel alignment plate in the u pp er internals assembly , and core su pport plate in the l ower support stru cture in PW R intern als designed by Combustion Engineering

. The reviewer determines whether a TLAA has been performed for each component, consiste nt with the actio n item docum ented in the staff's saf e ty evaluation for MRP

-227, Revision 0

. If a TLAA has not b een performed, the reviewer determi nes whether the applicant ha s performed an evaluation to identif y the potential location a nd extent of fatigue cracking f o r each component consist ent with the action item documented in the staff

's safety evaluation for MRP

-227 , Revision 0.

3.1.3.2.13 Re m o ved as a result of LR-ISG-201 1-04 Cracking due to Stress Corro sion Cracking and Fatigue The GALL Report recommends further evaluation of crackin g due to stre ss corrosion cracking and fatigue in the nicke l alloy control rod guide tu be assemblies, guide tu be support p i ns exposed to reactor coola n t, and neut ron flux. The reviewer re views the applicant's pro posed program on a case

-by-case basi s to ensure that an adequate program will be in pla c e for the managemen t of these ag ing effects co nsistent wit h the action item documented in the staff's safety evaluation for MRP

-227, Revi sion 0. 3.1.3.2.14 Re m o ved as a result of LR-ISG-201 1-04 Loss of Material du e to Wear The GALL Report recommends further evaluation of loss of material due to wear in nickel a lloy control rod g u ide tube assemblies, g u ide tube su pport pins a nd in Zircalo y-4 incore instrumentation lower thimble tubes exposed to reactor coola n t, and neut ron f lux. The reviewer reviews the applicant's p r oposed pro g ram on a case

-by-case basis to en sure that an adequate program will be in place for the management of these aging effects con s istent with th e action item documented in the staff's safety evaluation for MRP-227 , Revision 0

.

B-60 (4) Mark-up of changes to S R P-LR Tabl e 3.1-1 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 3 BW R/ PWR Stainl ess steel or nicke l allo y r eactor ve ssel inter nal compo nents e x pose d to reactor coo l ant and ne utron flux Cumul a tive fati gue d a ma ge due to fatig u e F a tigue is a T L AA eval uate d for the per iod of ext end ed o per ation (Se e SRP, Section 4.3 "Metal F a tigue," for ac ceptab le methods to co mpl y w i th 10 CF R 54.2 1 (c)(1) Yes, T L AA (See subsecti on 3.1.2.2.1) IV.B1.R-53 IV.B2.RP-303 IV.B3.RP-339 IV.B4.R-53 IV.B3.RP-389 IV.B3.RP-390 IV.B3.RP-391 IV.B1-14 (R-5 3) IV.B2-31 (R-5 3) IV.B3-24 (R-5 3) IV.B4-37 (R-5 3) N/A N/A N/A 15 PW R Stainl ess steel Babcock &

Wilcox (including CASS, martensitic SS, and PH SS) and n i ckel all o y react o r vessel i n terna l compo nents expos ed to rea c tor coola n t and n eutro n flu x Red u ction in of ductilit y a nd fracture toug hn ess due to neutro n irrad i at ion embrittlem ent, and for CASS, marten sitic SS, and PH SS due to thermal agi ng embrittlem ent Ductilit y - Re du ction in fF racture tT oughn ess is a T L AA to be evalu a ted for the peri od of e x te nde d oper ation., See the SRP, Section 4.7, "Other Plant-Specific T L AAs," for accepta b l e methods for of meetin g the re quir e ments of 10 CF R 54.2 1 (c)(1).). Yes, T L AA (See subsecti on 3.1.2.2.3.3) IV.B4.RP-376 N/A 23 PWR Stainl ess steel or nicke l allo y PWR re a c tor vessel intern al comp o nents (inacc e ssible locations) expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosio n crack i ng, an d irradi atio n-assisted stress corrosion crack i ng Cha p ter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" Yes, if accessible Pri m a r y , Ex pan si on o r Existi ng pr ogra m compo nents i n dicate agi ng effects that need mana geme n t (See subsecti on 3.1.2.2.9) IV.B2.RP-268 IV.B3.RP-309 IV.B4.RP-238 N/A N/A N/A 24 PWR Stainl ess steel or nicke l allo y PWR re a c tor vessel intern al comp o nents (inacc e ssible locations) expos ed to rea c tor coola n t and n eutro n flu x Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent; or chan ges i n dime nsio n du e to void s w e l l i n g; or los s of preloa d due to therm a l and irradi atio n en h ance d stress rela xation; or l o ss of material due to w e ar Chapter X I.M16A, "PWR Vessel Internals" Yes, if accessible Pri m a r y , Ex pan si on o r Existi ng pr ogra m compo nents i n dicate agi ng effects that need mana geme n t (See subsecti on 3.1.2.2.10) IV.B2.RP-269 IV.B3.RP-311 IV.B4.RP-239 N/A N/A N/A B-61 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 26 PWR Stainl ess steel Comb ustion Engi neer in g co re supp ort barrel assembly

lo w e r flang e w e l d e x pose d to reactor coo l ant and ne utron flux; U pper i n te rnals assembly: fuel alig nme n t plate (a ppl ica b l e to plants w i t h core shr o uds assemb led w i t h full he ig ht shrou d plat es) expos ed to reactor coo l ant and ne utron flux; Lo w e r su pport structure: core supp ort plate (a ppl ica b l e to plants w i t h a core su p port plate) expos ed to rea c tor coola n t and n eutro n flu x Cracking due to fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y," if fatigu e life cann ot be confi rmed b y TL A A Yes, evaluate to determi ne the potenti a l locati ons a nd e x te nt of fatigue crack i n g (See subsecti on 3.1.2.2.12) IV.B3.RP-333 IV.B3.RP-338 IV.B3.RP-343 N/A 27 PWR Nickel alloy W e stinghouse control ro d gui de tube assemb lies, gu ide tub e supp ort pins e x pose d to reactor coo l ant and ne utron flux Crackin g du e to stress corrosio n crack i ng a nd fatigue A plant-sp ecific agin g mana geme n t p r ogram is to be eval uate d Yes, plant

-sp e c ific (See subsecti on 3.1.2.2.13) IV.B2.RP-355 N/A 28 PWR Nickel alloy W e stinghouse control ro d gui de tube assemb lies, gu ide tub e supp ort pins, a nd Z i rcal o y-4 Comb ustion Engi neer in g incor e instrum entatio n thimbl e tubes e x p o se d to reactor coo l ant and ne utron flux Loss of materi al du e to w e ar A plant-sp ecific agin g mana geme n t p r ogram is to be eval uate d Yes, plant

-sp e c ific (See subsecti on 3.1.2.2.14) IV.B2.RP-356 IV.B3.RP-357 N/A N/A 28 PWR Stainl ess steel Comb ustion Engi neer in g "Existi ng Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Loss of materi al du e to w e ar; cracki ng due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-400 N/A B-62 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 32 PW R Stainl ess steel, nickel a llo y, or CASS react o r vessel intern als, core supp ort structure (n o t al re ady referenc ed as ASME Section XI Exa m inati on Categ o r y B-N-3 core su p port structure comp one nts in MRP-22 7-A), exp o se d to reactor coo l ant and ne utron flux Crackin g , or lo ss of material due to w e ar Chapter X I.M1, "ASME Section XI Inse rvice Inspection, Subsections IW B, IW C, and IW D" or Chapter X I.M16A, "PWR Vessel Internals," invoking app lica b le 10 CF R 50.55 a and ASME Sec t ion XI inservic e ins p e c tion requ ireme n ts for these compo nents No IV.B2.RP-382 IV.B3.RP-382 IV.B4.RP-382 IV.B2-26 (R-1 4 2) IV.B3-22 (R-1 7 0) IV.B4-42 (R-1 7 9) 51 PWR Stainl ess steel or nicke l-allo y B abcock & W ilcox reactor inter nal compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assisted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" No IV.B4.RP-236 IV.B4.RP-241 IV.B4.RP-244 IV.B4.RP-245 IV.B4.RP-246 IV.B4.RP-247 IV.B4.RP-248 IV.B4.RP-254 IV.B4.RP-256 IV.B4.RP-261 IV.B4.RP-262 IV.B4.RP-352 IV.B4.RP-375 N/A IV.B4-7(R-125) IV.B4-7(R-125) IV.B4-13(R-194) IV.B4-12(R-196) IV.B4-13(R-194) IV.B4-12(R-196) IV.B4-25(R-210) IV.B4-25(R-210) IV.B4-32(R-203) IV.B4-32(R-203) N/A N/A 52 PWR Stainl ess steel or nicke l-allo y C o mb usti on Engi neer in g re actor intern al compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assisted stress corrosio n crack i ng, or fatigue Cha p ter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" No IV.B3.RP-306 IV.B3.RP-312 IV.B3.RP-313 IV.B3.RP-314 IV.B3.RP-316 IV.B3.RP-320 IV.B3.RP-322 IV.B3.RP-323 IV.B3.RP-324 IV.B3.RP-325 IV.B3.RP-327 IV.B3.RP-328 IV.B3.RP-329 IV.B3.RP-330 IV.B3.RP-334 IV.B3.RP-335 IV.B3.RP-342 IV.B3.RP-358 N/A IV.B3-2(R-149) N/A IV.B3-9(R-162) IV.B3-9(R-162) IV.B3-9(R-162) N/A N/A N/A N/A IV.B3-15(R-155) IV.B3-15(R-155) IV.B3-15(R-155) IV.B3-23(R-167) IV.B3-23(R-167) IV.B3-23(R-167) N/A N/A B-63 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 53 PWR Stainl ess steel or nicke l-allo y W e sti ngh ouse re actor intern al comp o nents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assisted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Internal s," and Chapter X I.M2, "Water Chemistr y" No IV.B2.RP-265 IV.B2.RP-271 IV.B2.RP-273 IV.B2.RP-275 IV.B2.RP-276 IV.B2.RP-278 IV.B2.RP-280 IV.B2.RP-282 IV.B2.RP-286 IV.B2.RP-289 IV.B2.RP-291 IV.B2.RP-293 IV.B2.RP-294 IV.B2.RP-298 IV.B2.RP-301 IV.B2.RP-346 IV.B2.RP-387 N/A IV.B2-10(R-125) IV.B2-10(R-125) IV.B2-6(R-128) IV.B2-8(R-120) IV.B2-8(R-120) IV.B2-8(R-120) IV.B2-8(R-120) IV.B2-16(R-133) IV.B2-20(R-130) IV.B2-24(R-138) IV.B2-24(R-138) IV.B2-24(R-138) IV.B2-28(R-118) IV.B2-40(R-112) N/A N/A 51a PWR Stainl ess steel or nicke l allo y B abcock & W ilcox reactor inter nal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B4.RP-241 IV.B4.RP-241a IV.B4.RP-242a IV.B4.RP-247 IV.B4.RP-247a IV.B4.RP-248 IV.B4.RP-248a IV.B4.RP-249a IV.B4.RP-252a IV.B4.RP-256 IV.B4.RP-256a IV.B4.RP-258a IV.B4.RP-259a IV.B4.RP-261 IV.B4.RP-400 IV.B4-7 (R

-125) N/A N/A IV.B4-13 (R-194) N/A IV.B4-25 (R-210) N/A N/A N/A IV.B4-25 (R-210) N/A N/A N/A IV.B4-32 (R-203) N/A 51b PWR Stainl ess steel or nicke l allo y B abcock & W ilcox reactor inter nal "Exp ansi on" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, fatigu e, or overl oad Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B4.RP-244 IV.B4.RP-244a IV.B4.RP-245 IV.B4.RP-245a IV.B4.RP-246 IV.B4.RP-246a IV.B4.RP-254 IV.B4.RP-254a IV.B4-7 (R

-125) N/A IV.B4-13 (R-194) N/A IV.B4-12 (R-196) N/A IV.B4-25 (R-210) N/A B-64 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item IV.B4.RP-260a IV.B4.RP-262 IV.B4.RP-352 IV.B4.RP-250a IV.B4.RP-375 N/A IV.B4-32 (R-203) N/A N/A N/A 52a PWR Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g re actor intern al "Primar y" com ponents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-312 IV.B3.RP-314 IV.B3.RP-322 IV.B3.RP-324 IV.B3.RP-326a IV.B3.RP-327 IV.B3.RP-328 IV.B3.RP-342 IV.B3.RP-358 IV.B3.RP-362a IV.B3.RP-363 IV.B3.RP-338 IV.B3.RP-343 IV.B3-2 (R

-149) IV.B3-9 (R

-162) N/A N/A N/A IV.B3-15 (R-155) IV.B3-15 (R-155) N/A N/A N/A N/A N/A N/A 52b PWR Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g re actor intern al "Exp ans ion" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-313 IV.B3.RP-316 IV.B3.RP-323 IV.B3.RP-325 IV.B3.RP-329 IV.B3.RP-330 IV.B3.RP-333 IV.B3.RP-335 IV.B3.RP-362c NA IV.B3-9 (R

-162) N/A N/A IV.B3-12 (R-155) IV.B3-23 (R-167) N/A IV.B3-23 (R-167) N/A 52c PWR Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g reactor intern al "E xisti ng Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-320 IV.B3.RP-334 IV.B3-9 (R

-162) IV.B3-23 (R-167)

B-65 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 53a PWR Stainl ess steel or nicke l allo y W e sti ngh ouse reactor inter nal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B2.RP-270a IV.B2.RP-271 IV.B2.RP-275 IV.B2.RP-276 IV.B2.RP-280 IV.B2.RP-298 IV.B2.RP-302 IV.B2.RP-387 N/A IV.B2-10 (R-125) IV.B2-6 (R

-128) IV.B2-8 (R

-120) IV.B2-8 (R

-120) IV.B2-28 (R-118) N/A N/A 53b PWR Stainl ess steel W e stingh ouse reactor intern al "E xpa n s ion" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B2.RP-273 IV.B2.RP-278 IV.B2.RP-286 IV.B2.RP-291 IV.B2.RP-291a IV.B2.RP-291b IV.B2.RP-293 IV.B2.RP-294 IV.B2.RP-387a IV.B2-10 (R-125) IV.B2-8 (R

-120) IV.B2-16 (R-133) IV.B2-24 (R-138) N/A N/A IV.B2-24 (R-138) IV.B2-24 (R-138) N/A 53c PWR Stainl ess steel or nicke l allo y W e sti ngh ouse reactor inter nal "Existin g Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B2.RP-289 IV.B2.RP-301 IV.B2.RP-345 IV.B2.RP-346 IV.B2.RP-399 IV.B2.RP-355 IV.B2-20 (R-130) IV.B2-40 (R-112) N/A N/A N/A N/A 54 PW R Stainl ess steel bottom mounte d instru ment sy stem flux thimble tubes (w ith or w i t h o u t ch ro me pl a t i ng) e x po sed to reactor cool ant and neutro n flu x (W esting ho use "Ex i sting Progr a ms" compo nents) Loss of materi al du e to we a r Chapter X I.M16A, "PWR Vessel Internals,"

and or Chapter X I.M37, ""Flu x T h imble T ube Inspecti on"" No IV.B2.RP-284 IV.B2-12(R-143) IV.B2-13 (R-145) 55 PWR Stainl ess steel thermal shield assem b ly , thermal shiel d fle x ur es expos ed to reactor coo l ant and ne utron flux Crackin g du e to fatigue; Loss of materi al du e to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-302 N/A 55a PWR Stainl ess steel or nicke l allo y B abcock and W ilc o x reactor inter nal "No Additi ona l Mea s ures" N o a d d i ti on al ag i n g mana geme n t for reactor intern al "No Ad ditio nal Measur es" co mpon ents Chapter X I.M16A, "PWR Vessel Internals" No IV.B4.RP-236 NA B-66 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item compo nents e x pose d to reactor coo l ant and ne utron flux unl ess requ ire d b y ASME Section XI, Examin ation Categ o r y B-N-3 or relev ant oper ating e x pe rienc e inval i d a tes MR P-227-A. 55b PWR Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g re actor intern al "No Add i tion al Measur es" compo nents e x pose d to reactor coo l ant and ne utron flux N o a d d i ti on al ag i n g mana geme n t for reactor intern al "No Ad ditio nal Measur es" co mpon ents unl ess requ ire d b y ASME Section XI, Examin ation Categ o r y B-N-3 or relev ant oper ating e x pe rienc e inval i d a tes MR P-227-A. Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-306 NA 55c PWR Stainl ess steel or nicke l allo y W e sti ngh ouse reactor inter nal "No Additi ona l Mea s ures" compo nents e x pose d to reactor coo l ant and ne utron flux N o a d d i ti on al ag i n g mana geme n t for reactor intern al "No Ad ditio nal Measur es" co mpon ents unl ess requ ire d b y ASME Section XI, Examin ation Categ o r y B-N-3 or relev ant oper ating e x pe rienc e inval i d a tes MR P-227-A. Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-265 NA 56 PWR Stainl ess steel or nicke l-allo y C o mb usti on Engi neer in g re actor intern al compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent; or chan ges i n dime nsio n du e to void s w e l l i n g; or los s of preloa d due to therm a l and irradi atio n en h ance d stress rela xation; or l o ss of material due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-307 IV.B3.RP-315 IV.B3.RP-317 IV.B3.RP-318 IV.B3.RP-319 IV.B3.RP-326 IV.B3.RP-331 IV.B3.RP-332 IV.B3.RP-336 IV.B3.RP-359 IV.B3.RP-360 IV.B3.RP-361 IV.B3.RP-362 IV.B3.RP-363 IV.B3.RP-364 IV.B3.RP-365 N/A IV.B3-7(R-165) IV.B3-7(R-165) IV.B4-8(R-163) IV.B3-9(R-162) N/A N/A IV.B3-17(R-156) IV.B3-22(R-170) N/A N/A N/A N/A N/A N/A N/A B-67 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item IV.B3.RP-366 N/A 58 PWR Stainl ess steel or nicke l-allo y B abcock & W ilcox reactor inter nal compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent; or chan ges i n dime nsio n du e to void s w e l l i n g; or los s of preloa d due to therm a l and irradi atio n en h ance d stress rela xation; or l o ss of material due to w e ar Chapter X I.M16A, "PWR Ves sel Internals" No IV.B4.RP-237 IV.B4.RP-240 IV.B4.RP-242 IV.B4.RP-243 IV.B4.RP-249 IV.B4.RP-250 IV.B4.RP-251 IV.B4.RP-252 IV.B4.RP-253 IV.B4.RP-258 IV.B4.RP-259 IV.B4.RP-260 N/A IV.B4-1(R-128) IV.B4-4(R-183) IV.B4-1(R-128) IV.B4-12(R-196) IV.B4-12(R-196) IV.B4-15(R-190) IV.B4-16(R-188) IV.B4-21(R-191) IV.B4-4(R-183) IV.B4-31(R-205) IV.B4-31(R-205) 59 PWR Stainl ess steel or nicke l-allo y W e sti ngh ouse re actor intern al comp o nents expos ed to rea c tor coola n t and n eutro n flu x Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent; or chan ges i n dime nsio n du e to void s w e l l i n g; or los s of preloa d due to therm a l and irradi atio n en h ance d stress rela xation; or l o ss of material due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-267 IV.B2.RP-270 IV.B2.RP-272 IV.B2.RP-274 IV.B2.RP-281 IV.B2.RP-285 IV.B2.RP-287 IV.B2.RP-288 IV.B2.RP-290 IV.B2.RP-292 IV.B2.RP-295 IV.B2.RP-296 IV.B2.RP-297 IV.B2.RP-299 IV.B2.RP-300 IV.B2.RP-345 IV.B2.RP-354 IV.B2.RP-386 IV.B2.RP-388 N/A IV.B2-1(R-124) IV.B2-6(R-128) IV.B2-6(R-128) IV.B2-9(R-122) IV.B2-14(R-137) IV.B2-17(R-135) IV.B2-18(R-132) IV.B2-21(R-140) IV.B2-21(R-140) IV.B2-22(R-141) N/A N/A IV.B2-34(R-115) IV.B2-33(R-108) N/A N/A N/A N/A 56a PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y C o mb usti on Engi neer in g re actor intern al "Primar y" com ponents expos ed to rea c tor coola n t and n eutro n flu x Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-315 IV.B3.RP-318 IV.B3.RP-359 IV.B3.RP-360 IV.B3.RP-362 IV.B3.RP-364 IV.B3.RP-366 IV.B3.RP-365 IV.B3.RP-326 IV.B3-7 (R

-165) IV.B3-8 (R

-163) N/A N/A N/A N/A N/A N/A N/A B-68 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar 56b PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS)

Comb ustion E ngi neer in g "Exp ans ion" re actor intern al compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-317 IV.B3.RP-331 IV.B3.RP-359a IV.B3.RP-361 IV.B3.RP-362b IV.B3-7 (R

-165) N/A N/A N/A N/A 56c PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y C o mb usti on Engi neer in g re actor intern al "Ex i sting Progr a ms" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-319 IV.B3.RP-332 IV.B3.RP-334a IV.B3.RP-336 IV.B3.RP-357 IV.B3-9 (R

-162) IV.B3-17 (R-156) N/A IV.B3-22 (R-170) N/A 58a PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y B abcock & W ilcox reactor inter nal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to w e ar; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B4.RP-240 IV.B4.RP-240a IV.B4.RP-242 IV.B4.RP-247b IV.B4.RP-248b IV.B4.RP-249 IV.B4.RP-251 IV.B4.RP-251a IV.B4.RP-252 IV.B4.RP-254b IV.B4.RP-256b IV.B4.RP-258 IV.B4.RP-259 IV.B4.RP-401 IV.B4-1 (R

-128) N/A IV.B4-4 (R

-183) N/A N/A IV.B4-12 (R-196) IV.B4-15 (R-190) N/A IV.B4-16 (R-188) N/A N/A IV.B4-4 (R

-183) IV.B4-31 (R-205) N/A B-69 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 58b PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y B abcock & W ilcox reactor inter nal "Exp ansi on" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B4.RP-245b IV.B4.RP-246b IV.B4.RP-254b IV.B4.RP-260 IV.B4.RP-243 IV.B4.RP-243a IV.B4.RP-250 IV.B4.RP-375a N/A N/A N/A IV.B4-31 (R-205) IV.B4-1 (R

-128) N/A IV.B4-12 (R-196) N/A 59a PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y W e sti ngh ouse re actor internal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-270 IV.B2.RP-272 IV.B2.RP-296 IV.B2.RP-297 IV.B2.RP-302a IV.B2.RP-354 IV.B2.RP-388 IV.B2.RP-300 IV.B2-1 (R

-124) IV.B2-6 (R

-128) N/A N/A N/A N/A N/A N/A 59b PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS)

W e stingh ouse reactor intern al "E xpa n s ion" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-274 IV.B2.RP-278a IV.B2.RP-287 IV.B2.RP-290 IV.B2.RP-290a IV.B2.RP-290b IV.B2.RP-292 IV.B2.RP-295 IV.B2.RP-388a IV.B2-6 (R

-128) N/A IV.B2-17 (R-135) IV.B2-21 (R-140) N/A N/A IV.B2-21 (R-140) IV.B2-22 (R-141) N/A 59c PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y W e sti ngh ouse re actor Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-285 IV.B2.RP-288 IV.B2.RP-299 IV.B2.RP-356 IV.B2-14 (R-137) IV.B2-18 (R-132) IV.B2-34 (R-115) N/A B-70 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item intern al "E xisti ng Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar C-1 Appendix C STAFF RESPONSE TO PUBLIC COMMENTS ON DRAFT LICENSE RENEWAL IN TERIM STAFF GUIDANCE 2011-04 C-2 Source of Comments I. Comments from Jean Smith, Electric Power Research Institute Materials Reliability Program (EPRI-MRP) and the Pressurized Water Reactor Owners Group Materials Subcommittee (PWROG-MSC) (ADAMS Accession No. ML12146A267) II. Comments from Mark Richter, Nuclear Energy In stitute (NEI) (ADAMS Accession No. ML12144A147)

1. Source ID Summary of Comment Response 1 I-1 The NRC reviewed and approved with limitations MRP-227 Revision 0, and subsequently, MRP-227-A was published to incorporate the SER additions. All needed actions for licensees are contained in MRP-227-A. As a result, it is appropriate for the NRC to review a licensee's PWR reactor internals aging management program against the criteria contained in MRP-227-A. As such, it is not necessary to include all the details currently in NUREG-1800 and NUREG-1801 regarding PWR reactor internals, and instead, only a reference to MRP-227-A should be made. Outlining the requirements for reactor internals in the Interim Staff Guidance may lead to confusion with respect to the implementation of duplicate requirements, may cause undue NRC staff burden reconciling the documents each time MRP-227 is revised by the industry, and will likely lead to human errors in document alignment through future revisions. The staff agrees with the comment, in part, that it is not necessary to have the level of detail included in LR-ISG-2011-04 issued for public comment regarding PWR reactor vessel internal (RVI) components. However, the staff does not agree that the final LR-ISG-2011-04 should only reference MRP-227-A; instead reference to the topical report should be made only when it is appropriate. Revisions were made to eliminate duplication of information for RVIs that is detailed in MRP-227-A. The following is a summary of the revisions that have been incorporated into final LR-ISG-2011-04 as a result of this comment:

Revision to GALL Report Aging Management Program (AMP) XI.M16A In general, GALL Report AMP XI.M16A, "PWR Vessel Internals," in final LR-ISG-2011-04 references MRP-227-A in the program elements and does not delineate the MRP-227-A inspection and evaluation guidelines for PWR RVIs. In addition, areas resolved in the staff's safety evaluation (SE), Revision 1, for MRP-227 and Applicant/Licensee Action Items (A/LAI) are not addressed in GALL Report AMP XI.M16A in final LR-ISG-2011-04.

Revision to SRP-LR Table 3.1-1 Final LR-ISG-2011-04 does not incorporate specific reference to "Primary Category," "Expansion Category," or "Existing Program" inspection and evaluation guidelines into the "Rev. 2 Item" column in the aging management review (AMR) line items for PWR RVIs in SRP-LR Table 3.1-1. In addition, the "Component" column for PWR RVIs in SRP-LR Table 3.1-1 in final LR-ISG-2011-04 is based on the commodity groups and inspection categories in MRP-227-A.

Revision to GALL Tables IV.B2, IV.B3, and IV.B4 GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 do not reference inspection categories and MRP-227-A inspection and evaluation guidelines.

Revision to SRP-LR Further Evaluation Recommendations for PWR RVIs Areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04 (i.e., these SRP-LR sections were deleted and do appear in final LR-ISG-2011-04). In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the C-3 # Source ID Summary of Comment Response A/LAIs for MRP-227-A in Appendix C of the LRA.

2 I-2 Commenter referenced statement in Section 3.

1.2.2.9.A.1 of Appendix A of LR-ISG 2011-04.

This statement requires that licensees include responses to applicant action items in both Appendix C of the LRA and in appropriate further evaluation sections of the LRA. This duplication of information provides no significant value to the reviewers. It is recommended that all A/LAI responses be included only in Appendix C, so they are in an easily-referenced location. Any additional discussion of the A/LAIs in the further evaluation sections of the SRP should be limited to identifying each of the items requiring responses and any details necessary to ensure responses are adequate. Any other items requiring discussion of the A/LAI responses in further evaluation sections of the LRA should be deleted or reference made to Appendix C of the LRA.

The staff agrees with the comment that the responses to A/LAIs are to be provided in Appendix C of the LRAs. Thus, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. In addition, as a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of the staff's resolution of Source ID I-1, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.1. 3 I-3 In NUREG-1801 Revision 2 XI.M16A Program Description, last paragraph, as well as in ISG-LR-2011-04 Section 3.1.2.2.9.A.2, both an aging management program and an inspection plan are required to be submitted as part of an applicant's license renewal application.

However nowhere in these two documents is there any clear guidance on the information that should be included in an inspection plan. This ambiguity could lead to applicants submitting information that might not meet NRC needs in this area.

In order to address this situation it is requested that the aging management program and inspection plan for an applicant be clearly defined. It is proposed that the aging management program address the 10 program element recommendations for PWR RVI components in GALL AMP XI.M16A, PWR Vessel Internals (AMP XI.M16A in NUREG-1801, Revision 2). The inspection plan could be included within a program (i.e. a program/plan) or be a separate document if submitted with a license renewal application.

The industry believes these elements are satisfied by the applicable line items from Tables 4-1 through 4-9 and Tables 5-1 through 5-3 of MRP-227-A. The inspection plan submitted as part of a license renewal application (LRA) should be included in Appendix C of the LRA along with the responses to the A/LAI items since it is a requirement of A/LAI No. 8. The staff agrees, in part, with the comment in that better guidance regarding the inspection plan is needed to avoid confusion. Regulatory Issue Summary (RIS) 2011-07, "License Renewal Submittal Information For Pressurized Water Reactor Internals Aging Management," dated July 21, 2011, provides the staff's expectations for Category D plants (PWR plant licensees that had not submitted their LRAs but plan to submit an LRA in the future) to submit, for NRC staff review and approval, an AMP for vessel internals that is consistent with MRP-227-A.

An "inspection plan" is one aspect of satisfying A/LAI No. 8 of the staff's SE, Revision 1, for MRP-227. An "inspection plan" provides information about the RVI components to be inspected and a description of how they will be managed for age-related degradation (e.g., examination method, frequency, acceptance criteria, coverage, etc.). The staff expects that the details of an "inspection plan" for Category D plants will be incorporated into the LRA submittal as part of the 10-element AMP and AMR line items. Thus, consistent with RIS 2011-07, the staff does not expect Category D plants to provide a separate document that contains an "inspection plan" in response to A/LAI No. 8.

In order to avoid duplication and confusion, as part of the resolution to Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their C-4 # Source ID Summary of Comment Response responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. In doing this, the explicit reference to an "inspection plan" is avoided in the body of the AMP, and "inspection plan" is only referenced as part of A/LAI No. 8 in the staff's SE, Revision 1, for MRP-227.

However, the staff does not agree with the Commenter's general claim with respect to what satisfies an inspection plan per A/LAI No. 8, as additional guidance is outlined in the SE, Revision 1, for MRP-227, and fulfillment of that action item will depend on each applicant's plant-specific review.

4 I-4 The stipulation of appropriate inspection methodologies for these reactor internals components has already been addressed in the review of MRP-227-A. The recommended inspection methods have already been reviewed and found to be adequate to detect the relevant conditions. The AMP attribute that is at issue is not detection of aging effects; instead, the issue is the applicant's corrective action program, and the disposition of relevant conditions through supplemental examination or engineering evaluation, both of which are outside the scope of the Mandatory or Needed requirements of MRP-227-A. Standards for engineering evaluation are addressed in Section 6 of MRP-227-A and in the methodologies described in WCAP-17096. These recommendations are based on the practice used in Section XI of the ASME code and are consistent with existing aging management programs. Further justification for the use of the VT-3 examination is not necessary and should not be required by the ISG.

It is recommended that Acceptance Criteria Item 3.1.2.2.9.A.7 (Use of VT-3 Visual Inspection Techniques for Detection of Cracking) be completely eliminated and replaced by a limited requirement to address the acceptability of VT-3 as a management approach for components that 1) were not already considered for aging management in the development of MRP-227-A, 2) are evaluated to require active aging monitoring, and 3) are non-redundant. The Commenter provided justification for its recommendation.

The staff agrees with the comment, in part, that final LR-ISG-2011-04 address the acceptability of VT-3 as a management approach for certain components. Thus, final LR-ISG-2011-04 does not incorporate SRP-LR Sections 3.1.2.2.9.A.7, 3.1.2.2.9.C.1, and 3.1.2.2.9.C.4. However, the staff's position on the use of VT-3 to detect cracking will continue to be documented in the "Detection of Aging Effects" program element in GALL Report AMP XI.M16A, which states, in part, the following:

"...VT 3 visual methods may be applied for the detection of cracking in non-redundant RVI components only when the flaw tolerance of the component, as evaluated for reduced fracture toughness properties, is known and the component has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions." 5 I-5 Appendix A - Section 2, Acceptance Criteria Ite m 3.1.2.2.9.A.9 (Identification of TLAAs for PWR-Design RVI Components) on Page A-20 and A-21 stipulates that, "in order to satisfy the requirements of the ASME Code,Section III, Subsections NG-2160 and NG-3121, license renewal applicants demonstrating acceptability of RVI components with design-basis cumulative usage factor (CUF) analyses that are TLAAs should include the effects of the reactor The staff agrees with the comment, in part, that the evaluation of environmental effects for PWR RVI core support structures should not be incorporated in SRP-LR Section 3.1.2.2.9.A.9 in final LR-ISG-2011-04. However, the staff does not agree with the commenter's statement that the evaluation of time-limited aging analyses for the reactor internals should be addressed in accordance with the existing 10 CFR Part 54 requirements without the need to include environmental effects.

C-5 # Source ID Summary of Comment Response coolant system water environment in the fatigue CUF analyses." The Commenter provided its justification for removal of this last sentence.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. Final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of the staff's resolution of Source ID I-1, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.9.

To the extent that the commenter does not agree with the need to address evaluation of environmental effects, the staff's SE, Revision 1 for MRP-227-A documents the basis for limitations and conditions being placed on the use of MRP-227 as well as A/LAIs that shall be addressed by applicants/licensees who choose to implement the NRC-approved version of MRP-227. Specifically, the topic of environmentally-assisted fatigue for PWR RVIs is addressed in A/LAI No. 8, Item 5 of MRP-227-A. Thus, the intent of LR-ISG-2011-04 is not to supplement or modify the evaluation in the staff's SE, Revision 1.

6 I-6 The component-specific AMR items described in Appendix-A, Sections 4, 5 and 6 are based on migration from NUREG-1801. As a result the listing is more complex than the approved MRP-227-A tables. For example, there are approximately 25 items in Section 5 that classify as "Primary" component examinations, whereas the equivalent component list in MRP-227-A contains only 13 items. The component content is very similar but the breakdown is complex. A key advantage of aligning license renewal commitments to the MRP-227-A format is to facilitate important, industry-wide program updates based on Operating Experience through the NEI 03-08 process. The alignment between MRP-227-A and NUREG-1801 is compromised by embedding item detail in the ISG format. It is recommended that NUREG-1801 refer existing AMR items to "the applicable MRP-227-A table" and retain detail only for those items which may be beyond the scope of MRP-227-A. This will significantly reduce applicant and NRC staff burden, and improve integration of evolutionary changes through the NEI 03-08 process. The staff does not agree with the comment recommending that NUREG-1801 refer existing AMR line items to "the applicable MRP-227-A table" and retain detail only for those items which may be beyond the scope of MRP-227-A. In accordance with 10 CFR 54.21(a)(3) for each structure and component identified as part of the integrated plant assessment (IPA), the LRA is to demonstrate that the effects of aging will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. The IPA is independent of the line items in MRP-227-A and the GALL Report and may also result in additional components beyond the generic lists in these documents. This requires that the LRA provide a complete listing of AMR line items, which may include items consistent with MRP-227-A and the GALL Report and may also result in additional components beyond the generic lists in these documents. Thus, the number and content of AMR line items in the inspection tables of MRP-227-A are not the only basis for determining the AMR line items in the GALL Report. Similarly, the AMR line items in the GALL Report are not the only basis for determining the aging effects requiring management for components or establishing the AMR line items that are included in an LRA.

However, final LR-ISG-2011-04 incorporates revisions to SRP-LR Table 3.1-1 and GALL Tables IV.B2, IV.B3, and IV.B4 as summarized in the C-6 # Source ID Summary of Comment Response staff's resolution to Source ID I-1.

7 I-7 ISG implementation of Applicant/Licensee Action items from the MRP-227-A SER is by way of notes to AMR items listed in Sections 4, 5 and 6. This could be addressed by reference to the appropriate SER action items. It is recommended that the required evaluations would be documented in a single location specified by the ISG rather than associated with individual items. Associating these actions with each individual AMR item increases the burden for both the applicant and NRC staff reviewer. The staff agrees with the comment that associating A/LAIs with each individual AMR line item increases the burden for both the applicant and NRC staff reviewer.

As part of the resolution to Source ID I-1, final LR-ISG-2011-04 incorporates revisions to SRP-LR Table 3.1-1 and GALL Tables IV.B2, IV.B3, and IV.B4. Specifically, GALL Tables IV.B2, IV.B3, and IV.B4 were revised to be consistent with the format of AMR items in the GALL Report for non-RVI components and the footnotes in the "Further Evaluation" column of these tables were deleted.

8 I-8 The draft ISG requires Applicants to develop and submit evaluation of inaccessible Reactor Vessel Internal components in accordance with Note 3 to Sections 4 and 5, and Note 2 to Section 6. With the exception of A/LAI #6 of the MRP-227-A SER, these evaluations have been addressed during review and approval of the Industry program. The requirement to develop, submit and review the inspection basis is unnecessary. It is recommended that this note be eliminated. The staff agrees with the comment that it is not necessary to provide an evaluation of inaccessible RVI components, with the exception of A/LAI No. 6 of MRP-227-A. As part of the resolution to Source ID I-7, final LR-ISG-2011-04 incorporates revisions to delete the further evaluation footnotes from GALL Tables IV.B2, IV.B3, and IV.B4.

As a result of staff's reso lution to Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not redundantly addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04.

9 I-9 MRP-227-A provides applicants with an alternative to the defined inspection requirements when plant-specific analyses of accumulated fatigue usage are performed. Applicants may choose to either inspect in accordance with the approved MRP-227-A schedules, or perform analyses. In cases where Applicants perform analyses to relax MRP-227-A requirements, those analyses would be submitted for NRC staff approval in accordance with A/LAI 8. The ISG is unclear regarding these alternatives. For example item IV.B3.RP-343 appears to require physical examinations to support acceptance of the TLAA. The industry recommends that the ISG refer to MRP-227-A and the associated A/LAI requirement discussions. The staff agrees with the comment that LR-ISG-2011-04 refer to MRP-227-A and the associated A/LAI discussions for alternatives or deviations to the inspection and evaluation guidelines in MRP-227-A.

It is the responsibility of the license renewal applicant to demonstrate in accordance with 10 CFR 54.21(a)(3) that it can adequately manage aging of RVIs for the period of extended operation, whether through the use of MRP-227-A or alternatives. If a TLAA exists for a RVI, in accordance with 10 CFR 54.21(c)(1)(iii), an applicant may choose to demonstrate the effects of aging on the intended function of the component will be adequately managed for the period of extended operation. It is incumbent on the license renewal applicant to provide this demonstration of aging management, which can include the use of MRP-227-A or an appropriate alternative.

In order to avoid redundancy, areas reso lved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed again in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA.

C-7 # Source ID Summary of Comment Response 10 I-10 Item 9.B.1 of the ISG notes that Section 3.2.5.3 of the NRC SE (Revision 1) on MRP-227 Revision 0 recommends that the applicant consider replacement or inspection activities with regard to the Control Rod Guide Tube (CRGT) split pins if the pins are currently fabricated with Alloy X-750 or Type 316 stainless steel material. A review of the referenced section of the SE does not reach the conclusion that this specificity of action is required; the SE requirement is to evaluate the adequacy of the plant-specific existing program to ensure that the aging degradation is adequately managed during the extended period of operation. The SE direction is on evaluation of the performance of the existing program and does not suggest that it should be changed to include inspections. Therefore the industry considers the specificity of direction provided in the SE to be sufficient and the ISG should not provide alternate direction.

The staff agrees with the comment that there is an inconsistency between SRP-LR Section 3.1.2.2.

9.B.1 in draft LR-ISG-2011-04 and Section 3.2.5.3 of the staff's SE, Revision 1, for MRP-22

7. As part of the staff's resolution to Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of the staff's resolution of Source ID I-1, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.B.1.

Specific to the Westinghouse CRGT split pins, A/LAI No. 3 recommends an evaluation to consider the need to replace the Alloy X-750 split pins, if applicable, or an inspection of the replacement type 316 stainless steel split pins to ensure that cracking has been mitigated and that aging degradation is adequately monitored during the extended period of operation. Thus, the intent of LR-ISG-2011-04 is not to supplement or modify the evaluation in the staff's SE, Revision 1, but rather, to recommend that the response to A/LAI No. 3 of MRP-227-A be appropriately documented in Appendix C of the LRA.

11 I-11 Section C.3, page A23 of LR-ISG 2011-04 states that per MRP-227-A, "...EVT-1 inspections of certain CE-design components would be necessary if the design basis fatigue TLAAs for the components could not demonstrate that fatigue-induced cracking would be adequately managed..." This statement does not accurately represent MRP-227-A Table 4-2, because it assumes that the fatigue evaluations required by the MRP-227-A table item already exist and are part of the current licensing basis, and therefore are formally classifiable as TLAAs. In fact, many, if not all, of the older CE design reactor internals were not qualified to the fatigue rules of ASME III, so TLAAs as defined in 10 CFR Part 54 do not exist. Further, page A24 of the draft ISG states "Otherwise, CE-design applicants for renewal are requested to credit the MRP's EVT-1 basis in MRP-227-A as the applicable aging management basis if either: (1) the CLB does not include applicable CUF or It fatigue analyses for these components;-" This statement appears to compel the applicant who does not have a current licensing basis TLAA to perform EVT-1 inspections. MRP-227-A clearly does not require inspections based solely on the lack of a current licensing basis TLAA. In fact, it only requires that a fatigue evaluation be The staff agrees with the comment that the discussion related to CE-designed lower core flange welds, core support plates, and fuel alignment plates in SRP-LR Section 3.1.2.2.9.C.3 in draft LR-ISG-2011-04 is not clear.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.C.3.

10 CFR 54.21(a)(1) requires that license renewal application contain an IPA that must, for those systems, structures, and components within the scope of Part 54, identify and list those structures and components subject to an aging management review (AMR). The components evaluated in MRP-227-A may not fully encompass the components identified in an IPA, as required by 10 CFR 54.21(a)(1), and therefore, should not be considered a substitute for performance of an IPA.

The aging effects requiring management for RVIs are not governed by C-8 # Source ID Summary of Comment Response performed to determine if a fatigue issue might exist; and if so, where would inspection be focused to manage it. The method of the fatigue evaluation was intended to be the usual engineering practice, for example by comparison of the number expected operating transient cycles to those specified by design, or by stress analysis if required. MRP-227-A; rather the content of MRP-227-A serves to assist a PWR license renewal applicant. In accordance with 10 CFR 54.21(a)(3), the effects of aging are to be managed for all applicable aging effects for a particular component, which may be broader than the aging effects identified in MRP-227-A and the GALL Report for RVIs. It is the responsibility of the license renewal applicant to demonstrate that it can adequately manage aging of RVIs for the period of extended operation, whether through the use of MRP-227-A or alternatives.

Therefore, if the CE-designed lower core flange welds, core support plates, and fuel alignment plates are subject to an AMR and fatigue is an applicable aging effect, regardless if there is a TLAA, the LRA must demonstrate that fatigue will be adequately managed in accordance with 10 CFR 54.21(a)(3).

12 I-12 For A/LAI No. 2, when comparing the licensee renewal AMR from BAW-2248A to the tables in MRP-189, the locking devices for the vent valve were identified as a possible "Primary" component. The original vent valves located next to outlet nozzles failed due to flow induced vibration, and those valves next to the nozzles were replaced with locking devices made containing Alloy 600.

It is recommended that Table IV Reactor Vessel, Internals, and Reactor Coolant System, B4 Reactor Vessel Internals (PWR) - Babcock and Wilcox on page A-124 of LR-ISG 2011-04 be revised to include a line item addressing Alloy 600 replacement vent valve locking devices, which are subject to aging degradation due to primary water stress corrosion cracking (PWSCC).

The staff agrees with the comment to include an AMR line item for cracking of B&W vent valve locking devices made from Alloy 600 materials in GALL Table IV.B4 of draft LR-ISG-2011-04. Final LR-ISG-2011-04 incorporates the core support shield vent valve top and bottom retaining rings to be managed for cracking in GALL AMR Item IV.B4.RP-252a. 13 I-13 In Item 8 on page A-11 of the LR-ISG, the second sentence appears to be incomplete with respect to the statement pertaining to "-confirming that the quality of inspections, flaw evaluations, and corrective actions performed under this program." It is recommended that the revised statements be reviewed for completeness.

The staff agrees with the comment that the sentence in the "Confirmation Process" program element in GALL Report AMP XI.M16A of draft LR-ISG-2011-04 is incomplete. Final LR-ISG-2011-04 completes this sentence in the "Confirmation Process" program element.

14 I-14 Item 3 on page A-16 of the LR-ISG should reference NRC SE Section 3.2.5.1 and not Section 3.5.1. It is recommended that this reference be revised. The staff agrees with the comment that SRP-LR Section 3.1.2.2.9.A.3 in draft LR-ISG 2011-04 should reference NRC SE Section 3.2.5.1 and not Section 3.5.1.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a C-9 # Source ID Summary of Comment Response result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.

2.2.9.A.3.

15 I-15 Item D.1 on page A-25 discusses evaluation "Acceptance Criteria" recommendations applicable to Babcock and Wilcox reactor internals. In general, A/LAI 4 is not specific relative to the wording for the manner in which the items were stress relieved, and it was stated that a "stress relief process" was used. In Item D.1, the wording used in some cases implies a "post-weld heat treatment" process. The words "stress relief process" should be used consistently without the implication of a heat treatment process only. In addition, the requirements in Item D.1 appear to go beyond the requirements of the A/LAI as it was written and approved by the MRP-227-A SER.

The staff agrees with the comment to use the terminology "stress relief process" consistently throughout SRP-LR Section 3.1.2.

2.9.D.1 of draft LR-ISG-2011-04. Final LR-ISG-2011-04 does not use the term "post-weld heat treatment" and this term is replaced with the term "stress relief process." In addition, as a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.D.1.

16 II-1 Page A IV.B2.RP-300 Alignment and interfacing components, such as hold down springs, are addressed in MRP-227-A. Based on MRP-227-A, the intent of the GALL was only to apply to hold down springs made from Type 304 Stainless Steel (SS). The possibility of thermal embrittlement of hold down springs made from Type 403 martensitic SS is not addressed. The issue is, however, discussed in the proposed SRP section 3.1.2.2.9.A.6 and in applicant action item 7 of the SER (Revision1).

Proposed Change: Include the words "applicable to hold down springs fabricated from Type 304 SS" and add a line item to address thermal embrittlement for hold down springs fabricated for Type 403 stainless steel.

The staff agrees with the comment that the possibility of thermal embrittlement for Type 403 martensitic stainless steel hold down springs is not addressed. Final LR-ISG-2011-04 does not use the term "(Aust. SS Material)" in the "Material" column in GALL AMR Item IV.B2.RP-300. Furthermore, the use of the term "Stainless Steel" in GALL AMR Item IV.B2.RP-300 is generic and includes all grades of "stainless steel" as defined in GALL Table IX.C, "Selected Definitions & Use of Terms for Describing and Standardizing - MATERIALS." With these revisions hold down springs made from Type 403 martensitic SS are addressed in GALL AMR Item IV.B2.RP-300. 17 II-2 Page A Section 3.1.2.2.9.A.3, second paragraph There is little guidance on Applicant Action Item #2 related to additional RVI piece parts and what was used during the development of MRP 191. Utilities are left to draw a conclusion that unless the utility implemented a modification beyond the vendor's recommendation, all of the piece parts in the reactor vessel were considered during the development of MRP-189, 191 and 227-A.

Proposed Change: Add verbiage to provide additional guidance to allow utilities to make the assumption that unless a utility implemented modifications beyond that recommended by the The staff does not agree with the comment, in particular the inference that, unless a utility implemented modifications beyond that recommended by the vendor of the RVI, all of the piece parts of the RVI were considered during the development of MRP-189, 191 and 227-A. The methodology and results of a topical report, such as MRP-227-A, cannot be assumed to be generically bounding for every plant.

The IPA described in the response to Source ID I-11 is a plant-specific evaluation performed by a license renewal applicant. Thus, the components evaluated in MRP-189, 191 and 227-A may not fully encompass the components identified in an applicant's IPA and therefore, should not be considered a substitute for performance of an IPA. The C-10 # Source ID Summary of Comment Response vendor of the RVI, then all of the piece parts of the RVI were considered during the development of MRP-189, 191 and 227-A. aging effects requiring management for RVIs may be broader than the aging effects identified in MRP-189, 191 or 227-A. It is the responsibility of the license renewal applicant to demonstrate, in accordance with 10 CFR 54.21(a)(3), that it can adequately manage aging of RVIs for the period of extended operation, whether through the use of MRP-227-A or alternatives. The content in MRP-189, 191 or 227-A only serves to assist a PWR license renewal applicant.

However, as addressed in the staff's resolution to Source ID I-1, in order to avoid redundancy, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.3.

18 II-3 Page A Section 3.1.2.2.9.A.3 The words in the third paragraph are confusing and it is not clear what is meant by plant specific AMR line items or why Note E would be appropriate. For those applicants whose plant-specific review results in identification of additional components for inspection or different component inspection categories from those identified in MRP-227-A, the applicant is requested to identify the changes in the component inspection categories as either plant-specific AMR line items or NEI Note E consistent with GALL AMR items (whichever is applicable) in their Table 2 AMR line items for their PWR RVI components.

Proposed Change: It is suggested that if only a component line item or two that is not in GALL is being added then an exception can be taken to the program and justification be added that includes inspection specifics such as method and acceptance criteria such that the whole program doesn't have to be evaluated as a plant specific program.

The staff agrees with the comment that portions of SRP Section 3.1.2.2.9.A.3 are confusing.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.3.

19 II-4 Page A last paragraph Page A Section 3.1.2.2.9.1.2 The document does not provide clear direction as to what goes into an inspection plan.

The staff agrees that draft LR-ISG-2011-04 does not provide clear direction as to what goes into an inspection plan but does not agree with the commenter's proposed change. The staff does not agree with the Commenter's general claim with respect to what satisfies an inspection plan per A/LAI No. 8, as additional guidance is outlined in the SE, Revision 1, for MRP-227, and fulfillment of that action item will depend on each C-11 # Source ID Summary of Comment Response Proposed Change: Add verbiage to allow utilities to better determine what the inspection plan should consist of (e.g., A Westinghouse design plant should provide unit specific information in the Inspection Plan consistent with tables 4-3, 4-6 and 4-9 of MRP-227-A and the A/LAIs).

applicant's plant-specific review.

See the staff's resolution to Source ID I-3. RIS 2011-07 provides the staff's expectations for Category D plants (PWR plant licensees that have not submitted their LRAs but plan to submit an LRA in the future) to submit, for NRC staff review and approval, an AMP for vessel internals that is consistent with MRP-227-A. As an "inspection plan" is one aspect of satisfying A/LAI No. 8 of the staff's SE, Revision 1, for MRP-227. An "inspection plan" provides information about the RVI components to be inspected and a description of how they will be managed for age-related degradation (e.g., examination method, frequency, acceptance criteria, coverage, etc.). The staff expects that the details of an "inspection plan" for Category D plants will be incorporated into the LRA submittal as part of the 10-element AMP and AMR line items. Thus, consistent with RIS 2011-07, the staff does not expect Category D plants to provide a separate document that contains an "inspection plan" in response to A/LAI No. 8.

To avoid confusion, final LR-ISG-2011-04 avoids explicit reference to an "inspection plan" in the body of the AMP, and "inspection plan" is only referenced as part of A/LAI No. 8 in the staff's SE, Revision 1, for MRP-227. 20 II-5 Page A Table 3.1-1 Item 27a It is not clear that this line item is only applicable to hold down springs fabricated from Type 304 SS.

Proposed Change: Add Type 304 SS hold down springs.

The staff agrees with the comment that Table 3.1-1, Item 27a, of draft LR-ISG-2011-04 does not clearly address Type 304 stainless steel hold down springs.

Final LR-ISG-2011-04 does not include this item, but Westinghouse Type 304 stainless steel hold down springs were incorporated into Table 3.1-1, Item 59a, in final LR-ISG-2011-04, which uses the generic terminology "stainless steel."

21 II-6 Page A Table 3.1-1 Item 3 Under 'Further Evaluation Recommended' column, it is not clear what "It" stands for?

Proposed Change: Provide an explanation.

"It" refers to the parameter being calculated for the cyclical loading analyses. In later editions of the ASME Code Section III, these analyses were referred to as cumulative usage factor (CUF) analyses. Thus, the "I t" parameter is analogous to the CUF parameter required for Class 1 components designed to more recent editions of the ASME Code,Section III. The subscripted "t" was removed in the formatting duri ng the issuance of draft LR-ISG-2011-04 for public comment.

As a result of the staff's resolution of Source ID I-1, SRP-LR Table 3.1-1 Item 3 does not incorporate the reference to the "I t" parameter in final LR-ISG-2011-04.

C-12 # Source ID Summary of Comment Response 22 II-7 Page A IV.B2.RP-280 There is confusion regarding what comprises the lower core barrel flange weld for Westinghouse designed plants. This component is still listed in MRP-191, and 227-A for Westinghouse designed plants. MRP-227-A indicates it may be the weld between the core barrel and the lower support forging or casting.

Proposed Change: Provide an explanation regarding what this component is.

Page 3-11 of MRP-227-A states that "[t]he lower support forging is welded to and supported by the core barrel, which transmits vertical loads to the vessel through the core barrel flange." In addition, Table 5-3 of MRP-227-A provides the "Examination Acceptance Criteria and Expansion Criteria" for the "Core Barrel Assembly - Lower core barrel flange weld." Footnote 2 of Table 5-3 states that "[t]he lower core barrel flange weld may alternatively be designated as the core barrel-to-support plate weld in some Westinghouse plant designs."

No revisions were made as a result of this comment.

23 II-8 Page A Section 3.1.2.2.9.A.4 In the subject paragraph, it appears the NRC wanted an exception not an enhancement: For those component inspections that do not achieve the inspection coverage criteria stated in the NRC SE (Revision 1) on MRP-227, the applicant is requested to take a deviation from the MRP-defined inspection criteria and describe the process and type of evaluation that will be implemented to evaluate the impact of the aging effects on the inaccessible regions of the components. In this case, the applicant is requested to identify this process as an applicable enhancement of the "monitoring and trending" program element of its RVI Program.

Proposed Change: Clarify what is expected.

The staff agrees with the comment that the referenced text in SRP-LR Section 3.1.2.2.9.A.4 of draft LR-ISG-2011-04 is not clear. As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.4.

24 II-9 Page A IV.B2.RP-280 It is not clear how Note 3 in the "Further Evaluation" column is applicable to this GALL Line Item.

Proposed Change: Clarify the applicability.

The staff agrees with the comment that the applicability of Note 3 to GALL AMR Item IV.B2.RP-280 is not clear. As a result of the staff's resolution of Source ID I-1, Final LR-ISG-2011-04 incorporates revisions to GALL Tables IV.B2, IV.B3, and IV.B4 as summarized above.

25 II-10 Page 3 In the last paragraph of the Discussion section only table 3-1 is listed for justification of TE for the materials. Tables 3-2 and 3-3 should be mentioned since 3-1 is only for B&W internals.

Proposed Change: Add Tables 3-2 and 3-3.

The staff agrees with the comment that Tables 3-2 and 3-3 should be referenced in the last paragraph of the "Discussion" section. However, the "Discussion" section of final LR-ISG-2011-04 no longer references Table 3-1 in MRP-227-A.

26 II-11 Page A-7 The staff agrees with the comment to change the terminology to "Aging Management Requirement" tables in the "Parameters Monitored/Inspected" C-13 # Source ID Summary of Comment Response The second paragraph in this Section refers to condition monitoring tables in MRP-227-A. There are no tables with this title in MRP-227-A Proposed Change: Change to Aging Management Requirement tables. program element. The "Parameters Monitored/Inspected" program element in final LR-ISG-2011-04 states the following:

"Specifically, the program implements the parameters monitored/inspected criteria consistent with the applicable tables in Section 4, 'Aging Management Requirement,' in MRP-227-A-"

27 II-12 Page A-9 Only Table 5-1 is listed for acceptance criteria when MRP-227-A contains three tables, 5-1 thru 5-3 Proposed Change: Change to read " Section 5 and Tables 5-1 thru 5-3 of MRP-227" The staff agrees with the comment to add references to Table 5-2 and 5-3 of MRP-227-A for the "Acceptance Criteria" program element. The "Acceptance Criteria" program element of GALL Report AMP XI.M16A in final LR-ISG-2011-04 references Table 5-1 through 5-3 of MRP-227-A.

28 II-13 Page A-10 The first paragraph on the page says "The program adopts the acceptance criteria for the physical measurement monitoring methods recommended in MRP-227-A, as qualified in Section 3.3.5 and A/LAI No. 5 in Revision 1 of the NRC SE on MRP-227". Section 3.3.5 of the MRP does not specify acceptance criteria so there is nothing to be adopted. It only requires it be developed as discussed in footnote 3.

Proposed Change: Change sentence to read "The program includes acceptance criteria for the physical measurement monitoring methods as recommended in MRP-227-A, Section 3.3.5 and A/LAI No. 5 in Revision 1 of the NRC SE on MRP-227".

The staff agrees with the comment that Section 3.3.5 of MRP-227-A does not specify acceptance criteria for physical measurements. However, as a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in GALL Report AMP XI.M16A in final LR-ISG-2011-04.

The "Acceptance Criteria" program element in final LR-ISG-2011-04 states that, in general, the AMP establishes appropriate acceptance criteria for any physical measurement monitoring methods that are credited for aging management of RVIs.

29 II-14 Page A-12 The following sentence relates to notification criteria: "The evaluation in Section 3.5 of Revision 1 of the SE on MRP-227 provides the staff's basis for endorsing the NEI 03-08 implementation process for these programs. This includes NRC's endorsement of the NEI 03-08 criteria for notifying the NRC of any deviation from the I&E methodology in MRP-227-A and justification of the deviation no later than 45 days after approval by a licensee executive."

Proposed Change: Delete this sentence as it already is discussed in element 9 where it is appropriate. The staff agrees with the comment that the sentence associated to the notification criteria already exists in the "Administrative Controls" program element and does not need to be repeated in the "Operating Experience" program element of GALL Report AMP XI.M16A. The "Operating Experience" program element of GALL Report AMP XI.M16A in final LR-ISG-2011-04 does not incorporate this sentence associated with the notification criteria.

C-14 # Source ID Summary of Comment Response 30 II-15 Page A-8 The justification required for the use of VT-3 to detect cracking over that specified in MRP-227A and approved by the staff in the SE that allows its use without the additional limitations and analyses is not needed. Proposed Change: Eliminate need for additional justification if requirements as specified in SER and MRP are met.

The staff agrees with the comment that additional justification for the use of VT-3 to detect cracking is not needed if requirements specified in the SER and MRP are met. As a result of the staff's resolution of Source ID I-4, final LR-ISG-2011-04 does not incorporate SRP-LR Sections 3.1.2.2.9.A.7, 3.1.2.2.9.C.1, and 3.1.2.2.9.C.4. However, the staff's position on the use of VT-3 for the detection of cracking will continue to be documented in the "Detection of Aging Effects" program element in GALL Report AMP XI.M16A. 31 II-16 Page A Section 3.1.2.2.9.C.1 The justification required for the use of VT-3 to detect cracking over that specified in MRP-227A and approved by the staff in the SE that allows its use without the additional limitations and analyses is not needed. Proposed Change: Eliminate need for additional justification if requirements as specified in SER and MRP are met.

The staff agrees with the comment that additional justification for the use of VT-3 to detect cracking is not needed if requirements specified in the SER and MRP are met. As a result of the staff's resolutions to Source ID I-4 and ID II-15, final LR-ISG-2011-04 does not incorporate SRP-LR Sections 3.1.2.2.9.A.7, 3.1.

2.2.9.C.1, and 3.1.2.2.9.C.4. However, the staff's position on the use of VT-3 for the detection of cracking will continue to be documented in the "Detection of Aging Effects" program element in GALL Report AMP XI.M16A.

32 II-17 Page A Section 3.1.2.2.9.C.3 The option presented as (3), as an alternative basis for accepting the design basis fatigue analyses in accordance with the TLAA acceptance requirement in 10 CFR 54.21(c)(1)(iii) does not make sense when compared to options 1 and 2 Proposed Change: Add the word "the EVT-1 is used" at the beginning The staff agrees with the comment that the discussion related to CE-designed lower core flange welds, core support plates, and fuel alignment plates in SRP-LR Section 3.1.2.2.9.C.3 of draft LR-ISG-2011-04 is not clear. As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.C.3.

Also see the staff's resolution to Source ID I-11, in which the staff clarified that, if the CE-designed lower core flange welds, core support plates, and fuel alignment plates are subject to an AMR and fatigue is an applicable aging effect, regardless if there is a TLAA, then in accordance with 10 CFR 54.21(a)(3), the LRA must demonstrate that fatigue will be adequately managed. 33 II-18 Page A Section 3.1.2.2.9.D.1 There is no need for a plant-specific enhancement of the The staff agrees with the comment that there is not a need for a plant-specific enhancement of the "Preventive Actions" program element discussed in SRP-LR Section 3.1.2.2.9.D.1 of draft LR-ISG-2011-04, which C-15 # Source ID Summary of Comment Response "preventative actions" program element for their RVI Program enhancement to be identified if an applicant confirms that the welds were appropriately stress-relieved. An enhancement doesn't seem appropriate since the action has already been taken.

Proposed Change: Eliminate the need for an enhancement is associated with A/LAIs No. 4 of MRP-227-A.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.D.1.

34 II-19 Page A Table 3.0-1 There is no need for the words "or to applicable NRC further evaluation "acceptance criteria" recommendations in Section 3.1.2.2 of the SRP-LR (i.e., the latest NRC issued version of NUREG-1800)." Specific acceptance criteria do not need to be part of a SAR description. If it is an enhancement it will already be a commitment.

Proposed Change: Delete The staff agrees with the comment that the further evaluation acceptance criteria do not need to be specified as part of a Safety Analysis Report description. Final LR-ISG-2011-04 does not incorporate this second paragraph in the "Description of Program" column for GALL Report AMP XI.M16A in SRP-LR Table 3.0-1. However, 10 CFR 54.21(d) provides the requirements for a Final Safety Analysis Report supplement and states, in part, that it must contain a summary description of the programs and activities for managing the effects of aging. The specificity of such descriptions will depend on the program proposed by each license renewal applicant.

35 II-20 Page A Table IV.B2 There is no need for specifying the Examination technique in the Program column.

Proposed Change: Delete The staff agrees with the comment that there is no need for specifying the Examination Technique in the "Aging Management Program" column of GALL Table IV.B2. GALL Tables IV.B2, IV.B3 and IV.B4 in final LR-ISG-2011-04 do not incorporate a summary of the examination techniques from the "Aging Management Program" column.

36 II-21 A Footnotes For note 6, see comments 15 and 16 above on why no justification for using VT-3 exam is required when it was acceptable in SER for 227. This applies to CE and B&W tables that also contain a similar note. Proposed Change: Delete the note The staff agrees with the comment that additional justification for the use of VT-3 to detect cracking is not needed if requirements specified in the SER and MRP are met. As a result of the staff's resolutions of Source ID I-4 and ID II-15, final LR-ISG-2011-04 does not incorporate SRP-LR Sections 3.1.2.2.9.A.7, 3.1.

2.2.9.C.1, and 3.1.2.2.9.C.4. However, the staff's position on the use of VT-3 for the detection of cracking will continue to be documented in the "Detection of Aging Effects" program element in GALL Report AMP XI.M16A.

In addition, as part of the staff's resolution to Source ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-RVI components and does not incorporate the footnotes in the "Further Evaluation" column of these tables.

C-16 # Source ID Summary of Comment Response 37 II-22 Page A-102 - Footnote #1 "In conjunction" is repeated in the second sentence.

Proposed Change: Delete second in conjunction The staff agrees with the comment that "in conjunction with" was an editorial error in Note 1. However, as part of t he staff's resolution to Source ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-RVI components and does not incorporate the footnotes in the "Further Evaluation" column of these tables. As a result of these revisions, the referenced Note 1 is not incorporated in final LR-ISG-2011-04.

38 II-23 Page A-104 - Footnote #8 4th line "No.2 above, and is so" should be and if so.

Proposed Change: Correct The staff agrees with the comment that there is a typographical error in Note 8 of page A-103. However, as part of the staff's resolution to Source ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-RVI components and does not incorporate the footnotes in the "Further Evaluation" column of these tables. As a result of these revisions, the referenced Note 8 is not incorporated in final LR-ISG-2011-04.

39 II-24 Page A Table IV.B2 Water chemistry is not listed as an AMP, with the aging effect of stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC) such as in line items IV.B2.RP-270a, 345, 399, 299a. This mainly occurs in new line items and also shows up in Table IV.B3 and IV.B4 Proposed Change: Add XI.M2 as an AMP The staff agrees with the comment that GALL Report AMP X.M2, "Water Chemistry," is not listed in GALL Table IV.B2. GALL Table IV.B2, IV.B3 and IV.B4 of final LR-ISG-2011-04 include GALL Report AMP X.M2, "Water Chemistry," as a recommendation to manage cracking by SCC, PWSCC, or IASCC, or loss of material due to pitting or crevice corrosion of RVIs. 40 II-25 Page A IV.B2.RP-399 As indicated in Table 4-9 of MRP-227-A and the associated note 2, the clevis insert bolts are inspected for wear. To the extent cracking would be visible in the VT-3 inspection, it would of course be addressed; but, the intent of the inspection is to look for wear.

Proposed Change: Eliminate this line as an existing inspection program element, or change the AMP description to note the inspection is for gross effects of cracking The staff agrees with the comment that Table 4-9 of MRP-227-A did not identify cracking as an aging effect requiring management for Westinghouse-design clevis insert bolts of screws but does not agree with the commenter's proposed change.

Relevant operating experience associated with aging may exist that has not been accounted for in MRP-227-A. AMR item IV.B2.RP-399 for cracking of Westinghouse-design clevis insert bolts and screws was included in LR-ISG-2011-04 based on industry operating experience. Appendix A of MRP-227-A states, in part, that "[f]ailures of Alloy X-750 clevis insert bolts were reported by one Westinghouse-designed plant in 2010" and "[a]lthough the failed clevis insert bolts were not removed for metallurgical examination, it can be surmised that the most likely cause of failure was PWSCC." No revisions were made as a result of this comment.

41 II-26 Page A IV.B2.RP-285 The staff agrees with the comment to delete the aging mechanism of loss of fracture toughness from AMR item IV.B2.RP-285. Since the clevis bolts C-17 # Source ID Summary of Comment Response As described in MRP-191, the clevis bolts and inserts are not in a high flux region and irradiation embrittlement is not a significant aging mechanism. As indicated in Table 4-9 of MRP-227-A and the associated note 2, the clevis insert bolts are inspected for wear. Also, Note 5 is applied to the further evaluation column; however, Note 5 refers to reduction of fracture toughness due to thermal embrittlement in stainless steel components, while the material listed for this line is nickel alloy.

Proposed Change: Eliminate the aging mechanism of loss of fracture toughness from this line and remove note 5 from the further evaluation column.

and inserts are not in a high flux region, GALL AMR Item IV.B2.RP-285 in final LR-ISG-2011-04 does not incorporate the aging effect of loss of fracture toughness due to neutron irradiation embrittlement.

As a result of the staff's resolution of Source ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-RVI components and deleted the footnotes in the "Further Evaluation" column of these tables.

42 II-27 Page A IV.B2.RP-345 As indicated in Table 5-1 of MRP-191, cracking of the core barrel flange is a concern for the weld rather than the base metal. Table 4-3 specifically identifies the welds as primary components to be inspected for cracking. While inspections of the welds would identify cracking in the adjacent base metal, separately adding cracking as an aging effect to the base metal as an existing component is not consistent with MRP-227-A or existing inspections.

Proposed Change: Eliminate base metal cracking as an aging effect in this line.

The staff agrees with the comment to delete base metal cracking from GALL AMR Item IV.B2.RP-345 of draft LR-ISG-2011-04 since MRP-227-A identifies that the adjacent base metal is part of the examination coverage for the "Core Barrel Assembly - Lower core barrel flange weld."

Thus, GALL AMR Item IV.B2.RP-345 in final LR-ISG-2011-04 does not reference cracking of the core barrel flange (base metal). GALL AMR IV.B2.RP-345 continues to identify loss of material due to wear for the core barrel flange (base metal).

43 II-28 Page A-9 Flaw evaluation methods include recommendations for flaw depth sizing and for crack growth determinations as well for performing applicable load limit. It should read - "growth determinations as well as for performing."

Proposed Change: Change to include missing "as."

The staff agrees with the comment that the sentence is incomplete. This sentence in the "Monitoring and Trending" program element of GALL Report AMP XI.M16A in final LR-ISG-2011-04 is complete.

44 II-29 Page A-49 In the first paragraph under Systems, Structures, and Components thermal shield assembly should be changed to thermal shield or neutron pad assembly to address the newer Westinghouse plants. Also, the component type neutron pad is not addressed in Table B2 or MRP-227.

The staff acknowledges that the component type neutron pad assembly is not addressed in GALL Table IV.B2 or MRP-227-A. However, the intent of this LR-ISG is not to supplement such aspects that are not covered in MRP-227-A. Thus, no revisions were made as a result of this comment.

If a PWR license renewal applicant identifies during the I PA that its plant design contains a neutron pad assembly (instead of a thermal shield assembly) and is subject to an AMR, the license renewal applicant must C-18 # Source ID Summary of Comment Response Proposed Change: Address recommended change.

identify this assembly in its LRA and propose an adequate means of aging management.

45 II-30 Table IV.B2 The environment "Reactor coolant and neutron flux" is used for all line items/components in Table B2, however not all the components listed in Table B2 will experience a neutron fluence exceeding 10 17 n/cm2 (E>1MeV) at the end of the period of extended operation. The environment should be more specific based on the location (fluence) of the components.

Proposed Change: The Table should note exceptions to the neutron fluence level.

The staff does not agree with the comment to note exceptions with regard to use the term "neutron flux" in GALL AMR items in the GALL Report.

The GALL Report generically and conservatively assumes that PWR RVIs are exposed to an environment of "reactor coolant and neutron flux" regardless of the fluence level. The staff anticipates that applicants will address their plant-specific data in their IPA and identify appropriate AMR items. No revisions were made as a result of this comment.

46 II-31 Page A Table 3.1-1 Item 27 Component was changed to nickel alloy guide tube support pins, however associated Table B2 line items IV.B2.RP-355 and IV.B2.RP-356 were changed to include both nickel alloy and stainless steel.

Proposed Change: Clarify The component in SRP-LR Table 3.1-1, Item 27, which refers to control rod guide tube (CRGT) split pins (support pins), is applicable to both nickel alloy and stainless steel materials. SRP-LR Table 3.1-1 Item 27 in draft LR-ISG-2011-04 was removed and incorporated into Table 3.1-1 Item 53c in final LR-ISG-2011-04. 47 II-32 A Last paragraph Sentence "EPR MRP methodology left some..." should be changed.

Proposed Change: Should read "EPRI MRP methodology left some..."

The staff agrees with the proposed change, however, as a result of the staff's resolution of Source ID I-1 the referenced sentence is not incorporated in final LR-ISG-2011-04.

48 II-33 The following acronyms are used but not included in Appendix B of this ISG; CUF, NRC, SE, and USAR.

Proposed Change: Update Appendix B to include all acronyms.

The staff agrees with the comment; however, draft LR-ISG-2011-04 was revised to remove the full list of acronyms in LR-ISG-2011-04, Appendix B. Final LR-ISG-2011-04, Appendix B, was revised to document the mark-up of changes to the GALL Report and SRP-LR. Acronyms in final LR-ISG-2011-04 are defined the first time they are used.

49 II-34 The page numbers for Appendix B are A-165 and A-166, the last page of Appendix A is A-144.

Proposed Change: Verify correct pagination.

The staff agrees with the comment and final LR-ISG-2011-04 includes the correct page numbers.

C-19 # Source ID Summary of Comment Response 50 II-35 Page A Section 3.1.2.2.9.A.5 For re-inspection greater than 10 years, further evaluation is redundant and inconsistent with standard GALL AMR and AMP formatting and presentation. Inspection frequencies would be evaluated in AMP element 4 for consistency with MRP-227-A chapter 4 primary, expansion, and existing components inspection tables. If the inspection frequency is identified that is not consistent with MRP-227-A Chapter 4 tables, an exception must be identified and justified.

Proposed Change: Delete further evaluation 3.1.2.2.

9.A item 5. Item to be addressed by AMP element 4.

The staff agrees with the comment that if an inspection frequency is not consistent with MRP-227-A, an exception must be identified and justified.

Furthermore, Section 4.0 of the staff's SE, Revision 1, for MRP-227 provides the "Conditions And Limitations And Applicant/Licensee Plant-Specific Action Items," which specifically states that the re-examination frequency for "Primary" inspection category components shall be on a maximum 10-year interval, unless a plant-specific analysis providing justification for an extended examination frequency is submitted to and approved by the NRC.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.5.

51 II-36 Page A Section 3.1.2.2.9.A.7 For VT-3 Inspection, further evaluation is redundant and inconsistent with standard GALL AMR and AMP formatting and presentation. VT-3 inspection requirements should be addressed as part of AMP element 3 for consistency with MRP-227-A requirements. Potential enhancements noted by the ISG further evaluation would be addressed by an AMP enhancement.

Proposed Change: Delete further evaluation 3.1.2.2.

9.A item 7. Item to be addressed by AMP element 3.

The staff agrees with the comment that VT-3 inspection requirements should be addressed as part of GALL Report AMP XI.M16A.

As a result of the staff's resolution of Source ID I-4, final LR-ISG-2011-04 does not incorporate SRP-LR Sections 3.1.2.2.9.A.7, 3.1.2.2.9.

C.1, and 3.1.2.2.9.C.4 related to VT-3 inspections. In addition, the staff's position on the use of VT-3 for the detection of cracking will continue to be documented in the "Detection of Aging Effects" program element in GALL Report AMP XI.M16A.

52 II-37 Page A Section 3.1.2.2.9.B.2 For Westinghouse Hold Down Springs, further evaluation is redundant and inconsistent with standard GALL AMR and AMP formatting and presentation. Definition of physical measurement techniques for Westinghouse hold down springs should be addressed as part of AMP element 3. Acceptance criteria for the hold down spring inspections would be addressed by AMP element

6. Proposed Change: Delete further evaluation 3.1.2.2.

9.B item 2. Item to be addressed by AMP elements 3 and 6. The staff agrees with the comment that physical measurement techniques and the inspection acceptance criteria for Westinghouse hold down springs are to be defined in an AMP.

The staff's SE, Revision 1, for MRP-227 documents the basis for limitations and conditions placed on the use of MRP-227 as well as licensee/applicant action items that shall be addressed by applicants/licensees who choose to implement the NRC-approved version of MRP-227. Specifically, A/LAI No. 5 of MRP-227-A addresses physical measurements of Westinghouse hold down springs.

As a result of the staff's resolution of Source ID I-1, areas resolved in the C-20 # Source ID Summary of Comment Response staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.B.2.

53 II-38 Page A IV.B2.RP-297 For CASS CRGT Lower Flanges, the ISG revision to the stainless steel definition in GALL Section IX.C requires that CASS be specifically designated in an AMR line item when thermal and neutron embrittlement susceptibility are identified. MRP-227-A Table 3-3 identifies the material of construction for CRGT lower flanges as CF-8 and thermal and neutron embrittlement identified as considerations for primary component classification.

Proposed Change: Identify CASS as an additional material in GALL IB.B2.RP-297 The staff agrees with the comment to add cast austenitic stainless steel (CASS) as a material in GALL AMR Item IV.B2.RP-297. In final LR-ISG-2011-04 the "Material" colu mn of GALL AMR Item IV.B2.RP-297 states "stainless steel, including CASS" and the "Aging Effect/Mechanism" column states "Loss of preload due to neutron irradiation embrittlement, and for CASS due to thermal aging embrittlement."

54 II-39 Page A IV.B2.RP-268 It appears that the primary purpose for the Inaccessible Locations AMR line item is to provide a further evaluation of inaccessible locations in partially accessible components susceptible to cracking due to SCC and IASCC using further evaluation note 3 (SRP-LR Section 3.1.2.2.9A Part A). This further evaluation is redundant to the note 3 further evaluation required by other AMR lines.

Proposed Change: Delete IV.B2.RP-268 The staff agrees with the comment to delete IV.B2.RP-268. As a result of the staff's resolution of Source ID I-7 and ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-RVI components. In addition, the footnotes in the "Further Evaluation" column of these tables are not incorporated into final LR-ISG-201 1-04. GALL AMR Items IV.B2.RP-268, IV.B3.RP-309 and IV.B4.RP-238 for Westinghouse, Combustion Engineering and Babcock and Wilcox designed plants, respectively, are not incorporated in final LR-ISG-2011-04. 55 II-40 Page A IV.B2.RP-269 It appears that the primary purpose for the Inaccessible Locations AMR line item is to provide a further evaluation of inaccessible locations in partially accessible components susceptible to Loss of fracture toughness due to neutron and irradiation embrittlement using further evaluation note 3 (SRP-LR Section 3.1.2.2.9A Part A). This further evaluation is redundant to the note 3 further evaluation required by other AMR lines Proposed Change: Delete IV.B2.RP-269 The staff agrees with the comment to delete IV.B2.RP-269. As a result of the staff's resolution of Source ID I-7 and ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-RVI components. In addition, the footnotes in the "Further Evaluation" column of these tables are not incorporated into final LR-ISG-2011-04. As a result, GALL AMR Items IV.B2.RP-269, IV.B3.RP-311 and IV.B4.RP-239 for Westinghouse, Combustion Engineering and Babcock and Wilcox designed plants, respectively, are not incorporated into final LR-ISG-2011-04.

C-21 # Source ID Summary of Comment Response 56 II-41 Page A IV.B2.RP-265 No additional measures (Cracking due to SCC and IASCC) in Section 3.3.1 of MRP-227-A defines the no additional measures category as: those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria. Additional components were placed in the No Additional Measures group as a result of FMECA and the functionality assessment. No further action is required by the MRP-227-A for managing the aging of the No Additional Measures components. Simply put, there are no aging effects requiring aging management.

Proposed Change: Change the aging effect column and AMP column for IV.B2.RP-265 to be consistent with other GALL AMR "none-none" line items and move the lines to GALL Section IV.E, Common Miscellaneous Material Environment Combinations.

The staff does not agree with the comment to change GALL AMR Item IV.B2.RP-265 to be consistent with other GALL AMR "none-none" line items and the statement that there are no aging effects requirement management.

The "No Additional Measures" category of components in MRP-227-A does not equate to such components not having an aging effect requiring management; it only indicates that MRP-227-A does not include guidance to manage aging for components categorized as "No Additional Measures." Thus, the staff agrees with the commenter's following statement that "[n]o further action is required by MRP-227-A for managing the aging of the No Additional Measures components." The IPA is independent of MRP-227-A and may identify applicable aging effects to manage, which may be broader than the aging effects identified in MRP-227-A for RVIs. Thus, the "No Additional Measures" category of components in MRP-227-A does not alleviate the requirements in 10 CFR 54.21(a)(3).

In any event, the staff acknowledges that GALL AMR Items IV.B2.RP-265, IV.B2.RP-267, IV.B3.RP-306, IV.B3.RP-307, IV.B4.RP-236 and IV.B4.RP-237 caused confusion; thus, final LR-ISG-2011-04 does not incorporate GALL AMR Items IV.B3.RP-307, IV.B4.RP-236 and IV.B4.RP-237. In addition, GALL AMR Items IV.B2.RP-265, IV.B2.RP-267 and IV.B3.RP-306 in final LR-ISG-2011-04 clarify that there is no additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists.

57 II-42 Page A IV.B2.RP-267 No additional measures (Loss of fracture toughness due to neutron and irradiation embrittlement) in Section 3.3.1 of MRP-227-A defines the no additional measures category as: those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria. Additional components were placed in the No Additional Measures group as a result of FMECA and the functionality assessment. No further action is required by the MRP-227-A for managing the aging of the No Additional Measures components. Simply put, there are no aging effects requiring aging management.

Proposed Change: Change the aging effect column and AMP column for IV.B2.RP-267 to be consistent with other GALL AMR "none-none" line items and move the lines to GALL Section IV.E, The staff does not agree with the comment to change GALL AMR Item IV.B2.RP-267 to be consistent with other GALL AMR "none-none" line items and that there are no aging effects requirement management.

As a result of the staff's resolution for Source ID II-41, final LR-ISG-2011-04 does not incorporate GALL AMR Items IV.B3.RP-307, IV.B4.RP-236 and IV.B4.RP-237. In addition, GALL AMR Items IV.B2.RP-265, IV.B2.RP-267 and IV.B3.RP-306 in final LR-ISG-2011-04 clarify that there is no additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists.

C-22 # Source ID Summary of Comment Response Common Miscellaneous Material Environment Combinations.

58 II-43 Page A-6 Clarification is needed relative to the relationship between the SRP-LR and the GALL documents.

The staff noted draft LR-ISG-2011-04 caused confusion between the relationship of the SRP-LR and the GALL Report for PWR RVI components. As a result, final LR-ISG-2011-04 does not reference the SRP-LR in GALL Report AMP XI.M16A in order to be consistent with the format of other AMPs in the GALL Report. 59 II-44 Page A-11 Wording awkward Proposed Change: Delete "that" at the beginning of line 8.

The staff agrees with the comment to delete the word "that" from the "Confirmation Process" program element of GALL Report AMP XI.M16A in draft LR-ISG-2011-04. The staff revised this program element to state, in part, "- for confirming the quality of inspections, flaw evaluations, and corrective actions performed under this program." 60 II-45 Page A-17 There is a concern that "monitoring and trending" program elements and "corrective action" program elements are buried in the Acceptance Criteria section.

The staff agrees with the comment that there is a concern the "monitoring and trending" and "corrective actions" program elements are buried in the Acceptance Criteria section.

As a result of the staff's resolution of Source ID I-1 and II-8, the staff revised LR-ISG-2011-04 so that areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.4.

61 II-46 Page A-20 and A-21 The statement "To satisfy the requirements of ASME Code Section III-." is confusing if not all plants are committed to Subsection NG.

Proposed Change: The statement should be modified to include a qualifying statement like "if the plant is committed to Subsection NG." The staff does not agree with the comment to alter the referenced statement, as it comes from the staff's SE, Revision 1, on MRP-227. The topic of environmentally-assisted fatigue for PWR RVIs is addressed in A/LAI No. 8, Item 5 of MRP-227-A. Section 3.0 of the staff's SE, Revision 1, on MRP-227 documents the basis for limitations and conditions being placed on the use of MRP-227 as well as licensee/applicant action items that shall be addressed by applicants/licensees who choose to implement the NRC-approved version of MRP-227. Revisions to the conditions and limitations, applicant/licensee plant-specific action items, and conclusions of the staff's SE, Revision 1, for MRP-227 are not within the scope of LR-ISG-2011-04.

However, as a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.9.

C-23 # Source ID Summary of Comment Response 62 II-47 Page A Section 3.1.2.2.9.D.1 The intended meaning of the word "appropriately" in D.1, second paragraph. Is not clear.

Proposed Change: Clarify meaning The staff agrees that the referenced sentence in SRP-LR Section 3.1.2.2.9.D.1 is not clear. However, as a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.D.1.

63 II-48 Page A-142 Spell out variable name in 1.0 of "Further Evaluation Recommendations" The staff noted that as a result of the resolution to Source ID I-7, final LR-ISG-2011-04 does not incorporate the referenced variable names and further evaluation footnotes in GALL Tables IV.B2, IV.B3, and IV.B4.

64 II-49 Page A-5 Each of the following documents provide information for submittal of an AMP and inspection plan:

• GALL Revision 2 (page XI.M16A-2) and LR-ISG-2011-04 (draft, page A-5)

• RIS 2011-07 (page 6)

• Safety Evaluation Revision 1 for MRP-227 (page 34)

• Section 3.5.1 of the Safety Evaluation (page 25)

It is unclear what actually goes into the LRA and the format. The above verbiage implies that the AMP and inspection plan are separate documents that are submitted with the application but are reviewed and approved by the NRC as unique documents. A quick search of the GALL indicates that PWR Vessel Internals is the only program that requires the AMP and an inspection plan to be submitted for NRC review and approval.

Proposed Change: Commenter provided revisions to Section 3.5.1 of Safety Evaluation Revision 1 for MRP-227.

The staff disagrees with the comment because revisions to the conditions and limitations, applicant/licensee plant-specific action items and conclusions of the staff's SE, Revision 1, for MRP-227 are not within the scope of LR-ISG-2011-04.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. In addition, see the staff's resolution of Source ID I-3, in which the staff discusses the staff's position/guidance regarding inspection plans that is documented in RIS 2011-07 dated July 21, 2011. The staff expects that the details of an "inspection plan" for Category D plants (defined in RIS 2011-07) will be incorporated into the LRA submittal as part of the 10-element AMP and AMR line items. Thus, consistent with RIS 2011-07, the staff does not expect Category D plants to provide a separate document that contains an "inspection plan" in response to A/LAI No. 8. 65 II-50 Page A-6 GALL Rev 2 (page XI M16A-3) states: The responses to the LR A/LAIs on MRP-227 are provided in Appendix C of the LRA.

LR-ISG-2011-04 (page A-6) deleted this requirement, however LR-ISG-2011-04 (page A-14, 15) states to provide responses to the A/LAIs in Appendix C of the LRA, and to address SRP-LR further evaluation "acceptance criteria" that are based on these A/LAIs. It is The staff agrees with the comment that it is unclear where the A/LAIs should be addressed. As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.1 and also does not incorporate a discussion of A/LAIs in GALL AMP XI.M16A.

C-24 # Source ID Summary of Comment Response unclear where the licensee action items should be addressed. Wording implies that the applicant/licensee action items should be addressed in Appendix C and in the associated further evaluation section. Proposed Change: The Commenter provided revisions to LR-ISG-2011-04 (page A-6).

66 II-51 XI.M16A, PWR Vessel Internals elements 1. Scope of Program, 5. Monitoring and Trending, and 6. Acceptance Criteria refer to the latest NRC approved version of WCAP-17096-NP and the associated applicant/licensee action items. It is our understanding that WCAP-17096-NP has been submitted for approval however it has not been approved at this time. A program cannot be developed based on an unapproved document or unknown A/LAIs.

Proposed Change: Remove any reference to WCAP-17096-NP or delay issuance of LR-ISG-2011-04 until WCAP-17096-NP is approved by the NRC.

The staff agrees with the comment to delete any reference to WCAP-17096-NP since the report has been submitted for review but not approved by staff. Final LR-ISG-2011-04 does not reference WCAP-17096-NP. 67 II-52 Many of the A/LAIs specified in the Acceptance Criteria section of LR-ISG-2011-04 request that the applicant make enhancements or augmented enhancements to various program elements as a result of the responses to the "further evaluations." It would be simpler if the NRC specified an acceptable method of addressing an issue in the XI.M16A program elements and then if the licensee/applicant wanted to do something different take an exception rather than requiring each licensee/applicant to develop a unique set of enhancements for their program.

Proposed Change: Revise A/LAIs to clearly state that additional justification/information is only required to be included in the "further evaluation" responses if the applicant/licensee is deviating from the requirements of MRP-227-A.

The staff disagrees with the comment to revise A/LAIs in LR-ISG-2011-04, as it is a direct reference to the staff's SE, Revision 1, for MRP-227. See the staff's resolution to Source ID II-49, in which the staff explains that revisions to the conditions and limitations, applicant/licensee plant-specific action items and conclusions of the staff's SE, Revision 1, for MRP-227 are not within the scope of LR-ISG-2011-04. No revisions were made as a result of this comment.

68 II-53 A simplified method of addressing reactor internals in the GALL tables B.2, B.3, and B.4 would be to have line items based on component classification (Primary, Expansion, Existing, and No Additional Measures) as defined in MRP-227-A rather than individual component types (Alignment and Interfacing components: internals hold down spring, Alignment and interfacing components: upper core plate alignment pins, etc). Making this change would allow multiple line items to apply to several component types and The staff agrees with the comment, in part, that GALL AMR Tables IV.B2, IV.B3 and IV.B4 can be simplified. However, the staff does not agree with the commenter's proposed change.

As explained in the staff's resolution of Source ID I-6, the AMR line items in the GALL Report and MRP-227-A do not solely serve as the basis for determining components or aging effects that require management or establish the AMR line items to be included in an LRA. The IPA required C-25 # Source ID Summary of Comment Response reduce the number of Table 2 line items simplifying this section.

Proposed Change: Revise NUREG-1801 tables B.2, B.3, and B.4 to have line items associated with component classification (Primary, Expansion, Existing Program, and No Additional Measures) and refer to MRP-227-A for the specific components in the four classification groups.

by 10 CFR 54.21(a) is independent of the AMR line items provided in MRP-227-A and the GALL Report. It is not necessary for the staff to correlate the number and contents of AMR items in GALL Tables IV.B2, IV.B3, and IV.B4 exactly to the number and contents of inspection items in MRP-227-A.

In any event, as a result of the staff's resolution of Source ID I-1, GALL AMR Tables IV.B2, IV.B3 and IV.B4 were revised. See resolution of Source ID I-1 for a summary of the revisions.

69 II-54 Several of the applicant/licensee action items (A/LAI) identified in the Safety Evaluation for MRP-227 (pages 32 - 34) required plant-specified evaluations or analyses to be submitted as part of the application. A/LAI Number 5 requires the applicant/licensee include its proposed acceptance criteria and an explanation of how the proposed acceptance criteria are consistent with the plants' licensing basis and the need to maintain the functionality of the component being inspected under all licensing basis conditions of operation during the period of extended operation as part of their submittal to apply the approved version of MRP-227. A/LAI Number 7 requires a plant-specific analysis to be performed on Westinghouse lower support column bodies made of CASS be included as part of their submittal to apply the approved version of MRP-227.

Proposed Change: Revise A/LAI Numbers 5 and 7 to allow for the applicants/licensees to commit performing an analysis prior to the period of extended operation. This would allow applicants/licensees that are submitting in the near future (2013 timeframe) to perform the analyses on normal schedule rather than an expedited schedule.

The staff disagrees with the comment to revise A/LAIs in LR-ISG-2011-04, as it is a direct reference to the staff's SE on MRP-227-A. See the staff's resolution to Source ID II-49, in which the staff explains that revisions to the conditions and limitations, applicant/licensee plant-specific action items and conclusions of the staff's SE, Revision 1, for MRP-227 are not within the scope of LR-ISG-2011-04. No revisions were made as a result of this comment. 70 II-55 Page A-17 In the last paragraph it states: For those component inspections that do not achieve the inspection coverage criteria stated in the NRC SE (Revision 1) on MRP-227, the applicant is requested to identify a deviation from the MRP-defined inspection criteria and describe the process and type of evaluation that will be implemented to evaluate the impact of the aging effects on the integrity of those components in the population that were inaccessible to the inspection technique, and to identify this process as an applicable enhancement of the "monitoring and trending" program element of its RVI Program. The staff agrees with the comment that the referenced statement in SRP-LR Section 3.1.2.

2.9.A.4 of draft LR-ISG-2011-04 is more appropriately addressed in the AMP.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.4.

C-26 # Source ID Summary of Comment Response The SRP is specifying actions for applicants to perform as part of an aging management program which is more appropriately addressed within program requirements.

Proposed Change: These actions should be included as program elements, not in the further evaluation sections of the SRP.

71 II-56 Page A-19 and A Section 3.1.2.2.9.A.9 In the third paragraph the further evaluation states: "To satisfy the requirements of the ASME Code,Section III, Subsection NG-2160 and NG-3121, the existing fatigue CUF analysis shall include the effects of the reactor coolant water environment."

The December 26, 1999, Generic Safety Issue (GSI) 190 close-out memorandum from Ashok C. Thadani, Director of the Office of Regulatory Research, to William D. Travers, Executive Director for Operations, provides the basis for consideration of the effects of the reactor coolant water environment. It should be noted that GSI-190 concerns are limited to pipe leakage, which is not applicable to RVI components since they do not form a portion of the reactor coolant pressure boundary and are therefore not subject to leakage.

Proposed Change: Delete the referenced sentence in the third paragraph of Further Evaluation A.9.

The staff does not agree with the comment, in particular the inference that, the effects of the reactor coolant water environment on metal fatigue are not applicable to RVI components since they do not form a portion of the reactor coolant pressure boundary.

See the staff's resolution to Source ID II-46, in which the staff explains that environmentally-assisted fatigue for PWR RVIs is addressed specifically in A/LAI No. 8, Item 5 of the staff's SE, Revision 1, on MRP-227. Final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.9 72 II The comments submitted by EPRI-MRP, the PWROG-MSC, and NEI are extensive and involve complex issues. EPRI and PWROF, along with NEI, respectfully request a follow-up meeting with the NRC staff to discuss resolution of the comments and, if appropriate, an additional comment period.

The NRC staff acknowledges the complexity of the issues captured in MRP-227-A. However, LR-ISG-2011-04 is not intended to supplement, modify or further resolve the issues raised in MRP-227-A, but rather to reference MRP-227-A and the associated staff's SE, Revision 1, for MRP-227, in the usable format of a generic aging management program, as described in the GALL Report. To the extent that comments provided suggestions to clarify or simplify the format of LR-ISG-2011-04 for ease of use, the staff was able to incorporate those changes into the final document. However, to the extent that comments proposed changes to the actual content of the staff's SE, Revision 1, for MRP-227, the staff did not incorporate those comments, as it is beyond the scope and intent of LR-ISG-2011-04. The staff also does not perceive further benefits from an additional public meeting and comment period to resolve the latter set of comments, as it is beyond the scope of LR-ISG-2011-04.

FINAL LICENSE RENEWAL INTERIM STAFF GUIDANCE LR-ISG-2011-04 UPDATED AGING MANAGEMENT CRITERIA FOR REACTOR VESSEL INTERNAL CO MPONENTS FOR PRESSURIZED WATER REACTORS INTRODUCTION This license renewal interim staff guidance (LR-ISG) updates the U.S. Nuclear Regulatory Commission (NRC's) guidance in NUREG-1800, Revision 2, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants," (SRP-LR) and NUREG-1801, Revision 2, "Generic Aging Lessons Learned Report" (GALL Report). This LR-ISG is primarily based on the issuance of Revision 1 to the Final Safety Evaluation (SE) of Electric Power Research Institute (EPRI) Report, Materials Reliability Program (MRP) Report 1016596 (MRP-227), Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," by letter dated December 16, 2011 (SE, Revision 1, on MRP-227) (ADAMS Accession No. ML11308A770). After the issuance of the staff's SE, Revision 1, on MRP-227, EPRI Technical Report No. 1022863 (MRP-227-A) was published in January 2012. MRP-227-A

is the NRC-endorsed version of MRP-227, which incorporates the NRC staff's SE, Revision 1, on MRP-227. Specifically, this LR-ISG revises the recommendations in the GALL Report and the NRC staff's acceptance criteria and review procedures in the SRP-LR to ensure consistency with MRP-227-A. This LR-ISG also provides a framework to ensure that PWR license renewal applicants will adequately address age-related degradation and aging management of reactor vessel internal (RVI) components during the term of the renewed license.

DISCUSSION Current Regulatory Framework

Pursuant to Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants," Section 21(a)(3), of Title 10 of the Code of Federal Regulations (10 CFR 54.21(a)(3)), a license renewal applicant is required to perform an integrated plant assessment (IPA) that demonstrates that the effects of aging on structures and components subject to an aging management review (AMR) are adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis (CLB) for the period of extended operation.

The NRC's guidance in SRP-LR Section 3.0.1 defines the AMR as the identification of the structure and component materials, environments, aging effects, and aging management programs (AMPs) credited for managing the aging effects. In turn, SRP-LR Section A.1.2.3 defines an acceptable AMP as consisting of 10 elements. In addition, 10 CFR 54.21(d) requires the license renewal application (LRA) to contain a final safety analysis report (FSAR) supplement that includes a summary description of the programs and activities for managing the effects of aging and the evaluation of time-limited aging analyses (TLAAs) for the period of extended operation.

GALL Report AMP X I.M 16A, "PWR Vessel Inter nals," provides recomme ndations for an AMP to manage the effects of ag ing for PWR RVI compo nents. In ad dition, the GALL Report provides component-specific AMR items for PWR RVI c o mponents in the following tables:

• Table IV.B2 for Westing house-desig ned RVI components

• Table IV.B3 for CE-designed RVI components

• Table IV.B4 for B&W-designed RVI components SRP-LR Ta ble 3.1-1 pr ovides the specific commodity grou p-based AMR items for PWR RVI components. SRP-LR Sections 3.1.

2.2.1, 3.1.2.

2.3, 3.1.2.2.

9, 3.1.2.2.10 , 3.1.2.2.12, 3.1.2.2.13, and 3.1.2.2.

14 provide the aging management review results for which f u rther evaluation is recommend ed by the GALL Report for PWR RVI components. Finally, S R P-LR Tabl e 3.0-1 provides an example of t he type of informati on to be included in the FSAR Supplement for an AMP for PW R RVI comp onents. Basis for Issuing Interim Guidance On January 12, 2009, EPRI submitted Technica l Report No. 1016596, "Materials Reliability Program: Pressurized Water Reactor Internals I n spection a nd Evaluation Guidelines (MRP-227, Revision 0)," for NRC st aff review and approval. On June 22, 2011, the NRC staff issued its SE on MRP-22 7, Revision 0, which con t ained specific topical re port conditio n items (TRCIs) on the use of MRP-227 and Applicant/Licen see Action It ems (A/LAIs) that must be addressed by those applicants o r licensee s utilizing this topical repor

t. The staff i ssued a revi sion of it s SE on the report methodology (i.e., SE, Revision 1, on MRP-227) by letter dated December 16, 2011.

MRP-227-A, the NRC-en dorsed version of MRP-227, was later published in January 2012 and provides guidance for a PWR licensee or licen se renewal applicant to u s e in the de velopment and implementation of a n AMP for RVI components. MRP-2 27-A also in corporates A/LAIs that are to be ad dressed if th is report is referenced to satisfy the r equirements of 10 CFR 54.21(a)(3) for demonstrating that th e effects of aging on the RVI compo nents, within the scope o f MRP-227, will be adeq uately mana ged. The st aff recomme nds that a P W R license renewal applicant provide its responses to these A/LAIs in Appendix C of the LRA. The use of MRP-2 27-A by a PWR license renewal applicant is n o t a substitu te for performing a plant-specific IPA to identify those struct ures and co mponents subject to an aging management revi ew, in accor dance with 10 CFR 54.21(a)(1).

Regulatory Issue Summary (RIS) 20 11-07, "Lice n se Renewal Submittal Information For Pressurized Water Reactor Internals Aging Management," dated July 21, 2011, was issued, in part, to facilitate a predictable and co nsistent met hod for reviewing the aging management of RVI components for commercial PWR LRAs. An "inspection p l a n" is one aspect of the A/LAIs of the staff's SE, Revision 1, fo r MRP-227.

This "inspe ction plan" p r ovides information about the RVI components to be inspe c ted and a d e scription of how they wi ll be managed for age-related degradation.

Details of a n "inspectio n plan" for t hose PWR plant licen se es that have not submitted but plan to su bmit an LRA in the future will be incor porated into the LRA as part of the 10-element aging management program and aging management revie w line items. Thus, consiste nt with RIS 2011-07, these fu ture license renewal applicants n eed not submit a separate document that contains an "inspect i on plan" in r e sponse to t he A/LAIs of the staff's SE, Revision 1, for MRP-227. Prior to the completion of its review and issuan ce of the SE on MRP-22 7, the staff issued SRP-LR, Re vision 2, an d GALL Rep o rt, Revision 2, in Dece mber 2010. Since SRP-LR, Revision 2, and GALL Report, Revision 2, were based on MRP-227, Revi sion 0, the r e levant portions of t he SRP-LR, Revision 2, and GALL Report, Revision 2, are no w being updated with this LR-ISG to reconcile any differences with MRP-227-A.

ACTION This LR-ISG updates the GALL Repo rt, Revision 2, and SRP-LR, Revision 2, to ensur e consiste ncy with MRP-2 27-A for the aging management of age-related d egradation f o r PWR RVI components during the t e rm of a renewed operating licen se. Appendix A, "Revisions to the GALL Report and SRP-L R ," to this L R-ISG shows these cha nges. The majority of these changes result in the in corporation o f MRP-227-A within the SRP-LR, Re vision 2, an d the GALL Report, Revi sion 2. To b e tter show these chang es, a mark-up is shown in Appendix B, "Mark-Up of Changes to the GALL Report and SRP-LR," to this LR-ISG.

On March 2 0 , 2012, at Volume 77, page 16270, of the Fede ral Register (77 FR 16270), the NRC reques ted public comments on draft LR-ISG-2011-04. Subsequently, as noticed on April 19, 2012, at Volume 77, page 23513, of the Federal Register (77 FR 23513), the NRC issued an e d itorial corre ction to the original notice to specifically identify the ADAMS Accession Nos. for additional documents asso cia t ed with draft LR-ISG-20 11-04. The staff received comments on draft LR-ISG-20 11-04 by letters from EPRI and the Pressurized Water Reactor Owners Group Mate rials Subco mmittee (ADAMS Accession No. ML12146A267) and from the Nuclear Energy Institute (ADAMS Accession No. ML12144A147). The staff considered all comments, and its ev aluation of t hese comments is cont ained in App endix C, "Staff Response to Public Comment s on Draft License Rene wal Interim Staff Guidance 2011-04," of this LR-ISG. The guidance descr ibed in this f i nal LR-ISG supersedes the affected sections of the SRP-LR, Revision 2, and the GALL Report, Revision 2, and is appro v ed for use by the NRC staff and sta k eholders. NEWLY IDE N TIFIED SY STEMS, ST RUCT URES , AND C O MPON ENTS UNDE R 10 CFR 54.37(b) Any structures and components ide n tified in th is LR-ISG as requiring aging management, which were not previously identified in earlier versi ons of the SRP-LR, Revision 2, or GALL Report, Revision 2, are consider ed by the staff to be ne wly-identified structures and components under 10 CFR 54.37(b).

BAC KFITTING A ND IS SUE FIN A LITY This LR-ISG contains gu idance on o ne acceptab le approach for managing the associated aging effects durin g the PEO f o r components within th e scope of license rene wal. The st aff's discussion o n compliance with the requirements of the Backfit Rule, 10 CFR 50.109 is presented below. Compliance with the Backfit Rule and Issue Fina lity Issuance of this LR-ISG does not co nstitute backfitting a s d e fined in 10 CFR 50.109(a)(1), and the NRC staff did not prepare a backfit analysis for issuing this LR-ISG. T here are several rationales fo r this con c lu sion, depen ding on the status of the nuclear po wer plant licensee.

Licensees currently in th e license re newal process

- The ba ckfitting pro v isions in 10 CFR 50.109 do not prote c t an applicant, as backfitting policy consideratio ns are not a pplicable t o an applicant. T herefore, issuance of this LR-ISG do es not const i tute backf itting as define d in 10 CFR 50.109(a)(1). There current ly are no combined lice n ses (i.e., 1 0 CFR Part

52) license renewal applicants; ther efore, the ch anges and n e w positions presented in the LR-ISG ma y be made without considera t ion of the issue finality provisions in 10 CFR Part 52, "Licenses, Certification s , and Approvals for Nuclear Power Plants."

Licensees who already h o ld a renewed license

- This guidan ce is nonb in ding and the LR-ISG would not require curren t holders of r enewed lice n ses to take any action (i.e., progra mmatic or plant hardware changes for managing the associated aging e ffects for co mponents within the scope of this LR-ISG).

Current holders of renewed license s should treat this guidan ce as operating experience an d take actio n s as appro p riate to ensure that applicable agin g managemen t programs are, and will remain, effective.

If, in the future, the NRC decides to take additional a c tion and im pose require ments for managing the associat ed aging effect s for components within the scope of this LR-ISG, the n the NRC would follow the requirements of the Backfit Rule. Current operating licen se or com b in ed license h o lders who have not yet applied for r enewed license s - T he backfitt i n g provisions in 10 CFR 50.109 do n o t protect an y future applicant, as backfitting p o licy consid erations are not applicab le to a future applicant.

Therefore, issuance of this LR-ISG does not co nstitute backfitting a s d e fined in 10 CFR 50.109(a)(1). The issue fina lity provisions o f 10 CFR Pa rt 52 do not extend to the aging management ma tters covered by 10 CFR Part 54, as evidenced by the requirement in 10 CFR 52.107, "Application f o r Renewal,"

stating that application s for renewal of a combined license must be in accordance with 10 CFR Part

54. APPE NDIC E S Appendix A provides the staff's revisions to the S R P-LR, Re vi sion 2, and t he GALL Re port, Revision 2, for managing aging in P W R RVI co mponents and include s the following section s:

• Section 1 -

Revised version of the GALL Report

• Section 2 -

Revised version of the S R P-LR Appendix B provides a mark-up of the SRP-LR, Revision 2, and GALL Report, Revision 2, to better show the changes made as a result of LR-I SG-2011-04 and include s the following sections:

• Section 1 -

Mark-up of changes to th e GALL Rep o rt

• Section 2 -

Mark-up of changes to th e SRP-LR Appendix C provides the staff's ba se s for resolv ing comment s that were received on the draft LR-ISG-201 1-04. REFER E NC ES 1. U.S. Code o f Federal Regulations , "D omestic Licensing of P r oduction an d Utilization Facilit ies," P a rt 50, Chapter I, Title 10 , "Energy."

2. U.S. Code o f Federal Regulations , "L icenses, Certification s , a nd Approvals for Nuclea r Power Plants," Part 52, Chapter I, Title 10, "Energy."

3. U.S. Code o f Federal Regulations , "R equiremen t s for Renewal of Operating Licen s es for Nuclear Power Plants," Part 54, Chapter I, Title 10, "Energy.

" 4. U.S. Nuclear Regulatory Commissio n , "Generic Aging Lesso ns Learned (GALL) Rep o rt," NUREG-18 01, Revision 2, December 2010, ADAMS Accession No. ML103490041.

5. U.S. Nuclear Regulatory Commissio n , "Standard Review Pla n for Review of License Renewal Ap plications for Nuclear Power Plants," NUREG-18 00, Revision 2, December 2010, ADAMS Accession No. ML1 03490036.

6. U.S. Nuclear Regulatory Commissio n , Final Saf e ty Evaluati on of EPRI Report, Mat e rials Reliability Program Repo rt 1016596 (MRP-227), Revision 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, June 22, 20 11, ADAMS Accession No. ML111600498.

7. U.S. Nuclear Regulatory Commissio n , Revision 1 to the Fina l Safety Eval uation of Electric Power Research Institute (E PRI) Report, Materials R eliabili ty Pro g ram (MRP) Report 1016596 (MRP-2 27), Revisio n 0, Pressur i zed Water Reactor Internals Inspe c tion and Evaluation Guidelin es , December 16, 2011, ADAMS Ac cession No.

ML11308A7 70. 8. Electric Power Research Institute, E P RI Technical Report No. 1016596, Materials Reliability Program: Pre ssurized Wa ter Reactor Internals Inspection and Evaluation Guidelines (M RP-227 Revision 0), December 2008, ADAMS Accession No. ML090160204 (Cover le tter from EPRI MR P) an d ADAMS Accession No. ML090160206 (Final Report).

9. Electric Power Research Institute, E P RI Technical Report No. 1022863, Materials Reliability Program: Pre ssurized Wa ter Reactor Internals Inspection and Evaluation Guidelines (M RP-227-A), December 2011, ADAMS Accession No. ML1 2017A193 (Transmittal letter from the EPRI-MRP) and ADAMS Accession Nos. ML12017A194, ML12017A1 96, ML12017A197, ML1 2017A191, ML12017A1 92, ML12017A195 and ML12017A1 99 (Final Report).

10. U.S. Nuclear Regulatory Commissio n , "Updated Aging Management Crit eria for Reactor Vessel Inter nal Compon ents of Pressurized Wat e r Reactors," Federal Register, Vol. 7 7 , No. 54, March 20, 2012, pp. 16270-16271.

11. U.S. Nuclear Regulatory Commissio n , "Updated Aging Management Crit eria for Reactor Vessel Inter nal Compon ents of Pressurized Wat e r Reactors," Federal Register, Vol. 7 7 , No. 76, April 19, 2012, p

p. 23513. 12. T. Wells and E. Fernandez, Electri c Power Rese arch Institut e Materials Reliability Program an d the Pressurized Wate r Reactor Owners Grou p Materials Subcommitt ee, letter to Document Control Desk, U.S. Nuclear Regulatory Commission, May 21, 201 2, ADAMS Accession No.

ML12146A2 67. 13. M. Richter, Nuclear Energy Institute, letter to Cin d y K. Blade y, U.S. Nucl ear Regulatory Commission , May 21, 20 12, ADAMS Accession No. ML12144A147.

14. U.S. Nuclear Regulatory Commissio n , Nuclear Regulatory Commission Regulatory Issue Summary 2 011-07, Lice nse Renewal Sub m ittal Inform ation For Pressurized Water Reactor Internals Aging Manage m e n t , July 21, 2 011, ADAMS Accession No. ML1119 90086. 15. U.S. Nuclear Regulatory Commissio

n. 2008. Memorandum from Dale E. Klein, Chairman, t o Hubert T. Bell, Office of the Inspe c tor General, "Response to Recommen dation 8 of 9/6/07 Audit Report on NRC's License Renewal Program."

(April 1, 200 8). ADAMS Accession No. ML080870286.

A-1 Appendi x A REVISI O N S TO THE G A LL REPO RT AND SRP-LR A-2 Appendix A, Section 1 - Revised version of the GALL Re port (1) Revised ver s ion of GALL Report AMP XI.

M 16 A XI.M16A PWR VESSEL INTERNALS Program Description This progra m relies on implementati on of the Electric Power Research I n stitute (EPRI) Technical Report No. 1022863, "Materials Reliability Program:

Pressurized Wat e r Reactor (PWR) Internals Inspection a nd Evaluation Guidelines," (MRP-227-A) and EPRI Technical Report No. 1016609, "Materials Reliability Program: Inspection St anda rd for PWR Internals," (M RP-228) t o manage the aging eff e cts on the pressurized water reactor (PWR) rea c tor vessel internal (RVI) components. The reco mmended a c tivities in M R P-227-A a nd additiona l plant-spe cific activitie s not defined in MRP-227-A are implemented in accordance with Nuclear Energy Institute (NEI) 03-08, "Guideline for the Managemen t of Materials Issue s." T he staff appr oved the augmented inspection and evaluation (I&E) criteria fo r PWR RVI compon ents in NRC Safety E v aluation (SE), Re vision 1, on MRP-227 by letter dated December 16, 2011.

This progra m is used to manage the effects of a ge-related d egradation mechanisms that are applicable in general to t he PWR RVI components at the facility. These aging effects include: (a) cracking, including st ress corros i on crackin g (SCC), primary water stress corros i o n cracking (PWSCC), irradiation-assisted stress corrosion cr acking (IAS CC), and cracking due t o fatigue/cyclic loading; (b) loss of mat e rial induce d by wear; (c) loss o f fra c ture toughn ess due to either thermal aging or n eutron irradiation embri ttlement; (d) changes in dimensions d ue to void swelling or d i stortion; an d (e) loss of preload due to thermal and irradiatio n-enhanced stress relaxation or creep.

The program applies th e guidance in MRP-227-A for inspect i ng, evaluating, and, if a pplicable, disposit ionin g non-confo r ming RVI c o mponents at the facility. These examinations provide reasonable assurance t hat the effects of age-rel a ted degradation mechanisms will b e managed during the p e riod of extended operation. The program includes expanding periodic examination s and other inspection s, if the extent of the degra dation identified exceeds the expected levels.

MRP-227-A guidance for selectin g RVI components for inclu s ion in the inspection sample is based on a f our-step ran k ing proce s s. Through this proce s s, the RVIs for all three P W R designs wer e assigned t o one of the following fou r groups: "Primary," "Expansion," "E xisting Programs,"

and "No Additional Measures." Defini tions of ea ch group ar e provided in "Generic Aging Lesso ns Learned Report" (GALL Report), Revision 2, Chapter IX

.B. The result of this four-ste p sample se lection pro c ess is a set of "Primary" internals co mponent locations for each of the three plant d e signs that are inspecte d because t hey are exp e cted to show the le ading indica tions of the degradation effects, with another set of "Expansion" internals component locations tha t are specif ied to expand the sample should the indication s be more severe than anticipated.

The degradation effects in a third se t of internals location s ar e deemed to be adequat ely managed by "Existing Programs," such as American Society of Mechanical Engineers A-3 (ASME) Co de,Section X I , 11 Exa m ination Category B-N-3, e x a m inations of core suppor t structures.

A fourth set of internals locations are deemed to require "No Additional Measures."

Evaluation and Technical Basis

1. Scope of Program:

The scope o f the program includes all RVI components based on the plant's app licable nucle ar steam supply system design. The scope of th e program applies the methodology and guidance in MRP-227-A, which provides an augmented inspect i on and flaw evaluation methodology for assurin g the functio nal integrity of safety-related internals in commercial operating U.S. PWR nuclear power plants design ed by Babcock & Wilcox (B&W), Combustion Engineerin g (CE), and Westinghou se. The sco pe of components considered for inspection in MRP-227-A includes co re support st ructures, tho s e RVI components that serve an intended license renewal safety function pursua n t to criteria in 10 CFR 54.4(a)(1), and other RVI components whose failu re could pre v ent satisfacto ry accomplishment of any of the functions identified i n 10 CFR 54.4(a)(1)(i), (ii), or (iii). In a ddition, ASME Code,Section X I includes inspection r equirements for PWR removable core support str u ctures in T able IWB-2500-1, Exa m ination Category B-N-3, which are in additi on to any inspection s th at are implemented in accordance with MRP-2 27-A. The scope o f the program does not include con s umable items, such a s f uel assemblies, reactivity control assemblies, and n u clear instru mentation. The scope o f the program also does not include welded attachments to the internal surface of the reactor vessel becau se these components are conside r ed to be ASME Code Class 1 appurt enances to t he reactor vessel and are managed in accorda n ce with an applicant's AMP that co rresponds to GALL AMP X I.M 1, "ASME Cod e ,Section XI Inservice I n spection, S ubsection s I W B, IWC, and IWD."

2. Preventive Actions:

MRP-227-A relies on PWR water chemistry control to pre v ent or mitigate aging effects th at can be in duced by corrosive aging mechanisms (e.g., loss of material induced by general, pitting corrosion , crevice cor r osion, or str e ss corrosio n cracking o r any of its forms [SCC, PWSCC, or IASCC]). Reacto r coola n t water chemistry is mo nitored and maintained in accordance with the Water Chemistry Progra m , as describe d in GALL AMP X I.M 2, "Wat er Chemistry."

3. Parameters M onitored/Inspected:

The program manag es the follow i ng age-relat ed degradation effects and mechanisms that are ap plicable in g eneral to RVI components at the facility: (a) cracking ind u ced by SCC, PWSCC, IASCC, or fatigue/cyclic loading; (b) loss of material induced by wear; (c) loss o f fracture to ughness ind u ced by either thermal aging or neutron irra diation embrittlement; (d) changes in dimensions due to void swelling or d i stortion; and (e) loss of preload d ue to thermal and irradiation-enhan ced stress r e laxation or creep.

For the management of cracking, th e program moni tors for evidence of surface-bre a king linear discontinu i ties if a visual inspection t e chnique is used as the non-destruct i ve exa m ination (NDE) method, or for relevant flaw presentation signals if a volumetric ultrasonic testing (UT) method is used as the NDE metho d. For the managemen t of loss of material, the program monitors for gross or abn ormal surface condition s that may be indicative of loss of mat e rial occurr ing in the components. For the management of loss of pr eload, the pr ogram monit o rs for gross surface conditions t hat may be i ndicative of loosening in applicable b o lted, fasten ed, keyed, or pinned connection

s. The program does not directly monitor for loss of fracture t oughness th at is induced by thermal aging or neutron irradiation e m brittlement. Instead, th e impact of loss of

11 Refer to the GALL Re port, Chapter I, for ap plica b i lit y of various e d itio ns o f the ASME Code

,Section XI.

A-4 fracture tou ghness on component integrity is i ndirectly mana ged by: (1) using visual or volumetric examination techniques t o monitor for cracking in the components, and (2) applying applicable r educed fract u re toughness propertie s in the flaw evaluations, in case s where cracking is d e tected in th e components and is ex t ensive enough to necessitate a sup p lemental flaw growth or flaw tolerance evaluation.

The pr ogram uses physical measurements to monitor for any dime nsional ch a nges due to void swellin g or distortio

n. Specifically, the program implem ents the parameters monitored/inspecte d criteria co nsistent with the applicable tab l e s in Section 4, "Aging Management Requiremen t s," in MRP-227-A.

4. Detection of Aging Effects:

The inspect i on methods are defined an d establishe d in Section 4 of MRP-227-A.

Standards for impleme n ting the inspection methods are def ined and establishe d in MRP-228.

In all ca ses, well-established inspe c tion methods are sele ct ed. These methods include volume tric UT examination me thods for det ecting flaws in bolting an d various visual (VT-3, VT-1, and EVT-1) exa m inations fo r detecting e ffects rangin g from general conditions t o detection and sizing o f surface-br eaking disco n tinuities. Surface examinations may also be used as an alternative to visual examinations fo r detection a nd sizing of surface-breaking discontinuitie

s. Cracking ca used by SCC, IASCC, and fatigue is monitored/inspected by either VT-1 or EVT-1 examination (for internals other than bolting) or b y volu metric UT exa m ination (boltin g). VT-3 visual methods may be applied for t he detection of cracking in non-redu ndant RVI components only when the flaw toler ance of the component, as evaluated for reduce d fracture to ughness properties, is known an d the component has be en shown to be tolerant of easily det ected large flaws, even under reduced fracture t oughness conditions. V T-3 visual methods are a cceptable for the detection of cracking in redun dant RVI components (e.g., redunda nt bolts or pins used to secure a fastened RVI assembly).

In addition, VT-3 exami nations are used to monito r/inspect f o r loss of m a terial induced by wear and for general aging conditions, such as gross distortio n caused by void swelling and irradiation gr owth or by g r oss effect s of loss of pr eload cause d by thermal and irradiation-enhanced st ress relaxation and cree

p. The program adopts the guidance in MRP-227-A for definin g the "Expa n sion Criteria" that need to be applie d to the insp ection find in gs of "Primary" compone nts and for e x panding the examination s to include additional "E xpansion" components. RVI compo nent inspect i ons are performed consistent wit h the inspection frequen cy and sampling base s for "Primary" components, "Existing Programs" components, and "Expansion" components in MRP-227-A.

In some cases (as defin ed in MRP-2 27-A), phy sical measurements are used as supp lemental techniques t o manage for the gross e ffects of we ar, loss of pr eload due to stress relaxation, or for changes in dimensio ns due to void swelling o r distortion.

Inspection coverages for "Primary" a nd "Expansion" RVI components are implemente d consistent with Sections 3.3.1 and 3.3.2 of the NRC SE, Revision 1, on MRP-227. 5. M onitori ng and Trending:

The methods for monitoring, recording, e v aluating, and trending the data that result from the program's inspection s are given in Section 6 of MRP-227-A and its subsections.

Flaw evaluation methods, inclu d ing recommend ations for fla w depth sizing and for crack growt h determinations as well as for perfor m ing applicable limit loa d , linear ela s tic and elastic-p l astic fracture an alyses of relevant flaw indications, ar e defined in MRP-227-A.

The A-5 examination and re-examinations th at are implemented in accordance with MRP-2 27-A, together with the criteria specifie d in MRP-228 fo r inspectio n methodologies, inspe c tio n procedures, and inspect i on personne l, provide for timely detection, reporting, and implementat ion of corrective actions f o r the aging effects and mechanisms managed by the program.

The program applies a pplicable fra c ture toughn ess properties, inclu d ing reductions f o r thermal aging or neu tron embrittlement, in the flaw evaluations of the components in cases w here cracking is d e tected in a RVI compon ent and is e x tensive enough to warrant a supple m ental flaw growth or flaw tolerance evaluation.

For singly-represented components, the program incl udes crit eria to evaluate the agin g effects in the inaccessible por tions of the components and the resu lting impact on the inten ded function(s) o f the components. For r edundant co mponents (such as redu ndant bolts, screws, pins, keys, o r fasteners, some of which are acce ssible to inspection and some of which are not accessible t o inspect i on), the program includes criteria to evaluate the aging effects in the population o f components that are in accessible t o the applicable inspe c tion techniqu e and the resulting im pact on the intended fun c tion(s) of th e assembly containing t he components.

6. Acceptance Criteria

Section 5 of MRP-227-A, which includes Tab le 5-1 for B&W-designed RVIs, Table 5-2 for CE-designed RVIs, and Table 5-3 for Westinghouse-d e signed RVIs, provides the specific examination and flaw eval uation accept ance criteria for the "Primary" and "Expansion" RVI compo nent examin ation methods. For RVI components addressed by examination s performed in accordan ce with the ASME Code ,Section X I, the accepta n ce criteria in IWB-3500 are applica b le. For RVI components covered by other "Exis t ing Programs," the acceptance criteria are d e scribed wit h in the applicable refere nce document. As applicable, the program establishes a c ceptance crit eria for any physical measurement monitoring methods that are credited for aging management of particular R V I components.

7. Correctiv e Actions:

Any detected condition s that do not satisfy the e x amination acceptance criteria are r equired to b e disposition ed through t he plant corr ective action program, which may require repair, replacement, or analytical evaluation for contin ued service until the next inspection. The disposit ion will ensu r e that desig n basis fun c tions of the r eactor intern als components will continu e to be fulfil l ed for all li censing basis loads and e v ents. The implementat ion of the gu idance in M R P-227-A, p l us the implementation of any ASME Code requirements, provides an acceptab le level of a g ing management of safety-related components addressed in accordance with the corrective act i ons of 10 C F R Part 50, Appendix B or its equivalent, as applica b le. Other alternative correct ive actions b a ses may be us ed to disposition relevant conditions if they have been previously approved or endorsed by the NRC. Alt e rnative corrective actions not approved or endorsed by the NRC will be submitted for NRC approval prior to their implementat ion. 8. Confirmation Process:

Site quality assurance procedure s , review and approval processes, and administrative controls are im plemented in accordance with the recommendatio ns of NEI 03-08 and the requirements of 10 CFR Part 50, Appendix B, or their equivalent, as applicable.

The implementation of t he guidance in MRP-227-A, in conjun ction with NEI 03-08 and other guidance documents, reports, or methodologies r e ferenced in this AMP, provides an acceptable level of quality and an acceptable ba sis for confir ming the quality of inspe c tions, f l aw evaluations, and corrective actions.

A-6 9. Administrative Controls:

The administrative controls fo r these types of progra m s, includin g their implementing proce dures and r e view and approval processes, are implemented in accordance with the recommended i ndustry guidelines and criteria in NEI 03-08, and are under existing site 10 CFR 50 Appendix B, Quality Assurance Programs, or their equivalent, as applicable.

The evaluation in Section 3.5 of the NRC's SE, Revision 1, on MRP-22 7 provides the basis for endorsing NEI 03-08. This include s endorsement of the criteria in NEI 03-08 for notifying the NRC of any deviation from the I&E methodology in MRP-22 7-A and justifying the deviation no later than 4 5 days after it s approval by a license e executive.

10. Operati ng Experience:

The review and assessment of relevant operating experience for it s impacts on t he program, including implement ing procedures, are governed by NEI 03-08 and Appendix A of MRP-227-A. Consist ent with MRP-227-A, the reporting of inspect i o n results an d operating experience is t r eated as a "Needed" ca tegory item under the implementation of NEI 03-08. The program is informed and enhan ced when nece ssary thr ough the systematic an d ongoing review of both plant-spe cific and ind u stry operating experience, as discu ssed in App endix B of the GALL Report, which is documen ted in LR-ISG-2011-05.

References 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants , Office of the Federal Register, Nation al Archives and Records Administration, 2011.

10 CFR Part 50.55a, Codes and Sta ndards , Office of the Fe deral Regist er, National Archives and Records Administration, 2011.

ASME Boiler & Pressure Vessel Code,Section V, Nondestructive Exam ination , 2004 Edition, American Society of Mechanical En gineers, Ne w York, NY.

ASME Boiler & Pressure Vessel Code,Section XI, Rules for I n service I n spection of N u clear Power Plant Co m ponent s , The ASME Boiler and Pressure Vessel Code, 2004 edition as approved in 10 CFR 50.55a, The American Society of Mechanical Engin eers, New York, NY. EPRI Technical Report No. 1016596, Materials Reliability Program

Pressurized Wa ter Reactor Internals Inspection and Evaluation Guidelines (M RP-227, Revision 0), Electric Power Research In stitute, Palo Alto, CA: 2008.

EPRI Technical Report No.1022863, Materials Reliability Program

Pressurized Wa ter Reactor Internals Inspection and Evaluation Guidelines (M RP-227-A), December 2011, ADAMS Accession No. ML12017A193 (Transmittal letter f r om the EPRI-MRP) an d ADAMS Accession Nos. ML12017A194, ML 12017A196, ML12017A197, ML12017A191, ML 12017A192, ML12017A1 95 and ML12017A199, (Final Report).

EPRI 10166 09, Materials Reliability Progra m: In spection Sta ndard for PWR Internals (MRP-228), Electric Power Re search Inst it ute, Palo Alto, CA, July 2009 (Non-publicly available ADAMS Accession No.

ML092120574). The non-proprietary version of the report may be accesse d by members of the public a t ADAMS Ac cession No.

ML092750569.

A-7 NRC Interim Staff Guidance LR-ISG-2011-05, Ongoing Revie w Of Operat ing Experien c e , March 16, 2012, (ADAMS Accession No. ML12044A215).

Nuclear Energy Institute (NEI) Report No. 03-08, Revision 2, Guideline fo r the Manage m ent of Materials Issues , ADAMS Accession No. ML101 050334). NRC Safet y Evaluation from Robert A. Nelson (NRC) to Nei l Wilmshurst (EPRI), Re vision 1 to the Final Sa fety Evaluation of Electric Power Re search Inst it ute (EPRI)

Report, Mat e rials Reliability Progra m (MRP) Report 1016596 (MRP-227), Revision 0, Pressurized Wat e r Reactor Internals Inspe c tion and Eva l uation Guid elines , Dece mber 16, 2011, ADAMS Accession No. ML11308A770.

A-8 (2) Revised ver s ion of GALL Report Chapter IV.B2 B2. REACT O R VE SSE L INTERNALS (PWR) -

WESTINGHOUSE Sy stems, Structures, and Components This section addresses t he Westingh ouse pressu rized-water reactor (PWR) vessel internals, which consist of components in the upper intern als assembly, the control rod guide t ube assembly, the core barr e l assembly, the baffle/fo rmer assembly, the lower internals a ssembly, lower support assembly, thermal shield assembly, bottom mounted instr u mentation system, and alignment and interfacin g components.

Common mi scellaneou s material/environment combinations where aging effects are not expected to degrade the ability of th e structure o r component to perform its intended f unction for the period of extended operation are included in IV.E. Sy stem Interfaces The systems that interfa c e with the r eactor vessel internals include the r eactor pressure vessel (IV.A2).

A-9 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-300 IV.B2-33 (R-108) Alignment and interfacing components: internals hold down spring Stainless steel Reactor coolant and neutron flux Loss of preload due to thermal and irradiation enhanced stress relaxation; changes in dimensions due to void swelling or distortion; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-301 IV.B2-40 (R-112) Alignment and interfacing components: upper core plate alignment pins Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B2.RP-299 IV.B2-34 (R-115) Alignment and interfacing components: upper core plate alignment pins Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-271 IV.B2-10 (R-125) Baffle-to-former assembly: baffle-to-former bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-10 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-272 IV.B2-6 (R-128) Baffle-to-former assembly: baffle-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-270

IV.B2-1 (R-124) Baffle-to-former assembly: baffle and former plates Stainless steel Reactor coolant and neutron flux Changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-270a IV.B2-1 (R-124) Baffle-to-former assembly: baffle and former plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B2.RP-275 IV.B2-6 (R-128) Baffle-to-former assembly: baffle-edge bolts (all plants with baffle-edge bolts) Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-11 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-354 Baffle-to-former assembly: baffle-edge bolts (all plants with baffle-edge bolts) Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-273 IV.B2-10 (R-125) Baffle-to-former assembly: barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry (for SCC mechanisms only)

No IV.B2.RP-274 IV.B2-6 (R-128) Baffle-to-former assembly: barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No A-12 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-284 IV.B2-12 (R-143) Bottom mounted instrument system: flux thimble tubes Stainless steel (with or without chrome plating) Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" or Chapter XI.M37,"Flux Thimble Tube Inspection" No IV.B2.RP-293 IV.B2-24 (R-138) Bottom-mounted instrumentation system: bottom-mounted instrumentation (BMI) column bodies Stainless steel Reactor coolant and neutron flux Cracking due to fatigueChapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-292 Bottom-mounted instrumentation system: bottom-mounted instrument (BMI) column bodies Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-296

Control rod guide tube (CRGT) assemblies: CRGT guide plates (cards)

Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-298 IV.B2-28 (R-118) Control rod guide tube (CRGT) assemblies: CRGT lower flange welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-297 Control rod guide tube (CRGT) assemblies: CRGT lower flange welds Stainless steel (including CASS)Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging and neutron irradiation embrittlement and for CASS, due to thermal aging embrittlement Chapter XI.M16A, "PWR Vessel Internals" No A-13 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-355 Control rod guide tube (CRGT) assemblies: guide tube support pins (split pins)

Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-356 Control rod guide tube (CRGT) assemblies: guide tube support pins (split pins)

Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-387 Core barrel assembly: upper core barrel and lower core barrel circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-387a Core barrel assembly: upper core barrel and lower core barrel vertical (axial) welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-388 Core barrel assembly: upper core barrel and lower core barrel circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-388a Core barrel assembly: upper core barrel and lower core barrel vertical (axial) welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No A-14 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-345 Core barrel assembly: core barrel flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-278 IV.B2-8 (R-120) Core barrel assembly: core barrel outlet nozzle welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-278a Core barrel assembly: core barrel outlet nozzle welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-280 IV.B2-8 (R-120) Core barrel assembly: lower core barrel flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-276 IV.B2-8 (R-120) Core barrel assembly: upper core barrel flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking and irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B2.RP-285 IV.B2-14 (R-137) Lower internals assembly: clevis insert bolts or screws Nickel alloy Reactor coolant and neutron flux Loss of material due to wear; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No A-15 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-399 Lower internals assembly: clevis insert bolts or screws Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to primary water stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-289 IV.B2-20 (R-130) Lower internals assembly: lower core plate and extra-long (XL) lower core plate Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-288 IV.B2-18 (R-132) Lower internals assembly: lower core plate and extra-long (XL) lower core plate Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-291 IV.B2-24 (R-138) Lower support assembly: lower support column bodies (cast)

Cast austenitic stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B2.RP-290 IV.B2-21 (R-140) Lower support assembly: lower support column bodies (cast)

Cast austenitic stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging and neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No

A-16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-291a Lower support assembly: lower support forging or casting Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-290a Lower support assembly: lower support forging or casting Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement (and thermal aging embrittlement for CASS, PH SS, and martensitic SS)

Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-294 IV.B2-24 (R-138) Lower support assembly: lower support column bodies (non-cast)

Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B2.RP-295 Lower support assembly: lower support column bodies (non-cast)

Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-286 IV.B2-16 (R-133) Lower support assembly: lower support column bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-17 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-287 IV.B2-17 (R-135) Lower support assembly: lower support column bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-303 IV.B2-31 (R-53) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Cumulative fatigue damage due to fatigueFatigue is a time-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes - TLAA IV.B2.RP-24 IV.B2-32 (RP-24) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to pitting and crevice corrosion Chapter XI.M2, "Water Chemistry" No A-18 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-382 IV.B2-26 (R-142) Reactor vessel internals: ASME Section XI, Examination Category B-N-3 core support structure components (not already identified as "Existing Programs" components in MRP-227-A)

Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue, stress corrosion cracking, or irradiation-assisted stress corrosion cracking; loss of material due to wear Chapter XI.M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" or Chapter XI.M16A, "PWR Vessel Internals," by invoking applicable 10 CFR 50.55a and ASME Section XI inservice inspection requirements No IV.B2.RP-302 Thermal shield assembly: thermal shield flexures Stainless steel Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-302a Thermal shield assembly: thermal shield flexures Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-265 Reactor internal "No Additional Measures" components Stainless steel; nickel alloy Reactor coolant and neutron flux No additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-291b Upper Internals Assembly; upper core plate Stainless steel Reactor coolant and neutron flux Cracking due to fatigueChapter XI.M16A, "PWR Vessel Internals" No A-19 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-290b Upper Internals Assembly; upper core plate Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B2.RP-346 Upper Internals Assembly: upper support ring or skirt Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No A-20 (3) Revised ver s ion of GALL Report Chapter IV.B3 B3. REACT O R VE SSE L INTERNALS (PWR) -

COMBUSTI ON E N GI NEERING Sy stems, Structures, and Components This section addresses t he Combustion Enginee ring (CE) pressurized-w a ter reactor (PWR) vessel inter nals, which consist of co mponents in the upper in ternals asse mbly, the control element assembly (CEA), the core support barrel assembly, the core shro ud assembly, and the lower support structure assembly, and encore in strumentation (ICI) components.

Common mi scellaneou s material/environment combinations where aging effects are not expected to degrade the ability of th e structure o r component to perform its intended f unction for the period of extended operation are included in IV.E. Sy stem Interfaces The systems that interfa c e with the r eactor vessel internals include the r eactor pressure vessel (IV.A2).

A-21 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-312 IV.B3-2 (R-149) Control Element Assembly (CEA): instrument guide tubes in peripheral CEA assemblies Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No IV.B3.RP-313 Control Element Assembly (CEA):

remaining instrument guide tubes in CEA assemblies Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-320 IV.B3-9 (R-162) Core shroud assemblies (all plants): guide lugs; guide lug inserts and bolts Stainless steel Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-319 IV.B3-9 (R-162) Core shroud assemblies (all plants): guide lugs; guide lug inserts and bolts Stainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No A-22 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-358 Core shroud assemblies (for bolted core shroud assemblies): assembly components, including shroud plates and former plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-318

IV.B3-8 (R-163) Core shroud assemblies (for bolted core shroud assemblies): assembly components, including shroud plates and former plates Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-316

IV.B3-9 (R-162) Core shroud assemblies (for bolted core shroud assemblies): barrel-shroud bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-23 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-317

IV.B3-7 (R-165) Core shroud assemblies (for bolted core shroud assemblies): barrel-shroud bolts Stainless steel Reactor coolant and neutron flux Loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-314 IV.B3-9 (R-162) Core shroud assemblies (for bolted core shroud assemblies): core shroud bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-315 IV.B3-7 (R-165) Core shroud assemblies (for bolted core shroud assemblies): core shroud bolts Stainless steel Reactor coolant and neutron flux Loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No

A-24 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-359 Core shroud assembly (designs assembled in two vertical sections): core shroud plate-to-former plate welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-322

Core shroud assembly (designs assembled in two vertical sections): core shroud plate-to-former plate welds Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-326

Core shroud assembly (designs assembled in two vertical sections): assembly components, including monitoring of the gap opening at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Changes in dimensions due to void swelling or distortion; loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No A-25 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-326a Core shroud assembly (designs assembled in two vertical sections): assembly components, including monitoring of the gap opening at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-323

Core shroud assembly (designs assembled in two vertical sections): remaining axial welds Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-359a Core shroud assembly (designs assembled in two vertical sections): remaining axial welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No

A-26 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-324 Core shroud assembly (designs assembled with full-height shroud plates): shroud plate axial weld seams at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-360 Core shroud assembly (designs assembled with full-height shroud plates): shroud plate axial weld seams at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-325

Core shroud assembly (designs assembled with full-height shroud plates): remaining axial welds, ribs, and rings Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry""

No

A-27 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-361 Core shroud assembly (designs assembled with full-height shroud plates): remaining axial welds, ribs, and rings Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-362 Core support barrel assembly: lower cylinder circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-362a Core support barrel assembly: lower cylinder circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-362b Core support barrel assembly: lower cylinder vertical (axial) welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No

A-28 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-362c Core support barrel assembly: lower cylinder vertical (axial) welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-329 IV.B3-15 (R-155) Core support barrel assembly: upper cylinder (base metal and welds) and upper core barrel flange (flange base metal)

Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-333 Core support barrel assembly: lower flange Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-328

IV.B3-15 (R-155) Core support barrel assembly: lower core barrel flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-332 IV.B3-17 (R-156) Core support barrel assembly: upper core barrel flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-327 IV.B3-15 (R-155) Core support barrel assembly: upper core barrel flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No A-29 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-357 Incore instruments (ICI): ICI thimble tubes - lower Zircaloy-4 Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-336 IV.B3-22 (R-170) Lower support structure (designs assembled in two vertical sections): fuel alignment pinsStainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-334 IV.B3-23 (R-167) Lower support structure (designs assembled in two vertical sections or with full-height shroud plates): fuel alignment pinsStainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-30 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-334a IV.B3-22 (R-170) Lower support structure (designs assembled in two vertical sections or with full-height shroud plates): fuel alignment pinsStainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-364 Lower support structure (all plants): core support column welds Stainless steel (including CASS) Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement and for column welds made from CASS, thermal embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-363 Lower support structure (all plants): core support column welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-330 IV.B3-23 (R-167) Lower support structure: core support column bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-31 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-331 Lower support structure: core support column bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-335 IV.B3-23 (R-167) Lower support structure (designs except those assembled with full-height shroud plates): lower core support beams Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No IV.B3.RP-365 Lower support structure (designs with a core support plate): core support plate Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-343 Lower support structure (designs with a core support plate): core support plate Stainless steel Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No

A-32 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-342 Lower support structure (designs with core shrouds assembled with full height shroud plates): deep beams Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B3.RP-366 Lower support structure (designs with core shrouds assembled with full height shroud plates): deep beams Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-339 IV.B3-24 (R-53) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Cumulative fatigue damage due to fatigue Fatigue is a time-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes, TLAA IV.B3.RP-306 Reactor internal "No Additional Measures" components Stainless steel; nickel alloy Reactor coolant and neutron flux No additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists Chapter XI.M16A, "PWR Vessel Internals" No A-33 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-24 IV.B3-25 (RP-24) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to pitting and crevice corrosion Chapter XI.M2, "Water Chemistry" No IV.B3.RP-382 IV.B3-22 (R-170) Reactor vessel internals: ASME Section XI, Examination Category B-N-3 core support structure components (not already identified as "Existing Programs" components in MRP-227-A)

Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue, stress corrosion cracking, or irradiation-assisted stress corrosion cracking; loss of material due to wear Chapter XI.M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" or Chapter XI.M16A, "PWR Vessel Internals," by invoking applicable 10 CFR 50.55a and ASME Section XI inservice inspection requirements No IV.B3.RP-338 Upper internals assembly (designs with core shrouds assembled with full height shroud plates): fuel alignment plate Stainless steel Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B3.RP-400 Core Support Barrel Assembly: thermal shield positioning pins Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No A-34 (4) Revised ver s ion of GALL Report Chapter IV.B4 B4. REA C T O R VE SSE L INTERN A L S (PWR) -

BAB COCK AND WILC OX Sy stems, Structures, and Components This section addresses t he Babcock and Wilcox (B&W) pressurized-water reactor (PWR) vessel internals, w h ich con s ist of components in the plenum cover assembly, the upper grid assembly, the control r od guide tub e (CRGT) a ssembly, the core supp ort shield a s sembly, the core barrel assembly, the lower grid assembly, incore monitoring instru mentation (IMI) guide tube assembly, and the flow distributor a ssembly.

Common mi scellaneou s material/environment combinations where aging effects are not expected to degrade the ability of th e structure o r component to perform its intended f unction for the period of extended operation are included in IV.E. Sy stem Interfaces The systems that interfa c e with the r eactor vessel internals include the r eactor pressure vessel (IV.A2).

A-35 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-242 IV.B4-4 (R-183) Control rod guide tube (CRGT) assembly: CRGT spacer castings Cast austenitic stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-242a Control rod guide tube (CRGT) assembly: CRGT spacer castings Stainless steel (including CASS) Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No IV.B4.RP-245 IV.B4-13 (R-194) Core barrel assembly (applicable to Crystal River Unit 3 or Davis Besse only):

surveillance specimen holder tube (SSHT) studs/nuts or bolts Nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-245a Core barrel assembly (applicable to Crystal River Unit 3 or Davis Besse only):

surveillance specimen holder tube (SSHT) stud or bolt locking devices Nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No

A-36 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-245b Core barrel assembly (applicable to CR-3 or DB only): surveillance specimen holder tube (SSHT) stud or bolt locking devices Nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-247 IV.B4-13 (R-194) Core barrel assembly: lower core barrel (LCB) bolts Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-247a Core barrel assembly: lower core barrel (LCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "P WR Vessel Internals" No IV.B4.RP-247b Core barrel assembly: lower core barrel (LCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "P WR Vessel Internals" No IV.B4.RP-249 IV.B4-12 (R-196) Core barrel assembly: baffle plates Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-249a Core barrel assembly: baffle plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No A-37 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-241 IV.B4-7 (R-125) Core barrel assembly: baffle-to-former bolts and screws Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, fatigue, and overload Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-241a Core barrel assembly: locking devices (including locking welds) of baffle-to-former bolts and internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, fatigue, and overload Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-240 IV.B4-1 (R-128) Core barrel assembly: baffle-to-former bolts and screws Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No A-38 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-240a Core barrel assembly: locking devices (including locking welds) of baffle-to-former bolts and internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-250 IV.B4-12 (R-196) Core barrel assembly: core barrel cylinder (including vertical and circumferential seam welds); former plates Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-250a Core barrel assembly: core barrel cylinder (including vertical and circumferential seam welds); former plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-375 Core barrel assembly:

internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking, fatigue, or overload Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No A-39 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-375a Core barrel assembly:

internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear Chapter XI.M16A, "P WR Vessel Internals" No IV.B4.RP-244 IV.B4-7 (R-125) Core barrel assembly; external baffle-to-baffle bolts and core barrel-to-former bolts; Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking, fatigue, and overload Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-244a Core barrel assembly: locking devices (including welds) of external baffle-to-baffle bolts and core barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking, or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No A-40 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-243 IV.B4-1 (R-128) Core barrel assembly: external baffle-to-baffle bolts and core barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-243a Core barrel assembly: locking devices (including welds) of external baffle-to-baffle bolts and core barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of material due to wear Chapter XI.M16A, "P WR Vessel Internals" No IV.B4.RP-248 IV.B4-12 (R-196) Core support shield (CSS) assembly: upper core barrel (UCB) bolts Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-248a Core support shield (CSS) assembly: upper core barrel (UCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No A-41 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-248b Core support shield (CSS) assembly: upper core barrel (UCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-252 IV.B4-16 (R-188) Core support shield (CSS) assembly: CSS vent valve top and bottom retaining rings (valve body components) Stainless steel, including CASS and PH steels Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-252a IV.B4-16 (R-188) Core support shield (CSS) assembly: CSS vent valve top and bottom retaining rings; vent valve locking devices (valve body components) Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-251 IV.B4-15 (R-190) Core support shield (CSS) assembly: CSS top flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of preload (wear) Chapter XI.M16A, "PWR Vessel Internals"No IV.B4.RP-251a IV.B4-15 (R-190) Plenum cover assembly: plenum cover weldment rib pads and plenum cover support flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of preload (wear) Chapter XI.M16A, "PWR Vessel Internals" No A-42 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-256 IV.B4-25 (R-210) Flow distributor assembly: flow distributor bolts Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-256a Flow distributor assembly: flow distributor bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals"No IV.B4.RP-256b Flow distributor assembly: flow distributor bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to distortion or void swelling or distortion Chapter XI.M16A, "P WR Vessel Internals" No IV.B4.RP-259 IV.B4-31 (R-205) Incore Monitoring Instrument (IMI) guide tube assembly: IMI guide tube spider-to-lower grid rib section welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-259a Incore Monitoring Instrument (IMI) guide tube assembly: IMI guide tube spider-to-lower grid rib sections welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No.

A-43 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-258 IV.B4-4 (R-183) Incore Monitoring Instrument (IMI) guide tube assembly: IMI guide tube spiders (castings)

Cast austenitic stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging and neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals"No IV.B4.RP-258a Incore Monitoring Instrumentation (IMI) guide tube assembly: IMI guide tube spiders Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-254 IV.B4-25 (R-210) Lower grid assembly: alloy X-750 lower grid shock pad bolts (Three Mile Island Unit 1, only) Nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-254a Lower grid assembly: alloy X-750 lower grid shock pad bolt locking devices (Three Mile Island Unit 1, only) Nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-254b Lower grid assembly:

alloy X-750 lower grid shock pad bolt locking devices (Three Mile Island Unit 1, only) Nickel Alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals"

No A-44 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-246 IV.B4-12 (R-196) Lower grid assembly: upper thermal shield (UTS) bolts and lower thermal shield (LTS) bolts Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-246a Lower grid assembly: upper thermal shield (UTS) bolt locking devices and lower thermal shield (LTS) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "P WR Vessel Internals" No IV.B4.RP-246b Lower grid assembly: upper thermal shield (UTS) bolt locking devices and lower thermal shield (LTS) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-260 IV.B4-31 (R-205) Lower grid fuel assembly: (a) pads; (b) pad-to-rib section welds; (c) alloy X-750 dowels, cap screws and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" No

A-45 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-260a Lower grid fuel assembly: (a) pads; (b) pad-to-rib section welds; (c) alloy X-750 dowels, cap screws and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-262 IV.B4-32 (R-203) Lower grid assembly: alloy X-750 dowel-to-lower fuel assembly support pad locking welds Nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-261 IV.B4-32 (R-203) Lower grid assembly: alloy X-750 dowel-to-guide block welds Nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.R-53 IV.B4-37 (R-53) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Cumulative fatigue damage due to fatigue Fatigue is a time-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes, TLAA IV.B4.RP-24 IV.B4-38 (RP-24) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to pitting and crevice corrosion Chapter XI.M2, "Water Chemistry"

No A-46 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-376 Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Reduction in ductility and fracture toughness due to neutron irradiation Ductility - Reduction in Fracture Toughness is a TLAA (BAW-2248A) to be evaluated for the period of extended operation. See the SRP, Section 4.7, "Other Plant-Specific TLAAs," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes, TLAA IV.B4.RP-382 IV.B4-42 (R-179) Reactor vessel internals: ASME Section XI, Examination Category B-N-3 core support structure components Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue, stress corrosion cracking, or irradiation-assisted stress corrosion cracking; loss of material due to wear Chapter XI.M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" or Chapter XI.M16A, "PWR Vessel Internals," by invoking applicable 10 CFR 50.55a and ASME Section XI inservice inspection requirements No IV.B4.RP-352 Upper grid assembly: alloy X-750 dowel-to-upper fuel assembly support pad welds (all plants except Davis-Besse) Nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No A-47 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP)Further Evaluation IV.B4.RP-236 Reactor internal "No Additional Measures" components Stainless steel; nickel alloy Reactor coolant and neutron flux No additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-400 Core support shield assembly: upper (top) flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-401 Core support shield assembly: upper (top) flange weld Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals"No A-48 (5) Revised ver s ion of GALL Report Chapter IX.C and IX.G IX.C Selected Definitions & Use of Terms for Des c ribing and Standardizing MATE RIAL S Stainless st eel Products gr ouped under the term "stainless stee l" include austenitic, f e rritic, martensitic, pre c ip itation-harde ned (PH), or duplex stainless steel (Cr content >11%). These stainless ste e ls may be fabricated using a wrought or cast process. Th ese materials are susce p tible to a variety of aging effect s and mechanisms, including loss o f material due to pittin g and crevice corrosion, and crackin g due to stress c o rrosion crac kin g. In som e cases, when an aging effect is app licable to all of the various stainle ss steel categories, it can be assumed that the term "stainless steel" in the "Material" column of an AMR line-item in the GALL Report encompasses all stainless stee l types. Cast austenitic stainless steel (CASS) is quite susceptible to loss of fracture t oughness d ue to thermal and neutr on irradiatio n embrittleme n t. In addition, MRP-22 7-A indicate s that PH stainless ste e ls or martensitic stainle ss steels may be susceptible t o loss of fra c ture toughn ess by a thermal aging mechanism. Therefore, when loss of fracture tou ghness due to thermal and neutron irradiation embrittlement is an applicable a g ing effect a nd mechanism for a component in the GALL Report, the CASS, PH sta i nless steel, or martensitic stainless ste e l designat ion is spe c ifically identified in an AMR line

-item. Steel with st ainless steel cladding a l so may be co nsidered stainless ste e l when the aging effect is asso ciate d with the stainless ste e l surface of the material, rather than the composite volume of the material.

Exa m ples of stainless st eel designat ions that co mprise this category include A-286, SA193-Gr.

B8, SA193-Gr. B8M, Gr. 660 (A-286), SA193-6, SA193-Gr. B8 or B-8M, SA453, Type 416, Type 403, 410, 420 , and 431 martensitic st ainless steels, Type 15-5, 17-4, and 13-8-Mo PH stainle ss steels, and SA-193, Grade B8 a nd B8M bolting materials.

Exa m ples of wrought austenitic stain l ess materia l s that comprise this category include Type 304, 304NG, 304L, 308, 308L, 309, 309L, 316 and 347.

Exa m ples of CASS that comprise this categ o ry include CF3, CF3M, CF8 and CF8M. [Ref. 6, 7, 30]

A-49 IX.G References

30. Welding Handbook, Seventh Edition, Volume 4, Metals and Their Welda b ility, American Welding So ciety, 1984, p.76-145.

A-50 Appendix A, Section 2 - Revised version of the SRP-LR (1) Revised ver s ion of SRP-LR Table 3.0-1 Ta ble 3.0-1 FSA R Supple m ent for A g ing M a na ge me nt of A p plic a b le Sy s t e m s G A LL Chapter G A LL Progra m De sc ription of Progra m Imple m e n ta tion Sc he dule A p plicable GA L L Re port a nd S R P-LR Ch ap ter Ref e r e n ces X I.M16A PWR Vessel Internals The program relie s on impl ementation of the inspe c t i on and eval u a tion guidelin es in EPRI Tech nical Rep o rt No. 102 286 3 (MRP-22 7-A) and EPRI Tech nical Re port No. 10 16 609 (MRP-22 8) to manage the aging effects on the reacto r vessel internal comp one nts. This progra m is used to manag e (a) crackin g , inclu d ing stress corro s io n cra cki ng, prim ary water st re s s co rr osi on cr ac kin g , irra diat ion-a s sist e d st r e s s cor r o s io n c r ack i n g , a nd c r ack i n g d ue to fatigue/cycli c al loading; (b) loss of material in du ced by wear; (c) loss of fractu re toug hne ss d ue to either thermal a g ing , neutron irra diation embrittleme n t, or void swell i ng; (d) dimen s ion a l chang es d ue to void swelling o r di stortion; an d (e) loss of prelo ad du e to thermal a n d irradi ation-enha nced stress relaxatio n or cree p. Program sho u ld be impleme n ted prio r to perio d of extended operation GALL IV / SRP 3.1 (2) Revised ver s ion of SRP-LR Section 3.1.2, "Acce ptance Crite r ia" 3.1.2.2.9 Re m o ved as a result of LR-ISG-201 1-04 3.1.2.2.10 Re m o ved as a result of LR-ISG-201 1-04 3.1.2.2.12 Re m o ved as a result of LR-ISG-201 1-04 3.1.2.2.13 Re m o ved as a result of LR-ISG-201 1-04 3.1.2.2.14 Re m o ved as a result of LR-ISG-201 1-04 (3) Revised ver s ion of SRP-LR Section 3.1.3, "Review Procedures

" 3.1.3.2.9 Re m o ved as a result of LR-ISG-201 1-04 3.1.3.2.10 Re m o ved as a result of LR-ISG-201 1-04 3.1.3.2.12 Re m o ved as a result of LR-ISG-201 1-04 3.1.3.2.13 Re m o ved as a result of LR-ISG-201 1-04 3.1.3.2.14 Re m o ved as a result of LR-ISG-201 1-04 A-51 (5) Revised ver s ion of SRP-LR Table 3.1-1 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 3 BW R/ PWR Stainl ess steel or nicke l allo y r eactor ve ssel inter nal compo nents e x pose d to reactor coo l ant and ne utron flux Cumul a tive fati gue d a ma ge due to fatig u e F a tigue is a T L AA eval uate d for the per iod of ext end ed o per ation (Se e SRP, Section 4.3 "Metal F a tigue," for ac ceptab le methods to co mpl y w i th 10 CF R 54.2 1 (c)(1) Yes, T L AA (See subsecti on 3.1.2.2.1) IV.B1.R-53 IV.B2.RP-303 IV.B3.RP-339 IV.B4.R-53 IV.B1-14 (R-53) IV.B2-31 (R-53) IV.B3-24 (R-53) IV.B4-37 (R-53) 15 PW R Stainl ess steel Babcock &

Wilcox (including CASS, martensitic SS, and PH SS) and n i ckel all o y react o r vessel i n terna l compo nents expos ed to rea c tor coola n t and n eutro n flu x Red u ction of ductilit y an d fracture toug hn ess due to neutro n irrad i at ion embrittlem ent, and for CASS, marten sitic SS, and PH SS due to thermal agi ng embrittlem ent Ductilit y - Re du ction in fracture toug hn ess is a T L AA to be evalu a ted for the peri od of e x te nde d oper ation, See the SRP, Section 4.7, "Other Plant-Specific T L AAs," for accepta b l e methods of meetin g the re quir e ments of 10 CF R 54.2 1 (c). Yes, T L AA (See subsecti on 3.1.2.2.3.3)

IV.B4.RP-376 N/A 28 PW R Stainl ess steel Comb ustion Engi neer in g "Existi ng Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Loss of materi al du e to w e ar; cracki ng due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-400 N/A 32 PW R Stainl ess steel, nickel a llo y, or CASS react o r vessel intern als, core supp ort structure (not a l rea d y referenc ed as ASME Section XI Exa m inati on Categ o r y B-N-3 core supp ort structure compo nents i n MRP-22 7-A), exp o sed to reactor cool ant an d ne utron flu x Crackin g , or lo ss of material due to w e ar Chapter X I.M1, "ASME Section XI Inse rvice Inspection, Subsections IW B, IW C, and IW D" or Chapter X I.M16A, "PWR Vessel Internals," invoking app lica b le 10 CF R 50.55 a and ASME Sec t ion XI inservic e ins p e c tion requ ireme n ts for these compo nents No IV.B2.RP-382 IV.B3.RP-382 IV.B4.RP-382 IV.B2-26 (R-142) IV.B3-22 (R-170) IV.B4-42 (R-179)

A-52 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 51a PW R Stainl ess steel or nicke l allo y B abcock & W ilcox reactor inter nal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B4.RP-241 IV.B4.RP-241a IV.B4.RP-242a IV.B4.RP-247 IV.B4.RP-247a IV.B4.RP-248 IV.B4.RP-248a IV.B4.RP-249a IV.B4.RP-252a IV.B4.RP-256 IV.B4.RP-256a IV.B4.RP-258a IV.B4.RP-259a IV.B4.RP-261 IV.B4.RP-400 IV.B4-7 (R

-125) N/A N/A IV.B4-13 (R-194) N/A IV.B4-25 (R-210) N/A N/A N/A IV.B4-25 (R-210) N/A N/A N/A IV.B4-32 (R-203) N/A 51b PW R Stainl ess steel or nicke l allo y B abcock & W ilcox reactor inter nal "Exp ansi on" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, fatigu e, or overl oad Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B4.RP-244 IV.B4.RP-244a IV.B4.RP-245 IV.B4.RP-245a IV.B4.RP-246 IV.B4.RP-246a IV.B4.RP-254 IV.B4.RP-254a IV.B4.RP-260a IV.B4.RP-262 IV.B4.RP-352 IV.B4.RP-250a IV.B4.RP-375 IV.B4-7 (R

-125) N/A IV.B4-13 (R-194) N/A IV.B4-12 (R-196) N/A IV.B4-25 (R-210) N/A N/A IV.B4-32 (R-203) N/A N/A N/A 52a PW R Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g reactor internal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-312 IV.B3.RP-314 IV.B3.RP-322 IV.B3.RP-324 IV.B3.RP-326a IV.B3.RP-327 IV.B3.RP-328 IV.B3.RP-342 IV.B3.RP-358 IV.B3.RP-362a IV.B3.RP-363 IV.B3.RP-338 IV.B3.RP-343 IV.B3-2 (R

-149) IV.B3-9 (R

-162) N/A N/A N/A IV.B3-15 (R-155) IV.B3-15 (R-155) N/A N/A N/A N/A N/A N/A A-53 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 52b PW R Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g re actor intern al "Exp ans ion" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-313 IV.B3.RP-316 IV.B3.RP-323 IV.B3.RP-325 IV.B3.RP-329 IV.B3.RP-330 IV.B3.RP-333 IV.B3.RP-335 IV.B3.RP-362c NA IV.B3-9 (R

-162) N/A N/A IV.B3-12 (R-155) IV.B3-23 (R-167) N/A IV.B3-23 (R-167) N/A 52c PW R Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g re actor intern al "Ex i sting Progr a ms" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-320 IV.B3.RP-334 IV.B3-9 (R

-162) IV.B3-23 (R-167) 53a PW R Stainl ess steel or nicke l allo y W e sti ngh ouse reactor inter nal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B2.RP-270a IV.B2.RP-271 IV.B2.RP-275 IV.B2.RP-276 IV.B2.RP-280 IV.B2.RP-298 IV.B2.RP-302 IV.B2.RP-387 N/A IV.B2-10 (R-125) IV.B2-6 (R

-128) IV.B2-8 (R

-120) IV.B2-8 (R

-120) IV.B2-28 (R-118) N/A N/A 53b PW R Stainl ess steel W e stingh ouse reactor intern al "E xpa n s ion" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B2.RP-273 IV.B2.RP-278 IV.B2.RP-286 IV.B2.RP-291 IV.B2.RP-291a IV.B2.RP-291b IV.B2.RP-293 IV.B2.RP-294 IV.B2.RP-387a IV.B2-10 (R-125) IV.B2-8 (R

-120) IV.B2-16 (R-133) IV.B2-24 (R-138) N/A N/A IV.B2-24 (R-138) IV.B2-24 (R-138) N/A 53c PW R Stainl ess steel or nicke l allo y W e sti ngh ouse reactor inter nal "Existin g Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B2.RP-289 IV.B2.RP-301 IV.B2.RP-345 IV.B2.RP-346 IV.B2.RP-399 IV.B2.RP-355 IV.B2-20 (R-130) IV.B2-40 (R-112) N/A N/A N/A N/A A-54 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 54 PW R Stainl ess steel bottom mounte d instru ment sy stem flux thimble tubes (w ith or w i t h o u t ch ro me pl a t i ng) e x po sed to reactor cool ant and neutro n flu x (W esting ho use "Ex i sting Progr a ms" compo nents) Loss of materi al du e to we a r Chapter X I.M16A, "PWR Vessel Internals," or Chapter X I.M37, "Flux T h imble T ube Inspecti on" No IV.B2.RP-284 IV.B2-13 (R-145) 55a PW R Stainl ess steel or nicke l allo y B abcock and W ilc o x reactor inter nal "No Additi ona l Mea s ures" compo nents e x pose d to reactor coo l ant and ne utron flux N o a d d i ti on al ag i n g mana geme n t for reactor intern al "No Ad ditio nal Measur es" co mpon ents unl ess requ ire d b y ASME Section XI, Examin ation Categ o r y B-N-3 or relev ant oper ating e x pe rienc e e x ists Chapter X I.M16A, "PWR Vessel Internals" No IV.B4.RP-236 NA 55b PW R Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g re actor intern al "No Add i tion al Measur es" compo nents e x pose d to reactor coo l ant and ne utron flux N o a d d i ti on al ag i n g mana geme n t for reactor intern al "No Ad ditio nal Measur es" co mpon ents unl ess requ ire d b y ASME Section XI, Examin ation Categ o r y B-N-3 or relev ant oper ating e x pe rienc e e x ists Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-306 NA 55c PW R Stainl ess steel or nicke l allo y W e sti ngh ouse reactor inter nal "No Additi ona l Mea s ures" compo nents e x pose d to reactor coo l ant and ne utron flux N o a d d i ti on al ag i n g mana geme n t for reactor intern al "No Ad ditio nal Measur es" co mpon ents unl ess requ ire d b y ASME Section XI, Examin ation Categ o r y B-N-3 or relev ant oper ating e x pe rienc e e x ists Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-265 NA 56a PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y C o mb usti on Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-315 IV.B3.RP-318 IV.B3.RP-359 IV.B3.RP-360 IV.B3-7 (R

-165) IV.B3-8 (R

-163) N/A N/A A-55 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item Engi neer in g re actor intern al "Primar y" com ponents expos ed to rea c tor coola n t and n eutro n flu x due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar IV.B3.RP-362 IV.B3.RP-364 IV.B3.RP-366 IV.B3.RP-365 IV.B3.RP-326 N/A N/A N/A N/A N/A 56b PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS)

Comb ustion E ngi neer in g "Exp ans ion" re actor intern al compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-317 IV.B3.RP-331 IV.B3.RP-359a IV.B3.RP-361 IV.B3.RP-362b IV.B3-7 (R

-165) N/A N/A N/A N/A 56c PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y C o mb usti on Engi neer in g re actor intern al "Ex i sting Progr a ms" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-319 IV.B3.RP-332 IV.B3.RP-334a IV.B3.RP-336 IV.B3.RP-357 IV.B3-9 (R

-162) IV.B3-17 (R-156) N/A IV.B3-22 (R-170) N/A 58a PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y B abcock & W ilcox reactor inter nal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss Chapter X I.M16A, "PWR Vessel Internals" No IV.B4.RP-240 IV.B4.RP-240a IV.B4.RP-242 IV.B4.RP-247b IV.B4.RP-248b IV.B4.RP-249 IV.B4.RP-251 IV.B4.RP-251a IV.B4.RP-252 IV.B4-1 (R

-128) N/A IV.B4-4 (R

-183) N/A N/A IV.B4-12 (R-196) IV.B4-15 (R-190) N/A IV.B4-16 (R-188)

A-56 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item of prelo ad d ue to w e ar; or loss of materia l due to w e ar IV.B4.RP-254b IV.B4.RP-256b IV.B4.RP-258 IV.B4.RP-259 IV.B4.RP-401 N/A N/A IV.B4-4 (R

-183) IV.B4-31 (R-205) N/A 58b PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y B abcock & W ilcox reactor inter nal "Exp ansi on" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B4.RP-245b IV.B4.RP-246b IV.B4.RP-254b IV.B4.RP-260 IV.B4.RP-243 IV.B4.RP-243a IV.B4.RP-250 IV.B4.RP-375a N/A N/A N/A IV.B4-31 (R-205) IV.B4-1 (R

-128) N/A IV.B4-12 (R-196) N/A 59a PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y W e sti ngh ouse re actor internal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-270 IV.B2.RP-272 IV.B2.RP-296 IV.B2.RP-297 IV.B2.RP-302a IV.B2.RP-354 IV.B2.RP-388 IV.B2.RP-300 IV.B2-1 (R

-124) IV.B2-6 (R

-128) N/A N/A N/A N/A N/A N/A 59b PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS)

W e stingh ouse reactor intern al "E xpa n s ion" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-274 IV.B2.RP-278a IV.B2.RP-287 IV.B2.RP-290 IV.B2.RP-290a IV.B2.RP-290b IV.B2.RP-292 IV.B2.RP-295 IV.B2.RP-388a IV.B2-6 (R

-128) N/A IV.B2-17 (R-135) IV.B2-21 (R-140) N/A N/A IV.B2-21 (R-140) IV.B2-22 (R-141) N/A A-57 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ve s sel, Internals, an d Reac tor Coolant Sy stem Ev aluated in Chapte r IV of the GALL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item loss of materia l due to w e ar 59c PW R Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y W e sti ngh ouse re actor intern al "E xisti ng Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-285 IV.B2.RP-288 IV.B2.RP-299 IV.B2.RP-356 IV.B2-14 (R-137) IV.B2-18 (R-132) IV.B2-34 (R-115) N/A B-1 Appendix B MARK-UP OF CHANGES TO THE GALL REPORT AND SRP-LR

B-2 Appendix B, Section 1 - Mark-up of Changes to the GAL L Report In the mark-up, strikethr ough text indicates a de letion and u nderline text indicate s a n insertion.

Double strikethrough text indicate s t he original location of th e moved te xt and a doub le underline text indicate s the final lo cation of the moved te xt. (1) Mark-up of changes to GALL Report AMP XI.M16A XI.M16A PWR VESSEL INTERNALS Program Description This progra m relies on implementati on of the Electric Power Research I n stitute (EPRI)

Technical R eport No.

1016596 1022863, "Materials Reliability Program: Pressurized Water Reactor (PWR) Internals Inspe c tion and Evaluation Guidelines," (MRP-227-A) a nd EPRI Report Technica l No. 1016609 , "Materials Reliability Program: Inspection Stan dard for PWR Internals," (MRP-228) to manage the a g ing effects on the pressurized water reactor (PWR) reactor vessel int e rn al (RVI) components.

Th e recommen ded activitie s in MRP-227-A and addition al plant-spe cific a c tivitie s not define d in MRP-22 7-A are implemente d in accorda n ce with Nuclear Energ y Institute (NEI) 03-08, "Guideline f o r the Managemen t of Materials Issue s." T he staff appr oved the augmented inspection and evaluation (I&E) criteria for PWR RVI components in NRC Sa fety Evaluati on (SE), Revision 1, on MRP-22 7 by letter dated December 16, 2011.

This progra m is used to manage the effects of a ge-related d egradation mechanisms that are applicable in general to t he PWR RVI components at the facility.

These aging effects include : (a) various forms of cracking, including stress c o rrosion cra cking (SCC), which also encompasses primary water stress corrosion cracking (PW S CC), irradiation-assist ed stress corrosion cr acking (IAS CC), or and cracking d u e to fatigue/

cyclic al load ing; (b) loss of material induced by wear; (c) loss of fracture toughness d ue to either t hermal aging or neutron irradiation embrittleme n t; (d) chan ges in dime nsion s due t o void swelling or distortion

and (e) loss of preload due to thermal and irradiatio n-enhanced stress relax a tion or cree

p. The program applies th e guidance in MRP-227-A for inspect i ng, evaluating, and, if a pplicable, disposit ionin g non-confo r ming RVI c o mponents at the facility.

The program conforms to the definition of a sampling

-based condition monitoring program, as defined by the Bran ch Technical P o sition RSL B-1, with periodic examinations and other inspections of high ly-affected internals lo cations. These examinations provide reasonable assurance t hat the effects of age

--related deg radation mechanisms will be managed during th e period of e x tended operation.

The program includes expanding periodic ex aminations a nd other inspections

, if t he extent of the degradation effects identified exceeds the e x pected levels.

The MRP-2 2 7-A guidance for sele ct ing RVI components for inclusion in t he inspect i o n sample is based on a f our-step ran k ing proce s s. Through this proce s s, the reactor internals RVIs for all three PWR designs wer e assigned t o one of the following fou r groups: "Primary , ," "Expansion , ," "Existing Programs

, ," and "No Additional Measures components

." Definitions of e a ch group are provided in "Generic Aging Lesso ns Learned Report" (GALL Report), Revision 2, Chapter IX.B. The result of this four-ste p sample se lection pro c ess is a set of "Primary" I i nternals C c omponent locations for each of the three plant d e signs that are inspecte d because t hey are exp e cted to show the le ading indica tions of the degradation effects, with another set of "Expansion" B-3 I i nternals C c omponent locations that are specifie d to expand the sample should the indications be more severe than anticipated.

The degradation effects in a third se t of internals location s ar e deemed to be adequat ely managed by "Existing Programs," such as American Society of Mechanical Engineers (ASME) Code,Section X I , 11 Exa m ination Cate gory B-N-3 examination s of core su pport structu r es. A fourth set of internals lo cations are d eemed to require "Nn o a A dditional m M easures." As a result, the pr ogram typic a lly identifie s 5 to 15% o f the RVI locations as Primary Comp onent locations for inspections, with another 7 to 10% of the RVI locations to be inspected a s Expansion Component s, as warran t ed by the evaluation of the inspect i on results. A nother 5 to 15% of the internals lo cations are covered by Existing Programs, with the remainder requiring no additiona l measures.

This process thu s use s appropriate component functionality criteria, age-related degradation suscepti b ilit y criteria, an d failure con s equence cr iteria to iden tify the components that will be i n spected un der the program in a ma nner that co nforms to the sampling criteria fo r sampling

-based condition monitoring programs in Section A.1.2.3.4 of NRC Branch Position RLSB

-1. Consequently, the sample se lection pro c ess is a deq uate to assu re that the intended fun c tion(s) of th e PWR reactor internal components are maintained during t h e period of extended operation.

The program's use of visual examination methods in MRP

-2 27 for detect i on of relevant conditions (and the absence of rele vant conditions as a visual examina t ion accepta n ce criterion

) is consistent with the ASME Code,Section X I rules for visual examination. However, the program's adoption of th e MRP-227 guidance for visual examinations goe s beyond the ASME Code,Section X I visual examination criteria because additio nal guidance is incorp orated into MRP-227 to clarify how the particular visual examination methods will be u s ed to detect relevant conditions a nd describe s in more detail how the visual techniques relate t o the specif ic RVI components and how to detect their applicable a g e-related d egradation e ffects. The technical basis for detecting relevant conditions using volumetric ult r asonic te sting (UT) inspection t e chniques can be found in MRP-228, where the review of exi s ting bolt i ng UT examination technica l ju stificat ions h a s demonstrated the indication detection capabili t y of at least two vendors, and where vend or technica l ju stification is a requirement prior to any additional b o lting examinations. Spe c ifica lly, the capability of program's UT volumetric methods to detect loss of integrity of PWR internals bolt s , pins, an d fa steners, su ch as baffle

-fo rme r bolting in B

&W and Westinghouse units, has b een well de monstrated by operating experience.

In addition, t he program's adoption o f the MRP-2 27 guidance and process incorporat es the UT criteria in M R P-228, whi c h calls for t he technica l justification s that are ne eded for volumetric examination method demonstrations, required b y the ASME Code,Section V.

The program also inclu des future in dustry operating experience as in cor porated in p e riodic revisions to MRP-227. T he progra m thus provides reasonable assurance for the long

-term integrity and safe operation of reacto r internals in all commercial operatin g U.S. PWR nuclear power plants.

Age-related degradation in the reacto r internals is managed through an int egrated program.

S pecific feat ures of the integrated program are listed in the f o llowing ten program elements.

Degradation due to chan ges in material propertie s (e.g., loss of fracture t oughness) was considered in the determination of inspection re commendati ons and is manag ed by the requirement to use appr opriately degraded properties in the evaluation of identified d e fects. The integrated p r ogram is implemented by the applicant through an inspect i o n plan that is submitted to the NRC f o r review an d approval with the applicat ion f o r license rene wal. Evaluation and Technical Basis

11 Refer to the GALL Re po rt, Chapte r I, for applicability of various e d i t ions of the ASME Code,Section XI.

B-4 1. Scope of Program:

The scope o f the program includes all RVI components at t he [as an ad m i nistrati ve action ite m for the AMP, the app licant to f ill i n the nam e of the applicant' s nuclea r facilit y, inclu d ing applica b le units], w h ich [is/a re] built to a [

ap plicant to f ill in Westingh ouse, CE, or B&W, as applicable

] based on th e plant's ap plicable nu clear steam supply system NSSS design. The scope of th e program applies the m e thodology and guidance in MRP-22 7-Athe mos t recently NR C-endo rsed version of MRP

-22 7 , which pro v ides an augmented inspection and flaw evaluation methodology for assuring the functional integrity of safety-related intern als in commercial operating U.S. PWR nuclear power plants design ed by Babcock & Wilcox (B&W , ), Combustion Engineerin g (CE ,), and Westinghou se. The sco pe of components considered for inspection u nder in MRP-227 guidan ce include s-A include co re support st ructures (typically denoted as Exa m ination Category B

-N-3 by the ASME Code ,Section X I), , those RVI components that serve an intended license rene wal safety function pursu ant to criteria in 10 CFR 54.4(a)(1), and other RVI comp onents whose failure co uld prevent satisfactory accomplish m ent of any of the functions identif ie d in 10 CFR 54.4(a)(1)(i), (ii), or (iii).

In addition, ASME Code ,Section X I includes inspection requ irements for PWR remo vable core su pport structures in Table IWB-2500-1, Exa m ination Category B-N-3, which are in addition t o any inspection s t hat are implemented in accordance with MRP-2 27-A. The scope o f the program does not include con s umable items, such a s f uel assemblies, reactivity control assemblies, and n u clear instru mentation , because th ese components are not typically within the scop e of the components that are require d to be subject to an aging managemen t review (AMR), as defin ed by the criteria set in 1 0 CFR 54.21(a)(1).. The scope of the program also does n o t include w e lded attach ments to the internal surf ace of the re actor vessel beca u se these components are consider ed to be ASME Code Class 1 appurt enances to the reactor vessel and ar e adequately managed i n accordance with an applicant's AMP that corresponds to GALL AMP X I.M 1, "ASME Code ,Section X I Inservice In spection, Su bsection s IWB, IWC, and IWD."

The scope o f the program includes t he response bases to ap plicable lice n se renewal applicant a c tion items (L RAAIs) on the MRP

-22 7 methodology, and any additional pr ograms, actions, or a c tivities that are discussed in these LRAAI responses and credited for a g ing managemen t of the applicant's RVI components. The LRAAIs are identifie d in the staff

's safety evaluation on MRP

-227 and include applicable act i on items on meeting those assum p ti on s that formed the basis of the MRP' s augmented inspect i on and flaw evaluation methodology (as discussed in Section 2.4 of MRP

-227), and NSSS ven dor-specific or plant-specific LRAAIs as well.

The responses to the LRAAIs on MRP

-227 are provided in Appendix C of the LRA.

The guidance in MRP

-22 7 specifies a pplicability limitations to base-loaded plants and t he fuel loading managemen t assumptio n s upon which the functionality analyses were ba sed. These limita t ions and assumptions require a det ermination of appl icabilit y by the app licant for each reactor and are covered in Section 2.4 of MRP

-227. 2. Preventive Actions:

The guidan ce in MRP-2 2 7-A relies on PWR wat e r chemistry control to prevent or mitigate aging effects th at can be in duced by corrosive aging mechanisms (e.g., loss of material induced by general, pittin g corrosion, crevice corr osion, or str e ss corro si on crackin g or any of its forms [SCC, PWSCC, or IASCC]).

Reactor coola n t water chemistry is mo nitored and maintained in accor dance with t he Water Chemistry Program , as described. Th e program B-5 description , evaluation, and technical basis of water chemist r y are presented in GALL AMP X I.M 2, "Wat er Chemistry."

3. Parameters M onitored/Inspected:

The program manag es the follow i ng age-relat ed degradation effects and mechanisms that are ap plicable in g eneral to the RVI compo nents at the facility:

(a) cracking induced by SCC, PW SCC, IASCC, or fatigue/cyclic al load ing; (b) loss of material induced by wear; (c) loss of fractur e toughness induced by either thermal aging or neutron irra diation embrittlement; (d) changes in dimension s due to void swelling and irradiation growth, or distortion , or d e flection; an d (e) loss of preload cau s ed due tob y thermal an d irradiation

--enhanced st ress relaxation or creep.

For the management of cracking, th e program moni tors for evidence of surface brea king linear discontinu i ties if a visual inspecti on t e chnique is used as the non-destruct i on ve exami nation (NDE) meth od, or for relevant flaw presentation signals if a volumetric ult r asonic te sting (UT) method is used as the NDE metho d. For the managemen t of loss of material, the program monitors for gross or abn ormal surface condition s that may be indicative of loss of mat e rial occurring in the components.

For th e management of loss o f preload, th e program monitors for gross surfa c e condition s that may be indicative of loosening in applicable bolted, fast ened, keyed, or pinned conne ctions. The program does not direct ly monitor for loss of fra c ture toughness t hat is indu ced by thermal agi ng or n eutron irradiation embrittlement , or by void swelling and irradiation g r owth; inste a d. Instead

, the impact of loss of fra c ture toughn ess on component integrity is in directly managed by

(1) using visu al or volume tric examination techniques t o monitor for cracking in the components

, and by (2) applying applicable r educed fracture tou ghness prop erties in the flaw evaluations if , in cas e s where cracking is det ected in the components and is extensive enough to warrant necessitate a supplemental flaw growth or flaw toleran c e evaluatio n under the MRP-227 g u idance or ASME Code ,Section X I requirements.. The pro g ram uses physical measurements to monitor for any dimen s ional changes du e to void swelling or, irra diation growth, distortion , or deflectio n.. Specifically, the program impl ements the parameters monitored/inspecte d criteria for

[as an ad m i nistrative action ite m for the AMP, applicant is to se lect one of the f o llowing to finish the sentence, as applicable to its NSSS vendor for it s internals: "

f or B&W designed Primary Co m ponent s in Table 4

-1 of MRP-2 2 7"; "for CE designed Primary Co m p onents in Ta ble 4-2 of MRP-227"; and "for W e stinghouse designed P r im ary Co mpon ents in T able 4-3 of MRP-227"]. Additionally, the program impl ements the par ameters monitored/inspected cr it eria for [as an adm i n istrative action item for the AMP, ap plicant is to select one o f the followin g to finish the se ntence, as a pplicable t o its NSS S vendor for its internals: "fo r B&W designed Expansion Co m ponent s in Table 4

-4 of MRP-2 2 7"; "for CE designed Expansion Co mponents in Table 4

-5 of MRP-227"; and "for W e stinghouse designed E x pansion Co mponents in Table 4-6 of MRP-227"]. The p a rameters monit ored/inspected for Existing Program Co mp onents follow the b a ses for refe renced Existing Programs, such a s t he requirements for ASME Code Class RVI components in ASME Co de,Section X I, Table IW B-2500-1, Examination Categories B

-N-3, as implemented through t he applicant's ASME Code ,Section X I program, or the reco mmended p r ogram for inspecting W e stinghouse

-designed flux thimble tubes in GALL AMP X I.M 37, "Flux Thimble T ube Inspection." No insp ections, except for those specified in ASME Code ,Section X I, are re quire d for components that ar e identified as requiring "No Additional Measures," in accordance with the analyses rep o rted in MRP

-227. Specifically, the program implem ents the parameters monitored/inspecte d criteria co nsistent with the applicable tab l e s in Section 4, "Aging Management Requiremen t s," in MRP-227-A.

B-6 4. Detection of Aging Effects:

The detection of aging effe cts is covered in two places: (a) the guidance in Section 4 of MRP-227 provides an in troductory discussion an d justificatio n of the exa mination The inspect i on methods selected for detecting th e aging effe cts of intere st; and (b) standards fo r examinatio n are defined and established in Sect ion 4 of MRP-227-A. St andards for impleme n ting the inspection methods , proce dures, are d e fined and pe rsonnel are provided established in a companion document, MRP-228. In all ca ses, well-established inspection methods are were select ed. These methods include volume tric UT examination methods for detecting fla w s in bolting , physical measurements for detecting changes in dimension, and various visual (VT-3, VT-1, and EVT-1) exa m inations fo r detecting e ffects ranging fro m general conditions to detection an d sizing of surface-brea king discontinuities.

Surface examinations may also be used as an alte rnative to visual exami nations for d e tection and sizing o f surface-br eaking disco n tinuities.

Cracking ca used by SCC, IASCC, and fatigue is monitored/inspected by either VT-1 or EVT-1 examination (for internals other than bolting) or b y volu metric UT exa m ination (boltin g). The VT--3 visual methods may b e applied for the detectio n of crackin g in non-red undant RVI components only when the flaw toler ance of the component or affected a ssembly , as evaluated for reduced fracture tou ghness prop erties, is kn own and the component has been sh own to be tolerant of e a sily detecte d large flaw s, even under reduced f r acture toug hness cond itions. VT-3 visual methods are acceptable for t he detection of cracking in redundant RVI compo nents (e.g., redundant bolts or pin s used to secure a fasten ed RVI assembly).

In addition, VT-3 exami nations are used to monito r/inspect f o r loss of m a terial induced by wear and for general aging conditions, such as gross distortio n caused by void swelling and irradiation gr owth or by g r oss effect s of loss of pr eload cause d by thermal and irradiation-enhanced st ress relaxation and cree

p. In addition, t h e The program adopts the recomme nded guidan ce in MRP-2 2 7-A for defining the "Expansion criteria Criter ia" that need to be applie d to the insp ection find in g s of "Primary Component s and Existing Requirement Co mp onents " co mponents and for expanding the examination s to include additional "E x pansion Components. As a result," components. RVI component inspection s performed on the RVI components are performed consist ent with the inspection fr equency and sampling bases for "Pr i m a r y Co mponents, " components, "

Existing Requiremen t Componen ts Programs" components , and "Expansion Components

" co mponents in MRP-227-A, which have been demonstrated to be in conf orma nce with the inspection criteria, sampling basis cr iteria, and sample Expansion criteria in Section A.1.2.3.4 of NRC Br anch Position RLSB

-1. Specifically, the program impl ements the parameters monitored/inspecte d criteria an d bases for inspect i n g the relevant parameter condition s for [as an ad m i nistrative action item for the AMP, applicant is to se lect one of th e following t o finish the sentence, a s applicab le to its NSSS vend or for its internals: "B&W designed P r im ary Co mponents in Table 4

-1 of MRP-22 7"; "CE designed Primary Co m p onents in Ta ble 4-2 of MRP-227;" or "Westingho use designed Primary Co m p onents in Ta ble 4-3 of MRP-227"] and for [

as an ad m i nistrative action item for the AMP, applicant is to se lect one of th e following t o finish the sentence, as applicable t o its NSSS vendor for it s internals: "

f or B&W designed Expansion Com p onents in Table 4-4 of MRP-227;" "for CE designed expan sion com ponents in Tab le 4-5 of MRP-227;" and "for Westinghouse designed Expansion Co mponents in Table 4

-6 o f MRP-227"]. The program is supple m ented by t he following plant-specific Primary Component and Expansion Component inspection s f o r the program (as applicable): [

As a relevant license renewal applicant action item , the ap plicant is to list (using criteria in MRP

-22 7) each B-7 additional R V I com pone nt that need s to be insp ected as an additional p l ant-specific Pri m ary Co m ponent for the applicant' s progra m and each additional RVI co m pon ent that nee ds to be inspected a s an additio nal plant-sp ecific Exp a n s ion Com p o nent for t he applicant' s progra m. For each pla n t specif ic co m ponent added as an additional pr im ary or Ex pansion Co mponent, the list shoul d include th e applicabl e aging effect s that will be m onitored for, the inspe c tion method or methods used for m onit o ring, and th e sam p le size and frequencies for th e exam ination s]. In addition, in som e cas e s (as defin ed in M R P-227-A), physical measurements are used as supplemental technique s to manage for the gross effects o f wear, loss o f preload du e to stress relaxation, or for change s in dimension s due to void swelling

, deflection o r distortion.

The physical measurements methods applied in acco rdance with this program include [

Ap plicant to input physical m easure methods identified by th e MRP in re sponse to N RC RAI No. 11 in the NR C's Requ est for Additional Inform a t ion to Mr. Christen B. L a rson, EPRI MRP on To pical Report MRP

-227 dated Nove m ber 1 2 , 2009]. Inspection coverages for "Primary" a nd "Expansion" RVI components are implemente d consistent with Sections 3.3.1 and 3.3.2 of the NRC SE, Revision 1, on MRP-227. 5. M onitori ng and Trending:

The methods for monitoring, recording, e v aluating, and trending the data that result from the program's inspection s are given in Section 6 of MRP-227-A and its subsections.

The Flaw evaluation methods inclu d e , includin g recommen dations for flaw depth sizing and f o r crack gro w th determi nations as w e ll as for per forming applicable limit load, linear elastic a nd elastic-p l astic fracture an alyses of relevant flaw indications

., a r e defined in MRP-227-A. The examination s examination and re-examinations required by the that are implemented in accordan ce with MRP-227 guidance-A , together with the requirements criteria spe c ified in MRP-228 fo r inspectio n methodologies, inspe c tio n procedure s , and inspe c tion person nel, provide time ly detection, reporting, a nd corrective actions wit h respect to the effects o f the age-r elated degr adation mechanisms within the scop e of the pro g ram. The extent of the examination s , beginning with the sa mple of susceptible PW R internals component locations identified a s Primary Co mponent locations, with t he potential for inclusion of Expansi on Component locations if t he effects ar e greater than anticipat ed, plus the continuation of the Existing Programs activities, su ch as the ASM E Code,Section X I, Exa m ination Cate gory B-N-3 examination s for core su pport structu r es, provides a high deg ree of confid ence in the t o tal for timely detection, reportin g , and implementation of corrective a c tions for th e aging effe cts and mechanisms managed by the program.

The program applies a pplicable fra c ture toughn ess properties, inclu d ing reductions f o r thermal aging or neu tron embrittlement, in the flaw evaluations of the components in cases w here cracking is d e tected in a RVI compon ent and is e x tensive enough to warrant a supple m ental flaw growth or flaw tolerance evaluation.

For singly-represented components, the program incl udes crit eria to evaluate the agin g effects in the inaccessible por tions of the components and the resu lting impact on the inten ded function(s) o f the components. For r edundant co mponents (such as redu ndant bolts, screws, pins, keys, o r fasteners, some of which are acce ssible to inspection and some of which are not accessible t o inspect i on), the program includes criteria to evaluate the aging effects in the population o f components that are in accessible t o the applicable inspe c tion techniqu e and the resulting im pact on the intended fun c tion(s) of th e assembly containing t he components.

B-8 6. Acceptance Criteria

Section 5 of MRP-227-A, which includes Tab le 5-1 for B&W-designed RVIs, Table 5-2 for CE-designed RVIs, and Table 5-3 for Westinghouse-d e signed RVIs, provides the specif ic examination and flaw evaluation acceptance crit eria for the "Primary" an d "Expansion Component examination

s. For " RVI component examination methods. For RVI comp onents addressed b y exa m inations referenced to performed in acco rdance with the ASME Code,Section X I, the IWB-3500 acceptance criteria ap ply.in IWB-3500 are applicable.

For other RVI components covered by other "Existing Programs

, ," the examination acceptance cr iteria are described within the Existing Program applicable reference d o cument. The guidance in MRP

-22 7 contains t h ree types of examination As applica b le, the prog ram establishe s acceptance criteria

• For visual examination (and surface examination as an altern ative to visual examination), the examina t ion accepta n ce criterion is the abse n ce of any of the specif ic, descriptive relevant conditions; in a ddition, ther e are requir e ments to record and disposition surf ace breaking ind i cations th at are detecte d and sized for length by VT

-1/EVT-1 examinations;

• For volumetric examination, the examinati on acceptance cr i t erion is the capability fo r reliable dete c tion of ind i cations in bo lting, as demonstrated in the examina t ion Technical Justifi c ation; in addition, there are requirements for system

-level assessment of bolted or pinned asse mblies with unacceptable volumetric (UT) exa m ination indications that exceed specified limits; and For physical measureme n ts, the ex amination acceptance cr iterion for th e acceptable tolerance in the measured differe ntial height f r om the top of the plenu m rib pads to the vessel seating surface in B

&W plants a r e given in Table 5

-1 of MRP-227. T he acceptan ce criterion f o r physical measureme n ts performed on the he ight lim its of the Westing house-desig ned hold-do wn springs are [The incorporation o f this senten ce is a licen s e renewal applicant a c tion item for Westinghou se PWR applicants only - insert the applicable sentence incorporating t he specif ied any physical measureme n t criteria on ly if the app li cant's facility is based o n a Westing house NSSS design: the Westinghouse applicant is to incorporate t he applicable language and then specify the fit up lim its on the hold d o wn springs, as established on a pla nt-specific b a sis for the design of th e hold-down springs at t he applicant

's Westingh ouse-design ed facilit y].monitoring methods that are credite d for aging managemen t of particula r RVI compo nents. 7. Correctiv e Actions:

Corrective actions fo llowing th e dete c tion of una cceptable conditions are fundamentally provided for in ea ch plant's co rrective action program.

Any detected conditions t hat do not satisfy the examination acceptance criteria are re quired to be disposit ione d through th e plant corre ctive action program, wh ich may req u ire repair, replacement, or analytical evaluatio n for c ontinu ed service u n til the next inspection. The disposit ion will ensu r e that desig n basis fun c tions of the r eactor intern als compon ents will continue to be fulfilled f o r all li censi ng basis loa d s and events.

Example s of methodologies that can be used to analytically dispositio n unaccepta b le conditio n s are found in the ASME Code,Section X I o r in Section 6 of MRP-2 27. Section 6 of MRP-2 27 describe s the option s that are avail able for disposit ion of detected conditions t hat exceed the examinat ion accepta n ce criteria of Section 5 of the report. These include enginee ring evaluation methods, as well as supplementary exa m ina t ions to furth e r characterize the detected conditio n , or t he alte rnative of component repair and re placement procedures.

The latter ar e subject to the require ments of the ASME Code ,Section X I. The implementation of t he guidance in MRP-227 The implementation of the guida nce in MRP-227-A , plus the impleme n tation of an y ASME Co de requirements, provides an acceptable level of aging management of safety-related components add ressed in B-9 accordance with the corr ective action s of 10 CFR Part 50, Appendix B or its equivalen t, as applicable.

Other alternative correct ive action act i ons base s may be use d to disposition relevant conditions if they have been previously approved or endorsed by the NRC.

Exa m ples of previously NRC

-endorsed alt e rnative corrective action s bases include those corrective act i ons base s f o r Westinghou s e-design R V I components that are defined in T ables 4-1, 4-2, 4-3, 4-4, 4-5, 4-6, 4-7 and 4-8 of Westinghou se Report No. WCAP

-14 577-Rev. 1-A, or for B&W

-designed RVI components in B&W Report No. BAW

-2248. Westinghouse Report No. WCAP

-1457 7-Rev. 1-A was endo rsed for use in an NRC SE to the West inghouse Owners Grou p, dated February 10, 2001. B&W Report No. BAW

-2248 was endorsed for use in an SE to Framatome T e chnologie s on behalf of the B&W Owners Grou p, dated December 9, 1999.

Alternative correct ive action bases act i on s not approved or endorsed by the NRC will be submitted for NRC appr oval prior to their implementation.

8. Confirmation Process:

Site quality assurance procedure s , review and approval processes, and administrative controls are im plemented in accordance with the recommendatio ns of NEI 03-08 and the r equirements of 10 CFR Part 50, Appendix B, or their equivalent, as applicable.

It is expected that the The implementati on of the guidance in MRP-227 will provide-A, in conjunct i o n with NEI 03-08 and othe r guidance d o cuments, reports, or methodologie s referenced in this AMP, pro v ides an accep t able level o f quality and an accepta b le basis for inspection confirming the quality of inspect i ons , flaw evaluation, and oth e r elements of aging management of th e PWR internals tha t are addressed in accor dance with t he 10 CFR Part 50, Appendix B, or their equivalent (as applica b le), confirmation process, and administrative cont rols evaluations, and corrective actions. 9. Administrative Controls:

The administrative controls fo r such these types of programs, including t h eir implementing proced ures and review and approval processes, are imp l emented in accordan ce with the r e commende d industry guidelines a n d criteria in NEI 03-08, and are under existing site 10 C F R 50 Appendix B

, Qual ity Assurance Programs, or their equivalent, as applicable.

Such The evaluation in Section 3.5 of the NRC' s SE, Revision1, on MRP-227 provides the basis for en dorsing NEI 03-08. This includes e n dorsement of the criteria in NEI-03-08 for notifying the NRC of a n y deviation from the I&E methodology in MRP-227-A and justifying the deviation no later than 45 days after its approval by a program is thus expected to be establish ed with a su fficient level of documentation and ad mini strative controls to e n sure effective lon g-term imple m entation licensee executive

. 10. Operati ng Experience:

Relatively few incidents of PWR internals aging degra dation have been reported in operating U.S. commercial PWR plants. A summary of observations t o date is provided in Appendix A of MRP-227-A. The applicant is expected to review subsequ ent and assessment of relevant operating experience for impact its impacts on it s th e program or to participate in industry in itiatives that perform this function.

The applicat ion of the MRP

-227 guidance will est ablish a con s iderable a m ount of , including implementing procedure s , are governed by NEI 03-08 and Appendix A of MRP-227-A.

Consistent with MRP-2 27-A, the reporting of in spection re sults and ope rating experience over the next few years. Section 7 of MRP

-227 describ es the repor ting require ments for these application s , and the pla n for evaluating the accu mulated additional opera t ing experience is treated as a "Needed" category item under the implementation of NEI 03-08.

B-10 The program is informed and enhan ced when nece ssary thr ough the systematic an d ongoing review of both plant-spe cific and ind u stry operating experience, as discu ssed in App endix B of the GALL Report, which is documen ted in LR-ISG-2011-05.

References 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants , Office of the Federal Register, Nation al Archives and Records Administration, 2009 20 11. 10 CFR Part 50.55a, Codes and Sta ndards , Office of the Fe deral Regist er, National Archives and Records Administration, 2009.2 011. ASME Boiler & Pressure Vessel Code,Section V, Nondestructive Exam ination , 2004 Edition, American Society of Mechanical En gineers, Ne w York, NY.

ASME Boiler & Pressure Vessel Code,Section XI, Rules for I n service I n spection of N u clear Power Plant Co m ponent s , The ASME Boiler and Pressure Vessel Code, 2004 edition as approved in 10 CFR 50.55a, The American Society of Mechanical Engin eers, New York, NY. B&W Report No. BAW-2 248, De m o n s tration of th e Manage ment of Aging Effects for t he Reactor Vessel Inter nals , Framatome Technologies (now AREVA Technologies), Lynchburg VA, July 1997. (NRC Microfiche Accession Number A0076 , Microfiche Pages 001

- 108). EPRI 10149 86, PWR Primary Water Chem istry Guidelines , V o lume 1, Revision 6, Ele c tric Power Rese arch Institut e, Palo Alto, CA, December 2007. (Non

-publicly available ADAMS Accession Number ML0 81140278). The non

-proprietary vers ion of the re port may ac cessed by me mbers of th e public at ADAMS Accessio n Number ML 081230449 EPRI 10165 96, Materials Reliability Progra m: Pressurized Water Reactor Internals I n spection and Evaluation Guidelin es (MRP-22 7 , -Rev ision. 0), Electric Power Rese arch Institut e, Palo Alto, CA:

2008. EPRI Technical Report No. 1022863, Materials Reliability Program

Pressurized Wa ter Reactor Internals Inspection and Evaluation Guidelines (M RP-227-A), December 2011, ADAMS Accession No. ML12017A193 (Transmittal letter f r om the EPRI-MRP) an d ADAMS Accession Nos. ML12017A194, ML 12017A196, ML12017A197, ML12017A191, ML 12017A192, ML12017A1 95 and ML12017A199, (Final Report).

EPRI Technical Report No. 1016609, Materials Reliability Program: Inspection Stand ard for PWR Internals (MRP-22 8), Electric Power Rese arch Institut e, Palo Alto, CA, July 2009

. (Non-publicly available ADAMS Accession N o.u m ber ML09 2120574). The non-proprietary version of the report may accessed by me mbers of the public at ADAMS Accessio n N o.umber ML092750569.

NRC RAI N o. 11 in the NRC's Requ est for Additional Inform a t ion to the Mr. Christen B. Larson, EPRI MRP on Topical Report MRP-227 dated Nove mber 1 2 , 2009. NRC Safet y Evaluation from C. I. Gri m es [NRC] t o R. A, Newton [Chairman, Westinghouse Owners Gro up], Accepta n ce for Referencing of Generi c Licen s e Renewal Progra m To pical Report Entitled "License Renewal Evaluation: Aging Management for Re actor Intern als," WCAP-1457 7, Revision 1, February 10, 2001. (ADAMS Accession Number ML010430375).

B-11 NRC Safet y Evaluation from C. I. Gri m es [NRC] t o W. R. Gra y [Framato me Technologies], Acceptance for Referencing of Generic Licen s e Renewal Progra m Topical Report Entitled "De m onstrat ion of the Managem ent of Aging Effects for the Reactor Vessel Internals," February 10, 2001. (ADAMS Accession Number ML993490288

). NURE G-18 00, Revision 2, Standard Review Pla n for Review of License Renewal Ap plication s for Nuclear Power Plants, Appendix A.1, "Aging Managemen t Review - Generic (Branch Technical P o sition RLS B-1)," U.S.

Nuclear Regulatory Commission, Wa shington, D C , 2010. Westinghou se Non-Proprietary Class 3 Report No. WCAP-14577-Rev. 1-A, License Renewal Evaluation:

Aging Manage m ent for Reactor Internals , Westinghouse Ele c tric Company, Pittsburgh, PA [March 2 001]. Report was submitted to the NRC Docume nt Control Desk in a letter dated April 9, 200

1. (ADAMS Accession Number ML0 11080790).

NRC Interim Staff Guidance LR-ISG-2011-05, Ongoing Revie w Of Operat ing Experien c e , March 16, 2012, (ADAMS Accession No. ML12044A215).

Nuclear Energy Institute (NEI) Report No. 03-08, Revision 2, Guideline fo r the Manage m ent of Materials Issues , ADAMS Accession No. ML101 050334). NRC Safet y Evaluation from Robert A. Nelson (NRC) to Nei l Wilmshurst (EPRI), Re vision 1 to the Final Sa fety Evaluation of Electric Power Re search Inst it ute (EPRI)

Report, Mat e rials Reliability Progra m (MRP) Report 1016596 (MRP-227), Revision 0, Pressurized Wat e r Reactor Internals Inspe c tion and Eva l uation Guid elines , Dece mber 16, 2011, ADAMS Accession No. ML11308A770.

B-12 (2) Mark-up of changes to GALL Report Chapter IV.B2 B2. REACT OR VESSEL I N TERNAL S (P WR) - WE STINGHOU SE Sy stems, Structures, and Components This section addresses t he Westingh ouse pressu rized -water reactor (PWR) vessel internals and consist s of , which consist of components in the upper intern als assembly, the control rod guide tube assem b lies assembly , the core barrel asse m b ly , the ba ffle/former assembly, the lower internal assembly, and t h e internals assembly, lower support assembly, thermal shield assembly, bottom mou n ted instrumentation sup port structur es. Based o n Regulatory Guide 1.2 6 , "Quality Group Classification s an d Standards for Water

-, Steam-, and Radioactive

-Waste-Containing Component s of Nuclear Power Plants," all struc t u res and co mponents that comprise the reactor vessel are g o verned by Group A or B Quality St andards.system, and alignment and interfacing components.

Common mi scellaneou s material/environment combinations where aging effects are not expected to degrade the ability of th e structure o r component to perform its intended f unction for the period of extended operation are included in IV.E. Sy stem Interfaces The systems that interfa c e with the r eactor vessel internals include the r eactor pressure vessel (IV.A2). Inspection Plan An applicant will submit an inspect i on plan for reactor internals to the NRC for review and approval with the application for lice nse renewal in accordan ce with Chapter X I.M 16 A , "PWR Vessel Inter nals."

B-13 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-300 IV.B2-33 (R-108) Alignment and interfacing components:

internals hold down spring Stainless steel Reactor coolant and neutron flux Loss of preload due to thermal and irradiation enhanced stress relaxation; changes in dimensions due to void swelling or distortion; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals"

Primary components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-301 IV.B2-40 (R-112) Alignment and interfacing components:

upper core plate alignment pins Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking 'Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-299 IV.B2-34 (R-115) Alignment and interfacing components:

upper core plate alignment pins Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-271 IV.B2-10 (R-125) Baffle-to-former assembly: accessible baffle-to-former bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Expansion components see AMR Items IV.B2.RP-273 and IV.B2.RP

-286)SCC mechanisms only)

No B-14 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-272 IV.B2-6 (R-128) Baffle-to-former assembly: accessible baffle-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; change changes in dimensions due to void swelling or distortion

loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) (for Expansion components see AMR Items IV.B2.RP

-274 and IV.B2.RP-287) No IV.B2.RP-270 IV.B2-1 (R-124) Baffle-to-former assembly: baffle and former plates Stainless steel Reactor coolant and neutron flux Changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-270a IV.B2-1 (R-124) Baffle-to-former assembly: baffle and former plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B2.RP-275 IV.B2-6 (R-128) Baffle-to-former assembly: baffle-edge bolts (all plants with baffle-edge bolts)

Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Primary components (identified in the "Structure and Components" column) no Expansion components No B-15 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-354 Baffle-to-former assembly: baffle-edge

bolts (all plants with baffle-edge bolts)

Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion

loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-273 IV.B2-10 (R-125) Baffle-to-former assembly

barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry (for Primary components see AMR Item IV.B2.RP

-271)SCC mechanisms only)

No IV.B2.RP-274 IV.B2-6 (R-128) Baffle-to-former assembly: barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation

embrittlement; changes in dimensions due to void swelling or distortion

loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-272) No B-16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-284 IV.B2-12 (R-143) Bottom mounted instrument system: flux

thimble tubes Stainless steel (with or without

chrome plating)

Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column)

No expansion components; and or Chapter XI.M37 , " ,"Flux Thimble Tube Inspection

"" No IV.B2.RP-293 IV.B2-24 (R-138) Bottom-mounted instrumentation system:

bottom-mounted instrumentation (BMI) column bodies Stainless steel Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-298) No IV.B2.RP-292 IV.B2-21(R-140) Bottom-mounted instrumentation system:

bottom-mounted instrumentationinstrume nt (BMI) column bodies Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals"

Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-297) No IV.B2.RP-296 Control rod guide tube (CRGT) assemblies: CRGT guide plates (cards) Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Primary Components (identified in the "Structure and Components" column) (for Expansion components see AMR Line Item IV.B2.RP-386) No B-17 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-298 IV.B2-28 (R-118) Control rod guide tube (CRGT) assemblies:

CRGT lower flange welds (accessible)

Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Expansion components see AMR Items IV.B2.RP-291 and IV.B2.RP-293)SCC mechanisms only)

No IV.B2.RP-297 Control rod guide tube (CRGT) assemblies: CRGT lower flange welds (accessible)

Stainless steel (including CASS)

Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging and neutron irradiation

embrittlement and for CASS, due to thermal aging embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) (for Expansion components see AMR Items IV.B2.RP

-290 and IV.B2.RP-292) No IV.B2.RP-386 Control rod guide tube (CRGT) assemblies: C

-tubes and sheaths Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) are only the components associated with a primary component that exceeded the acceptance limit.

(for Primary components see AMR Item IV.B2.RP

-296) No IV.B2.RP-355 IV.B2.RP-355 Control rod guide tube (CRGT) assemblies: guide tube support pins (split pins)

NickelStainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking and or fatigue A plant-specific aging management program is to be evaluatedChapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) Yes, plant

-specific No B-18 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-356 Control rod guide tube (CRGT) assemblies: guide tube support pins (split pins)

NickelStainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear A plant-specific aging management program is to be evaluatedChapter XI.M16A, "PWR Vessel Internals" Yes, plant

-specific No IV.B2.RP-387 Core barrel assembly:

upper core barrel axialand lower core barrel circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking , and or irradiation-assisted stress

-corrosion cracking or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B2.RP-276SCC mechanisms only) No IV.B2.RP-387a Core barrel assembly: upper core barrel and lower core barrel vertical (axial) welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-388 Core barrel assembly:

upper core barrel axialand lower core barrel circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-276) No IV.B2.RP-282 388a IV.B2-8(R-120) Core barrel assembly:

upper core barrel flangeand lower core barrel vertical (axial) welds Stainless steel Reactor coolant and neutron flux Cracking Loss of fracture toughness due to stress corrosion cracking and fatigueneutron irradiation embrittlement Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-276) No B-19 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-345 Core barrel assembly: core barrel flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-278 IV.B2-8 (R-120) Core barrel assembly: core barrel outlet nozzle welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking and fatigue Cracking due to stress corrosion cracking or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion component (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B2.RP-276)SCC mechanisms only) No IV.B2.RP-280 278a IV.B2-8(R-120) Core barrel assembly: lower core barrel flange weldoutlet nozzle welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking and irradiation

-assisted stress corrosion cracking Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion component (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-276) No IV.B2.RP-281 280 IV.B2-98 (R-122 120) Core barrel assembly: lower core barrel flange weld Stainless steel Reactor coolant and neutron flux Loss of fracture toughnessCracking due to neutron irradiation embrittlement stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" Expansion Components (identified in the "Structure and Components" column) Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B2.RP

-276)SCC mechanisms only)

No B-20 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-276 IV.B2-8 (R-120) Core barrel assembly: upper core barrel flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking and irradiation-assisted stress corrosion cracking Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Primary components (identified in the "Structure and Components" column) (for Expansion components see AMR Items IV.B2.RP

-278, IV.B2.RP-280, IV.B2.RP

-282, IV.B2.RP-294, IV.B2.RP

-295,IV.B2.RP

-281, IV.B2.RP

-387, and IV.B2.RP

-388) No IV.B2.RP-285 IV.B2-14 (R-137) Lower internals assembly: clevis insert

bolts or screws Nickel alloy Reactor coolant and neutron flux Loss of material due to weardue to wear; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals"

Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-399 Lower internals assembly: clevis insert bolts or screws Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to primary water stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-289 IV.B2-20 (R-130) Lower internals assembly: lower core plate and extra-long (XL) lower core plate Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion cracking , and or fatigue 'Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Existing Program components (identified in the "Structure and Components" column) no Expansion components No B-21 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-288 IV.B2-18 (R-132) Lower internals assembly: lower core plate and extra-long (XL) lower core plate Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals"

Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-291 IV.B2-24 (R-138) Lower support assembly: lower support column bodies (cast)

Cast austenitic stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion cracking Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-298) No IV.B2.RP-290 IV.B2-21 (R-140) Lower support assembly: lower support column bodies (cast)

Cast austenitic stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging and neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-297) No IV.B2.RP-291a Lower support assembly: lower support forging or casting Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B2.RP-290a Lower support assembly: lower support forging or casting Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement (and thermal aging embrittlement for CASS, PH SS, and martensitic SS) Chapter XI.M16A, "PWR Vessel Internals" No B-22 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-294 IV.B2-24 (R-138) Lower support assembly: lower support

column bodies (non-cast) Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion crackingChapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-276) No IV.B2.RP-295 IV.B2-22(R-141) Lower support assembly: lower support column bodies (non-cast) Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion Components (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-276) No IV.B2.RP-286 IV.B2-16 (R-133) Lower support assembly: lower support column bolts Stainless steel

nickel alloy Reactor coolant and neutron flux Cracking due to irradiation-assisted stress -corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" col umn) and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B2.RP-271)SCC mechanisms only) No IV.B2.RP-287 IV.B2-17 (R-135) Lower support assembly: lower support column bolts Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation

embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals" Expansion component (identified in the "Structure and Components" column) (for Primary components see AMR Item IV.B2.RP

-272) No B-23 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-303 IV.B2-31 (R-53) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Cumulative fatigue damage due to fatigue Fatigue is a time-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes , - TLAA IV.B2.RP-24 IV.B2-32 (RP-24) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to pitting and crevice corrosion Chapter XI.M2, "Water Chemistry" No IV.B2.RP-268 382 IV.B2-26 (R-142) Reactor vessel internal internals: ASME Section XI, Examination Category B-N-3 core support structure components (inaccessible locations)not already identified as "Existing Programs" components in MRP-227-A)

Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue, stress corrosion cracking, and or irradiation

- assisted stress corrosion cracking

loss of material due to wear Chapter XI.M2, "Water Chemistry,"

M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" or Chapter XI.M16A, "PWR Vessel Internals

" ," by invoking applicable 10 CFR 50.55a and ASME Section XI inservice inspection requirements Yes, if accessible Primary, Expansion or Existing program components indicate aging effects that nee d management No IV.B2.RP-302 Thermal shield assembly: thermal shield flexures Stainless steel Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No B-24 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-269 302a Reactor vessel internal components (inaccessible locations)Thermal shield assembly: thermal shield flexures Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; change in dimension due to void swelling; loss of preload due to thermal and irradiation enhanced stress relaxation; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Yes, if accessible Primary, Expansion or Existing program components indicate aging effects that need manag ement NoIV.B2.RP-265 Reactor internal "No Additional Measures" componentsReactor vessel internal components with no additional measures Stainless steel; nickel alloy Reactor coolant and neutron flux No additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience existsCracking due to stress corrosion cracking, and irradiation

-assisted stress corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Note: Components with no additional measures are not uniquely identified in GALL tables

- Components with no additional measures are defined in Section 3.3.1 of MRP

-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" No IV.B2.RP-267 291b Reactor vessel internal components with no additional measures Upper Internals Assembly; upper core plate Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of fracture toughness Cracking due to neutron irradiation embrittlement; change in dimension due to void swelling; loss of preload due to thermal and irradiation enhanced stress relaxationfatigue; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Note

Components with no additional measures are not uniquely identified in GALL tables

- Components with no additional measures are defined in Section 3.3.1 of MRP

-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" No IV.B2.RP-382 IV.B2-26(R-142) Reactor vessel internals: core support structure Stainless steel; nickel alloy; cast austenitic stainless steel Reactor coolant and neutron flux Cracking, or Loss of material due to wear Chapter XI.M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" No B-25 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B2.RP-302 290b Upper Internals Assembly; upper core plateThermal shield assembly: thermal shield flexures Stainless steel Reactor coolant and neutron flux Cracking due to fatigue; loss Loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals"

Primary components (identified in the "Structure and Components" column) no Expansion components No IV.B2.RP-346 Upper i Internals a Assembly: upper support ring or skirt Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking and or fatigue 'Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only) Existing Program components (identified in the "Structure and Components" column) no Expansion components No B-26 (3) Mark-up of changes to GALL Report Chapter IV.B3 B3. REACT OR VESSEL I N TERNAL S (P WR) - C O MBU S TION ENGI NEER ING Sy stems, Structures, and Components This section addresses t he Combustion Enginee ring (CE) pressurized

-water reactor (PWR) vessel inter nals and consists of , which con s ist of components in the up per internals assembly, the control e l ement asse mbly (CEA) shrouds,), the core sup port barrel a ssem b ly , the core shroud asse mbly, and th e lower internal assembly. Based on Regulatory Guide 1.26, "Quality Group Class ifications and Standards for Water

-, Steam-, and Radioa ctive-Waste-Containing Components of Nuclear Power Pl ants," all structures and components that compri se the reactor vessel are governed by Group A or B Quality Standards support structure assembly, and encore instr u mentation (ICI) components

. Common mi scellaneou s material/environment combinations where aging effects are not expected to degrade the ability of th e structure o r component to perform its intended f unction for the period of extended operation are included in IV.E. Sy stem Interfaces The systems that interfa c e with the r eactor vessel internals include the r eactor pressure vessel (IV.A2). Inspection P l an An applicant will submit an inspect i on plan for reactor internals to the NRC for review and approval with the applica tion for lice nse renewal in accordan ce with Chapter X I.M 16 A, "PWR Vessel Inter nals."

B-27 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-312 IV.B3-2 (R-149) Control Element Assembly (CEA):

shroud assemblies:

instrument guide tubes in peripheral CEA assemblies Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Expansion components see AMR Item IV.B3.RP-313)SCC mechanisms only)

No IV.B3.RP-313 Control Element Assembly (CEA):

shroud assemblies: remaining instrument guide tubes in CEA

assemblies Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, ""PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

" and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B3.RP-312)SCC mechanisms only)

No IV.B3.RP-320 IV.B3-9 (R-162) Core shroud assemblies (all plants): guide lugs and; guide lug insertinserts and bolts Stainless steel Reactor coolant and neutron flux Cracking due to fatigue

'Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B3.RP-319 IV.B3-9 (R-162) Core shroud assemblies (all plants): guide lugs and; guide lug insertinserts and bolts Stainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of preload due to thermal and irradiation

enhanced stress relaxation or creep Chapter XI.M16A, "PWR Vessel Internals

"" Existing Program components (identified in the "Structure and Components" column) no Expansion components No B-28 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-358 Core shroud assemblies (for bolted core shroud assemblies): (a)assembly components, including shroud plates and (b) former plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column)

(for Primary component see AMR Item IV.B3.RP-314) No IV.B3.RP-318 IV.B3-8 (R-163) Core shroud assemblies (for bolted core shroud assemblies): (a)assembly components, including shroud plates and (b) former plates Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement;

changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion components No IV.B3.RP-316 IV.B3-9 (R-162) Core shroud assemblies (for bolted core shroud assemblies): barrel-

shroud bolts with neutron exposures greater than 3 dpa Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, ""PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

" and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B3.RP-314)SCC mechanisms only)

No IV.B3.RP-317 IV.B3-7 (R-165) Core shroud assemblies (for bolted core shroud assemblies): barrel-shroud bolts with neutron exposures greater than 3 dpa Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of fracture toughness due to neutron

irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B3.RP-315) No B-29 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-314 IV.B3-9 (R-162) Core shroud assemblies (for bolted core shroud assemblies): core shroud bolts (accessible)

Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking and or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Expansion components see AMR Items IV.B3.RP-316, IV.B3.RP

-330, and IV.B3.RP

-358)SCC mechanisms only)

No IV.B3.RP-315 IV.B3-7(R-165) Core shroud assemblies (for bolted core shroud assemblies): core shroud bolts (accessible)

Stainless steel Reactor coolant and neutron flux Loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of fracture toughness due to neutron

irradiation embrittlement; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals

," Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Items IV.B3.RP-317, and IV.B3.RP

-331)" No IV.B3.RP-359 Core shroud assemblies (welded):

(assembly (designs assembled in two vertical sections): core shroud plates and (b) plate-to-former platesplate welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals

," Primary components (identified in the "Structure and Components" column) no Expansion components

" No B-30 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-322 Core shroud assembly (for welded core shrouds designs assembled in two vertical sections): Core shroud plate

-former plate weld (a) The axial and horizontal weld seams at the core shroud re-entrant corners as visible from the core side of the shroud, within six inches of the central flange and horizontal stiffeners, and (b) the horizontal stiffen ers in core shroud plate-to-former plate weld welds Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Item IV.B3.RP-323)and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-326 Core shroud assembly (for welded core shrouds designs assembled in two vertical sections):

gap betweenassembly components, including monitoring of the upper and lower plates gap opening at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Changes in dimensions due to void swelling or distortion; loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion components No B-31 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-323326a Core shroud assembly (for welded core shrouds designs assembled in two vertical sections): remaining axial welds in assembly components, including monitoring of the gap opening at the core shroud plate-to-former platere-entrant corners Stainless steel Reactor coolant and neutron flux Cracking due to irradiation

-assisted stress -corrosion cracking or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B3.RP-322)SCC mechanisms only)

No IV.B3.RP-324323 Core shroud assembly (for welded core shrouds with full-height shroud plates): axial weld seams at the core shroud re-entrant corners, at the core mid-plane (+3 feet in height) as visible from the core side of the shroud Core shroud assembly (designs assembled in two vertical sections): remaining axial welds Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Item IV.B3.RP-325)and Chapter XI.M2, "Water Chemistry" No B-32 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-360 359a Core shroud assembly (for welded core shrouds with full-height shroud plates): axial weld seams at the core shroud re-entrant corners, at the core mid-plane (+3 feet in height) as visible from the core side of the shroud Core shroud assembly (designs assembled in two vertical sections): remaining axial welds Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement

change s in dimension s due to void swelling or distortion Chapter XI.M16A, ""PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Item IV.B3.RP-361)" No IV.B3.RP-325324 Core shroud assembly (for welded core shrouds designs assembled with full-height shroud plates): remainingshroud plate axial welds, ribs, and ringsweld seams at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B3.RP-324) No B-33 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-361360 Core shroud assembly (for welded core shrouds designs assembled with full-height shroud plates): remainingshroud plate axial welds, ribs, and ringsweld seams at the core shroud re-entrant corners Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B3.RP-360) No IV.B3.RP-362325 Core support barrelshroud assembly: lower cylinder (designs assembled with full-height shroud plates):

remaining axial welds, ribs, and rings Stainless steel Reactor coolant and neutron flux Loss of fracture toughnessCracking due to neutron irradiation embrittlement-assisted stress corrosion cracking Chapter XI.M16A, ""PWR Vessel Internals" Expansion components (identified in the "Structure

" and Components" column)Chapter XI.M2, "Water Chemistry""

(for Primary components see AMR Item IV.B3.RP-327) No IV.B3.RP-329361 IV.B3-15(R-155) Core support barrelshroud assembly: lower cylinder (designs assembled with full-height shroud plates):

remaining axial welds , ribs, and remaining core barrel assembly welds rings Stainless steel Reactor coolant and neutron flux Cracking Loss of fracture toughness due to stress corrosion cracking neutron irradiation embrittlement Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "" PWR Vessel Internals

"" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B3.RP-327) No IV.B3.RP-333362 Core support barrel assembly: lower flange weld, if fatigue life cannot be demonstrated by TLAAcylinder circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Cracking Loss of fracture toughness due to fatigueneutron irradiation embrittlement Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "" PWR Vessel Internals

"" Primary components (identified in the "Structure and Components" column) no Expansion components Yes, evaluate to determine the potential locations and extent of fatigue cracking No B-34 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-389 362a Core support barrel assembly: lower flange weld (if fatigue analysis exists)cylinder circumferential (girth) welds Stainless steel Reactor coolant and neutron flux Cumulative fatigue damage due to fatigueCracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Yes, TLAANo IV.B3.RP-328362b IV.B3-15(R-155) Core support barrel assembly: surfaces of the lower core barrel flange weld (accessible surfaces)cylinder vertical (axial) welds Stainless steel Reactor coolant and neutron flux Cracking Loss of fracture toughness due to stress corrosion cracking and fatigueneutron irradiation embrittlement Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "" PWR Vessel Internals

" Primary components (identified in the "Structure and Components" column) no Expansion components" No IV.B3.RP-332 362c IV.B3-17(R-156) Core support barrel assembly: upper core barrel flangelower cylinder vertical (axial) welds Stainless steel Reactor coolant and neutron flux Loss of material due to wearCracking due to stress corrosion cracking or irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column) no Expansion componentsChapter XI.M2, "Water Chemistry" No IV.B3.RP-327329 IV.B3-15(R-155) Core support barrel assembly: upper cylinder (base metal and welds) and upper core support barrel flange weld (accessible surfaces)(flange base metal) Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Items IV.B3.RP-329, IV.B3.RP

-335, IV.B3.RP

-362, IV.B3.RP-363, IV.B3.RP

-364)and Chapter XI.M2, "Water Chemistry" No IV.B3.RP-357333 Incore instrumentation (ICI):

ICI thimble tubes

- lowerCore support barrel assembly: lower flange Zircaloy-4 Stainless steel Reactor coolant and neutron flux Loss of materialCracking due to wearstress corrosion cracking or fatigue A plant-specific aging management program is to be evaluatedChapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Yes, plant

-specificNo B-35 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-336328 IV.B3-22 15(R-170)155) Lower support structure: A286 fuel alignment pins (all plants with core shroud assembled in two vertical sections)Core support barrel assembly: lower core barrel flange weld Stainless steel Reactor coolant and neutron flux Cracking Loss of material due to wear; loss of fracture toughness due to neutr on irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxationcorrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column) no Expansion componentsChapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No IV.B3.RP-334332 IV.B3-23 17(R-167)156) Lower support structure: A286 fuel alignment pins (all plants with core shroud assembled with full-height shroud plates)Core support barrel assembly: upper core barrel flange Stainless steel Reactor coolant and neutron flux Cracking Loss of material due to irradiation

-assisted stress corrosion cracking and fatiguewear 'Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Existing Program components (identified in the "Structure and Components" column) no Expansion components No IV.B3.RP-364327 IV.B3-15(R-155) LowerCore support structure:barrel assembly: upper core support columnbarrel flange weld Cast austenitic stainlessStainless steel Reactor coolant and neutron flux Loss of fracture toughnessCracking due to neutron irradiation and thermal embrittlement stress corrosion cracking Chapter XI.M16A, ""PWR Vessel Internals" Expansion components (identified in the "Structure

" and Components" column)

(for Primary components see AMR Item IV.B3RP-327)Chapter XI.M2, "Water Chemistry" No IV.B3.RP-363 357 Lower support structure: core support columnIncoreinstruments (ICI): ICI thimble tubes - lower Stainless steel Zircaloy-4 Reactor coolant and neutron flux Loss of fracture toughnessmaterial due to neutron irradiation embrittlementwear Chapter XI.M16A, ""PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B3RP-327)" No B-36 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-330336 IV.B3-23 22(R-167)170) Lower support structure: core support column bolts (designs assembled in two vertical sections): fuel alignment pins Stainless steel Reactor coolant and neutron flux Loss of materialCracking due to wear; loss of fracture toughness due to neutron irradiation

-assisted embrittlement; loss of preload due to thermal and irradiation enhanced stress corrosion cracking and fatigue relaxation or creep Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item 'IV.B3.RP-314) No IV.B3.RP-331334 IV.B3-23(R-167) Lower support structure: core support column bolts (designs assembled in two vertical sections or with full-height shroud plates):

fuel alignment pins Stainless steel Reactor coolant and neutron flux Loss of fracture toughnessCracking due to neutron stress corrosion cracking, irradiation embrittlement-assisted stress corrosion cracking, or fatigue Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item 'IV.B3.RP-315)SCC mechanisms only)

No IV.B3.RP-335 334a IV.B3-23 22(R-167)170) Lower support structure: core support column welds, applicable to all plants except those (designs assembled in two vertical sections or with full-height shroud plates): fuel alignment pins Stainless steel Reactor coolant and neutron flux Cracking Loss of material due to stress corrosion cracking,wear; loss of fracture toughness due to neutron irradiation

-assisted stress corrosion cracking, embrittlement; loss of preload due to thermal and fatigue irradiation enhanced stress relaxation or creep Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B3.RP-327) No B-37 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-365364 Lower support structure: (all plants):

core support platecolumn welds Stainless steel (including CASS)Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement and for column welds made from CASS, thermal embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary component (identified in the "Structure and Components" column) no Expansion components No IV.B3.RP-343363 Lower support structure: (all plants):

core support plate (applicable to plants with a core support plate), if fatigue life cannot be demonstrated by TLAAcolumn welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, or fatigue Chapter XI.M2, "Water Chemistry", and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Primary components (identified in the "Structure and Components" column) no Expansion components Yes, evaluate t o determine the potential locations and extent of fatigue cracking No IV.B3.RP-390330 IV.B3-23(R-167) Lower support structure: core

support plate (applicable to plants with a core support plate), if fatigue analysis exists column bolts Stainless steel Reactor coolant and neutron flux Cumulative Cracking due to irradiation-assisted stress corrosion cracking or fatigue damage due to fatigue Fatigue is a time

-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Yes, TLAA No IV.B3.RP-342331 Lower support structure:

deep beams (applicable assemblies with full height shroud plates)core support column bolts Stainless steel Reactor coolant and neutron flux Cracking Loss of fracture toughness due to stress corrosion cracking, neutron irradiation

-assisted stress corrosion cracking, and fatigue embrittlement Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion components No B-38 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-366335 IV.B3-23(R-167) Lower support structure: deep beams (applicable assemblies (designs except those assembled with full

-height shroud plates)): lower core support beams Stainless steel Reactor coolant and neutron flux Loss of fracture toughnessCracking due to neutron irradiation embrittlement stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion componentsChapter XI.M2, "Water Chemistry" (for SCC mechanisms only) No IV.B3.RP-365

Lower support structure (designs with a core support plate): core support plate Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary component (identified in the "Structure and Components" column) no Expansion components No IV.B3.RP-24343 IV.B3-25(RP-24) Reactor vessel internal componentsLower support structure (designs with a core support plate): core support plate Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of materialCracking due to pitting and crevice corrosion fatigue Chapter XI.M2, "Water Chemistry" M16A, "PWR Vessel Internals" No IV.B3.RP-309342 Reactor vessel internal components (inaccessible locations)Lower support structure (designs with core shrouds assembled with full height shroud plates)

deep beams Stainless steel

nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking, and irradiation-assisted stress

-corrosion cracking , or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Yes, if accessible Primary, Expansion or Existing program components indicate aging effects that need management No B-39 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-311366 Reactor vessel internal components (inaccessible locations)Lower support structure (designs with core shrouds assembled with full height shroud plates): deep beams Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement

change in dimension due to void swelling

loss of preload due to thermal and irradiation enhanced stress relaxation; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Yes, if accessible Primary, Expansion or Existing program components indicate aging effects that need management No IV.B3.RP-339 IV.B3-24(R-53) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Cumulative fatigue damage due to fatigue Fatigue is a time-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes, TLAA IV.B3.RP-306 Reactor internal "No Additional Measures" componentsReactor vessel internal components with no additional measures Stainless steel; nickel alloy Reactor coolant and neutron flux No additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience existsCracking due to stress corrosion cracking, and irradiation

-assisted stress corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Note: Components with no additional measures are not uniquely identified in GALL tables - Components with no additional measures are defined in Section 3.3.1 of MRP-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluat ion Guidelines" No B-40 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-30724 IV.B3-25(RP-24)

Reactor vessel internal components with no additional measures Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of fracture toughnessmaterial due to neutron irradiation embrittlement; change in dimension due to void swelling; loss of preload due to thermal and irradiation enhanced stress relaxation; loss of material due to wear pitting and crevice corrosion Chapter XI.M16A, "PWR Vessel Internals" Note: Components with no additional measures are not uniquely identified in GALL tables - Components with no additional measures are defined in Section 3.3.1 of MRP-227, "Materials Reliability Program: Pressurized M2, "Water Reactor Inte rnals Inspection and Evaluation Guidelines"Chemistry" No IV.B3.RP-382 IV.B3-22(R-170) Reactor vessel internals:

ASME Section XI, Examination Category B-N-3 core support structure components (not already identified as "Existing Programs" components in MRP-227-A)

Stainless steel; nickel alloy

cast austenitic stainless steel Reactor coolant and neutron flux Cracking due to fatigue, stress corrosion cracking , or irradiation-assisted stress corrosion cracking; Loss loss of material due to wear Chapter XI.M1, "ASME Section XI Inservice Inspection, Subsecti ons IWB, IWC, and IWD" or Chapter XI.M16A, "PWR Vessel Internals," by invoking applicable 10 CFR 50.55a and ASME Section XI inservice inspection requirements No IV.B3.RP-338 Upper internals assembly

fuel alignment plate (applicable to plants (designs with core shrouds assembled with full height

shroud plates), if fatigue life cannot be demonstrated by TLAA): fuel alignment plate Stainless steel Reactor coolant and neutron flux Cracking due to fatigue

'Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) no Expansion components Yes, evaluate to determine the potential locations and extent of fatigue cracking No B-41 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B3.RP-391 400 Upper internals assembly: fuel alignment plate (applicable to plants with core shrouds assembled with full height shroud plates), if fatigue analysis exists Core Support Barrel Assembly: thermal shield positioning pins Stainless steel Reactor coolant and neutron flux Cumulative fatigue damage due to fatigueCracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue; loss of material due to wear Fatigue is a time

-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

Yes, TLAANo B-42 (4) Mark-up of changes to GALL Report Chapter IV.B4 B4. REACT OR VESSEL I N TERNAL S (P WR) - BA BCOCK A N D WILCOX Sy stems, Structures, and Components This section addresses t he Babcock and Wilcox (B&W) pressurized

-water reactor (PWR) vessel internals and consists , which consist of components in the plenum cover and plenum cylind e r assembly , the u pper grid assembly, the control rod g u ide tube (CRGT) assembly, the core support shield a sse mbly, the core barrel assembly, the lower grid assembly, and the flow distributor assembly. Based on Regulatory Guide 1.26, "Quality Group Classifications an d Standards for Water

-, Steam-, and Radioactive

-Waste-Containing Components of Nuclear Power Plants," all structures and co mponents that comprise the reactor vessel are governed by Group A or B Quality Standards.

incor e monitoring instrumentation (IMI) guide tube a ssembly, and the flow distributor assembly.

Common mi scellaneou s material/environment combinations where aging effects are not expected to degrade the ability of th e structure o r component to perform its intended f unction for the period of extended operation are included in IV.E. Sy stem Interfaces The systems that interfa c e with the r eactor vessel internals include the r eactor pressure vessel (IV.A2). Inspection Plan An applica n t will sub m it an inspection plan for reactor i n ternals to the NRC for review and approv al with the applicati on for license renewal in accordanc e with Chapter XI.M16 A , "P WR Vessel Internals."

B-43 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-242 IV.B4-4 (R-183) Control rod guide tube (CRGT) assembly:

accessible surfaces at four screw locations (every 90 degrees) for CRGT spacer castings Cast austenitic stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Items IV.B4.RP-253 and IV.B4.RP

-258) No IV.B4.RP-242a Control rod guide tube (CRGT) assembly: CRGT spacer castings Stainless steel (including CASS)

Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-245 IV.B4-13 (R-194) Core barrel assembly: (a) upper thermal shield bolts; (b) (applicable to Crystal River Unit 3 or Davis Besse only):

surveillance specimen holder tube bolts (Davis

-Besse, only); (c) surveillance specimen tube holder (SSHT) studs , and /nuts (Crystal River Unit 3, only)or bolts Stainless steel; nickelNickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Items IV.B4.RP-247 and IV.B4.RP

-248) No IV.B4.RP-245a Core barrel assembly (applicable to Crystal River Unit 3 or Davis Besse only): surveillance specimen holder tube (SSHT) stud or bolt locking devices Nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No B-44 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-245b Core barrel assembly (applicable to CR-3 or DB only): surveillance specimen holder tube (SSHT) stud or bolt locking devices Nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-247 IV.B4-13 (R-194) Core barrel assembly: accessible lower core barrel (LCB) bolts and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Items IV.B4.RP

-245, IV.B4.RP

-246, IV.B4.RP-254, and IV.B4.RP

-256) No IV.B4.RP-247a Core barrel assembly: lower core barrel (LCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-247b Core barrel assembly: lower core barrel (LCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-249 IV.B4-12 (R-196) Core barrel assembly:

baffle plate accessible surfaces within one inch around each baffle plate flow and bolt hole plates Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Item IV.B4.RP-250) No IV.B4.RP-249a Core barrel assembly:

baffle plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No B-45 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-241 IV.B4-7 (R-125) Core barrel assembly: baffle-to-former bolts and screws Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking, fatigue, and overload Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-241 241a IV.B4-7(R-125) Core barrel assembly: baffle/former assembly: (a) accessible baffle

-to-former bolts and screws; (b) accessible locking devices (including locking welds) of baffle-to-former bolts and internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking, irradiation-assisted stress -corrosion cracking , fatigue, and overload Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary Components (identified in the "Structure and Components" colu mn) and Chapter XI.M2, "Water Chemistry" (for Expansion components see AMR Items IV.B4.RP-244 and IV.B4.RP

-375)SCC mechanisms only)

No IV.B4.RP-240 IV.B4-1 (R-128) Core barrel assembly:

baffle/former assembly: (a) accessible baffle-to-former bolts and screws

(b) accessible locking devices (including welds) of baffle-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals."" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Item IV.B4.RP-243.) No IV.B4.RP-240a Core barrel assembly:

locking devices (including locking welds) of baffle-to-former bolts and internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No B-46 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-250 IV.B4-12 (R-196) Core barrel assembly: core barrel cylinder (including vertical and circumferential seam welds); former plates Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B4.RP-249) No IV.B4.RP-250a Core barrel assembly: core barrel cylinder (including vertical and circumferential seam welds); former plates Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-375 Core barrel assembly:

internal baffle-to-baffle

bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking, fatigue, or overload Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column) and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B4.RP-241SCC mechanisms only) No IV.B4.RP-375a Core barrel assembly:

internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-244 IV.B4-7 (R-125) Core barrel assembly; external baffle-to-baffle bolts and core barrel-to-former bolts; Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress corrosion cracking, fatigue, and overload Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No B-47 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-244 244a IV.B4-7(R-125) Core barrel assembly; (a) external baffle

-to-baffle bolts; (b) core barrel

-to-former bolts; (c)

locking devices (including welds) of external baffle-to-baffle bolts and core barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Cracking due to irradiation-assisted stress

-corrosion cracking , or fatigue Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, ""PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

" and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Item IV.B4.RP-241)SCC mechanisms only)

No IV.B4.RP-243 IV.B4-1 (R-128) Core barrel assembly; (a) external baffle

-to-baffle bolts; (b) core barrel

-to-former bolts; (c) locking devices (including welds) of: external baffle-to-baffle bolts and core barrel-to-former bolts

(d) internal baffle-to-baffle bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B4.RP-240) No IV.B4.RP-243a Core barrel assembly:

locking devices (including welds) of external baffle-to-baffle bolts and core barrel-to-former bolts Stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-248 IV.B4-12 (R-196) Core support shield (CSS) assembly:

accessible upper core barrel (UCB) bolts and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Items IV.B4.

RP-245, IV.B4.RP

-246, IV.B4.RP-254, IV.B4.RP

-247, and IV.B4.RP-256) No B-48 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-248a Core support shield (CSS) assembly: upper core barrel (UCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-248b Core support shield (CSS) assembly: upper core barrel (UCB) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-253 252 IV.B4-2116 (R-191)188) Core support shield (CSS) assembly: (a) CSS cast outlet nozzles (Oconee Unit 3 and Davis-Besse, only); (b)

CSS vent valve discs top and bottom retaining rings (valve body components)

Cast austenitic stainlessStainless steel , including CASS and PH steels Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging

embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Item IV.B4.RP-242) No IV.B4.RP-252 252a IV.B4-16 (R-188) Core support shield (CSS) assembly: (a) CSS vent valve disc shaft or hinge pin (b)

CSS vent valve top retaining ring (c) CSS vent valve and bottom retaining ringrings; vent valve locking devices (valve body components)Stainless steel Reactor coolant and neutron flux Loss of fracture toughness Cracking due to thermal aging embrittlement stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column) Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No Expansion components No IV.B4.RP-251 IV.B4-15 (R-190) Core support shield (CSS) assembly: CSS top flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear; loss of preload (wear) Chapter XI.M16A, "PWR Vessel Internals" No B-49 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-251 251a IV.B4-15 (R-190) Core support shield (CSS) assembly: CSS top flange; plenum Plenum cover assembly: plenum cover weldment rib pads

and plenum cover support flange Stainless steel Reactor coolant and neutron flux Loss of material due to wear

loss of preload (wear)

Chapter XI.M16A, "PWR Vessel Internals" Primary component (identified in the "Structure and Components" column)

No Expansion components No IV.B4.RP-256 IV.B4-25 (R-210) Flow distributor assembly: flow distributor bolts and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals

," Expansion components (identified in the "Structure and Components" column)" and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Items IV.B4.RP-247 and IV.B4.RP

-248) No IV.B4.RP-256a Flow distributor assembly: flow distributor bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-256b Flow distributor assembly: flow distributor bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to distortion or void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-259 IV.B4-31 (R-205) Incore Monitoring InstrumentationInstrument(IMI) guide tube assembly: accessible top surfaces of IMI guide tube spider-to-lower grid rib sectionssection welds Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of fracture toughness

due to thermal aging, neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see Item IV.B4.RP-260) No IV.B4.RP-259a Incore Monitoring Instrument (IMI) guide tube assembly: IMI guide tube spider-to-lower grid rib sections welds Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No.

B-50 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-258 IV.B4-4 (R-183) Incore Monitoring InstrumentationInstrument(IMI) guide tube assembly: accessible top surfaces of IMI Incore guide tube spider spiders (castings ) Cast austenitic stainless steel Reactor coolant and neutron flux Loss of fracture toughness due to thermal aging

, and neutron irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see Item IV.B4.RP-242) No IV.B4.RP-258a Incore Monitoring Instrumentation (IMI) guide tube assembly: IMI guide tube spiders Stainless steel Reactor coolant and neutron flux Cracking due to stress corrosion cracking, irradiation-assisted stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No IV.B4.RP-254 IV.B4-25 (R-210) Lower grid assembly: alloy X-750 lower grid shock pad bolts and locking devices (T hree M ile I sland Unit -1, only) Nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals

," Expansion components (identified in the "Structure and Components" column) " and Chapter XI.M2, "Water Chemistry" (for Primary components see AMR Items IV.B4.RP-247 and IV.B4.R P-248) No IV.B4.RP-254a Lower grid assembly: alloy X-750 lower grid shock pad bolt locking devices (Three Mile Island Unit 1, only) Nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-254b Lower grid assembly: alloy X-750 lower grid shock pad bolt locking devices (Three Mile Island Unit 1, only) Nickel Alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-246 IV.B4-12 (R-196) Lower grid assembly: upper thermal shield (UTS) bolts and lower thermal shield (LTS) bolts Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking

'Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Items IV.B4.RP-247 and IV.B4.RP

-248) No B-51 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-246a Lower grid assembly: upper thermal shield (UTS) bolt locking devices and lower thermal shield (LTS) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-246b Lower grid assembly: upper thermal shield (UTS) bolt locking devices and lower thermal shield (LTS) bolt locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to wear; changes in dimensions due to void swelling or distortion Chapter XI.M16A, "PWR Vessel Internals" No IV.B4.RP-260 IV.B4-31 (R-205) Lower grid fuel assembly: (a) accessible pads; (b) accessible pad-to-rib section welds; (c) accessible alloy X-750 dowels, cap screws and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of fracture toughness due to neutron

irradiation embrittlement Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B4.RP-259) No IV.B4.RP-260a Lower grid fuel assembly: (a) pads; (b) pad-to-rib section welds; (c) alloy X-750 dowels, cap screws and locking devices Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking or fatigue Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" (for SCC mechanisms only)

No IV.B4.RP-262 IV.B4-32 (R-203) Lower grid assembly: accessible alloy X-750 dowel-to-lower fuel assembly support pad locking welds Nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B4.RP-261) No IV.B4.RP-261 IV.B4-32 (R-203) Lower grid assembly: alloy X-750 dowel-to-guide block welds Nickel alloy Reactor coolant and neutron flux Cracking due to stress

-corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Items IV.B4.RP

-262 and IV.B4.RP

-352)and Chapter XI.M2, "Water Chemistry" No B-52 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.R-53 IV.B4-37 (R-53) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Cumulative fatigue damage due to fatigue Fatigue is a time-limited aging analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes, TLAA IV.B4.RP-24 IV.B4-38 (RP-24) Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Loss of material due to pitting and crevice corrosion Chapter XI.M2, "Water Chemistry" No IV.B4.RP-376 Reactor vessel internal components Stainless steel; nickel alloy Reactor coolant and neutron flux Reduction in ductility and fracture

toughness due to neutron irradiation Ductility - Reduction in Fracture Toughness is a TLAA (BAW-2248A) to be evaluated for the period of extended

operation.

See the SRP, Section 4.7, "Other Plant-Specific TLAAs," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

Yes, TLAA IV.B4.RP-238 382 IV.B4-42 (R-179) Reactor vessel internal internals: ASME Section XI, Examination Category B-N-3 core support structure components (inaccessible locations)

Stainless steel; nickel alloy Reactor coolant and neutron flux Cracking due to fatigue, stress corrosion cracking, and or irradiation

- assisted stress corrosion cracking

loss of material due to wear Chapter XI.M2, "Water Chemistry,"

M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" or Chapter XI.M16A, "PWR Vessel Internals"," by invoking applicable 10 CFR 50.55a and ASME Section XI inservice inspection requirements Yes, if accessible Primary, Expansion or Existing program components indicate aging effects that need management No B-53 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-239 352 Upper grid assembly: alloy X-750 dowel-to-upper fuel assembly support pad welds (all plants except Davis-Besse)Reactor vessel internal components (inaccessible locations)

Stainless steel; nickelNickel alloy Reactor coolant and neutron flux Loss of fracture toughnessCracking due to neutron irradiation embrittlement; change in dimension due to void swelling; loss of preload due to thermal and irradiation enhanced stress relaxation; loss of material due to wear-corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" Yes, if accessible Primary, Expansion or Existing program components indicate aging effects that need manag ement No IV.B4.RP-236 Reactor internal "No Additional Measures" componentsReactor vessel internal components with no additional measures Stainless steel; nickel alloy Reactor coolant and neutron flux No additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience existsCracking due to stress corrosion cracking , and irradiation

-assisted stress corrosion cracking Chapter XI.M2, "Water Chemistry" and Chapter XI.M16A, "PWR Vessel Internals" Note: Components with no additional measures are not uniquely identifies in GALL tables

- Components with no additional measures are defined in Section 3.3.1 of MRP

-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" No IV.B4.RP-400 Core support shield assembly: upper (top) flange weld Stainless steel Reactor coolant and neutron flux Cracking due to stress-corrosion cracking Chapter XI.M16A, "PWR Vessel Internals" and Chapter XI.M2, "Water Chemistry" No B-54 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Item Link Structure and/or Component Material Environment Aging Effect/ Mechanism Aging Management Program (AMP) Further Evaluation IV.B4.RP-237401 Reactor vessel internal components with no additional measures Core support shield assembly: upper (top) flange weld Stainless steel

nickel alloy Reactor coolant and neutron flux Loss of fracture toughness due to neutron irradiation embrittlement

change in dimension due to void swelling; loss of preload due to thermal and irradiation enhanced stress relaxation; loss of material due to wear Chapter XI.M16A, "PWR Vessel Internals" Note

Components with no additional measures are not uniquely identified in GALL tables

- Components with no additional measures are defined in Section 3.3.1 of MRP

-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" No IV.B4.RP-382 IV.B4-42(R-179) Reactor vessel internals: core support structure Stainless steel; nickel alloy; cast austenitic stainless steel Reactor coolant and neutron flux Cracking, or Loss of material due to wear Chapter XI.M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" No IV.B4.RP-352 Upper grid assembly: alloy X-750 dowel-to-upper fuel assembly support pad welds (all plants except Davis

-Besse) Nickel alloy Reactor coolant and neutron flux Cracking due to stress corrosion cracking Chapter XI.M2, "Water Chemistry," and Chapter XI.M16A, "PWR Vessel Internals" Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B4.RP-261) No B-55 (5) Mark-up of changes to GALL Report Chapter IX

.C and IX.G IX.C Selected Definitions & Use of Terms for Des c ribing and Standardizing MATE RIAL S Stainless st eel Products gr ouped under the term "stainless stee l" include wrought or forged auste nitic, ferrit i c, martensitic, precipitation

-hardened (PH), or duplex stainless steel (Cr content >11

%). These stainless stee ls may be fabricated using a wro ught or cast process. Th ese materials are susceptible t o a variety o f aging effects and mechanisms, including loss of material due to pittin g and crevice corrosion, a nd crackin g due to stress corrosion cracking.

In some cases, when the recommen ded AMP an aging effect is app licable to all of the same for PH variou s stainless ste e l or cast cat egories, it can be assu med that the term "stainless steel" in the "Material" colum n of an AMR line-item in the GALL Report encompasse s all stainless ste e l types. Cast austenit i c stainless ste e l (CASS) as f o r stainless steel, PH stainless steel or CASS are included as a part of the stainle s s steel classif i cation. However, C ASS is quite suscept ible to loss of fra c ture toughness d ue to thermal and neutr on irradiatio n embrittleme n t. Therefore, wh en this aging effect is being considered, CASS In ad dition, MRP-227-A indica tes that PH stainless steels or m a rtensitic st ainless steels may be susceptible t o loss of fra c ture toughn ess by a thermal aging mechanism. Therefore, when loss of fracture tou ghness due to thermal and neutron irradiation embrittlement is an applicable a g ing effect a nd mechanism for a component in the GALL Report, the CASS, PH sta i nless steel, or martensitic stainless ste e l designat ion is spe c ifi c ally identified de signated in an AMR line-item.

Steel with st ainless steel cladding a l so may be co nsidered stainless ste e l when the aging effect is asso ciate d with the stainless ste e l surface of the material, rather than the composite volume of the material.

Exa m ples of stainless st eel designat ions that co mprise this category include A-286, SA193-Gr.

B8, SA193-Gr. B8M, Gr. 660 (A-286), SA193-6, SA193-Gr. B8 or B-8M, SA453, Type 416, Type 403, 410, 420 , and Types 431 martensitic stainless ste e ls, Type 15-5, 17-4, and 13-8-Mo PH stainless steels, and SA-193, Gra de B8 and B8M bolting materials.

Exa m ples of wrought austenitic stain l ess materia l s that comprise this category include Type 304, 304NG, 304L, 308, 308L, 3 09, 309L, 31 6 , and 347 , 403, and 41 6.. Exa m ples of CASS designations that comprise th is category

B-56 include CF-3, -8, -3M, CF3, CF3M, CF8 and -8M.CF8 M. [Ref. 6, 7

], 30] IX.G References

30. Welding Handbook, Seventh Edition, Volume 4, Metals and Their Welda b ility, American Welding So ciety, 1984, p.76-145.

B-57 Appendix B, Section 2 - Mark-up of Changes to the SRP-LR In the mark-up, red or green strikethr ough text indicates a de letion and blue underline text indicates an insertion.

Green text i ndicates a move, where a double strikethrough indicates th e original lo ca tion of the te xt and a double underlin e indicate s t he final lo ca tion of the moved text.

(1) Mark-up of changes to S R P-LR Tabl e 3.0-1 Ta ble 3.0-1 FSA R Supple m ent for A g ing M a na ge me nt of A p plic a b le Sy s t e m s G A LL Chapter G A LL Progra m De sc ription of Progra m Imple m e n ta tion Sc he dule A p plicable GA L L Re port a nd S R P-LR Chapter Refer e nce s X I.M16A PWR Vessel Internals The program relie s on impl ementation of the inspe c t i on and eval u a tion guidelin es in EPRI Tech nical Rep o rt No. 101 659 6 1022 863 (MRP-227-A) and EPRI Te chni cal Repo rt No. 1016 609 (MRP-228) to ma nage the aging effe cts on the rea c to r vessel internal com p onent s. This prog ram i s use d to mana ge (a) var i ous for m s of cra c king, in cl uding st re ss cor r o s io n cra c kingS C C , primary wate r stre ss cor r o s io n cr a cki ngP WS C C , irradiatio n-as sist e d st re s s co rro sio n c r ac kin g (IASCC), or an d crackin g d ue to fatigue/cycli c al loading; (b) loss of material in du ced by wear; (c) loss of fractu re toug hne ss d ue to either thermal aging or , neutro n irradiatio n embrittleme n t , or void swell i ng; (d) dimen s ion a l chang es a nd p o tential loss of fractu re toughn ess due to void swelling an d irra diation g r o w th or distortio n; an d (e) lo ss of preloa d due to thermal an d irra diation-e nhan ce d stre ss relaxat i on or cre ep. Program sho u ld be implem ent ed prio r to perio d of extended operation GALL IV / SRP 3.1 (2) Mark-up of changes to S R P-LR Secti on 3.1.2, "Acceptance C r iteria" 3.1.2.2.9 Re m o ved as a result of LR-ISG-201 1-04 Cracking due to Stress Corro sion Cracking and Irradiation

-Assisted Stress Co rrosion Cra cking Cracking du e to SCC an d irradiation

-assisted str e ss corros i o n cracking (I ASCC) could occur in inaccessible location s fo r stainless st eel and nickel

-alloy Primary and Exp ansion PWR reactor v e ssel inter nal components. If agin g effects are identified in accessible locations, th e GALL Report recommends furt her evaluation of the aging effects in inaccessible location s on a plant

-specific basis to ensure t hat this agin g effect is a dequately managed. Acc eptance crit eria are described in Branch Technical Position RLSB

-1 (Appendix A.

1 of this SRP

-LR).

B-58 3.1.2.2.10 Re m o ved as a result of LR-ISG-201 1-04 Loss of Fracture Toughness due to Neutron Irradiation Embrittlement; Change in Dimension due to Void Sw ell i n g; Loss of Preload due to Stress Relaxation; or Loss of Material du e to Wear Loss of fract u re toughness due to ne utron irradia t ion embrittlement, change in dimension due to void swellin g, loss of pr eload due to stress rel a xation, or loss of materia l due to wear could occur in inacce ssible locatio n s for stainle s s steel and nickel-a lloy Primary and Expansion PWR reactor vessel inter nal components. If agin g effects are identified in accessible locations, th e GALL Report recommends furt her evaluation of the aging effects in inaccessible location s on a plant

-specific basis to ensure t hat this agin g effe ct is a dequately managed. Acceptance crit eria are described in Branch Technical Position RLSB

-1 (Appendix A.

1 of this SRP

-LR). 3.1.2.2.12 Re m o ved as a result of LR-ISG-201 1-04 Cracking due to Fatigue EPRI 10165 96, Materials Reliability Progra m: Pressurized W a ter Reactor Internals In spection and Evaluation Guidelin es (MRP-22 7-Rev. 0) id entifies cracking due to f a tigue as an aging effect that can occur for the lo wer flange weld in the core support barrel asse mbly, fuel alignment plate in the upper internals a s sembly, and core suppor t plate lower support stru cture in PW R internals designed by Combustion Engineering. The GALL Report recommends that inspect i on for cracking in t h is compon ent be perfo rmed if acce ptable fatigu e life cannot be demonstrated by TLAA through the perio d of extended operation as defined in 10 CFR 5 4.3. 3.1.2.2.13 Re m o ved as a result of LR-ISG-201 1-04 Cracking due to Stress Corro sion Cracking and Fatigue Cracking du e to stress corrosion cra cking and fa tigue could occur in nickel alloy cont rol rod guide tube assemblies, guid e tube support pins expose d to reacto r coolant, an d neutron flu

x. The GALL Report, AMR Ite m IV.B2.RP-35 5, recomme nds further evaluation of a plant

-spe c ific AMP to ensure this aging effe ct is adequa tely manage

d. Acceptan ce criteria ar e described in Branch Technical P o sition RLS B-1 (Appendix A.1 of this SRP

-LR). 3.1.2.2.14 Re m o ved as a result of LR-ISG-201 1-04 Loss of Material du e to Wear Loss of material due to wear could occur in nickel alloy cont rol rod guide tube assemblies, guid e tube support pins and in Zircaloy-4 in core instrumentation low e r thimble tubes exposed to reactor coolant, and neutron flux. The GALL Report, AMR Items IV.

B 2.RP-356 and IV.B3.RP-357, recommend s further evaluation of a plant-specific AMP to en sure this ag ing effect is adequately managed. Acceptance criteria are d e scribed in Branch Technical Position RLSB

-1 (Appendix A.

1 of this SRP

-LR). (3) Mark-up of changes to S R P-LR Secti on 3.1.3, "Review Procedures

" 3.1.3.2.9 Re m o ved as a result of LR-ISG-201 1-04 Cracking due to Stress Corro sion Cracking and Irradiation

-Assisted Stress Co rrosion Cra cking The GALL Report recommends further evaluation of crackin g due to SCC and IASCC for inaccessible location s fo r P rimary an d E xpansion PWR reactor vessel int e rnal components if aging effect s are identif ied for these components in acce ssible location

s. The reviewe r reviews the applican t's proposed program on a case

-by-case basis to ensure that an adequate program B-59 will be in pla c e for the management of these agi ng effects consi stent wit h the action item documented in the staff

's safety evaluation for MRP

-227, Re vision 0.. 3.1.3.2.10 Re m o ved as a result of LR-ISG-201 1-04 Loss of Fracture Toughness due to Neutron Irradiation Embrittlement; Change in Dimension due to Void Sw ell i n g; Loss of Preload due to Stress Rel axation; or Loss of Material due to Wear The GALL Report recommends further evaluation of loss of f r acture toug hness due t o neutron irradiation e m brittlement, change in dimension due to void swelling, loss of preload d ue to stress relaxation, or loss of mat e rial due to wear for inaccessible lo cat ions for P rimary and E xpansion PWR reactor vessel inte rnal components , if agin g effects are identified fo r these components in access ible l o cations. Th e reviewer reviews the applicant's pr oposed prog ram on a case

-by-case basis to en sure that an adequate pr ogram will be in place fo r the management of these aging effects con s istent with th e action ite m document ed in the sta ff's safety evaluation for MRP-227, Revision 0

. 3.1.3.2.12 Re m o ved as a result of LR-ISG-201 1-04 Cracking due to Fatigue The GALL Report recommends further evaluation of crackin g due to fatigue in the lo wer flange weld in the core support barrel asse mbly , fuel alignment plate in the u pp er internals assembly , and core su pport plate in the l ower support stru cture in PW R intern als designed by Combustion Engineering

. The reviewer determines whether a TLAA has been performed for each component, consiste nt with the actio n item docum ented in the staff's saf e ty evaluation for MRP

-227, Revision 0

. If a TLAA has not b een performed, the reviewer determi nes whether the applicant ha s performed an evaluation to identif y the potential location a nd extent of fatigue cracking f o r each component consist ent with the action item documented in the staff

's safety evaluation for MRP

-227 , Revision 0.

3.1.3.2.13 Re m o ved as a result of LR-ISG-201 1-04 Cracking due to Stress Corro sion Cracking and Fatigue The GALL Report recommends further evaluation of crackin g due to stre ss corrosion cracking and fatigue in the nicke l alloy control rod guide tu be assemblies, guide tu be support p i ns exposed to reactor coola n t, and neut ron flux. The reviewer re views the applicant's pro posed program on a case

-by-case basi s to ensure that an adequate program will be in pla c e for the managemen t of these ag ing effects co nsistent wit h the action item documented in the staff's safety evaluation for MRP

-227, Revi sion 0. 3.1.3.2.14 Re m o ved as a result of LR-ISG-201 1-04 Loss of Material du e to Wear The GALL Report recommends further evaluation of loss of material due to wear in nickel a lloy control rod g u ide tube assemblies, g u ide tube su pport pins a nd in Zircalo y-4 incore instrumentation lower thimble tubes exposed to reactor coola n t, and neut ron f lux. The reviewer reviews the applicant's p r oposed pro g ram on a case

-by-case basis to en sure that an adequate program will be in place for the management of these aging effects con s istent with th e action item documented in the staff's safety evaluation for MRP-227 , Revision 0

.

B-60 (4) Mark-up of changes to S R P-LR Tabl e 3.1-1 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 3 BW R/ PWR Stainl ess steel or nicke l allo y r eactor ve ssel inter nal compo nents e x pose d to reactor coo l ant and ne utron flux Cumul a tive fati gue d a ma ge due to fatig u e F a tigue is a T L AA eval uate d for the per iod of ext end ed o per ation (Se e SRP, Section 4.3 "Metal F a tigue," for ac ceptab le methods to co mpl y w i th 10 CF R 54.2 1 (c)(1) Yes, T L AA (See subsecti on 3.1.2.2.1) IV.B1.R-53 IV.B2.RP-303 IV.B3.RP-339 IV.B4.R-53 IV.B3.RP-389 IV.B3.RP-390 IV.B3.RP-391 IV.B1-14 (R-5 3) IV.B2-31 (R-5 3) IV.B3-24 (R-5 3) IV.B4-37 (R-5 3) N/A N/A N/A 15 PW R Stainl ess steel Babcock &

Wilcox (including CASS, martensitic SS, and PH SS) and n i ckel all o y react o r vessel i n terna l compo nents expos ed to rea c tor coola n t and n eutro n flu x Red u ction in of ductilit y a nd fracture toug hn ess due to neutro n irrad i at ion embrittlem ent, and for CASS, marten sitic SS, and PH SS due to thermal agi ng embrittlem ent Ductilit y - Re du ction in fF racture tT oughn ess is a T L AA to be evalu a ted for the peri od of e x te nde d oper ation., See the SRP, Section 4.7, "Other Plant-Specific T L AAs," for accepta b l e methods for of meetin g the re quir e ments of 10 CF R 54.2 1 (c)(1).). Yes, T L AA (See subsecti on 3.1.2.2.3.3) IV.B4.RP-376 N/A 23 PWR Stainl ess steel or nicke l allo y PWR re a c tor vessel intern al comp o nents (inacc e ssible locations) expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosio n crack i ng, an d irradi atio n-assisted stress corrosion crack i ng Cha p ter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" Yes, if accessible Pri m a r y , Ex pan si on o r Existi ng pr ogra m compo nents i n dicate agi ng effects that need mana geme n t (See subsecti on 3.1.2.2.9) IV.B2.RP-268 IV.B3.RP-309 IV.B4.RP-238 N/A N/A N/A 24 PWR Stainl ess steel or nicke l allo y PWR re a c tor vessel intern al comp o nents (inacc e ssible locations) expos ed to rea c tor coola n t and n eutro n flu x Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent; or chan ges i n dime nsio n du e to void s w e l l i n g; or los s of preloa d due to therm a l and irradi atio n en h ance d stress rela xation; or l o ss of material due to w e ar Chapter X I.M16A, "PWR Vessel Internals" Yes, if accessible Pri m a r y , Ex pan si on o r Existi ng pr ogra m compo nents i n dicate agi ng effects that need mana geme n t (See subsecti on 3.1.2.2.10) IV.B2.RP-269 IV.B3.RP-311 IV.B4.RP-239 N/A N/A N/A B-61 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 26 PWR Stainl ess steel Comb ustion Engi neer in g co re supp ort barrel assembly

lo w e r flang e w e l d e x pose d to reactor coo l ant and ne utron flux; U pper i n te rnals assembly: fuel alig nme n t plate (a ppl ica b l e to plants w i t h core shr o uds assemb led w i t h full he ig ht shrou d plat es) expos ed to reactor coo l ant and ne utron flux; Lo w e r su pport structure: core supp ort plate (a ppl ica b l e to plants w i t h a core su p port plate) expos ed to rea c tor coola n t and n eutro n flu x Cracking due to fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y," if fatigu e life cann ot be confi rmed b y TL A A Yes, evaluate to determi ne the potenti a l locati ons a nd e x te nt of fatigue crack i n g (See subsecti on 3.1.2.2.12) IV.B3.RP-333 IV.B3.RP-338 IV.B3.RP-343 N/A 27 PWR Nickel alloy W e stinghouse control ro d gui de tube assemb lies, gu ide tub e supp ort pins e x pose d to reactor coo l ant and ne utron flux Crackin g du e to stress corrosio n crack i ng a nd fatigue A plant-sp ecific agin g mana geme n t p r ogram is to be eval uate d Yes, plant

-sp e c ific (See subsecti on 3.1.2.2.13) IV.B2.RP-355 N/A 28 PWR Nickel alloy W e stinghouse control ro d gui de tube assemb lies, gu ide tub e supp ort pins, a nd Z i rcal o y-4 Comb ustion Engi neer in g incor e instrum entatio n thimbl e tubes e x p o se d to reactor coo l ant and ne utron flux Loss of materi al du e to w e ar A plant-sp ecific agin g mana geme n t p r ogram is to be eval uate d Yes, plant

-sp e c ific (See subsecti on 3.1.2.2.14) IV.B2.RP-356 IV.B3.RP-357 N/A N/A 28 PWR Stainl ess steel Comb ustion Engi neer in g "Existi ng Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Loss of materi al du e to w e ar; cracki ng due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-400 N/A B-62 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 32 PW R Stainl ess steel, nickel a llo y, or CASS react o r vessel intern als, core supp ort structure (n o t al re ady referenc ed as ASME Section XI Exa m inati on Categ o r y B-N-3 core su p port structure comp one nts in MRP-22 7-A), exp o se d to reactor coo l ant and ne utron flux Crackin g , or lo ss of material due to w e ar Chapter X I.M1, "ASME Section XI Inse rvice Inspection, Subsections IW B, IW C, and IW D" or Chapter X I.M16A, "PWR Vessel Internals," invoking app lica b le 10 CF R 50.55 a and ASME Sec t ion XI inservic e ins p e c tion requ ireme n ts for these compo nents No IV.B2.RP-382 IV.B3.RP-382 IV.B4.RP-382 IV.B2-26 (R-1 4 2) IV.B3-22 (R-1 7 0) IV.B4-42 (R-1 7 9) 51 PWR Stainl ess steel or nicke l-allo y B abcock & W ilcox reactor inter nal compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assisted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" No IV.B4.RP-236 IV.B4.RP-241 IV.B4.RP-244 IV.B4.RP-245 IV.B4.RP-246 IV.B4.RP-247 IV.B4.RP-248 IV.B4.RP-254 IV.B4.RP-256 IV.B4.RP-261 IV.B4.RP-262 IV.B4.RP-352 IV.B4.RP-375 N/A IV.B4-7(R-125) IV.B4-7(R-125) IV.B4-13(R-194) IV.B4-12(R-196) IV.B4-13(R-194) IV.B4-12(R-196) IV.B4-25(R-210) IV.B4-25(R-210) IV.B4-32(R-203) IV.B4-32(R-203) N/A N/A 52 PWR Stainl ess steel or nicke l-allo y C o mb usti on Engi neer in g re actor intern al compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assisted stress corrosio n crack i ng, or fatigue Cha p ter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" No IV.B3.RP-306 IV.B3.RP-312 IV.B3.RP-313 IV.B3.RP-314 IV.B3.RP-316 IV.B3.RP-320 IV.B3.RP-322 IV.B3.RP-323 IV.B3.RP-324 IV.B3.RP-325 IV.B3.RP-327 IV.B3.RP-328 IV.B3.RP-329 IV.B3.RP-330 IV.B3.RP-334 IV.B3.RP-335 IV.B3.RP-342 IV.B3.RP-358 N/A IV.B3-2(R-149) N/A IV.B3-9(R-162) IV.B3-9(R-162) IV.B3-9(R-162) N/A N/A N/A N/A IV.B3-15(R-155) IV.B3-15(R-155) IV.B3-15(R-155) IV.B3-23(R-167) IV.B3-23(R-167) IV.B3-23(R-167) N/A N/A B-63 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 53 PWR Stainl ess steel or nicke l-allo y W e sti ngh ouse re actor intern al comp o nents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assisted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Internal s," and Chapter X I.M2, "Water Chemistr y" No IV.B2.RP-265 IV.B2.RP-271 IV.B2.RP-273 IV.B2.RP-275 IV.B2.RP-276 IV.B2.RP-278 IV.B2.RP-280 IV.B2.RP-282 IV.B2.RP-286 IV.B2.RP-289 IV.B2.RP-291 IV.B2.RP-293 IV.B2.RP-294 IV.B2.RP-298 IV.B2.RP-301 IV.B2.RP-346 IV.B2.RP-387 N/A IV.B2-10(R-125) IV.B2-10(R-125) IV.B2-6(R-128) IV.B2-8(R-120) IV.B2-8(R-120) IV.B2-8(R-120) IV.B2-8(R-120) IV.B2-16(R-133) IV.B2-20(R-130) IV.B2-24(R-138) IV.B2-24(R-138) IV.B2-24(R-138) IV.B2-28(R-118) IV.B2-40(R-112) N/A N/A 51a PWR Stainl ess steel or nicke l allo y B abcock & W ilcox reactor inter nal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B4.RP-241 IV.B4.RP-241a IV.B4.RP-242a IV.B4.RP-247 IV.B4.RP-247a IV.B4.RP-248 IV.B4.RP-248a IV.B4.RP-249a IV.B4.RP-252a IV.B4.RP-256 IV.B4.RP-256a IV.B4.RP-258a IV.B4.RP-259a IV.B4.RP-261 IV.B4.RP-400 IV.B4-7 (R

-125) N/A N/A IV.B4-13 (R-194) N/A IV.B4-25 (R-210) N/A N/A N/A IV.B4-25 (R-210) N/A N/A N/A IV.B4-32 (R-203) N/A 51b PWR Stainl ess steel or nicke l allo y B abcock & W ilcox reactor inter nal "Exp ansi on" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, fatigu e, or overl oad Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B4.RP-244 IV.B4.RP-244a IV.B4.RP-245 IV.B4.RP-245a IV.B4.RP-246 IV.B4.RP-246a IV.B4.RP-254 IV.B4.RP-254a IV.B4-7 (R

-125) N/A IV.B4-13 (R-194) N/A IV.B4-12 (R-196) N/A IV.B4-25 (R-210) N/A B-64 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item IV.B4.RP-260a IV.B4.RP-262 IV.B4.RP-352 IV.B4.RP-250a IV.B4.RP-375 N/A IV.B4-32 (R-203) N/A N/A N/A 52a PWR Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g re actor intern al "Primar y" com ponents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-312 IV.B3.RP-314 IV.B3.RP-322 IV.B3.RP-324 IV.B3.RP-326a IV.B3.RP-327 IV.B3.RP-328 IV.B3.RP-342 IV.B3.RP-358 IV.B3.RP-362a IV.B3.RP-363 IV.B3.RP-338 IV.B3.RP-343 IV.B3-2 (R

-149) IV.B3-9 (R

-162) N/A N/A N/A IV.B3-15 (R-155) IV.B3-15 (R-155) N/A N/A N/A N/A N/A N/A 52b PWR Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g re actor intern al "Exp ans ion" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-313 IV.B3.RP-316 IV.B3.RP-323 IV.B3.RP-325 IV.B3.RP-329 IV.B3.RP-330 IV.B3.RP-333 IV.B3.RP-335 IV.B3.RP-362c NA IV.B3-9 (R

-162) N/A N/A IV.B3-12 (R-155) IV.B3-23 (R-167) N/A IV.B3-23 (R-167) N/A 52c PWR Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g reactor intern al "E xisti ng Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B3.RP-320 IV.B3.RP-334 IV.B3-9 (R

-162) IV.B3-23 (R-167)

B-65 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 53a PWR Stainl ess steel or nicke l allo y W e sti ngh ouse reactor inter nal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B2.RP-270a IV.B2.RP-271 IV.B2.RP-275 IV.B2.RP-276 IV.B2.RP-280 IV.B2.RP-298 IV.B2.RP-302 IV.B2.RP-387 N/A IV.B2-10 (R-125) IV.B2-6 (R

-128) IV.B2-8 (R

-120) IV.B2-8 (R

-120) IV.B2-28 (R-118) N/A N/A 53b PWR Stainl ess steel W e stingh ouse reactor intern al "E xpa n s ion" compo nents e x pose d to reactor coo l ant and ne utron flux Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B2.RP-273 IV.B2.RP-278 IV.B2.RP-286 IV.B2.RP-291 IV.B2.RP-291a IV.B2.RP-291b IV.B2.RP-293 IV.B2.RP-294 IV.B2.RP-387a IV.B2-10 (R-125) IV.B2-8 (R

-120) IV.B2-16 (R-133) IV.B2-24 (R-138) N/A N/A IV.B2-24 (R-138) IV.B2-24 (R-138) N/A 53c PWR Stainl ess steel or nicke l allo y W e sti ngh ouse reactor inter nal "Existin g Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x Cracking due to stress corrosion crack i ng, irradi atio n-assi sted stress corrosio n crack i ng, or fatigue Chapter X I.M16A, "PWR Vessel Intern al s," and Chapter X I.M2, "Water Chemistr y" (for SCC mechanisms only

) No IV.B2.RP-289 IV.B2.RP-301 IV.B2.RP-345 IV.B2.RP-346 IV.B2.RP-399 IV.B2.RP-355 IV.B2-20 (R-130) IV.B2-40 (R-112) N/A N/A N/A N/A 54 PW R Stainl ess steel bottom mounte d instru ment sy stem flux thimble tubes (w ith or w i t h o u t ch ro me pl a t i ng) e x po sed to reactor cool ant and neutro n flu x (W esting ho use "Ex i sting Progr a ms" compo nents) Loss of materi al du e to we a r Chapter X I.M16A, "PWR Vessel Internals,"

and or Chapter X I.M37, ""Flu x T h imble T ube Inspecti on"" No IV.B2.RP-284 IV.B2-12(R-143) IV.B2-13 (R-145) 55 PWR Stainl ess steel thermal shield assem b ly , thermal shiel d fle x ur es expos ed to reactor coo l ant and ne utron flux Crackin g du e to fatigue; Loss of materi al du e to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-302 N/A 55a PWR Stainl ess steel or nicke l allo y B abcock and W ilc o x reactor inter nal "No Additi ona l Mea s ures" N o a d d i ti on al ag i n g mana geme n t for reactor intern al "No Ad ditio nal Measur es" co mpon ents Chapter X I.M16A, "PWR Vessel Internals" No IV.B4.RP-236 NA B-66 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item compo nents e x pose d to reactor coo l ant and ne utron flux unl ess requ ire d b y ASME Section XI, Examin ation Categ o r y B-N-3 or relev ant oper ating e x pe rienc e inval i d a tes MR P-227-A. 55b PWR Stainl ess steel or nicke l allo y C o mb usti on Engi neer in g re actor intern al "No Add i tion al Measur es" compo nents e x pose d to reactor coo l ant and ne utron flux N o a d d i ti on al ag i n g mana geme n t for reactor intern al "No Ad ditio nal Measur es" co mpon ents unl ess requ ire d b y ASME Section XI, Examin ation Categ o r y B-N-3 or relev ant oper ating e x pe rienc e inval i d a tes MR P-227-A. Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-306 NA 55c PWR Stainl ess steel or nicke l allo y W e sti ngh ouse reactor inter nal "No Additi ona l Mea s ures" compo nents e x pose d to reactor coo l ant and ne utron flux N o a d d i ti on al ag i n g mana geme n t for reactor intern al "No Ad ditio nal Measur es" co mpon ents unl ess requ ire d b y ASME Section XI, Examin ation Categ o r y B-N-3 or relev ant oper ating e x pe rienc e inval i d a tes MR P-227-A. Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-265 NA 56 PWR Stainl ess steel or nicke l-allo y C o mb usti on Engi neer in g re actor intern al compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent; or chan ges i n dime nsio n du e to void s w e l l i n g; or los s of preloa d due to therm a l and irradi atio n en h ance d stress rela xation; or l o ss of material due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-307 IV.B3.RP-315 IV.B3.RP-317 IV.B3.RP-318 IV.B3.RP-319 IV.B3.RP-326 IV.B3.RP-331 IV.B3.RP-332 IV.B3.RP-336 IV.B3.RP-359 IV.B3.RP-360 IV.B3.RP-361 IV.B3.RP-362 IV.B3.RP-363 IV.B3.RP-364 IV.B3.RP-365 N/A IV.B3-7(R-165) IV.B3-7(R-165) IV.B4-8(R-163) IV.B3-9(R-162) N/A N/A IV.B3-17(R-156) IV.B3-22(R-170) N/A N/A N/A N/A N/A N/A N/A B-67 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item IV.B3.RP-366 N/A 58 PWR Stainl ess steel or nicke l-allo y B abcock & W ilcox reactor inter nal compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent; or chan ges i n dime nsio n du e to void s w e l l i n g; or los s of preloa d due to therm a l and irradi atio n en h ance d stress rela xation; or l o ss of material due to w e ar Chapter X I.M16A, "PWR Ves sel Internals" No IV.B4.RP-237 IV.B4.RP-240 IV.B4.RP-242 IV.B4.RP-243 IV.B4.RP-249 IV.B4.RP-250 IV.B4.RP-251 IV.B4.RP-252 IV.B4.RP-253 IV.B4.RP-258 IV.B4.RP-259 IV.B4.RP-260 N/A IV.B4-1(R-128) IV.B4-4(R-183) IV.B4-1(R-128) IV.B4-12(R-196) IV.B4-12(R-196) IV.B4-15(R-190) IV.B4-16(R-188) IV.B4-21(R-191) IV.B4-4(R-183) IV.B4-31(R-205) IV.B4-31(R-205) 59 PWR Stainl ess steel or nicke l-allo y W e sti ngh ouse re actor intern al comp o nents expos ed to rea c tor coola n t and n eutro n flu x Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent; or chan ges i n dime nsio n du e to void s w e l l i n g; or los s of preloa d due to therm a l and irradi atio n en h ance d stress rela xation; or l o ss of material due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-267 IV.B2.RP-270 IV.B2.RP-272 IV.B2.RP-274 IV.B2.RP-281 IV.B2.RP-285 IV.B2.RP-287 IV.B2.RP-288 IV.B2.RP-290 IV.B2.RP-292 IV.B2.RP-295 IV.B2.RP-296 IV.B2.RP-297 IV.B2.RP-299 IV.B2.RP-300 IV.B2.RP-345 IV.B2.RP-354 IV.B2.RP-386 IV.B2.RP-388 N/A IV.B2-1(R-124) IV.B2-6(R-128) IV.B2-6(R-128) IV.B2-9(R-122) IV.B2-14(R-137) IV.B2-17(R-135) IV.B2-18(R-132) IV.B2-21(R-140) IV.B2-21(R-140) IV.B2-22(R-141) N/A N/A IV.B2-34(R-115) IV.B2-33(R-108) N/A N/A N/A N/A 56a PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y C o mb usti on Engi neer in g re actor intern al "Primar y" com ponents expos ed to rea c tor coola n t and n eutro n flu x Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-315 IV.B3.RP-318 IV.B3.RP-359 IV.B3.RP-360 IV.B3.RP-362 IV.B3.RP-364 IV.B3.RP-366 IV.B3.RP-365 IV.B3.RP-326 IV.B3-7 (R

-165) IV.B3-8 (R

-163) N/A N/A N/A N/A N/A N/A N/A B-68 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar 56b PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS)

Comb ustion E ngi neer in g "Exp ans ion" re actor intern al compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-317 IV.B3.RP-331 IV.B3.RP-359a IV.B3.RP-361 IV.B3.RP-362b IV.B3-7 (R

-165) N/A N/A N/A N/A 56c PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y C o mb usti on Engi neer in g re actor intern al "Ex i sting Progr a ms" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B3.RP-319 IV.B3.RP-332 IV.B3.RP-334a IV.B3.RP-336 IV.B3.RP-357 IV.B3-9 (R

-162) IV.B3-17 (R-156) N/A IV.B3-22 (R-170) N/A 58a PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y B abcock & W ilcox reactor inter nal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to w e ar; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B4.RP-240 IV.B4.RP-240a IV.B4.RP-242 IV.B4.RP-247b IV.B4.RP-248b IV.B4.RP-249 IV.B4.RP-251 IV.B4.RP-251a IV.B4.RP-252 IV.B4.RP-254b IV.B4.RP-256b IV.B4.RP-258 IV.B4.RP-259 IV.B4.RP-401 IV.B4-1 (R

-128) N/A IV.B4-4 (R

-183) N/A N/A IV.B4-12 (R-196) IV.B4-15 (R-190) N/A IV.B4-16 (R-188) N/A N/A IV.B4-4 (R

-183) IV.B4-31 (R-205) N/A B-69 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item 58b PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y B abcock & W ilcox reactor inter nal "Exp ansi on" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B4.RP-245b IV.B4.RP-246b IV.B4.RP-254b IV.B4.RP-260 IV.B4.RP-243 IV.B4.RP-243a IV.B4.RP-250 IV.B4.RP-375a N/A N/A N/A IV.B4-31 (R-205) IV.B4-1 (R

-128) N/A IV.B4-12 (R-196) N/A 59a PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y W e sti ngh ouse re actor internal "Primar y" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-270 IV.B2.RP-272 IV.B2.RP-296 IV.B2.RP-297 IV.B2.RP-302a IV.B2.RP-354 IV.B2.RP-388 IV.B2.RP-300 IV.B2-1 (R

-124) IV.B2-6 (R

-128) N/A N/A N/A N/A N/A N/A 59b PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS)

W e stingh ouse reactor intern al "E xpa n s ion" compo nents e x pose d to reactor coo l ant and ne utron flux Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-274 IV.B2.RP-278a IV.B2.RP-287 IV.B2.RP-290 IV.B2.RP-290a IV.B2.RP-290b IV.B2.RP-292 IV.B2.RP-295 IV.B2.RP-388a IV.B2-6 (R

-128) N/A IV.B2-17 (R-135) IV.B2-21 (R-140) N/A N/A IV.B2-21 (R-140) IV.B2-22 (R-141) N/A 59c PWR Stainl ess steel (SS, including CAS S , PH SS or martensitic SS) or nickel allo y W e sti ngh ouse re actor Loss of fracture tough ness due to n eutron irradi atio n embrittlem ent and for CASS, martensitic SS, and PH SS Chapter X I.M16A, "PWR Vessel Internals" No IV.B2.RP-285 IV.B2.RP-288 IV.B2.RP-299 IV.B2.RP-356 IV.B2-14 (R-137) IV.B2-18 (R-132) IV.B2-34 (R-115) N/A B-70 Table 3.1-1 Summar y of Aging Man a gement Prog rams for Rea c tor Ves sel, Internals, an d Reac tor Coolan t Sy stem Ev aluated in Chapte r IV of the GA LL Repor t ID T y pe Compon ent Aging Effec t/Mec h a nism Aging Man a gement Programs Furth er Ev aluation Recomme nd ed Rev 2 Item Rev 1 Item intern al "E xisti ng Programs" co mpon ents expos ed to rea c tor coola n t and n eutro n flu x due to therm a l agi ng embrittlem ent; or chan ges i n dime nsio ns du e to void s w e l l i n g or dist ortion; or loss of prelo ad d ue to thermal and irr adi atio n enh anc ed stress rela xati o n or creep; or loss of materia l due to w e ar C-1 Appendix C STAFF RESPONSE TO PUBLIC COMMENTS ON DRAFT LICENSE RENEWAL IN TERIM STAFF GUIDANCE 2011-04 C-2 Source of Comments I. Comments from Jean Smith, Electric Power Research Institute Materials Reliability Program (EPRI-MRP) and the Pressurized Water Reactor Owners Group Materials Subcommittee (PWROG-MSC) (ADAMS Accession No. ML12146A267) II. Comments from Mark Richter, Nuclear Energy In stitute (NEI) (ADAMS Accession No. ML12144A147)

1. Source ID Summary of Comment Response 1 I-1 The NRC reviewed and approved with limitations MRP-227 Revision 0, and subsequently, MRP-227-A was published to incorporate the SER additions. All needed actions for licensees are contained in MRP-227-A. As a result, it is appropriate for the NRC to review a licensee's PWR reactor internals aging management program against the criteria contained in MRP-227-A. As such, it is not necessary to include all the details currently in NUREG-1800 and NUREG-1801 regarding PWR reactor internals, and instead, only a reference to MRP-227-A should be made. Outlining the requirements for reactor internals in the Interim Staff Guidance may lead to confusion with respect to the implementation of duplicate requirements, may cause undue NRC staff burden reconciling the documents each time MRP-227 is revised by the industry, and will likely lead to human errors in document alignment through future revisions. The staff agrees with the comment, in part, that it is not necessary to have the level of detail included in LR-ISG-2011-04 issued for public comment regarding PWR reactor vessel internal (RVI) components. However, the staff does not agree that the final LR-ISG-2011-04 should only reference MRP-227-A; instead reference to the topical report should be made only when it is appropriate. Revisions were made to eliminate duplication of information for RVIs that is detailed in MRP-227-A. The following is a summary of the revisions that have been incorporated into final LR-ISG-2011-04 as a result of this comment:

Revision to GALL Report Aging Management Program (AMP) XI.M16A In general, GALL Report AMP XI.M16A, "PWR Vessel Internals," in final LR-ISG-2011-04 references MRP-227-A in the program elements and does not delineate the MRP-227-A inspection and evaluation guidelines for PWR RVIs. In addition, areas resolved in the staff's safety evaluation (SE), Revision 1, for MRP-227 and Applicant/Licensee Action Items (A/LAI) are not addressed in GALL Report AMP XI.M16A in final LR-ISG-2011-04.

Revision to SRP-LR Table 3.1-1 Final LR-ISG-2011-04 does not incorporate specific reference to "Primary Category," "Expansion Category," or "Existing Program" inspection and evaluation guidelines into the "Rev. 2 Item" column in the aging management review (AMR) line items for PWR RVIs in SRP-LR Table 3.1-1. In addition, the "Component" column for PWR RVIs in SRP-LR Table 3.1-1 in final LR-ISG-2011-04 is based on the commodity groups and inspection categories in MRP-227-A.

Revision to GALL Tables IV.B2, IV.B3, and IV.B4 GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 do not reference inspection categories and MRP-227-A inspection and evaluation guidelines.

Revision to SRP-LR Further Evaluation Recommendations for PWR RVIs Areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04 (i.e., these SRP-LR sections were deleted and do appear in final LR-ISG-2011-04). In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the C-3 # Source ID Summary of Comment Response A/LAIs for MRP-227-A in Appendix C of the LRA.

2 I-2 Commenter referenced statement in Section 3.

1.2.2.9.A.1 of Appendix A of LR-ISG 2011-04.

This statement requires that licensees include responses to applicant action items in both Appendix C of the LRA and in appropriate further evaluation sections of the LRA. This duplication of information provides no significant value to the reviewers. It is recommended that all A/LAI responses be included only in Appendix C, so they are in an easily-referenced location. Any additional discussion of the A/LAIs in the further evaluation sections of the SRP should be limited to identifying each of the items requiring responses and any details necessary to ensure responses are adequate. Any other items requiring discussion of the A/LAI responses in further evaluation sections of the LRA should be deleted or reference made to Appendix C of the LRA.

The staff agrees with the comment that the responses to A/LAIs are to be provided in Appendix C of the LRAs. Thus, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. In addition, as a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of the staff's resolution of Source ID I-1, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.1. 3 I-3 In NUREG-1801 Revision 2 XI.M16A Program Description, last paragraph, as well as in ISG-LR-2011-04 Section 3.1.2.2.9.A.2, both an aging management program and an inspection plan are required to be submitted as part of an applicant's license renewal application.

However nowhere in these two documents is there any clear guidance on the information that should be included in an inspection plan. This ambiguity could lead to applicants submitting information that might not meet NRC needs in this area.

In order to address this situation it is requested that the aging management program and inspection plan for an applicant be clearly defined. It is proposed that the aging management program address the 10 program element recommendations for PWR RVI components in GALL AMP XI.M16A, PWR Vessel Internals (AMP XI.M16A in NUREG-1801, Revision 2). The inspection plan could be included within a program (i.e. a program/plan) or be a separate document if submitted with a license renewal application.

The industry believes these elements are satisfied by the applicable line items from Tables 4-1 through 4-9 and Tables 5-1 through 5-3 of MRP-227-A. The inspection plan submitted as part of a license renewal application (LRA) should be included in Appendix C of the LRA along with the responses to the A/LAI items since it is a requirement of A/LAI No. 8. The staff agrees, in part, with the comment in that better guidance regarding the inspection plan is needed to avoid confusion. Regulatory Issue Summary (RIS) 2011-07, "License Renewal Submittal Information For Pressurized Water Reactor Internals Aging Management," dated July 21, 2011, provides the staff's expectations for Category D plants (PWR plant licensees that had not submitted their LRAs but plan to submit an LRA in the future) to submit, for NRC staff review and approval, an AMP for vessel internals that is consistent with MRP-227-A.

An "inspection plan" is one aspect of satisfying A/LAI No. 8 of the staff's SE, Revision 1, for MRP-227. An "inspection plan" provides information about the RVI components to be inspected and a description of how they will be managed for age-related degradation (e.g., examination method, frequency, acceptance criteria, coverage, etc.). The staff expects that the details of an "inspection plan" for Category D plants will be incorporated into the LRA submittal as part of the 10-element AMP and AMR line items. Thus, consistent with RIS 2011-07, the staff does not expect Category D plants to provide a separate document that contains an "inspection plan" in response to A/LAI No. 8.

In order to avoid duplication and confusion, as part of the resolution to Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their C-4 # Source ID Summary of Comment Response responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. In doing this, the explicit reference to an "inspection plan" is avoided in the body of the AMP, and "inspection plan" is only referenced as part of A/LAI No. 8 in the staff's SE, Revision 1, for MRP-227.

However, the staff does not agree with the Commenter's general claim with respect to what satisfies an inspection plan per A/LAI No. 8, as additional guidance is outlined in the SE, Revision 1, for MRP-227, and fulfillment of that action item will depend on each applicant's plant-specific review.

4 I-4 The stipulation of appropriate inspection methodologies for these reactor internals components has already been addressed in the review of MRP-227-A. The recommended inspection methods have already been reviewed and found to be adequate to detect the relevant conditions. The AMP attribute that is at issue is not detection of aging effects; instead, the issue is the applicant's corrective action program, and the disposition of relevant conditions through supplemental examination or engineering evaluation, both of which are outside the scope of the Mandatory or Needed requirements of MRP-227-A. Standards for engineering evaluation are addressed in Section 6 of MRP-227-A and in the methodologies described in WCAP-17096. These recommendations are based on the practice used in Section XI of the ASME code and are consistent with existing aging management programs. Further justification for the use of the VT-3 examination is not necessary and should not be required by the ISG.

It is recommended that Acceptance Criteria Item 3.1.2.2.9.A.7 (Use of VT-3 Visual Inspection Techniques for Detection of Cracking) be completely eliminated and replaced by a limited requirement to address the acceptability of VT-3 as a management approach for components that 1) were not already considered for aging management in the development of MRP-227-A, 2) are evaluated to require active aging monitoring, and 3) are non-redundant. The Commenter provided justification for its recommendation.

The staff agrees with the comment, in part, that final LR-ISG-2011-04 address the acceptability of VT-3 as a management approach for certain components. Thus, final LR-ISG-2011-04 does not incorporate SRP-LR Sections 3.1.2.2.9.A.7, 3.1.2.2.9.C.1, and 3.1.2.2.9.C.4. However, the staff's position on the use of VT-3 to detect cracking will continue to be documented in the "Detection of Aging Effects" program element in GALL Report AMP XI.M16A, which states, in part, the following:

"...VT 3 visual methods may be applied for the detection of cracking in non-redundant RVI components only when the flaw tolerance of the component, as evaluated for reduced fracture toughness properties, is known and the component has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions." 5 I-5 Appendix A - Section 2, Acceptance Criteria Ite m 3.1.2.2.9.A.9 (Identification of TLAAs for PWR-Design RVI Components) on Page A-20 and A-21 stipulates that, "in order to satisfy the requirements of the ASME Code,Section III, Subsections NG-2160 and NG-3121, license renewal applicants demonstrating acceptability of RVI components with design-basis cumulative usage factor (CUF) analyses that are TLAAs should include the effects of the reactor The staff agrees with the comment, in part, that the evaluation of environmental effects for PWR RVI core support structures should not be incorporated in SRP-LR Section 3.1.2.2.9.A.9 in final LR-ISG-2011-04. However, the staff does not agree with the commenter's statement that the evaluation of time-limited aging analyses for the reactor internals should be addressed in accordance with the existing 10 CFR Part 54 requirements without the need to include environmental effects.

C-5 # Source ID Summary of Comment Response coolant system water environment in the fatigue CUF analyses." The Commenter provided its justification for removal of this last sentence.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. Final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of the staff's resolution of Source ID I-1, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.9.

To the extent that the commenter does not agree with the need to address evaluation of environmental effects, the staff's SE, Revision 1 for MRP-227-A documents the basis for limitations and conditions being placed on the use of MRP-227 as well as A/LAIs that shall be addressed by applicants/licensees who choose to implement the NRC-approved version of MRP-227. Specifically, the topic of environmentally-assisted fatigue for PWR RVIs is addressed in A/LAI No. 8, Item 5 of MRP-227-A. Thus, the intent of LR-ISG-2011-04 is not to supplement or modify the evaluation in the staff's SE, Revision 1.

6 I-6 The component-specific AMR items described in Appendix-A, Sections 4, 5 and 6 are based on migration from NUREG-1801. As a result the listing is more complex than the approved MRP-227-A tables. For example, there are approximately 25 items in Section 5 that classify as "Primary" component examinations, whereas the equivalent component list in MRP-227-A contains only 13 items. The component content is very similar but the breakdown is complex. A key advantage of aligning license renewal commitments to the MRP-227-A format is to facilitate important, industry-wide program updates based on Operating Experience through the NEI 03-08 process. The alignment between MRP-227-A and NUREG-1801 is compromised by embedding item detail in the ISG format. It is recommended that NUREG-1801 refer existing AMR items to "the applicable MRP-227-A table" and retain detail only for those items which may be beyond the scope of MRP-227-A. This will significantly reduce applicant and NRC staff burden, and improve integration of evolutionary changes through the NEI 03-08 process. The staff does not agree with the comment recommending that NUREG-1801 refer existing AMR line items to "the applicable MRP-227-A table" and retain detail only for those items which may be beyond the scope of MRP-227-A. In accordance with 10 CFR 54.21(a)(3) for each structure and component identified as part of the integrated plant assessment (IPA), the LRA is to demonstrate that the effects of aging will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. The IPA is independent of the line items in MRP-227-A and the GALL Report and may also result in additional components beyond the generic lists in these documents. This requires that the LRA provide a complete listing of AMR line items, which may include items consistent with MRP-227-A and the GALL Report and may also result in additional components beyond the generic lists in these documents. Thus, the number and content of AMR line items in the inspection tables of MRP-227-A are not the only basis for determining the AMR line items in the GALL Report. Similarly, the AMR line items in the GALL Report are not the only basis for determining the aging effects requiring management for components or establishing the AMR line items that are included in an LRA.

However, final LR-ISG-2011-04 incorporates revisions to SRP-LR Table 3.1-1 and GALL Tables IV.B2, IV.B3, and IV.B4 as summarized in the C-6 # Source ID Summary of Comment Response staff's resolution to Source ID I-1.

7 I-7 ISG implementation of Applicant/Licensee Action items from the MRP-227-A SER is by way of notes to AMR items listed in Sections 4, 5 and 6. This could be addressed by reference to the appropriate SER action items. It is recommended that the required evaluations would be documented in a single location specified by the ISG rather than associated with individual items. Associating these actions with each individual AMR item increases the burden for both the applicant and NRC staff reviewer. The staff agrees with the comment that associating A/LAIs with each individual AMR line item increases the burden for both the applicant and NRC staff reviewer.

As part of the resolution to Source ID I-1, final LR-ISG-2011-04 incorporates revisions to SRP-LR Table 3.1-1 and GALL Tables IV.B2, IV.B3, and IV.B4. Specifically, GALL Tables IV.B2, IV.B3, and IV.B4 were revised to be consistent with the format of AMR items in the GALL Report for non-RVI components and the footnotes in the "Further Evaluation" column of these tables were deleted.

8 I-8 The draft ISG requires Applicants to develop and submit evaluation of inaccessible Reactor Vessel Internal components in accordance with Note 3 to Sections 4 and 5, and Note 2 to Section 6. With the exception of A/LAI #6 of the MRP-227-A SER, these evaluations have been addressed during review and approval of the Industry program. The requirement to develop, submit and review the inspection basis is unnecessary. It is recommended that this note be eliminated. The staff agrees with the comment that it is not necessary to provide an evaluation of inaccessible RVI components, with the exception of A/LAI No. 6 of MRP-227-A. As part of the resolution to Source ID I-7, final LR-ISG-2011-04 incorporates revisions to delete the further evaluation footnotes from GALL Tables IV.B2, IV.B3, and IV.B4.

As a result of staff's reso lution to Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not redundantly addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04.

9 I-9 MRP-227-A provides applicants with an alternative to the defined inspection requirements when plant-specific analyses of accumulated fatigue usage are performed. Applicants may choose to either inspect in accordance with the approved MRP-227-A schedules, or perform analyses. In cases where Applicants perform analyses to relax MRP-227-A requirements, those analyses would be submitted for NRC staff approval in accordance with A/LAI 8. The ISG is unclear regarding these alternatives. For example item IV.B3.RP-343 appears to require physical examinations to support acceptance of the TLAA. The industry recommends that the ISG refer to MRP-227-A and the associated A/LAI requirement discussions. The staff agrees with the comment that LR-ISG-2011-04 refer to MRP-227-A and the associated A/LAI discussions for alternatives or deviations to the inspection and evaluation guidelines in MRP-227-A.

It is the responsibility of the license renewal applicant to demonstrate in accordance with 10 CFR 54.21(a)(3) that it can adequately manage aging of RVIs for the period of extended operation, whether through the use of MRP-227-A or alternatives. If a TLAA exists for a RVI, in accordance with 10 CFR 54.21(c)(1)(iii), an applicant may choose to demonstrate the effects of aging on the intended function of the component will be adequately managed for the period of extended operation. It is incumbent on the license renewal applicant to provide this demonstration of aging management, which can include the use of MRP-227-A or an appropriate alternative.

In order to avoid redundancy, areas reso lved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed again in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA.

C-7 # Source ID Summary of Comment Response 10 I-10 Item 9.B.1 of the ISG notes that Section 3.2.5.3 of the NRC SE (Revision 1) on MRP-227 Revision 0 recommends that the applicant consider replacement or inspection activities with regard to the Control Rod Guide Tube (CRGT) split pins if the pins are currently fabricated with Alloy X-750 or Type 316 stainless steel material. A review of the referenced section of the SE does not reach the conclusion that this specificity of action is required; the SE requirement is to evaluate the adequacy of the plant-specific existing program to ensure that the aging degradation is adequately managed during the extended period of operation. The SE direction is on evaluation of the performance of the existing program and does not suggest that it should be changed to include inspections. Therefore the industry considers the specificity of direction provided in the SE to be sufficient and the ISG should not provide alternate direction.

The staff agrees with the comment that there is an inconsistency between SRP-LR Section 3.1.2.2.

9.B.1 in draft LR-ISG-2011-04 and Section 3.2.5.3 of the staff's SE, Revision 1, for MRP-22

7. As part of the staff's resolution to Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of the staff's resolution of Source ID I-1, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.B.1.

Specific to the Westinghouse CRGT split pins, A/LAI No. 3 recommends an evaluation to consider the need to replace the Alloy X-750 split pins, if applicable, or an inspection of the replacement type 316 stainless steel split pins to ensure that cracking has been mitigated and that aging degradation is adequately monitored during the extended period of operation. Thus, the intent of LR-ISG-2011-04 is not to supplement or modify the evaluation in the staff's SE, Revision 1, but rather, to recommend that the response to A/LAI No. 3 of MRP-227-A be appropriately documented in Appendix C of the LRA.

11 I-11 Section C.3, page A23 of LR-ISG 2011-04 states that per MRP-227-A, "...EVT-1 inspections of certain CE-design components would be necessary if the design basis fatigue TLAAs for the components could not demonstrate that fatigue-induced cracking would be adequately managed..." This statement does not accurately represent MRP-227-A Table 4-2, because it assumes that the fatigue evaluations required by the MRP-227-A table item already exist and are part of the current licensing basis, and therefore are formally classifiable as TLAAs. In fact, many, if not all, of the older CE design reactor internals were not qualified to the fatigue rules of ASME III, so TLAAs as defined in 10 CFR Part 54 do not exist. Further, page A24 of the draft ISG states "Otherwise, CE-design applicants for renewal are requested to credit the MRP's EVT-1 basis in MRP-227-A as the applicable aging management basis if either: (1) the CLB does not include applicable CUF or It fatigue analyses for these components;-" This statement appears to compel the applicant who does not have a current licensing basis TLAA to perform EVT-1 inspections. MRP-227-A clearly does not require inspections based solely on the lack of a current licensing basis TLAA. In fact, it only requires that a fatigue evaluation be The staff agrees with the comment that the discussion related to CE-designed lower core flange welds, core support plates, and fuel alignment plates in SRP-LR Section 3.1.2.2.9.C.3 in draft LR-ISG-2011-04 is not clear.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.C.3.

10 CFR 54.21(a)(1) requires that license renewal application contain an IPA that must, for those systems, structures, and components within the scope of Part 54, identify and list those structures and components subject to an aging management review (AMR). The components evaluated in MRP-227-A may not fully encompass the components identified in an IPA, as required by 10 CFR 54.21(a)(1), and therefore, should not be considered a substitute for performance of an IPA.

The aging effects requiring management for RVIs are not governed by C-8 # Source ID Summary of Comment Response performed to determine if a fatigue issue might exist; and if so, where would inspection be focused to manage it. The method of the fatigue evaluation was intended to be the usual engineering practice, for example by comparison of the number expected operating transient cycles to those specified by design, or by stress analysis if required. MRP-227-A; rather the content of MRP-227-A serves to assist a PWR license renewal applicant. In accordance with 10 CFR 54.21(a)(3), the effects of aging are to be managed for all applicable aging effects for a particular component, which may be broader than the aging effects identified in MRP-227-A and the GALL Report for RVIs. It is the responsibility of the license renewal applicant to demonstrate that it can adequately manage aging of RVIs for the period of extended operation, whether through the use of MRP-227-A or alternatives.

Therefore, if the CE-designed lower core flange welds, core support plates, and fuel alignment plates are subject to an AMR and fatigue is an applicable aging effect, regardless if there is a TLAA, the LRA must demonstrate that fatigue will be adequately managed in accordance with 10 CFR 54.21(a)(3).

12 I-12 For A/LAI No. 2, when comparing the licensee renewal AMR from BAW-2248A to the tables in MRP-189, the locking devices for the vent valve were identified as a possible "Primary" component. The original vent valves located next to outlet nozzles failed due to flow induced vibration, and those valves next to the nozzles were replaced with locking devices made containing Alloy 600.

It is recommended that Table IV Reactor Vessel, Internals, and Reactor Coolant System, B4 Reactor Vessel Internals (PWR) - Babcock and Wilcox on page A-124 of LR-ISG 2011-04 be revised to include a line item addressing Alloy 600 replacement vent valve locking devices, which are subject to aging degradation due to primary water stress corrosion cracking (PWSCC).

The staff agrees with the comment to include an AMR line item for cracking of B&W vent valve locking devices made from Alloy 600 materials in GALL Table IV.B4 of draft LR-ISG-2011-04. Final LR-ISG-2011-04 incorporates the core support shield vent valve top and bottom retaining rings to be managed for cracking in GALL AMR Item IV.B4.RP-252a. 13 I-13 In Item 8 on page A-11 of the LR-ISG, the second sentence appears to be incomplete with respect to the statement pertaining to "-confirming that the quality of inspections, flaw evaluations, and corrective actions performed under this program." It is recommended that the revised statements be reviewed for completeness.

The staff agrees with the comment that the sentence in the "Confirmation Process" program element in GALL Report AMP XI.M16A of draft LR-ISG-2011-04 is incomplete. Final LR-ISG-2011-04 completes this sentence in the "Confirmation Process" program element.

14 I-14 Item 3 on page A-16 of the LR-ISG should reference NRC SE Section 3.2.5.1 and not Section 3.5.1. It is recommended that this reference be revised. The staff agrees with the comment that SRP-LR Section 3.1.2.2.9.A.3 in draft LR-ISG 2011-04 should reference NRC SE Section 3.2.5.1 and not Section 3.5.1.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a C-9 # Source ID Summary of Comment Response result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.

2.2.9.A.3.

15 I-15 Item D.1 on page A-25 discusses evaluation "Acceptance Criteria" recommendations applicable to Babcock and Wilcox reactor internals. In general, A/LAI 4 is not specific relative to the wording for the manner in which the items were stress relieved, and it was stated that a "stress relief process" was used. In Item D.1, the wording used in some cases implies a "post-weld heat treatment" process. The words "stress relief process" should be used consistently without the implication of a heat treatment process only. In addition, the requirements in Item D.1 appear to go beyond the requirements of the A/LAI as it was written and approved by the MRP-227-A SER.

The staff agrees with the comment to use the terminology "stress relief process" consistently throughout SRP-LR Section 3.1.2.

2.9.D.1 of draft LR-ISG-2011-04. Final LR-ISG-2011-04 does not use the term "post-weld heat treatment" and this term is replaced with the term "stress relief process." In addition, as a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.D.1.

16 II-1 Page A IV.B2.RP-300 Alignment and interfacing components, such as hold down springs, are addressed in MRP-227-A. Based on MRP-227-A, the intent of the GALL was only to apply to hold down springs made from Type 304 Stainless Steel (SS). The possibility of thermal embrittlement of hold down springs made from Type 403 martensitic SS is not addressed. The issue is, however, discussed in the proposed SRP section 3.1.2.2.9.A.6 and in applicant action item 7 of the SER (Revision1).

Proposed Change: Include the words "applicable to hold down springs fabricated from Type 304 SS" and add a line item to address thermal embrittlement for hold down springs fabricated for Type 403 stainless steel.

The staff agrees with the comment that the possibility of thermal embrittlement for Type 403 martensitic stainless steel hold down springs is not addressed. Final LR-ISG-2011-04 does not use the term "(Aust. SS Material)" in the "Material" column in GALL AMR Item IV.B2.RP-300. Furthermore, the use of the term "Stainless Steel" in GALL AMR Item IV.B2.RP-300 is generic and includes all grades of "stainless steel" as defined in GALL Table IX.C, "Selected Definitions & Use of Terms for Describing and Standardizing - MATERIALS." With these revisions hold down springs made from Type 403 martensitic SS are addressed in GALL AMR Item IV.B2.RP-300. 17 II-2 Page A Section 3.1.2.2.9.A.3, second paragraph There is little guidance on Applicant Action Item #2 related to additional RVI piece parts and what was used during the development of MRP 191. Utilities are left to draw a conclusion that unless the utility implemented a modification beyond the vendor's recommendation, all of the piece parts in the reactor vessel were considered during the development of MRP-189, 191 and 227-A.

Proposed Change: Add verbiage to provide additional guidance to allow utilities to make the assumption that unless a utility implemented modifications beyond that recommended by the The staff does not agree with the comment, in particular the inference that, unless a utility implemented modifications beyond that recommended by the vendor of the RVI, all of the piece parts of the RVI were considered during the development of MRP-189, 191 and 227-A. The methodology and results of a topical report, such as MRP-227-A, cannot be assumed to be generically bounding for every plant.

The IPA described in the response to Source ID I-11 is a plant-specific evaluation performed by a license renewal applicant. Thus, the components evaluated in MRP-189, 191 and 227-A may not fully encompass the components identified in an applicant's IPA and therefore, should not be considered a substitute for performance of an IPA. The C-10 # Source ID Summary of Comment Response vendor of the RVI, then all of the piece parts of the RVI were considered during the development of MRP-189, 191 and 227-A. aging effects requiring management for RVIs may be broader than the aging effects identified in MRP-189, 191 or 227-A. It is the responsibility of the license renewal applicant to demonstrate, in accordance with 10 CFR 54.21(a)(3), that it can adequately manage aging of RVIs for the period of extended operation, whether through the use of MRP-227-A or alternatives. The content in MRP-189, 191 or 227-A only serves to assist a PWR license renewal applicant.

However, as addressed in the staff's resolution to Source ID I-1, in order to avoid redundancy, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.3.

18 II-3 Page A Section 3.1.2.2.9.A.3 The words in the third paragraph are confusing and it is not clear what is meant by plant specific AMR line items or why Note E would be appropriate. For those applicants whose plant-specific review results in identification of additional components for inspection or different component inspection categories from those identified in MRP-227-A, the applicant is requested to identify the changes in the component inspection categories as either plant-specific AMR line items or NEI Note E consistent with GALL AMR items (whichever is applicable) in their Table 2 AMR line items for their PWR RVI components.

Proposed Change: It is suggested that if only a component line item or two that is not in GALL is being added then an exception can be taken to the program and justification be added that includes inspection specifics such as method and acceptance criteria such that the whole program doesn't have to be evaluated as a plant specific program.

The staff agrees with the comment that portions of SRP Section 3.1.2.2.9.A.3 are confusing.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.3.

19 II-4 Page A last paragraph Page A Section 3.1.2.2.9.1.2 The document does not provide clear direction as to what goes into an inspection plan.

The staff agrees that draft LR-ISG-2011-04 does not provide clear direction as to what goes into an inspection plan but does not agree with the commenter's proposed change. The staff does not agree with the Commenter's general claim with respect to what satisfies an inspection plan per A/LAI No. 8, as additional guidance is outlined in the SE, Revision 1, for MRP-227, and fulfillment of that action item will depend on each C-11 # Source ID Summary of Comment Response Proposed Change: Add verbiage to allow utilities to better determine what the inspection plan should consist of (e.g., A Westinghouse design plant should provide unit specific information in the Inspection Plan consistent with tables 4-3, 4-6 and 4-9 of MRP-227-A and the A/LAIs).

applicant's plant-specific review.

See the staff's resolution to Source ID I-3. RIS 2011-07 provides the staff's expectations for Category D plants (PWR plant licensees that have not submitted their LRAs but plan to submit an LRA in the future) to submit, for NRC staff review and approval, an AMP for vessel internals that is consistent with MRP-227-A. As an "inspection plan" is one aspect of satisfying A/LAI No. 8 of the staff's SE, Revision 1, for MRP-227. An "inspection plan" provides information about the RVI components to be inspected and a description of how they will be managed for age-related degradation (e.g., examination method, frequency, acceptance criteria, coverage, etc.). The staff expects that the details of an "inspection plan" for Category D plants will be incorporated into the LRA submittal as part of the 10-element AMP and AMR line items. Thus, consistent with RIS 2011-07, the staff does not expect Category D plants to provide a separate document that contains an "inspection plan" in response to A/LAI No. 8.

To avoid confusion, final LR-ISG-2011-04 avoids explicit reference to an "inspection plan" in the body of the AMP, and "inspection plan" is only referenced as part of A/LAI No. 8 in the staff's SE, Revision 1, for MRP-227. 20 II-5 Page A Table 3.1-1 Item 27a It is not clear that this line item is only applicable to hold down springs fabricated from Type 304 SS.

Proposed Change: Add Type 304 SS hold down springs.

The staff agrees with the comment that Table 3.1-1, Item 27a, of draft LR-ISG-2011-04 does not clearly address Type 304 stainless steel hold down springs.

Final LR-ISG-2011-04 does not include this item, but Westinghouse Type 304 stainless steel hold down springs were incorporated into Table 3.1-1, Item 59a, in final LR-ISG-2011-04, which uses the generic terminology "stainless steel."

21 II-6 Page A Table 3.1-1 Item 3 Under 'Further Evaluation Recommended' column, it is not clear what "It" stands for?

Proposed Change: Provide an explanation.

"It" refers to the parameter being calculated for the cyclical loading analyses. In later editions of the ASME Code Section III, these analyses were referred to as cumulative usage factor (CUF) analyses. Thus, the "I t" parameter is analogous to the CUF parameter required for Class 1 components designed to more recent editions of the ASME Code,Section III. The subscripted "t" was removed in the formatting duri ng the issuance of draft LR-ISG-2011-04 for public comment.

As a result of the staff's resolution of Source ID I-1, SRP-LR Table 3.1-1 Item 3 does not incorporate the reference to the "I t" parameter in final LR-ISG-2011-04.

C-12 # Source ID Summary of Comment Response 22 II-7 Page A IV.B2.RP-280 There is confusion regarding what comprises the lower core barrel flange weld for Westinghouse designed plants. This component is still listed in MRP-191, and 227-A for Westinghouse designed plants. MRP-227-A indicates it may be the weld between the core barrel and the lower support forging or casting.

Proposed Change: Provide an explanation regarding what this component is.

Page 3-11 of MRP-227-A states that "[t]he lower support forging is welded to and supported by the core barrel, which transmits vertical loads to the vessel through the core barrel flange." In addition, Table 5-3 of MRP-227-A provides the "Examination Acceptance Criteria and Expansion Criteria" for the "Core Barrel Assembly - Lower core barrel flange weld." Footnote 2 of Table 5-3 states that "[t]he lower core barrel flange weld may alternatively be designated as the core barrel-to-support plate weld in some Westinghouse plant designs."

No revisions were made as a result of this comment.

23 II-8 Page A Section 3.1.2.2.9.A.4 In the subject paragraph, it appears the NRC wanted an exception not an enhancement: For those component inspections that do not achieve the inspection coverage criteria stated in the NRC SE (Revision 1) on MRP-227, the applicant is requested to take a deviation from the MRP-defined inspection criteria and describe the process and type of evaluation that will be implemented to evaluate the impact of the aging effects on the inaccessible regions of the components. In this case, the applicant is requested to identify this process as an applicable enhancement of the "monitoring and trending" program element of its RVI Program.

Proposed Change: Clarify what is expected.

The staff agrees with the comment that the referenced text in SRP-LR Section 3.1.2.2.9.A.4 of draft LR-ISG-2011-04 is not clear. As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.4.

24 II-9 Page A IV.B2.RP-280 It is not clear how Note 3 in the "Further Evaluation" column is applicable to this GALL Line Item.

Proposed Change: Clarify the applicability.

The staff agrees with the comment that the applicability of Note 3 to GALL AMR Item IV.B2.RP-280 is not clear. As a result of the staff's resolution of Source ID I-1, Final LR-ISG-2011-04 incorporates revisions to GALL Tables IV.B2, IV.B3, and IV.B4 as summarized above.

25 II-10 Page 3 In the last paragraph of the Discussion section only table 3-1 is listed for justification of TE for the materials. Tables 3-2 and 3-3 should be mentioned since 3-1 is only for B&W internals.

Proposed Change: Add Tables 3-2 and 3-3.

The staff agrees with the comment that Tables 3-2 and 3-3 should be referenced in the last paragraph of the "Discussion" section. However, the "Discussion" section of final LR-ISG-2011-04 no longer references Table 3-1 in MRP-227-A.

26 II-11 Page A-7 The staff agrees with the comment to change the terminology to "Aging Management Requirement" tables in the "Parameters Monitored/Inspected" C-13 # Source ID Summary of Comment Response The second paragraph in this Section refers to condition monitoring tables in MRP-227-A. There are no tables with this title in MRP-227-A Proposed Change: Change to Aging Management Requirement tables. program element. The "Parameters Monitored/Inspected" program element in final LR-ISG-2011-04 states the following:

"Specifically, the program implements the parameters monitored/inspected criteria consistent with the applicable tables in Section 4, 'Aging Management Requirement,' in MRP-227-A-"

27 II-12 Page A-9 Only Table 5-1 is listed for acceptance criteria when MRP-227-A contains three tables, 5-1 thru 5-3 Proposed Change: Change to read " Section 5 and Tables 5-1 thru 5-3 of MRP-227" The staff agrees with the comment to add references to Table 5-2 and 5-3 of MRP-227-A for the "Acceptance Criteria" program element. The "Acceptance Criteria" program element of GALL Report AMP XI.M16A in final LR-ISG-2011-04 references Table 5-1 through 5-3 of MRP-227-A.

28 II-13 Page A-10 The first paragraph on the page says "The program adopts the acceptance criteria for the physical measurement monitoring methods recommended in MRP-227-A, as qualified in Section 3.3.5 and A/LAI No. 5 in Revision 1 of the NRC SE on MRP-227". Section 3.3.5 of the MRP does not specify acceptance criteria so there is nothing to be adopted. It only requires it be developed as discussed in footnote 3.

Proposed Change: Change sentence to read "The program includes acceptance criteria for the physical measurement monitoring methods as recommended in MRP-227-A, Section 3.3.5 and A/LAI No. 5 in Revision 1 of the NRC SE on MRP-227".

The staff agrees with the comment that Section 3.3.5 of MRP-227-A does not specify acceptance criteria for physical measurements. However, as a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in GALL Report AMP XI.M16A in final LR-ISG-2011-04.

The "Acceptance Criteria" program element in final LR-ISG-2011-04 states that, in general, the AMP establishes appropriate acceptance criteria for any physical measurement monitoring methods that are credited for aging management of RVIs.

29 II-14 Page A-12 The following sentence relates to notification criteria: "The evaluation in Section 3.5 of Revision 1 of the SE on MRP-227 provides the staff's basis for endorsing the NEI 03-08 implementation process for these programs. This includes NRC's endorsement of the NEI 03-08 criteria for notifying the NRC of any deviation from the I&E methodology in MRP-227-A and justification of the deviation no later than 45 days after approval by a licensee executive."

Proposed Change: Delete this sentence as it already is discussed in element 9 where it is appropriate. The staff agrees with the comment that the sentence associated to the notification criteria already exists in the "Administrative Controls" program element and does not need to be repeated in the "Operating Experience" program element of GALL Report AMP XI.M16A. The "Operating Experience" program element of GALL Report AMP XI.M16A in final LR-ISG-2011-04 does not incorporate this sentence associated with the notification criteria.

C-14 # Source ID Summary of Comment Response 30 II-15 Page A-8 The justification required for the use of VT-3 to detect cracking over that specified in MRP-227A and approved by the staff in the SE that allows its use without the additional limitations and analyses is not needed. Proposed Change: Eliminate need for additional justification if requirements as specified in SER and MRP are met.

The staff agrees with the comment that additional justification for the use of VT-3 to detect cracking is not needed if requirements specified in the SER and MRP are met. As a result of the staff's resolution of Source ID I-4, final LR-ISG-2011-04 does not incorporate SRP-LR Sections 3.1.2.2.9.A.7, 3.1.2.2.9.C.1, and 3.1.2.2.9.C.4. However, the staff's position on the use of VT-3 for the detection of cracking will continue to be documented in the "Detection of Aging Effects" program element in GALL Report AMP XI.M16A. 31 II-16 Page A Section 3.1.2.2.9.C.1 The justification required for the use of VT-3 to detect cracking over that specified in MRP-227A and approved by the staff in the SE that allows its use without the additional limitations and analyses is not needed. Proposed Change: Eliminate need for additional justification if requirements as specified in SER and MRP are met.

The staff agrees with the comment that additional justification for the use of VT-3 to detect cracking is not needed if requirements specified in the SER and MRP are met. As a result of the staff's resolutions to Source ID I-4 and ID II-15, final LR-ISG-2011-04 does not incorporate SRP-LR Sections 3.1.2.2.9.A.7, 3.1.

2.2.9.C.1, and 3.1.2.2.9.C.4. However, the staff's position on the use of VT-3 for the detection of cracking will continue to be documented in the "Detection of Aging Effects" program element in GALL Report AMP XI.M16A.

32 II-17 Page A Section 3.1.2.2.9.C.3 The option presented as (3), as an alternative basis for accepting the design basis fatigue analyses in accordance with the TLAA acceptance requirement in 10 CFR 54.21(c)(1)(iii) does not make sense when compared to options 1 and 2 Proposed Change: Add the word "the EVT-1 is used" at the beginning The staff agrees with the comment that the discussion related to CE-designed lower core flange welds, core support plates, and fuel alignment plates in SRP-LR Section 3.1.2.2.9.C.3 of draft LR-ISG-2011-04 is not clear. As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.C.3.

Also see the staff's resolution to Source ID I-11, in which the staff clarified that, if the CE-designed lower core flange welds, core support plates, and fuel alignment plates are subject to an AMR and fatigue is an applicable aging effect, regardless if there is a TLAA, then in accordance with 10 CFR 54.21(a)(3), the LRA must demonstrate that fatigue will be adequately managed. 33 II-18 Page A Section 3.1.2.2.9.D.1 There is no need for a plant-specific enhancement of the The staff agrees with the comment that there is not a need for a plant-specific enhancement of the "Preventive Actions" program element discussed in SRP-LR Section 3.1.2.2.9.D.1 of draft LR-ISG-2011-04, which C-15 # Source ID Summary of Comment Response "preventative actions" program element for their RVI Program enhancement to be identified if an applicant confirms that the welds were appropriately stress-relieved. An enhancement doesn't seem appropriate since the action has already been taken.

Proposed Change: Eliminate the need for an enhancement is associated with A/LAIs No. 4 of MRP-227-A.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.D.1.

34 II-19 Page A Table 3.0-1 There is no need for the words "or to applicable NRC further evaluation "acceptance criteria" recommendations in Section 3.1.2.2 of the SRP-LR (i.e., the latest NRC issued version of NUREG-1800)." Specific acceptance criteria do not need to be part of a SAR description. If it is an enhancement it will already be a commitment.

Proposed Change: Delete The staff agrees with the comment that the further evaluation acceptance criteria do not need to be specified as part of a Safety Analysis Report description. Final LR-ISG-2011-04 does not incorporate this second paragraph in the "Description of Program" column for GALL Report AMP XI.M16A in SRP-LR Table 3.0-1. However, 10 CFR 54.21(d) provides the requirements for a Final Safety Analysis Report supplement and states, in part, that it must contain a summary description of the programs and activities for managing the effects of aging. The specificity of such descriptions will depend on the program proposed by each license renewal applicant.

35 II-20 Page A Table IV.B2 There is no need for specifying the Examination technique in the Program column.

Proposed Change: Delete The staff agrees with the comment that there is no need for specifying the Examination Technique in the "Aging Management Program" column of GALL Table IV.B2. GALL Tables IV.B2, IV.B3 and IV.B4 in final LR-ISG-2011-04 do not incorporate a summary of the examination techniques from the "Aging Management Program" column.

36 II-21 A Footnotes For note 6, see comments 15 and 16 above on why no justification for using VT-3 exam is required when it was acceptable in SER for 227. This applies to CE and B&W tables that also contain a similar note. Proposed Change: Delete the note The staff agrees with the comment that additional justification for the use of VT-3 to detect cracking is not needed if requirements specified in the SER and MRP are met. As a result of the staff's resolutions of Source ID I-4 and ID II-15, final LR-ISG-2011-04 does not incorporate SRP-LR Sections 3.1.2.2.9.A.7, 3.1.

2.2.9.C.1, and 3.1.2.2.9.C.4. However, the staff's position on the use of VT-3 for the detection of cracking will continue to be documented in the "Detection of Aging Effects" program element in GALL Report AMP XI.M16A.

In addition, as part of the staff's resolution to Source ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-RVI components and does not incorporate the footnotes in the "Further Evaluation" column of these tables.

C-16 # Source ID Summary of Comment Response 37 II-22 Page A-102 - Footnote #1 "In conjunction" is repeated in the second sentence.

Proposed Change: Delete second in conjunction The staff agrees with the comment that "in conjunction with" was an editorial error in Note 1. However, as part of t he staff's resolution to Source ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-RVI components and does not incorporate the footnotes in the "Further Evaluation" column of these tables. As a result of these revisions, the referenced Note 1 is not incorporated in final LR-ISG-2011-04.

38 II-23 Page A-104 - Footnote #8 4th line "No.2 above, and is so" should be and if so.

Proposed Change: Correct The staff agrees with the comment that there is a typographical error in Note 8 of page A-103. However, as part of the staff's resolution to Source ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-RVI components and does not incorporate the footnotes in the "Further Evaluation" column of these tables. As a result of these revisions, the referenced Note 8 is not incorporated in final LR-ISG-2011-04.

39 II-24 Page A Table IV.B2 Water chemistry is not listed as an AMP, with the aging effect of stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC) such as in line items IV.B2.RP-270a, 345, 399, 299a. This mainly occurs in new line items and also shows up in Table IV.B3 and IV.B4 Proposed Change: Add XI.M2 as an AMP The staff agrees with the comment that GALL Report AMP X.M2, "Water Chemistry," is not listed in GALL Table IV.B2. GALL Table IV.B2, IV.B3 and IV.B4 of final LR-ISG-2011-04 include GALL Report AMP X.M2, "Water Chemistry," as a recommendation to manage cracking by SCC, PWSCC, or IASCC, or loss of material due to pitting or crevice corrosion of RVIs. 40 II-25 Page A IV.B2.RP-399 As indicated in Table 4-9 of MRP-227-A and the associated note 2, the clevis insert bolts are inspected for wear. To the extent cracking would be visible in the VT-3 inspection, it would of course be addressed; but, the intent of the inspection is to look for wear.

Proposed Change: Eliminate this line as an existing inspection program element, or change the AMP description to note the inspection is for gross effects of cracking The staff agrees with the comment that Table 4-9 of MRP-227-A did not identify cracking as an aging effect requiring management for Westinghouse-design clevis insert bolts of screws but does not agree with the commenter's proposed change.

Relevant operating experience associated with aging may exist that has not been accounted for in MRP-227-A. AMR item IV.B2.RP-399 for cracking of Westinghouse-design clevis insert bolts and screws was included in LR-ISG-2011-04 based on industry operating experience. Appendix A of MRP-227-A states, in part, that "[f]ailures of Alloy X-750 clevis insert bolts were reported by one Westinghouse-designed plant in 2010" and "[a]lthough the failed clevis insert bolts were not removed for metallurgical examination, it can be surmised that the most likely cause of failure was PWSCC." No revisions were made as a result of this comment.

41 II-26 Page A IV.B2.RP-285 The staff agrees with the comment to delete the aging mechanism of loss of fracture toughness from AMR item IV.B2.RP-285. Since the clevis bolts C-17 # Source ID Summary of Comment Response As described in MRP-191, the clevis bolts and inserts are not in a high flux region and irradiation embrittlement is not a significant aging mechanism. As indicated in Table 4-9 of MRP-227-A and the associated note 2, the clevis insert bolts are inspected for wear. Also, Note 5 is applied to the further evaluation column; however, Note 5 refers to reduction of fracture toughness due to thermal embrittlement in stainless steel components, while the material listed for this line is nickel alloy.

Proposed Change: Eliminate the aging mechanism of loss of fracture toughness from this line and remove note 5 from the further evaluation column.

and inserts are not in a high flux region, GALL AMR Item IV.B2.RP-285 in final LR-ISG-2011-04 does not incorporate the aging effect of loss of fracture toughness due to neutron irradiation embrittlement.

As a result of the staff's resolution of Source ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-RVI components and deleted the footnotes in the "Further Evaluation" column of these tables.

42 II-27 Page A IV.B2.RP-345 As indicated in Table 5-1 of MRP-191, cracking of the core barrel flange is a concern for the weld rather than the base metal. Table 4-3 specifically identifies the welds as primary components to be inspected for cracking. While inspections of the welds would identify cracking in the adjacent base metal, separately adding cracking as an aging effect to the base metal as an existing component is not consistent with MRP-227-A or existing inspections.

Proposed Change: Eliminate base metal cracking as an aging effect in this line.

The staff agrees with the comment to delete base metal cracking from GALL AMR Item IV.B2.RP-345 of draft LR-ISG-2011-04 since MRP-227-A identifies that the adjacent base metal is part of the examination coverage for the "Core Barrel Assembly - Lower core barrel flange weld."

Thus, GALL AMR Item IV.B2.RP-345 in final LR-ISG-2011-04 does not reference cracking of the core barrel flange (base metal). GALL AMR IV.B2.RP-345 continues to identify loss of material due to wear for the core barrel flange (base metal).

43 II-28 Page A-9 Flaw evaluation methods include recommendations for flaw depth sizing and for crack growth determinations as well for performing applicable load limit. It should read - "growth determinations as well as for performing."

Proposed Change: Change to include missing "as."

The staff agrees with the comment that the sentence is incomplete. This sentence in the "Monitoring and Trending" program element of GALL Report AMP XI.M16A in final LR-ISG-2011-04 is complete.

44 II-29 Page A-49 In the first paragraph under Systems, Structures, and Components thermal shield assembly should be changed to thermal shield or neutron pad assembly to address the newer Westinghouse plants. Also, the component type neutron pad is not addressed in Table B2 or MRP-227.

The staff acknowledges that the component type neutron pad assembly is not addressed in GALL Table IV.B2 or MRP-227-A. However, the intent of this LR-ISG is not to supplement such aspects that are not covered in MRP-227-A. Thus, no revisions were made as a result of this comment.

If a PWR license renewal applicant identifies during the I PA that its plant design contains a neutron pad assembly (instead of a thermal shield assembly) and is subject to an AMR, the license renewal applicant must C-18 # Source ID Summary of Comment Response Proposed Change: Address recommended change.

identify this assembly in its LRA and propose an adequate means of aging management.

45 II-30 Table IV.B2 The environment "Reactor coolant and neutron flux" is used for all line items/components in Table B2, however not all the components listed in Table B2 will experience a neutron fluence exceeding 10 17 n/cm2 (E>1MeV) at the end of the period of extended operation. The environment should be more specific based on the location (fluence) of the components.

Proposed Change: The Table should note exceptions to the neutron fluence level.

The staff does not agree with the comment to note exceptions with regard to use the term "neutron flux" in GALL AMR items in the GALL Report.

The GALL Report generically and conservatively assumes that PWR RVIs are exposed to an environment of "reactor coolant and neutron flux" regardless of the fluence level. The staff anticipates that applicants will address their plant-specific data in their IPA and identify appropriate AMR items. No revisions were made as a result of this comment.

46 II-31 Page A Table 3.1-1 Item 27 Component was changed to nickel alloy guide tube support pins, however associated Table B2 line items IV.B2.RP-355 and IV.B2.RP-356 were changed to include both nickel alloy and stainless steel.

Proposed Change: Clarify The component in SRP-LR Table 3.1-1, Item 27, which refers to control rod guide tube (CRGT) split pins (support pins), is applicable to both nickel alloy and stainless steel materials. SRP-LR Table 3.1-1 Item 27 in draft LR-ISG-2011-04 was removed and incorporated into Table 3.1-1 Item 53c in final LR-ISG-2011-04. 47 II-32 A Last paragraph Sentence "EPR MRP methodology left some..." should be changed.

Proposed Change: Should read "EPRI MRP methodology left some..."

The staff agrees with the proposed change, however, as a result of the staff's resolution of Source ID I-1 the referenced sentence is not incorporated in final LR-ISG-2011-04.

48 II-33 The following acronyms are used but not included in Appendix B of this ISG; CUF, NRC, SE, and USAR.

Proposed Change: Update Appendix B to include all acronyms.

The staff agrees with the comment; however, draft LR-ISG-2011-04 was revised to remove the full list of acronyms in LR-ISG-2011-04, Appendix B. Final LR-ISG-2011-04, Appendix B, was revised to document the mark-up of changes to the GALL Report and SRP-LR. Acronyms in final LR-ISG-2011-04 are defined the first time they are used.

49 II-34 The page numbers for Appendix B are A-165 and A-166, the last page of Appendix A is A-144.

Proposed Change: Verify correct pagination.

The staff agrees with the comment and final LR-ISG-2011-04 includes the correct page numbers.

C-19 # Source ID Summary of Comment Response 50 II-35 Page A Section 3.1.2.2.9.A.5 For re-inspection greater than 10 years, further evaluation is redundant and inconsistent with standard GALL AMR and AMP formatting and presentation. Inspection frequencies would be evaluated in AMP element 4 for consistency with MRP-227-A chapter 4 primary, expansion, and existing components inspection tables. If the inspection frequency is identified that is not consistent with MRP-227-A Chapter 4 tables, an exception must be identified and justified.

Proposed Change: Delete further evaluation 3.1.2.2.

9.A item 5. Item to be addressed by AMP element 4.

The staff agrees with the comment that if an inspection frequency is not consistent with MRP-227-A, an exception must be identified and justified.

Furthermore, Section 4.0 of the staff's SE, Revision 1, for MRP-227 provides the "Conditions And Limitations And Applicant/Licensee Plant-Specific Action Items," which specifically states that the re-examination frequency for "Primary" inspection category components shall be on a maximum 10-year interval, unless a plant-specific analysis providing justification for an extended examination frequency is submitted to and approved by the NRC.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.5.

51 II-36 Page A Section 3.1.2.2.9.A.7 For VT-3 Inspection, further evaluation is redundant and inconsistent with standard GALL AMR and AMP formatting and presentation. VT-3 inspection requirements should be addressed as part of AMP element 3 for consistency with MRP-227-A requirements. Potential enhancements noted by the ISG further evaluation would be addressed by an AMP enhancement.

Proposed Change: Delete further evaluation 3.1.2.2.

9.A item 7. Item to be addressed by AMP element 3.

The staff agrees with the comment that VT-3 inspection requirements should be addressed as part of GALL Report AMP XI.M16A.

As a result of the staff's resolution of Source ID I-4, final LR-ISG-2011-04 does not incorporate SRP-LR Sections 3.1.2.2.9.A.7, 3.1.2.2.9.

C.1, and 3.1.2.2.9.C.4 related to VT-3 inspections. In addition, the staff's position on the use of VT-3 for the detection of cracking will continue to be documented in the "Detection of Aging Effects" program element in GALL Report AMP XI.M16A.

52 II-37 Page A Section 3.1.2.2.9.B.2 For Westinghouse Hold Down Springs, further evaluation is redundant and inconsistent with standard GALL AMR and AMP formatting and presentation. Definition of physical measurement techniques for Westinghouse hold down springs should be addressed as part of AMP element 3. Acceptance criteria for the hold down spring inspections would be addressed by AMP element

6. Proposed Change: Delete further evaluation 3.1.2.2.

9.B item 2. Item to be addressed by AMP elements 3 and 6. The staff agrees with the comment that physical measurement techniques and the inspection acceptance criteria for Westinghouse hold down springs are to be defined in an AMP.

The staff's SE, Revision 1, for MRP-227 documents the basis for limitations and conditions placed on the use of MRP-227 as well as licensee/applicant action items that shall be addressed by applicants/licensees who choose to implement the NRC-approved version of MRP-227. Specifically, A/LAI No. 5 of MRP-227-A addresses physical measurements of Westinghouse hold down springs.

As a result of the staff's resolution of Source ID I-1, areas resolved in the C-20 # Source ID Summary of Comment Response staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.B.2.

53 II-38 Page A IV.B2.RP-297 For CASS CRGT Lower Flanges, the ISG revision to the stainless steel definition in GALL Section IX.C requires that CASS be specifically designated in an AMR line item when thermal and neutron embrittlement susceptibility are identified. MRP-227-A Table 3-3 identifies the material of construction for CRGT lower flanges as CF-8 and thermal and neutron embrittlement identified as considerations for primary component classification.

Proposed Change: Identify CASS as an additional material in GALL IB.B2.RP-297 The staff agrees with the comment to add cast austenitic stainless steel (CASS) as a material in GALL AMR Item IV.B2.RP-297. In final LR-ISG-2011-04 the "Material" colu mn of GALL AMR Item IV.B2.RP-297 states "stainless steel, including CASS" and the "Aging Effect/Mechanism" column states "Loss of preload due to neutron irradiation embrittlement, and for CASS due to thermal aging embrittlement."

54 II-39 Page A IV.B2.RP-268 It appears that the primary purpose for the Inaccessible Locations AMR line item is to provide a further evaluation of inaccessible locations in partially accessible components susceptible to cracking due to SCC and IASCC using further evaluation note 3 (SRP-LR Section 3.1.2.2.9A Part A). This further evaluation is redundant to the note 3 further evaluation required by other AMR lines.

Proposed Change: Delete IV.B2.RP-268 The staff agrees with the comment to delete IV.B2.RP-268. As a result of the staff's resolution of Source ID I-7 and ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-RVI components. In addition, the footnotes in the "Further Evaluation" column of these tables are not incorporated into final LR-ISG-201 1-04. GALL AMR Items IV.B2.RP-268, IV.B3.RP-309 and IV.B4.RP-238 for Westinghouse, Combustion Engineering and Babcock and Wilcox designed plants, respectively, are not incorporated in final LR-ISG-2011-04. 55 II-40 Page A IV.B2.RP-269 It appears that the primary purpose for the Inaccessible Locations AMR line item is to provide a further evaluation of inaccessible locations in partially accessible components susceptible to Loss of fracture toughness due to neutron and irradiation embrittlement using further evaluation note 3 (SRP-LR Section 3.1.2.2.9A Part A). This further evaluation is redundant to the note 3 further evaluation required by other AMR lines Proposed Change: Delete IV.B2.RP-269 The staff agrees with the comment to delete IV.B2.RP-269. As a result of the staff's resolution of Source ID I-7 and ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-RVI components. In addition, the footnotes in the "Further Evaluation" column of these tables are not incorporated into final LR-ISG-2011-04. As a result, GALL AMR Items IV.B2.RP-269, IV.B3.RP-311 and IV.B4.RP-239 for Westinghouse, Combustion Engineering and Babcock and Wilcox designed plants, respectively, are not incorporated into final LR-ISG-2011-04.

C-21 # Source ID Summary of Comment Response 56 II-41 Page A IV.B2.RP-265 No additional measures (Cracking due to SCC and IASCC) in Section 3.3.1 of MRP-227-A defines the no additional measures category as: those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria. Additional components were placed in the No Additional Measures group as a result of FMECA and the functionality assessment. No further action is required by the MRP-227-A for managing the aging of the No Additional Measures components. Simply put, there are no aging effects requiring aging management.

Proposed Change: Change the aging effect column and AMP column for IV.B2.RP-265 to be consistent with other GALL AMR "none-none" line items and move the lines to GALL Section IV.E, Common Miscellaneous Material Environment Combinations.

The staff does not agree with the comment to change GALL AMR Item IV.B2.RP-265 to be consistent with other GALL AMR "none-none" line items and the statement that there are no aging effects requirement management.

The "No Additional Measures" category of components in MRP-227-A does not equate to such components not having an aging effect requiring management; it only indicates that MRP-227-A does not include guidance to manage aging for components categorized as "No Additional Measures." Thus, the staff agrees with the commenter's following statement that "[n]o further action is required by MRP-227-A for managing the aging of the No Additional Measures components." The IPA is independent of MRP-227-A and may identify applicable aging effects to manage, which may be broader than the aging effects identified in MRP-227-A for RVIs. Thus, the "No Additional Measures" category of components in MRP-227-A does not alleviate the requirements in 10 CFR 54.21(a)(3).

In any event, the staff acknowledges that GALL AMR Items IV.B2.RP-265, IV.B2.RP-267, IV.B3.RP-306, IV.B3.RP-307, IV.B4.RP-236 and IV.B4.RP-237 caused confusion; thus, final LR-ISG-2011-04 does not incorporate GALL AMR Items IV.B3.RP-307, IV.B4.RP-236 and IV.B4.RP-237. In addition, GALL AMR Items IV.B2.RP-265, IV.B2.RP-267 and IV.B3.RP-306 in final LR-ISG-2011-04 clarify that there is no additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists.

57 II-42 Page A IV.B2.RP-267 No additional measures (Loss of fracture toughness due to neutron and irradiation embrittlement) in Section 3.3.1 of MRP-227-A defines the no additional measures category as: those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria. Additional components were placed in the No Additional Measures group as a result of FMECA and the functionality assessment. No further action is required by the MRP-227-A for managing the aging of the No Additional Measures components. Simply put, there are no aging effects requiring aging management.

Proposed Change: Change the aging effect column and AMP column for IV.B2.RP-267 to be consistent with other GALL AMR "none-none" line items and move the lines to GALL Section IV.E, The staff does not agree with the comment to change GALL AMR Item IV.B2.RP-267 to be consistent with other GALL AMR "none-none" line items and that there are no aging effects requirement management.

As a result of the staff's resolution for Source ID II-41, final LR-ISG-2011-04 does not incorporate GALL AMR Items IV.B3.RP-307, IV.B4.RP-236 and IV.B4.RP-237. In addition, GALL AMR Items IV.B2.RP-265, IV.B2.RP-267 and IV.B3.RP-306 in final LR-ISG-2011-04 clarify that there is no additional aging management for reactor internal "No Additional Measures" components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists.

C-22 # Source ID Summary of Comment Response Common Miscellaneous Material Environment Combinations.

58 II-43 Page A-6 Clarification is needed relative to the relationship between the SRP-LR and the GALL documents.

The staff noted draft LR-ISG-2011-04 caused confusion between the relationship of the SRP-LR and the GALL Report for PWR RVI components. As a result, final LR-ISG-2011-04 does not reference the SRP-LR in GALL Report AMP XI.M16A in order to be consistent with the format of other AMPs in the GALL Report. 59 II-44 Page A-11 Wording awkward Proposed Change: Delete "that" at the beginning of line 8.

The staff agrees with the comment to delete the word "that" from the "Confirmation Process" program element of GALL Report AMP XI.M16A in draft LR-ISG-2011-04. The staff revised this program element to state, in part, "- for confirming the quality of inspections, flaw evaluations, and corrective actions performed under this program." 60 II-45 Page A-17 There is a concern that "monitoring and trending" program elements and "corrective action" program elements are buried in the Acceptance Criteria section.

The staff agrees with the comment that there is a concern the "monitoring and trending" and "corrective actions" program elements are buried in the Acceptance Criteria section.

As a result of the staff's resolution of Source ID I-1 and II-8, the staff revised LR-ISG-2011-04 so that areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.4.

61 II-46 Page A-20 and A-21 The statement "To satisfy the requirements of ASME Code Section III-." is confusing if not all plants are committed to Subsection NG.

Proposed Change: The statement should be modified to include a qualifying statement like "if the plant is committed to Subsection NG." The staff does not agree with the comment to alter the referenced statement, as it comes from the staff's SE, Revision 1, on MRP-227. The topic of environmentally-assisted fatigue for PWR RVIs is addressed in A/LAI No. 8, Item 5 of MRP-227-A. Section 3.0 of the staff's SE, Revision 1, on MRP-227 documents the basis for limitations and conditions being placed on the use of MRP-227 as well as licensee/applicant action items that shall be addressed by applicants/licensees who choose to implement the NRC-approved version of MRP-227. Revisions to the conditions and limitations, applicant/licensee plant-specific action items, and conclusions of the staff's SE, Revision 1, for MRP-227 are not within the scope of LR-ISG-2011-04.

However, as a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.9.

C-23 # Source ID Summary of Comment Response 62 II-47 Page A Section 3.1.2.2.9.D.1 The intended meaning of the word "appropriately" in D.1, second paragraph. Is not clear.

Proposed Change: Clarify meaning The staff agrees that the referenced sentence in SRP-LR Section 3.1.2.2.9.D.1 is not clear. However, as a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.D.1.

63 II-48 Page A-142 Spell out variable name in 1.0 of "Further Evaluation Recommendations" The staff noted that as a result of the resolution to Source ID I-7, final LR-ISG-2011-04 does not incorporate the referenced variable names and further evaluation footnotes in GALL Tables IV.B2, IV.B3, and IV.B4.

64 II-49 Page A-5 Each of the following documents provide information for submittal of an AMP and inspection plan:

• GALL Revision 2 (page XI.M16A-2) and LR-ISG-2011-04 (draft, page A-5)

• RIS 2011-07 (page 6)

• Safety Evaluation Revision 1 for MRP-227 (page 34)

• Section 3.5.1 of the Safety Evaluation (page 25)

It is unclear what actually goes into the LRA and the format. The above verbiage implies that the AMP and inspection plan are separate documents that are submitted with the application but are reviewed and approved by the NRC as unique documents. A quick search of the GALL indicates that PWR Vessel Internals is the only program that requires the AMP and an inspection plan to be submitted for NRC review and approval.

Proposed Change: Commenter provided revisions to Section 3.5.1 of Safety Evaluation Revision 1 for MRP-227.

The staff disagrees with the comment because revisions to the conditions and limitations, applicant/licensee plant-specific action items and conclusions of the staff's SE, Revision 1, for MRP-227 are not within the scope of LR-ISG-2011-04.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. In addition, see the staff's resolution of Source ID I-3, in which the staff discusses the staff's position/guidance regarding inspection plans that is documented in RIS 2011-07 dated July 21, 2011. The staff expects that the details of an "inspection plan" for Category D plants (defined in RIS 2011-07) will be incorporated into the LRA submittal as part of the 10-element AMP and AMR line items. Thus, consistent with RIS 2011-07, the staff does not expect Category D plants to provide a separate document that contains an "inspection plan" in response to A/LAI No. 8. 65 II-50 Page A-6 GALL Rev 2 (page XI M16A-3) states: The responses to the LR A/LAIs on MRP-227 are provided in Appendix C of the LRA.

LR-ISG-2011-04 (page A-6) deleted this requirement, however LR-ISG-2011-04 (page A-14, 15) states to provide responses to the A/LAIs in Appendix C of the LRA, and to address SRP-LR further evaluation "acceptance criteria" that are based on these A/LAIs. It is The staff agrees with the comment that it is unclear where the A/LAIs should be addressed. As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.1 and also does not incorporate a discussion of A/LAIs in GALL AMP XI.M16A.

C-24 # Source ID Summary of Comment Response unclear where the licensee action items should be addressed. Wording implies that the applicant/licensee action items should be addressed in Appendix C and in the associated further evaluation section. Proposed Change: The Commenter provided revisions to LR-ISG-2011-04 (page A-6).

66 II-51 XI.M16A, PWR Vessel Internals elements 1. Scope of Program, 5. Monitoring and Trending, and 6. Acceptance Criteria refer to the latest NRC approved version of WCAP-17096-NP and the associated applicant/licensee action items. It is our understanding that WCAP-17096-NP has been submitted for approval however it has not been approved at this time. A program cannot be developed based on an unapproved document or unknown A/LAIs.

Proposed Change: Remove any reference to WCAP-17096-NP or delay issuance of LR-ISG-2011-04 until WCAP-17096-NP is approved by the NRC.

The staff agrees with the comment to delete any reference to WCAP-17096-NP since the report has been submitted for review but not approved by staff. Final LR-ISG-2011-04 does not reference WCAP-17096-NP. 67 II-52 Many of the A/LAIs specified in the Acceptance Criteria section of LR-ISG-2011-04 request that the applicant make enhancements or augmented enhancements to various program elements as a result of the responses to the "further evaluations." It would be simpler if the NRC specified an acceptable method of addressing an issue in the XI.M16A program elements and then if the licensee/applicant wanted to do something different take an exception rather than requiring each licensee/applicant to develop a unique set of enhancements for their program.

Proposed Change: Revise A/LAIs to clearly state that additional justification/information is only required to be included in the "further evaluation" responses if the applicant/licensee is deviating from the requirements of MRP-227-A.

The staff disagrees with the comment to revise A/LAIs in LR-ISG-2011-04, as it is a direct reference to the staff's SE, Revision 1, for MRP-227. See the staff's resolution to Source ID II-49, in which the staff explains that revisions to the conditions and limitations, applicant/licensee plant-specific action items and conclusions of the staff's SE, Revision 1, for MRP-227 are not within the scope of LR-ISG-2011-04. No revisions were made as a result of this comment.

68 II-53 A simplified method of addressing reactor internals in the GALL tables B.2, B.3, and B.4 would be to have line items based on component classification (Primary, Expansion, Existing, and No Additional Measures) as defined in MRP-227-A rather than individual component types (Alignment and Interfacing components: internals hold down spring, Alignment and interfacing components: upper core plate alignment pins, etc). Making this change would allow multiple line items to apply to several component types and The staff agrees with the comment, in part, that GALL AMR Tables IV.B2, IV.B3 and IV.B4 can be simplified. However, the staff does not agree with the commenter's proposed change.

As explained in the staff's resolution of Source ID I-6, the AMR line items in the GALL Report and MRP-227-A do not solely serve as the basis for determining components or aging effects that require management or establish the AMR line items to be included in an LRA. The IPA required C-25 # Source ID Summary of Comment Response reduce the number of Table 2 line items simplifying this section.

Proposed Change: Revise NUREG-1801 tables B.2, B.3, and B.4 to have line items associated with component classification (Primary, Expansion, Existing Program, and No Additional Measures) and refer to MRP-227-A for the specific components in the four classification groups.

by 10 CFR 54.21(a) is independent of the AMR line items provided in MRP-227-A and the GALL Report. It is not necessary for the staff to correlate the number and contents of AMR items in GALL Tables IV.B2, IV.B3, and IV.B4 exactly to the number and contents of inspection items in MRP-227-A.

In any event, as a result of the staff's resolution of Source ID I-1, GALL AMR Tables IV.B2, IV.B3 and IV.B4 were revised. See resolution of Source ID I-1 for a summary of the revisions.

69 II-54 Several of the applicant/licensee action items (A/LAI) identified in the Safety Evaluation for MRP-227 (pages 32 - 34) required plant-specified evaluations or analyses to be submitted as part of the application. A/LAI Number 5 requires the applicant/licensee include its proposed acceptance criteria and an explanation of how the proposed acceptance criteria are consistent with the plants' licensing basis and the need to maintain the functionality of the component being inspected under all licensing basis conditions of operation during the period of extended operation as part of their submittal to apply the approved version of MRP-227. A/LAI Number 7 requires a plant-specific analysis to be performed on Westinghouse lower support column bodies made of CASS be included as part of their submittal to apply the approved version of MRP-227.

Proposed Change: Revise A/LAI Numbers 5 and 7 to allow for the applicants/licensees to commit performing an analysis prior to the period of extended operation. This would allow applicants/licensees that are submitting in the near future (2013 timeframe) to perform the analyses on normal schedule rather than an expedited schedule.

The staff disagrees with the comment to revise A/LAIs in LR-ISG-2011-04, as it is a direct reference to the staff's SE on MRP-227-A. See the staff's resolution to Source ID II-49, in which the staff explains that revisions to the conditions and limitations, applicant/licensee plant-specific action items and conclusions of the staff's SE, Revision 1, for MRP-227 are not within the scope of LR-ISG-2011-04. No revisions were made as a result of this comment. 70 II-55 Page A-17 In the last paragraph it states: For those component inspections that do not achieve the inspection coverage criteria stated in the NRC SE (Revision 1) on MRP-227, the applicant is requested to identify a deviation from the MRP-defined inspection criteria and describe the process and type of evaluation that will be implemented to evaluate the impact of the aging effects on the integrity of those components in the population that were inaccessible to the inspection technique, and to identify this process as an applicable enhancement of the "monitoring and trending" program element of its RVI Program. The staff agrees with the comment that the referenced statement in SRP-LR Section 3.1.2.

2.9.A.4 of draft LR-ISG-2011-04 is more appropriately addressed in the AMP.

As a result of the staff's resolution of Source ID I-1, areas resolved in the staff's SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.4.

C-26 # Source ID Summary of Comment Response The SRP is specifying actions for applicants to perform as part of an aging management program which is more appropriately addressed within program requirements.

Proposed Change: These actions should be included as program elements, not in the further evaluation sections of the SRP.

71 II-56 Page A-19 and A Section 3.1.2.2.9.A.9 In the third paragraph the further evaluation states: "To satisfy the requirements of the ASME Code,Section III, Subsection NG-2160 and NG-3121, the existing fatigue CUF analysis shall include the effects of the reactor coolant water environment."

The December 26, 1999, Generic Safety Issue (GSI) 190 close-out memorandum from Ashok C. Thadani, Director of the Office of Regulatory Research, to William D. Travers, Executive Director for Operations, provides the basis for consideration of the effects of the reactor coolant water environment. It should be noted that GSI-190 concerns are limited to pipe leakage, which is not applicable to RVI components since they do not form a portion of the reactor coolant pressure boundary and are therefore not subject to leakage.

Proposed Change: Delete the referenced sentence in the third paragraph of Further Evaluation A.9.

The staff does not agree with the comment, in particular the inference that, the effects of the reactor coolant water environment on metal fatigue are not applicable to RVI components since they do not form a portion of the reactor coolant pressure boundary.

See the staff's resolution to Source ID II-46, in which the staff explains that environmentally-assisted fatigue for PWR RVIs is addressed specifically in A/LAI No. 8, Item 5 of the staff's SE, Revision 1, on MRP-227. Final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.9 72 II The comments submitted by EPRI-MRP, the PWROG-MSC, and NEI are extensive and involve complex issues. EPRI and PWROF, along with NEI, respectfully request a follow-up meeting with the NRC staff to discuss resolution of the comments and, if appropriate, an additional comment period.

The NRC staff acknowledges the complexity of the issues captured in MRP-227-A. However, LR-ISG-2011-04 is not intended to supplement, modify or further resolve the issues raised in MRP-227-A, but rather to reference MRP-227-A and the associated staff's SE, Revision 1, for MRP-227, in the usable format of a generic aging management program, as described in the GALL Report. To the extent that comments provided suggestions to clarify or simplify the format of LR-ISG-2011-04 for ease of use, the staff was able to incorporate those changes into the final document. However, to the extent that comments proposed changes to the actual content of the staff's SE, Revision 1, for MRP-227, the staff did not incorporate those comments, as it is beyond the scope and intent of LR-ISG-2011-04. The staff also does not perceive further benefits from an additional public meeting and comment period to resolve the latter set of comments, as it is beyond the scope of LR-ISG-2011-04.