Text

Official Exhibit - NYS000532-PUB-00-BD01 - Pre-filled Written Supplemental Testimony of Dr. David J. Duquette Regarding Contention NYS-38/RK-TC-5
	ML15334A218
Person / Time
Site:	Indian Point
Issue date:	06/08/2015
From:	; State of NY
To:	; Atomic Safety and Licensing Board Panel
	SECY RAS
References
	50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01, RAS 27936
	Download: ML15334A218 (30)
	v • d • e

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.

June 8, 2015 9

x 10 PRE-FILED WRITTEN SUPPLEMENTAL TESTIMONY OF 11 DR. DAVID J. DUQUETTE 12 REGARDING CONTENTION NYS-38 / RK-TC-5 13 On behalf of the State of New York (NYS or the State),

14 the Office of the Attorney General hereby submits the following 15 rebuttal testimony by David J. Duquette, Ph.D. regarding 16 Contention NYS-38/RK-TC-5.

17 Q.

Please state your full name.

18 A.

David J. Duquette.

19 Q.

What is the purpose of this testimony you are now 20 providing?

21 A.

This testimony supplements my initial and rebuttal 22 testimony on Contention NYS-38/RK-TC-5. It has been 23 1

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 #:

Identified:

Admitted:

Withdrawn:

Rejected:

Stricken:

Other:

NYS000532-PUB-00-BD01 11/5/2015 11/5/2015 NYS000532 Submitted: June 9, 2015

approximately three years since I provided my initial pre-filed 1

testimony in this matter and two and a half years since I 2

provided rebuttal testimony. The State of New York has asked me 3

to review the record on Contention NYS-38/RK-TC-5 and respond to 4

recent information and events.

5 Q.

What documents did you review in preparation for this 6

supplemental testimony?

7 A.

I reviewed again Entergys August 20, 2012 Statement 8

of Position Regarding Contention NYS-38/RK-TC-5 (ENT000520),

9 Entergys Pre-filed Testimony of Entergy witnesses Nelson 10 Azevedo, Robert Dolansky, Alan Cox, Jack Strosnider, Robert 11 Nickel, Ph.D., and Mark Gray regarding Contention NYS-38/RK-TC-5 12 (ENT000521), and the accompanying exhibits. I also reviewed the 13 NRC Staffs August 20, 2012 Statement of Position on Contention 14 NYS-38/RK-TC-5 (NRC000147), NRCs Pre-filed Testimony of NRC 15 Witnesses Dr. Allen Hiser and Kenneth Karwoski Concerning 16 Portions of Contention NYS-38/RK-TC-5 (NRC000161), which focuses 17 on steam generator issues, and the accompanying exhibits.1 18 1 NRC Staff also submitted pre-filed testimony on another aspect of Contention NYS-38/RK-TC-5, namely NRC000148. That testimony focused on metal fatigue issues and did not discuss my June 2012 testimony or report on steam generator issues. Accordingly, my testimony here does not discuss NRC000148.

2

In addition, I also re-reviewed documents previously 1

submitted by the State on this contention including my previous 2

pre-filed testimony and report (NYS000372, NYS000373, NYS000452) 3 and exhibits (including, without limitation, NYS000375 to 4

NYS000394 and NYS000454 to NYS000463, NYS000472, NYS000146, 5

NYS000147, NYS000160). These documents include a presentation 6

from the EPRI Steam Generator Task Force (SGTF) to the NRC 7

entitled NRC/EPRI Steam Generator Task Force Meeting, dated 8

August 21, 2012 (NYS000463), an NRC chart identifying original 9

and replacement steam generators at U.S. plants prepared in 2009 10 (NYS000458), a paper numbered ICONE18-29457 entitled Inspection 11 of the Steam Generator Divider Plate, presented at the 18th 12 International Conference on Nuclear Engineering, authored by D.

13 DAnnucci and E. Lecour of Westinghouse for the May 2010 ICONE 14 meeting (ENT000526), EPRI Report 1025133, Steam Generator 15 Management Program: Assessment of Channel Head Susceptibility to 16 Primary Water Stress Corrosion Cracking, dated June 2012 17 (ENT000524), and various summary or demonstrative exhibits 18 prepared by the State (NYS000454 to NYS000456).

19 In addition, I reviewed a summary chart identifying the 20 materials used in the eight steam generators at Indian Point 21 Unit 2 and Unit 3 (NYS000560), a 2014 EPRI report of cracking in 22 3

steam generator channel head assemblies (NYS000544A-D), a 2012 1

Westinghouse Nuclear Safety Advisory Letter (NYS000549); various 2

NRC/EPRI Steam Generator Task Force presentations (NYS000546 and 3

NYS000550); steam generator tube inspection reports (NYS000543 4

and NYS000537); integrated inspection reports (NYS000536 and 5

NYS541); an in-service inspection summary (NYS000540); steam 6

management program documents (NYS000533, NYS000534, NYS000554, 7

NYS000555); commitment closure and verification forms (NYS000535 8

and NYS000553); NRC information notices and reports (NYS000551 9

and NYS000538); license amendment requests and approval letters 10 and related documents (NYS000539, NYS000542, NYS000556, and 11 NYS000547); responses to NRC requests for information 12 (NYS000545); and an NRC report on Lessons learned from San 13 Onofre (NYS000552).

14 Q.

What are your overall conclusions having reviewed that 15 information?

16 A.

First, I disagree with Entergy and the NRC staffs 17 suggestion set out in their testimony that divider plate 18 cracking is unlikely to occur in the future because it has not 19 been observed to date in United States-based steam generators.

20 I likewise disagree with Entergy and NRC staff's position that 21 Entergys general approach to aging management issues will 22 4

effectively provide adequate safety measures if cracking were to 1

occur. Entergys testimony reflects a trust us approach in 2

the absence of real data on the condition of the eight Indian 3

Point steam generators. Second, it is my opinion that in order 4

to adequately address aging degradation in the Indian Point 5

steam generators Entergy must unequivocally commit to and 6

establish a sufficiently detailed aging management program that 7

includes baseline and follow-up inspections of the steam 8

generator channel head and divider plate assemblies, including 9

the tube-to-tubesheet welds.

10 As discussed in my 2012 testimony, as well as that of 11 Entergy and NRC Staff, the EPRI-sponsored Steam Generator Task 12 Force is conducting an extensive research program into the 13 propagation of cracks in the divider plate assembly. I 14 understand that in October 2014, EPRI 15 16 17

. I have 18 reviewed the report, and it does not change my view that Entergy 19 must address potential primary water stress corrosion cracking 20 and fatigue cracking in the eight steam generators at Indian 21 Point before relicensing occurs. Thus, it is still my opinion 22 5

that inspections of the steam generator channel head and divider 1

plate assemblies and tube-to-tubesheet welds should be conducted 2

before Indian Point Unit 3 begins its period of extended 3

operation, and that such inspections should be conducted 4

promptly at Indian Point Unit 2, since they have not yet been 5

conducted at that facility.

6 In addition, while no industry-qualified technique for 7

inspection of the lower channel head and divider plate assembly 8

currently exists in the United States, any license renewal given 9

to Entergy for the Indian Point facilities should be contingent 10 on the companys expeditious qualification of an inspection 11 technique capable of identifying and evaluating primary water 12 stress corrosion cracking and fatigue-related cracks. Entergy 13 has identified a remote inspection technique that relies on 14 ultrasonic, visual and liquid penetrant technologies developed 15 by Westinghouse that has been used to successfully inspect 16 divider plates in French steam generators. ICONE Westinghouse 17 Paper (ENT000526). Instead of relying on the current absence of 18 a U.S. industry-qualified inspection technique as an excuse to 19 delay inspections at Indian Point 2 and Indian Point 3, Entergy 20 should conduct the necessary inspections using techniques 21 available now for detecting and evaluating cracks in the lower 22 6

channel assembly. For example, Entergy can employ the 1

Westinghouse technique pending future industry qualification, or 2

some other similarly effective technique. I note that remote 3

visual and ultrasonic inspections were used to inspect for 4

possible flaws in the tubesheet to channel head transition 5

region in Westinghouse steam generators at Wolf Creek Generating 6

Station and Surry Power Station Unit 2. NRC Information Notice 7

13-20 (NYS000538).

8 Q.

Why is it important that Entergy inspect the lower 9

assemblies and the tube-to-tubesheet welds of the Indian Point 10 steam generators?

11 A.

Both Entergy and the NRC staff agree that the Indian 12 Point Unit 2 and Indian Point Unit 3 steam generators have 13 divider plates that are constructed from Alloy 600 and that the 14 weld materials are also an Alloy 600 derivative (Alloy 82/182).

15 It is well known that Alloy 600 is susceptible to PWSCC. As of 16 mid-2015, the four steam generators at Indian Point Unit 2 have 17 been in use for approximately 15 years. They were installed 18 following the steam generator accident at Unit 2 in 2000. The 19 four steam generators at Indian Point Unit 3 have been in use 20 for approximately 26 years. They were installed in 1989 to 21 7

replace the original Unit 3 steam generators, which had been in 1

use for approximately 14 years at that time.

2 However, the current state of the divider plates, the stub 3

runners, the channel heads, as well as the tube-to-tubesheet 4

welds at Indian Point is largely unknown. Over the past few 5

years, based on reports of cracking in divider plate assemblies 6

in French steam generators, EPRIs SGTF has been examining the 7

susceptibility of divider plate assemblies to PWSCC and 8

investigating the possibility that stress corrosion cracking or 9

fatigue induced cracks could propagate into the pressure 10 boundary components. Entergy has stated that its approach to 11 the divider plate assembly cracking problems is not dependent on 12 the results of EPRI research, but that inspections being 13 committed to by plants with renewed licenses will occur at an 14 appropriate time, and that the Indian Point Quality Assurance 15 Program will drive appropriate safety evaluations. Without 16 specific criteria for determining appropriateness, Entergys 17 plan remains a hollow assurance that aging degradation of its 18 steam generators will be adequately managed.

19 In my June 2012 report, I pointed out that EPRI has 20 generically stated that the divider plates in United States 21 steam generators are thicker than those that have experienced 22 8

cracking in French steam generators, and that that factor alone 1

may mitigate against PWSCC initiation in United States steam 2

generators. Even if that conclusion proved to be true for some 3

or most United States steam generators, the divider plates at 4

Indian Point Unit 2 and Unit 3 are an exception to this general 5

rule. While the majority of steam generators in the United 6

States have divider plate thicknesses of approximately 1.9 7

inches, the Westinghouse Model 44F steam generators at Indian 8

Point Unit 2 and Unit 3 have plate thicknesses of 1.26 inches, 9

essentially the equivalent of the 1.3 inch thick divider plates 10 used in the French steam generators where PWSCC cracking was 11 first discovered. Thus, barring the possibility of differences 12 in loading or pre-assembly processing of the divider plates and 13 associated assemblies, the steam generators at Indian Point have 14 essentially the same sensitivity to PWSCC as the French steam 15 generators.

16 In my initial June 2012 testimony in this proceeding I 17 referred to cracking that had occurred in the steam generator at 18 Indian Point Unit 2. I agree with Entergy that replacement of 19 mill annealed Alloy 600 tubing with thermally treated Alloy 600 20 tubing may reduce (but not eliminate) the potential for PWSCC in 21 9

steam generator tubes.2 However, no evidence has been presented 1

that the divider plate assemblies are constructed from thermally 2

treated alloys. Even if they are, the geometry of cracking that 3

has been observed in the European steam generators has occurred 4

near the welds joining the divider plates to the stub runners.

5 Welding of these components can be expected to lead to 6

dissolution of the grain boundary precipitates that are believed 7

to provide a degree of PWSCC resistance in thermally treated 8

alloys. Accordingly, the Entergy comments concerning the lack 9

of cracking in the steam generator Alloy 600TT tubes has little 10 or no relevance to the possibility of PWSCC in the divider 11 plates or stub runners - or for that matter in the tube-to-12 tubesheet welds.

13 Q.

I show you what has been marked as Exhibit NYS000549.

14 Are you familiar with this document?

15 A.

Yes. It is a 16 17 18 were designated proprietary 19 2 Often times the abbreviation TT is used to designate thermally treated components, e.g., Alloy 600 TT tubes.

10

by Westinghouse; Entergy subsequently disclosed these documents 1

to the State in this proceeding.

2 Q.

What is the purpose of the 3

?

4 A.

5 6

7 8

9 10 11 12 13 It was later 14 disclosed that these conditions were identified at a foreign 15 plant.

16 17 18 19 20 21 11

1 2

Q.

What recommendations, if any, did 3

4 5

A.

6 7

8 9

10 11 12 13 14 Q.

Has Entergy performed the inspections recommended by 15 16 A.

Yes. I understand, based on Entergys 2013 and 2014 17 Integrated Inspection Reports for IP2 and IP3 (NYS000536 and 18 NYS000541), that Entergy performed remote video camera 19 inspections of the lower channel head and divider plate to 20 channel head welds for six of the eight Indian Point steam 21 12

generators (21, 22, 24, 31, 33, 34) following 1

2 Q.

Did the NRC take any action to follow up on 3

4 A.

In October 2013, the NRC issued Information Notice 5

2013-20 entitled, SG Channel Head and Tubesheet Degradation 6

(NYS000538) which addressed issues of potential corrosion and 7

degradation in channel heads and tubesheets.

8 Q.

Directing your attention to Exhibit NYS000545A-D, do 9

you recognize that document?

10 A.

Yes. It is a copy of 11 12 13 14 15 provided the report 16 to Entergy and possibly other reactor operators. As I noted 17 earlier, EPRI designated the document as containing proprietary 18 information.

19 Q.

Does the EPRI report resolve your concerns about the 20 Indian Point steam generators?

21 A.

No, it does not.

22 13

Q.

Why is that?

1 A.

There are several reasons. To begin with, 2

3 4

5 6

. This is a serious omission, since the steam 7

generators at IP3, installed in 1989, will be operating beyond 8

their 40 year life span towards the end of IP3s period of 9

extended operation. Cracks can experience exponential growth 10 rates in cyclically stressed materials. For example, a small 11 crack that develops during the first 25 years of an IP3 steam 12 generators life may rapidly develop into a crack that 13 compromises the integrity of a reactor pressure boundary or 14 other safety related component before the renewed licensing 15 period ends.

does not provide any assurance 16 whatsoever that this scenario would not occur.

17 In addition, it appears that 18 analysis may be non-conservative because it did not take into 19 account the specific environmental conditions within the Indian 20 Point steam generators, such as high temperatures and 21 corrosivity, which are widely known to accelerate crack growth.

22 14

Any conclusions in the report based on this analysis would 1

therefore have little to no relevance to the issue of crack 2

growth in the Indian Point steam generators.

3 Furthermore, components made 4

of thermally-treated Alloy 690 (Alloy 690TT), which is more 5

PWSCC resistant than thermally-treated Alloy 600 (Alloy 600TT).

6 Since the IP2 steam generators tubes, tube-to-tube sheet welds 7

and divider plate assembly components are composed of Alloy 8

600TT, the report findings are simply inapplicable to IP2. I 9

have concerns about the condition of IP2s steam generators 10 precisely because these components are constructed of materials 11 known to be susceptible to PWSCC.

12 I am also concerned about PWSCC in Alloy 600TT components 13 and parts in IP3 steam generators. Although the tubes at IP3 14 steam generators are constructed of Alloy 690TT, the divider 15 plate assemblies are conservatively assumed to be Alloy 600TT.

16 Thus, the 17 18 19 20 Q.

?

21 15

A.

1 2

3 4

5 6

Q.

7 8

9 A.

Yes, it did.

10 11 12 13 14 Q.

Do you wish to comment on that?

15 A.

Yes. First, I want to point out that Entergy has not 16 confirmed that the steam generators at Indian Point do not have 17 a layer of cold-work potentially susceptible to cracking. There 18 is some evidence that the tube-to-tubesheet welds in IP2 have 19 been cold-worked. For example, Westinghouses Alternative 20 Repair Criteria Analysis (WCAP-17828-NP) at pp. 2-9 and 3-7 21 (NYS000547) describes the fabrication and material properties of 22 16

the tube and tubesheet welds and states that [t]he 1

manufacturing process used to assemble a steam generator creates 2

a strain-hardened condition in the tubes. These tubes are then 3

inserted into the tubesheet bores and tack-expanded by hydraulic 4

expansion or mechanical hard rolling before being welded to the 5

tubesheet. Therefore, any cold-worked surfaces of the steam 6

generators could be vulnerable to the same conditions 7

experienced by the European reactors.

8 Moreover, I understand that the French operating experience 9

differs in various ways from the U.S. operating experience which 10 may account for slower crack growth rates observed in these 11 foreign plants. My experience with presentations by Electricite 12 de France (EdF), the operator of the steam generators in which 13 cracking of the divider plate assembly was initially observed, 14 is that, when a reactor in France encounters a limiting problem 15 with a steam generator tube, the French typically de-rate the 16 generator, meaning that they reduce the power of the system. In 17 contrast, U.S. nuclear system operators typically plug a tube, 18 meaning that the tube is taken out of service by blocking the 19 entry and exit openings, but do not reduce the power rating.

20 This means that, all other things being equal, U.S. pressurized 21 water reactor steam generators may run hotter and be subject to 22 17

greater stresses than their French counterparts. This 1

difference in operating environments can affect steam generator 2

susceptibility to PWSCC, as well as the growth rate of any 3

cracks that develop. At Indian Point, steam generators with a 4

number of plugged tubes may be more susceptible to PWSCC and 5

fatigue induced cracking than steam generators at French 6

reactors. Thus, while the French experience helped alert 7

industry and government to the potential for divider plate 8

assembly cracking under normal operating conditions in those 9

plants, the lack of significant crack growth observed at the 10 French reactors since the cracks were first reported should not 11 be interpreted to suggest that any cracks found in a U.S. plant 12 today would not propagate.

13 Q.

14 15 16 A.

17 18 19 20 21 22 18

1 2

3 I believe any decision to abandon inspection 4

plans is misguided. As I stated 5

earlier, regular inspections provide licensees and the NRC an 6

opportunity to gather baseline data for benchmarking objective 7

evidence of degradation and are a critical part of ensuring that 8

systems operate safely. From an engineering perspective, it 9

would be irresponsible to rely exclusively on mathematical 10 modeling data, particularly since we have seen, in both the 11 fracture toughness context (i.e., recently identified non-12 conservatism of BTP-5-3)(NYS000518-NYS000519) and the San Onofre 13 steam generator tube rupture context (NRC Review of Lessons 14 Learned at San Onofre, March 2014)(NYS000552), that models can 15 be non-conservative, unreliable or just plain wrong.

16 Q.

To your knowledge, has Entergy inspected the divider 17 plate assemblies and tube-to-tubesheet welds of the Indian Point 18 steam generators?

19 A.

While Entergy has performed remote video inspections 20 of the channel heads and divider plate-to-channel head welds for 21 cladding degradation and PWSCC based on Westinghouses NSAL 12-22 19

1, it appears that inspections were performed for only six of 1

the eight steam generators at Indian Point. Moreover, those 2

inspections were limited in scope and did not employ techniques 3

qualified to detect and measure cracks or flaws due to PWSCC.

4 NRC Integrated Inspection Report, May 9, 2014 at 10 (NYS000541).

5 Indeed, I do not believe Entergy used any magnification as a 6

part of its NSAL 12-1 channel head inspection, as its focus was 7

to identify NYS000549). It 8

is my opinion that and the operating experiences at 9

Wolf Creek and Surry referenced in the NRCs Information Notice 10 13-20 (NYS000538) suggest that failure of corrosion-resistant 11 cladding in steam generators like those in use at Indian Point 12 is a potential problem requiring detailed inspection and 13 monitoring. Given the limited information available regarding 14 the current condition of the lower channel head assembly areas 15 of the eight steam generators at Indian Point, Entergy should, 16 as soon as possible, perform an initial baseline inspection of 17 IP2 and IP3 steam generator divider plate and channel head 18 assemblies and tube-to-tubesheet welds as part of the companys 19 One Time Inspection Program in order to confirm that its water 20 chemistry program is in fact effective and that primary water 21 stress corrosion cracking is not occurring. Generic Aging 22 20

Lessons Learned, Rev. 2 (2010), IV D 1-3,8. Similar to Entergys 1

In-Service Inspections, subsequent inspections of these steam 2

generator locations should be performed at least once every 10 3

years.

4 5

underscores the vulnerability of these 6

steam generators to corrosion and cracking, and the need for 7

regular inspections to maintain safe operations.

8 Finally, recent documents report that Indian Points steam 9

generators have experienced age-related degradation as a result 10 of wear associated with steam generator tube vibration, and that 11 a number of tubes have been plugged and taken out of service as 12 a result. (IP2 Steam Generator Examination Program Results 2014 13 Refueling Outage (2R21)(September 8, 2014)(NYS000543). I am 14 concerned about the numerous indications of vibration-induced 15 wear in the steam generator tubes at IP2, as documented in the 16 plants most recent tube inspection report. During the last 17 outage, Entergy plugged five tubes due to wear. We learned from 18 the San Onofre steam generator tube rupture event that wear, in 19 that case caused by fluid-elastic instability, can quickly 20 progress from flaw or crack initiation to tube failure. Unlike 21 other, longer-acting degradation mechanisms that may be 22 21

identified before they progress to a critical stage, wear can 1

under certain circumstances rapidly progress between inspection 2

intervals.

3 I also note that foreign objects were identified during 4

Entergys steam generator tube inspections. During the most 5

recent inspection, Entergy plugged at least nine tubes due to 6

foreign objects trapped inside the tubes. Foreign objects in 7

the steam generator can cause dents and dings. For example, in 8

1990, only one year after Steam Generator 34 was installed at 9

IP3, a fuel alignment pin was found partially lodged in a tube 10 end in the generator. 2007 Indian Point 3 Steam Generator 11 Program (NYS000533) at p. 13, 14. Visual examination revealed 12 that the foreign object made numerous indentations on the 13 channel head surfaces. Follow up inspections indicated that 14 impacts from loose parts resulted in deformities of some tube 15 ends. The presence of foreign objects in the Indian Point steam 16 generators and their potential to cause damage to the reactor 17 coolant pressure boundary is an important concern. According to 18 the NRCs Information Notice 2013-11 (NYS000551), cracking in 19 dented or dinged regions of Alloy 600TT tubing has been 20 reported, and this operating experience highlights the 21 importance of, and the challenges to, inspecting locations 22 22

susceptible to degradation and identifying inspection methods 1

capable of detecting that degradation. It is therefore 2

imperative that Entergy remain vigilant in its inspections of 3

the steam generator tubes, tube-to-tubesheet welds, and divider 4

plate and channel head assemblies at IP2 and IP3.

5 Q.

Can you describe Entergys proposed inspection and 6

aging management program for the lower assembly area and tube-7 to-tubesheet welds in the steam generators?

8 A.

It is difficult to tell exactly what Entergy has 9

unequivocally committed to do. As Ive discussed, in 2011, 10 Entergy presented two commitments regarding the steam 11 generators, Commitment 41 and Commitment 42. These commitments 12 are set out in Appendix A of the NRC Staffs 2011 Supplemental 13 Safety Evaluation Report (NYS000160), at pages A-23 and A-24.

14 Q.

Can you read Commitment 41?

15 A.

Commitment 41 states that, IPEC will inspect steam 16 generators for both units to assess the condition of the divider 17 plate assembly. The examination technique used will be capable 18 of detecting PWSCC in the steam generator divider plate 19 assembly. The IP2 steam generator divider plate inspections 20 will be completed within the first ten years of the period of 21 extended operation (PEO). The IP3 steam generator divider plate 22 23

inspections will be completed within the first refueling outage 1

following the beginning of the PEO.

2 Q.

What is the implementation schedule for Commitment 41?

3 A.

For IP2, it is after the beginning of the PEO and 4

prior to September 28, 2023. For IP3, it is prior to the end 5

of the first refueling outage following the beginning of the 6

PEO, which I understand to be around March or April 2017.

7 Q.

Can you please read Commitment 42?

8 A.

Commitment 42 provides that IPEC will develop a plan 9

for each unit to address the potential for cracking of the 10 primary to secondary pressure boundary due to PWSCC of tube-to-11 tubesheet welds using one of the following two options.

12 Q.

What is Option 1?

13 A.

Option 1, which is also referred to as the analysis 14 option, states that IPEC will perform an analytical evaluation 15 of the steam generator tube-to-tubesheet welds in order to 16 establish a technical basis for either determining that the 17 tubesheet cladding and welds are not susceptible to PWSCC, or 18 redefining the pressure boundary in which the tube-to-tubesheet 19 weld is no longer included and, therefore, is not required for 20 reactor coolant pressure boundary function. The redefinition of 21 24

the reactor coolant pressure boundary must be approved by the 1

NRC as a license amendment request.

2 Q.

What is the implementation schedule for Option 1?

3 A.

For IP2, implementation is prior to March 2024, and 4

for IP3, prior to the end of the first refueling outage 5

following the beginning of the PEO.

6 Q.

What is Option 2?

7 A.

Option 2, which is also referred to as the 8

inspection option, provides that IPEC will perform a one-time 9

inspection of a representative number of tube-to-tubesheet welds 10 in each steam generator to determine if PWSCC cracking is 11 present. If weld cracking is identified:

12

a. The condition will be resolved through repair or 13 engineering evaluation to justify continued service, as 14 appropriate, and 15

b. An ongoing monitoring program will be established to 16 perform routine tube-to-tubesheet weld inspections for the 17 remaining life of the steam generators.

18 Q.

What is the implementation schedule for Option 2?

19 A.

For IP2, the implementation schedule is between March 20 2020 and March 2024, and for IP3, prior to the end of the 21 25

first refueling outages following the beginning of the PEO, 1

which again, I understand to be around March or April of 2017.

2 Q.

Can you summarize what Entergy has agreed to do under 3

those Commitments?

4 Q.

Under Commitment 41, Entergy committed to inspect and 5

assess the condition of the divider plate assemblies in the IP2 6

and IP3 steam generators. Under Commitment 42, Entergy 7

committed to either perform an analytical evaluation or an 8

inspection of the tube-to-tubesheet welds.

9 Q.

What is the status of those Commitments today?

10 A.

I understand that on September 5, 2014, NRC staff 11 approved an amendment to Entergys operating license for Indian 12 Point Unit 2 so as to redefine the reactor coolant pressure 13 boundary to exclude tube-to-tubesheet welds (Amendment 277) and 14 thereby relieved Entergy of the obligation to inspect the tube-15 to-tubesheet welds. (Technical Specification Amendment 16 277)(NYS000542). As a result of that license amendment, on 17 September 17, 2014 Entergy deemed its Commitment 42 complete 18 for IP2. Commitment Closure Verification Form/ Corrective 19 Action (LR-LAR-2011-00174)(NYS000553). Based on the data 20 available today, I believe the NRC Staff was premature in 21 granting Amendment 277. The NRC and the nuclear industrys 22 26

understanding of PWSCC in the steam generator environment 1

continues to evolve. In fact, the NRC recently committed over 2

$2.3 million to fund research at Pacific Northwest National 3

Laboratories for the purpose of evaluating PWSCC in nickel-based 4

alloys used in steam generator and reactor components. NRC 5

Weekly Information Report, May 15 2015 (NYS000557).

6 For now, it appears that it is Entergys position that 7

Commitment 41 relating to the divider plate assembly inspections 8

is still open for IP2 and IP3 (Commitment 41 Closure 9

Verification Form (NYS000535), but that Commitment 42 relating 10 to tube-to-tubesheet welds is open for IP3 only (NYS000553).

11 However, it is unclear what impact will have 12 on these remaining open commitments. As I noted earlier, the 13 14 15 16 17 Although Entergy disclosed it is not 18 clear what use, if any, Entergy has made, or will make, of the 19 document. In my opinion, Entergy should not -- and cannot --

20 rely on the to avoid inspecting the channel head and 21 divider plate assemblies, including the tube-to-tubesheet welds 22 27

in the eight Indian Point steam generators. To the extent that 1

Entergy remains committed to performing inspections after 2

license renewal, potentially well into the plants periods of 3

extended operation, that is an inadequate assurance for managing 4

aging steam generators at Indian Point. Rather, Entergy should 5

affirmatively and clearly commit to performing inspections as 6

soon as possible for IP2, and certainly before the period of 7

extended operation for IP3. Additionally, Entergy must identify 8

the inspection techniques it intends to use, develop acceptance 9

criteria, and provide a detailed plan for addressing any flaws 10 or indications that it may encounter. I also recommend that 11 Entergy conduct follow-up inspections at least every 10 years, 12 given the primarily Alloy 600TT construction of IP2 steam 13 generator components and assemblies and the age of the IP3 steam 14 generators.

15 In conclusion, from my perspective in 2011 and 2012 there 16 was substantial uncertainty about what pathway Entergy would 17 pursue with respect to steam generators; moreover, essential 18 details were lacking in the various optional pathways Entergy 19 identified. The recent EPRI Report and the operating license 20 amendment have not resolved these uncertainties and unknowns.

21 Q.

Does this conclude your supplemental testimony?

22 28

A.

Yes. However, I reserve the right to offer further 1

opinions if new information is presented.

2 3

29

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.

June 8, 2015 9

x 10 DECLARATION OF DAVID J. DUQUETTE 11 I, David J. Duquette, do hereby declare under penalty of 12 perjury that my statements in the foregoing rebuttal testimony 13 and my statement of professional qualifications are true and 14 correct to the best of my knowledge and belief.

15 Executed in Accord with 10 C.F.R. § 2.304(d) 16 17 David J. Duquette, Ph.D.

18 Materials Engineering Consulting Services 4 North Lane Loudonville, New York 12211 Tel: 518 276 6490 Fax: 518 462 1206 Email: duqued@rpi.edu June 8, 2015 19 30