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 J 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 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. September 9, 2015 9 -----------------------------------x 10 PRE-FILED WRITTEN SUPPLEMENTAL REPLY 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. I have previously provided testimony in this 22 proceeding regarding primary water stress corrosion cracking 23 (PWSCC) in steam generators and the need for baseline 1 inspections of the Indian Point steam generator divider plate 2 assemblies, tube-to-tubesheet welds, and channel head assemblies 3 prior to license renewal. (NYS000372, NYS000373, NYS000452, 4 NYS000532). This testimony supplements, and incorporates by 5 reference, my prior testimony in this proceeding. 6 Q. I show you what has been marked as Exhibit ENT000699. 7 Do you recognize that document? 8 A. Yes. It is a copy of the pre-filed testimony of the 9 witnesses for Entergy on Contention NYS-38/RK-TC-5 that were 10 submitted in August 2015. 11 Q. I show you what has been marked as Exhibit NRCR000161, 12 NRC000197 and NRC000168. Do you recognize those documents? 13 A. Yes. They are copies of the pre-filed testimony of 14 NRC Staff witnesses that were submitted in August 2015. They 15 concern Contention NYS-38/RK-TC-5. I note that NRC000168 and 16 NRC000197 primarily address Contentions NYS-25 and NYS-26B/RK-17 TC-1B. 18 Q. Have you had an opportunity to review ENT000699, 19 NRCR000161, NRC000168, and NRC000197? 20 A. Yes. 21 Q. Has Entergy's and NRC Staff's August pre-filed 1 testimony caused you to change the testimony and opinions that 2 you have previously submitted in this proceeding in connection 3 with Contention NYS-38? 4 A. No. It is still my opinion that Entergy should 5 perform visual baseline inspections of the eight Indian Point 6 steam generators prior to license renewal, that is, before 7 Entergy receives renewal 20-year operating licenses for Indian 8 Point Unit 2 and Unit 3. 9 Q. Does Entergy currently perform inspections of the 10 steam generators? 11 A. It is my understanding that Entergy currently performs 12 periodic visual inspections of the steam generator bowls, 13 tubesheets, and plugs using remote camera techniques. Entergy 14 Testimony at A175 (ENT000699). Entergy could readily expand the 15 scope of those inspections to include the divider plate 16 assemblies and the tube-to-tubesheet welds. 17 Q. What is Entergy's position with respect to performing 18 divider plate and tube-to-tubesheet weld inspections? 19 A. It is non-committal, at present. Although Entergy 20 committed to inspect for PWSCC in the IP2 and IP3 steam 21 generator divider plates as part of Commitment 41 (see, NL-11-22 074, Attach. 1 at 14 (NYS000152), its most recent testimony 1 confirms that the company is considering retraction of that 2 commitment in light of 12 Entergy had also committed, via Commitment 42, to address 13 PWSCC in tube-to-tubesheet welds by either performing 14 inspections of those steam generator component locations, or by 15 demonstrating through analyses that they are not susceptible to 16 PWSCC or are not part of the reactor coolant boundary. NL-11-17 032, Attach. 1 at 22-23 (NYS000151); NL-11-074, Attach. 2 at 15 18 (NYS000152). 19 Q. What is the current status of Commitment 42? 20 A. In late 2014, Entergy redefined the reactor coolant 21 pressure boundary so as to exclude tube-to-tubesheet weldments 22 at the IP2 steam generators. NYS000542. Apparently, Entergy 1 considers Commitment 42 fulfilled as to IP2. NYS000553. 2 Q. How is Entergy proposing to fulfill Commitment 42 for 3 IP3 steam generator tube-to-tubesheet welds? 4 A. Entergy has indicated that it prefers the "analysis" 5 option over the "inspection" option for IP3, and is currently 6 evaluating whether Entergy Testimony at A166 (ENT000699). 9 Q. To your knowledge, have the IP2 and IP3 steam 10 generator divider plates and tube-to-tubesheet welds ever been 11 inspected for PWSCC? 12 A. No, I do not believe they have ever been inspected. In 13 its testimony, Entergy does not mention any previous inspections 14 of these components and locations. Moreover, given Entergy's 15 support for and the company's stated 16 preference for analysis rather than inspection, those 17 components/locations may never be inspected. 18 Q. Do you have any concerns about Entergy's reliance on 19 as a basis to retract Commitment 41 and/or 20 close Commitment 42? 21 A. As I have previously indicated, I do not believe the 1 results of EPRI's investigations into steam generator component 2 cracking eliminate the need for actual inspections at Indian 3 Point. To the contrary, Entergy Testimony at A188 (ENT000699). 17 As an initial matter, I would like to emphasize that 22 14 By this standard, Entergy has not shown that the chromium 15 contents of its divider plates and tube-to-tubesheet welds are 16 sufficient to mitigate PWSCC initiation. For example, Entergy 17 acknowledges that IP2 and IP3 steam generator divider plates are 18 constructed of Alloy 600 with Alloy 82/182 cladding, a low-19 chromium combination that leaves the component vulnerable to 20 PWSCC. 21 Alloy 600 has a nominal chromium content of 14 to 17 wt.%, 1 while the welding alloys, 182 and 82, have chromium contents of 2 13 to 17 wt.% and 18 to 22 wt.% respectively. Any combination of Alloy 600, Alloy 182, and 4 Alloy 82 therefore would result in a combined chromium content 5 of less than Accordingly, the cladding of the alloy 6 steel tubesheet, the divider plate and the divider plate-to- 7 tubesheet cladding welds are all, susceptible to PWSCC 9 initiation, independent of the ratio of ratio of the Alloy 10 82/182 in the cladding or weldments (the ratios of these alloys 11 in the cladding and in the weldments are apparently unknown). 12 Q. Please discuss the chromium content of the tube-to-13 tubesheet welds at Indian Point. 14 A. The chromium content of a tube-to-tubesheet weld -- 15 and therefore its resistance to PWSCC -- depends on the type of 16 tube (Alloy 600 vs.690) and the tubesheet cladding (Alloy 82 vs. 17 182) involved in the weldment. 22 8 9 1 2 Q. Do the Indian Point tubesheets contain Alloy 82 and 3 Alloy 182 cladding? 4 A. Yes. Entergy acknowledges the presence of both Alloy 5 82 and Alloy 182 cladding on its tubesheets, and has not 6 disputed 7 8 Q. How does this affect the PWSCC susceptibility of tube-1 to-tubesheet welds at Indian Point? 2 A. The IP2 steam generators utilize Alloy 600 tubes and 3 Alloy 82/182 tubesheet cladding. Thus, all of the tube-to-tubesheet welds at IP2 9 are potentially susceptible to PWSCC based on reduced chromium 10 content. 11 For IP3 steam generators, that a significant number of tube-to-tubesheet 13 welds also lack sufficient chromium levels to mitigate PWSCC 14 initiation. The IP3 15 steam generators utilize higher chromium content Alloy 690TT 16 tubing. However, the chromium content of welds is diluted 17 through the lower chromium-containing Alloy 82/182 tubesheet 18 cladding. 21

initiation or propagation of cracks in the tubesheet, cladding 1 or tube-to-tubesheet welds. 4 If a crack does initiate in any region of residual tensile 5 stresses, the leading edge of a crack is a stress intensifier, 6 locally increasing the tensile stresses. Thus, once a crack 7 initiates, it has an autocatalytic nature that tends to increase 8 the crack propagation rate (longer cracks grow faster in a 9 constant global stress field). For example, if a crack initiates 12 in the divider plate assembly or tube-to-tubesheet welds, and 13 propagates through the cladding to the tubesheet, corrosion of 14 the alloy steel tubesheet may occur during periods when the 15 divider plate assembly and the steam generator bowl are exposed 16 to air and water (maintenance periods). For example, the 17 wastage observed in the channel head bowl drain at Wolf Creek 18 illustrates the type and extent of corrosion that may occur 19 notwithstanding the presence of cladding intended to prevent 20 corrosion or cracking of components exposed to water. 21 Westinghouse Nuclear Safety Advisory Letter 12-1 (NYS000549). 22 Corrosion of alloy steels results in corrosion products 1 that are considerably more voluminous than the alloy from which 2 they are produced. This is especially important since the 3 tubesheets are known to undergo cyclic loading (fatigue). 4 Entergy Testimony at A189, A199. However, if corrosion occurs at 7 the leading edges of primary water stress corrosion cracks, the 8 local environment will be in tension due to the expansion 9 created by the corrosion product and fatigue cracks will be free 10 to propagate under local tensile stresses. 11 Q. Can PWSCC affect the growth of cracks initiated by 12 fatigue? 13 A. Cracking can occur as a result of fatigue or PWSCC, or 14 a combination of the two (sometimes expressed as stress 15 corrosion fatigue). A crack that originates by fatigue can 16 propagate by PWSCC, just as a crack that originates by PWSCC can 17 propagate by fatigue. Notably, Westinghouse's fatigue analysis 18 for the IP3 divider plates indicates that (ENT000683). This indicates 22 that Prompt inspection 10 of the divider plate assemblies and other components of Indian 11 Point's aging steam generators would afford early detection of 12 cracks resulting from fatigue, PWSCC or a combination of fatigue 13 and PWSCC. 14 Q. Do you have closing remarks? 15 A. Both domestic and foreign operational experience 16 indicate that nuclear system components experience corrosion, 17 cracking and/or other modes of degradation and failure, 18 particularly within the context of aging fleets, despite models, 19 calculations, simulations and/or projections that would indicate 20 otherwise. These include failures of piping, steam generator 21 tubes, corrosion of clad steels, cracking of divider plate 22 assemblies, etc. Entergy's own expert, Barry M. Gordon, in a 1 recent article on corrosion in light water reactors (BWR's and 2 PWR's), stated: "Although corrosion was somewhat considered in 3 both plant designs, corrosion was not considered as a serious 4 concern...The problem was that the 'qualifying' laboratory tests 5 did not necessarily reproduce the reactor operating conditions 6 (e.g., especially the high residual tensile stresses from 7 welding and cold work) and the test times were of short duration 8 relative to the initial plant design lifetime of 40 years, which 9 is currently being extended to 60 to 80 years. For example, the 10 initiation time for environmentally-assisted cracking (EAC), 11 i.e., primary water stress corrosion cracking (PWSCC) of nickel-12 base alloys in PWRs, which is the primary corrosion concern is 13 this design LWR, can be a long as 25 years! [sic]" See, B.M. 14 Gordon, "Corrosion and Corrosion Control in Light Water 15 Reactors," Journal of Metals, Vol. 65, Issue 8, August 2013 at 5 16 (ENT000713). The following table is an excerpt from Gordon's 17 Table I entitled "Partial Summary of the Corrosion History of 18 LWRs" (id. at 6), and indicates the myriad problems of 19 unexpected corrosion-related events encountered in the PWR 20 fleet: 21 22 1 Corrosion Event Time of Detection Alloy 600 IGSCC in a laboratory study Late 1950's IGSCC in U-bend region of PWR steam generator Early 1970s Denting of PWR Alloy 600 steam generator tubing Mid 1970s PWSCC of PWR Alloy 600 steam generator tubing Mid 1970s PWSCC in PWR pressurizer heater sleeves Early 1980s General corrosion of carbon steel containments Early 1980s FAC of single phase carbon steel systems in PWRs Mid 1980s PWSCC in PWR pressurizer instrument nozzles Late 1980s Axial PWSCC of Alloy 600 of PWR top head penetration Early 1900s Circumferential PWSCC of j-groove welds Early 1900s PWSCC of PWR hot leg nozzle Alloy 182/82 Early 2000s PWSCC induced severe boric acid corrosion of a PWR head Early 2000s SCC of stainless steels in PWRs Early 2000s 2 Given this history of unpredicted corrosion events, the use of 3 laboratory simulations and computational approaches to predict 4 the performance of the divider plate assemblies and associated 5 steam generator components is problematic at best. In my 6 opinion, baseline inspections with follow-up periodic 7 inspections of steam generators are the only effective means to 1 ensure that unexpected cracks or defects neither occur, nor 2 otherwise grow undetected to become failures. 3 As I previously testified, I believe Entergy should 4 affirmatively and clearly commit to performing inspections as 5 soon as possible for IP2, and certainly before the period of 6 extended operation for IP3. Instead of inspecting 7 "representative welds" Entergy should specifically target tube-8 to-tubesheet welds in areas where Additionally, Entergy should identify the 11 inspection techniques it intends to use, develop acceptance 12 criteria, and provide a detailed plan for addressing any flaws 13 or indications that it may encounter. Follow-up inspections 14 should be performed at least every 10 years, given the primarily 15 Alloy 600 construction of IP2 steam generator components and 16 assemblies and the age of the IP3 steam generators. 17 In 2011, as part of this relicensing proceeding, Entergy 18 "conservatively committed to confirm the absence of PWSCC 19 indications during the PEO." Entergy Testimony at A147 20 (ENT000699). NRC should condition license renewal upon Entergy 21 fulfilling that commitment. 22 Finally, I reserve the right to supplement my testimony if 1 new information is disclosed or introduced. 2 3 4 5 6 7 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. September 9, 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 16 1 Executed in Accord with 10 C.F.R. § 2.304(d) 2 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 September 9, 2015 5 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. September 9, 2015 9 -----------------------------------x 10 PRE-FILED WRITTEN SUPPLEMENTAL REPLY 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. I have previously provided testimony in this 22 proceeding regarding primary water stress corrosion cracking 23 (PWSCC) in steam generators and the need for baseline 1 inspections of the Indian Point steam generator divider plate 2 assemblies, tube-to-tubesheet welds, and channel head assemblies 3 prior to license renewal. (NYS000372, NYS000373, NYS000452, 4 NYS000532). This testimony supplements, and incorporates by 5 reference, my prior testimony in this proceeding. 6 Q. I show you what has been marked as Exhibit ENT000699. 7 Do you recognize that document? 8 A. Yes. It is a copy of the pre-filed testimony of the 9 witnesses for Entergy on Contention NYS-38/RK-TC-5 that were 10 submitted in August 2015. 11 Q. I show you what has been marked as Exhibit NRCR000161, 12 NRC000197 and NRC000168. Do you recognize those documents? 13 A. Yes. They are copies of the pre-filed testimony of 14 NRC Staff witnesses that were submitted in August 2015. They 15 concern Contention NYS-38/RK-TC-5. I note that NRC000168 and 16 NRC000197 primarily address Contentions NYS-25 and NYS-26B/RK-17 TC-1B. 18 Q. Have you had an opportunity to review ENT000699, 19 NRCR000161, NRC000168, and NRC000197? 20 A. Yes. 21 Q. Has Entergy's and NRC Staff's August pre-filed 1 testimony caused you to change the testimony and opinions that 2 you have previously submitted in this proceeding in connection 3 with Contention NYS-38? 4 A. No. It is still my opinion that Entergy should 5 perform visual baseline inspections of the eight Indian Point 6 steam generators prior to license renewal, that is, before 7 Entergy receives renewal 20-year operating licenses for Indian 8 Point Unit 2 and Unit 3. 9 Q. Does Entergy currently perform inspections of the 10 steam generators? 11 A. It is my understanding that Entergy currently performs 12 periodic visual inspections of the steam generator bowls, 13 tubesheets, and plugs using remote camera techniques. Entergy 14 Testimony at A175 (ENT000699). Entergy could readily expand the 15 scope of those inspections to include the divider plate 16 assemblies and the tube-to-tubesheet welds. 17 Q. What is Entergy's position with respect to performing 18 divider plate and tube-to-tubesheet weld inspections? 19 A. It is non-committal, at present. Although Entergy 20 committed to inspect for PWSCC in the IP2 and IP3 steam 21 generator divider plates as part of Commitment 41 (see, NL-11-22 074, Attach. 1 at 14 (NYS000152), its most recent testimony 1 confirms that the company is considering retraction of that 2 commitment in light of 12 Entergy had also committed, via Commitment 42, to address 13 PWSCC in tube-to-tubesheet welds by either performing 14 inspections of those steam generator component locations, or by 15 demonstrating through analyses that they are not susceptible to 16 PWSCC or are not part of the reactor coolant boundary. NL-11-17 032, Attach. 1 at 22-23 (NYS000151); NL-11-074, Attach. 2 at 15 18 (NYS000152). 19 Q. What is the current status of Commitment 42? 20 A. In late 2014, Entergy redefined the reactor coolant 21 pressure boundary so as to exclude tube-to-tubesheet weldments 22 at the IP2 steam generators. NYS000542. Apparently, Entergy 1 considers Commitment 42 fulfilled as to IP2. NYS000553. 2 Q. How is Entergy proposing to fulfill Commitment 42 for 3 IP3 steam generator tube-to-tubesheet welds? 4 A. Entergy has indicated that it prefers the "analysis" 5 option over the "inspection" option for IP3, and is currently 6 evaluating whether Entergy Testimony at A166 (ENT000699). 9 Q. To your knowledge, have the IP2 and IP3 steam 10 generator divider plates and tube-to-tubesheet welds ever been 11 inspected for PWSCC? 12 A. No, I do not believe they have ever been inspected. In 13 its testimony, Entergy does not mention any previous inspections 14 of these components and locations. Moreover, given Entergy's 15 support for and the company's stated 16 preference for analysis rather than inspection, those 17 components/locations may never be inspected. 18 Q. Do you have any concerns about Entergy's reliance on 19 as a basis to retract Commitment 41 and/or 20 close Commitment 42? 21 A. As I have previously indicated, I do not believe the 1 results of EPRI's investigations into steam generator component 2 cracking eliminate the need for actual inspections at Indian 3 Point. To the contrary, Entergy Testimony at A188 (ENT000699). 17 As an initial matter, I would like to emphasize that 22 14 By this standard, Entergy has not shown that the chromium 15 contents of its divider plates and tube-to-tubesheet welds are 16 sufficient to mitigate PWSCC initiation. For example, Entergy 17 acknowledges that IP2 and IP3 steam generator divider plates are 18 constructed of Alloy 600 with Alloy 82/182 cladding, a low-19 chromium combination that leaves the component vulnerable to 20 PWSCC. 21 Alloy 600 has a nominal chromium content of 14 to 17 wt.%, 1 while the welding alloys, 182 and 82, have chromium contents of 2 13 to 17 wt.% and 18 to 22 wt.% respectively. Any combination of Alloy 600, Alloy 182, and 4 Alloy 82 therefore would result in a combined chromium content 5 of less than Accordingly, the cladding of the alloy 6 steel tubesheet, the divider plate and the divider plate-to- 7 tubesheet cladding welds are all, susceptible to PWSCC 9 initiation, independent of the ratio of ratio of the Alloy 10 82/182 in the cladding or weldments (the ratios of these alloys 11 in the cladding and in the weldments are apparently unknown). 12 Q. Please discuss the chromium content of the tube-to-13 tubesheet welds at Indian Point. 14 A. The chromium content of a tube-to-tubesheet weld -- 15 and therefore its resistance to PWSCC -- depends on the type of 16 tube (Alloy 600 vs.690) and the tubesheet cladding (Alloy 82 vs. 17 182) involved in the weldment. 22 8 9 1 2 Q. Do the Indian Point tubesheets contain Alloy 82 and 3 Alloy 182 cladding? 4 A. Yes. Entergy acknowledges the presence of both Alloy 5 82 and Alloy 182 cladding on its tubesheets, and has not 6 disputed 7 8 Q. How does this affect the PWSCC susceptibility of tube-1 to-tubesheet welds at Indian Point? 2 A. The IP2 steam generators utilize Alloy 600 tubes and 3 Alloy 82/182 tubesheet cladding. Thus, all of the tube-to-tubesheet welds at IP2 9 are potentially susceptible to PWSCC based on reduced chromium 10 content. 11 For IP3 steam generators, that a significant number of tube-to-tubesheet 13 welds also lack sufficient chromium levels to mitigate PWSCC 14 initiation. The IP3 15 steam generators utilize higher chromium content Alloy 690TT 16 tubing. However, the chromium content of welds is diluted 17 through the lower chromium-containing Alloy 82/182 tubesheet 18 cladding. 21

initiation or propagation of cracks in the tubesheet, cladding 1 or tube-to-tubesheet welds. 4 If a crack does initiate in any region of residual tensile 5 stresses, the leading edge of a crack is a stress intensifier, 6 locally increasing the tensile stresses. Thus, once a crack 7 initiates, it has an autocatalytic nature that tends to increase 8 the crack propagation rate (longer cracks grow faster in a 9 constant global stress field). For example, if a crack initiates 12 in the divider plate assembly or tube-to-tubesheet welds, and 13 propagates through the cladding to the tubesheet, corrosion of 14 the alloy steel tubesheet may occur during periods when the 15 divider plate assembly and the steam generator bowl are exposed 16 to air and water (maintenance periods). For example, the 17 wastage observed in the channel head bowl drain at Wolf Creek 18 illustrates the type and extent of corrosion that may occur 19 notwithstanding the presence of cladding intended to prevent 20 corrosion or cracking of components exposed to water. 21 Westinghouse Nuclear Safety Advisory Letter 12-1 (NYS000549). 22 Corrosion of alloy steels results in corrosion products 1 that are considerably more voluminous than the alloy from which 2 they are produced. This is especially important since the 3 tubesheets are known to undergo cyclic loading (fatigue). 4 Entergy Testimony at A189, A199. However, if corrosion occurs at 7 the leading edges of primary water stress corrosion cracks, the 8 local environment will be in tension due to the expansion 9 created by the corrosion product and fatigue cracks will be free 10 to propagate under local tensile stresses. 11 Q. Can PWSCC affect the growth of cracks initiated by 12 fatigue? 13 A. Cracking can occur as a result of fatigue or PWSCC, or 14 a combination of the two (sometimes expressed as stress 15 corrosion fatigue). A crack that originates by fatigue can 16 propagate by PWSCC, just as a crack that originates by PWSCC can 17 propagate by fatigue. Notably, Westinghouse's fatigue analysis 18 for the IP3 divider plates indicates that (ENT000683). This indicates 22 that Prompt inspection 10 of the divider plate assemblies and other components of Indian 11 Point's aging steam generators would afford early detection of 12 cracks resulting from fatigue, PWSCC or a combination of fatigue 13 and PWSCC. 14 Q. Do you have closing remarks? 15 A. Both domestic and foreign operational experience 16 indicate that nuclear system components experience corrosion, 17 cracking and/or other modes of degradation and failure, 18 particularly within the context of aging fleets, despite models, 19 calculations, simulations and/or projections that would indicate 20 otherwise. These include failures of piping, steam generator 21 tubes, corrosion of clad steels, cracking of divider plate 22 assemblies, etc. Entergy's own expert, Barry M. Gordon, in a 1 recent article on corrosion in light water reactors (BWR's and 2 PWR's), stated: "Although corrosion was somewhat considered in 3 both plant designs, corrosion was not considered as a serious 4 concern...The problem was that the 'qualifying' laboratory tests 5 did not necessarily reproduce the reactor operating conditions 6 (e.g., especially the high residual tensile stresses from 7 welding and cold work) and the test times were of short duration 8 relative to the initial plant design lifetime of 40 years, which 9 is currently being extended to 60 to 80 years. For example, the 10 initiation time for environmentally-assisted cracking (EAC), 11 i.e., primary water stress corrosion cracking (PWSCC) of nickel-12 base alloys in PWRs, which is the primary corrosion concern is 13 this design LWR, can be a long as 25 years! [sic]" See, B.M. 14 Gordon, "Corrosion and Corrosion Control in Light Water 15 Reactors," Journal of Metals, Vol. 65, Issue 8, August 2013 at 5 16 (ENT000713). The following table is an excerpt from Gordon's 17 Table I entitled "Partial Summary of the Corrosion History of 18 LWRs" (id. at 6), and indicates the myriad problems of 19 unexpected corrosion-related events encountered in the PWR 20 fleet: 21 22 1 Corrosion Event Time of Detection Alloy 600 IGSCC in a laboratory study Late 1950's IGSCC in U-bend region of PWR steam generator Early 1970s Denting of PWR Alloy 600 steam generator tubing Mid 1970s PWSCC of PWR Alloy 600 steam generator tubing Mid 1970s PWSCC in PWR pressurizer heater sleeves Early 1980s General corrosion of carbon steel containments Early 1980s FAC of single phase carbon steel systems in PWRs Mid 1980s PWSCC in PWR pressurizer instrument nozzles Late 1980s Axial PWSCC of Alloy 600 of PWR top head penetration Early 1900s Circumferential PWSCC of j-groove welds Early 1900s PWSCC of PWR hot leg nozzle Alloy 182/82 Early 2000s PWSCC induced severe boric acid corrosion of a PWR head Early 2000s SCC of stainless steels in PWRs Early 2000s 2 Given this history of unpredicted corrosion events, the use of 3 laboratory simulations and computational approaches to predict 4 the performance of the divider plate assemblies and associated 5 steam generator components is problematic at best. In my 6 opinion, baseline inspections with follow-up periodic 7 inspections of steam generators are the only effective means to 1 ensure that unexpected cracks or defects neither occur, nor 2 otherwise grow undetected to become failures. 3 As I previously testified, I believe Entergy should 4 affirmatively and clearly commit to performing inspections as 5 soon as possible for IP2, and certainly before the period of 6 extended operation for IP3. Instead of inspecting 7 "representative welds" Entergy should specifically target tube-8 to-tubesheet welds in areas where Additionally, Entergy should identify the 11 inspection techniques it intends to use, develop acceptance 12 criteria, and provide a detailed plan for addressing any flaws 13 or indications that it may encounter. Follow-up inspections 14 should be performed at least every 10 years, given the primarily 15 Alloy 600 construction of IP2 steam generator components and 16 assemblies and the age of the IP3 steam generators. 17 In 2011, as part of this relicensing proceeding, Entergy 18 "conservatively committed to confirm the absence of PWSCC 19 indications during the PEO." Entergy Testimony at A147 20 (ENT000699). NRC should condition license renewal upon Entergy 21 fulfilling that commitment. 22 Finally, I reserve the right to supplement my testimony if 1 new information is disclosed or introduced. 2 3 4 5 6 7 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. September 9, 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 16 1 Executed in Accord with 10 C.F.R. § 2.304(d) 2 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 September 9, 2015 5