Text

NYS000590 - NYS Supplemental Written Testimony of Richard Lahey - Public
	ML16064A512
Person / Time
Site:	Indian Point
Issue date:	03/04/2016
From:	Lahey R; Rensselaer Polytechnic Institute
To:	; Atomic Safety and Licensing Board Panel
	SECY RAS
References
	50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01, RAS 50983
	Download: ML16064A512 (50)
	v • d • e

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement UNITED STATES NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

x In re:

Docket Nos. 50-247-LR; 50-286-LR License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 Entergy Nuclear Indian Point 3, LLC, and Entergy Nuclear Operations, Inc.

March 4, 2016

x SUPPLEMENTAL WRITTEN TESTIMONY OF DR. RICHARD T. LAHEY, JR.

REGARDING CONTENTIONS NYS-25, NYS-26B/RK-TC-1B, AND NYS-38/RK-TC-5 NYS000590 (Public Version)

Submitted: March 4, 2016

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 i

Table of Contents Page I.

Introduction.......................................... 1 II.

The Inspection Response Plan and Flaw Acceptance Criteria.............................................. 4

1. The Inspection Response Plan...................... 6

2. The Flaw Acceptance Criteria for IP-2 and IP-3.... 7 III.

Westinghouses Baffle-Former Bolt Analysis........... 12 IV.

Entergys Flaw Acceptance Criteria & the Eason and Pathania Paper....................................... 29 V.

The IP-3 Inspection Report........................... 42 VI.

Conclusion........................................... 44

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 ii List of Proposed New Exhibits Proposed Exhibit No.

Document NYS000583 Indian Point Units 2 and 3 Inspection Response Plan for Aging Management of MRP-227-A Primary and Expansion Components (WCAP-17941-P, Rev. 1) (IPECPROP00085900),

January 12, 2016 Short title:

Inspection Response Plan NYS000584 Background and Technical Basis Supporting Engineering Flaw Acceptance Criteria for Indian Point Unit-2 Reactor Vessel Internals MRP-227-A Primary and Expansion Components (WCAP-17949-P, Rev. 0)

(IPECPROP00085359), January 12, 2016 Short title:

Flaw Acceptance Criteria for IP-2 NYS000585 Background and Technical Basis Supporting Engineering Flaw Acceptance Criteria for Indian Point Unit-3 Reactor Vessel Internals MRP-227-A Primary and Expansion Components (WCAP-17951-P, Rev. 0)

(IPECPROP000856629), January 12, 2016 Short title:

Flaw Acceptance Criteria for IP-3 NYS000586 Determination of Acceptable Baffle-Former Bolting for Indian Point Units 2 and 3 (WCAP-18048-P, Rev. 0) (IPECPROP00086434),

January 12, 2016 Short title:

Bolting Analysis

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 iii Proposed Exhibit No.

Document NYS000587 E. Eason and R. Pathania, Disposition Curves for Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in Light Water Reactor Environments, ASME PVP2015-4532, Proceedings of the ASME 2015 Pressure Vessels and Piping Conference, July 19-23, 2015 Short title:

Eason and Pathania Paper NYS000588 USNRC Indian Point Nuclear Generating Unit License Renewal Inspection Report 05000286/2015011, Nov. 19, 2015 (ML15323A026)

Short title:

IP-3 Inspection Report

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 iv List of Acronyms Acronym Meaning AMP aging management plan ASLB, the Board Atomic Safety Licensing Board BWR boiling water nuclear reactor CGR crack growth rate DBA design basis accident EPRI Electric Power Research Institute FMEA failure modes and effects analysis IASCC irradiation-induced stress corrosion cracking IP-2 Indian Point nuclear facility Unit-2 IP-3 Indian Point nuclear facility Unit-3 LBB leak-before-break LOCA loss-of-coolant accident LRA license renewal application NYS, the State the State of New York PEO period of extended operation PWR pressurized water nuclear reactor RTA real-time analysis RVI reactor vessel internal SSE safe-shutdown earthquake UFSAR Updated Final Safety Analysis Report

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 1

I.

Introduction 1

On behalf of the State of New York (NYS or the State),

2 the Office of the Attorney General hereby submits the following 3

testimony by RICHARD T. LAHEY, JR., Ph.D., regarding Contentions 4

NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5.

5 Q.

Please state your full name.

6 A.

Richard T. Lahey, Jr.

7 Q.

Dr. Lahey, you have previously provided your 8

educational and professional qualifications and submitted 9

testimony in this proceeding, correct?

10 A.

Yes. I have previously submitted testimony and reports 11 in this proceeding. My education and professional qualifications 12 and experience are described in my previously submitted 13 Curricula Vitae and were summarized in my previous testimony.

14 Q.

I show you proposed exhibits NYS000583 through 15 NYS000588. Do you recognize those documents?

16 A.

Yes. These documents consist of four technical reports 17 authored by Westinghouse on behalf of Entergy, one technical 18 paper authored by E. Eason and R. Pathania, and one USNRC 19 inspection report for Indian Point Unit-3.

20 Q. Have you reviewed proposed exhibits NYS000583 through 21 NYS000588?

22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 2

A. Yes.

1 Q. Have these documents caused you to change the 2

testimony and opinions that you have previously submitted in 3

this proceeding in connection with Contentions NYS-25, NYS-4 26B/RK-TC-1B and NYS-38/RK-TC-5?

5 A. No. In fact, I believe that these documents support 6

the concerns, opinions, and testimony that I have presented 7

previously in this proceeding. It is my opinion that the 8

documents are relevant to the technical and legal matters at 9

issue in this proceeding, and that they provide substantial 10 support for the opinions that I have presented to the Atomic 11 Safety and Licensing Board (the Board) in my testimony and 12 written reports in this proceeding. In my opinion, these 13 documents provide further bases for the Board to conclude that 14 Entergys license renewal application (LRA) for the Indian 15 Point facilities fails to adequately manage the effects of aging 16 degradation such as embrittlement, fatigue, and irradiation 17 assisted stress corrosion cracking (IASCC) on reactor vessel 18 internal (RVI) systems, structures and components. They also 19 describe the facilities approach to fatigue analysis and 20 confirm that certain aging management program plan (AMP) 21

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 3

details will be deferred until well into the period of extended 1

operation (PEO).

2 Q. In your opinion, what is the significance of the 3

proposed new exhibits?

4 A.

The documents call into question whether Entergys 5

currently proposed amended and revised AMP for RVIs and the 6

associated Inspection Plan for the Indian Point nuclear 7

facilities, e.g., exhibits NYS000496-NYS000507 (Amended and 8

Revised RVI AMP), provide reasonable assurance against the 9

failure of various RVI systems, structures, and components. The 10 documents also suggest that Westinghouses approach to 11 developing flaw acceptance criteria and the associated 12 inspection intervals under Entergys Amended and Revised RVI AMP 13 may be inadequate and non-conservative. For example, 14 15 16 17 18 19

. In my testimony below, I will 20 begin by describing how three of the proposed new exhibits 21 provide specific details on the implementation of Entergys 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 4

Amended and Revised RVI AMP and Inspection Plan. Next, I will 1

discuss my concern regarding Westinghouses plant-specific 2

bolting pattern analysis for the baffle-former bolts at Indian 3

Point Unit-2 (IP-2) and Unit-3 (IP-3). Thereafter, I will 4

describe my concern that Entergys Amended and Revised RVI AMP 5

contains several non-conservative assumptions that may cause 6

IASCC cracks to grow faster at IP-2 and IP-3 than would be 7

accounted for by Entergy under its Amended and Revised RVI AMP.

8 Finally, I discuss USNRCs recent inspection report for IP-3.

9 II.

The Inspection Response Plan and Flaw Acceptance Criteria 10 Q. I show you three of the proposed new exhibits, each of 11 which relates to Entergys Amended and Revised RVI AMP. The 12 documents are:

13

Indian Point Units 2 and 3 Inspection Response Plan 14 for Aging Management of MRP-227-A Primary and Expansion 15 Components, a document that has been marked as proposed exhibit 16 NYS000583, and identified by Westinghouse as WCAP-17941-P, Rev.

17 1 (Inspection Response Plan);

18

Background and Technical Basis Supporting Engineering 19 Flaw Acceptance Criteria for Indian Point Unit-2 Reactor Vessel 20 Internals MRP-227-A Primary and Expansion Components, a 21 document that has been marked as proposed exhibit NYS000584, and 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 5

identified by Westinghouse as WCAP-17949-P, Rev. 0 (Flaw 1

Acceptance Criteria for IP-2); and, 2

Background and Technical Basis Supporting Engineering 3

Flaw Acceptance Criteria for Indian Point Unit-3 Reactor Vessel 4

Internals MRP-227-A Primary and Expansion Components, a 5

document that has been marked as proposed exhibit NYS000585, and 6

identified by Westinghouse as WCAP-17951-P, Rev. 0 (Flaw 7

Acceptance Criteria for IP-3).

8 Please describe how the Inspection Response Plan 9

(NYS000583) and Flaw Acceptance Criteria for IP-2 (NYS000584) 10 and IP-3 (NYS000585) each relate to the implementation of 11 Entergys Amended and Revised RVI AMP?

12 A. According to the Inspection Response Plan and the Flaw 13 Acceptance Criteria for IP-2 and IP-3, these documents, 14 15 16 17 18 19

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 6

1.

The Inspection Response Plan 1

Q. I show you the Inspection Response Plan (NYS000583).

2 Would you please explain your understanding of how this document 3

relates to Entergys Amended and Revised RVI AMP?

4 A.

Yes.

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 7

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16

2.

The Flaw Acceptance Criteria for IP-2 and IP-3 17 Q.

I now show you two documents, the Flaw Acceptance 18 Criteria for IP-2 (NYS000584), and the Flaw Acceptance Criteria 19 for IP-3 (NYS000585). Would you explain your understanding of 20 how these documents relate to Entergys Amended and Revised RVI 21 AMP?

22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 8

A.

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 10

and Entergy 1

License Renewal Application - Revised Reactor Vessel Internals 2

Program and Inspection Plan, NL-12-037, Attach. 2, at Table 5-2 3

(Primary components) and Table 5-3 (Expansion components).

4 In my view, waiting 10 years between inspections is not a 5

reasonable or prudent approach for plants operating beyond their 6

originally intended design life. As I have discussed in my prior 7

testimony, many systems, structures, and components at Indian 8

Point will be significantly degraded as they approach the end of 9

their PEO. In particular, some RVIs will have accumulated 10 sufficient neutron fluence such that the effects of various age-11 related degradation mechanisms such as irradiation-induced 12 embrittlement and IASCC will be a concern. Because degradation 13 is expected to accelerate with aging, I believe more frequent 14 inspections are warranted, particularly as these RVI systems, 15 structures, and components reach the later stages of the 16 proposed PEO when exposure to radiation is expected to meet or 17 exceed thresholds for the development of significant 18 embrittlement and IASCC.

19 Q.

Do you have any other concerns about Entergys Amended 20 and Revised RVI AMP, as described in the Inspection Response 21 Plan and the Flaw Acceptance Criteria for IP-2 and IP-3?

22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 11 A.

Yes. As I have discussed in my prior testimony, any 1

inspection-based aging management program has some probability 2

that an existing flaw will go undetected. For example, 3

4 5

6 7

8

. Undetected cracks are a concern because 9

inspections will not occur for another 10 years. Not only can 10 existing cracks -- both detected and undetected -- grow, but new 11 cracks can develop and grow in the period between inspections.

12 Embrittlement results in decreased fracture toughness, with a 13 corresponding decrease in the critical flaw size that can lead 14 to accelerated component failure. Thus my concern is that with 15 embrittled components, cracks may grow to critical size, and the 16 components may fail before corrective action is taken to ensure 17 component integrity or assembly functionality. Although 18 Westinghouses flaw evaluation procedures under Entergys 19 Amended and Revised RVI AMP 20 21

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 12

, Entergy cannot possibly evaluate and monitor 1

any cracks that it has failed to identify.

2 III. Westinghouses Baffle-Former Bolt Analysis 3

Q.

Dr. Lahey, as described above, one of your concerns 4

relates to Entergys and Westinghouses approach to evaluating 5

baffle-former bolt failures at the Indian Point facilities. How 6

does Entergy plan to address baffle-former bolt failures in the 7

two Indian Point reactors?

8 A.

Proposed exhibit NYS000586, a Westinghouse report 9

entitled Determination of Acceptable Baffle-Former Bolting for 10 Indian Point Units 2 and 3 (WCAP-18048-P, Rev. 0) (Bolting 11 Analysis),

12 13 14 15 16

. This 17 document is the site-specific bolting analysis that Entergy 18 witnesses referenced during the November 2015 evidentiary 19 hearing. See Transcript of ASLB Hearing, at 5239, 5314 (Nov. 17, 20 2016). At the time of this ASLB hearing, my understanding was 21 that this site-specific bolting analysis was not available.

22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 13 Q.

In your opinion, are bolting failures a concern at 1

Indian Point?

2 A.

Yes. I have previously testified that I am concerned 3

about the possible failure of baffle-former bolts during an 4

extended 20-year period of operation, particularly because 5

failures of baffle-former bolts have been observed in other 6

similar nuclear power plants, and due to my prevailing concern 7

that Westinghouses analyses do not consider the combined or 8

synergistic effects of multiple aging degradation mechanisms 9

acting together, and the effect of very energetic shock loads.

10 Indeed, the failure of baffle-former bolts is a significant 11 issue for aging pressurized water nuclear reactor (PWR) plants 12 worldwide, and certainly for the two Indian Point plants.

13 Initially identified in European PWRs in 1989, numerous flaws in 14 baffle-former bolts have, in recent years, been observed in a 15 number of domestic PWRs as well. For example, failed baffle-16 former bolts have been found in domestic PWRs at Ginna, D.C.

17 Cook, Surry, Prairie Island, Point Beach, and Robinson. Among 18 other things, bolt failures can contribute to fuel damage and 19 loose parts. Moreover, bolt failures can greatly degrade core 20 cooling capabilities and compromise safe plant operations.

21

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 15 1

2 Q. So, in your opinion, is a bolting analysis that is 3

based on 4

an adequate approach?

5 A.

No. In my opinion, analysis of the single and two-6 phase the shock loads typically associated with the original 7

design basis accident (DBA) LOCA (i.e., as specified in 10 8

C.F.R. 50 Appendix K, an instantaneous double-ended guillotine 9

break of the most limiting primary coolant line, in particular, 10 the cold leg), is much more appropriate to help ensure that the 11 RVI components will perform their intended functions during the 12 20 years of extended operation. It is therefore my opinion that 13 Entergys and Westinghouses proposed methodology 14 is non-15 conservative and ill-advised from a safety perspective for a 20-16 year period of extended operation beyond the original 40-year 17 license term.

18 Q. How do the shock loads differ between, for example, 19

?

20 A.

Large energetic, instantaneous DBA LOCA cold leg 21 breaks generate substantially higher in-vessel shock loads than 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 16 1

. To 2

understand the implications of this, I note that in a typical 3

structural dynamics analysis, a structure may be modeled as a 4

spring-mass-dashpot system:

5

+ 2 +

6 where and are, respectively, the second and first ordinary 7

derivatives with respect to time (t) of the displacement (x) of 8

a structure. The term =

is the natural frequency of the 9

structure (which has an equivalent spring constant of k, a 10 damping rate of, and a mass of M), and the transient force (F) 11 acting on a structure is given by F = pA, where p is the 12 transient pressure differential acting on a structure of area A.

13 During the subcooled decompression stage of a DBA LOCA, the 14 initial p across the baffle plate/bolting complex (and, for 15 that matter, other in-vessel structures) can be given by a 16 Heaviside step function (U). In particular, if P is the 17 magnitude of the largest transient p, then, F(t) = PAU(t) =

18 PA if t 0.0, and zero for t < 0.0.

19 For this type of impulsive forcing function, an underdamped 20 system, such as the baffle/bolting complex in a PWR, responds as 21 shown schematically in the hand-drawn diagram I prepared during 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 18 original state, xss = 0.0. Naturally, the smaller the size of the 1

pipe break, the less the inertial overshoot. Additionally, if 2

the p(t) occurs over a finite period of time (t) 3 4

the induced overshoot for these ramps (as opposed to steps) in 5

p(t) will also be reduced. In any event, the magnitude of the 6

transient overshoot in the displacement, x(t) of the structure 7

determines the maximum strain (and stress) that the structure is 8

exposed to. Because the baffle-former bolts during the period of 9

extended operations will be highly fatigued and embrittled, 10 their stress-strain curve will initially be quite steep (i.e.,

11 the linear elastic part will have a large slope due to 12 irradiation-induced hardening), and the ultimate stress (which 13 for highly embrittled stainless steel bolts will be almost the 14 same as the yield stress) will be larger than for a 15 corresponding ductile structure. However, physical failure of 16 the structure is expected once the strain associated with the 17 ultimate stress has been exceeded. Thus, for the large 18 displacements (and thus, for the large strains) associated with 19 energetic shock loads such as those for a DBA LOCA, failure of 20 many of the baffle-former bolts in IP-2 and IP-3 may occur. If 21 so, there may no longer be an intact core geometry and thus 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 19 adequate core cooling will no longer be assured. In addition, 1

the two-phase LOCA loads, which occur after the subcooled 2

decompression loads, may lead to crushing of the fuel rod grids 3

and the inability to insert the control rods (which are required 4

to safely shut down the reactor). See Westinghouse Methodology 5

for Evaluating the Acceptability of Baffle-Former-Barrel Bolting 6

Distributions under Faulted Load Conditions, WCAP-15030-NP-A 7

(Mar. 2, 1999), § K at 2-6 (ENT000655).

8 Because Entergys evaluations (performed by Westinghouse) 9 assumes 10 11 12 13 14 15

. As such, these 16 transient loads are non-conservative and serve to undermine the 17 validity of any future real-time analyses of acceptable bolting 18 patterns performed for IP-2 and IP-3. Significantly, 19 20 21

, such as the DBA LOCA which 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 20 was required by 10 C.F.R. 50 Appendix K in the original design 1

of the Indian Point reactors. See Updated Final Safety Analysis 2

Report (UFSAR) for IP-2, § 14.3.4.3.1 at 113 (NRC000206); § 3

4.1.2.4.2 (ENT000634).

4 5

6 Q. In your opinion, should Entergys and Westinghouses 7

consideration of dynamic shock loads in the Bolting Analysis 8

include DBA LOCA events such as an instantaneous double-ended 9

guillotine break in the main coolant piping?

10 A.

Yes. I am aware that USNRC regulations require an 11 applicant to consider the impact of various size pipe ruptures, 12 including an instantaneous double-ended guillotine rupture of 13 the cold leg in the primary system of a PWR when assessing plant 14 safety under LOCA conditions. See, e.g., 10 C.F.R. 50, 15 Appendices A, K. However, I also understand that in 1998, the 16 USNRC Staff authorized the analysis of acceptable baffle-former-17 barrel bolting distributions based on smaller auxiliary line 18 breaks and slower LBB break opening times. See Westinghouse 19 Methodology for Evaluating the Acceptability of Baffle-Former-20 Barrel Bolting Distributions under Faulted Load Conditions, 21 WCAP-15030-NP-A (Mar. 2, 1999) (ENT000655),

22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 21 1

Nevertheless, as I have 2

indicated in my previously filed testimony, 3

were never intended to be applied to 4

evaluations designed to demonstrate core cooling capability and 5

internal component integrity, and certainly not for aged and 6

degraded components that have been in use for over 40 years, the 7

original design life of these plants. Instead, this methodology 8

was developed for use in pipe whip evaluations within the 9

containment. Indeed, it is totally inconsistent to use 10 for the analysis of potential RVI failures because, 11 for example, significant baffle-former bolt failures may lead to 12 an uncoolable core geometry, while the DBA LOCA analysis 13 required by 10 C.F.R. 50, Appendix K implicitly assumes an 14 intact core geometry. Thus, the PWR plant safety analysis 15 requires the assumption of an instantaneous double-ended 16 guillotine break of the cold leg for both the transient loads on 17 RVIs and the transient core thermal-hydraulics. Anything else 18 would be inadequate, and, in my opinion, Entergy and 19 Westinghouse 20 21

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 22 Q. In your opinion, does Westinghouses Bolting Analysis 1

(NYS000586) demonstrate that the possible failure of baffle-2 former bolts at IP-2 and IP-3 will be adequately addressed?

3 A. Absolutely not. Interestingly, the report reveals that 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 23 1

2 3

4 5

Q.

Throughout your testimony in this proceeding, you have 6

expressed your concern that Entergy is not considering adequate 7

shock loads in the various analyses it has conducted under its 8

LRA.

9

, are there other shock loads you believe Entergy 10 should consider in its analysis of acceptable baffle-former 11 bolting at the Indian Point facilities?

12 A.

Yes. The safe operation of a nuclear plant requires 13 analyses of the plants responses to various postulated 14 equipment failures or malfunctions. It is important to select a 15 sufficiently broad spectrum of accident and transient events to 16 evaluate. To be conservative, it is important to identify the 17 accidents or events that give rise to the most limiting or 18 challenging conditions. As a nuclear safety expert, my opinion 19 is that it is vital to consider the most limiting event -- which 20 for the Indian Point plants is a double-ended guillotine break 21 of the cold leg between the reactor coolant pump and the reactor 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 25 core cooling. Does the existence of this system alleviate your 1

concern?

2 A. No. Although Entergy has emergency core cooling 3

systems in IP-2 and IP-3, they are not fail-safe. These 4

engineered safety coolant injection systems are intended to 5

maintain core cooling in the event of various pipe ruptures, 6

including a DBA LOCA, but my concern is that due to the 7

associated shock loads generated during a DBA LOCA some of the 8

highly embrittled and fatigued reactor vessel internals (e.g.,

9 the baffle-former bolting) may fail such that there will no 10 longer be a coolable core geometry. Moreover, as I have 11 previously testified, important primary coolant pressure 12 boundary components (e.g., the accumulator line and nozzles) 13 that are degraded due to fatigue and stress corrosion cracking 14 and/or embrittled due to thermal embrittlement of the welds, may 15 fail under large LOCA-induced shock loads.

16 17 18

, it is unclear 19 whether a coolable core can be maintained in the event of a 20 large, energetic line break in IP-2 and IP-3. In my opinion, 21 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 26 full dynamic effects of a blowdown following a DBA LOCA in order 1

to confirm that reactor vessel internals and reactor coolant 2

piping can withstand the most limiting LOCA load in combination 3

with a SSE, which may initiate the LOCA event. That is, if a DBA 4

LOCA event ever occurs, it is most likely during the PEO when an 5

aged and degraded PWR experiences a significant seismic event.

6 Thus the safety evaluation of this postulated event is 7

essential.

8 Q.

In your opinion, does the Bolting Analysis (NYS000586) 9 adequately address your previous concerns regarding the impact 10 of embrittlement, or the loss of ductility, on the ability of 11 baffle-former bolts to withstand loads?

12 A.

No.

13 14 15 16 17 18 rather than outline corrective actions such as 19 repair or replacement to address cracked bolts.

20 21 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 27 1

2 3

4 5

6 7

8 My concern is that 9

, it would likely be 10 insufficient to enable the bolts to sustain an energetic 11 impulsive shock load, such as that associated with a DBA LOCA.

12 Q:

Does Westinghouse address undetected cracks in bolts?

13 A:

14 15 16 Q.

How do Entergy and Westinghouse propose to address 17 these undetected flaws and the risk that failed bolts pose to 18 the integrity of the structure and core coolability?

19 A.

20 21 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 29 1

2 3

In my opinion, it is necessary to maintain 4

a substantial margin throughout the PEO because the number of 5

baffle-former bolts exceeding the stress and IASCC thresholds is 6

expected to increase -- not decrease -- over time.

7 Q.

Did the Bolting Analysis identify conditions where 8

additional evaluation is required when a flaw is detected?

9 A.

10 11 12 13 14 15 16 IV.

Entergys Flaw Acceptance Criteria & the Eason and Pathania 17 Paper 18 Q. Dr. Lahey, as discussed above, another concern you 19 raise relates to potential non-conservatisms in the flaw 20 acceptance criteria for certain RVI structures and components, 21 and particularly to the crack growth rate model used by 22 Westinghouse in developing these acceptance criteria. How does 23

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 30 Westinghouse address crack growth in its flaw acceptance 1

criteria?

2 A. As indicated in the Flaw Acceptance Criteria for IP-2 3

(NYS000584) and the Flaw Acceptance Criteria for IP-3 4

(NYS000585), for certain components, 5

6 7

8 9

10 11 12 13

14 15 16 17

18 19 20 21

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 31

1 2

3 4

Q. I present to you a document marked as exhibit 5

NYS000587, a paper by E. Eason and R. Pathania, entitled 6

Disposition Curves for Irradiation-Assisted Stress Corrosion 7

Cracking of Austenitic Stainless Steels in Light Water Reactor 8

Environments, PVP2015-4532, from the Proceedings of the ASME 9

2015 Pressure Vessels and Piping Conference (Eason and Pathania 10 Paper). Have you reviewed this document?

11 A. Yes.

12 Q. Are you familiar with the authors of this paper?

13 A. I have never met them but I understand that Mr.

14 Pathania, one of the co-authors, is a technical executive with 15 the Electric Power Research Institute (EPRI). Additionally, he 16 is EPRIs Roadmap Owner for its BWR and PWR irradiated 17 materials testing and the degradation models for the reactor 18 internals project. See http://mydocs.epri.com/docs/Portfolio/

19 P2016/Roadmaps/NUC_MAT_01-BWR-PWR-Irradiated-Materials-20 Testing.pdf.

21

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 32 Q. In your opinion, what in the Eason and Pathania Paper 1

is relevant to issues in this proceeding?

2 A. The Eason and Pathania Paper presents new IASCC crack 3

growth rate disposition curves for both boiling water reactors 4

(BWRs) and PWRs. The paper summarizes the results of a multi-5 year international effort sponsored by EPRI to collect, rank, 6

and model IASCC crack growth rate data. See Eason and Pathania 7

Paper at 1. The paper summarizes over 800 IASCC crack growth 8

rate data points collected from six laboratories worldwide, and 9

reviewed and ranked by an international panel of known IASCC 10 experts. Id. at 2.

11 According to the Eason and Pathania Paper, the new crack 12 growth rate disposition curves presented in the paper reflect 13 an improvement over the earlier BWRVIP-99-A and MRP-227-A 14 disposition curves. Id. at 9. The earlier BWRVIP-99-A and MRP-15 227-A crack growth rate disposition curves were developed circa 16 2001 from then-available data, and were primarily used for 17 estimating crack growth rates in BWR environments. Id. at 2.

18 However, according to the Eason and Pathania Paper, 19 substantially more data is now available, and is reflected in 20 their paper. Id. at 2.

21

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 33 Q.

What information regarding IASCC crack growth rate 1

disposition curves does the Eason and Pathania Paper disclose?

2 A.

Figure-3 of the Eason and Pathania Paper presents a 3

PWR primary water crack growth rate disposition curve based on 4

the newly available data. Id. at 4. The new curve is plotted at 5

a temperature of 325°C and an irradiated yield stress of 700 6

MPa. Id. at 4. For comparison, Figure-3 also presents a plot of 7

the older MRP-227-A disposition curve. Id. at 4.

8 The Eason and Pathania Paper states that the new PWR 9

primary water crack growth rate disposition curve is about a 10 factor of 5.6 higher than the dashed MRP-227-A curve. Id. at 4.

11 The Eason and Pathania Paper goes on to state that a higher 12 irradiated yield stress of 970 MPa (rather than 700 MPa) would 13 shift the new PWR primary water crack growth rate disposition 14 curve in Figure-3 further upward by a factor of 2.3. Id. at 5.

15 Taken together, the Eason and Pathania Paper suggests that 16 the PWR primary water crack growth rate disposition curve could 17 increase the older MRP-227-A IASCC crack growth rate disposition 18 curve by a factor of at least 10, depending on the operating 19 temperatures and irradiated yield stress. Id. at 4-5, 9. Stated 20 another way, the Eason and Pathania Paper indicates that IASCC 21 cracks can grow up to an order of magnitude faster in the PWR 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 34 primary water operating environment compared to earlier (mostly 1

BWR) data that was used in the MRP-227-A disposition curves.

2 Q. How does this finding relate to Entergys Amended and 3

Revised RVI AMP?

4 A. The Eason and Pathania Paper suggests that the crack 5

growth rates 6

under Entergys Amended 7

and Revised RVI AMP are non-conservative. As I explained above, 8

the Eason and Pathania Paper indicates, based on over 800 IASCC 9

crack growth rate data points, that IASCC crack growth rates in 10 PWR primary water environments are between five and 10 times 11 greater than those set forth in MRP-227-A.

12 13 14 15 Q. To the extent that 16 IASCC crack growth rates that are non-17 conservative, what are the implications for the adequacy of 18 Entergys Amended and Revised RVI AMP?

19 A. In my opinion, the new IASCC crack growth rate 20 disposition curves developed by EPRI, and discussed in the Eason 21 and Pathania Paper, call into serious question the conservatism 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 35 of Entergys Amended and Revised RVI AMP.

1 2

3

, the plan is non-conservative and fails to 4

provide reasonable assurance that RVI component cracking due to 5

aging degradation mechanisms such as IASCC will be adequately 6

managed.

7 17 19 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 36 I am concerned, however, that possible 5

non-conservatism in these flaw acceptance criteria means that 6

both identified and undetected cracks may grow faster than 7

Entergy assumes but will not be identified during 8

follow-up inspections.

9 There are at least three specific circumstances under 10 Entergys Amended and Revised RVI AMP that, in my opinion, may 11 result in the failure of RVIs due to the faster PWR IASCC crack 12 growth rates identified in the Eason and Pathania Paper:

13

First, a flaw or surface crack in an RVI system, 14 structure, or component that is not detected during baseline 15 visual inspections. This is a real possibility because any 16 visual inspection technique has an inherent limit of detection, 17 as I discussed above and in my prior testimony. Under Entergys 18 Amended and Revised RVI AMP, the next inspection will not occur 19 for 10 years. Should the flaw or crack grow to critical size 20 before this next inspection, the component may fail. In 21 particular, given the results presented in the recent Eason and 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 37 Pathania Paper, it is possible, and even likely, that the flaw 1

or surface crack will grow at a faster rate 2

3 4

Second, a new flaw or surface crack in an RVI system, 5

structure, or component that develops after the baseline visual 6

inspection. Again, because the next inspection will not occur 7

for 10 years, I am concerned that the component may fail if the 8

flaw or crack grows to critical size before the next inspection.

9 In particular, it is possible, even likely, that such a crack 10 will grow faster 11 in light of the accelerated PWR IASCC 12 crack growth rate disposition curve presented in the Eason and 13 Pathania Paper.

14

Third, a flaw or surface crack that is detected by 15 visual inspection and is indicated for Entergys Corrective 16 Action Program, 17 18 19 20 21 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 38 1

2 3

. I am concerned that 4

such a crack may meet Entergys and Westinghouses calculated 5

allowable crack size set forth under the Flaw 6

Acceptance Criteria for IP-2 and IP-3, but then grow faster than 7

anticipated. In particular, towards the end of the PEO, when the 8

fluence exposure of RVI systems, structures, and components will 9

be in the range at which IASCC is a serious concern, it is 10 possible that Entergy will not be monitoring or inspecting a 11 developing crack at all. Additional inspections at the end of 12 the PEO are important because the estimated fluence threshold 13 for IASCC is 3 dpa, or 2 x 1021 n/cm2 (E > 1 MeV). See MRP-191 at 14 3-3, Table 3-2 (NYS000321). For many RVI components that 15 threshold may be reached in the latter part of the PEO, when no 16 inspections are expected to take place. Id. at 4-22 to 4-29, 17 Table 4-6. For example, with respect to girth welds, predicted 18 neutron fluence is expected to be 4 dpa towards the end of the 19 PEO. See USNRC Supplemental Safety Evaluation Report (SSER2) 20 (NYS000507), at 3-46.

21

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 39 In addition, Entergys use of 1

for developing its flaw 2

acceptance criteria underscores my prevailing concern, as 3

discussed above and in my previous testimony, that Entergys 4

Amended and Revised RVI AMP does not adequately protect against 5

the failure of fatigued and embrittled components that 6

experience significant shock loads (i.e., those due to a DBA 7

LOCA).

8 9

10 11 12 13 14 15

. In my 16 opinion, Entergy should, at a minimum, implement an AMP that 17 takes into account the latest information on crack growth rates, 18 apply appropriate DBA LOCA loads, and incorporate planned 19 inspections in the years approaching the end of the PEO, when 20 IASCC may become a very serious concern. Additionally, Entergy 21 should implement shorter inspection intervals combined with 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 40 regular, follow-up examinations to minimize the number of cracks 1

that escape detection and to properly monitor crack growth.

2 Q. Do you have any other opinions with regard to the new 3

PWR primary water crack growth rate disposition curve presented 4

in the Eason and Pathania Paper?

5 A.

Yes. As Entergy concedes, in the event that its LRA 6

for the Indian Point reactors is granted, it will not be 7

compelled to undertake future actions to incorporate the new 8

crack growth rate curves for PWRs into its Amended and Revised 9

RVI AMP. See Testimony of Entergy Witnesses Nelson F. Azevedo, 10 Robert J. Dolansky, Alan B. Cox, Jack R. Strosnider, Timothy J.

11 Griesbach, Randy G. Lott, and Mark A. Gray Regarding Contention 12 NYS-25 at A136 (ENT000616). Rather, such actions by Entergy 13 would be entirely voluntary. Id.

14 Q.

Does Entergys Amended and Revised RVI AMP and 15 Inspection Plan incorporate margin into its flaw acceptance 16 methodology?

17 A.

18 19 20 21 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 41 1

2 3

I am also concerned that this possible non-conservatism in 4

margin in Westinghouses flaw acceptance criteria creates a 5

source of uncertainty in addition to the non-conservatisms that 6

I have discussed previously. According to Entergy, the ASME 7

codes adjustment factors of 2 on stress and 20 on cycles in the 8

fatigue design curves provide substantial margin. See *Entergy 9

Revised Testimony on NYS-26B/RK-TC-1B (ENT000679), 43 (A70).

10 However, as explained in NUREG-6909, Revision 1, the factors of 11 2 on stress and 20 on cycles used in the ASME Code Section III 12 air fatigue design curves were intended to cover the effects of 13 variables that influence fatigue lives (i.e., material 14 variability, different heats, surface finish, size, mean stress, 15 and loading sequence) and are not per se safety margins. See 16 Effect of LWR Coolant Environments on the Fatigue Life of 17 Reactor Materials - Draft Report, NUREG-6909, Rev. 1 18 (NYS000490A), at XXV, 5. Because these factors were not 19 investigated in the laboratory tests that provided the data for 20 the curves, it is therefore not clear how much conservatism is 21 actually embedded in the ASME code. See NYS000490B at 147-148.

22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 42 Studies on vessel steels, piping and other components have shown 1

that ASME Code Section III fatigue design procedures do not 2

always contain large conservatisms and that cracks can initiate 3

before they are predicted to occur. Id. In my opinion, Entergys 4

reliance on the inherent margins in the ASME code for fatigue 5

evaluations is misplaced.

6 V.

The IP-3 Inspection Report 7

Q. I now show you a document marked as exhibit NYS000588, 8

which has the title, USNRC Indian Point Nuclear Generating 9

Unit License Renewal Inspection Report 05000286/2015011, 10 and is dated November 19, 2015 (IP-3 Inspection Report). Can 11 you describe this document?

12 A.

The IP-3 Inspection Report appears to be a document 13 authored by USNRC Staff that summarizes the Staffs view 14 concerning the status of Entergys commitments for IP-3. More 15 specifically, the IP-3 Inspection Report sets forth Staffs 16 conclusion that Entergy has fulfilled its remaining commitments 17 with respect to metal fatigue (Commitment 49), but that its 18 commitment to develop a plant-specific safety analysis for 19 reactor vessel plate B2803-03 three years prior to reaching the 20 pressurized thermal shock reference temperature (Commitment 32) 21 still remains outstanding.

22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 43 With respect to Commitment 49, which requires Entergy to 1

recalculate cumulative usage factors for certain reactor vessel 2

internals taking into account environmental effects, the IP-3 3

Inspection Report notes that initial application of the maximum 4

stainless steel Fen of 15.348 to the CUF values for those 5

components resulted in CUFen values in excess of the limit of 1.0 6

for five components and locations (i.e., upper support plate 7

assembly, upper support plate flange, lower core plate, lower 8

core support plate, and lower support columns) for IP-2 and 9

three components and locations (i.e., upper support assembly, 10 instrumentation columns, and lower support columns) for IP-3.

11 The report states that refined fatigue calculations were 12 performed for these components and locations to qualify them for 13 service. Significantly, the report summarizes the manner in 14 which Entergy has systematically removed conservatisms in order 15 to reduce the CUFen values to below 1.0. The document also 16 presents the USNRC Inspectors view that such refinement through 17 reduction of conservatisms obviates the need for further CUF 18 re-analysis, and/or repair or replacement. See IP-3 Inspection 19 Report (NYS000583), at 7. Clearly this iterative process is 20 troubling since the so-called conservatisms which were 21 eliminated may well be needed design margins.

22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 44 VI.

Conclusion 1

Q.

Would you summarize your testimony?

2 A.

In summary, it is my opinion that these new documents 3

provide additional bases for the Board to conclude that 4

Entergys Amended and Revised RVI AMP for IP-2 and IP-3 fails to 5

adequately manage the effects of aging, and does not provide 6

reasonable assurance that the safety functions of these plants 7

will be maintained during the period of extended operation.

8 Q. It that the end of your supplementary testimony today?

9 A. Yes it is. However, I reserve the right to supplement 10 my testimony if new information is disclosed or introduced.

11

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 45 UNITED STATES 1

NUCLEAR REGULATORY COMMISSION 2

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 3

x 4

In re:

Docket Nos. 50-247-LR; 50-286-LR 5

License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 6

Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 7

Entergy Nuclear Indian Point 3, LLC, and 8

Entergy Nuclear Operations, Inc.

March 4, 2016 9

x 10 DECLARATION OF RICHARD T. LAHEY, JR.

11 I, Richard T. Lahey, Jr., do hereby declare under penalty 12 of perjury that my statements in the foregoing testimony and my 13 statement of professional qualifications are true and correct to 14 the best of my knowledge and belief.

15 Executed in Accord with 10 C.F.R. § 2.304(d) 16 17 18 Dr. Richard T. Lahey, Jr.

19 The Edward E. Hood Professor Emeritus of Engineering 20 Rensselaer Polytechnic Institute, Troy, NY 12180 21 (518)495-3884, laheyr@rpi.edu 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement UNITED STATES NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

x In re:

Docket Nos. 50-247-LR; 50-286-LR License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 Entergy Nuclear Indian Point 3, LLC, and Entergy Nuclear Operations, Inc.

March 4, 2016

x SUPPLEMENTAL WRITTEN TESTIMONY OF DR. RICHARD T. LAHEY, JR.

REGARDING CONTENTIONS NYS-25, NYS-26B/RK-TC-1B, AND NYS-38/RK-TC-5 NYS000590 (Public Version)

Submitted: March 4, 2016

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 i

Table of Contents Page I.

Introduction.......................................... 1 II.

The Inspection Response Plan and Flaw Acceptance Criteria.............................................. 4

1. The Inspection Response Plan...................... 6

2. The Flaw Acceptance Criteria for IP-2 and IP-3.... 7 III.

Westinghouses Baffle-Former Bolt Analysis........... 12 IV.

Entergys Flaw Acceptance Criteria & the Eason and Pathania Paper....................................... 29 V.

The IP-3 Inspection Report........................... 42 VI.

Conclusion........................................... 44

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 ii List of Proposed New Exhibits Proposed Exhibit No.

Document NYS000583 Indian Point Units 2 and 3 Inspection Response Plan for Aging Management of MRP-227-A Primary and Expansion Components (WCAP-17941-P, Rev. 1) (IPECPROP00085900),

January 12, 2016 Short title:

Inspection Response Plan NYS000584 Background and Technical Basis Supporting Engineering Flaw Acceptance Criteria for Indian Point Unit-2 Reactor Vessel Internals MRP-227-A Primary and Expansion Components (WCAP-17949-P, Rev. 0)

(IPECPROP00085359), January 12, 2016 Short title:

Flaw Acceptance Criteria for IP-2 NYS000585 Background and Technical Basis Supporting Engineering Flaw Acceptance Criteria for Indian Point Unit-3 Reactor Vessel Internals MRP-227-A Primary and Expansion Components (WCAP-17951-P, Rev. 0)

(IPECPROP000856629), January 12, 2016 Short title:

Flaw Acceptance Criteria for IP-3 NYS000586 Determination of Acceptable Baffle-Former Bolting for Indian Point Units 2 and 3 (WCAP-18048-P, Rev. 0) (IPECPROP00086434),

January 12, 2016 Short title:

Bolting Analysis

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 iii Proposed Exhibit No.

Document NYS000587 E. Eason and R. Pathania, Disposition Curves for Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in Light Water Reactor Environments, ASME PVP2015-4532, Proceedings of the ASME 2015 Pressure Vessels and Piping Conference, July 19-23, 2015 Short title:

Eason and Pathania Paper NYS000588 USNRC Indian Point Nuclear Generating Unit License Renewal Inspection Report 05000286/2015011, Nov. 19, 2015 (ML15323A026)

Short title:

IP-3 Inspection Report

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 iv List of Acronyms Acronym Meaning AMP aging management plan ASLB, the Board Atomic Safety Licensing Board BWR boiling water nuclear reactor CGR crack growth rate DBA design basis accident EPRI Electric Power Research Institute FMEA failure modes and effects analysis IASCC irradiation-induced stress corrosion cracking IP-2 Indian Point nuclear facility Unit-2 IP-3 Indian Point nuclear facility Unit-3 LBB leak-before-break LOCA loss-of-coolant accident LRA license renewal application NYS, the State the State of New York PEO period of extended operation PWR pressurized water nuclear reactor RTA real-time analysis RVI reactor vessel internal SSE safe-shutdown earthquake UFSAR Updated Final Safety Analysis Report

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 1

I.

Introduction 1

On behalf of the State of New York (NYS or the State),

2 the Office of the Attorney General hereby submits the following 3

testimony by RICHARD T. LAHEY, JR., Ph.D., regarding Contentions 4

NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5.

5 Q.

Please state your full name.

6 A.

Richard T. Lahey, Jr.

7 Q.

Dr. Lahey, you have previously provided your 8

educational and professional qualifications and submitted 9

testimony in this proceeding, correct?

10 A.

Yes. I have previously submitted testimony and reports 11 in this proceeding. My education and professional qualifications 12 and experience are described in my previously submitted 13 Curricula Vitae and were summarized in my previous testimony.

14 Q.

I show you proposed exhibits NYS000583 through 15 NYS000588. Do you recognize those documents?

16 A.

Yes. These documents consist of four technical reports 17 authored by Westinghouse on behalf of Entergy, one technical 18 paper authored by E. Eason and R. Pathania, and one USNRC 19 inspection report for Indian Point Unit-3.

20 Q. Have you reviewed proposed exhibits NYS000583 through 21 NYS000588?

22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 2

A. Yes.

1 Q. Have these documents caused you to change the 2

testimony and opinions that you have previously submitted in 3

this proceeding in connection with Contentions NYS-25, NYS-4 26B/RK-TC-1B and NYS-38/RK-TC-5?

5 A. No. In fact, I believe that these documents support 6

the concerns, opinions, and testimony that I have presented 7

previously in this proceeding. It is my opinion that the 8

documents are relevant to the technical and legal matters at 9

issue in this proceeding, and that they provide substantial 10 support for the opinions that I have presented to the Atomic 11 Safety and Licensing Board (the Board) in my testimony and 12 written reports in this proceeding. In my opinion, these 13 documents provide further bases for the Board to conclude that 14 Entergys license renewal application (LRA) for the Indian 15 Point facilities fails to adequately manage the effects of aging 16 degradation such as embrittlement, fatigue, and irradiation 17 assisted stress corrosion cracking (IASCC) on reactor vessel 18 internal (RVI) systems, structures and components. They also 19 describe the facilities approach to fatigue analysis and 20 confirm that certain aging management program plan (AMP) 21

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 3

details will be deferred until well into the period of extended 1

operation (PEO).

2 Q. In your opinion, what is the significance of the 3

proposed new exhibits?

4 A.

The documents call into question whether Entergys 5

currently proposed amended and revised AMP for RVIs and the 6

associated Inspection Plan for the Indian Point nuclear 7

facilities, e.g., exhibits NYS000496-NYS000507 (Amended and 8

Revised RVI AMP), provide reasonable assurance against the 9

failure of various RVI systems, structures, and components. The 10 documents also suggest that Westinghouses approach to 11 developing flaw acceptance criteria and the associated 12 inspection intervals under Entergys Amended and Revised RVI AMP 13 may be inadequate and non-conservative. For example, 14 15 16 17 18 19

. In my testimony below, I will 20 begin by describing how three of the proposed new exhibits 21 provide specific details on the implementation of Entergys 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 4

Amended and Revised RVI AMP and Inspection Plan. Next, I will 1

discuss my concern regarding Westinghouses plant-specific 2

bolting pattern analysis for the baffle-former bolts at Indian 3

Point Unit-2 (IP-2) and Unit-3 (IP-3). Thereafter, I will 4

describe my concern that Entergys Amended and Revised RVI AMP 5

contains several non-conservative assumptions that may cause 6

IASCC cracks to grow faster at IP-2 and IP-3 than would be 7

accounted for by Entergy under its Amended and Revised RVI AMP.

8 Finally, I discuss USNRCs recent inspection report for IP-3.

9 II.

The Inspection Response Plan and Flaw Acceptance Criteria 10 Q. I show you three of the proposed new exhibits, each of 11 which relates to Entergys Amended and Revised RVI AMP. The 12 documents are:

13

Indian Point Units 2 and 3 Inspection Response Plan 14 for Aging Management of MRP-227-A Primary and Expansion 15 Components, a document that has been marked as proposed exhibit 16 NYS000583, and identified by Westinghouse as WCAP-17941-P, Rev.

17 1 (Inspection Response Plan);

18

Background and Technical Basis Supporting Engineering 19 Flaw Acceptance Criteria for Indian Point Unit-2 Reactor Vessel 20 Internals MRP-227-A Primary and Expansion Components, a 21 document that has been marked as proposed exhibit NYS000584, and 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 5

identified by Westinghouse as WCAP-17949-P, Rev. 0 (Flaw 1

Acceptance Criteria for IP-2); and, 2

Background and Technical Basis Supporting Engineering 3

Flaw Acceptance Criteria for Indian Point Unit-3 Reactor Vessel 4

Internals MRP-227-A Primary and Expansion Components, a 5

document that has been marked as proposed exhibit NYS000585, and 6

identified by Westinghouse as WCAP-17951-P, Rev. 0 (Flaw 7

Acceptance Criteria for IP-3).

8 Please describe how the Inspection Response Plan 9

(NYS000583) and Flaw Acceptance Criteria for IP-2 (NYS000584) 10 and IP-3 (NYS000585) each relate to the implementation of 11 Entergys Amended and Revised RVI AMP?

12 A. According to the Inspection Response Plan and the Flaw 13 Acceptance Criteria for IP-2 and IP-3, these documents, 14 15 16 17 18 19

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 6

1.

The Inspection Response Plan 1

Q. I show you the Inspection Response Plan (NYS000583).

2 Would you please explain your understanding of how this document 3

relates to Entergys Amended and Revised RVI AMP?

4 A.

Yes.

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 7

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16

2.

The Flaw Acceptance Criteria for IP-2 and IP-3 17 Q.

I now show you two documents, the Flaw Acceptance 18 Criteria for IP-2 (NYS000584), and the Flaw Acceptance Criteria 19 for IP-3 (NYS000585). Would you explain your understanding of 20 how these documents relate to Entergys Amended and Revised RVI 21 AMP?

22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 8

A.

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22

Contains Westinghouse and Entergy Designated Proprietary Information Subject to Nondisclosure Agreement Supplemental Written Testimony of Richard T. Lahey, Jr.

Contentions NYS-25, NYS-26B/RK-TC-1B, and NYS-38/RK-TC-5 10

and Entergy 1