ML15337A314

Official Exhibit - ENTR00615-PUB-00-BD01 - Entergy'S Statement of Position Re Contention NYS-25 (Embrittlement) - Redacted

ML15337A314
Person / Time
Site:	Indian Point
Issue date:	09/04/2015
From:	Entergy Nuclear Operations
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 28300, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15337A314 (63)

Text

United States Nuclear Regulatory Commission Official Hearing Exhibit In the Matter of: Entergy Nuclear Operations, Inc.

(Indian Point Nuclear Generating Units 2 and 3)

ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #: ENTR00615-PUB-00-BD01 Identified: 11/5/2015 ENTR00615 Admitted: 11/5/2015 Rejected:

Withdrawn:

Stricken:

Revised: September 4, 2015 Other:

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) Docket Nos. 50-247-LR and

) 50-286-LR ENTERGY NUCLEAR OPERATIONS, INC. )

)

(Indian Point Nuclear Generating Units 2 and 3) )

) September 4, 2015 ENTERGYS STATEMENT OF POSITION REGARDING CONTENTION NYS-25 (EMBRITTLEMENT)

William B. Glew, Esq. Kathryn M. Sutton, Esq.

Entergy Nuclear Operations, Inc. Paul M. Bessette, Esq.

440 Hamilton Avenue Raphael P. Kuyler, Esq.

White Plains, NY 10601 MORGAN, LEWIS & BOCKIUS LLP Phone: (914) 272-3202 1111 Pennsylvania Avenue, N.W.

Fax: (914) 272-3205 Washington, D.C. 20004 E-mail: wglew@entergy.com Phone: (202) 739-3000 Fax: (202) 739-3001 E-mail: ksutton@morganlewis.com E-mail: pbessette@morganlewis.com E-mail: rkuyler@morganlewis.com Counsel for Entergy Nuclear Operations, Inc.

TABLE OF CONTENTS Page I. PRELIMINARY STATEMENT ....................................................................................... 1 A. Dr. Laheys Claims Regarding Synergistic Effects Lack Merit ......................... 4 B. The RVI AMP Appropriately Addresses Shock Loads...................................... 6 C. The RVI AMP Is Fully Adequate as It Specifies Appropriate Preventative Actions, Corrective Actions, and Acceptance Criteria .......................................... 6 D. The IPEC Fatigue Analyses Are Conservative and Support the Finding that the Effects of Fatigue Will Be Adequately Managed ..................................... 8 E. NYS Does Not Challenge the Adequacy of the LRA with Respect to RPVs........ 8 II. PROCEDURAL HISTORY OF CONTENTION NYS-25 ............................................... 9 A. Original Contention ............................................................................................ 10 B. First Amended Contention ................................................................................... 11 C. Revised RVI AMP and Inspection Plan............................................................... 13 D. 2011 NYS Testimony .......................................................................................... 14 E. Second Amended Contention .............................................................................. 14 III. APPLICABLE LEGAL AND REGULATORY STANDARDS ................................... 16 A. NYS Continues Its Attempts to Impermissibly Expand the Scope of this License Renewal Proceeding ............................................................................... 16

1. The License Renewal Review Is a Limited One ...................................... 16

2. The Reasonable Assurance Standard ....................................................... 19 B. License Renewal Guidance .................................................................................. 20

1. NUREG-1801 Is Entitled to Special Weight in This Proceeding ............ 22

2. Revisions to NUREG-1801...................................................................... 23 C. Burden of Proof.................................................................................................... 25 IV. ENTERGYS WITNESSES ............................................................................................ 26 A. Mr. Nelson F. Azevedo ........................................................................................ 26 B. Mr. Robert J. Dolansky ........................................................................................ 27 C. Mr. Alan B. Cox................................................................................................... 28 D. Mr. Jack R. Strosnider, Jr..................................................................................... 29 E. Mr. Timothy J. Griesbach .................................................................................... 30 F. Dr. Randy G. Lott ................................................................................................ 30 G. Mr. Mark A. Gray ................................................................................................ 31 V. ENTERGYS EVIDENCE AND ARGUMENTS ......................................................... 32

TABLE OF CONTENTS Page A. Technical Background on the Aging Management of RVIs and RPVs............... 32 B. Regulatory Guidance Addressing Aging Management of RVIs.......................... 35 C. Entergys LRA Effectively Addresses Aging Management of RVIs and RPVs .................................................................................................................... 39

1. The IPEC RVI AMP ................................................................................ 40

a. Overview of the RVI AMP and Inspection Plan ........................ 40

b. The RVI AMP Describes Inspections in Detail ........................... 40

c. The RVI AMP Manages the Effects of Aging on RVIs Regardless of the Underlying Aging Mechanism ........................ 41

d. The RVIs Are Robust and Highly Failure Tolerant ..................... 44

e. The RVI AMP Addresses Combinations of Aging Effects from Multiple Degradation Mechanisms ..................................... 46

f. The RVI AMP Addresses Appropriate Design Basis Loads, Including Seismic and LOCA Loads ........................................... 49

g. The RVI AMP Uses Appropriate Inspection Techniques............ 51

h. The RVI AMP Includes Appropriate Acceptance Criteria, Corrective Actions, and Preventive Actions ................................ 52

i. The IPEC Fatigue Evaluations Appropriately Analyze Environmentally-Assisted Fatigue ............................................... 55

j. The RVI AMP Experience Addresses Operating Experience ................................................................................... 57

2. Entergys Aging Management Activities for RPVs................................. 58 VI. CONCLUSION ................................................................................................................ 59

- ii -

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) Docket Nos. 50-247-LR and

) 50-286-LR ENTERGY NUCLEAR OPERATIONS, INC. )

)

(Indian Point Nuclear Generating Units 2 and 3) )

) September 4, 2015 ENTERGYS STATEMENT OF POSITION REGARDING CONTENTION NYS-25 (EMBRITTLEMENT)

Pursuant to 10 C.F.R. § 2.1207(a)(1) and the Atomic Safety and Licensing Boards (Board) Revised Scheduling Order,1 Entergy Nuclear Operations, Inc. (Entergy) submits this Statement of Position (SOP) regarding Contention NYS-25 proffered by New York State (NYS or the State). This Statement is supported by the Testimony of Nelson F. Azevedo, Robert J. Dolansky, Alan B. Cox, Jack R. Strosnider, Jr., Timothy J. Griesbach, Randy G. Lott, and Mark A. Gray Regarding Contention NYS-25 (Embrittlement) (Entergys Testimony)

(ENT000616), and the exhibits thereto (ENT000617 through ENT000721). For the reasons discussed below, NYS-25 lacks merit and should be resolved in Entergys favor.

I. PRELIMINARY STATEMENT NYS-25 is a safety contention, asserting that Entergys License Renewal Application does not include an adequate plan to monitor and manage the effects of aging due to embrittlement of the reactor pressure vessels (RPVs) and the associated internals.2 The 1

Licensing Board Revised Scheduling Order at 2 (Dec. 9, 2014) (unpublished) (Revised Scheduling Order).

2 New York State, Notice of Intention to Participate and Petition to Intervene at 223 (Nov. 30, 2007) (NYS Petition).

testimony of the States sole witnessDr. Richard T. Laheyfocuses on purported deficiencies in the aging management program (AMP) for reactor vessel internals (RVIs) at Indian Point Nuclear Generating Units 2 and 3 (IP2 and IP3, collectively Indian Point Energy Center or IPEC).

The States claims and testimony in NYS-25 are cumulative and overlapping with other contentions, redundant in some areas and contradictory in others. Such an approach is not only undisciplined, but also contrary to the Commissions intent in requiring intervenors to bring forward well-defined and adequately-supported contentions so that other parties to the proceeding are given full and fair notice of the intervenors actual claims.3 In NYS-25, the State claims that the license renewal application (LRA) is deficient for four reasons: (1) the RVI AMP is not based on an analysis that addresses synergistic degradation of RVIs caused by combinations of degradation mechanisms; (2) Entergy fails to consider the full range of transient shock loads to which RVIs may be subjected in the event of various postulated accidents; (3) the RVI AMP does not include a commitment to take preventative actions or to implement corrective actions, or provide specific enforceable acceptance criteria for some components; and (4) the AMP relies on fatigue predictions which are non-conservative and may not accurately predict fatigue-induced component failures.4 As demonstrated below, all four of these claims lack merit because Dr. Lahey and the State unfortunately have disregardedrather than disputedthe substantial body of technical 3

Pub. Serv. Co. of N.H. (Seabrook Station, Units 1 & 2), ALAB-899, 28 NRC 93, 97 (1988), aff'd sub nom.

Massachusetts v. NRC, 924 F.2d 311 (D.C. Cir. 1991), cert. denied, 502 U.S. 899 (1991).

4 See State of New York, Revised Statement of Position, Contention NYS-25 at 17 (June 9, 2015) (NYS Revised SOP) (NYS000481).

work that supports the IPEC RVI AMP. They have done so, despite the longstanding availability of the underlying technical work and supporting documentation, to the peril of their arguments.5 The IPEC RVI AMP is based on sound, state-of-the-art science and is fully compliant with the applicable criteria in NUREG-1801, as updated in the latest interim staff guidance. Dr.

Lahey demands a systematic safety evaluation to support the RVI AMP.6 He ignores, however, the systematic evaluation that the Electric Power Research Institute (EPRI) has already performed, and that Entergy relies upon. This horse-blinder approach falls woefully short of the requisite specific and substantial support7 necessary to overcome the special weight the Board must accord to Nuclear Regulatory Commission (NRC) Staff guidance, which endorses the approach Entergy has taken at IPEC.8 Moreover, several aspects of contention NYS-25 impermissibly challenge the current licensing basis (CLB) for IPEC.9 In summary, the State has failed to meet its burden of moving forward with sufficient evidence to show a deficiency in the RVI AMP,10 and Entergys Testimony fully refutes the States claims.

5 As explained further below, NYS retains the burden of going forward with its contention, even at the hearing stage. See AmerGen Energy Co., LLC (Oyster Creek Nuclear Generating Station), CLI-09-07, 69 NRC 235, 268-70 (2009), affd sub nom. N.J. Envtl. Fedn v. NRC, 645 F.3d 220 (2011).

6 Revised Pre-Filed Written Testimony of Dr. Richard T. Lahey, Jr., Regarding Contention NYS-25 at 51 (June 9, 2015) (NYS000482) (Revised Lahey Testimony).

7 See Entergy Nuclear Vt. Yankee, L.L.C. & Entergy Nuclear Operations, Inc. (Vt. Yankee Nuclear Power Station), CLI-10-17, 72 NRC 1, 33 n.185 (2010).

8 See Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 & 3), CLI-15-6, 81 NRC __,

slip op. at 19 (Mar. 9, 2015); NextEra Energy Seabrook, LLC (Seabrook Station, Unit 1), CLI-12-5, 75 NRC 301, at 314 n.78 (2012).

9 See Entergy Nuclear Generation Co. & Entergy Nuclear Operations, Inc. (Pilgrim Nuclear Power Station),

CLI-10-14, 71 NRC 449, 461 (2010); Oyster Creek, CLI-09-7, 69 NRC at 270.

10 See Oyster Creek, CLI-09-7, 69 NRC at 269.

Accordingly, Entergy has met its burden of showing, by a preponderance of the evidence,11 that NYS-25 lacks merit and should be resolved in Entergys favor.

A. Dr. Laheys Claims Regarding Synergistic Effects Lack Merit The States first claim is Dr. Laheys purported discover[y] that the IPEC RVI AMP fails to address potential synergistic degradation caused by combinations aging mechanisms.12 The IPEC RVI AMP, however, is based on a decade-long systematic expert evaluation of known and potential degradation mechanisms, resulting aging effects, and consequences of those effects for RVIs.13 This evaluation considered the relevant aging mechanisms, including multiple aging mechanisms which can produce combined effects on RVI components.14 NYS and Dr. Lahey have largely ignored the substantial state-of-the-art engineering and technical basis for the RVI AMP contained in the EPRI Materials Reliability Programs (MRP) MRP-227-A, Pressurized Water Reactor Internal Inspection and Evaluation Guidelines,15 its numerous supporting technical reports, and the plant-specific technical analyses submitted by Entergy for Indian Point and reviewed by the NRC Staff in SSER 2.16 11 See Pac. Gas & Elec. Co. (Diablo Canyon Nuclear Power Plant, Units 1 & 2), ALAB-763, 19 NRC 571, 577 (1984); Oyster Creek, CLI-09-07, 69 NRC at 263.

12 Revised Lahey Testimony at 78 (NYS000482).

13 See Testimony of Entergy Witnesses Nelson F. Azevedo, Robert J. Dolansky, Alan B. Cox, Jack R. Strosnider, Timothy J. Griesbach, Randy G. Lott, and Mark A. Gray Regarding Contention NYS-25 (Embrittlement)

§ VI.B (Aug. 10, 2015) (Entergys Testimony) (ENT000616).

14 See id.

15 MRP-227-A, EPRI Materials Reliability Program: Pressurized Water Reactor Internal Inspection and Evaluation Guidelines (Dec. 23, 2011) (MRP-227-A) (NRC000114A-F).

16 NUREG-1930, Supp. 2, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3 (Nov. 6, 2014) (SSER 2), available at ADAMS Accession No. ML14310A803.

In their testimony, Entergys expert witnesses explain that the IPEC RVI AMP is consistent with MRP-227-A, which is endorsed in current NRC guidance17 and incorporated in AMP XI.M16A in NUREG-1801 (the GALL Report).18 As a matter of law, the Commission has held that:

[A] license renewal applicant who commits to implement an AMP that is consistent with the corresponding AMP in the GALL Report has demonstrated reasonable assurance under 10 C.F.R. § 54.29(a) that the aging effects will be adequately managed during the period of extended operation.19 Because the IPEC RVI AMP is consistent with the current GALL Report AMP,20 it satisfies the regulatory requirements in 10 C.F.R. Part 54. Although an intervenor may challenge whether an applicants AMP is consistent with NRC Staff guidance,21 the State and Dr. Lahey do not make such an allegation. Instead, NYS and Dr. Lahey only proffer generic attacks against the NRC guidance and MRP-227-A on the topic of allegedly synergistic aging effects.22 But as explained further below, NYS must provide specific and substantial support23 to overcome the special weight accorded to this guidance in this proceeding.24 As Entergys testimony shows, NYS and Dr. Lahey have failed to clear that high hurdle.

17 LR-ISG-2011-04, Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors (Mar. 20, 2012) (LR-ISG-2011-04) (ENT000641).

18 See generally NUREG-1801, Generic Aging Lessons Learned Report, Revision 2 (Dec. 2010) (NUREG-1801, Rev. 2) (NYS00147A-D); NUREG-1801, Generic Aging Lessons Learned Report, Revision 1 (Sept.

2005) (NUREG-1801, Rev. 1) (NYS00146A-C).

19 Seabrook, CLI-12-5, 75 NRC at 315 see also AmerGen Energy Co., LLC (Oyster Creek Nuclear Generating Station), CLI-08-23, 68 NRC 461, 468 (2008).

20 See Entergys Testimony § VII.A (ENT000616).

21 Vt. Yankee, CLI-10-17, 72 NRC at 37.

22 Revised Lahey Testimony at 39 (NYS000482) (MRP-227-A is an inspection-based aging management plan, which I believe is inadequate).

23 See Vt. Yankee, CLI-10-17, 72 NRC at 33 n.185, 37.

24 Seabrook, CLI-12-5, 75 NRC at 314 n.78.

B. The RVI AMP Appropriately Addresses Shock Loads As to the States second claim, that Entergy has failed to consider potential shock loads that could impact RVI components, the inspection guidelines in MRP-227-A are designed to provide reasonable assurance that the RVIs will continue to perform their intended functions, consistent with the CLB.25 This includes maintaining functionality under CLB accident and seismic loads (or shock loads, as Dr. Lahey describes them).26 The design basis loads are established in accordance with the CLB and do not change because the units will operate during the period of extended operation (PEO).27 The guidelines in MRP-227-A, however, include inspections to provide reasonable assurance that components are not degraded due to the effects of aging, and provide instructions to guide engineering evaluations to determine the functionality of RVI componentsunder CLB loadsif any degradation is discovered.28 Accordingly, the IPEC RVI AMP fully accounts for CLB accident and transient loads.

Dr. Lahey has not discovered anything new.29 He instead impermissibly challenges the CLB.

To the extent his concerns are with shock loads caused by beyond design-basis accidents, this is an impermissible challenge to the CLB.30 C. The RVI AMP Is Fully Adequate as It Specifies Appropriate Preventive Actions, Corrective Actions, and Acceptance Criteria Next, NYS argues that the IPEC RVI AMP does not include a commitment to take preventative actions or to implement corrective actions, or provide specific enforceable 25 See Entergys Testimony § VII.A.6 (ENT000616).

26 See id. As explained further below, the adequacy of the CLB is not subject to attack in this proceeding. See Pilgrim, CLI-10-14, 71 NRC at 461; Oyster Creek, CLI-09-7, 69 NRC at 270.

27 See Entergys Testimony at A115 (ENT000616).

28 See id. § VI.B.

29 Revised Lahey Testimony at 78 (NYS000482).

30 See Pilgrim, CLI-10-14, 71 NRC at 461; Oyster Creek, CLI-09-7, 69 NRC at 270.

acceptance criteria for some components.31 Entergys witnesses demonstrate that these claims also lack merit.32 For example, Dr. Lahey and the State claim that the IPEC RVI AMP manifestly does not include preventative actions.33 They base this claim, however, on a crabbed interpretation of a single phrase lifted from the IPEC RVI AMP and taken out of context. In actuality, the remainder of the section discusses the preventive actions Entergy is taking related to RVIs some of which are being taken under the aegis of other AMPs at IPEC.34 In response to NYS other claims, Entergys witnesses fully demonstrate the adequacy and specificity of the inspection schedules, corrective actions, and acceptance criteria in the RVI AMP.35 As for Dr.

Laheys and the States concerns about the scope of the programthe components they identify are either not RVIs, and are covered by other aging management programs,36 or are active components outside the scope of aging management review (AMR) altogether.37 So yet again, NYS has not shouldered its evidentiary burden with respect to its claims in NYS-25.

31 NYS Revised SOP at 17 (NYS000481).

32 See Entergys Testimony § VII.A.7-8 (ENT000616).

33 NYS Revised SOP at 26 ¶ 21 (quoting NL-12-037, Letter from F. Dacimo, Vice President, Entergy, to NRC Document Control Desk, License Renewal Application - Revised Reactor Vessel Internals Program and Inspection Plan Compliant with MRP-227-A, Attach. 1 at 5 (Feb. 17, 2012) (NL-12-037) (NYS000496));

see also Revised Lahey Testimony at 53 (NYS000482).

34 See NL-12-037, Attach. 1 at 5 (NYS000496).

35 See Section V.C.1, infra.

36 For example, Dr. Laheys concerns about the j-groove welds, Revised Lahey Testimony at 45 (NYS000482),

are about a component that is managed under the Reactor Vessel Head Penetration Inspection AMP, not the RVI AMP. See Entergys Testimony at A101 (ENT000616).

37 For example, as explained below, control rods are active, short-lived components. See Indian Point, CLI-15-6, slip op. at 8; 10 C.F.R. § 54.21(a)(1).

D. The IPEC Fatigue Analyses Are Conservative and Support the Finding that the Effects of Fatigue Will Be Adequately Managed Finally, the States and Dr. Laheys claim that the environmentally-assisted fatigue (EAF) evaluations prepared in support of the IPEC LRA are not conservative and may be inaccurate lacks merit.38 Entergys EAF evaluations, including EAF evaluations of RVI components, are fully documented, conservative engineering analyses that support a finding that the effects of fatigue, including the effects of the reactor water environment, will be adequately managed.39 There is no technical basis to conclude that an additional correction factor is necessary to account for the effects of irradiation embrittlement on fatigue life.40 The RVI AMP, moreover, includes inspections intended to identify potential cracking caused by fatigue in susceptible RVI components, including irradiated RVI components.41 The inspection activities under the RVI AMP are in addition to, not in lieu of, the review of EAF for RVI components under the fatigue management program (FMP).42 E. NYS Does Not Challenge the Adequacy of the LRA with Respect to RPVs In addition to these claims, Entergy notes that NYS-25, as stated, alleges that Entergys LRA is inadequate with respect to the RPVs. Indeed, the original focus of NYS-25 in 2007 was on the RPVs themselves.43 But since 2011, NYS has focused almost exclusively on Entergys 38 NYS Revised SOP at 17 (NYS000481).

39 See generally Revised Testimony of Entergy Witnesses Nelson F. Azevedo, Alan B. Cox, Jack R. Strosnider, Randy G. Lott, Mark A. Gray and Barry M. Gordon Regarding Contention NYS-26B/RK-TC-1B (Aug. 10, 2015) (Entergys NYS-26B/RK-TC-1B Testimony) (ENT000679).

40 See id. at A76.

41 See Entergys Testimony § VII.A.7 (ENT000616).

42 See id. at A111.

43 See NYS Petition at 224, 226.

AMP for RVIs.44 While Dr. Laheys current testimony briefly alludes to some of his prior claims regarding the RPVs, both Dr. Lahey and the State stop short of alleging any specific deficiency in Entergys LRA regarding the RPVs.45 To ensure a complete record, however, Entergys expert witnesses demonstrate that the information regarding RPVs in the IPEC LRA complies fully with 10 C.F.R. Parts 50 and 54 and is consistent with NRC Staff guidance regarding the management of the effects of aging and the evaluation of time-limited aging analyses (TLAAs) for RPVs. The LRA therefore provides reasonable assurance that the effects of aging on the IPEC RPVs will be adequately managed, such that they will continue to perform their intended functions, consistent with the CLB.46 Neither the State nor Dr. Lahey provides any specific challenge to this information.

II. PROCEDURAL HISTORY OF CONTENTION NYS-25 As noted above, the claims in NYS-25 are cumulative and overlapping with other contentions. Specifically, the claims in NYS-25 substantially overlap those in contentions NYS-26B/RK-TC-1B (the metal fatigue contention) and NYS-38/RK-TC-5 (the safety commitments contention).47 Indeed, Dr. Laheys testimony regarding RVIs across the three 44 NYS Revised SOP at 17 (NYS000481); see also State of New York, Initial Statement of Position, Contention NYS-25 at 10 (Dec. 22, 2011) (NYS Initial SOP) (NYS000293).

45 See Report of Dr. Richard T. Lahey, Jr. in Support of Contentions NYS-25 and NYS-26B/RK-TC-1B at 13 (Dec. 20, 2011) (Report) (NYS000296); see also Pre-Filed Written Testimony of Dr. Richard T. Lahey, Jr.

Regarding Contention NYS-25 at 28-31 (Dec. 22, 2011) (Lahey 2011 Testimony) (NYS000294); Revised Lahey Testimony at 74 (NYS000482).

46 See 10 C.F.R. §§ 54.21, 54.29.

47 In objecting to the proposed amendments to NYS-25 and NYS-38/RK-TC-5 earlier this year, Entergy noted there was no discernible distinction between the two amended contentions, and asked the Board to separate the various claims in the interest of adjudicatory economy. Entergys Consolidated Answer Opposing Intervenors Motions to Amend Contentions NYS-25 and NYS-38/RK-TC-5 at 13 (Mar. 10, 2015), available at ADAMS Accession No. ML15069A677. The Board acknowledged that there is significant overlap, but found the States actions permissible. Memorandum and Order (Granting Motions for Leave to File Amendments to Contentions NYS-25 and NYS-38/RK-TC-5), at 14 (Mar. 31, 2015) (Second Order Amending NYS-25), available at ADAMS Accession No. ML15090A771.

contentions is substantively identical.48 Moreover, despite the significant developments and new information that has become available over the past three years or more, NYS has not replaced its 2011 SOP, testimony, or report with updated materials; it has merely added new information into the record in 2015 despite the fact that several prior positions and claims have been superseded by intervening events.49 Accordingly, Entergys Testimony and Statement of Position focus on the States most recent statement of position, testimony, and exhibits, filed on June 9, 2015. To assist the Boards review of the record, Entergy addresses challenges related to RVI and RPV aging management in its testimony, here, on contention NYS-25. Where there is an irreconcilable inconsistency, we focus on the most recent filings. Entergy addresses challenges related to metal fatigue (including EAF evaluations of RVI components) in its testimony on contention NYS-26B/RK-TC-1B.50 And it addresses specific challenges related to safety commitments (including RVI-related commitments) in its testimony on contentions NYS-38/RK-TC-5.51 A. Original Contention NYS first proffered Contention NYS-25 in 2007, as part of its initial Petition to Intervene.52 The States initial pleadings focused almost entirely on the RPV, rather than the RVIs, claiming that the information in the LRA on the TLAAs associated with the RPVs did not 48 Compare Revised Lahey Testimony (NYS00482) with Revised Pre-filed Written Testimony of Dr. Richard T.

Lahey, Jr. Regarding Consolidated Contention NYS-26B/RK-TC-1B, (June 9, 2015) (NYS000530) and Revised Pre-filed Written Testimony of Dr. Richard T. Lahey, Jr. Regarding Joint Contention NYS-38/RK-TC-5, (June 9, 2015) (NYS000562).

49 See Entergys Testimony at A65 (ENT000616).

50 See Entergys NYS-26B/RK-TC-1B Testimony (ENT000679).

51 See Revised Testimony of Entergy Witnesses Nelson F. Azevedo, Robert J. Dolansky, Alan B. Cox, Jack R.

Strosnider, Timothy J. Griesbach, Barry M. Gordon, Randy G. Lott, and Mark A. Gray Regarding Contention NYS-38/RK-TC-5 (Aug. 10, 2015) (Entergys NYS-38/RK-TC-5 Testimony) (ENT000699) .

52 See NYS Petition at 223-27.

include information on age-related accident analyses53 and that an intermediate shell in IP2 will not meet the upper shelf energy acceptance criterion of 50ft-lb.54 The original contention did not provide any basis for the States concerns about the associated internals of the RPVs, except to list the names of certain RVI components where Dr. Laheys [c]oncerns over embrittlement applied, and offer the unclear observation that RPV internals in IP3 imply operational limits for extended life operations due to the high [nil ductility temperature] NDT associated with the predicted irradiation-induced embrittlement.55 Entergy and the NRC Staff opposed admission of NYS-25.56 Entergys objections included that the proposed contention repeatedly confused the RPV and RVIs, was inadequately supported by the bare assertions in Dr. Laheys declaration, and failed to raise a genuine dispute with any information in the LRA.57 The Board admitted NYS-25 in 2008.58 B. First Amended Contention Consistent with Commitment 30 in the original LRA,59 Entergy submitted a detailed AMP for IP2 and IP3 RVIs on July 14, 2010.60 This AMP fully described Entergys program to 53 Id. at 224.

54 Id. at 226.

55 Id. at 224-226.

56 See Answer of Entergy Nuclear Operations, Inc. Opposing New York State Notice of Intention to Participate and Petition to Intervene at 135-41 (Jan. 22, 2008), available at ADAMS Accession No. ML080300149 (Entergy Answer); NRC Staffs Response to Petitions for Leave to Intervene Filed by [NYS] at 75-77 (Jan.

22, 2008), available at ADAMS Accession No. ML080230543.

57 See Entergy Answer at 135-41.

58 Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 & 3), LBP-08-13, 68 NRC 43, 131 (2008).

59 In Commitment 30, consistent with the then-current NRC guidance in NUREG-1801, Revision 1, Entergy committed to participate in industry programs for investigating and managing aging effects on RVIs, to evaluate and implement industry programs applicable to RVIs, and to submit an RVI inspection plan not less than 24 months before entering the PEO. See IPEC License Renewal Application at 3.1-7 to 3.1-8, 3.1-9 to 3.1-10 (Apr. 2007) (LRA) (ENT000015A); see also NUREG-1801, Generic Aging Lessons Learned Report, Rev. 1 at 7-30, tbl. 1 (Sept. 2005) (NUREG-1801) (NYS00146A).

manage the effects of aging on RVIs using guidance developed from nearly a decade of extensive industry research and set forth in EPRI Materials Reliability Program documents MRP-227 and MRP-228.61 Following Entergys submittal of its RVI AMP, NYS filed a motion to submit additional bases for NYS-25.62 The amended contention alleged, among other things, that the RVI AMP did not consider synergistic aging effects or potential shock loads, did not provide sufficient details on baseline and periodic inspections, did not provide sufficient details on corrective actions, including repair or replacement, disavow[ed] preventive actions, and relied on vague future commitments.63 Entergy and the NRC Staff objected on both timeliness and substantive grounds.64 Entergy, among other things, argued that NYS failed to challenge directly-relevant information in the RVI AMP.65 On July 6, 2011, the Board admitted the amended NYS-25.66 60 See NL-10-063, Letter from F. Dacimo to NRC Document Control Desk, Amendment 9 to License Renewal Application (LRA) - Reactor Vessel Internals Program (July 14, 2010) (NL-10-063) (NYS000313).

61 See NL-10-063, Attach. 1 at 82-84 (NYS000313).

62 See State of New Yorks Motion for Leave to File Additional Bases for Previously-Admitted Contention NYS-25 in Response to Entergys July 14, 2010 Proposed Aging Management Program for Reactor Pressure Vessels and Internal Components (Sept. 15, 2010), available at ADAMS Accession No. ML103050402.

63 See id.; Decl. of Richard T. Lahey, Jr., ¶¶ 13-15 (Sept. 15, 2010) (attached to motion), available at ML12335A461.

64 See Applicants Answer to Amended Contention New York State 25 Concerning Aging Management of Embrittlement of Reactor Pressure Vessel Internals (Oct. 12, 2010), available at ADAMS Accession No. ML103010104 (Entergys 2010 Answer); NRC Staffs Answer to State of New Yorks Motion for Leave to File Additional Bases for Previously-Admitted Contention NYS-25 (Oct. 12, 2010), available at ADAMS Accession No. ML102850764.

65 See generally Entergys 2010 Answer. As Entergys witnesses show throughout their testimony, for the past five years Dr. Lahey and the State have continued to disregard, rather than dispute the technical basis for the RVI AMP.

66 Licensing Board Order (Ruling on Pending Motions for Leave to File New and Amended Contentions) at 27 (July 6, 2011) (unpublished).

C. Revised RVI AMP and Inspection Plan Again, consistent with Commitment 30, Entergy submitted its RVI Inspection Plan on September 28, 2011, two years prior to entering the PEO for IP2.67 The Inspection Plan was based on detailed inspection guidance in MRP-227, and fully addressed the NRC Staffs action items and conditions in the Safety Evaluation for MRP-227, Revision 0.68 It also included a comprehensive schedule for inspections of RVI components at IPEC.69 The RVI Inspection Plan governed both IP2 and IP3.70 After EPRI issued the NRC-approved aging management guidance for RVIs in MRP-227-A (discussed further below), Entergy submitted a revised RVI AMP and Inspection Plan for both IP2 and IP3 based on MRP-227-A on February 17, 2012.71 IP2 and IP3 were among the first units in the U.S. fleet to prepare RVI AMPs based on the state-of-the-art NRC Staff-approved guidance in MRP-227-A and to have such an AMP reviewed by the NRC Staff as part of an LRA.72 From 2012 through 2014, the NRC Staff issued detailed requests for additional information (RAI) to Entergy on this first-of-a-kind AMP.73 Following Entergys submission of significant additional technical information in response to these RAIs, the NRC Staff approved Entergys revised RVI AMP and Inspection 67 See NL-11-107, Letter from F. Dacimo, Vice President, Entergy, to NRC Document Control Desk, License Renewal Application - Completion of Commitment # 30 Regarding the Reactor Vessel Internals Inspection Plan (Sept. 28, 2011) (NYS000314).

68 See id.

69 See id., Attach. 1 at 35-39, tbl.5-2.

70 See Entergys Testimony at A58 (ENT000616).

71 NL-12-037 (NYS000496).

72 See Entergys Testimony at A60 (ENT000616).

73 See NUREG-1930, Supp. 2, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3 at B-2 to B-7 (Nov. 2014) (SSER 2) (NYS000507).

Plan as documented in SSER 2 issued on November 6, 2014.74 NRC Staff concluded that Entergys LRA for IP2 and IP3 demonstrates that the effects of aging on RVI components will be adequately managed, as required under 10 C.F.R. §§ 54.21(a)(3) and (c)(1)(iii), and the RVI Inspection Plan implements the elements of the RVI AMP in an acceptable manner.75 D. 2011 NYS Testimony Along with the other Track 1 contentions, the State submitted its initial statement of position, prefiled testimony, and exhibits on NYS-25 in December 2011.76 Before any further testimony was filed on NYS-25, however, the Board placed NYS-25 on the schedule for the second set of hearings in this proceeding (i.e., Track 2).77 E. Second Amended Contention Following the publication of SSER 2,78 the State filed a Motion for Leave to Supplement Previously-Admitted Contention NYS-25 focusing the IPEC RVIs AMP.79 The States Second Amended Contention alleged that the RVI AMP remained deficient because it does not: (1) address or manage the combined synergistic aging effects of embrittlement, fatigue, and other aging mechanisms; (2) maintain safety margins during the PEO by, for example, repair or replacement of the RVIs, and does not account for the full range of transient shock loads; and 74 See id. at 3-26, 3-59; see also id. at B-2 to B-7.

75 SSER 2 at 3-26, 3-59 (NYS000507).

76 See NYS Initial SOP at 10; Report at 13; see also Pre-Filed Written Testimony of Dr. Richard T. Lahey, Jr.

Regarding Contention NYS-25 at 28-31 (Dec. 22, 2011) (Lahey 2011 Testimony) (NYS000294);

Supplemental Report of Dr. Richard T. Lahey, Jr. in Support of Contention NYS-25 and NYS-26B/RK-TC-1B (Supplemental Lahey Report) (NYS000297).

77 See Licensing Board Order (Granting NRC Staffs Unopposed Time Extension Motion and Directing Filing of Status Updates) at 2 (Feb. 16, 2012) (unpublished).

78 NUREG-1930, Supp. 2, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3 (Nov. 6, 2014) (SSER 2), available at ADAMS Accession No. ML14310A803.

79 State of New Yorks Motion for Leave to Supplement Previously-Admitted Contention NYS-25 (Feb. 13, 2015) (Second Motion to Amend), available at ADAMS Accession No. ML15044A493.

(3) include required preventative or corrective actions or acceptance criteria for the baffle-former bolt inspections. The State further alleged that the Westinghouse EAF calculations prepared for Indian Point are allegedly inadequate.80 The Second Amended NYS-25 did not allege any deficiencies in the IPEC LRA regarding the RPVs.81 Entergy objected on both timeliness and substantive grounds.82 Once again, Entergy objected on the grounds that the State continued to disregard, rather than dispute the technical basis for the RVI AMP.83 In particular, Entergy showed that it had disclosed to the State a substantial body of technical documentation supporting MRP-27-A and, the IPEC RVI AMP,84 yet the State did not even mention this information in its proposed amended contention.85 On March 31, 2015, the Board granted the States motion without altering or amending the contention.86 Thereafter, under the Boards scheduling orders, NYS filed its revised statement of position, testimony, and additional exhibits on June 9, 2015.87 80 New York State February 2015 Supplement to Previously-Admitted Contention NYS-25 at 1-3 (Feb. 13, 2015)

(Second Supplement to NYS-25), available at ADAMS Accession No. ML15044A491.

81 Although the State does not allege any deficiencies in the LRA related to RPVs, Dr. Lahey did note the suggestion of a potential non-conservatism in BTP 5-3, related to RPVs, which is discussed in further detail below.

82 Entergys Consolidated Answer Opposing Intervenors Motions to Amend Contention NYS-25 and NYS-38/RK-TC-5 (Mar. 10, 2015), available at ADAMS Accession No. ML15069A677.

83 See id. at 18-23.

84 See id. at 8 & n.30.

85 See id. at 19.

86 Second Order Amending NYS-25 at 10.

87 See generally NYS Revised SOP (NYS000481); Revised Lahey Testimony (NYS000482); exhibits NYS000483 through NYS000528.

III. APPLICABLE LEGAL AND REGULATORY STANDARDS As demonstrated below, the IPEC RVI AMP fully meets the applicable legal and regulatory requirements in 10 C.F.R. Part 54. In addition to the lack of technical merit, the States claims in NYS-25 are legally deficient in that they are contrary to the limited scope of the license renewal rule in 10 C.F.R. Part 54 as well as the NRCs reasonable assurance standard.

NYS has, moreover, failed to carry its burden of going forward on its contention and overcoming the special weight accorded to NRC Staff guidance documents.

A. NYS Continues Its Attempts to Impermissibly Expand the Scope of this License Renewal Proceeding

1. The License Renewal Review Is a Limited One The State continues to bring forward claims that attempt to expand the scope of this license renewal proceeding beyond the bounds that are clearly established by 10 C.F.P. Part 54 and Commission precedent. For example, the States claims related to the consideration of shock loads involve concerns about postulated accidents or events that are beyond the design basis of IP2 and IP388 and which are clearly outside the limited scope of this license renewal proceeding. Similarly, the States claims regarding alleged deficiencies in the seismic hazard curves for IP2 and IP3,89 demands for wholesale repair or replacement of RVIs in lieu of an AMP,90 and claims regarding active components such as control rods and control rod drive mechanisms,91 also fall beyond the bounds of this proceeding and must be rejected by the Board.

88 E.g., NYS Revised SOP at 17 (NYS000481).

89 See id. at 41.

90 See id. at 31.

91 See id. at 25 (Entergys 2011 AMP for RPVIs was inadequate with respect to the embrittlement of the control rod drives . . . .).

Specifically, 10 C.F.R. Part 54 is focused on managing the effects of aging on passive, long-lived components. It does not include a review of the adequacy of a plants CLB, including its design basis. Nor does it include a review of ongoing regulatory matters that are fully addressed under 10 C.F.R. Part 50 and by NRC inspection and enforcement activities.92 The Commissions license renewal regulations clearly reflect this distinction between 10 C.F.R. Part 54 aging management issues on the one hand, and ongoing 10 C.F.R. Part 50 regulatory process (e.g., the adequacy of the plants design basis) on the other.93 The underlying adequacy of the CLB itself is outside the scope of license renewal and is not open to challenge in this proceeding.94 The license renewal review is premised upon the determination that, with the exception of aging management issues, the NRCs ongoing regulatory process is adequate to ensure that the CLB of an operating plant provides and maintains an acceptable level of safety.95 Thus, the States challenges to the adequacy of design basis loads on plant components, including seismic loads, must be rejected on legal grounds aloneputting aside their factual inadequacies. Likewise, as further explained below, the Board must also reject the States concerns regarding control rods, which are active components and consumables, as beyond the scope of this proceeding.96 The State further challenges the license renewal rule when it argues that, instead of implementing an AMP, Entergy should take proactive steps to repair or replace aging RVI 92 See Fla. Power & Light Co. (Turkey Point Nuclear Generating Plant, Units 3 and 4), CLI-01-17, 54 NRC 3, 7-9 (2001); see also Indian Point, CLI-15-6, slip op. at 8; 10 C.F.R. § 54.21(a)(1).

93 Turkey Point, CLI-01-17, 54 NRC at 7; see also id. at 9 (The current licensing basis . . . includes the plant-specific design basis information documented in the plants most recent Final Safety Analysis Report . . . and any orders, exemptions, and licensee commitments that are part of the docket for the plants license . . . .).

94 See Pilgrim, CLI-10-14, 71 NRC at 461; Oyster Creek, CLI-09-7, 69 NRC at 270.

95 See Final Rule, Nuclear Power Plant License Renewal; Revisions, 56 Fed. Reg. 64,943, 64,946 (Dec. 13, 1991).

96 See Indian Point, CLI-15-6, 81 NRC at __, slip op. at 8; 10 C.F.R. § 54.21(a)(1).

components.97 The license renewal rule requires the applicant to show that there is reasonable assurance that the effects of aging will be adequately managednot that aging effects will be precluded.98 The review of TLAAs for license renewal is also well-defined in Part 54 and not subject to challenge in this proceeding. Certain in-scope components are subject to time-limited calculations or analyses that are part of the CLB, known as TLAAs. TLAAs must be evaluated for the PEO.99 In doing so, an applicant must: (i) show that the original TLAAs will remain valid for the PEO; (ii) revise and extend the TLAAs to be valid for a longer term, such as 60 years; or (iii) otherwise demonstrate that the effects of aging will be adequately managed during the renewal term.100 As they relate to this contention, the EAF evaluations prepared by Westinghouse for IPEC address all components with a CLB cumulative usage factor (CUF) analysis.101 The EAF evaluations are part of the Fatigue Management Program (FMP)the program that Entergy is using to resolve the CUF TLAAs under 10 C.F.R. § 54.21(c)(iii).102 But the CLB CUF analysis is a fatigue analysis, not a general analysis of all aging effects.103 Thus, to the extent the State and Dr. Lahey argue that irradiation embrittlement or other degradation mechanisms be considered in EAF evaluations,104 their claims are a challenge to the CLB and the license renewal rule. As further explained below, Entergy uses the RVI AMP to manage the 97 NYS Revised SOP at 31 (NYS000481).

98 See Seabrook, CLI-12-5, 75 NRC at 314-15.

99 See 10 C.F.R. § 54.21(c)(1).

100 See id.

101 See Entergys NYS-26B/RK-TC-1B Testimony § V.C (ENT000679).

102 See id. at A97.

103 See id. § IV.A.

104 See NYS Revised SOP at 41 (NYS000481); Revised Lahey Testimony at 19-20 (NYS000482).

effects of aging on RVI components caused by all pertinent aging mechanisms, including the effects of fatigue, embrittlement, and stress corrosion cracking.

2. The Reasonable Assurance Standard Pursuant to Section 54.29(a), the NRC will issue a renewed license if it finds that the applicant has identified actions that have been taken or will be taken such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the CLB.105 In addition to the limitations on the scope of this proceeding set forth in 10 C.F.R. § 54.21(a)(1), the reasonable assurance standard does not require Entergy to show protection against speculative, postulated events that are beyond the design basis of the plant,106 or to preclude all potential aging effects by replacing the RVIs.107 It also requires Dr.

Lahey to provide more than the speculation he repeatedly presents in support of his claims.108 Longstanding precedent makes clear that the reasonable assurance standard does not require an applicant to meet an absolute or beyond a reasonable doubt standard.109 Rather, the Commission takes a case-by-case approach, applying sound technical judgment and verifying the applicants compliance with Commission regulations.110 Branch Technical Position RLSB-1, in the Standard Review Plan for Review of License Renewal (SRP-LR), explains that the 105 10 C.F.R. § 54.29(a).

106 NYS Revised SOP at 17 (NYS000481).

107 Id. at 31.

108 See, e.g., Revised Lahey Testimony at 16 (seriously embrittled and fatigued RPV internals may not be able to survive the shock loads), 16-17 (multiple aging mechanisms that occur in a reactor core (including fatigue, irradiation embrittlement, and corrosion) may result in cumulative material degradation), 40 (highly embrittled and fatigued RVI components may not have signs of degradation that can be detected by an inspection, but such weakened components could nonetheless fail) (emphasis added) (NYS000482).

109 Oyster Creek, CLI-09-7, 69 NRC at 262 n.142; Commonwealth Edison Co. (Zion Station, Units 1 & 2),

ALAB-616, 12 NRC 419, 421 (1980); N. Anna Envtl. Coal. v. NRC, 533 F.2d 655, 667-68 (D.C. Cir. 1976)

(rejecting the argument that reasonable assurance requires proof beyond a reasonable doubt and noting that the licensing board equated reasonable assurance with a clear preponderance of the evidence).

110 See Oyster Creek, CLI-09-7, 69 NRC at 262, n.143, 263; Pilgrim, CLI-10-14, 71 NRC at 465-66.

license renewal process is not intended to demonstrate absolute assurance that structures and components will not fail, but rather that there is reasonable assurance that they will continue to perform their intended functions consistent with the CLB during the PEO.111 Indeed, the plain language of the regulations, and Commission decisions interpreting those regulations, state that the central question for a license renewal applicant is whether aging management activities have been identified and actions have been or will be taken to provide reasonable assurance of continued safety.112 Importantly, these regulations do not require the applicant to demonstrate that aging effects be precluded,113 but are oriented in large part toward identifying actions that will be taken in the future.114 B. License Renewal Guidance As previously noted, nowhere in their filings do the State or Dr. Lahey allege that IPEC RVI AMP is inconsistent with NRC Staff guidance. Instead, they attack the NRC guidance itself, which endorses the industrys detailed guidelines in MRP-227-A.115 While the Commission has not forbidden such arguments, the State and Dr. Lahey face a high bar to overcome the special weight accorded to the NRC Staffs guidance on license renewal.116 As 111 SRP-LR, Revision 1, Appx. A, at A.1-1 (NYS000195); NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Rev. 2 (Dec. 2010) (SRP-LR, Revision 2)

(NYS000161).

112 See 10 C.F.R. §§ 54.21(a)(3), 54.29(a)(1).

113 See Seabrook, CLI-12-5, 75 NRC at 314-15.

114 See Vt. Yankee, CLI-10-17, 72 NRC at 36.

115 Revised Lahey Testimony, at 39 (NYS000482) (MRP-227-A is an inspection-based aging management plan, which I believe is inadequate).

116 See, e.g., Indian Point, CLI-15-6, slip op. at 19; Seabrook, CLI-12-5, 75 NRC 314 n.78 (quoting Private Fuel Storage, L.L.C. (Indep. Spent Fuel Storage Installation), CLI-01-22, 54 NRC 255, 264 (2001)); see also id.

(We recognize, of course, that guidance documents do not have the force and effect of law. Nonetheless, guidance is at least implicitly endorsed by the Commission and therefore is entitled to correspondingly special weight) (quoting Yankee Atomic Elec. Co. (Yankee Nuclear Power Station), CLI-05-15, 61 NRC 365, 375 n.26 (2005)).

explained throughout this Statement of Position, and as Entergys witnesses have demonstrated, the State and Dr. Lahey have not cleared that bar.

The two primary license renewal guidance documents issued by the NRC Staff are NUREG-1801, the Generic Aging Lessons Learned Report or GALL Report,117 and NUREG-1800, the Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, or SRP-LR.118 The SRP-LR provides guidance to NRC staff for conducting their review of LRAs and provides acceptance criteria for determining whether the applicant has met the regulatory requirements for license renewal.119 NUREG-1801 provides the technical basis for the SRP-LR and contains the NRC Staffs generic evaluation of programs that manage the effects of aging during the PEO, and meet the requirements of 10 C.F.R. Part 54.120 NUREG-1801 indicates that many existing, current-term programs are also adequate to manage the aging effects for particular structures or components for license renewal. Thus, programs that are consistent with NUREG-1801 are accepted by the Staff as adequate to meet the requirements of the license renewal rule.121 The Commission has endorsed NUREG-1801 because it is based on extensive research and evaluation of operating 117 See generally NUREG-1801, Rev. 1 (NYS00146A-C); NUREG-1801, Rev. 2 (NYS00147A-D).

118 See generally Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Rev. 1 (Sept. 2005) (SRP-LR, Rev. 1) (NYS000195); NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Rev. 2 (Dec. 2010) (SRP-LR, Rev. 2)

(NYS000161).

119 See SRP-LR, Rev. 2 at 1-3 (NYS00146A).

120 See NUREG-1801, Rev 1, at 3-4 (NYS00146A).

121 See id. at 3.

experience derived from a comprehensive set of sources.122 NUREG-1801 is also subject to stakeholder review and comment.123

1. NUREG-1801 Is Entitled to Special Weight in This Proceeding The Commission has held that a license renewal applicants use of the guidance in NUREG-1801 satisfies regulatory requirements under 10 C.F.R. Part 54.124 Also, where the NRC develops a guidance documentsuch as NUREG-1801to assist in compliance with applicable regulations, that document is entitled to special weight in NRC proceedings.125 In particular, for license renewal safety issues, an applicants use of an AMP identified in NUREG-1801 constitutes reasonable assurance that it will manage the targeted aging effect during the renewal period.126 The Commission has reiterated this principle, holding that a commitment to implement an AMP that the NRC finds is consistent with NUREG-1801 constitutes an acceptable method for compliance with 10 C.F.R. § 54.21(c)(1)(iii).127 Accordingly, to challenge the adequacy of an NRC-approved guidance document, an intervenor must provide specificity and substantial support128 to overcome the special weight accorded to a guidance document that has been 122 See NUREG-1801, Rev. 2, at 2 (NYS00147A).

123 See id. Neither NYS nor Riverkeeper, however, submitted comments to the NRC for consideration in NUREG-1801, Rev. 2. See NUREG-1950, Disposition of Public Comments and Technical Bases for Changes in the License Renewal Guidance Documents NUREG-1801 and NUREG-1800, at IV-1 to IV-21 (Apr. 2011)

(ENT000528) (listing public comments on changes to NUREG-1801 and NUREG-1800).

124 See, e.g., Oyster Creek, CLI-08-23, 68 NRC at 468.

125 Indian Point, CLI-15-6, slip op. at 19; Seabrook, CLI-12-5, 75 NRC 314 n.78.

126 See Oyster Creek, CLI-08-23, 68 NRC at 468 (emphasis added); see also Seabrook, CLI-12-05, 75 NRC at 304 (If the NRC concludes that an aging management program (AMP) is consistent with the GALL Report, then it accepts the applicants commitment to implement that AMP, finding the commitment itself to be an adequate demonstration of reasonable assurance under section 54.29(a).).

127 Vt. Yankee, CLI-10-17, 72 NRC at 36.

128 See id. at 33 n.185, 37.

implicitly endorsed by the Commission.129 As demonstrated by Entergys testimony, the State has not done so here.

In light of the foregoing, a finding that an applicants AMP is consistent with NUREG-1801 carries special weight130 and constitutes a finding of reasonable assurance under 10 C.F.R.

§§ 54.21(a), 54.21(c)(1)(iii), and 54.29(a).131

2. Revisions to NUREG-1801

a. NUREG-1801, Revision 1 The IPEC LRA was prepared using the guidance of NUREG-1801, Revision 1. In 2010, more than three years after the LRA was submitted, and more than a year after the NRC Staffs original SER was published, the Staff issued NUREG-1801, Revision 2.132 Subsequently, in 2013, the NRC Staff published interim staff guidance to revise and update NUREG-1801, Rev. 2 based on the NRCs approval of industry guidance on the aging management of RVIs MRP-227-A.133 As discussed further below, the RVI AMP fully meets this most recent guidance.

b. NUREG-1801, Revision 2 and MRP-227 The NRC Staff issued NUREG-1801, Rev. 2 in December 2010.134 As explained in Section V.B. of Entergys Testimony, NUREG-1801, Rev. 2 contained a new AMP (XI.M16A) addressing pressurized water reactor (PWR) RVIs. The new AMP relied on the 129 Seabrook, CLI-12-5, 75 NRC at 314 n.78.

130 Id.

131 Vt. Yankee, CLI-10-17, 72 NRC at 36.

132 See generally NUREG-1801, Rev. 2 (NYS00147A-D).

133 Final License Renewal Interim Staff Guidance LR-ISG-2011-04, Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors at 1 (May 28, 2013) (LR-ISG-2011-04)

(ENT000641).

134 See generally NUREG-1801, Rev. 2 (NYS00147A-D).

implementation of industry guidance from EPRI MRP in MRP-227, Revision 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines.135 As discussed in Sections II.B and C above, in 2010 Entergy submitted its original RVI AMP, and in 2011 submitted its RVI Inspection Plan, both of which were based on MRP-227, Revision 0.

c. Interim Staff Guidance and Endorsement of MRP-227-A As explained further below, in 2011, the NRC Staff issued its safety evaluation (SE) on MRP-227, Revision 0.136 The SE contained specific topical report conditions and applicant/licensee action items (A/LAIs) that were to be addressed by applicants or licensees utilizing the report. MRP-227-A, the NRC-endorsed version of MRP-227, Revision 0, was published in January 2012 to incorporate the Staffs topical report conditions and A/LAIs.137 Thereafter, the NRC Staff issued interim staff guidance, LR-ISG-2011-04, to amended AMP XI.M16A to reflect its endorsement of MRP-227-A.138 As discussed in Section II.C, above, Entergy submitted a revised RVI AMP and Inspection Plan based on the NRC-endorsed MRP-227-A guidelines. Thus, the IPEC RVI AMP meets the intent of the latest Staff guidance on management of the effects of aging on PWR RVIs. Again, NYS and Dr. Lahey do not challenge that fact, but rather challenge the guidance itself. To challenge the adequacy of the underlying guidance, NYS must to overcome the special weight with specificity and substantial support139 for its arguments. As the Commission has held 135 Id. at XI M16A-1.

136 Letter from R. Nelson, NRC, to N. Wilmshurst, EPRI, Revision 1 to the Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Prog[ra]m (MRP) Report 1016596 (MRP-227),

Revision 0, Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines (Dec. 16, 2011)

(SE for MRP-227-A) (ENT000230).

137 See generally MRP-227-A (NRC0014A-F).

138 LR-ISG-2011-04 at 2-3 (ENT000641).

139 See Vt. Yankee, CLI-10-17, 72 NRC at 33 n.185, 37.

in this proceeding, the State must show there are unusual circumstances in this case that would justify setting aside the applicable guidance.140 The State has made no such showing.

C. Burden of Proof At the hearing stage, an intervenor has the initial burden of going forward; that is, it must provide sufficient, probative evidence to establish a prima facie case for the claims made in the admitted contention.141 The mere admission of a contention does not satisfy this burden.142 If the Intervenors do establish a prima facie case on a particular claim, then the burden shifts to Applicant to provide sufficient evidence to rebut the intervenors contention.143 At the admissibility stage, the petitioner has the ironclad obligation to examine the available documentation with sufficient care to support the foundation for a contention.144 This obligation applies with equal, if not greater, force at the hearing stage.145 As will be further explained below, the State and its witness, Dr. Lahey, disregard, rather than dispute, the technical 140 Indian Point, CLI-15-6, slip op. at 21-22.

141 Oyster Creek, CLI-09-07, 69 NRC at 269 (quoting Consumers Power Co. (Midland Plant, Units 1 & 2),

ALAB-123, 6 AEC 331, 345 (1973) (The ultimate burden of proof on the question of whether the permit or license should be issued is . . . upon the applicant. But where . . . one of the other parties contends that, for a specific reason . . . the permit or license should be denied, that party has the burden of going forward with evidence to buttress that contention. Once he has introduced sufficient evidence to establish a prima facie case, the burden then shifts to the applicant who, as part of his overall burden of proof, must provide a sufficient rebuttal to satisfy the Board that it should reject the contention as a basis for denial of the permit or license.) (emphasis in original)); see also Vt. Yankee Nuclear Power Corp. v. Natural Res. Def. Council, 435 U.S. 519, 554 (1978) (upholding this threshold test for intervenor participation in licensing proceedings);

Phila. Elec. Co. (Limerick Generating Station, Units 1 & 2), ALAB-262, 1 NRC 163, 191 (1975) (holding that the intervenors had the burden of introducing evidence to demonstrate that the basis for their contention was more than theoretical).

142 See Oyster Creek, CLI-09-07, 69 NRC at 268-70.

143 See, e.g., id. at 269; La. Power & Light Co. (Waterford Steam Electric Station, Unit 3), ALAB-732, 17 NRC 1076, 1093 (1983) (citing Midland, ALAB-123, 6 AEC at 345); see also 10 C.F.R. § 2.325.

144 See Duke Power Co. (Catawba Nuclear Station, Units 1 & 2), ALAB-687, 16 NRC 460, 468 (1982), vacated in part on other grounds, CLI-83-19, 17 NRC 1041 (1983).

145 See Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 & 3), LBP-13-13, 78 NRC 246, 301 n.308 (2013) (rejecting an experts claims based on some averages and a gut feeling, rather than a thorough a review of available documentation).

basis for the RVI AMP. The State, therefore, has failed to meet its burden of going forward with evidence to support NYS-25. Considering the States and Dr. Laheys disregard of the substantial available documentation in direct testimony, Entergy reserves its right to object and to seek to strike any new critique of these studies that Dr. Lahey or NYS may offer in rebuttal, or, in the alternative, to seek to file sur-rebuttal testimony.

To prevail, the Applicants position must be supported by a preponderance of the evidence.146 IV. ENTERGYS WITNESSES Entergys testimony on NYS-25 is sponsored by the witnesses identified below. The testimony, opinions, and evidence presented by these witnesses are based on their substantial technical and regulatory expertise, professional experience, and personal knowledge of the issues raised in NYS-25. Collectively, these witnesses will demonstrate that NYS-25 lacks merit.

A. Mr. Nelson F. Azevedo Nelson Azevedos professional and educational qualifications are summarized in his curriculum vitae147 and in Section I.A of Entergys testimony. Mr. Azevedo is employed by Entergy as the Supervisor of Code Programs at IPEC. He holds a Bachelor of Science degree in Mechanical and Materials Engineering from the University of Connecticut, and a Master of Science in Mechanical Engineering and Master of Business Administration (M.B.A.) degrees from the Rensselaer Polytechnic Institute (RPI) in Troy, New York. Mr. Azevedo has more than 30 years of professional experience in the nuclear power industry. In his current position, he oversees the IPEC engineering section responsible for implementing American Society of 146 See Pac. Gas & Elec. Co., ALAB-763, 19 NRC at 577; Oyster Creek, CLI-09-07, 69 NRC at 263.

147 See Curriculum Vitae for Nelson F. Azevedo (ENT000032).

Mechanical Engineers (ASME) Code programs, including the fatigue monitoring, inservice inspection, inservice testing, boric acid corrosion control, non-destructive examination, steam generators, alloy 600 cracking, RPV embrittlement, and RVI programs. In addition to those duties he is responsible for ensuring compliance with the ASME Code,Section XI requirements for repair and replacement activities at IPEC and represents IPEC before industry organizations, including the PWR Owners Group Management Committee. Accordingly, Mr. Azevedo is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on the Entergy RVI AMP, and Entergys aging management activities and TLAAs for RPVs.

B. Mr. Robert J. Dolansky Bob Dolanskys professional and educational qualifications are summarized in his curriculum vitae148 and in Section I.B of Entergys testimony. Mr. Dolansky is employed by Entergy as a Code Programs Engineer at IPEC. He holds a Bachelor of Science degree in Aeronautical Engineering from RPI in Troy, New York. Mr. Dolansky has more than 25 years of professional experience as an ASME Code Programs Engineer at IPEC. He has been the program owner for, among other programs, the RVI, inservice inspection (ISI), inservice testing, steam generator, and alloy 600 cracking programs. In his current position, he is the program owner of the IPEC RVI AMPs for both units. He is also a member of the PWR Owners Group materials subcommittee.

Accordingly, Mr. Dolansky is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on the Entergy RVI AMP.

148 See Curriculum Vitae for Robert J. Dolansky (ENT000522).

C. Mr. Alan B. Cox Alan Coxs professional and educational qualifications are summarized in his curriculum vitae149 and in Section I.C of Entergys testimony. In brief, he holds a Bachelor of Science degree in Nuclear Engineering from the University of Oklahoma and a Master of Business Administration (M.B.A.) from the University of Arkansas at Little Rock. Prior to his retirement from Entergy in 2015, he was the Technical Manager of License Renewal. Presently, he continues to work with Entergy as an independent consultant. Mr. Cox has more than 37 years of experience in the nuclear power industry, having served in various positions related to engineering and operations of nuclear power plants, including several years as a licensed reactor operator and a senior reactor operator. From 2001 to 2015, he worked full-time on license renewal matters, supporting the integrated plant assessment and LRA development for Entergy license renewal projects, as well as projects for other utilities.

Mr. Cox was directly involved in preparing the LRA and developing or reviewing AMP descriptions for IP2 and IP3, including the IPEC RVI AMPs. He has also been directly involved in developing or reviewing Entergy responses to NRC Staff RAIs concerning the LRA and necessary amendments or revisions to the application. Accordingly, he has extensive knowledge of IPEC aging management activities, including the descriptions in the LRA and other related documentation discussed below. Thus, Mr. Cox is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on the Entergy RVI AMP, and Entergys aging management activities and TLAAs for RPVs.

149 See Curriculum Vitae for Alan B. Cox (ENTR00031).

D. Mr. Jack R. Strosnider, Jr.

Jack Strosniders professional and educational qualifications are summarized in his curriculum vitae150 and in Section I.D of Entergys testimony. Mr. Strosnider holds a Bachelor of Science degree and a Master of Science degree, both in Engineering Mechanics from the University of Missouri at Rolla, and an M.B.A. degree from the University of Maryland. Mr.

Strosnider is a Senior Nuclear Safety Consultant with Talisman International, LLC. Prior to April 2007, he was employed for 31 years by the NRC. During that time, he held numerous senior management positions at the NRC, including Director of the Office of Nuclear Material Safety and Safeguards, Deputy Director of the Office of Nuclear Regulatory Research, and Director of the Division of Engineering in the Office of Nuclear Reactor Regulation (NRR).

On technical matters, he was, for example, involved in the development of the technical bases for 10 C.F.R. § 50.61, which provides fracture toughness requirements for protection against pressurized thermal shock (PTS) events, and was responsible for licensing reviews associated with the integrity of the RPV and monitoring of RVIs.

Mr. Strosnider has extensive experience in developing and applying NRC regulations and programs addressing the aging of nuclear power plant structures and components. He has directed engineering reviews and the preparation of SERs for license renewal. With respect to aging effects on RPVs, Mr. Strosnider was involved in the development of the technical bases for the requirements in 10 C.F.R. § 50.61. Thus, Mr. Strosnider is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on the NRC regulatory requirements relating to RVI AMPs, aging management activities, and TLAAs for RPVs, and criteria necessary to satisfy those requirements.

150 See Curriculum Vitae for Jack R. Strosnider, Jr. (ENTR00184).

E. Mr. Timothy J. Griesbach Tim Griesbachs professional and educational qualifications are summarized in his curriculum vitae151 and in Section I.E of Entergys testimony. In brief, he holds Bachelor of Science and Master of Science degrees in Metallurgy and Materials Science from Case Western Reserve University. Currently, he is a Senior Associate at Structural Integrity Associates, Inc.

Mr. Griesbach has more than 40 years of experience in metallurgy and materials engineering, primarily in the nuclear field.

He is a member of the American Nuclear Society and the American Society of Mechanical Engineers (ASME), where he has served on various ASME Boiler and Pressure Vessel Code committees for over 33 years, chairs the ASME Section XI Working Group on Operating Plant Criteria, and is currently a member of the ASME Section XI Standards Committee. He has worked closely with the EPRI Materials Reliability Program to develop and implement the MRP-227 inspection and evaluation guidelines for the safety and long-term operation of PWR vessel internals. Thus, Mr. Griesbach is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on the Entergy RVI AMP, and Entergys aging management activities and TLAAs for RPVs.

F. Dr. Randy G. Lott Dr. Randy Lotts professional and educational qualifications are summarized in his curriculum vitae152 and in Section I.F of Entergys testimony. Dr. Lott holds a Bachelor of Science in Engineering degree in nuclear engineering from the University of Michigan, and Master of Science and Doctor of Philosophy degrees in nuclear engineering from the University 151 See Curriculum Vitae for Timothy J. Griesbach (ENT000617).

152 See Curriculum Vitae for Randy G. Lott (ENT000618).

of Wisconsin. Currently, he is a Consulting Engineer at Westinghouse and has more than 35 years of experience in nuclear materials and radiation effects.

Dr. Lott has extensive experience with post-irradiation evaluation of reactor components, and has been directly involved in the design and implementation of aging management programs for reactor internals. He has supervised testing of RPV surveillance capsules and conducted research programs on irradiation embrittlement and annealing of RPV steels, and he has conducted numerous test programs on highly irradiated stainless steels, including measurement of tensile, fracture toughness and irradiation-assisted stress corrosion cracking (IASCC) properties. As a member of the MRP Reactor Internals Inspection and Evaluation Guidelines Core Group, he was a contributor to the U.S. industry Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227). Specifically, he worked on aging management strategies for the Westinghouse and Combustion Engineering plants to provide the basis for the RVI inspection guidelines in MRP-227. Thus, Dr. Lott is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on the Entergy RVI AMP.

G. Mr. Mark A. Gray Mark Grays professional and educational qualifications are summarized in his curriculum vitae153 and in Section I.G of Entergys testimony. Mr. Gray is a Principal Engineer in the Primary Systems Design and Repair group at Westinghouse. He holds Master of Science and Bachelor of Science degrees in Mechanical Engineering from the University of Pittsburgh and has over 34 years of experience in the nuclear power industry. His principal work activities include the evaluation of the structural integrity of primary system piping and components, 153 See Curriculum Vitae for Mark A. Gray (ENTR00186).

including the development of plant life extension and monitoring programs and analysis. He participated in the development and application of transient and fatigue monitoring algorithms and software for the WESTEMS' Transient and Fatigue Monitoring System, and collaborated with vendors outside Westinghouse in the development of transient and fatigue monitoring systems.

During the preparation of the EAF analyses for IPEC license renewal, Mr. Gray provided general technical direction for the engineers performing the EAF analyses, and either co-authored or reviewed the resulting Westinghouse environmental fatigue reports, referred to as WCAP reports. For these reasons, Mr. Gray is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on fatigue analysis of RVIs.

V. ENTERGYS EVIDENCE AND ARGUMENTS In their testimony, Entergys experts explain why Entergys IPEC RVI AMPtogether with substantial supporting informationprovides reasonable assurance that the effects of aging will be adequately managed throughout the PEO as required by 10 C.F.R. §§ 54.21(a)(3),

54.21(c)(1)(iii), and 54.29(a). In so doing, Entergys experts refute the States and Dr. Laheys assertions point-by-point, thereby demonstrating that the issues raised in NYS-25 lack merit from regulatory and technical perspectives.

A. Technical Background on the Aging Management of RVIs and RPVs In Section V of Entergys testimony, Entergys expert witnesses describe the layout and functions of the RVIs and RPVs, the scope of components covered by the RVI AMP, the materials used in the RVIs and RPVs at IPEC, and the design basis loads those materials are subjected to.

Entergys witnesses first explain that the RPV contains the reactor core and RVIs, and is a key part of the reactor coolant pressure boundary.154 The RVIs, located inside the RPV, direct the coolant flow, support the reactor core, and guide the control rods, but do not form part of the reactor coolant pressure boundary.155 The RVI AMP provides a complete and correct list of the PWR RVI sub-assemblies and components at IPEC.156 Contrary to Dr. Laheys belief,157 the RVIs do not include control rods.

Further, the control rods are active components that perform their intended function with moving parts or a change in configuration.158 They also are consumables, subject to replacement based on a qualified life or specified time period under 10 C.F.R. § 54.21(a)(1).159 They are therefore excluded from AMR pursuant to Part 54.160 Dr. Lahey also incorrectly asserts that Entergy has not addressed other control rod-related components,161 but they are in fact included in AMR and the RVI AMP.162 And while Dr. Lahey raises concerns about the control rod stub tube welds or J-groove welds, and about RPV head penetrations, the effects of aging on those components are managed under the Reactor Vessel Head Penetration Inspection AMP, not the RVI AMP.163 Accordingly, several of Dr.

Laheys complaints are at odds with accepted industry definitions of RVIs and Part 54, or are 154 See Entergys Testimony at A90 (ENT000616).

155 See id. at A94.

156 See id. at A98 (citing NL-12-037 (NYS000496)).

157 See Revised Lahey Testimony at 13 (NYS000482).

158 See Indian Point, CLI-15-6, 81 NRC at __, slip op. at 8.

159 See Entergys Testimony at A99 (ENT000616); see also 10 C.F.R. § 54.21(a)(1).

160 See Entergys Testimony at A99 (ENT000616).

161 See Revised Lahey Testimony at 12-13 (incorrectly asserting that the guide tubes, plates, pins, and welds associated with the control rods are omitted from the RVI AMP) (NYS000482).

162 See Entergys Testimony at A100 (ENT000616).

163 See id. at A101.

adequately addressed by other aging management programs that he either ignored or chose not to review.

With respect to materials, Entergys witnesses explain that, contrary to Dr. Laheys testimony, the materials used in the IPEC RVIs and RPVs are fundamentally different, and have very different mechanical properties and behavior under irradiation.164 Therefore, many of his arguments and assertions regarding the behavior of RVI materials under irradiation and the potential for RVIs to undergo a transition from ductile to brittle behavior and are both incorrect and unsupported. Specifically, IP2 and IP3 RPVs are constructed primarily of low-alloy (carbon) steel, with stainless steel cladding, while the RVIs are made of wrought austenitic stainless steel, other stainless steels including Cast Austenitic Stainless Steel (CASS), or nickel-based alloys.165 As a result, the IPEC RVI materials exhibit less temperature-dependent changes in unirradiated mechanical properties than the RPV materials.166 Also, the mechanical properties of the IPEC RVI beltline materials, including the cast austenitic stainless steel (CASS) lower support column caps (LSCCs), do not change with irradiation to the same extent as low-alloy RPV materials do; i.e., they do not exhibit a shift in the ductile-to-brittle transition temperature.167 Overall, the RVI materials are far less susceptible to irradiation effects than the RPV.168 164 See id. § V.C.

165 See id. at A104.

166 See id.

167 See id. at A107.

168 See id. at A117.

Dr. Lahey also asserts that RVIs could be subject to pressure and/or thermal shock loads.169 It is not entirely clear what Dr. Laheys concerns are, however, to the extent his concern is that RVI components could fail due to PTS, it lacks basis.170 The RVIs have no pressure retaining function.171 For this reason, a PTS transient does not subject the RVI components to the stresses characteristic of the effects of a PTS event on an RPV.172 The design basis transients and loads on the RVIs are defined in the CLB for IPEC, and identified on a plant-specific basis in Chapter 4 of the Updated Final Safety Analysis Reports (UFSAR) for IP2 and IP3.173 As fully explained in Entergys testimony, the RVI AMP considers the full range of design basis loads, which are established in accordance with the CLB.174 B. Regulatory Guidance Addressing Aging Management of RVIs In Section VI, Entergys expert witnesses summarize Entergys full compliance with the specific regulatory guidance addressing the management of the effects of aging on RVIs. In particular, MRP-227-A is the current, NRC-approved version of EPRIs guidance on the aging management of RVIs.175 The NRC Staff thoroughly reviewed this guidance, approved it in a safety evaluation,176 and issued interim staff guidance updating the NUREG-1801, Revision 2 169 Revised Lahey Testimony at 16 (NYS000482).

170 See Entergys Testimony § V.C.3 (ENT000616).

171 See id. at A114.

172 See id.

173 See id. at A115.

174 See id. § V.C.3.

175 See id. §§ VI.B-C.

176 See id. § VI.C; see also NRC Staff Safety Evaluation for MRP-227-A (Letter from R. Nelson, NRC, to N.

Wilmshurst, EPRI, Revision 1 to the Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Prog[ra]m (MRP) Report 1016596 (MRP-227), Revision 0, Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines (Dec. 16, 2011) (SE for MRP-227-A) (ENT000230).

AMP (XI.M16A) to incorporate MRP-227-A.177 While Dr. Lahey believe[s] that the NRC-approved industry guidance is inadequate,178 Entergys witnesses show that he has largely disregarded and failed to challenge the substantial technical basis supporting that guidance.179 MRP-227-A is the result of a decade-long systematic evaluation of the effects of aging on RVIs.180 MRP-227-A was developed in four steps: (1) development of screening criteria for the applicable aging mechanisms; (2) screening of RVI components based on susceptibility to degradation; (3) functionality analysis and failure modes, effects, and criticality analyses (FMECA), which resulted in the binning of components into different risk severity and inspection categories; and (4) development of the inspection and evaluation guidelines and flaw evaluation methodology.181 The screening process explicitly considered potential combinations of aging effects, including all of the effects mentioned by Dr. Lahey.182 The aging management guidelines in MRP-227-A are supported by numerous underlying EPRI MRP technical studies, covering topics from aging degradation mechanisms and resulting effects, categorization of components, aging management strategies, acceptance criteria, and other topics.183 These technical studies document the considerable body of operating experience, state-of-the art research, and laboratory experiments that underpin the MRP-227-A guidelines.184 177 See Entergys Testimony § VI.A (ENT000616); see also LR-ISG-2011-04 (NYS000524).

178 See Revised Lahey Testimony at 39 (NYS000482).

179 See Entergys Testimony § VII (ENT000616).

180 See id. § VI.B.

181 See id. at A124.

182 See id. at A125.

183 See id. at A126.

184 See id.

The principal documents, along with MRP-227-A, total over 1600 pages of research and analysis and include:

• MRP-232: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals;185

• MRP-230: Functionality Analysis for Westinghouse and Combustion Engineering Representative PWR Internals;186

• MRP-228 Inspection Standard for PWR Internals;187

• MRP-210: Fracture Toughness Evaluation of Highly Irradiated PWR Stainless Steel Internal Components;188

• MRP-191: Screening, Categorization and Ranking of Reactor Internals of Westinghouse and Combustion Engineering PWR Designs;189

• MRP-175: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values;190

• MRP-134: Framework and Strategies for Managing Aging Effects in PWR Internals;191 and

• WCAP-17096-NP, Rev. 2, Reactor Internals Acceptance Criteria Methodology and Data Requirements.192 185 See MRP-232, EPRI Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals (Dec. 2008) (MRP-232) (ENT000642A-C); see also MRP-232, Revision 1, EPRI Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals (Dec. 2012) (MRP-232, Rev. 1) (ENT000643).

186 See MRP-230, EPRI Materials Reliability Program: Functionality Analysis for Westinghouse and Combustion Engineering Representative PWR Internals (Oct. 2009) (MRP-230) (ENT000644).

187 See MRP-228, EPRI Materials Reliability Program: Inspection Standard for PWR Internals (July 2009)

(NYS000323); see also MRP-228, Rev. 1, EPRI Materials Reliability Program: Inspection Standard for PWR Internals (Dec. 2012) (ENT000645).

188 See MRP-210, EPRI Materials Reliability Program: Fracture Toughness Evaluation of Highly Irradiated PWR Stainless Steel Internal Components (Dec. 2007) (MRP-210) (ENT000646).

189 See MRP-191, EPRI Materials Reliability Program: Screening, Categorization and Ranking of Reactor Internals of Westinghouse and Combustion Engineering PWR Designs (NYS000321).

190 See MRP-175 EPRI Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (ENT000631).

191 See MRP-134, EPRI Materials Reliability Program: Framework and Strategies for Managing Aging Effects in PWR Internals (June 2005) (ENT000647).

Based on these supporting reports, MRP-227-A provides comprehensive aging management guidelines, detailing inspections to detect the effects of aging (individually or in combination), methods to evaluate such aging effects, and considerations for repair or replacement of degraded components.193 MRP-227-A also defines risk-prioritized inspections to detect the effects of aging, and recommends methods to evaluate aging effects.194 Dr. Lahey generally disregards all of these analyses, and explains no disagreements with any information in them, despite the fact that they have nearly all been available to the State for several years through the mandatory disclosure process.195 This approach does not meet the States burden of moving forward with providing sufficient probative evidence to support its contention at the hearing stage.196 Instead, the State offers baseless legalistic justifications in an apparent attempt to excuse Dr. Lahey from the obligation to actually review the technical basis for the IPEC RVI AMP.

First, the State asserts that Entergy can only rely on its own RVI AMP documentation:

The RPVI AMP that is currently before the Board for review consists of the Revised and Amended RVI Plan, developed between 2012 and 2014 and approved by NRC Staff in the SSER2.

Thus, the adequacy of the AMP for RPVIs must stand or fall on the adequacy of these documents.197 192 See WCAP-17096-NP, Rev. 2, Reactor Internals Acceptance Criteria Methodology and Data Requirements (Dec. 2009) (ENT000635).

193 See Entergys Testimony at A125 (ENT000616).

194 See id.

195 See id. at A127. Notably, Entergy has disclosed all of these documents to the State under the mandatory disclosure process in 10 C.F.R. § 2.336and nearly all of them several years ago.

196 See Oyster Creek, CLI-09-07, 69 NRC at 269.

197 NYS Revised SOP at 19 (NYS000481).

Similarly, the State claims that Entergy has failed to submit an analysis of the allegedly synergistic effects of embrittlement, fatigue, and stress corrosion on RVIs.198 But these objections are groundless, as nothing in the NRCs regulations prevents Entergy from relying on an NRC-approved topical report as the basis for its RVI AMP, or requires Entergy to submit to the NRC Staff in this proceeding the MRP reports reviewed on a generic basis by the Staff. On the contrary, the Commission has endorsed and encouraged the practice of using generic guidance to improve efficiency, as has the Board in this proceeding.199 Overall, the States weak attempt to avoid its experts obligation to review the available technical basis for the RVI AMP fails.

Throughout Entergys Testimony, the witnesses explain that the IPEC RVI AMP is consistent with MRP-227-Aa point the State does not dispute. To challenge the adequacy of an NRC-approved guidance document, an intervenor must provide specificity and substantial support for such a challenge, in order to show unusual circumstances200 are present to overcome the special weight it is accorded in NRC proceedings.201 The State has not done so.

C. Entergys LRA Effectively Addresses Aging Management of RVIs and RPVs In Section VII, Entergys expert witnesses show that the IPEC RVI AMP provides reasonable assurance that the RVI components will continue to perform their intended functions, 198 Id. at 23 ¶ 10.

199 See Vt. Yankee, CLI-10-17, 72 NRC at 19 (noting that the GALL Report may be referenced in a license renewal application in the same manner as an approved topical report); Indian Point, CLI-15-6, slip op. at 21-22 (requiring unusual circumstances to be present to justify setting Staff guidance aside); Indian Point, LBP-13-13, 78 NRC at 297 (allowing Entergy to rely on the guidance in NSAC-202L as the basis for its flow-accelerated corrosion (FAC) AMP).

200 Indian Point, CLI-15-6, slip op. at 21.

201 Vt. Yankee, CLI-10-17, 72 NRC at 32-33, n.185.

consistent with the CLB, during the PEO, as required by 10 C.F.R. §§ 54.21(a)(3), (c)(1)(iii), and 54.29(a).

1. The IPEC RVI AMP

a. Overview of the RVI AMP and Inspection Plan Consistent with the guidance in MRP-227-A, the IPEC RVI AMP is divided into three main areas: (1) examinations and other inspections, along with a comparison of data to examination acceptance criteria, as defined in MRP-227-A and MRP-228; (2) process for resolution of indications that exceed examination acceptance criteria by entering them into the applicants Corrective Action Program; and (3) monitoring and control of reactor primary coolant water chemistry based on industry guidelines.202 The RVI Inspection Plan provides additional details on the inspections to be conducted under the RVI AMP, including: (1) the type of examinations; (2) the level of examination qualification; (3) the schedule of initial inspection and frequency of subsequent inspections; (4) the criteria for sampling and coverage; (5) the criteria for expansion of scope if unanticipated indications are found; (6) the acceptance criteria; (6) the methods for evaluation of examination results that do not meet the acceptance criteria; (7) provisions to update the program based on industry-wide results; and (8) contingency measures to repair, replace, or mitigate, beyond the information set forth in the RVI AMP.203

b. The RVI AMP Describes Inspections in Detail Dr. Lahey alleges that Entergy has not provided sufficient details about its inspection schedule.204 The IPEC RVI AMP, however, provides a comprehensive inspection schedule 202 See Entergys Testimony at A139 (ENT000616) (citing NL-12-037, Attach. 1 (NYS000496)).

203 See Entergys Testimony at A137 (ENT000616).

204 See Revised Lahey Testimony at 48-49 (NYS000482).

based on the guidance in MRP-227-A.205 RVI components are separated into four groups with aging management strategies specified for each group (Primary, Expansion, Existing Programs, and No Additional Measures) depending on: (1) the relative susceptibility to and tolerance of applicable aging effects; and (2) the existence of other programs that manage the effects of aging on those components.206 The inspections are specified in Table 5-2 (primary components), Table 5-3 (expansion components), and Table 5-4 (existing program components) of the Inspection Plan.207 Importantly, the inspection categorization is not dependent on analyzing the behavior of the individual components under accident loads.208 Rather, the EPRI MRP evaluated possible component failure under accident loads, and if the assumed failure could impact a design basis function the component was assigned to an inspection category using the appropriate inspection techniques and frequency of inspections.209 Dr. Lahey provides no critique of EPRIs categorization methodologyindeed, he does not mention it at all. Therefore, contrary to Dr.

Laheys bare assertions, the inspection schedule is comprehensive and adequate.210

c. The RVI AMP Manages the Effects of Aging on RVIs Regardless of the Underlying Aging Mechanism Dr. Lahey asserts that Entergy has considered various aging mechanisms that could affect the RVIs in silos, without considering the synergistic interactions between mechanisms.211 As a threshold matter, this position is directly contrary to the approach the Commission specified 205 See Entergys Testimony at A139 (ENT000616).

206 See id. at A138.

207 See id. at A139 (citing NL-12-037, Attach. 2, at 37-51 (NYS000496)).

208 See id. at A137.

209 See id.

210 Revised Lahey Testimony at 48-49 (NYS000482).

211 See id. at 14-15.

when it promulgated the license renewal rules in 10 C.F.R. Part 54. As described in Sections VII.A.3 and 5 of Entergys Testimony, the NRCs license renewal process has long focused on aging effects, rather than aging mechanisms.212 Since 1995, when the NRC promulgated its revised license renewal rules, the NRC has emphasized that the identification of individual aging mechanisms is not required as part of the license renewal review.213 Instead, the regulations in 10 C.F.R. Part 54 concentrate on ensuring that important structures, systems and components (SSCs) will continue to perform their intended functions during the PEO regardless of the particular aging mechanism.214 Consistent with this principle, the inspections conducted under the RVI AMP will look for evidence of any of the aging effects of concern, and appropriate action is taken if any relevant conditions related to those effects are discovered, regardless of their cause.215 In any event, to the extent Dr. Laheys claim is that the combined effects of multiple aging mechanisms are not addressed in the RVI AMP, he is mistaken. As specified in MRP-227-A, Section 3.2, the RVI AMP addresses the following eight age-related degradation mechanisms and their associated effects, each of which are further described in MRP-227-A:

• Stress corrosion cracking (SCC);

• IASCC;

• Wear;

• Fatigue;

• Thermal aging embrittlement;

• Irradiation embrittlement (also referred to as neutron embrittlement);

• Void swelling and irradiation growth; and 212 See Entergys Testimony §§ VII.A.3, 5 (ENT000616).

213 See id. at A143; see also Nuclear Power Plant License Renewal; Revisions, 60 Fed. Reg. 22,461, 22,463 (May 8, 1995) (Part 54 SOC) (NYS000016).

214 See Entergys Testimony at A143 (ENT000616); see also Part 54 SOC, 60 Fed. Reg. at 22,463 (NYS000016).

215 See Entergys Testimony at A143 (ENT000616).

• Thermal and irradiation-enhanced stress relaxation or irradiation-enhanced creep.216 For each of the eight mechanisms, MRP-227-A identifies the resulting aging effect, which will then be managed through inspections under the MRP-227-A guidelines.217 Notably, in most cases, the key effects are cracking, dimensional changes, or wear, but in all cases, as explained below, the inspections specified in MRP-227-A are designed to detect potential aging effects applicable to each RVI component, regardless of the underlying mechanism.218 Therefore, contrary to Dr. Laheys claims, the IPEC RVI AMP does not fail[] to consider how those interacting degradation mechanisms will impact the . . . RPV internals.219 With respect to the effects of embrittlement, no recommendations for inspection to determine embrittlement level are contained in the guidance because these mechanisms cannot be directly observed.220 But, as Entergys witnesses show, embrittlement is only an issue for RVI components if there is a crack.221 Therefore, while it is not possible to detect the level of embrittlement directly through visual inspection, MRP-227-A provides for inspections that detect the manifestation of significant thermal aging or neutron-irradiation embrittlement specifically, the potential growth of a pre-existing defect.222 Once a defect is discovered, its ability to withstand fatigue and combinations of both normal and accident shock loads is evaluated by either fracture mechanics analysis or a structural analysis (an engineering evaluation) using the lower bound fracture toughness; i.e., the 216 See id. at A144 (citing NL-12-037, Attach. 1 at 6 (NYS000496)).

217 See Entergys Testimony at A144 (ENT000616).

218 See id.

219 Revised Lahey Testimony at 49 (NYS000482).

220 See Entergys Testimony at A146 (ENT000616).

221 See id.

222 See id.

evaluation assumes the maximum level of embrittlement of the material.223 Thus, the program has compensated for any inability to directly determine the level of embrittlement through a conservative assumption employed during evaluation of inspection findings.224 Thus, reasonable assurance that the effects of aging will be adequately managed is provided without the need for direct observation or analysis of the level of embrittlement.225

d. The RVIs Are Robust and Highly Failure Tolerant Next, Entergys witnesses show that the RVI materials are constructed of damage-resistant and flaw-tolerant materials that have performed well in service at many plants for thousands of reactor years, with very little adverse operating experience.226 For example, ASME Code Section XI periodic inspections for PWR RVIs to date have included inspections of baffle-former bolts at several plants.227 Although these bolts are leading indicators for Westinghouse RVIs for the combination of irradiation-induced stress relaxation, void swelling, and IASCC, very few cracked or failed baffle-former bolts have been detected during these examinations and, in most cases, no cracked or failed bolts were detected at all.228

.229 223 See id. at A146 (citing MRP-227-A at 6-4 (NRC000114B)).

224 See Entergys Testimony at A146 (ENT000616).

225 See id.

226 See id. at A148 (citing MRP-227-A, App. A (NRC000114C)).

227 See Entergys Testimony A148, A150 (ENT000616).

228 See id. at A150.

229 See id. at A156.

Dr. Lahey disregards the overall operating experience when he labels Entergys plans regarding the baffle-former bolts as a wait-and-see approach.230 Entergys plans to inspect the baffle-former bolts are adequately and sufficiently specified in the record to provide the requisite reasonable assurance that the effects of aging on baffle-former bolts will be adequately managed.231 Specifically, at IPEC, Entergy is appropriately planning to inspect 100% of the baffle former bolts at IP2 in Spring 2016 and at IP3 in Spring 2019, with subsequent examinations on ten-year intervals.232 In preparation for these inspections, Entergy is preparing a technical justification (TJ) which will demonstrate that the ultrasonic testing (UT) inspections at IPEC will be capable of detecting defects exceeding 30% of the bolt cross-sectional area, as specified in Westinghouses evaluations of the baffle-former assembly.233 In addition, Entergy has contracted with Westinghouse to perform a more realistic plant-specific minimum bolting pattern analysis for IPEC.234 This evaluation will consider design basis loads for IP2 and IP3, including the dynamic effects and blowdown loads from pipe breaks of various sizes, low cycle thermal fatigue loads, high cycle flow induced vibration loads, and seismic loads.235 If inspections reveal degradation in baffle-former bolts, then this minimum bolting pattern will be used as the basis for engineering evaluations to determine the acceptability of the bolts following the required UT examinations from MRP-227-A.236 230 Revised Lahey Testimony at 55-56 (NYS000482).

231 See id. § VII.A.4.b.

232 See id. at A152.

233 See id. at A154.

234 See id. at A158.

235 See id.

236 See id. at A158.

Considering the operating experience, Dr. Laheys demand for wholesale replacement of the clevis insert bolts is likewise baseless.237 The clevis insert bolts are treated as an Existing Program component under the RVI AMP, because they are periodically inspected once every ten year interval under the ASME Code,Section XI program per Table IWB-2500.238 Entergy last inspected the clevis bolt inserts at IP2 in 2006 and at IP3 in 2009.239 Entergy has further evaluated recent operating experience regarding clevis insert bolts, and demonstrated that the existing ASME Code inspections are adequate at IPEC.240 Therefore, Entergys planned inspections of clevis insert bolts provide reasonable assurance that the effects of aging will be adequately managed.241

e. The RVI AMP Addresses Combinations of Aging Effects from Multiple Degradation Mechanisms Entergys experts further explain that MRP-227-A guidelines and the IPEC RVI AMP properly address applicable aging effects, including combinations of effects.242 Specifically, during the development of MRP-227-A, the EPRI MRP experts developed a set of standard screening criteria that were used to identify components with one or more potential aging mechanisms and how those effects could combine to affect functionality.243 That work, documented in MRP-175, identified thresholds for aging effects which were then used to develop the screening and categorization results documented in MRP-191 (NYS000321). These results, 237 Revised Lahey Testimony at 56-57 (NYS000482).

238 See Entergys Testimony at A163 (ENT000616).

239 See id.

240 See id. at A164.

241 See id. § VII.A.4.c.

242 See id. §§ VII.A.3, 5.

243 See id. at A168.

in turn, provide the technical basis for the functionality analysis in MRP-230 (ENT000644) and ultimately the examinations specified in MRP-227-A.244 Dr. Lahey raises no dispute with this information. Instead, he incorrectly assumes that his concerns regarding synergistic aging effects have never been addressed,245 and remarkably states he has discovered this important new issue.246 Entergys witnesses show that the RVI AMPand the license renewal process in generalappropriately consider the combined effects resulting from multiple aging mechanisms that could impact the IPEC RVIs.247 In short, Dr.

Lahey does not dispute how EPRI addressed his over-arching concern.248 This misconception is a fundamental reason why NYS-25 lacks merit.

Instead of discussing the substantial work of the EPRI MRP, Dr. Lahey claims that the Department of Energy (DOE), the NRC, and national laboratories have recently embarked on an ambitious R&D program to understand and resolve his concerns regarding the synergistic aging effects on nuclear plant components.249 But Dr. Lahey is mistaken, as this program is intended to address the long-term challenges and research needs for operating nuclear plants beyond 60 years, not beyond 40 years.250 Therefore, Dr. Laheys purported evidence falls short of identifying any deficiency regarding the PEO for IP2 and IP3 at issue here. In any event, the MRP-227 inspection and evaluation guidelines were based on state-of-the art engineering, and designed to accommodate the uncertainties associated with areas where research remains 244 See id.

245 See id. at A169.

246 Revised Lahey Testimony at 78 (NYS000482).

247 See Entergys Testimony §§ VII.A.3, 5 (ENT000616).

248 Revised Lahey Testimony at 14 (NYS000482).

249 See, e.g., id. at 17 (citing DOE, Light Water Sustainability Program, Material Aging and Degradation Technical Program Plan (Aug. 2014) (MAaD Program Plan) (NYS000485)).

250 See Entergys Testimony at A171 (ENT000616).

ongoing.251 Ultimately, the fact that certain research is ongoing is not an indication of any deficiency in an AMP.252 In fact, it is a sign of a healthy, constantly-improving program.253 What Dr. Lahey and the State are proposingto prevent an applicant from using a state-of-the art AMP because some research related to it remains ongoingwould transform the license renewal process into an open-ended research project, which the Commission explicitly intended to avoid when it promulgated 10 C.F.R. Part 54.254 Another misconception from Dr. Lahey is that he broadly implies that the synergy between combined aging effects may have a greater (i.e., worsening) effect than the sum of the individual mechanisms alone.255 This overlooks the fact that a combination of aging effects may in some cases have less of an effect, or even an improvement, in the materials resistance to aging.256 Fatigue and irradiation embrittlement, for example, do not interact synergistically, and in some cases irradiation can increase the fatigue life of RVI materials.257 The RVI AMP accounts for this recognized complexity, but Dr. Lahey does not.258 Dr. Lahey raises concerns about the potential combined effects of thermal and irradiation embrittlement on the CASS LSCCs.259 Entergys witnesses do not dispute that that there is ongoing research on this topic, but this is precisely why the NRC Staff identified the need for 251 See id.

252 See id.

253 See id.

254 See Part 54 SOC, 60 Fed. Reg. at 22,469.

255 See Revised Lahey Testimony at 17 (NYS000482).

256 See Entergys Testimony at A173 (ENT000616).

257 See id.

258 See id. at A174.

259 See Revised Lahey Testimony at 18 & 20 (NYS000482).

further evaluation of CASS components in its Safety Evaluation for MRP-227-A.260 Entergy, in response, demonstrated to the NRC Staff that the potential combined effects of thermal and irradiation embrittlement of CASS components is not an issue for the specific materials used at IPEC because the LSCCs do not have a high percentage of delta ferrite.261 Dr. Lahey does not mention or dispute this information. As a result, he has failed to carry the States burden of moving forward with evidence to support this contention at hearing.

f. The RVI AMP Addresses Appropriate Design Basis Loads, Including Seismic and LOCA Loads Dr. Lahey asserts that the RVI AMP has not adequately addressed potential shock loads on RVI components, but it is not entirely clear what Dr. Lahey means by the term shock loads.262 If the concern is with loads caused by postulated events that are greater than or different from those specified in the CLB for IP2 and IP3, or with scenarios that are beyond the plants licensing bases,263 then there is no requirement to address such loads in the RVI AMP.264 As previously noted, the adequacy of the CLB itself is not open to challenge in this proceeding.265 In particular, the State argues in its Revised SOP that the potential seismic hazard curves for the Indian Point site are higher than the seismic spectra developed in the 1970s during the proceedings concerning the initial operating licenses.266 To the extent the State is arguing that the seismic hazards considered in the CLB for IP2 and IP3 should be reconsidered in 260 See Entergys Testimony at A175 (ENT000616).

261 See id. at A176; NL-14-013, Letter from F. Dacimo to NRC Document Control Desk, Additional Information Regarding the License Renewal Application - Action Item 7 from MRP-227-A, Attach. 1 at 3, 4 (Jan. 28, 2014) (NYS000503).

262 See Entergys Testimony at A179 (ENT000616).

263 NYS Revised SOP at 17 (NYS000481).

264 See Entergys Testimony at A179 (ENT000616).

265 See Pilgrim, CLI-10-14, 71 NRC at 461; Oyster Creek, CLI-09-7, 69 NRC at 270.

266 NYS Revised SOP at 41.

this proceeding, such arguments are a collateral attack the license renewal rules in 10 C.F.R. Part 54.

If Dr. Lahey and the States concern is with loads that are within the CLB of IP2 and IP3, then such loads are appropriately addressed in the RVI AMP.267 The MRP-227-A inspection and evaluation guidelines are intended to detect conditions that may impair the continued functionality of the RVIs, under CLB loadsincluding loss-of-coolant accident (LOCA) and seismic loads.268 First, the MRP-227-A guidelines specify inspections of key irradiated components to assure that there are no cracks that could lead to failure and loss of functionality under transient loads.269 Without the presence of cracking, the ability of the irradiation-strengthened RVI-material to withstand shock loads is not degraded.270 Second, if a degraded component is discovered, then MRP-227-A requires the explicit evaluation of CLB loads, including accident and transient loads such as acoustic loads and rarefraction waves due to a LOCA in an engineering evaluation, to the extent such loads are part of the IP2 or IP3 CLB.271 For potentially irradiated components, the engineering evaluation assumes the component is embrittled.272 Dr. Lahey disregards rather than disputes this well-established analytical approach to accounting for design basis loads, and instead merely speculates that synergistic aging effects 267 See Entergys Testimony § V.A.6 (ENT000616).

268 See id. at A180.

269 See id.

270 See id.

271 See id. (citing MRP-227-A § 6 (NRC00014C)).

272 See Entergys Testimony at A171 (ENT000616).

and shock loads have not been considered.273 This is simply not enough at this stage of the proceeding for NYS to meet its burden of going forward with evidence to support its case.274

g. The RVI AMP Uses Appropriate Inspection Techniques MRP-227-A and its companion document, MRP-228, specify inspection techniques for those PWR RVI components that are most susceptible to the aging effects of concern and have the highest risk associated with failure.275 The standards for deployment of these inspection techniques and the necessary qualification requirements for both equipment and personnel are given in MRP-228.276 The NRC Staff reviewed and approved the selected inspection techniques in its Safety Evaluation for MRP-227-A.277 Dr. Lahey criticizes the use of VT-3 visual inspections as inadequate for use in inspections for cracking,278 but the adequacy of these techniques is explained in extensive detail in MRP-228, and further explained in the response to an NRC Staff non-concurrence on this topic.279 Again, the State and Dr. Lahey disregard the available information, rather than dispute the adequacy of the record on the use of VT-3 examinations. Moreover, the State and Dr. Lahey cannot simply rely on an NRC Staff nonconcurrence as the basis for their challenge to the use of VT-3 inspections. In particular, Dr. Lahey does not assert any expertise in this area and he offers no opinion of his own regarding the strength or weakness of the VT-3 inspections, or even offer 273 Revised Lahey Testimony at 15-16 (NYS000482); see also, e.g., Declaration of Richard T. Lahey, Jr. at 13 ¶ 19 (Feb. 13, 2015) (NYS000483) (New Yorks main concerns . . . have simply been ignored).

274 See Oyster Creek, CLI-09-07, 69 NRC at 268-70.

275 See Entergys Testimony at A186 (ENT000616).

276 See id.

277 See id. at A188.

278 See Revised Lahey Testimony at 62 (NYS000482).

279 See Entergys Testimony at A188, A132 (ENT000616).

a preferred alternative of his own.280 Again, Dr. Lahey must do much more at this stage of the proceeding.281

h. The RVI AMP Includes Appropriate Acceptance Criteria, Corrective Actions, and Preventive Actions Dr. Lahey criticizes the RVI AMP for allegedly failing to include objective criteria . . .

for corrective actions . . . .282 Entergys witnesses demonstrate that the RVI AMP includes appropriate acceptance criteria, corrective actions, and preventive actions, consistent with the applicable guidance and current operating practice. First, the RVI AMP contains specific, conservative examination acceptance criteria,283 based on the acceptance criteria in MRP-227-A.284 The inspections required in Section 4 of MRP-227-A and relied upon in the IPEC RVI AMP and Inspection Plan are designed to detect all of the pertinent aging effects described above, with conservative examination acceptance criteria.285 In most cases the examination acceptance criterion is any detectable degradation.286 The specific acceptance criteria will be carried forward into the program procedural documents, including the Pre-Inspection Engineering Packages prepared prior to each inspection.287 If examinations reveal conditions that do not meet the examination acceptance criteria set forth in the IPEC RVI Inspection Plan, then the discovery of the condition is entered into the 280 Revised Lahey Testimony at 62 (NYS000482); cf. USEC, Inc. (Am. Centrifuge Plant), CLI-06-10, 63 NRC 451, 472 (2006) (holding that mere references to documents without explanation or analysis does not supply an adequate basis for admitting a contention, and that conclusory statements proffered by an alleged expert do not provide sufficient support for a contention).

281 See Oyster Creek, CLI-09-07, 69 NRC at 268-70.

282 Revised Lahey Testimony at 49 (NYS000482).

283 See Entergys Testimony at A189 (ENT000616) (citing NL-12-037, Attach. 2 at 52-57 (NYS000496)).

284 See id. (citing MRP-227-A at 5-1 to 5-23 (NRC000114B)).

285 See Entergys Testimony at A189 (ENT000616).

286 See NL-12-037, Attach. 2 at 52-57 (NYS000496)).

287 See Entergys Testimony at A189 (ENT000616).

IPEC corrective action program for resolution.288 This could lead to: (1) a more detailed inspection; (2) an engineering evaluation; (3) repair; or (4) replacement of the affected component.289 MRP-227-A Section 6 provides an overview of the methodologies to be used for the development of engineering evaluations, which consider CLB loading and the characteristics of the material including the potential effects of embrittlement.290 In addition, if an inspection of a Primary component detects aging effects that exceed the Expansion Criteria specified in the tables of Section 5 of MRP-227-A, then inspections of corresponding Expansion components must take place.291 Contrary to Dr. Laheys demand, there is no further regulatory requirement for objective criteria for corrective actions.292 On the contrary, it would be impractical to establish pre-defined criteria in advance for all potential unsatisfactory examination results for all components.

Instead, such issues are handled on a case-by-case basis through engineering evaluations conducted under Entergys corrective action and quality assurance programs.293 This is fully consistent with how such matters are managed for operating plants under Part 50.

On the topic of Primary and Expansion components, Dr. Lahey criticizes the linkage between the core barrel girth weld (which is a leading indicator for IASCC and irradiation embrittlement) to the LSCCs because of the alleged differences between these components.294 288 See id. at A190.

289 See id.

290 See id. at A191, A192.

291 See id. at A193.

292 Revised Lahey Testimony at 49 (NYS000482).

293 See Entergys Testimony at A190 (ENT000616).

294 See Revised Lahey Testimony at 60 (NYS000482).

Entergys witnesses show that Dr. Laheys criticisms lack merit.295 In MRP-227-A, the LSCCs are an Expansion component linked to the control rod guide tube lower flanges as a Primary component.296 In response to NRC Staff RAIs, Entergy appropriately modified the RVI AMP to link the LSCCs to an additional Primary component that is an appropriate predictor of IASCC and irradiation embrittlement (IE) in the LSCCs: the core barrel girth weld.297 Although Dr.

Lahey disputes this linkage, he only asserts that the core barrel girth weld may be exposed to different aging mechanisms and shock loads than the LSCCs.298 This unsupported speculation does not directly challenge Entergys detailed plant-specific technical evaluation of the susceptibility of LSCCs to thermal embrittlement (TE), IE, and IASCC in support of its RVI AMP.299 Dr. Lahey and the State next assert that the IPEC RVI AMP manifestly does not include preventative actions.300 But this claim is based on a single phrase, lifted from the RVI AMP out of context, and in disregard of the remainder of the section which explains the preventive actions Entergy is taking related to RVIseven if they are being taken in the context of programs other than the RVI AMP.301 In any event, Entergys witnesses readily show that this claim is incorrect, as shown in the LRA itself and the RVI AMP. Specifically, the IPEC Water Chemistry Control program provides for preventive and mitigative action by maintaining primary water chemistry in 295 See Entergys Testimony § VII.A.8.c (ENT000616).

296 See MRP-227-A at 4-26 (NRC000114B).

297 See Entergys Testimony at A194 (ENT000616).

298 Revised Lahey Testimony at 60 (NYS000482).

299 See Entergys Testimony at A196 (ENT000616).

300 Revised SOP at 26 ¶ 21 (quoting NL-12-037, Attach. 1 at 5 (NYS000496)); see also Revised Lahey Testimony at 53 (NYS000482).

301 See NL-12-037, Attach. 1 at 5 (NYS000496).

accordance with EPRI guidelines.302 In addition, as part of the RVI aging management activities, Entergy replaced the IP2 split pins in 1995, the IP3 split pins in 2009, and will replace the IP2 split pins again in 2016.303 Further, Entergy will use the Fatigue Monitoring Program to track fatigue usage of RVI components with CUF analyses, thereby ensuring that the number of transients does not exceed the assumptions in the Westinghouse fatigue analyses.304 In addition, Entergy has implemented neutron flux reduction programs to minimize radiation effects and the resulting potential for degradation.305 Finally, the State and Dr. Lahey argue that Entergy must proactively replace RVIs, rather than managing the effects of aging through inspections and appropriate corrective actions (i.e.,

an AMP).306 This position is entirely unsupported and disregards the technical basis for the RVI AMP as documented in all of the supporting reports for MRP-227-A. Moreover, the States position is contrary to Commission precedent, because it amounts to a demand that aging effects be precluded, and seeks to negate the regulatory standard in 10 C.F.R. § 54.21(a)(3), which requires the applicant to show that there is reasonable assurance that the effects of aging will be adequately managed.307 Thus, the States view is entirely without merit.

i. The IPEC Fatigue Evaluations Appropriately Analyze Environmentally-Assisted Fatigue The State also claims, in NYS-25, that the EAF evaluations prepared by Westinghouse in support of the IPEC LRA, including EAF evaluations of RVI components, may be non-302 See Entergys Testimony at A203 (ENT000616) (citing NL-12-037, Attach. 1 at 5 (NYS000496); SSER 2 at 3-66 (NYS000507) (emphasis added)).

303 See id. (citing SSER 2 at A-15 (NYS000507)).

304 See id.

305 See id. (citing NL-12-037, Attach. 2 at 21-22 (NYS000496)).

306 See NYS Revised SOP at 31 (NYS000481); Revised Lahey Testimony at 79 (NYS000482).

307 See Seabrook, CLI-12-5, 75 NRC at 315.

conservative.308 In contrast to this speculation, Entergys witnesses show that the Westinghouse EAF evaluations are fully-documented, conservative engineering analyses that support a finding that the effects of fatigue, including the effects of the reactor water environment, will be adequately managed.309 Entergys witnesses explain this in their testimony on this contention,310 and on the metal fatigue contention (NYS-26B/RK-TC-1B),311 which is incorporated by reference into their NYS-25 testimony. Specifically, consistent with Entergys commitments and with standard ASME Code methods, Westinghouse recalculated each of the limiting CLB CUFs provided in Section 4.3 of the LRA for the RVIs to include reactor coolant environmental effects.312 Entergys witnesses show that there is no technical basis to apply any additional correction factors to account for the potential effects of embrittlement on fatigue life, beyond the correction factors specified in NRC guidance.313 Moreover, fatigue is one of the eight age-related degradation mechanisms evaluated during the development of the guidelines in MRP-227-A.314 As a result, the RVI AMP includes inspections intended to identify potential cracking caused by fatigue in susceptible RVI components.315 These inspection activities are in addition to, not in lieu of, the review of EAF for RVI components under the FMP.316 Thus, taken together, the RVI AMP and FMP provide 308 NYS Revised SOP at 17 (NYS000481).

309 See Entergys Testimony § VII.A.9 (ENT000616).

310 See id.

311 See generally Entergys NYS-26B/RK-TC-1B Testimony (ENT000679).

312 See Entergys Testimony at A206 (ENT000616).

313 See Entergys NYS-26B/RK-TC-1B Testimony at A76 (ENT000679).

314 See id. at A208.

315 See id.

316 See id.

reasonable assurance that the effects of aging due to fatigue on RVI components will be adequately managed throughout the PEO.

j. The RVI AMP Addresses Operating Experience Entergys witnesses describe how the industry has engaged in a decade-long effort to evaluate aging management of PWR RVIs, implement plant-specific AMPs for aging management of internals, develop a detailed RVI inspection program that has been approved by the NRC, and continues to collect and share relevant inspection results and operating experience for improved reliability.317 Consistent with the operating experience element of the RVI AMP and Commitment 40, Entergy will continue to review domestic and international operating experience during the PEO, and appropriately apply that operating experience in the IPEC RVI AMP, including updated inspection methods and improved methods of evaluating aging effects.318 In sum, Entergys experts demonstrate that the IPEC RVI AMP is consistent with MRP-227-A, as it uses state-of-the-art engineering and operating experience and demonstrated inspection techniques. Dr. Lahey and the State have overlooked rather than disputed the substantial technical basis developed by the EPRI MRP. Overall, the RVI AMP provides reasonable assurance that the effects of aging on the IP2 and IP3 RVIs will be adequately managed such that the intended functions of the IP2 and IP3 RVIs will be maintained consistent with the CLB throughout the PEO, as required by 10 C.F.R. §§ 54.21(a)(3), 54.21(c), and 54.29(a).

317 See id. at A211.

318 See id. § VII.A.10.

2. Entergys Aging Management Activities for RPVs As noted above, the States initial pleadings in 2007 on this contention focused primarily on the RPVs, rather than the RVIs, claiming that the information in the LRA on the RPV TLAAs did not include information on age-related accident analyses,319 and that an intermediate shell in IP2 will not meet the upper shelf energy acceptance criterion of 50ft-lb.320 Following the admission of contention NYS-25, Entergy submitted several RPV-related amendments to clarify its LRA, revise the description of how Entergy would address the then-proposed alternate PTS rule, and note the closure of certain RPV-related commitments.321 The State, however, has never amended NYS-25 to address or challenge these updates.322 Instead, the State has shifted its focus to RVIs.323 In particular, in Dr. Laheys prefiled testimony and the States statements of position on this contention, Dr. Lahey and the State do not allege any specific deficiency in Entergys LRA regarding the RPVs.324 To ensure a complete record, however, Entergys expert witnesses summarize the information regarding RPVs in the IPEC LRA and show that the LRA complies fully with 10 C.F.R. Parts 50 and 54 and is consistent with NRC Staff guidance regarding the management of the effects of aging and the evaluation of TLAAs for RPVs.325 In his most recent testimony, Dr. Lahey refers to certain documents discussing Branch Technical Position (BTP) 5-3, which is longstanding NRC guidance for estimating the initial, 319 NYS Petition at 224.

320 Id. at 226.

321 See Entergys Testimony at A50 (ENT000616).

322 See id. at A51.

323 NYS Revised SOP at 17 (NYS000481) ([t]he focus of Contention 25 is Entergys deficient AMP for RPVIs); see also Position Statement at 10 (NYS000293).

324 See Report (NYS000296); Lahey 2011 Testimony (NYS000294); Revised Lahey Testimony at 74 (NYS000482).

325 See Entergys Testimony §§ V.A,C, VII.B (ENT000616).

unirradiated transition temperature for certain RPVs when some of the required testing information is not available, and suggests that certain RPV embrittlement analyses may be non-conservative.326 But he cites this example only for the general principle that it is very important to preserve - rather than erode - operational safety margins as reactors age.327 Therefore, the States testimony contains no valid challenge to Entergys LRA with regard to the management of the effects of aging on RPVs. In any event, Entergys witnesses explain that this is actually a good example of the level of inherent conservatisms in embrittlement evaluations for RPVs. Specifically, the industry has now shown that other conservatisms and margin in RPV embrittlement calculations were more than sufficient to offset the potential non-conservatism identified in the BTP 5-3 methodology.328 VI. CONCLUSION For the foregoing reasons, the IPEC RVI and RPV aging management activities are consistent with NRC guidance, which is entitled to special weight, and satisfy all regulatory requirements. Therefore, Entergys LRA provides reasonable assurance that the effects of aging will be adequately managed throughout the PEO. The Intervenors have not carried their burden of providing sufficient evidence to support the claims made in NYS-25. Accordingly, NYS-25 should be resolved in Entergys favor.

326 Revised Lahey Testimony at 74 (NYS000482).

327 Id.

328 See Entergys Testimony at A79 (ENT000616).

Respectfully submitted, Executed in Accord with 10 C.F.R. § 2.304(d)

William B. Glew, Esq. Kathryn M. Sutton, Esq.

Entergy Nuclear Operations, Inc. Paul M. Bessette, Esq.

440 Hamilton Avenue Raphael P. Kuyler, Esq.

White Plains, NY 10601 MORGAN, LEWIS & BOCKIUS LLP Phone: (914) 272-3202 1111 Pennsylvania Avenue, N.W.

Fax: (914) 272-3205 Washington, D.C. 20004 E-mail: wglew@entergy.com Phone: (202) 739-3000 Fax: (202) 739-3001 E-mail: ksutton@morganlewis.com E-mail: pbessette@morganlewis.com E-mail: rkuyler@morganlewis.com Counsel for Entergy Nuclear Operations, Inc.

Dated in Washington, D.C.

this 4th day of September 2015 DB1/ 84307038 United States Nuclear Regulatory Commission Official Hearing Exhibit In the Matter of: Entergy Nuclear Operations, Inc.

(Indian Point Nuclear Generating Units 2 and 3)

ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #: ENTR00615-PUB-00-BD01 Identified: 11/5/2015 ENTR00615 Admitted: 11/5/2015 Rejected:

Withdrawn:

Stricken:

Revised: September 4, 2015 Other:

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) Docket Nos. 50-247-LR and

) 50-286-LR ENTERGY NUCLEAR OPERATIONS, INC. )

)

(Indian Point Nuclear Generating Units 2 and 3) )

) September 4, 2015 ENTERGYS STATEMENT OF POSITION REGARDING CONTENTION NYS-25 (EMBRITTLEMENT)

William B. Glew, Esq. Kathryn M. Sutton, Esq.

Entergy Nuclear Operations, Inc. Paul M. Bessette, Esq.

440 Hamilton Avenue Raphael P. Kuyler, Esq.

White Plains, NY 10601 MORGAN, LEWIS & BOCKIUS LLP Phone: (914) 272-3202 1111 Pennsylvania Avenue, N.W.

Fax: (914) 272-3205 Washington, D.C. 20004 E-mail: wglew@entergy.com Phone: (202) 739-3000 Fax: (202) 739-3001 E-mail: ksutton@morganlewis.com E-mail: pbessette@morganlewis.com E-mail: rkuyler@morganlewis.com Counsel for Entergy Nuclear Operations, Inc.

TABLE OF CONTENTS Page I. PRELIMINARY STATEMENT ....................................................................................... 1 A. Dr. Laheys Claims Regarding Synergistic Effects Lack Merit ......................... 4 B. The RVI AMP Appropriately Addresses Shock Loads...................................... 6 C. The RVI AMP Is Fully Adequate as It Specifies Appropriate Preventative Actions, Corrective Actions, and Acceptance Criteria .......................................... 6 D. The IPEC Fatigue Analyses Are Conservative and Support the Finding that the Effects of Fatigue Will Be Adequately Managed ..................................... 8 E. NYS Does Not Challenge the Adequacy of the LRA with Respect to RPVs........ 8 II. PROCEDURAL HISTORY OF CONTENTION NYS-25 ............................................... 9 A. Original Contention ............................................................................................ 10 B. First Amended Contention ................................................................................... 11 C. Revised RVI AMP and Inspection Plan............................................................... 13 D. 2011 NYS Testimony .......................................................................................... 14 E. Second Amended Contention .............................................................................. 14 III. APPLICABLE LEGAL AND REGULATORY STANDARDS ................................... 16 A. NYS Continues Its Attempts to Impermissibly Expand the Scope of this License Renewal Proceeding ............................................................................... 16

1. The License Renewal Review Is a Limited One ...................................... 16

2. The Reasonable Assurance Standard ....................................................... 19 B. License Renewal Guidance .................................................................................. 20

1. NUREG-1801 Is Entitled to Special Weight in This Proceeding ............ 22

2. Revisions to NUREG-1801...................................................................... 23 C. Burden of Proof.................................................................................................... 25 IV. ENTERGYS WITNESSES ............................................................................................ 26 A. Mr. Nelson F. Azevedo ........................................................................................ 26 B. Mr. Robert J. Dolansky ........................................................................................ 27 C. Mr. Alan B. Cox................................................................................................... 28 D. Mr. Jack R. Strosnider, Jr..................................................................................... 29 E. Mr. Timothy J. Griesbach .................................................................................... 30 F. Dr. Randy G. Lott ................................................................................................ 30 G. Mr. Mark A. Gray ................................................................................................ 31 V. ENTERGYS EVIDENCE AND ARGUMENTS ......................................................... 32

TABLE OF CONTENTS Page A. Technical Background on the Aging Management of RVIs and RPVs............... 32 B. Regulatory Guidance Addressing Aging Management of RVIs.......................... 35 C. Entergys LRA Effectively Addresses Aging Management of RVIs and RPVs .................................................................................................................... 39

1. The IPEC RVI AMP ................................................................................ 40

a. Overview of the RVI AMP and Inspection Plan ........................ 40

b. The RVI AMP Describes Inspections in Detail ........................... 40

c. The RVI AMP Manages the Effects of Aging on RVIs Regardless of the Underlying Aging Mechanism ........................ 41

d. The RVIs Are Robust and Highly Failure Tolerant ..................... 44

e. The RVI AMP Addresses Combinations of Aging Effects from Multiple Degradation Mechanisms ..................................... 46

f. The RVI AMP Addresses Appropriate Design Basis Loads, Including Seismic and LOCA Loads ........................................... 49

g. The RVI AMP Uses Appropriate Inspection Techniques............ 51

h. The RVI AMP Includes Appropriate Acceptance Criteria, Corrective Actions, and Preventive Actions ................................ 52

i. The IPEC Fatigue Evaluations Appropriately Analyze Environmentally-Assisted Fatigue ............................................... 55

j. The RVI AMP Experience Addresses Operating Experience ................................................................................... 57

2. Entergys Aging Management Activities for RPVs................................. 58 VI. CONCLUSION ................................................................................................................ 59

- ii -

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) Docket Nos. 50-247-LR and

) 50-286-LR ENTERGY NUCLEAR OPERATIONS, INC. )

)

(Indian Point Nuclear Generating Units 2 and 3) )

) September 4, 2015 ENTERGYS STATEMENT OF POSITION REGARDING CONTENTION NYS-25 (EMBRITTLEMENT)

Pursuant to 10 C.F.R. § 2.1207(a)(1) and the Atomic Safety and Licensing Boards (Board) Revised Scheduling Order,1 Entergy Nuclear Operations, Inc. (Entergy) submits this Statement of Position (SOP) regarding Contention NYS-25 proffered by New York State (NYS or the State). This Statement is supported by the Testimony of Nelson F. Azevedo, Robert J. Dolansky, Alan B. Cox, Jack R. Strosnider, Jr., Timothy J. Griesbach, Randy G. Lott, and Mark A. Gray Regarding Contention NYS-25 (Embrittlement) (Entergys Testimony)

(ENT000616), and the exhibits thereto (ENT000617 through ENT000721). For the reasons discussed below, NYS-25 lacks merit and should be resolved in Entergys favor.

I. PRELIMINARY STATEMENT NYS-25 is a safety contention, asserting that Entergys License Renewal Application does not include an adequate plan to monitor and manage the effects of aging due to embrittlement of the reactor pressure vessels (RPVs) and the associated internals.2 The 1

Licensing Board Revised Scheduling Order at 2 (Dec. 9, 2014) (unpublished) (Revised Scheduling Order).

2 New York State, Notice of Intention to Participate and Petition to Intervene at 223 (Nov. 30, 2007) (NYS Petition).

testimony of the States sole witnessDr. Richard T. Laheyfocuses on purported deficiencies in the aging management program (AMP) for reactor vessel internals (RVIs) at Indian Point Nuclear Generating Units 2 and 3 (IP2 and IP3, collectively Indian Point Energy Center or IPEC).

The States claims and testimony in NYS-25 are cumulative and overlapping with other contentions, redundant in some areas and contradictory in others. Such an approach is not only undisciplined, but also contrary to the Commissions intent in requiring intervenors to bring forward well-defined and adequately-supported contentions so that other parties to the proceeding are given full and fair notice of the intervenors actual claims.3 In NYS-25, the State claims that the license renewal application (LRA) is deficient for four reasons: (1) the RVI AMP is not based on an analysis that addresses synergistic degradation of RVIs caused by combinations of degradation mechanisms; (2) Entergy fails to consider the full range of transient shock loads to which RVIs may be subjected in the event of various postulated accidents; (3) the RVI AMP does not include a commitment to take preventative actions or to implement corrective actions, or provide specific enforceable acceptance criteria for some components; and (4) the AMP relies on fatigue predictions which are non-conservative and may not accurately predict fatigue-induced component failures.4 As demonstrated below, all four of these claims lack merit because Dr. Lahey and the State unfortunately have disregardedrather than disputedthe substantial body of technical 3

Pub. Serv. Co. of N.H. (Seabrook Station, Units 1 & 2), ALAB-899, 28 NRC 93, 97 (1988), aff'd sub nom.

Massachusetts v. NRC, 924 F.2d 311 (D.C. Cir. 1991), cert. denied, 502 U.S. 899 (1991).

4 See State of New York, Revised Statement of Position, Contention NYS-25 at 17 (June 9, 2015) (NYS Revised SOP) (NYS000481).

work that supports the IPEC RVI AMP. They have done so, despite the longstanding availability of the underlying technical work and supporting documentation, to the peril of their arguments.5 The IPEC RVI AMP is based on sound, state-of-the-art science and is fully compliant with the applicable criteria in NUREG-1801, as updated in the latest interim staff guidance. Dr.

Lahey demands a systematic safety evaluation to support the RVI AMP.6 He ignores, however, the systematic evaluation that the Electric Power Research Institute (EPRI) has already performed, and that Entergy relies upon. This horse-blinder approach falls woefully short of the requisite specific and substantial support7 necessary to overcome the special weight the Board must accord to Nuclear Regulatory Commission (NRC) Staff guidance, which endorses the approach Entergy has taken at IPEC.8 Moreover, several aspects of contention NYS-25 impermissibly challenge the current licensing basis (CLB) for IPEC.9 In summary, the State has failed to meet its burden of moving forward with sufficient evidence to show a deficiency in the RVI AMP,10 and Entergys Testimony fully refutes the States claims.

5 As explained further below, NYS retains the burden of going forward with its contention, even at the hearing stage. See AmerGen Energy Co., LLC (Oyster Creek Nuclear Generating Station), CLI-09-07, 69 NRC 235, 268-70 (2009), affd sub nom. N.J. Envtl. Fedn v. NRC, 645 F.3d 220 (2011).

6 Revised Pre-Filed Written Testimony of Dr. Richard T. Lahey, Jr., Regarding Contention NYS-25 at 51 (June 9, 2015) (NYS000482) (Revised Lahey Testimony).

7 See Entergy Nuclear Vt. Yankee, L.L.C. & Entergy Nuclear Operations, Inc. (Vt. Yankee Nuclear Power Station), CLI-10-17, 72 NRC 1, 33 n.185 (2010).

8 See Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 & 3), CLI-15-6, 81 NRC __,

slip op. at 19 (Mar. 9, 2015); NextEra Energy Seabrook, LLC (Seabrook Station, Unit 1), CLI-12-5, 75 NRC 301, at 314 n.78 (2012).

9 See Entergy Nuclear Generation Co. & Entergy Nuclear Operations, Inc. (Pilgrim Nuclear Power Station),

CLI-10-14, 71 NRC 449, 461 (2010); Oyster Creek, CLI-09-7, 69 NRC at 270.

10 See Oyster Creek, CLI-09-7, 69 NRC at 269.

Accordingly, Entergy has met its burden of showing, by a preponderance of the evidence,11 that NYS-25 lacks merit and should be resolved in Entergys favor.

A. Dr. Laheys Claims Regarding Synergistic Effects Lack Merit The States first claim is Dr. Laheys purported discover[y] that the IPEC RVI AMP fails to address potential synergistic degradation caused by combinations aging mechanisms.12 The IPEC RVI AMP, however, is based on a decade-long systematic expert evaluation of known and potential degradation mechanisms, resulting aging effects, and consequences of those effects for RVIs.13 This evaluation considered the relevant aging mechanisms, including multiple aging mechanisms which can produce combined effects on RVI components.14 NYS and Dr. Lahey have largely ignored the substantial state-of-the-art engineering and technical basis for the RVI AMP contained in the EPRI Materials Reliability Programs (MRP) MRP-227-A, Pressurized Water Reactor Internal Inspection and Evaluation Guidelines,15 its numerous supporting technical reports, and the plant-specific technical analyses submitted by Entergy for Indian Point and reviewed by the NRC Staff in SSER 2.16 11 See Pac. Gas & Elec. Co. (Diablo Canyon Nuclear Power Plant, Units 1 & 2), ALAB-763, 19 NRC 571, 577 (1984); Oyster Creek, CLI-09-07, 69 NRC at 263.

12 Revised Lahey Testimony at 78 (NYS000482).

13 See Testimony of Entergy Witnesses Nelson F. Azevedo, Robert J. Dolansky, Alan B. Cox, Jack R. Strosnider, Timothy J. Griesbach, Randy G. Lott, and Mark A. Gray Regarding Contention NYS-25 (Embrittlement)

§ VI.B (Aug. 10, 2015) (Entergys Testimony) (ENT000616).

14 See id.

15 MRP-227-A, EPRI Materials Reliability Program: Pressurized Water Reactor Internal Inspection and Evaluation Guidelines (Dec. 23, 2011) (MRP-227-A) (NRC000114A-F).

16 NUREG-1930, Supp. 2, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3 (Nov. 6, 2014) (SSER 2), available at ADAMS Accession No. ML14310A803.

In their testimony, Entergys expert witnesses explain that the IPEC RVI AMP is consistent with MRP-227-A, which is endorsed in current NRC guidance17 and incorporated in AMP XI.M16A in NUREG-1801 (the GALL Report).18 As a matter of law, the Commission has held that:

[A] license renewal applicant who commits to implement an AMP that is consistent with the corresponding AMP in the GALL Report has demonstrated reasonable assurance under 10 C.F.R. § 54.29(a) that the aging effects will be adequately managed during the period of extended operation.19 Because the IPEC RVI AMP is consistent with the current GALL Report AMP,20 it satisfies the regulatory requirements in 10 C.F.R. Part 54. Although an intervenor may challenge whether an applicants AMP is consistent with NRC Staff guidance,21 the State and Dr. Lahey do not make such an allegation. Instead, NYS and Dr. Lahey only proffer generic attacks against the NRC guidance and MRP-227-A on the topic of allegedly synergistic aging effects.22 But as explained further below, NYS must provide specific and substantial support23 to overcome the special weight accorded to this guidance in this proceeding.24 As Entergys testimony shows, NYS and Dr. Lahey have failed to clear that high hurdle.

17 LR-ISG-2011-04, Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors (Mar. 20, 2012) (LR-ISG-2011-04) (ENT000641).

18 See generally NUREG-1801, Generic Aging Lessons Learned Report, Revision 2 (Dec. 2010) (NUREG-1801, Rev. 2) (NYS00147A-D); NUREG-1801, Generic Aging Lessons Learned Report, Revision 1 (Sept.

2005) (NUREG-1801, Rev. 1) (NYS00146A-C).

19 Seabrook, CLI-12-5, 75 NRC at 315 see also AmerGen Energy Co., LLC (Oyster Creek Nuclear Generating Station), CLI-08-23, 68 NRC 461, 468 (2008).

20 See Entergys Testimony § VII.A (ENT000616).

21 Vt. Yankee, CLI-10-17, 72 NRC at 37.

22 Revised Lahey Testimony at 39 (NYS000482) (MRP-227-A is an inspection-based aging management plan, which I believe is inadequate).

23 See Vt. Yankee, CLI-10-17, 72 NRC at 33 n.185, 37.

24 Seabrook, CLI-12-5, 75 NRC at 314 n.78.

B. The RVI AMP Appropriately Addresses Shock Loads As to the States second claim, that Entergy has failed to consider potential shock loads that could impact RVI components, the inspection guidelines in MRP-227-A are designed to provide reasonable assurance that the RVIs will continue to perform their intended functions, consistent with the CLB.25 This includes maintaining functionality under CLB accident and seismic loads (or shock loads, as Dr. Lahey describes them).26 The design basis loads are established in accordance with the CLB and do not change because the units will operate during the period of extended operation (PEO).27 The guidelines in MRP-227-A, however, include inspections to provide reasonable assurance that components are not degraded due to the effects of aging, and provide instructions to guide engineering evaluations to determine the functionality of RVI componentsunder CLB loadsif any degradation is discovered.28 Accordingly, the IPEC RVI AMP fully accounts for CLB accident and transient loads.

Dr. Lahey has not discovered anything new.29 He instead impermissibly challenges the CLB.

To the extent his concerns are with shock loads caused by beyond design-basis accidents, this is an impermissible challenge to the CLB.30 C. The RVI AMP Is Fully Adequate as It Specifies Appropriate Preventive Actions, Corrective Actions, and Acceptance Criteria Next, NYS argues that the IPEC RVI AMP does not include a commitment to take preventative actions or to implement corrective actions, or provide specific enforceable 25 See Entergys Testimony § VII.A.6 (ENT000616).

26 See id. As explained further below, the adequacy of the CLB is not subject to attack in this proceeding. See Pilgrim, CLI-10-14, 71 NRC at 461; Oyster Creek, CLI-09-7, 69 NRC at 270.

27 See Entergys Testimony at A115 (ENT000616).

28 See id. § VI.B.

29 Revised Lahey Testimony at 78 (NYS000482).

30 See Pilgrim, CLI-10-14, 71 NRC at 461; Oyster Creek, CLI-09-7, 69 NRC at 270.

acceptance criteria for some components.31 Entergys witnesses demonstrate that these claims also lack merit.32 For example, Dr. Lahey and the State claim that the IPEC RVI AMP manifestly does not include preventative actions.33 They base this claim, however, on a crabbed interpretation of a single phrase lifted from the IPEC RVI AMP and taken out of context. In actuality, the remainder of the section discusses the preventive actions Entergy is taking related to RVIs some of which are being taken under the aegis of other AMPs at IPEC.34 In response to NYS other claims, Entergys witnesses fully demonstrate the adequacy and specificity of the inspection schedules, corrective actions, and acceptance criteria in the RVI AMP.35 As for Dr.

Laheys and the States concerns about the scope of the programthe components they identify are either not RVIs, and are covered by other aging management programs,36 or are active components outside the scope of aging management review (AMR) altogether.37 So yet again, NYS has not shouldered its evidentiary burden with respect to its claims in NYS-25.

31 NYS Revised SOP at 17 (NYS000481).

32 See Entergys Testimony § VII.A.7-8 (ENT000616).

33 NYS Revised SOP at 26 ¶ 21 (quoting NL-12-037, Letter from F. Dacimo, Vice President, Entergy, to NRC Document Control Desk, License Renewal Application - Revised Reactor Vessel Internals Program and Inspection Plan Compliant with MRP-227-A, Attach. 1 at 5 (Feb. 17, 2012) (NL-12-037) (NYS000496));

see also Revised Lahey Testimony at 53 (NYS000482).

34 See NL-12-037, Attach. 1 at 5 (NYS000496).

35 See Section V.C.1, infra.

36 For example, Dr. Laheys concerns about the j-groove welds, Revised Lahey Testimony at 45 (NYS000482),

are about a component that is managed under the Reactor Vessel Head Penetration Inspection AMP, not the RVI AMP. See Entergys Testimony at A101 (ENT000616).

37 For example, as explained below, control rods are active, short-lived components. See Indian Point, CLI-15-6, slip op. at 8; 10 C.F.R. § 54.21(a)(1).

D. The IPEC Fatigue Analyses Are Conservative and Support the Finding that the Effects of Fatigue Will Be Adequately Managed Finally, the States and Dr. Laheys claim that the environmentally-assisted fatigue (EAF) evaluations prepared in support of the IPEC LRA are not conservative and may be inaccurate lacks merit.38 Entergys EAF evaluations, including EAF evaluations of RVI components, are fully documented, conservative engineering analyses that support a finding that the effects of fatigue, including the effects of the reactor water environment, will be adequately managed.39 There is no technical basis to conclude that an additional correction factor is necessary to account for the effects of irradiation embrittlement on fatigue life.40 The RVI AMP, moreover, includes inspections intended to identify potential cracking caused by fatigue in susceptible RVI components, including irradiated RVI components.41 The inspection activities under the RVI AMP are in addition to, not in lieu of, the review of EAF for RVI components under the fatigue management program (FMP).42 E. NYS Does Not Challenge the Adequacy of the LRA with Respect to RPVs In addition to these claims, Entergy notes that NYS-25, as stated, alleges that Entergys LRA is inadequate with respect to the RPVs. Indeed, the original focus of NYS-25 in 2007 was on the RPVs themselves.43 But since 2011, NYS has focused almost exclusively on Entergys 38 NYS Revised SOP at 17 (NYS000481).

39 See generally Revised Testimony of Entergy Witnesses Nelson F. Azevedo, Alan B. Cox, Jack R. Strosnider, Randy G. Lott, Mark A. Gray and Barry M. Gordon Regarding Contention NYS-26B/RK-TC-1B (Aug. 10, 2015) (Entergys NYS-26B/RK-TC-1B Testimony) (ENT000679).

40 See id. at A76.

41 See Entergys Testimony § VII.A.7 (ENT000616).

42 See id. at A111.

43 See NYS Petition at 224, 226.

AMP for RVIs.44 While Dr. Laheys current testimony briefly alludes to some of his prior claims regarding the RPVs, both Dr. Lahey and the State stop short of alleging any specific deficiency in Entergys LRA regarding the RPVs.45 To ensure a complete record, however, Entergys expert witnesses demonstrate that the information regarding RPVs in the IPEC LRA complies fully with 10 C.F.R. Parts 50 and 54 and is consistent with NRC Staff guidance regarding the management of the effects of aging and the evaluation of time-limited aging analyses (TLAAs) for RPVs. The LRA therefore provides reasonable assurance that the effects of aging on the IPEC RPVs will be adequately managed, such that they will continue to perform their intended functions, consistent with the CLB.46 Neither the State nor Dr. Lahey provides any specific challenge to this information.

II. PROCEDURAL HISTORY OF CONTENTION NYS-25 As noted above, the claims in NYS-25 are cumulative and overlapping with other contentions. Specifically, the claims in NYS-25 substantially overlap those in contentions NYS-26B/RK-TC-1B (the metal fatigue contention) and NYS-38/RK-TC-5 (the safety commitments contention).47 Indeed, Dr. Laheys testimony regarding RVIs across the three 44 NYS Revised SOP at 17 (NYS000481); see also State of New York, Initial Statement of Position, Contention NYS-25 at 10 (Dec. 22, 2011) (NYS Initial SOP) (NYS000293).

45 See Report of Dr. Richard T. Lahey, Jr. in Support of Contentions NYS-25 and NYS-26B/RK-TC-1B at 13 (Dec. 20, 2011) (Report) (NYS000296); see also Pre-Filed Written Testimony of Dr. Richard T. Lahey, Jr.

Regarding Contention NYS-25 at 28-31 (Dec. 22, 2011) (Lahey 2011 Testimony) (NYS000294); Revised Lahey Testimony at 74 (NYS000482).

46 See 10 C.F.R. §§ 54.21, 54.29.

47 In objecting to the proposed amendments to NYS-25 and NYS-38/RK-TC-5 earlier this year, Entergy noted there was no discernible distinction between the two amended contentions, and asked the Board to separate the various claims in the interest of adjudicatory economy. Entergys Consolidated Answer Opposing Intervenors Motions to Amend Contentions NYS-25 and NYS-38/RK-TC-5 at 13 (Mar. 10, 2015), available at ADAMS Accession No. ML15069A677. The Board acknowledged that there is significant overlap, but found the States actions permissible. Memorandum and Order (Granting Motions for Leave to File Amendments to Contentions NYS-25 and NYS-38/RK-TC-5), at 14 (Mar. 31, 2015) (Second Order Amending NYS-25), available at ADAMS Accession No. ML15090A771.

contentions is substantively identical.48 Moreover, despite the significant developments and new information that has become available over the past three years or more, NYS has not replaced its 2011 SOP, testimony, or report with updated materials; it has merely added new information into the record in 2015 despite the fact that several prior positions and claims have been superseded by intervening events.49 Accordingly, Entergys Testimony and Statement of Position focus on the States most recent statement of position, testimony, and exhibits, filed on June 9, 2015. To assist the Boards review of the record, Entergy addresses challenges related to RVI and RPV aging management in its testimony, here, on contention NYS-25. Where there is an irreconcilable inconsistency, we focus on the most recent filings. Entergy addresses challenges related to metal fatigue (including EAF evaluations of RVI components) in its testimony on contention NYS-26B/RK-TC-1B.50 And it addresses specific challenges related to safety commitments (including RVI-related commitments) in its testimony on contentions NYS-38/RK-TC-5.51 A. Original Contention NYS first proffered Contention NYS-25 in 2007, as part of its initial Petition to Intervene.52 The States initial pleadings focused almost entirely on the RPV, rather than the RVIs, claiming that the information in the LRA on the TLAAs associated with the RPVs did not 48 Compare Revised Lahey Testimony (NYS00482) with Revised Pre-filed Written Testimony of Dr. Richard T.

Lahey, Jr. Regarding Consolidated Contention NYS-26B/RK-TC-1B, (June 9, 2015) (NYS000530) and Revised Pre-filed Written Testimony of Dr. Richard T. Lahey, Jr. Regarding Joint Contention NYS-38/RK-TC-5, (June 9, 2015) (NYS000562).

49 See Entergys Testimony at A65 (ENT000616).

50 See Entergys NYS-26B/RK-TC-1B Testimony (ENT000679).

51 See Revised Testimony of Entergy Witnesses Nelson F. Azevedo, Robert J. Dolansky, Alan B. Cox, Jack R.

Strosnider, Timothy J. Griesbach, Barry M. Gordon, Randy G. Lott, and Mark A. Gray Regarding Contention NYS-38/RK-TC-5 (Aug. 10, 2015) (Entergys NYS-38/RK-TC-5 Testimony) (ENT000699) .

52 See NYS Petition at 223-27.

include information on age-related accident analyses53 and that an intermediate shell in IP2 will not meet the upper shelf energy acceptance criterion of 50ft-lb.54 The original contention did not provide any basis for the States concerns about the associated internals of the RPVs, except to list the names of certain RVI components where Dr. Laheys [c]oncerns over embrittlement applied, and offer the unclear observation that RPV internals in IP3 imply operational limits for extended life operations due to the high [nil ductility temperature] NDT associated with the predicted irradiation-induced embrittlement.55 Entergy and the NRC Staff opposed admission of NYS-25.56 Entergys objections included that the proposed contention repeatedly confused the RPV and RVIs, was inadequately supported by the bare assertions in Dr. Laheys declaration, and failed to raise a genuine dispute with any information in the LRA.57 The Board admitted NYS-25 in 2008.58 B. First Amended Contention Consistent with Commitment 30 in the original LRA,59 Entergy submitted a detailed AMP for IP2 and IP3 RVIs on July 14, 2010.60 This AMP fully described Entergys program to 53 Id. at 224.

54 Id. at 226.

55 Id. at 224-226.

56 See Answer of Entergy Nuclear Operations, Inc. Opposing New York State Notice of Intention to Participate and Petition to Intervene at 135-41 (Jan. 22, 2008), available at ADAMS Accession No. ML080300149 (Entergy Answer); NRC Staffs Response to Petitions for Leave to Intervene Filed by [NYS] at 75-77 (Jan.

22, 2008), available at ADAMS Accession No. ML080230543.

57 See Entergy Answer at 135-41.

58 Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 & 3), LBP-08-13, 68 NRC 43, 131 (2008).

59 In Commitment 30, consistent with the then-current NRC guidance in NUREG-1801, Revision 1, Entergy committed to participate in industry programs for investigating and managing aging effects on RVIs, to evaluate and implement industry programs applicable to RVIs, and to submit an RVI inspection plan not less than 24 months before entering the PEO. See IPEC License Renewal Application at 3.1-7 to 3.1-8, 3.1-9 to 3.1-10 (Apr. 2007) (LRA) (ENT000015A); see also NUREG-1801, Generic Aging Lessons Learned Report, Rev. 1 at 7-30, tbl. 1 (Sept. 2005) (NUREG-1801) (NYS00146A).

manage the effects of aging on RVIs using guidance developed from nearly a decade of extensive industry research and set forth in EPRI Materials Reliability Program documents MRP-227 and MRP-228.61 Following Entergys submittal of its RVI AMP, NYS filed a motion to submit additional bases for NYS-25.62 The amended contention alleged, among other things, that the RVI AMP did not consider synergistic aging effects or potential shock loads, did not provide sufficient details on baseline and periodic inspections, did not provide sufficient details on corrective actions, including repair or replacement, disavow[ed] preventive actions, and relied on vague future commitments.63 Entergy and the NRC Staff objected on both timeliness and substantive grounds.64 Entergy, among other things, argued that NYS failed to challenge directly-relevant information in the RVI AMP.65 On July 6, 2011, the Board admitted the amended NYS-25.66 60 See NL-10-063, Letter from F. Dacimo to NRC Document Control Desk, Amendment 9 to License Renewal Application (LRA) - Reactor Vessel Internals Program (July 14, 2010) (NL-10-063) (NYS000313).

61 See NL-10-063, Attach. 1 at 82-84 (NYS000313).

62 See State of New Yorks Motion for Leave to File Additional Bases for Previously-Admitted Contention NYS-25 in Response to Entergys July 14, 2010 Proposed Aging Management Program for Reactor Pressure Vessels and Internal Components (Sept. 15, 2010), available at ADAMS Accession No. ML103050402.

63 See id.; Decl. of Richard T. Lahey, Jr., ¶¶ 13-15 (Sept. 15, 2010) (attached to motion), available at ML12335A461.

64 See Applicants Answer to Amended Contention New York State 25 Concerning Aging Management of Embrittlement of Reactor Pressure Vessel Internals (Oct. 12, 2010), available at ADAMS Accession No. ML103010104 (Entergys 2010 Answer); NRC Staffs Answer to State of New Yorks Motion for Leave to File Additional Bases for Previously-Admitted Contention NYS-25 (Oct. 12, 2010), available at ADAMS Accession No. ML102850764.

65 See generally Entergys 2010 Answer. As Entergys witnesses show throughout their testimony, for the past five years Dr. Lahey and the State have continued to disregard, rather than dispute the technical basis for the RVI AMP.

66 Licensing Board Order (Ruling on Pending Motions for Leave to File New and Amended Contentions) at 27 (July 6, 2011) (unpublished).

C. Revised RVI AMP and Inspection Plan Again, consistent with Commitment 30, Entergy submitted its RVI Inspection Plan on September 28, 2011, two years prior to entering the PEO for IP2.67 The Inspection Plan was based on detailed inspection guidance in MRP-227, and fully addressed the NRC Staffs action items and conditions in the Safety Evaluation for MRP-227, Revision 0.68 It also included a comprehensive schedule for inspections of RVI components at IPEC.69 The RVI Inspection Plan governed both IP2 and IP3.70 After EPRI issued the NRC-approved aging management guidance for RVIs in MRP-227-A (discussed further below), Entergy submitted a revised RVI AMP and Inspection Plan for both IP2 and IP3 based on MRP-227-A on February 17, 2012.71 IP2 and IP3 were among the first units in the U.S. fleet to prepare RVI AMPs based on the state-of-the-art NRC Staff-approved guidance in MRP-227-A and to have such an AMP reviewed by the NRC Staff as part of an LRA.72 From 2012 through 2014, the NRC Staff issued detailed requests for additional information (RAI) to Entergy on this first-of-a-kind AMP.73 Following Entergys submission of significant additional technical information in response to these RAIs, the NRC Staff approved Entergys revised RVI AMP and Inspection 67 See NL-11-107, Letter from F. Dacimo, Vice President, Entergy, to NRC Document Control Desk, License Renewal Application - Completion of Commitment # 30 Regarding the Reactor Vessel Internals Inspection Plan (Sept. 28, 2011) (NYS000314).

68 See id.

69 See id., Attach. 1 at 35-39, tbl.5-2.

70 See Entergys Testimony at A58 (ENT000616).

71 NL-12-037 (NYS000496).

72 See Entergys Testimony at A60 (ENT000616).

73 See NUREG-1930, Supp. 2, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3 at B-2 to B-7 (Nov. 2014) (SSER 2) (NYS000507).

Plan as documented in SSER 2 issued on November 6, 2014.74 NRC Staff concluded that Entergys LRA for IP2 and IP3 demonstrates that the effects of aging on RVI components will be adequately managed, as required under 10 C.F.R. §§ 54.21(a)(3) and (c)(1)(iii), and the RVI Inspection Plan implements the elements of the RVI AMP in an acceptable manner.75 D. 2011 NYS Testimony Along with the other Track 1 contentions, the State submitted its initial statement of position, prefiled testimony, and exhibits on NYS-25 in December 2011.76 Before any further testimony was filed on NYS-25, however, the Board placed NYS-25 on the schedule for the second set of hearings in this proceeding (i.e., Track 2).77 E. Second Amended Contention Following the publication of SSER 2,78 the State filed a Motion for Leave to Supplement Previously-Admitted Contention NYS-25 focusing the IPEC RVIs AMP.79 The States Second Amended Contention alleged that the RVI AMP remained deficient because it does not: (1) address or manage the combined synergistic aging effects of embrittlement, fatigue, and other aging mechanisms; (2) maintain safety margins during the PEO by, for example, repair or replacement of the RVIs, and does not account for the full range of transient shock loads; and 74 See id. at 3-26, 3-59; see also id. at B-2 to B-7.

75 SSER 2 at 3-26, 3-59 (NYS000507).

76 See NYS Initial SOP at 10; Report at 13; see also Pre-Filed Written Testimony of Dr. Richard T. Lahey, Jr.

Regarding Contention NYS-25 at 28-31 (Dec. 22, 2011) (Lahey 2011 Testimony) (NYS000294);

Supplemental Report of Dr. Richard T. Lahey, Jr. in Support of Contention NYS-25 and NYS-26B/RK-TC-1B (Supplemental Lahey Report) (NYS000297).

77 See Licensing Board Order (Granting NRC Staffs Unopposed Time Extension Motion and Directing Filing of Status Updates) at 2 (Feb. 16, 2012) (unpublished).

78 NUREG-1930, Supp. 2, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3 (Nov. 6, 2014) (SSER 2), available at ADAMS Accession No. ML14310A803.

79 State of New Yorks Motion for Leave to Supplement Previously-Admitted Contention NYS-25 (Feb. 13, 2015) (Second Motion to Amend), available at ADAMS Accession No. ML15044A493.

(3) include required preventative or corrective actions or acceptance criteria for the baffle-former bolt inspections. The State further alleged that the Westinghouse EAF calculations prepared for Indian Point are allegedly inadequate.80 The Second Amended NYS-25 did not allege any deficiencies in the IPEC LRA regarding the RPVs.81 Entergy objected on both timeliness and substantive grounds.82 Once again, Entergy objected on the grounds that the State continued to disregard, rather than dispute the technical basis for the RVI AMP.83 In particular, Entergy showed that it had disclosed to the State a substantial body of technical documentation supporting MRP-27-A and, the IPEC RVI AMP,84 yet the State did not even mention this information in its proposed amended contention.85 On March 31, 2015, the Board granted the States motion without altering or amending the contention.86 Thereafter, under the Boards scheduling orders, NYS filed its revised statement of position, testimony, and additional exhibits on June 9, 2015.87 80 New York State February 2015 Supplement to Previously-Admitted Contention NYS-25 at 1-3 (Feb. 13, 2015)

(Second Supplement to NYS-25), available at ADAMS Accession No. ML15044A491.

81 Although the State does not allege any deficiencies in the LRA related to RPVs, Dr. Lahey did note the suggestion of a potential non-conservatism in BTP 5-3, related to RPVs, which is discussed in further detail below.

82 Entergys Consolidated Answer Opposing Intervenors Motions to Amend Contention NYS-25 and NYS-38/RK-TC-5 (Mar. 10, 2015), available at ADAMS Accession No. ML15069A677.

83 See id. at 18-23.

84 See id. at 8 & n.30.

85 See id. at 19.

86 Second Order Amending NYS-25 at 10.

87 See generally NYS Revised SOP (NYS000481); Revised Lahey Testimony (NYS000482); exhibits NYS000483 through NYS000528.

III. APPLICABLE LEGAL AND REGULATORY STANDARDS As demonstrated below, the IPEC RVI AMP fully meets the applicable legal and regulatory requirements in 10 C.F.R. Part 54. In addition to the lack of technical merit, the States claims in NYS-25 are legally deficient in that they are contrary to the limited scope of the license renewal rule in 10 C.F.R. Part 54 as well as the NRCs reasonable assurance standard.

NYS has, moreover, failed to carry its burden of going forward on its contention and overcoming the special weight accorded to NRC Staff guidance documents.

A. NYS Continues Its Attempts to Impermissibly Expand the Scope of this License Renewal Proceeding

1. The License Renewal Review Is a Limited One The State continues to bring forward claims that attempt to expand the scope of this license renewal proceeding beyond the bounds that are clearly established by 10 C.F.P. Part 54 and Commission precedent. For example, the States claims related to the consideration of shock loads involve concerns about postulated accidents or events that are beyond the design basis of IP2 and IP388 and which are clearly outside the limited scope of this license renewal proceeding. Similarly, the States claims regarding alleged deficiencies in the seismic hazard curves for IP2 and IP3,89 demands for wholesale repair or replacement of RVIs in lieu of an AMP,90 and claims regarding active components such as control rods and control rod drive mechanisms,91 also fall beyond the bounds of this proceeding and must be rejected by the Board.

88 E.g., NYS Revised SOP at 17 (NYS000481).

89 See id. at 41.

90 See id. at 31.

91 See id. at 25 (Entergys 2011 AMP for RPVIs was inadequate with respect to the embrittlement of the control rod drives . . . .).

Specifically, 10 C.F.R. Part 54 is focused on managing the effects of aging on passive, long-lived components. It does not include a review of the adequacy of a plants CLB, including its design basis. Nor does it include a review of ongoing regulatory matters that are fully addressed under 10 C.F.R. Part 50 and by NRC inspection and enforcement activities.92 The Commissions license renewal regulations clearly reflect this distinction between 10 C.F.R. Part 54 aging management issues on the one hand, and ongoing 10 C.F.R. Part 50 regulatory process (e.g., the adequacy of the plants design basis) on the other.93 The underlying adequacy of the CLB itself is outside the scope of license renewal and is not open to challenge in this proceeding.94 The license renewal review is premised upon the determination that, with the exception of aging management issues, the NRCs ongoing regulatory process is adequate to ensure that the CLB of an operating plant provides and maintains an acceptable level of safety.95 Thus, the States challenges to the adequacy of design basis loads on plant components, including seismic loads, must be rejected on legal grounds aloneputting aside their factual inadequacies. Likewise, as further explained below, the Board must also reject the States concerns regarding control rods, which are active components and consumables, as beyond the scope of this proceeding.96 The State further challenges the license renewal rule when it argues that, instead of implementing an AMP, Entergy should take proactive steps to repair or replace aging RVI 92 See Fla. Power & Light Co. (Turkey Point Nuclear Generating Plant, Units 3 and 4), CLI-01-17, 54 NRC 3, 7-9 (2001); see also Indian Point, CLI-15-6, slip op. at 8; 10 C.F.R. § 54.21(a)(1).

93 Turkey Point, CLI-01-17, 54 NRC at 7; see also id. at 9 (The current licensing basis . . . includes the plant-specific design basis information documented in the plants most recent Final Safety Analysis Report . . . and any orders, exemptions, and licensee commitments that are part of the docket for the plants license . . . .).

94 See Pilgrim, CLI-10-14, 71 NRC at 461; Oyster Creek, CLI-09-7, 69 NRC at 270.

95 See Final Rule, Nuclear Power Plant License Renewal; Revisions, 56 Fed. Reg. 64,943, 64,946 (Dec. 13, 1991).

96 See Indian Point, CLI-15-6, 81 NRC at __, slip op. at 8; 10 C.F.R. § 54.21(a)(1).

components.97 The license renewal rule requires the applicant to show that there is reasonable assurance that the effects of aging will be adequately managednot that aging effects will be precluded.98 The review of TLAAs for license renewal is also well-defined in Part 54 and not subject to challenge in this proceeding. Certain in-scope components are subject to time-limited calculations or analyses that are part of the CLB, known as TLAAs. TLAAs must be evaluated for the PEO.99 In doing so, an applicant must: (i) show that the original TLAAs will remain valid for the PEO; (ii) revise and extend the TLAAs to be valid for a longer term, such as 60 years; or (iii) otherwise demonstrate that the effects of aging will be adequately managed during the renewal term.100 As they relate to this contention, the EAF evaluations prepared by Westinghouse for IPEC address all components with a CLB cumulative usage factor (CUF) analysis.101 The EAF evaluations are part of the Fatigue Management Program (FMP)the program that Entergy is using to resolve the CUF TLAAs under 10 C.F.R. § 54.21(c)(iii).102 But the CLB CUF analysis is a fatigue analysis, not a general analysis of all aging effects.103 Thus, to the extent the State and Dr. Lahey argue that irradiation embrittlement or other degradation mechanisms be considered in EAF evaluations,104 their claims are a challenge to the CLB and the license renewal rule. As further explained below, Entergy uses the RVI AMP to manage the 97 NYS Revised SOP at 31 (NYS000481).

98 See Seabrook, CLI-12-5, 75 NRC at 314-15.

99 See 10 C.F.R. § 54.21(c)(1).

100 See id.

101 See Entergys NYS-26B/RK-TC-1B Testimony § V.C (ENT000679).

102 See id. at A97.

103 See id. § IV.A.

104 See NYS Revised SOP at 41 (NYS000481); Revised Lahey Testimony at 19-20 (NYS000482).

effects of aging on RVI components caused by all pertinent aging mechanisms, including the effects of fatigue, embrittlement, and stress corrosion cracking.

2. The Reasonable Assurance Standard Pursuant to Section 54.29(a), the NRC will issue a renewed license if it finds that the applicant has identified actions that have been taken or will be taken such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the CLB.105 In addition to the limitations on the scope of this proceeding set forth in 10 C.F.R. § 54.21(a)(1), the reasonable assurance standard does not require Entergy to show protection against speculative, postulated events that are beyond the design basis of the plant,106 or to preclude all potential aging effects by replacing the RVIs.107 It also requires Dr.

Lahey to provide more than the speculation he repeatedly presents in support of his claims.108 Longstanding precedent makes clear that the reasonable assurance standard does not require an applicant to meet an absolute or beyond a reasonable doubt standard.109 Rather, the Commission takes a case-by-case approach, applying sound technical judgment and verifying the applicants compliance with Commission regulations.110 Branch Technical Position RLSB-1, in the Standard Review Plan for Review of License Renewal (SRP-LR), explains that the 105 10 C.F.R. § 54.29(a).

106 NYS Revised SOP at 17 (NYS000481).

107 Id. at 31.

108 See, e.g., Revised Lahey Testimony at 16 (seriously embrittled and fatigued RPV internals may not be able to survive the shock loads), 16-17 (multiple aging mechanisms that occur in a reactor core (including fatigue, irradiation embrittlement, and corrosion) may result in cumulative material degradation), 40 (highly embrittled and fatigued RVI components may not have signs of degradation that can be detected by an inspection, but such weakened components could nonetheless fail) (emphasis added) (NYS000482).

109 Oyster Creek, CLI-09-7, 69 NRC at 262 n.142; Commonwealth Edison Co. (Zion Station, Units 1 & 2),

ALAB-616, 12 NRC 419, 421 (1980); N. Anna Envtl. Coal. v. NRC, 533 F.2d 655, 667-68 (D.C. Cir. 1976)

(rejecting the argument that reasonable assurance requires proof beyond a reasonable doubt and noting that the licensing board equated reasonable assurance with a clear preponderance of the evidence).

110 See Oyster Creek, CLI-09-7, 69 NRC at 262, n.143, 263; Pilgrim, CLI-10-14, 71 NRC at 465-66.

license renewal process is not intended to demonstrate absolute assurance that structures and components will not fail, but rather that there is reasonable assurance that they will continue to perform their intended functions consistent with the CLB during the PEO.111 Indeed, the plain language of the regulations, and Commission decisions interpreting those regulations, state that the central question for a license renewal applicant is whether aging management activities have been identified and actions have been or will be taken to provide reasonable assurance of continued safety.112 Importantly, these regulations do not require the applicant to demonstrate that aging effects be precluded,113 but are oriented in large part toward identifying actions that will be taken in the future.114 B. License Renewal Guidance As previously noted, nowhere in their filings do the State or Dr. Lahey allege that IPEC RVI AMP is inconsistent with NRC Staff guidance. Instead, they attack the NRC guidance itself, which endorses the industrys detailed guidelines in MRP-227-A.115 While the Commission has not forbidden such arguments, the State and Dr. Lahey face a high bar to overcome the special weight accorded to the NRC Staffs guidance on license renewal.116 As 111 SRP-LR, Revision 1, Appx. A, at A.1-1 (NYS000195); NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Rev. 2 (Dec. 2010) (SRP-LR, Revision 2)

(NYS000161).

112 See 10 C.F.R. §§ 54.21(a)(3), 54.29(a)(1).

113 See Seabrook, CLI-12-5, 75 NRC at 314-15.

114 See Vt. Yankee, CLI-10-17, 72 NRC at 36.

115 Revised Lahey Testimony, at 39 (NYS000482) (MRP-227-A is an inspection-based aging management plan, which I believe is inadequate).

116 See, e.g., Indian Point, CLI-15-6, slip op. at 19; Seabrook, CLI-12-5, 75 NRC 314 n.78 (quoting Private Fuel Storage, L.L.C. (Indep. Spent Fuel Storage Installation), CLI-01-22, 54 NRC 255, 264 (2001)); see also id.

(We recognize, of course, that guidance documents do not have the force and effect of law. Nonetheless, guidance is at least implicitly endorsed by the Commission and therefore is entitled to correspondingly special weight) (quoting Yankee Atomic Elec. Co. (Yankee Nuclear Power Station), CLI-05-15, 61 NRC 365, 375 n.26 (2005)).

explained throughout this Statement of Position, and as Entergys witnesses have demonstrated, the State and Dr. Lahey have not cleared that bar.

The two primary license renewal guidance documents issued by the NRC Staff are NUREG-1801, the Generic Aging Lessons Learned Report or GALL Report,117 and NUREG-1800, the Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, or SRP-LR.118 The SRP-LR provides guidance to NRC staff for conducting their review of LRAs and provides acceptance criteria for determining whether the applicant has met the regulatory requirements for license renewal.119 NUREG-1801 provides the technical basis for the SRP-LR and contains the NRC Staffs generic evaluation of programs that manage the effects of aging during the PEO, and meet the requirements of 10 C.F.R. Part 54.120 NUREG-1801 indicates that many existing, current-term programs are also adequate to manage the aging effects for particular structures or components for license renewal. Thus, programs that are consistent with NUREG-1801 are accepted by the Staff as adequate to meet the requirements of the license renewal rule.121 The Commission has endorsed NUREG-1801 because it is based on extensive research and evaluation of operating 117 See generally NUREG-1801, Rev. 1 (NYS00146A-C); NUREG-1801, Rev. 2 (NYS00147A-D).

118 See generally Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Rev. 1 (Sept. 2005) (SRP-LR, Rev. 1) (NYS000195); NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Rev. 2 (Dec. 2010) (SRP-LR, Rev. 2)

(NYS000161).

119 See SRP-LR, Rev. 2 at 1-3 (NYS00146A).

120 See NUREG-1801, Rev 1, at 3-4 (NYS00146A).

121 See id. at 3.

experience derived from a comprehensive set of sources.122 NUREG-1801 is also subject to stakeholder review and comment.123

1. NUREG-1801 Is Entitled to Special Weight in This Proceeding The Commission has held that a license renewal applicants use of the guidance in NUREG-1801 satisfies regulatory requirements under 10 C.F.R. Part 54.124 Also, where the NRC develops a guidance documentsuch as NUREG-1801to assist in compliance with applicable regulations, that document is entitled to special weight in NRC proceedings.125 In particular, for license renewal safety issues, an applicants use of an AMP identified in NUREG-1801 constitutes reasonable assurance that it will manage the targeted aging effect during the renewal period.126 The Commission has reiterated this principle, holding that a commitment to implement an AMP that the NRC finds is consistent with NUREG-1801 constitutes an acceptable method for compliance with 10 C.F.R. § 54.21(c)(1)(iii).127 Accordingly, to challenge the adequacy of an NRC-approved guidance document, an intervenor must provide specificity and substantial support128 to overcome the special weight accorded to a guidance document that has been 122 See NUREG-1801, Rev. 2, at 2 (NYS00147A).

123 See id. Neither NYS nor Riverkeeper, however, submitted comments to the NRC for consideration in NUREG-1801, Rev. 2. See NUREG-1950, Disposition of Public Comments and Technical Bases for Changes in the License Renewal Guidance Documents NUREG-1801 and NUREG-1800, at IV-1 to IV-21 (Apr. 2011)

(ENT000528) (listing public comments on changes to NUREG-1801 and NUREG-1800).

124 See, e.g., Oyster Creek, CLI-08-23, 68 NRC at 468.

125 Indian Point, CLI-15-6, slip op. at 19; Seabrook, CLI-12-5, 75 NRC 314 n.78.

126 See Oyster Creek, CLI-08-23, 68 NRC at 468 (emphasis added); see also Seabrook, CLI-12-05, 75 NRC at 304 (If the NRC concludes that an aging management program (AMP) is consistent with the GALL Report, then it accepts the applicants commitment to implement that AMP, finding the commitment itself to be an adequate demonstration of reasonable assurance under section 54.29(a).).

127 Vt. Yankee, CLI-10-17, 72 NRC at 36.

128 See id. at 33 n.185, 37.

implicitly endorsed by the Commission.129 As demonstrated by Entergys testimony, the State has not done so here.

In light of the foregoing, a finding that an applicants AMP is consistent with NUREG-1801 carries special weight130 and constitutes a finding of reasonable assurance under 10 C.F.R.

§§ 54.21(a), 54.21(c)(1)(iii), and 54.29(a).131

2. Revisions to NUREG-1801

a. NUREG-1801, Revision 1 The IPEC LRA was prepared using the guidance of NUREG-1801, Revision 1. In 2010, more than three years after the LRA was submitted, and more than a year after the NRC Staffs original SER was published, the Staff issued NUREG-1801, Revision 2.132 Subsequently, in 2013, the NRC Staff published interim staff guidance to revise and update NUREG-1801, Rev. 2 based on the NRCs approval of industry guidance on the aging management of RVIs MRP-227-A.133 As discussed further below, the RVI AMP fully meets this most recent guidance.

b. NUREG-1801, Revision 2 and MRP-227 The NRC Staff issued NUREG-1801, Rev. 2 in December 2010.134 As explained in Section V.B. of Entergys Testimony, NUREG-1801, Rev. 2 contained a new AMP (XI.M16A) addressing pressurized water reactor (PWR) RVIs. The new AMP relied on the 129 Seabrook, CLI-12-5, 75 NRC at 314 n.78.

130 Id.

131 Vt. Yankee, CLI-10-17, 72 NRC at 36.

132 See generally NUREG-1801, Rev. 2 (NYS00147A-D).

133 Final License Renewal Interim Staff Guidance LR-ISG-2011-04, Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors at 1 (May 28, 2013) (LR-ISG-2011-04)

(ENT000641).

134 See generally NUREG-1801, Rev. 2 (NYS00147A-D).

implementation of industry guidance from EPRI MRP in MRP-227, Revision 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines.135 As discussed in Sections II.B and C above, in 2010 Entergy submitted its original RVI AMP, and in 2011 submitted its RVI Inspection Plan, both of which were based on MRP-227, Revision 0.

c. Interim Staff Guidance and Endorsement of MRP-227-A As explained further below, in 2011, the NRC Staff issued its safety evaluation (SE) on MRP-227, Revision 0.136 The SE contained specific topical report conditions and applicant/licensee action items (A/LAIs) that were to be addressed by applicants or licensees utilizing the report. MRP-227-A, the NRC-endorsed version of MRP-227, Revision 0, was published in January 2012 to incorporate the Staffs topical report conditions and A/LAIs.137 Thereafter, the NRC Staff issued interim staff guidance, LR-ISG-2011-04, to amended AMP XI.M16A to reflect its endorsement of MRP-227-A.138 As discussed in Section II.C, above, Entergy submitted a revised RVI AMP and Inspection Plan based on the NRC-endorsed MRP-227-A guidelines. Thus, the IPEC RVI AMP meets the intent of the latest Staff guidance on management of the effects of aging on PWR RVIs. Again, NYS and Dr. Lahey do not challenge that fact, but rather challenge the guidance itself. To challenge the adequacy of the underlying guidance, NYS must to overcome the special weight with specificity and substantial support139 for its arguments. As the Commission has held 135 Id. at XI M16A-1.

136 Letter from R. Nelson, NRC, to N. Wilmshurst, EPRI, Revision 1 to the Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Prog[ra]m (MRP) Report 1016596 (MRP-227),

Revision 0, Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines (Dec. 16, 2011)

(SE for MRP-227-A) (ENT000230).

137 See generally MRP-227-A (NRC0014A-F).

138 LR-ISG-2011-04 at 2-3 (ENT000641).

139 See Vt. Yankee, CLI-10-17, 72 NRC at 33 n.185, 37.

in this proceeding, the State must show there are unusual circumstances in this case that would justify setting aside the applicable guidance.140 The State has made no such showing.

C. Burden of Proof At the hearing stage, an intervenor has the initial burden of going forward; that is, it must provide sufficient, probative evidence to establish a prima facie case for the claims made in the admitted contention.141 The mere admission of a contention does not satisfy this burden.142 If the Intervenors do establish a prima facie case on a particular claim, then the burden shifts to Applicant to provide sufficient evidence to rebut the intervenors contention.143 At the admissibility stage, the petitioner has the ironclad obligation to examine the available documentation with sufficient care to support the foundation for a contention.144 This obligation applies with equal, if not greater, force at the hearing stage.145 As will be further explained below, the State and its witness, Dr. Lahey, disregard, rather than dispute, the technical 140 Indian Point, CLI-15-6, slip op. at 21-22.

141 Oyster Creek, CLI-09-07, 69 NRC at 269 (quoting Consumers Power Co. (Midland Plant, Units 1 & 2),

ALAB-123, 6 AEC 331, 345 (1973) (The ultimate burden of proof on the question of whether the permit or license should be issued is . . . upon the applicant. But where . . . one of the other parties contends that, for a specific reason . . . the permit or license should be denied, that party has the burden of going forward with evidence to buttress that contention. Once he has introduced sufficient evidence to establish a prima facie case, the burden then shifts to the applicant who, as part of his overall burden of proof, must provide a sufficient rebuttal to satisfy the Board that it should reject the contention as a basis for denial of the permit or license.) (emphasis in original)); see also Vt. Yankee Nuclear Power Corp. v. Natural Res. Def. Council, 435 U.S. 519, 554 (1978) (upholding this threshold test for intervenor participation in licensing proceedings);

Phila. Elec. Co. (Limerick Generating Station, Units 1 & 2), ALAB-262, 1 NRC 163, 191 (1975) (holding that the intervenors had the burden of introducing evidence to demonstrate that the basis for their contention was more than theoretical).

142 See Oyster Creek, CLI-09-07, 69 NRC at 268-70.

143 See, e.g., id. at 269; La. Power & Light Co. (Waterford Steam Electric Station, Unit 3), ALAB-732, 17 NRC 1076, 1093 (1983) (citing Midland, ALAB-123, 6 AEC at 345); see also 10 C.F.R. § 2.325.

144 See Duke Power Co. (Catawba Nuclear Station, Units 1 & 2), ALAB-687, 16 NRC 460, 468 (1982), vacated in part on other grounds, CLI-83-19, 17 NRC 1041 (1983).

145 See Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 & 3), LBP-13-13, 78 NRC 246, 301 n.308 (2013) (rejecting an experts claims based on some averages and a gut feeling, rather than a thorough a review of available documentation).

basis for the RVI AMP. The State, therefore, has failed to meet its burden of going forward with evidence to support NYS-25. Considering the States and Dr. Laheys disregard of the substantial available documentation in direct testimony, Entergy reserves its right to object and to seek to strike any new critique of these studies that Dr. Lahey or NYS may offer in rebuttal, or, in the alternative, to seek to file sur-rebuttal testimony.

To prevail, the Applicants position must be supported by a preponderance of the evidence.146 IV. ENTERGYS WITNESSES Entergys testimony on NYS-25 is sponsored by the witnesses identified below. The testimony, opinions, and evidence presented by these witnesses are based on their substantial technical and regulatory expertise, professional experience, and personal knowledge of the issues raised in NYS-25. Collectively, these witnesses will demonstrate that NYS-25 lacks merit.

A. Mr. Nelson F. Azevedo Nelson Azevedos professional and educational qualifications are summarized in his curriculum vitae147 and in Section I.A of Entergys testimony. Mr. Azevedo is employed by Entergy as the Supervisor of Code Programs at IPEC. He holds a Bachelor of Science degree in Mechanical and Materials Engineering from the University of Connecticut, and a Master of Science in Mechanical Engineering and Master of Business Administration (M.B.A.) degrees from the Rensselaer Polytechnic Institute (RPI) in Troy, New York. Mr. Azevedo has more than 30 years of professional experience in the nuclear power industry. In his current position, he oversees the IPEC engineering section responsible for implementing American Society of 146 See Pac. Gas & Elec. Co., ALAB-763, 19 NRC at 577; Oyster Creek, CLI-09-07, 69 NRC at 263.

147 See Curriculum Vitae for Nelson F. Azevedo (ENT000032).

Mechanical Engineers (ASME) Code programs, including the fatigue monitoring, inservice inspection, inservice testing, boric acid corrosion control, non-destructive examination, steam generators, alloy 600 cracking, RPV embrittlement, and RVI programs. In addition to those duties he is responsible for ensuring compliance with the ASME Code,Section XI requirements for repair and replacement activities at IPEC and represents IPEC before industry organizations, including the PWR Owners Group Management Committee. Accordingly, Mr. Azevedo is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on the Entergy RVI AMP, and Entergys aging management activities and TLAAs for RPVs.

B. Mr. Robert J. Dolansky Bob Dolanskys professional and educational qualifications are summarized in his curriculum vitae148 and in Section I.B of Entergys testimony. Mr. Dolansky is employed by Entergy as a Code Programs Engineer at IPEC. He holds a Bachelor of Science degree in Aeronautical Engineering from RPI in Troy, New York. Mr. Dolansky has more than 25 years of professional experience as an ASME Code Programs Engineer at IPEC. He has been the program owner for, among other programs, the RVI, inservice inspection (ISI), inservice testing, steam generator, and alloy 600 cracking programs. In his current position, he is the program owner of the IPEC RVI AMPs for both units. He is also a member of the PWR Owners Group materials subcommittee.

Accordingly, Mr. Dolansky is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on the Entergy RVI AMP.

148 See Curriculum Vitae for Robert J. Dolansky (ENT000522).

C. Mr. Alan B. Cox Alan Coxs professional and educational qualifications are summarized in his curriculum vitae149 and in Section I.C of Entergys testimony. In brief, he holds a Bachelor of Science degree in Nuclear Engineering from the University of Oklahoma and a Master of Business Administration (M.B.A.) from the University of Arkansas at Little Rock. Prior to his retirement from Entergy in 2015, he was the Technical Manager of License Renewal. Presently, he continues to work with Entergy as an independent consultant. Mr. Cox has more than 37 years of experience in the nuclear power industry, having served in various positions related to engineering and operations of nuclear power plants, including several years as a licensed reactor operator and a senior reactor operator. From 2001 to 2015, he worked full-time on license renewal matters, supporting the integrated plant assessment and LRA development for Entergy license renewal projects, as well as projects for other utilities.

Mr. Cox was directly involved in preparing the LRA and developing or reviewing AMP descriptions for IP2 and IP3, including the IPEC RVI AMPs. He has also been directly involved in developing or reviewing Entergy responses to NRC Staff RAIs concerning the LRA and necessary amendments or revisions to the application. Accordingly, he has extensive knowledge of IPEC aging management activities, including the descriptions in the LRA and other related documentation discussed below. Thus, Mr. Cox is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on the Entergy RVI AMP, and Entergys aging management activities and TLAAs for RPVs.

149 See Curriculum Vitae for Alan B. Cox (ENTR00031).

D. Mr. Jack R. Strosnider, Jr.

Jack Strosniders professional and educational qualifications are summarized in his curriculum vitae150 and in Section I.D of Entergys testimony. Mr. Strosnider holds a Bachelor of Science degree and a Master of Science degree, both in Engineering Mechanics from the University of Missouri at Rolla, and an M.B.A. degree from the University of Maryland. Mr.

Strosnider is a Senior Nuclear Safety Consultant with Talisman International, LLC. Prior to April 2007, he was employed for 31 years by the NRC. During that time, he held numerous senior management positions at the NRC, including Director of the Office of Nuclear Material Safety and Safeguards, Deputy Director of the Office of Nuclear Regulatory Research, and Director of the Division of Engineering in the Office of Nuclear Reactor Regulation (NRR).

On technical matters, he was, for example, involved in the development of the technical bases for 10 C.F.R. § 50.61, which provides fracture toughness requirements for protection against pressurized thermal shock (PTS) events, and was responsible for licensing reviews associated with the integrity of the RPV and monitoring of RVIs.

Mr. Strosnider has extensive experience in developing and applying NRC regulations and programs addressing the aging of nuclear power plant structures and components. He has directed engineering reviews and the preparation of SERs for license renewal. With respect to aging effects on RPVs, Mr. Strosnider was involved in the development of the technical bases for the requirements in 10 C.F.R. § 50.61. Thus, Mr. Strosnider is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on the NRC regulatory requirements relating to RVI AMPs, aging management activities, and TLAAs for RPVs, and criteria necessary to satisfy those requirements.

150 See Curriculum Vitae for Jack R. Strosnider, Jr. (ENTR00184).

E. Mr. Timothy J. Griesbach Tim Griesbachs professional and educational qualifications are summarized in his curriculum vitae151 and in Section I.E of Entergys testimony. In brief, he holds Bachelor of Science and Master of Science degrees in Metallurgy and Materials Science from Case Western Reserve University. Currently, he is a Senior Associate at Structural Integrity Associates, Inc.

Mr. Griesbach has more than 40 years of experience in metallurgy and materials engineering, primarily in the nuclear field.

He is a member of the American Nuclear Society and the American Society of Mechanical Engineers (ASME), where he has served on various ASME Boiler and Pressure Vessel Code committees for over 33 years, chairs the ASME Section XI Working Group on Operating Plant Criteria, and is currently a member of the ASME Section XI Standards Committee. He has worked closely with the EPRI Materials Reliability Program to develop and implement the MRP-227 inspection and evaluation guidelines for the safety and long-term operation of PWR vessel internals. Thus, Mr. Griesbach is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on the Entergy RVI AMP, and Entergys aging management activities and TLAAs for RPVs.

F. Dr. Randy G. Lott Dr. Randy Lotts professional and educational qualifications are summarized in his curriculum vitae152 and in Section I.F of Entergys testimony. Dr. Lott holds a Bachelor of Science in Engineering degree in nuclear engineering from the University of Michigan, and Master of Science and Doctor of Philosophy degrees in nuclear engineering from the University 151 See Curriculum Vitae for Timothy J. Griesbach (ENT000617).

152 See Curriculum Vitae for Randy G. Lott (ENT000618).

of Wisconsin. Currently, he is a Consulting Engineer at Westinghouse and has more than 35 years of experience in nuclear materials and radiation effects.

Dr. Lott has extensive experience with post-irradiation evaluation of reactor components, and has been directly involved in the design and implementation of aging management programs for reactor internals. He has supervised testing of RPV surveillance capsules and conducted research programs on irradiation embrittlement and annealing of RPV steels, and he has conducted numerous test programs on highly irradiated stainless steels, including measurement of tensile, fracture toughness and irradiation-assisted stress corrosion cracking (IASCC) properties. As a member of the MRP Reactor Internals Inspection and Evaluation Guidelines Core Group, he was a contributor to the U.S. industry Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227). Specifically, he worked on aging management strategies for the Westinghouse and Combustion Engineering plants to provide the basis for the RVI inspection guidelines in MRP-227. Thus, Dr. Lott is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on the Entergy RVI AMP.

G. Mr. Mark A. Gray Mark Grays professional and educational qualifications are summarized in his curriculum vitae153 and in Section I.G of Entergys testimony. Mr. Gray is a Principal Engineer in the Primary Systems Design and Repair group at Westinghouse. He holds Master of Science and Bachelor of Science degrees in Mechanical Engineering from the University of Pittsburgh and has over 34 years of experience in the nuclear power industry. His principal work activities include the evaluation of the structural integrity of primary system piping and components, 153 See Curriculum Vitae for Mark A. Gray (ENTR00186).

including the development of plant life extension and monitoring programs and analysis. He participated in the development and application of transient and fatigue monitoring algorithms and software for the WESTEMS' Transient and Fatigue Monitoring System, and collaborated with vendors outside Westinghouse in the development of transient and fatigue monitoring systems.

During the preparation of the EAF analyses for IPEC license renewal, Mr. Gray provided general technical direction for the engineers performing the EAF analyses, and either co-authored or reviewed the resulting Westinghouse environmental fatigue reports, referred to as WCAP reports. For these reasons, Mr. Gray is qualified through knowledge, skill, directly-relevant experience, training, and education to provide expert witness testimony on fatigue analysis of RVIs.

V. ENTERGYS EVIDENCE AND ARGUMENTS In their testimony, Entergys experts explain why Entergys IPEC RVI AMPtogether with substantial supporting informationprovides reasonable assurance that the effects of aging will be adequately managed throughout the PEO as required by 10 C.F.R. §§ 54.21(a)(3),

54.21(c)(1)(iii), and 54.29(a). In so doing, Entergys experts refute the States and Dr. Laheys assertions point-by-point, thereby demonstrating that the issues raised in NYS-25 lack merit from regulatory and technical perspectives.

A. Technical Background on the Aging Management of RVIs and RPVs In Section V of Entergys testimony, Entergys expert witnesses describe the layout and functions of the RVIs and RPVs, the scope of components covered by the RVI AMP, the materials used in the RVIs and RPVs at IPEC, and the design basis loads those materials are subjected to.

Entergys witnesses first explain that the RPV contains the reactor core and RVIs, and is a key part of the reactor coolant pressure boundary.154 The RVIs, located inside the RPV, direct the coolant flow, support the reactor core, and guide the control rods, but do not form part of the reactor coolant pressure boundary.155 The RVI AMP provides a complete and correct list of the PWR RVI sub-assemblies and components at IPEC.156 Contrary to Dr. Laheys belief,157 the RVIs do not include control rods.

Further, the control rods are active components that perform their intended function with moving parts or a change in configuration.158 They also are consumables, subject to replacement based on a qualified life or specified time period under 10 C.F.R. § 54.21(a)(1).159 They are therefore excluded from AMR pursuant to Part 54.160 Dr. Lahey also incorrectly asserts that Entergy has not addressed other control rod-related components,161 but they are in fact included in AMR and the RVI AMP.162 And while Dr. Lahey raises concerns about the control rod stub tube welds or J-groove welds, and about RPV head penetrations, the effects of aging on those components are managed under the Reactor Vessel Head Penetration Inspection AMP, not the RVI AMP.163 Accordingly, several of Dr.

Laheys complaints are at odds with accepted industry definitions of RVIs and Part 54, or are 154 See Entergys Testimony at A90 (ENT000616).

155 See id. at A94.

156 See id. at A98 (citing NL-12-037 (NYS000496)).

157 See Revised Lahey Testimony at 13 (NYS000482).

158 See Indian Point, CLI-15-6, 81 NRC at __, slip op. at 8.

159 See Entergys Testimony at A99 (ENT000616); see also 10 C.F.R. § 54.21(a)(1).

160 See Entergys Testimony at A99 (ENT000616).

161 See Revised Lahey Testimony at 12-13 (incorrectly asserting that the guide tubes, plates, pins, and welds associated with the control rods are omitted from the RVI AMP) (NYS000482).

162 See Entergys Testimony at A100 (ENT000616).

163 See id. at A101.

adequately addressed by other aging management programs that he either ignored or chose not to review.

With respect to materials, Entergys witnesses explain that, contrary to Dr. Laheys testimony, the materials used in the IPEC RVIs and RPVs are fundamentally different, and have very different mechanical properties and behavior under irradiation.164 Therefore, many of his arguments and assertions regarding the behavior of RVI materials under irradiation and the potential for RVIs to undergo a transition from ductile to brittle behavior and are both incorrect and unsupported. Specifically, IP2 and IP3 RPVs are constructed primarily of low-alloy (carbon) steel, with stainless steel cladding, while the RVIs are made of wrought austenitic stainless steel, other stainless steels including Cast Austenitic Stainless Steel (CASS), or nickel-based alloys.165 As a result, the IPEC RVI materials exhibit less temperature-dependent changes in unirradiated mechanical properties than the RPV materials.166 Also, the mechanical properties of the IPEC RVI beltline materials, including the cast austenitic stainless steel (CASS) lower support column caps (LSCCs), do not change with irradiation to the same extent as low-alloy RPV materials do; i.e., they do not exhibit a shift in the ductile-to-brittle transition temperature.167 Overall, the RVI materials are far less susceptible to irradiation effects than the RPV.168 164 See id. § V.C.

165 See id. at A104.

166 See id.

167 See id. at A107.

168 See id. at A117.

Dr. Lahey also asserts that RVIs could be subject to pressure and/or thermal shock loads.169 It is not entirely clear what Dr. Laheys concerns are, however, to the extent his concern is that RVI components could fail due to PTS, it lacks basis.170 The RVIs have no pressure retaining function.171 For this reason, a PTS transient does not subject the RVI components to the stresses characteristic of the effects of a PTS event on an RPV.172 The design basis transients and loads on the RVIs are defined in the CLB for IPEC, and identified on a plant-specific basis in Chapter 4 of the Updated Final Safety Analysis Reports (UFSAR) for IP2 and IP3.173 As fully explained in Entergys testimony, the RVI AMP considers the full range of design basis loads, which are established in accordance with the CLB.174 B. Regulatory Guidance Addressing Aging Management of RVIs In Section VI, Entergys expert witnesses summarize Entergys full compliance with the specific regulatory guidance addressing the management of the effects of aging on RVIs. In particular, MRP-227-A is the current, NRC-approved version of EPRIs guidance on the aging management of RVIs.175 The NRC Staff thoroughly reviewed this guidance, approved it in a safety evaluation,176 and issued interim staff guidance updating the NUREG-1801, Revision 2 169 Revised Lahey Testimony at 16 (NYS000482).

170 See Entergys Testimony § V.C.3 (ENT000616).

171 See id. at A114.

172 See id.

173 See id. at A115.

174 See id. § V.C.3.

175 See id. §§ VI.B-C.

176 See id. § VI.C; see also NRC Staff Safety Evaluation for MRP-227-A (Letter from R. Nelson, NRC, to N.

Wilmshurst, EPRI, Revision 1 to the Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Prog[ra]m (MRP) Report 1016596 (MRP-227), Revision 0, Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines (Dec. 16, 2011) (SE for MRP-227-A) (ENT000230).

AMP (XI.M16A) to incorporate MRP-227-A.177 While Dr. Lahey believe[s] that the NRC-approved industry guidance is inadequate,178 Entergys witnesses show that he has largely disregarded and failed to challenge the substantial technical basis supporting that guidance.179 MRP-227-A is the result of a decade-long systematic evaluation of the effects of aging on RVIs.180 MRP-227-A was developed in four steps: (1) development of screening criteria for the applicable aging mechanisms; (2) screening of RVI components based on susceptibility to degradation; (3) functionality analysis and failure modes, effects, and criticality analyses (FMECA), which resulted in the binning of components into different risk severity and inspection categories; and (4) development of the inspection and evaluation guidelines and flaw evaluation methodology.181 The screening process explicitly considered potential combinations of aging effects, including all of the effects mentioned by Dr. Lahey.182 The aging management guidelines in MRP-227-A are supported by numerous underlying EPRI MRP technical studies, covering topics from aging degradation mechanisms and resulting effects, categorization of components, aging management strategies, acceptance criteria, and other topics.183 These technical studies document the considerable body of operating experience, state-of-the art research, and laboratory experiments that underpin the MRP-227-A guidelines.184 177 See Entergys Testimony § VI.A (ENT000616); see also LR-ISG-2011-04 (NYS000524).

178 See Revised Lahey Testimony at 39 (NYS000482).

179 See Entergys Testimony § VII (ENT000616).

180 See id. § VI.B.

181 See id. at A124.

182 See id. at A125.

183 See id. at A126.

184 See id.

The principal documents, along with MRP-227-A, total over 1600 pages of research and analysis and include:

• MRP-232: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals;185

• MRP-230: Functionality Analysis for Westinghouse and Combustion Engineering Representative PWR Internals;186

• MRP-228 Inspection Standard for PWR Internals;187

• MRP-210: Fracture Toughness Evaluation of Highly Irradiated PWR Stainless Steel Internal Components;188

• MRP-191: Screening, Categorization and Ranking of Reactor Internals of Westinghouse and Combustion Engineering PWR Designs;189

• MRP-175: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values;190

• MRP-134: Framework and Strategies for Managing Aging Effects in PWR Internals;191 and

• WCAP-17096-NP, Rev. 2, Reactor Internals Acceptance Criteria Methodology and Data Requirements.192 185 See MRP-232, EPRI Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals (Dec. 2008) (MRP-232) (ENT000642A-C); see also MRP-232, Revision 1, EPRI Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals (Dec. 2012) (MRP-232, Rev. 1) (ENT000643).

186 See MRP-230, EPRI Materials Reliability Program: Functionality Analysis for Westinghouse and Combustion Engineering Representative PWR Internals (Oct. 2009) (MRP-230) (ENT000644).

187 See MRP-228, EPRI Materials Reliability Program: Inspection Standard for PWR Internals (July 2009)

(NYS000323); see also MRP-228, Rev. 1, EPRI Materials Reliability Program: Inspection Standard for PWR Internals (Dec. 2012) (ENT000645).

188 See MRP-210, EPRI Materials Reliability Program: Fracture Toughness Evaluation of Highly Irradiated PWR Stainless Steel Internal Components (Dec. 2007) (MRP-210) (ENT000646).

189 See MRP-191, EPRI Materials Reliability Program: Screening, Categorization and Ranking of Reactor Internals of Westinghouse and Combustion Engineering PWR Designs (NYS000321).

190 See MRP-175 EPRI Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (ENT000631).

191 See MRP-134, EPRI Materials Reliability Program: Framework and Strategies for Managing Aging Effects in PWR Internals (June 2005) (ENT000647).

Based on these supporting reports, MRP-227-A provides comprehensive aging management guidelines, detailing inspections to detect the effects of aging (individually or in combination), methods to evaluate such aging effects, and considerations for repair or replacement of degraded components.193 MRP-227-A also defines risk-prioritized inspections to detect the effects of aging, and recommends methods to evaluate aging effects.194 Dr. Lahey generally disregards all of these analyses, and explains no disagreements with any information in them, despite the fact that they have nearly all been available to the State for several years through the mandatory disclosure process.195 This approach does not meet the States burden of moving forward with providing sufficient probative evidence to support its contention at the hearing stage.196 Instead, the State offers baseless legalistic justifications in an apparent attempt to excuse Dr. Lahey from the obligation to actually review the technical basis for the IPEC RVI AMP.

First, the State asserts that Entergy can only rely on its own RVI AMP documentation:

The RPVI AMP that is currently before the Board for review consists of the Revised and Amended RVI Plan, developed between 2012 and 2014 and approved by NRC Staff in the SSER2.

Thus, the adequacy of the AMP for RPVIs must stand or fall on the adequacy of these documents.197 192 See WCAP-17096-NP, Rev. 2, Reactor Internals Acceptance Criteria Methodology and Data Requirements (Dec. 2009) (ENT000635).

193 See Entergys Testimony at A125 (ENT000616).

194 See id.

195 See id. at A127. Notably, Entergy has disclosed all of these documents to the State under the mandatory disclosure process in 10 C.F.R. § 2.336and nearly all of them several years ago.

196 See Oyster Creek, CLI-09-07, 69 NRC at 269.

197 NYS Revised SOP at 19 (NYS000481).

Similarly, the State claims that Entergy has failed to submit an analysis of the allegedly synergistic effects of embrittlement, fatigue, and stress corrosion on RVIs.198 But these objections are groundless, as nothing in the NRCs regulations prevents Entergy from relying on an NRC-approved topical report as the basis for its RVI AMP, or requires Entergy to submit to the NRC Staff in this proceeding the MRP reports reviewed on a generic basis by the Staff. On the contrary, the Commission has endorsed and encouraged the practice of using generic guidance to improve efficiency, as has the Board in this proceeding.199 Overall, the States weak attempt to avoid its experts obligation to review the available technical basis for the RVI AMP fails.

Throughout Entergys Testimony, the witnesses explain that the IPEC RVI AMP is consistent with MRP-227-Aa point the State does not dispute. To challenge the adequacy of an NRC-approved guidance document, an intervenor must provide specificity and substantial support for such a challenge, in order to show unusual circumstances200 are present to overcome the special weight it is accorded in NRC proceedings.201 The State has not done so.

C. Entergys LRA Effectively Addresses Aging Management of RVIs and RPVs In Section VII, Entergys expert witnesses show that the IPEC RVI AMP provides reasonable assurance that the RVI components will continue to perform their intended functions, 198 Id. at 23 ¶ 10.

199 See Vt. Yankee, CLI-10-17, 72 NRC at 19 (noting that the GALL Report may be referenced in a license renewal application in the same manner as an approved topical report); Indian Point, CLI-15-6, slip op. at 21-22 (requiring unusual circumstances to be present to justify setting Staff guidance aside); Indian Point, LBP-13-13, 78 NRC at 297 (allowing Entergy to rely on the guidance in NSAC-202L as the basis for its flow-accelerated corrosion (FAC) AMP).

200 Indian Point, CLI-15-6, slip op. at 21.

201 Vt. Yankee, CLI-10-17, 72 NRC at 32-33, n.185.

consistent with the CLB, during the PEO, as required by 10 C.F.R. §§ 54.21(a)(3), (c)(1)(iii), and 54.29(a).

1. The IPEC RVI AMP

a. Overview of the RVI AMP and Inspection Plan Consistent with the guidance in MRP-227-A, the IPEC RVI AMP is divided into three main areas: (1) examinations and other inspections, along with a comparison of data to examination acceptance criteria, as defined in MRP-227-A and MRP-228; (2) process for resolution of indications that exceed examination acceptance criteria by entering them into the applicants Corrective Action Program; and (3) monitoring and control of reactor primary coolant water chemistry based on industry guidelines.202 The RVI Inspection Plan provides additional details on the inspections to be conducted under the RVI AMP, including: (1) the type of examinations; (2) the level of examination qualification; (3) the schedule of initial inspection and frequency of subsequent inspections; (4) the criteria for sampling and coverage; (5) the criteria for expansion of scope if unanticipated indications are found; (6) the acceptance criteria; (6) the methods for evaluation of examination results that do not meet the acceptance criteria; (7) provisions to update the program based on industry-wide results; and (8) contingency measures to repair, replace, or mitigate, beyond the information set forth in the RVI AMP.203

b. The RVI AMP Describes Inspections in Detail Dr. Lahey alleges that Entergy has not provided sufficient details about its inspection schedule.204 The IPEC RVI AMP, however, provides a comprehensive inspection schedule 202 See Entergys Testimony at A139 (ENT000616) (citing NL-12-037, Attach. 1 (NYS000496)).

203 See Entergys Testimony at A137 (ENT000616).

204 See Revised Lahey Testimony at 48-49 (NYS000482).

based on the guidance in MRP-227-A.205 RVI components are separated into four groups with aging management strategies specified for each group (Primary, Expansion, Existing Programs, and No Additional Measures) depending on: (1) the relative susceptibility to and tolerance of applicable aging effects; and (2) the existence of other programs that manage the effects of aging on those components.206 The inspections are specified in Table 5-2 (primary components), Table 5-3 (expansion components), and Table 5-4 (existing program components) of the Inspection Plan.207 Importantly, the inspection categorization is not dependent on analyzing the behavior of the individual components under accident loads.208 Rather, the EPRI MRP evaluated possible component failure under accident loads, and if the assumed failure could impact a design basis function the component was assigned to an inspection category using the appropriate inspection techniques and frequency of inspections.209 Dr. Lahey provides no critique of EPRIs categorization methodologyindeed, he does not mention it at all. Therefore, contrary to Dr.

Laheys bare assertions, the inspection schedule is comprehensive and adequate.210

c. The RVI AMP Manages the Effects of Aging on RVIs Regardless of the Underlying Aging Mechanism Dr. Lahey asserts that Entergy has considered various aging mechanisms that could affect the RVIs in silos, without considering the synergistic interactions between mechanisms.211 As a threshold matter, this position is directly contrary to the approach the Commission specified 205 See Entergys Testimony at A139 (ENT000616).

206 See id. at A138.

207 See id. at A139 (citing NL-12-037, Attach. 2, at 37-51 (NYS000496)).

208 See id. at A137.

209 See id.

210 Revised Lahey Testimony at 48-49 (NYS000482).

211 See id. at 14-15.

when it promulgated the license renewal rules in 10 C.F.R. Part 54. As described in Sections VII.A.3 and 5 of Entergys Testimony, the NRCs license renewal process has long focused on aging effects, rather than aging mechanisms.212 Since 1995, when the NRC promulgated its revised license renewal rules, the NRC has emphasized that the identification of individual aging mechanisms is not required as part of the license renewal review.213 Instead, the regulations in 10 C.F.R. Part 54 concentrate on ensuring that important structures, systems and components (SSCs) will continue to perform their intended functions during the PEO regardless of the particular aging mechanism.214 Consistent with this principle, the inspections conducted under the RVI AMP will look for evidence of any of the aging effects of concern, and appropriate action is taken if any relevant conditions related to those effects are discovered, regardless of their cause.215 In any event, to the extent Dr. Laheys claim is that the combined effects of multiple aging mechanisms are not addressed in the RVI AMP, he is mistaken. As specified in MRP-227-A, Section 3.2, the RVI AMP addresses the following eight age-related degradation mechanisms and their associated effects, each of which are further described in MRP-227-A:

• Stress corrosion cracking (SCC);

• IASCC;

• Wear;

• Fatigue;

• Thermal aging embrittlement;

• Irradiation embrittlement (also referred to as neutron embrittlement);

• Void swelling and irradiation growth; and 212 See Entergys Testimony §§ VII.A.3, 5 (ENT000616).

213 See id. at A143; see also Nuclear Power Plant License Renewal; Revisions, 60 Fed. Reg. 22,461, 22,463 (May 8, 1995) (Part 54 SOC) (NYS000016).

214 See Entergys Testimony at A143 (ENT000616); see also Part 54 SOC, 60 Fed. Reg. at 22,463 (NYS000016).

215 See Entergys Testimony at A143 (ENT000616).

• Thermal and irradiation-enhanced stress relaxation or irradiation-enhanced creep.216 For each of the eight mechanisms, MRP-227-A identifies the resulting aging effect, which will then be managed through inspections under the MRP-227-A guidelines.217 Notably, in most cases, the key effects are cracking, dimensional changes, or wear, but in all cases, as explained below, the inspections specified in MRP-227-A are designed to detect potential aging effects applicable to each RVI component, regardless of the underlying mechanism.218 Therefore, contrary to Dr. Laheys claims, the IPEC RVI AMP does not fail[] to consider how those interacting degradation mechanisms will impact the . . . RPV internals.219 With respect to the effects of embrittlement, no recommendations for inspection to determine embrittlement level are contained in the guidance because these mechanisms cannot be directly observed.220 But, as Entergys witnesses show, embrittlement is only an issue for RVI components if there is a crack.221 Therefore, while it is not possible to detect the level of embrittlement directly through visual inspection, MRP-227-A provides for inspections that detect the manifestation of significant thermal aging or neutron-irradiation embrittlement specifically, the potential growth of a pre-existing defect.222 Once a defect is discovered, its ability to withstand fatigue and combinations of both normal and accident shock loads is evaluated by either fracture mechanics analysis or a structural analysis (an engineering evaluation) using the lower bound fracture toughness; i.e., the 216 See id. at A144 (citing NL-12-037, Attach. 1 at 6 (NYS000496)).

217 See Entergys Testimony at A144 (ENT000616).

218 See id.

219 Revised Lahey Testimony at 49 (NYS000482).

220 See Entergys Testimony at A146 (ENT000616).

221 See id.

222 See id.

evaluation assumes the maximum level of embrittlement of the material.223 Thus, the program has compensated for any inability to directly determine the level of embrittlement through a conservative assumption employed during evaluation of inspection findings.224 Thus, reasonable assurance that the effects of aging will be adequately managed is provided without the need for direct observation or analysis of the level of embrittlement.225

d. The RVIs Are Robust and Highly Failure Tolerant Next, Entergys witnesses show that the RVI materials are constructed of damage-resistant and flaw-tolerant materials that have performed well in service at many plants for thousands of reactor years, with very little adverse operating experience.226 For example, ASME Code Section XI periodic inspections for PWR RVIs to date have included inspections of baffle-former bolts at several plants.227 Although these bolts are leading indicators for Westinghouse RVIs for the combination of irradiation-induced stress relaxation, void swelling, and IASCC, very few cracked or failed baffle-former bolts have been detected during these examinations and, in most cases, no cracked or failed bolts were detected at all.228

.229 223 See id. at A146 (citing MRP-227-A at 6-4 (NRC000114B)).

224 See Entergys Testimony at A146 (ENT000616).

225 See id.

226 See id. at A148 (citing MRP-227-A, App. A (NRC000114C)).

227 See Entergys Testimony A148, A150 (ENT000616).

228 See id. at A150.

229 See id. at A156.

Dr. Lahey disregards the overall operating experience when he labels Entergys plans regarding the baffle-former bolts as a wait-and-see approach.230 Entergys plans to inspect the baffle-former bolts are adequately and sufficiently specified in the record to provide the requisite reasonable assurance that the effects of aging on baffle-former bolts will be adequately managed.231 Specifically, at IPEC, Entergy is appropriately planning to inspect 100% of the baffle former bolts at IP2 in Spring 2016 and at IP3 in Spring 2019, with subsequent examinations on ten-year intervals.232 In preparation for these inspections, Entergy is preparing a technical justification (TJ) which will demonstrate that the ultrasonic testing (UT) inspections at IPEC will be capable of detecting defects exceeding 30% of the bolt cross-sectional area, as specified in Westinghouses evaluations of the baffle-former assembly.233 In addition, Entergy has contracted with Westinghouse to perform a more realistic plant-specific minimum bolting pattern analysis for IPEC.234 This evaluation will consider design basis loads for IP2 and IP3, including the dynamic effects and blowdown loads from pipe breaks of various sizes, low cycle thermal fatigue loads, high cycle flow induced vibration loads, and seismic loads.235 If inspections reveal degradation in baffle-former bolts, then this minimum bolting pattern will be used as the basis for engineering evaluations to determine the acceptability of the bolts following the required UT examinations from MRP-227-A.236 230 Revised Lahey Testimony at 55-56 (NYS000482).

231 See id. § VII.A.4.b.

232 See id. at A152.

233 See id. at A154.

234 See id. at A158.

235 See id.

236 See id. at A158.

Considering the operating experience, Dr. Laheys demand for wholesale replacement of the clevis insert bolts is likewise baseless.237 The clevis insert bolts are treated as an Existing Program component under the RVI AMP, because they are periodically inspected once every ten year interval under the ASME Code,Section XI program per Table IWB-2500.238 Entergy last inspected the clevis bolt inserts at IP2 in 2006 and at IP3 in 2009.239 Entergy has further evaluated recent operating experience regarding clevis insert bolts, and demonstrated that the existing ASME Code inspections are adequate at IPEC.240 Therefore, Entergys planned inspections of clevis insert bolts provide reasonable assurance that the effects of aging will be adequately managed.241

e. The RVI AMP Addresses Combinations of Aging Effects from Multiple Degradation Mechanisms Entergys experts further explain that MRP-227-A guidelines and the IPEC RVI AMP properly address applicable aging effects, including combinations of effects.242 Specifically, during the development of MRP-227-A, the EPRI MRP experts developed a set of standard screening criteria that were used to identify components with one or more potential aging mechanisms and how those effects could combine to affect functionality.243 That work, documented in MRP-175, identified thresholds for aging effects which were then used to develop the screening and categorization results documented in MRP-191 (NYS000321). These results, 237 Revised Lahey Testimony at 56-57 (NYS000482).

238 See Entergys Testimony at A163 (ENT000616).

239 See id.

240 See id. at A164.

241 See id. § VII.A.4.c.

242 See id. §§ VII.A.3, 5.

243 See id. at A168.

in turn, provide the technical basis for the functionality analysis in MRP-230 (ENT000644) and ultimately the examinations specified in MRP-227-A.244 Dr. Lahey raises no dispute with this information. Instead, he incorrectly assumes that his concerns regarding synergistic aging effects have never been addressed,245 and remarkably states he has discovered this important new issue.246 Entergys witnesses show that the RVI AMPand the license renewal process in generalappropriately consider the combined effects resulting from multiple aging mechanisms that could impact the IPEC RVIs.247 In short, Dr.

Lahey does not dispute how EPRI addressed his over-arching concern.248 This misconception is a fundamental reason why NYS-25 lacks merit.

Instead of discussing the substantial work of the EPRI MRP, Dr. Lahey claims that the Department of Energy (DOE), the NRC, and national laboratories have recently embarked on an ambitious R&D program to understand and resolve his concerns regarding the synergistic aging effects on nuclear plant components.249 But Dr. Lahey is mistaken, as this program is intended to address the long-term challenges and research needs for operating nuclear plants beyond 60 years, not beyond 40 years.250 Therefore, Dr. Laheys purported evidence falls short of identifying any deficiency regarding the PEO for IP2 and IP3 at issue here. In any event, the MRP-227 inspection and evaluation guidelines were based on state-of-the art engineering, and designed to accommodate the uncertainties associated with areas where research remains 244 See id.

245 See id. at A169.

246 Revised Lahey Testimony at 78 (NYS000482).

247 See Entergys Testimony §§ VII.A.3, 5 (ENT000616).

248 Revised Lahey Testimony at 14 (NYS000482).

249 See, e.g., id. at 17 (citing DOE, Light Water Sustainability Program, Material Aging and Degradation Technical Program Plan (Aug. 2014) (MAaD Program Plan) (NYS000485)).

250 See Entergys Testimony at A171 (ENT000616).

ongoing.251 Ultimately, the fact that certain research is ongoing is not an indication of any deficiency in an AMP.252 In fact, it is a sign of a healthy, constantly-improving program.253 What Dr. Lahey and the State are proposingto prevent an applicant from using a state-of-the art AMP because some research related to it remains ongoingwould transform the license renewal process into an open-ended research project, which the Commission explicitly intended to avoid when it promulgated 10 C.F.R. Part 54.254 Another misconception from Dr. Lahey is that he broadly implies that the synergy between combined aging effects may have a greater (i.e., worsening) effect than the sum of the individual mechanisms alone.255 This overlooks the fact that a combination of aging effects may in some cases have less of an effect, or even an improvement, in the materials resistance to aging.256 Fatigue and irradiation embrittlement, for example, do not interact synergistically, and in some cases irradiation can increase the fatigue life of RVI materials.257 The RVI AMP accounts for this recognized complexity, but Dr. Lahey does not.258 Dr. Lahey raises concerns about the potential combined effects of thermal and irradiation embrittlement on the CASS LSCCs.259 Entergys witnesses do not dispute that that there is ongoing research on this topic, but this is precisely why the NRC Staff identified the need for 251 See id.

252 See id.

253 See id.

254 See Part 54 SOC, 60 Fed. Reg. at 22,469.

255 See Revised Lahey Testimony at 17 (NYS000482).

256 See Entergys Testimony at A173 (ENT000616).

257 See id.

258 See id. at A174.

259 See Revised Lahey Testimony at 18 & 20 (NYS000482).

further evaluation of CASS components in its Safety Evaluation for MRP-227-A.260 Entergy, in response, demonstrated to the NRC Staff that the potential combined effects of thermal and irradiation embrittlement of CASS components is not an issue for the specific materials used at IPEC because the LSCCs do not have a high percentage of delta ferrite.261 Dr. Lahey does not mention or dispute this information. As a result, he has failed to carry the States burden of moving forward with evidence to support this contention at hearing.

f. The RVI AMP Addresses Appropriate Design Basis Loads, Including Seismic and LOCA Loads Dr. Lahey asserts that the RVI AMP has not adequately addressed potential shock loads on RVI components, but it is not entirely clear what Dr. Lahey means by the term shock loads.262 If the concern is with loads caused by postulated events that are greater than or different from those specified in the CLB for IP2 and IP3, or with scenarios that are beyond the plants licensing bases,263 then there is no requirement to address such loads in the RVI AMP.264 As previously noted, the adequacy of the CLB itself is not open to challenge in this proceeding.265 In particular, the State argues in its Revised SOP that the potential seismic hazard curves for the Indian Point site are higher than the seismic spectra developed in the 1970s during the proceedings concerning the initial operating licenses.266 To the extent the State is arguing that the seismic hazards considered in the CLB for IP2 and IP3 should be reconsidered in 260 See Entergys Testimony at A175 (ENT000616).

261 See id. at A176; NL-14-013, Letter from F. Dacimo to NRC Document Control Desk, Additional Information Regarding the License Renewal Application - Action Item 7 from MRP-227-A, Attach. 1 at 3, 4 (Jan. 28, 2014) (NYS000503).

262 See Entergys Testimony at A179 (ENT000616).

263 NYS Revised SOP at 17 (NYS000481).

264 See Entergys Testimony at A179 (ENT000616).

265 See Pilgrim, CLI-10-14, 71 NRC at 461; Oyster Creek, CLI-09-7, 69 NRC at 270.

266 NYS Revised SOP at 41.

this proceeding, such arguments are a collateral attack the license renewal rules in 10 C.F.R. Part 54.

If Dr. Lahey and the States concern is with loads that are within the CLB of IP2 and IP3, then such loads are appropriately addressed in the RVI AMP.267 The MRP-227-A inspection and evaluation guidelines are intended to detect conditions that may impair the continued functionality of the RVIs, under CLB loadsincluding loss-of-coolant accident (LOCA) and seismic loads.268 First, the MRP-227-A guidelines specify inspections of key irradiated components to assure that there are no cracks that could lead to failure and loss of functionality under transient loads.269 Without the presence of cracking, the ability of the irradiation-strengthened RVI-material to withstand shock loads is not degraded.270 Second, if a degraded component is discovered, then MRP-227-A requires the explicit evaluation of CLB loads, including accident and transient loads such as acoustic loads and rarefraction waves due to a LOCA in an engineering evaluation, to the extent such loads are part of the IP2 or IP3 CLB.271 For potentially irradiated components, the engineering evaluation assumes the component is embrittled.272 Dr. Lahey disregards rather than disputes this well-established analytical approach to accounting for design basis loads, and instead merely speculates that synergistic aging effects 267 See Entergys Testimony § V.A.6 (ENT000616).

268 See id. at A180.

269 See id.

270 See id.

271 See id. (citing MRP-227-A § 6 (NRC00014C)).

272 See Entergys Testimony at A171 (ENT000616).

and shock loads have not been considered.273 This is simply not enough at this stage of the proceeding for NYS to meet its burden of going forward with evidence to support its case.274

g. The RVI AMP Uses Appropriate Inspection Techniques MRP-227-A and its companion document, MRP-228, specify inspection techniques for those PWR RVI components that are most susceptible to the aging effects of concern and have the highest risk associated with failure.275 The standards for deployment of these inspection techniques and the necessary qualification requirements for both equipment and personnel are given in MRP-228.276 The NRC Staff reviewed and approved the selected inspection techniques in its Safety Evaluation for MRP-227-A.277 Dr. Lahey criticizes the use of VT-3 visual inspections as inadequate for use in inspections for cracking,278 but the adequacy of these techniques is explained in extensive detail in MRP-228, and further explained in the response to an NRC Staff non-concurrence on this topic.279 Again, the State and Dr. Lahey disregard the available information, rather than dispute the adequacy of the record on the use of VT-3 examinations. Moreover, the State and Dr. Lahey cannot simply rely on an NRC Staff nonconcurrence as the basis for their challenge to the use of VT-3 inspections. In particular, Dr. Lahey does not assert any expertise in this area and he offers no opinion of his own regarding the strength or weakness of the VT-3 inspections, or even offer 273 Revised Lahey Testimony at 15-16 (NYS000482); see also, e.g., Declaration of Richard T. Lahey, Jr. at 13 ¶ 19 (Feb. 13, 2015) (NYS000483) (New Yorks main concerns . . . have simply been ignored).

274 See Oyster Creek, CLI-09-07, 69 NRC at 268-70.

275 See Entergys Testimony at A186 (ENT000616).

276 See id.

277 See id. at A188.

278 See Revised Lahey Testimony at 62 (NYS000482).

279 See Entergys Testimony at A188, A132 (ENT000616).

a preferred alternative of his own.280 Again, Dr. Lahey must do much more at this stage of the proceeding.281

h. The RVI AMP Includes Appropriate Acceptance Criteria, Corrective Actions, and Preventive Actions Dr. Lahey criticizes the RVI AMP for allegedly failing to include objective criteria . . .

for corrective actions . . . .282 Entergys witnesses demonstrate that the RVI AMP includes appropriate acceptance criteria, corrective actions, and preventive actions, consistent with the applicable guidance and current operating practice. First, the RVI AMP contains specific, conservative examination acceptance criteria,283 based on the acceptance criteria in MRP-227-A.284 The inspections required in Section 4 of MRP-227-A and relied upon in the IPEC RVI AMP and Inspection Plan are designed to detect all of the pertinent aging effects described above, with conservative examination acceptance criteria.285 In most cases the examination acceptance criterion is any detectable degradation.286 The specific acceptance criteria will be carried forward into the program procedural documents, including the Pre-Inspection Engineering Packages prepared prior to each inspection.287 If examinations reveal conditions that do not meet the examination acceptance criteria set forth in the IPEC RVI Inspection Plan, then the discovery of the condition is entered into the 280 Revised Lahey Testimony at 62 (NYS000482); cf. USEC, Inc. (Am. Centrifuge Plant), CLI-06-10, 63 NRC 451, 472 (2006) (holding that mere references to documents without explanation or analysis does not supply an adequate basis for admitting a contention, and that conclusory statements proffered by an alleged expert do not provide sufficient support for a contention).

281 See Oyster Creek, CLI-09-07, 69 NRC at 268-70.

282 Revised Lahey Testimony at 49 (NYS000482).

283 See Entergys Testimony at A189 (ENT000616) (citing NL-12-037, Attach. 2 at 52-57 (NYS000496)).

284 See id. (citing MRP-227-A at 5-1 to 5-23 (NRC000114B)).

285 See Entergys Testimony at A189 (ENT000616).

286 See NL-12-037, Attach. 2 at 52-57 (NYS000496)).

287 See Entergys Testimony at A189 (ENT000616).

IPEC corrective action program for resolution.288 This could lead to: (1) a more detailed inspection; (2) an engineering evaluation; (3) repair; or (4) replacement of the affected component.289 MRP-227-A Section 6 provides an overview of the methodologies to be used for the development of engineering evaluations, which consider CLB loading and the characteristics of the material including the potential effects of embrittlement.290 In addition, if an inspection of a Primary component detects aging effects that exceed the Expansion Criteria specified in the tables of Section 5 of MRP-227-A, then inspections of corresponding Expansion components must take place.291 Contrary to Dr. Laheys demand, there is no further regulatory requirement for objective criteria for corrective actions.292 On the contrary, it would be impractical to establish pre-defined criteria in advance for all potential unsatisfactory examination results for all components.

Instead, such issues are handled on a case-by-case basis through engineering evaluations conducted under Entergys corrective action and quality assurance programs.293 This is fully consistent with how such matters are managed for operating plants under Part 50.

On the topic of Primary and Expansion components, Dr. Lahey criticizes the linkage between the core barrel girth weld (which is a leading indicator for IASCC and irradiation embrittlement) to the LSCCs because of the alleged differences between these components.294 288 See id. at A190.

289 See id.

290 See id. at A191, A192.

291 See id. at A193.

292 Revised Lahey Testimony at 49 (NYS000482).

293 See Entergys Testimony at A190 (ENT000616).

294 See Revised Lahey Testimony at 60 (NYS000482).

Entergys witnesses show that Dr. Laheys criticisms lack merit.295 In MRP-227-A, the LSCCs are an Expansion component linked to the control rod guide tube lower flanges as a Primary component.296 In response to NRC Staff RAIs, Entergy appropriately modified the RVI AMP to link the LSCCs to an additional Primary component that is an appropriate predictor of IASCC and irradiation embrittlement (IE) in the LSCCs: the core barrel girth weld.297 Although Dr.

Lahey disputes this linkage, he only asserts that the core barrel girth weld may be exposed to different aging mechanisms and shock loads than the LSCCs.298 This unsupported speculation does not directly challenge Entergys detailed plant-specific technical evaluation of the susceptibility of LSCCs to thermal embrittlement (TE), IE, and IASCC in support of its RVI AMP.299 Dr. Lahey and the State next assert that the IPEC RVI AMP manifestly does not include preventative actions.300 But this claim is based on a single phrase, lifted from the RVI AMP out of context, and in disregard of the remainder of the section which explains the preventive actions Entergy is taking related to RVIseven if they are being taken in the context of programs other than the RVI AMP.301 In any event, Entergys witnesses readily show that this claim is incorrect, as shown in the LRA itself and the RVI AMP. Specifically, the IPEC Water Chemistry Control program provides for preventive and mitigative action by maintaining primary water chemistry in 295 See Entergys Testimony § VII.A.8.c (ENT000616).

296 See MRP-227-A at 4-26 (NRC000114B).

297 See Entergys Testimony at A194 (ENT000616).

298 Revised Lahey Testimony at 60 (NYS000482).

299 See Entergys Testimony at A196 (ENT000616).

300 Revised SOP at 26 ¶ 21 (quoting NL-12-037, Attach. 1 at 5 (NYS000496)); see also Revised Lahey Testimony at 53 (NYS000482).

301 See NL-12-037, Attach. 1 at 5 (NYS000496).

accordance with EPRI guidelines.302 In addition, as part of the RVI aging management activities, Entergy replaced the IP2 split pins in 1995, the IP3 split pins in 2009, and will replace the IP2 split pins again in 2016.303 Further, Entergy will use the Fatigue Monitoring Program to track fatigue usage of RVI components with CUF analyses, thereby ensuring that the number of transients does not exceed the assumptions in the Westinghouse fatigue analyses.304 In addition, Entergy has implemented neutron flux reduction programs to minimize radiation effects and the resulting potential for degradation.305 Finally, the State and Dr. Lahey argue that Entergy must proactively replace RVIs, rather than managing the effects of aging through inspections and appropriate corrective actions (i.e.,

an AMP).306 This position is entirely unsupported and disregards the technical basis for the RVI AMP as documented in all of the supporting reports for MRP-227-A. Moreover, the States position is contrary to Commission precedent, because it amounts to a demand that aging effects be precluded, and seeks to negate the regulatory standard in 10 C.F.R. § 54.21(a)(3), which requires the applicant to show that there is reasonable assurance that the effects of aging will be adequately managed.307 Thus, the States view is entirely without merit.

i. The IPEC Fatigue Evaluations Appropriately Analyze Environmentally-Assisted Fatigue The State also claims, in NYS-25, that the EAF evaluations prepared by Westinghouse in support of the IPEC LRA, including EAF evaluations of RVI components, may be non-302 See Entergys Testimony at A203 (ENT000616) (citing NL-12-037, Attach. 1 at 5 (NYS000496); SSER 2 at 3-66 (NYS000507) (emphasis added)).

303 See id. (citing SSER 2 at A-15 (NYS000507)).

304 See id.

305 See id. (citing NL-12-037, Attach. 2 at 21-22 (NYS000496)).

306 See NYS Revised SOP at 31 (NYS000481); Revised Lahey Testimony at 79 (NYS000482).

307 See Seabrook, CLI-12-5, 75 NRC at 315.

conservative.308 In contrast to this speculation, Entergys witnesses show that the Westinghouse EAF evaluations are fully-documented, conservative engineering analyses that support a finding that the effects of fatigue, including the effects of the reactor water environment, will be adequately managed.309 Entergys witnesses explain this in their testimony on this contention,310 and on the metal fatigue contention (NYS-26B/RK-TC-1B),311 which is incorporated by reference into their NYS-25 testimony. Specifically, consistent with Entergys commitments and with standard ASME Code methods, Westinghouse recalculated each of the limiting CLB CUFs provided in Section 4.3 of the LRA for the RVIs to include reactor coolant environmental effects.312 Entergys witnesses show that there is no technical basis to apply any additional correction factors to account for the potential effects of embrittlement on fatigue life, beyond the correction factors specified in NRC guidance.313 Moreover, fatigue is one of the eight age-related degradation mechanisms evaluated during the development of the guidelines in MRP-227-A.314 As a result, the RVI AMP includes inspections intended to identify potential cracking caused by fatigue in susceptible RVI components.315 These inspection activities are in addition to, not in lieu of, the review of EAF for RVI components under the FMP.316 Thus, taken together, the RVI AMP and FMP provide 308 NYS Revised SOP at 17 (NYS000481).

309 See Entergys Testimony § VII.A.9 (ENT000616).

310 See id.

311 See generally Entergys NYS-26B/RK-TC-1B Testimony (ENT000679).

312 See Entergys Testimony at A206 (ENT000616).

313 See Entergys NYS-26B/RK-TC-1B Testimony at A76 (ENT000679).

314 See id. at A208.

315 See id.

316 See id.

reasonable assurance that the effects of aging due to fatigue on RVI components will be adequately managed throughout the PEO.

j. The RVI AMP Addresses Operating Experience Entergys witnesses describe how the industry has engaged in a decade-long effort to evaluate aging management of PWR RVIs, implement plant-specific AMPs for aging management of internals, develop a detailed RVI inspection program that has been approved by the NRC, and continues to collect and share relevant inspection results and operating experience for improved reliability.317 Consistent with the operating experience element of the RVI AMP and Commitment 40, Entergy will continue to review domestic and international operating experience during the PEO, and appropriately apply that operating experience in the IPEC RVI AMP, including updated inspection methods and improved methods of evaluating aging effects.318 In sum, Entergys experts demonstrate that the IPEC RVI AMP is consistent with MRP-227-A, as it uses state-of-the-art engineering and operating experience and demonstrated inspection techniques. Dr. Lahey and the State have overlooked rather than disputed the substantial technical basis developed by the EPRI MRP. Overall, the RVI AMP provides reasonable assurance that the effects of aging on the IP2 and IP3 RVIs will be adequately managed such that the intended functions of the IP2 and IP3 RVIs will be maintained consistent with the CLB throughout the PEO, as required by 10 C.F.R. §§ 54.21(a)(3), 54.21(c), and 54.29(a).

317 See id. at A211.

318 See id. § VII.A.10.

2. Entergys Aging Management Activities for RPVs As noted above, the States initial pleadings in 2007 on this contention focused primarily on the RPVs, rather than the RVIs, claiming that the information in the LRA on the RPV TLAAs did not include information on age-related accident analyses,319 and that an intermediate shell in IP2 will not meet the upper shelf energy acceptance criterion of 50ft-lb.320 Following the admission of contention NYS-25, Entergy submitted several RPV-related amendments to clarify its LRA, revise the description of how Entergy would address the then-proposed alternate PTS rule, and note the closure of certain RPV-related commitments.321 The State, however, has never amended NYS-25 to address or challenge these updates.322 Instead, the State has shifted its focus to RVIs.323 In particular, in Dr. Laheys prefiled testimony and the States statements of position on this contention, Dr. Lahey and the State do not allege any specific deficiency in Entergys LRA regarding the RPVs.324 To ensure a complete record, however, Entergys expert witnesses summarize the information regarding RPVs in the IPEC LRA and show that the LRA complies fully with 10 C.F.R. Parts 50 and 54 and is consistent with NRC Staff guidance regarding the management of the effects of aging and the evaluation of TLAAs for RPVs.325 In his most recent testimony, Dr. Lahey refers to certain documents discussing Branch Technical Position (BTP) 5-3, which is longstanding NRC guidance for estimating the initial, 319 NYS Petition at 224.

320 Id. at 226.

321 See Entergys Testimony at A50 (ENT000616).

322 See id. at A51.

323 NYS Revised SOP at 17 (NYS000481) ([t]he focus of Contention 25 is Entergys deficient AMP for RPVIs); see also Position Statement at 10 (NYS000293).

324 See Report (NYS000296); Lahey 2011 Testimony (NYS000294); Revised Lahey Testimony at 74 (NYS000482).

325 See Entergys Testimony §§ V.A,C, VII.B (ENT000616).

unirradiated transition temperature for certain RPVs when some of the required testing information is not available, and suggests that certain RPV embrittlement analyses may be non-conservative.326 But he cites this example only for the general principle that it is very important to preserve - rather than erode - operational safety margins as reactors age.327 Therefore, the States testimony contains no valid challenge to Entergys LRA with regard to the management of the effects of aging on RPVs. In any event, Entergys witnesses explain that this is actually a good example of the level of inherent conservatisms in embrittlement evaluations for RPVs. Specifically, the industry has now shown that other conservatisms and margin in RPV embrittlement calculations were more than sufficient to offset the potential non-conservatism identified in the BTP 5-3 methodology.328 VI. CONCLUSION For the foregoing reasons, the IPEC RVI and RPV aging management activities are consistent with NRC guidance, which is entitled to special weight, and satisfy all regulatory requirements. Therefore, Entergys LRA provides reasonable assurance that the effects of aging will be adequately managed throughout the PEO. The Intervenors have not carried their burden of providing sufficient evidence to support the claims made in NYS-25. Accordingly, NYS-25 should be resolved in Entergys favor.

326 Revised Lahey Testimony at 74 (NYS000482).

327 Id.

328 See Entergys Testimony at A79 (ENT000616).

Respectfully submitted, Executed in Accord with 10 C.F.R. § 2.304(d)

William B. Glew, Esq. Kathryn M. Sutton, Esq.

Entergy Nuclear Operations, Inc. Paul M. Bessette, Esq.

440 Hamilton Avenue Raphael P. Kuyler, Esq.

White Plains, NY 10601 MORGAN, LEWIS & BOCKIUS LLP Phone: (914) 272-3202 1111 Pennsylvania Avenue, N.W.

Fax: (914) 272-3205 Washington, D.C. 20004 E-mail: wglew@entergy.com Phone: (202) 739-3000 Fax: (202) 739-3001 E-mail: ksutton@morganlewis.com E-mail: pbessette@morganlewis.com E-mail: rkuyler@morganlewis.com Counsel for Entergy Nuclear Operations, Inc.

Dated in Washington, D.C.

this 4th day of September 2015 DB1/ 84307038