ML15337A304

Official Exhibit - NYS000571-PUB-00-BD01 - Pre-filed Supplemental Reply Testimony of David J. Duquette in Support of Contention NYS-38/RK-TC-5 (Public, Redacted) (September 9, 2015)

ML15337A304
Person / Time
Site:	Indian Point
Issue date:	09/09/2015
From:	Duquette D State of NY, Office of the Attorney General
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 28274, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15337A304 (21)

Text

United States Nuclear Regulatory Commission Official Hearing Exhibit In the Matter of: Entergy Nuclear Operations, Inc. NYS000571 (Public, Redacted)

(Indian Point Nuclear Generating Units 2 and 3)

Submitted: September 9, 2015 ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #: NYS000571-PUB-00-BD01 Identified: 11/5/2015 Admitted: 11/5/2015 Withdrawn:

Rejected: Stricken:

Other:

1 UNITED STATES 2 NUCLEAR REGULATORY COMMISSION 3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4 -----------------------------------x 5 In re: Docket Nos. 50-247-LR; 50-286-LR 6 License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 7 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 8 Entergy Nuclear Indian Point 3, LLC, and 9 Entergy Nuclear Operations, Inc. September 9, 2015 10 -----------------------------------x 11 PRE-FILED WRITTEN SUPPLEMENTAL REPLY TESTIMONY OF 12 DR. DAVID J. DUQUETTE 13 REGARDING CONTENTION NYS-38 / RK-TC-5 14 On behalf of the State of New York (NYS or the State),

15 the Office of the Attorney General hereby submits the following 16 rebuttal testimony by David J. Duquette, Ph.D. regarding 17 Contention NYS-38/RK-TC-5.

18 Q. Please state your full name.

19 A. David J. Duquette.

20 Q. What is the purpose of this testimony you are now 21 providing?

22 A. I have previously provided testimony in this 23 proceeding regarding primary water stress corrosion cracking

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 (PWSCC) in steam generators and the need for baseline 2 inspections of the Indian Point steam generator divider plate 3 assemblies, tube-to-tubesheet welds, and channel head assemblies 4 prior to license renewal. (NYS000372, NYS000373, NYS000452, 5 NYS000532). This testimony supplements, and incorporates by 6 reference, my prior testimony in this proceeding.

7 Q. I show you what has been marked as Exhibit ENT000699.

8 Do you recognize that document?

9 A. Yes. It is a copy of the pre-filed testimony of the 10 witnesses for Entergy on Contention NYS-38/RK-TC-5 that were 11 submitted in August 2015.

12 Q. I show you what has been marked as Exhibit NRCR000161, 13 NRC000197 and NRC000168. Do you recognize those documents?

14 A. Yes. They are copies of the pre-filed testimony of 15 NRC Staff witnesses that were submitted in August 2015. They 16 concern Contention NYS-38/RK-TC-5. I note that NRC000168 and 17 NRC000197 primarily address Contentions NYS-25 and NYS-26B/RK-18 TC-1B.

19 Q. Have you had an opportunity to review ENT000699, 20 NRCR000161, NRC000168, and NRC000197?

21 A. Yes.

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 Q. Has Entergys and NRC Staffs August pre-filed 2 testimony caused you to change the testimony and opinions that 3 you have previously submitted in this proceeding in connection 4 with Contention NYS-38?

5 A. No. It is still my opinion that Entergy should 6 perform visual baseline inspections of the eight Indian Point 7 steam generators prior to license renewal, that is, before 8 Entergy receives renewal 20-year operating licenses for Indian 9 Point Unit 2 and Unit 3.

10 Q. Does Entergy currently perform inspections of the 11 steam generators?

12 A. It is my understanding that Entergy currently performs 13 periodic visual inspections of the steam generator bowls, 14 tubesheets, and plugs using remote camera techniques. Entergy 15 Testimony at A175 (ENT000699). Entergy could readily expand the 16 scope of those inspections to include the divider plate 17 assemblies and the tube-to-tubesheet welds.

18 Q. What is Entergys position with respect to performing 19 divider plate and tube-to-tubesheet weld inspections?

20 A. It is non-committal, at present. Although Entergy 21 committed to inspect for PWSCC in the IP2 and IP3 steam 22 generator divider plates as part of Commitment 41 (see, NL NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 074, Attach. 1 at 14 (NYS000152), its most recent testimony 2 confirms that the company is considering retraction of that commitment in light of 12 13 Entergy had also committed, via Commitment 42, to address 14 PWSCC in tube-to-tubesheet welds by either performing 15 inspections of those steam generator component locations, or by 16 demonstrating through analyses that they are not susceptible to 17 PWSCC or are not part of the reactor coolant boundary. NL 18 032, Attach. 1 at 22-23 (NYS000151); NL-11-074, Attach. 2 at 15 19 (NYS000152).

20 Q. What is the current status of Commitment 42?

21 A. In late 2014, Entergy redefined the reactor coolant 22 pressure boundary so as to exclude tube-to-tubesheet weldments

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 at the IP2 steam generators. NYS000542. Apparently, Entergy 2 considers Commitment 42 fulfilled as to IP2. NYS000553.

3 Q. How is Entergy proposing to fulfill Commitment 42 for 4 IP3 steam generator tube-to-tubesheet welds?

5 A. Entergy has indicated that it prefers the analysis 6 option over the inspection option for IP3, and is currently evaluating whether 9 Entergy Testimony at A166 (ENT000699).

10 Q. To your knowledge, have the IP2 and IP3 steam 11 generator divider plates and tube-to-tubesheet welds ever been 12 inspected for PWSCC?

13 A. No, I do not believe they have ever been inspected. In 14 its testimony, Entergy does not mention any previous inspections 15 of these components and locations. Moreover, given Entergys 16 support for and the companys stated 17 preference for analysis rather than inspection, those 18 components/locations may never be inspected.

19 Q. Do you have any concerns about Entergys reliance on 20 as a basis to retract Commitment 41 and/or 21 close Commitment 42?

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 A. As I have previously indicated, I do not believe the 2 results of EPRIs investigations into steam generator component 3 cracking eliminate the need for actual inspections at Indian Point. To the contrary, 17 Entergy Testimony at A188 (ENT000699).

As an initial matter, I would like to emphasize that 22

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 14 15 By this standard, Entergy has not shown that the chromium 16 contents of its divider plates and tube-to-tubesheet welds are 17 sufficient to mitigate PWSCC initiation. For example, Entergy 18 acknowledges that IP2 and IP3 steam generator divider plates are 19 constructed of Alloy 600 with Alloy 82/182 cladding, a low-20 chromium combination that leaves the component vulnerable to 21 PWSCC.

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 Alloy 600 has a nominal chromium content of 14 to 17 wt.%,

2 while the welding alloys, 182 and 82, have chromium contents of 13 to 17 wt.% and 18 to 22 wt.% respectively.

4 Any combination of Alloy 600, Alloy 182, and 5 Alloy 82 therefore would result in a combined chromium content 6 of less than Accordingly, the cladding of the alloy 7 steel tubesheet, the divider plate and the divider plate-to-tubesheet cladding welds are all, 9 susceptible to PWSCC 10 initiation, independent of the ratio of ratio of the Alloy 11 82/182 in the cladding or weldments (the ratios of these alloys 12 in the cladding and in the weldments are apparently unknown).

13 Q. Please discuss the chromium content of the tube-to-14 tubesheet welds at Indian Point.

15 A. The chromium content of a tube-to-tubesheet weld --

16 and therefore its resistance to PWSCC -- depends on the type of 17 tube (Alloy 600 vs.690) and the tubesheet cladding (Alloy 82 vs.

182) involved in the weldment.

22

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 8

9

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1

2 3 Q. Do the Indian Point tubesheets contain Alloy 82 and 4 Alloy 182 cladding?

5 A. Yes. Entergy acknowledges the presence of both Alloy 6 82 and Alloy 182 cladding on its tubesheets, and has not 7 disputed 8

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 Q. How does this affect the PWSCC susceptibility of tube-2 to-tubesheet welds at Indian Point?

3 A. The IP2 steam generators utilize Alloy 600 tubes and Alloy 82/182 tubesheet cladding.

9 Thus, all of the tube-to-tubesheet welds at IP2 10 are potentially susceptible to PWSCC based on reduced chromium 11 content.

For IP3 steam generators, 13 that a significant number of tube-to-tubesheet 14 welds also lack sufficient chromium levels to mitigate PWSCC 15 initiation. The IP3 16 steam generators utilize higher chromium content Alloy 690TT 17 tubing. However, the chromium content of welds is diluted 18 through the lower chromium-containing Alloy 82/182 tubesheet cladding.

21

NYS000571 (Public, Redacted)

Submitted: September 9, 2015

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 initiation or propagation of cracks in the tubesheet, cladding or tube-to-tubesheet welds.

4 5 If a crack does initiate in any region of residual tensile 6 stresses, the leading edge of a crack is a stress intensifier, 7 locally increasing the tensile stresses. Thus, once a crack 8 initiates, it has an autocatalytic nature that tends to increase 9 the crack propagation rate (longer cracks grow faster in a constant global stress field).

12 For example, if a crack initiates 13 in the divider plate assembly or tube-to-tubesheet welds, and 14 propagates through the cladding to the tubesheet, corrosion of 15 the alloy steel tubesheet may occur during periods when the 16 divider plate assembly and the steam generator bowl are exposed 17 to air and water (maintenance periods). For example, the 18 wastage observed in the channel head bowl drain at Wolf Creek 19 illustrates the type and extent of corrosion that may occur 20 notwithstanding the presence of cladding intended to prevent 21 corrosion or cracking of components exposed to water.

22 Westinghouse Nuclear Safety Advisory Letter 12-1 (NYS000549).

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 Corrosion of alloy steels results in corrosion products 2 that are considerably more voluminous than the alloy from which 3 they are produced. This is especially important since the 4 tubesheets are known to undergo cyclic loading (fatigue).

7 Entergy Testimony at A189, A199. However, if corrosion occurs at 8 the leading edges of primary water stress corrosion cracks, the 9 local environment will be in tension due to the expansion 10 created by the corrosion product and fatigue cracks will be free 11 to propagate under local tensile stresses.

12 Q. Can PWSCC affect the growth of cracks initiated by 13 fatigue?

14 A. Cracking can occur as a result of fatigue or PWSCC, or 15 a combination of the two (sometimes expressed as stress 16 corrosion fatigue). A crack that originates by fatigue can 17 propagate by PWSCC, just as a crack that originates by PWSCC can 18 propagate by fatigue. Notably, Westinghouses fatigue analysis for the IP3 divider plates indicates that 22 (ENT000683). This indicates

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 that 10 Prompt inspection 11 of the divider plate assemblies and other components of Indian 12 Points aging steam generators would afford early detection of 13 cracks resulting from fatigue, PWSCC or a combination of fatigue 14 and PWSCC.

15 Q. Do you have closing remarks?

16 A. Both domestic and foreign operational experience 17 indicate that nuclear system components experience corrosion, 18 cracking and/or other modes of degradation and failure, 19 particularly within the context of aging fleets, despite models, 20 calculations, simulations and/or projections that would indicate 21 otherwise. These include failures of piping, steam generator 22 tubes, corrosion of clad steels, cracking of divider plate

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 assemblies, etc. Entergys own expert, Barry M. Gordon, in a 2 recent article on corrosion in light water reactors (BWRs and 3 PWRs), stated: Although corrosion was somewhat considered in 4 both plant designs, corrosion was not considered as a serious 5 concern...The problem was that the qualifying laboratory tests 6 did not necessarily reproduce the reactor operating conditions 7 (e.g., especially the high residual tensile stresses from 8 welding and cold work) and the test times were of short duration 9 relative to the initial plant design lifetime of 40 years, which 10 is currently being extended to 60 to 80 years. For example, the 11 initiation time for environmentally-assisted cracking (EAC),

12 i.e., primary water stress corrosion cracking (PWSCC) of nickel-13 base alloys in PWRs, which is the primary corrosion concern is 14 this design LWR, can be a long as 25 years! [sic] See, B.M.

15 Gordon, Corrosion and Corrosion Control in Light Water 16 Reactors, Journal of Metals, Vol. 65, Issue 8, August 2013 at 5 17 (ENT000713). The following table is an excerpt from Gordons 18 Table I entitled Partial Summary of the Corrosion History of 19 LWRs (id. at 6), and indicates the myriad problems of 20 unexpected corrosion-related events encountered in the PWR 21 fleet:

22

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1

Corrosion Event Time of Detection Alloy 600 IGSCC in a laboratory study Late 1950s IGSCC in U-bend region of PWR steam Early 1970s generator Denting of PWR Alloy 600 steam Mid 1970s generator tubing PWSCC of PWR Alloy 600 steam generator Mid 1970s tubing PWSCC in PWR pressurizer heater sleeves Early 1980s General corrosion of carbon steel Early 1980s containments FAC of single phase carbon steel Mid 1980s systems in PWRs PWSCC in PWR pressurizer instrument Late 1980s nozzles Axial PWSCC of Alloy 600 of PWR top Early 1900s head penetration Circumferential PWSCC of j-groove welds Early 1900s PWSCC of PWR hot leg nozzle Alloy Early 2000s 182/82 PWSCC induced severe boric acid Early 2000s corrosion of a PWR head SCC of stainless steels in PWRs Early 2000s 2

3 Given this history of unpredicted corrosion events, the use of 4 laboratory simulations and computational approaches to predict 5 the performance of the divider plate assemblies and associated 6 steam generator components is problematic at best. In my 7 opinion, baseline inspections with follow-up periodic

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 inspections of steam generators are the only effective means to 2 ensure that unexpected cracks or defects neither occur, nor 3 otherwise grow undetected to become failures.

4 As I previously testified, I believe 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. Instead of inspecting 8 representative welds Entergy should specifically target tube-to-tubesheet welds in areas where 11 Additionally, Entergy should identify the 12 inspection techniques it intends to use, develop acceptance 13 criteria, and provide a detailed plan for addressing any flaws 14 or indications that it may encounter. Follow-up inspections 15 should be performed at least every 10 years, given the primarily 16 Alloy 600 construction of IP2 steam generator components and 17 assemblies and the age of the IP3 steam generators.

18 In 2011, as part of this relicensing proceeding, Entergy 19 conservatively committed to confirm the absence of PWSCC 20 indications during the PEO. Entergy Testimony at A147 21 (ENT000699). NRC should condition license renewal upon Entergy 22 fulfilling that commitment.

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 Finally, I reserve the right to supplement my testimony if 2 new information is disclosed or introduced.

3 4

5 6

7

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 UNITED STATES 2 NUCLEAR REGULATORY COMMISSION 3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4 -----------------------------------x 5 In re: Docket Nos. 50-247-LR; 50-286-LR 6 License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 7 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 8 Entergy Nuclear Indian Point 3, LLC, and 9 Entergy Nuclear Operations, Inc. September 9, 2015 10 -----------------------------------x 11 DECLARATION OF DAVID J. DUQUETTE 12 I, David J. Duquette, do hereby declare under penalty of 13 perjury that my statements in the foregoing rebuttal testimony 14 and my statement of professional qualifications are true and 15 correct to the best of my knowledge and belief.

16

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1

2 Executed in Accord with 10 C.F.R. § 2.304(d) 3 4

David J. Duquette, Ph.D.

Materials Engineering Consulting Services 4 North Lane Loudonville, New York 12211 Tel: 518 276 6490 Fax: 518 462 1206 Email: duqued@rpi.edu 5 September 9, 2015

United States Nuclear Regulatory Commission Official Hearing Exhibit In the Matter of: Entergy Nuclear Operations, Inc. NYS000571 (Public, Redacted)

(Indian Point Nuclear Generating Units 2 and 3)

Submitted: September 9, 2015 ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #: NYS000571-PUB-00-BD01 Identified: 11/5/2015 Admitted: 11/5/2015 Withdrawn:

Rejected: Stricken:

Other:

1 UNITED STATES 2 NUCLEAR REGULATORY COMMISSION 3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4 -----------------------------------x 5 In re: Docket Nos. 50-247-LR; 50-286-LR 6 License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 7 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 8 Entergy Nuclear Indian Point 3, LLC, and 9 Entergy Nuclear Operations, Inc. September 9, 2015 10 -----------------------------------x 11 PRE-FILED WRITTEN SUPPLEMENTAL REPLY TESTIMONY OF 12 DR. DAVID J. DUQUETTE 13 REGARDING CONTENTION NYS-38 / RK-TC-5 14 On behalf of the State of New York (NYS or the State),

15 the Office of the Attorney General hereby submits the following 16 rebuttal testimony by David J. Duquette, Ph.D. regarding 17 Contention NYS-38/RK-TC-5.

18 Q. Please state your full name.

19 A. David J. Duquette.

20 Q. What is the purpose of this testimony you are now 21 providing?

22 A. I have previously provided testimony in this 23 proceeding regarding primary water stress corrosion cracking

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 (PWSCC) in steam generators and the need for baseline 2 inspections of the Indian Point steam generator divider plate 3 assemblies, tube-to-tubesheet welds, and channel head assemblies 4 prior to license renewal. (NYS000372, NYS000373, NYS000452, 5 NYS000532). This testimony supplements, and incorporates by 6 reference, my prior testimony in this proceeding.

7 Q. I show you what has been marked as Exhibit ENT000699.

8 Do you recognize that document?

9 A. Yes. It is a copy of the pre-filed testimony of the 10 witnesses for Entergy on Contention NYS-38/RK-TC-5 that were 11 submitted in August 2015.

12 Q. I show you what has been marked as Exhibit NRCR000161, 13 NRC000197 and NRC000168. Do you recognize those documents?

14 A. Yes. They are copies of the pre-filed testimony of 15 NRC Staff witnesses that were submitted in August 2015. They 16 concern Contention NYS-38/RK-TC-5. I note that NRC000168 and 17 NRC000197 primarily address Contentions NYS-25 and NYS-26B/RK-18 TC-1B.

19 Q. Have you had an opportunity to review ENT000699, 20 NRCR000161, NRC000168, and NRC000197?

21 A. Yes.

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 Q. Has Entergys and NRC Staffs August pre-filed 2 testimony caused you to change the testimony and opinions that 3 you have previously submitted in this proceeding in connection 4 with Contention NYS-38?

5 A. No. It is still my opinion that Entergy should 6 perform visual baseline inspections of the eight Indian Point 7 steam generators prior to license renewal, that is, before 8 Entergy receives renewal 20-year operating licenses for Indian 9 Point Unit 2 and Unit 3.

10 Q. Does Entergy currently perform inspections of the 11 steam generators?

12 A. It is my understanding that Entergy currently performs 13 periodic visual inspections of the steam generator bowls, 14 tubesheets, and plugs using remote camera techniques. Entergy 15 Testimony at A175 (ENT000699). Entergy could readily expand the 16 scope of those inspections to include the divider plate 17 assemblies and the tube-to-tubesheet welds.

18 Q. What is Entergys position with respect to performing 19 divider plate and tube-to-tubesheet weld inspections?

20 A. It is non-committal, at present. Although Entergy 21 committed to inspect for PWSCC in the IP2 and IP3 steam 22 generator divider plates as part of Commitment 41 (see, NL NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 074, Attach. 1 at 14 (NYS000152), its most recent testimony 2 confirms that the company is considering retraction of that commitment in light of 12 13 Entergy had also committed, via Commitment 42, to address 14 PWSCC in tube-to-tubesheet welds by either performing 15 inspections of those steam generator component locations, or by 16 demonstrating through analyses that they are not susceptible to 17 PWSCC or are not part of the reactor coolant boundary. NL 18 032, Attach. 1 at 22-23 (NYS000151); NL-11-074, Attach. 2 at 15 19 (NYS000152).

20 Q. What is the current status of Commitment 42?

21 A. In late 2014, Entergy redefined the reactor coolant 22 pressure boundary so as to exclude tube-to-tubesheet weldments

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 at the IP2 steam generators. NYS000542. Apparently, Entergy 2 considers Commitment 42 fulfilled as to IP2. NYS000553.

3 Q. How is Entergy proposing to fulfill Commitment 42 for 4 IP3 steam generator tube-to-tubesheet welds?

5 A. Entergy has indicated that it prefers the analysis 6 option over the inspection option for IP3, and is currently evaluating whether 9 Entergy Testimony at A166 (ENT000699).

10 Q. To your knowledge, have the IP2 and IP3 steam 11 generator divider plates and tube-to-tubesheet welds ever been 12 inspected for PWSCC?

13 A. No, I do not believe they have ever been inspected. In 14 its testimony, Entergy does not mention any previous inspections 15 of these components and locations. Moreover, given Entergys 16 support for and the companys stated 17 preference for analysis rather than inspection, those 18 components/locations may never be inspected.

19 Q. Do you have any concerns about Entergys reliance on 20 as a basis to retract Commitment 41 and/or 21 close Commitment 42?

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 A. As I have previously indicated, I do not believe the 2 results of EPRIs investigations into steam generator component 3 cracking eliminate the need for actual inspections at Indian Point. To the contrary, 17 Entergy Testimony at A188 (ENT000699).

As an initial matter, I would like to emphasize that 22

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 14 15 By this standard, Entergy has not shown that the chromium 16 contents of its divider plates and tube-to-tubesheet welds are 17 sufficient to mitigate PWSCC initiation. For example, Entergy 18 acknowledges that IP2 and IP3 steam generator divider plates are 19 constructed of Alloy 600 with Alloy 82/182 cladding, a low-20 chromium combination that leaves the component vulnerable to 21 PWSCC.

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 Alloy 600 has a nominal chromium content of 14 to 17 wt.%,

2 while the welding alloys, 182 and 82, have chromium contents of 13 to 17 wt.% and 18 to 22 wt.% respectively.

4 Any combination of Alloy 600, Alloy 182, and 5 Alloy 82 therefore would result in a combined chromium content 6 of less than Accordingly, the cladding of the alloy 7 steel tubesheet, the divider plate and the divider plate-to-tubesheet cladding welds are all, 9 susceptible to PWSCC 10 initiation, independent of the ratio of ratio of the Alloy 11 82/182 in the cladding or weldments (the ratios of these alloys 12 in the cladding and in the weldments are apparently unknown).

13 Q. Please discuss the chromium content of the tube-to-14 tubesheet welds at Indian Point.

15 A. The chromium content of a tube-to-tubesheet weld --

16 and therefore its resistance to PWSCC -- depends on the type of 17 tube (Alloy 600 vs.690) and the tubesheet cladding (Alloy 82 vs.

182) involved in the weldment.

22

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 8

9

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1

2 3 Q. Do the Indian Point tubesheets contain Alloy 82 and 4 Alloy 182 cladding?

5 A. Yes. Entergy acknowledges the presence of both Alloy 6 82 and Alloy 182 cladding on its tubesheets, and has not 7 disputed 8

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 Q. How does this affect the PWSCC susceptibility of tube-2 to-tubesheet welds at Indian Point?

3 A. The IP2 steam generators utilize Alloy 600 tubes and Alloy 82/182 tubesheet cladding.

9 Thus, all of the tube-to-tubesheet welds at IP2 10 are potentially susceptible to PWSCC based on reduced chromium 11 content.

For IP3 steam generators, 13 that a significant number of tube-to-tubesheet 14 welds also lack sufficient chromium levels to mitigate PWSCC 15 initiation. The IP3 16 steam generators utilize higher chromium content Alloy 690TT 17 tubing. However, the chromium content of welds is diluted 18 through the lower chromium-containing Alloy 82/182 tubesheet cladding.

21

NYS000571 (Public, Redacted)

Submitted: September 9, 2015

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 initiation or propagation of cracks in the tubesheet, cladding or tube-to-tubesheet welds.

4 5 If a crack does initiate in any region of residual tensile 6 stresses, the leading edge of a crack is a stress intensifier, 7 locally increasing the tensile stresses. Thus, once a crack 8 initiates, it has an autocatalytic nature that tends to increase 9 the crack propagation rate (longer cracks grow faster in a constant global stress field).

12 For example, if a crack initiates 13 in the divider plate assembly or tube-to-tubesheet welds, and 14 propagates through the cladding to the tubesheet, corrosion of 15 the alloy steel tubesheet may occur during periods when the 16 divider plate assembly and the steam generator bowl are exposed 17 to air and water (maintenance periods). For example, the 18 wastage observed in the channel head bowl drain at Wolf Creek 19 illustrates the type and extent of corrosion that may occur 20 notwithstanding the presence of cladding intended to prevent 21 corrosion or cracking of components exposed to water.

22 Westinghouse Nuclear Safety Advisory Letter 12-1 (NYS000549).

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 Corrosion of alloy steels results in corrosion products 2 that are considerably more voluminous than the alloy from which 3 they are produced. This is especially important since the 4 tubesheets are known to undergo cyclic loading (fatigue).

7 Entergy Testimony at A189, A199. However, if corrosion occurs at 8 the leading edges of primary water stress corrosion cracks, the 9 local environment will be in tension due to the expansion 10 created by the corrosion product and fatigue cracks will be free 11 to propagate under local tensile stresses.

12 Q. Can PWSCC affect the growth of cracks initiated by 13 fatigue?

14 A. Cracking can occur as a result of fatigue or PWSCC, or 15 a combination of the two (sometimes expressed as stress 16 corrosion fatigue). A crack that originates by fatigue can 17 propagate by PWSCC, just as a crack that originates by PWSCC can 18 propagate by fatigue. Notably, Westinghouses fatigue analysis for the IP3 divider plates indicates that 22 (ENT000683). This indicates

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 that 10 Prompt inspection 11 of the divider plate assemblies and other components of Indian 12 Points aging steam generators would afford early detection of 13 cracks resulting from fatigue, PWSCC or a combination of fatigue 14 and PWSCC.

15 Q. Do you have closing remarks?

16 A. Both domestic and foreign operational experience 17 indicate that nuclear system components experience corrosion, 18 cracking and/or other modes of degradation and failure, 19 particularly within the context of aging fleets, despite models, 20 calculations, simulations and/or projections that would indicate 21 otherwise. These include failures of piping, steam generator 22 tubes, corrosion of clad steels, cracking of divider plate

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 assemblies, etc. Entergys own expert, Barry M. Gordon, in a 2 recent article on corrosion in light water reactors (BWRs and 3 PWRs), stated: Although corrosion was somewhat considered in 4 both plant designs, corrosion was not considered as a serious 5 concern...The problem was that the qualifying laboratory tests 6 did not necessarily reproduce the reactor operating conditions 7 (e.g., especially the high residual tensile stresses from 8 welding and cold work) and the test times were of short duration 9 relative to the initial plant design lifetime of 40 years, which 10 is currently being extended to 60 to 80 years. For example, the 11 initiation time for environmentally-assisted cracking (EAC),

12 i.e., primary water stress corrosion cracking (PWSCC) of nickel-13 base alloys in PWRs, which is the primary corrosion concern is 14 this design LWR, can be a long as 25 years! [sic] See, B.M.

15 Gordon, Corrosion and Corrosion Control in Light Water 16 Reactors, Journal of Metals, Vol. 65, Issue 8, August 2013 at 5 17 (ENT000713). The following table is an excerpt from Gordons 18 Table I entitled Partial Summary of the Corrosion History of 19 LWRs (id. at 6), and indicates the myriad problems of 20 unexpected corrosion-related events encountered in the PWR 21 fleet:

22

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1

Corrosion Event Time of Detection Alloy 600 IGSCC in a laboratory study Late 1950s IGSCC in U-bend region of PWR steam Early 1970s generator Denting of PWR Alloy 600 steam Mid 1970s generator tubing PWSCC of PWR Alloy 600 steam generator Mid 1970s tubing PWSCC in PWR pressurizer heater sleeves Early 1980s General corrosion of carbon steel Early 1980s containments FAC of single phase carbon steel Mid 1980s systems in PWRs PWSCC in PWR pressurizer instrument Late 1980s nozzles Axial PWSCC of Alloy 600 of PWR top Early 1900s head penetration Circumferential PWSCC of j-groove welds Early 1900s PWSCC of PWR hot leg nozzle Alloy Early 2000s 182/82 PWSCC induced severe boric acid Early 2000s corrosion of a PWR head SCC of stainless steels in PWRs Early 2000s 2

3 Given this history of unpredicted corrosion events, the use of 4 laboratory simulations and computational approaches to predict 5 the performance of the divider plate assemblies and associated 6 steam generator components is problematic at best. In my 7 opinion, baseline inspections with follow-up periodic

NYS000571 (Public, Redacted)

Submitted: September 9, 2015 1 inspections of steam generators are the only effective means to 2 ensure that unexpected cracks or defects neither occur, nor 3 otherwise grow undetected to become failures.

4 As I previously testified, I believe 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. Instead of inspecting 8 representative welds Entergy should specifically target tube-to-tubesheet welds in areas where 11 Additionally, Entergy should identify the 12 inspection techniques it intends to use, develop acceptance 13 criteria, and provide a detailed plan for addressing any flaws 14 or indications that it may encounter. Follow-up inspections 15 should be performed at least every 10 years, given the primarily 16 Alloy 600 construction of IP2 steam generator components and 17 assemblies and the age of the IP3 steam generators.

18 In 2011, as part of this relicensing proceeding, Entergy 19 conservatively committed to confirm the absence of PWSCC 20 indications during the PEO. Entergy Testimony at A147 21 (ENT000699). NRC should condition license renewal upon Entergy 22 fulfilling that commitment.

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Submitted: September 9, 2015 1 Finally, I reserve the right to supplement my testimony if 2 new information is disclosed or introduced.

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Submitted: September 9, 2015 1 UNITED STATES 2 NUCLEAR REGULATORY COMMISSION 3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4 -----------------------------------x 5 In re: Docket Nos. 50-247-LR; 50-286-LR 6 License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 7 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 8 Entergy Nuclear Indian Point 3, LLC, and 9 Entergy Nuclear Operations, Inc. September 9, 2015 10 -----------------------------------x 11 DECLARATION OF DAVID J. DUQUETTE 12 I, David J. Duquette, do hereby declare under penalty of 13 perjury that my statements in the foregoing rebuttal testimony 14 and my statement of professional qualifications are true and 15 correct to the best of my knowledge and belief.

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Submitted: September 9, 2015 1

2 Executed in Accord with 10 C.F.R. § 2.304(d) 3 4

David J. Duquette, Ph.D.

Materials Engineering Consulting Services 4 North Lane Loudonville, New York 12211 Tel: 518 276 6490 Fax: 518 462 1206 Email: duqued@rpi.edu 5 September 9, 2015