ML15163A154
| ML15163A154 | |
| 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: ML15163A154 (30) | |
Text
NYS000532 Submitted: June 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. June 8, 2015 10 -----------------------------------x 11 PRE-FILED WRITTEN SUPPLEMENTAL 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. This testimony supplements my initial and rebuttal 23 testimony on Contention NYS-38/RK-TC-5. It has been 1
1 approximately three years since I provided my initial pre-filed 2 testimony in this matter and two and a half years since I 3 provided rebuttal testimony. The State of New York has asked me 4 to review the record on Contention NYS-38/RK-TC-5 and respond to 5 recent information and events.
6 Q. What documents did you review in preparation for this 7 supplemental testimony?
8 A. I reviewed again Entergys August 20, 2012 Statement 9 of Position Regarding Contention NYS-38/RK-TC-5 (ENT000520),
10 Entergys Pre-filed Testimony of Entergy witnesses Nelson 11 Azevedo, Robert Dolansky, Alan Cox, Jack Strosnider, Robert 12 Nickel, Ph.D., and Mark Gray regarding Contention NYS-38/RK-TC-5 13 (ENT000521), and the accompanying exhibits. I also reviewed the 14 NRC Staffs August 20, 2012 Statement of Position on Contention 15 NYS-38/RK-TC-5 (NRC000147), NRCs Pre-filed Testimony of NRC 16 Witnesses Dr. Allen Hiser and Kenneth Karwoski Concerning 17 Portions of Contention NYS-38/RK-TC-5 (NRC000161), which focuses 18 on steam generator issues, and the accompanying exhibits. 1 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
1 In addition, I also re-reviewed documents previously 2 submitted by the State on this contention including my previous 3 pre-filed testimony and report (NYS000372, NYS000373, NYS000452) 4 and exhibits (including, without limitation, NYS000375 to 5 NYS000394 and NYS000454 to NYS000463, NYS000472, NYS000146, 6 NYS000147, NYS000160). These documents include a presentation 7 from the EPRI Steam Generator Task Force (SGTF) to the NRC 8 entitled NRC/EPRI Steam Generator Task Force Meeting, dated 9 August 21, 2012 (NYS000463), an NRC chart identifying original 10 and replacement steam generators at U.S. plants prepared in 2009 11 (NYS000458), a paper numbered ICONE18-29457 entitled Inspection 12 of the Steam Generator Divider Plate, presented at the 18th 13 International Conference on Nuclear Engineering, authored by D.
14 DAnnucci and E. Lecour of Westinghouse for the May 2010 ICONE 15 meeting (ENT000526), EPRI Report 1025133, Steam Generator 16 Management Program: Assessment of Channel Head Susceptibility to 17 Primary Water Stress Corrosion Cracking, dated June 2012 18 (ENT000524), and various summary or demonstrative exhibits 19 prepared by the State (NYS000454 to NYS000456).
20 In addition, I reviewed a summary chart identifying the 21 materials used in the eight steam generators at Indian Point 22 Unit 2 and Unit 3 (NYS000560), a 2014 EPRI report of cracking in 3
1 steam generator channel head assemblies (NYS000544A-D), a 2012 2 Westinghouse Nuclear Safety Advisory Letter (NYS000549); various 3 NRC/EPRI Steam Generator Task Force presentations (NYS000546 and 4 NYS000550); steam generator tube inspection reports (NYS000543 5 and NYS000537); integrated inspection reports (NYS000536 and 6 NYS541); an in-service inspection summary (NYS000540); steam 7 management program documents (NYS000533, NYS000534, NYS000554, 8 NYS000555); commitment closure and verification forms (NYS000535 9 and NYS000553); NRC information notices and reports (NYS000551 10 and NYS000538); license amendment requests and approval letters 11 and related documents (NYS000539, NYS000542, NYS000556, and 12 NYS000547); responses to NRC requests for information 13 (NYS000545); and an NRC report on Lessons learned from San 14 Onofre (NYS000552).
15 Q. What are your overall conclusions having reviewed that 16 information?
17 A. First, I disagree with Entergy and the NRC staffs 18 suggestion set out in their testimony that divider plate 19 cracking is unlikely to occur in the future because it has not 20 been observed to date in United States-based steam generators.
21 I likewise disagree with Entergy and NRC staff's position that 22 Entergys general approach to aging management issues will 4
1 effectively provide adequate safety measures if cracking were to 2 occur. Entergys testimony reflects a trust us approach in 3 the absence of real data on the condition of the eight Indian 4 Point steam generators. Second, it is my opinion that in order 5 to adequately address aging degradation in the Indian Point 6 steam generators Entergy must unequivocally commit to and 7 establish a sufficiently detailed aging management program that 8 includes baseline and follow-up inspections of the steam 9 generator channel head and divider plate assemblies, including 10 the tube-to-tubesheet welds.
11 As discussed in my 2012 testimony, as well as that of 12 Entergy and NRC Staff, the EPRI-sponsored Steam Generator Task 13 Force is conducting an extensive research program into the 14 propagation of cracks in the divider plate assembly. I 15 understand that in October 2014, EPRI 16 17 18 . I have 19 reviewed the report, and it does not change my view that Entergy 20 must address potential primary water stress corrosion cracking 21 and fatigue cracking in the eight steam generators at Indian 22 Point before relicensing occurs. Thus, it is still my opinion 5
1 that inspections of the steam generator channel head and divider 2 plate assemblies and tube-to-tubesheet welds should be conducted 3 before Indian Point Unit 3 begins its period of extended 4 operation, and that such inspections should be conducted 5 promptly at Indian Point Unit 2, since they have not yet been 6 conducted at that facility.
7 In addition, while no industry-qualified technique for 8 inspection of the lower channel head and divider plate assembly 9 currently exists in the United States, any license renewal given 10 to Entergy for the Indian Point facilities should be contingent 11 on the companys expeditious qualification of an inspection 12 technique capable of identifying and evaluating primary water 13 stress corrosion cracking and fatigue-related cracks. Entergy 14 has identified a remote inspection technique that relies on 15 ultrasonic, visual and liquid penetrant technologies developed 16 by Westinghouse that has been used to successfully inspect 17 divider plates in French steam generators. ICONE Westinghouse 18 Paper (ENT000526). Instead of relying on the current absence of 19 a U.S. industry-qualified inspection technique as an excuse to 20 delay inspections at Indian Point 2 and Indian Point 3, Entergy 21 should conduct the necessary inspections using techniques 22 available now for detecting and evaluating cracks in the lower 6
1 channel assembly. For example, Entergy can employ the 2 Westinghouse technique pending future industry qualification, or 3 some other similarly effective technique. I note that remote 4 visual and ultrasonic inspections were used to inspect for 5 possible flaws in the tubesheet to channel head transition 6 region in Westinghouse steam generators at Wolf Creek Generating 7 Station and Surry Power Station Unit 2. NRC Information Notice 8 13-20 (NYS000538).
9 Q. Why is it important that Entergy inspect the lower 10 assemblies and the tube-to-tubesheet welds of the Indian Point 11 steam generators?
12 A. Both Entergy and the NRC staff agree that the Indian 13 Point Unit 2 and Indian Point Unit 3 steam generators have 14 divider plates that are constructed from Alloy 600 and that the 15 weld materials are also an Alloy 600 derivative (Alloy 82/182).
16 It is well known that Alloy 600 is susceptible to PWSCC. As of 17 mid-2015, the four steam generators at Indian Point Unit 2 have 18 been in use for approximately 15 years. They were installed 19 following the steam generator accident at Unit 2 in 2000. The 20 four steam generators at Indian Point Unit 3 have been in use 21 for approximately 26 years. They were installed in 1989 to 7
1 replace the original Unit 3 steam generators, which had been in 2 use for approximately 14 years at that time.
3 However, the current state of the divider plates, the stub 4 runners, the channel heads, as well as the tube-to-tubesheet 5 welds at Indian Point is largely unknown. Over the past few 6 years, based on reports of cracking in divider plate assemblies 7 in French steam generators, EPRIs SGTF has been examining the 8 susceptibility of divider plate assemblies to PWSCC and 9 investigating the possibility that stress corrosion cracking or 10 fatigue induced cracks could propagate into the pressure 11 boundary components. Entergy has stated that its approach to 12 the divider plate assembly cracking problems is not dependent on 13 the results of EPRI research, but that inspections being 14 committed to by plants with renewed licenses will occur at an 15 appropriate time, and that the Indian Point Quality Assurance 16 Program will drive appropriate safety evaluations. Without 17 specific criteria for determining appropriateness, Entergys 18 plan remains a hollow assurance that aging degradation of its 19 steam generators will be adequately managed.
20 In my June 2012 report, I pointed out that EPRI has 21 generically stated that the divider plates in United States 22 steam generators are thicker than those that have experienced 8
1 cracking in French steam generators, and that that factor alone 2 may mitigate against PWSCC initiation in United States steam 3 generators. Even if that conclusion proved to be true for some 4 or most United States steam generators, the divider plates at 5 Indian Point Unit 2 and Unit 3 are an exception to this general 6 rule. While the majority of steam generators in the United 7 States have divider plate thicknesses of approximately 1.9 8 inches, the Westinghouse Model 44F steam generators at Indian 9 Point Unit 2 and Unit 3 have plate thicknesses of 1.26 inches, 10 essentially the equivalent of the 1.3 inch thick divider plates 11 used in the French steam generators where PWSCC cracking was 12 first discovered. Thus, barring the possibility of differences 13 in loading or pre-assembly processing of the divider plates and 14 associated assemblies, the steam generators at Indian Point have 15 essentially the same sensitivity to PWSCC as the French steam 16 generators.
17 In my initial June 2012 testimony in this proceeding I 18 referred to cracking that had occurred in the steam generator at 19 Indian Point Unit 2. I agree with Entergy that replacement of 20 mill annealed Alloy 600 tubing with thermally treated Alloy 600 21 tubing may reduce (but not eliminate) the potential for PWSCC in 9
1 steam generator tubes. 2 However, no evidence has been presented 2 that the divider plate assemblies are constructed from thermally 3 treated alloys. Even if they are, the geometry of cracking that 4 has been observed in the European steam generators has occurred 5 near the welds joining the divider plates to the stub runners.
6 Welding of these components can be expected to lead to 7 dissolution of the grain boundary precipitates that are believed 8 to provide a degree of PWSCC resistance in thermally treated 9 alloys. Accordingly, the Entergy comments concerning the lack 10 of cracking in the steam generator Alloy 600TT tubes has little 11 or no relevance to the possibility of PWSCC in the divider 12 plates or stub runners - or for that matter in the tube-to-13 tubesheet welds.
14 Q. I show you what has been marked as Exhibit NYS000549.
15 Are you familiar with this document?
16 A. Yes. It is a 17 18 19 were designated proprietary 2
Often times the abbreviation TT is used to designate thermally treated components, e.g., Alloy 600 TT tubes.
10
1 by Westinghouse; Entergy subsequently disclosed these documents 2 to the State in this proceeding.
3 Q. What is the purpose of the 4 ?
5 A.
6 7
8 9
10 11 12 13 14 It was later 15 disclosed that these conditions were identified at a foreign 16 plant.
17 18 19 20 21 11
1 2 .
3 Q. What recommendations, if any, did 4
5 6 A.
7 8
9 10 11 12 13 14 .
15 Q. Has Entergy performed the inspections recommended by 16 17 A. Yes. I understand, based on Entergys 2013 and 2014 18 Integrated Inspection Reports for IP2 and IP3 (NYS000536 and 19 NYS000541), that Entergy performed remote video camera 20 inspections of the lower channel head and divider plate to 21 channel head welds for six of the eight Indian Point steam 12
1 generators (21, 22, 24, 31, 33, 34) following 2 .
3 Q. Did the NRC take any action to follow up on 4
5 A. In October 2013, the NRC issued Information Notice 6 2013-20 entitled, SG Channel Head and Tubesheet Degradation 7 (NYS000538) which addressed issues of potential corrosion and 8 degradation in channel heads and tubesheets.
9 Q. Directing your attention to Exhibit NYS000545A-D, do 10 you recognize that document?
11 A. Yes. It is a copy of 12 13 14 15 16 provided the report 17 to Entergy and possibly other reactor operators. As I noted 18 earlier, EPRI designated the document as containing proprietary 19 information.
20 Q. Does the EPRI report resolve your concerns about the 21 Indian Point steam generators?
22 A. No, it does not.
13
1 Q. Why is that?
2 A. There are several reasons. To begin with, 3
4 5
6 7 . This is a serious omission, since the steam 8 generators at IP3, installed in 1989, will be operating beyond 9 their 40 year life span towards the end of IP3s period of 10 extended operation. Cracks can experience exponential growth 11 rates in cyclically stressed materials. For example, a small 12 crack that develops during the first 25 years of an IP3 steam 13 generators life may rapidly develop into a crack that 14 compromises the integrity of a reactor pressure boundary or 15 other safety related component before the renewed licensing 16 period ends. does not provide any assurance 17 whatsoever that this scenario would not occur.
18 In addition, it appears that 19 analysis may be non-conservative because it did not take into 20 account the specific environmental conditions within the Indian 21 Point steam generators, such as high temperatures and 22 corrosivity, which are widely known to accelerate crack growth.
14
1 Any conclusions in the report based on this analysis would 2 therefore have little to no relevance to the issue of crack 3 growth in the Indian Point steam generators.
4 Furthermore, components made 5 of thermally-treated Alloy 690 (Alloy 690TT), which is more 6 PWSCC resistant than thermally-treated Alloy 600 (Alloy 600TT).
7 Since the IP2 steam generators tubes, tube-to-tube sheet welds 8 and divider plate assembly components are composed of Alloy 9 600TT, the report findings are simply inapplicable to IP2. I 10 have concerns about the condition of IP2s steam generators 11 precisely because these components are constructed of materials 12 known to be susceptible to PWSCC.
13 I am also concerned about PWSCC in Alloy 600TT components 14 and parts in IP3 steam generators. Although the tubes at IP3 15 steam generators are constructed of Alloy 690TT, the divider 16 plate assemblies are conservatively assumed to be Alloy 600TT.
17 Thus, the 18 19 20 21 Q. ?
15
1 A.
2 3
4 5
6 7 Q.
8 9
10 A. Yes, it did.
11 12 13 14 15 Q. Do you wish to comment on that?
16 A. Yes. First, I want to point out that Entergy has not 17 confirmed that the steam generators at Indian Point do not have 18 a layer of cold-work potentially susceptible to cracking. There 19 is some evidence that the tube-to-tubesheet welds in IP2 have 20 been cold-worked. For example, Westinghouses Alternative 21 Repair Criteria Analysis (WCAP-17828-NP) at pp. 2-9 and 3-7 22 (NYS000547) describes the fabrication and material properties of 16
1 the tube and tubesheet welds and states that [t]he 2 manufacturing process used to assemble a steam generator creates 3 a strain-hardened condition in the tubes. These tubes are then 4 inserted into the tubesheet bores and tack-expanded by hydraulic 5 expansion or mechanical hard rolling before being welded to the 6 tubesheet. Therefore, any cold-worked surfaces of the steam 7 generators could be vulnerable to the same conditions 8 experienced by the European reactors.
9 Moreover, I understand that the French operating experience 10 differs in various ways from the U.S. operating experience which 11 may account for slower crack growth rates observed in these 12 foreign plants. My experience with presentations by Electricite 13 de France (EdF), the operator of the steam generators in which 14 cracking of the divider plate assembly was initially observed, 15 is that, when a reactor in France encounters a limiting problem 16 with a steam generator tube, the French typically de-rate the 17 generator, meaning that they reduce the power of the system. In 18 contrast, U.S. nuclear system operators typically plug a tube, 19 meaning that the tube is taken out of service by blocking the 20 entry and exit openings, but do not reduce the power rating.
21 This means that, all other things being equal, U.S. pressurized 22 water reactor steam generators may run hotter and be subject to 17
1 greater stresses than their French counterparts. This 2 difference in operating environments can affect steam generator 3 susceptibility to PWSCC, as well as the growth rate of any 4 cracks that develop. At Indian Point, steam generators with a 5 number of plugged tubes may be more susceptible to PWSCC and 6 fatigue induced cracking than steam generators at French 7 reactors. Thus, while the French experience helped alert 8 industry and government to the potential for divider plate 9 assembly cracking under normal operating conditions in those 10 plants, the lack of significant crack growth observed at the 11 French reactors since the cracks were first reported should not 12 be interpreted to suggest that any cracks found in a U.S. plant 13 today would not propagate.
14 Q.
15 16 17 A.
18 19 20 21 22 .
18
1 2
3 4 I believe any decision to abandon inspection 5 plans is misguided. As I stated 6 earlier, regular inspections provide licensees and the NRC an 7 opportunity to gather baseline data for benchmarking objective 8 evidence of degradation and are a critical part of ensuring that 9 systems operate safely. From an engineering perspective, it 10 would be irresponsible to rely exclusively on mathematical 11 modeling data, particularly since we have seen, in both the 12 fracture toughness context (i.e., recently identified non-13 conservatism of BTP-5-3)(NYS000518-NYS000519) and the San Onofre 14 steam generator tube rupture context (NRC Review of Lessons 15 Learned at San Onofre, March 2014)(NYS000552), that models can 16 be non-conservative, unreliable or just plain wrong.
17 Q. To your knowledge, has Entergy inspected the divider 18 plate assemblies and tube-to-tubesheet welds of the Indian Point 19 steam generators?
20 A. While Entergy has performed remote video inspections 21 of the channel heads and divider plate-to-channel head welds for 22 cladding degradation and PWSCC based on Westinghouses NSAL 12-19
1 1, it appears that inspections were performed for only six of 2 the eight steam generators at Indian Point. Moreover, those 3 inspections were limited in scope and did not employ techniques 4 qualified to detect and measure cracks or flaws due to PWSCC.
5 NRC Integrated Inspection Report, May 9, 2014 at 10 (NYS000541).
6 Indeed, I do not believe Entergy used any magnification as a 7 part of its NSAL 12-1 channel head inspection, as its focus was 8 to identify . NYS000549). It 9 is my opinion that and the operating experiences at 10 Wolf Creek and Surry referenced in the NRCs Information Notice 11 13-20 (NYS000538) suggest that failure of corrosion-resistant 12 cladding in steam generators like those in use at Indian Point 13 is a potential problem requiring detailed inspection and 14 monitoring. Given the limited information available regarding 15 the current condition of the lower channel head assembly areas 16 of the eight steam generators at Indian Point, Entergy should, 17 as soon as possible, perform an initial baseline inspection of 18 IP2 and IP3 steam generator divider plate and channel head 19 assemblies and tube-to-tubesheet welds as part of the companys 20 One Time Inspection Program in order to confirm that its water 21 chemistry program is in fact effective and that primary water 22 stress corrosion cracking is not occurring. Generic Aging 20
1 Lessons Learned, Rev. 2 (2010), IV D 1-3,8. Similar to Entergys 2 In-Service Inspections, subsequent inspections of these steam 3 generator locations should be performed at least once every 10 4 years.
5 6 underscores the vulnerability of these 7 steam generators to corrosion and cracking, and the need for 8 regular inspections to maintain safe operations.
9 Finally, recent documents report that Indian Points steam 10 generators have experienced age-related degradation as a result 11 of wear associated with steam generator tube vibration, and that 12 a number of tubes have been plugged and taken out of service as 13 a result. (IP2 Steam Generator Examination Program Results 2014 14 Refueling Outage (2R21)(September 8, 2014)(NYS000543). I am 15 concerned about the numerous indications of vibration-induced 16 wear in the steam generator tubes at IP2, as documented in the 17 plants most recent tube inspection report. During the last 18 outage, Entergy plugged five tubes due to wear. We learned from 19 the San Onofre steam generator tube rupture event that wear, in 20 that case caused by fluid-elastic instability, can quickly 21 progress from flaw or crack initiation to tube failure. Unlike 22 other, longer-acting degradation mechanisms that may be 21
1 identified before they progress to a critical stage, wear can 2 under certain circumstances rapidly progress between inspection 3 intervals.
4 I also note that foreign objects were identified during 5 Entergys steam generator tube inspections. During the most 6 recent inspection, Entergy plugged at least nine tubes due to 7 foreign objects trapped inside the tubes. Foreign objects in 8 the steam generator can cause dents and dings. For example, in 9 1990, only one year after Steam Generator 34 was installed at 10 IP3, a fuel alignment pin was found partially lodged in a tube 11 end in the generator. 2007 Indian Point 3 Steam Generator 12 Program (NYS000533) at p. 13, 14. Visual examination revealed 13 that the foreign object made numerous indentations on the 14 channel head surfaces. Follow up inspections indicated that 15 impacts from loose parts resulted in deformities of some tube 16 ends. The presence of foreign objects in the Indian Point steam 17 generators and their potential to cause damage to the reactor 18 coolant pressure boundary is an important concern. According to 19 the NRCs Information Notice 2013-11 (NYS000551), cracking in 20 dented or dinged regions of Alloy 600TT tubing has been 21 reported, and this operating experience highlights the 22 importance of, and the challenges to, inspecting locations 22
1 susceptible to degradation and identifying inspection methods 2 capable of detecting that degradation. It is therefore 3 imperative that Entergy remain vigilant in its inspections of 4 the steam generator tubes, tube-to-tubesheet welds, and divider 5 plate and channel head assemblies at IP2 and IP3.
6 Q. Can you describe Entergys proposed inspection and 7 aging management program for the lower assembly area and tube-8 to-tubesheet welds in the steam generators?
9 A. It is difficult to tell exactly what Entergy has 10 unequivocally committed to do. As Ive discussed, in 2011, 11 Entergy presented two commitments regarding the steam 12 generators, Commitment 41 and Commitment 42. These commitments 13 are set out in Appendix A of the NRC Staffs 2011 Supplemental 14 Safety Evaluation Report (NYS000160), at pages A-23 and A-24.
15 Q. Can you read Commitment 41?
16 A. Commitment 41 states that, IPEC will inspect steam 17 generators for both units to assess the condition of the divider 18 plate assembly. The examination technique used will be capable 19 of detecting PWSCC in the steam generator divider plate 20 assembly. The IP2 steam generator divider plate inspections 21 will be completed within the first ten years of the period of 22 extended operation (PEO). The IP3 steam generator divider plate 23
1 inspections will be completed within the first refueling outage 2 following the beginning of the PEO.
3 Q. What is the implementation schedule for Commitment 41?
4 A. For IP2, it is after the beginning of the PEO and 5 prior to September 28, 2023. For IP3, it is prior to the end 6 of the first refueling outage following the beginning of the 7 PEO, which I understand to be around March or April 2017.
8 Q. Can you please read Commitment 42?
9 A. Commitment 42 provides that IPEC will develop a plan 10 for each unit to address the potential for cracking of the 11 primary to secondary pressure boundary due to PWSCC of tube-to-12 tubesheet welds using one of the following two options.
13 Q. What is Option 1?
14 A. Option 1, which is also referred to as the analysis 15 option, states that IPEC will perform an analytical evaluation 16 of the steam generator tube-to-tubesheet welds in order to 17 establish a technical basis for either determining that the 18 tubesheet cladding and welds are not susceptible to PWSCC, or 19 redefining the pressure boundary in which the tube-to-tubesheet 20 weld is no longer included and, therefore, is not required for 21 reactor coolant pressure boundary function. The redefinition of 24
1 the reactor coolant pressure boundary must be approved by the 2 NRC as a license amendment request.
3 Q. What is the implementation schedule for Option 1?
4 A. For IP2, implementation is prior to March 2024, and 5 for IP3, prior to the end of the first refueling outage 6 following the beginning of the PEO.
7 Q. What is Option 2?
8 A. Option 2, which is also referred to as the 9 inspection option, provides that IPEC will perform a one-time 10 inspection of a representative number of tube-to-tubesheet welds 11 in each steam generator to determine if PWSCC cracking is 12 present. If weld cracking is identified:
13 a. The condition will be resolved through repair or 14 engineering evaluation to justify continued service, as 15 appropriate, and 16 b. An ongoing monitoring program will be established to 17 perform routine tube-to-tubesheet weld inspections for the 18 remaining life of the steam generators.
19 Q. What is the implementation schedule for Option 2?
20 A. For IP2, the implementation schedule is between March 21 2020 and March 2024, and for IP3, prior to the end of the 25
1 first refueling outages following the beginning of the PEO, 2 which again, I understand to be around March or April of 2017.
3 Q. Can you summarize what Entergy has agreed to do under 4 those Commitments?
5 Q. Under Commitment 41, Entergy committed to inspect and 6 assess the condition of the divider plate assemblies in the IP2 7 and IP3 steam generators. Under Commitment 42, Entergy 8 committed to either perform an analytical evaluation or an 9 inspection of the tube-to-tubesheet welds.
10 Q. What is the status of those Commitments today?
11 A. I understand that on September 5, 2014, NRC staff 12 approved an amendment to Entergys operating license for Indian 13 Point Unit 2 so as to redefine the reactor coolant pressure 14 boundary to exclude tube-to-tubesheet welds (Amendment 277) and 15 thereby relieved Entergy of the obligation to inspect the tube-16 to-tubesheet welds. (Technical Specification Amendment 17 277)(NYS000542). As a result of that license amendment, on 18 September 17, 2014 Entergy deemed its Commitment 42 complete 19 for IP2. Commitment Closure Verification Form/ Corrective 20 Action (LR-LAR-2011-00174)(NYS000553). Based on the data 21 available today, I believe the NRC Staff was premature in 22 granting Amendment 277. The NRC and the nuclear industrys 26
1 understanding of PWSCC in the steam generator environment 2 continues to evolve. In fact, the NRC recently committed over 3 $2.3 million to fund research at Pacific Northwest National 4 Laboratories for the purpose of evaluating PWSCC in nickel-based 5 alloys used in steam generator and reactor components. NRC 6 Weekly Information Report, May 15 2015 (NYS000557).
7 For now, it appears that it is Entergys position that 8 Commitment 41 relating to the divider plate assembly inspections 9 is still open for IP2 and IP3 (Commitment 41 Closure 10 Verification Form (NYS000535), but that Commitment 42 relating 11 to tube-to-tubesheet welds is open for IP3 only (NYS000553).
12 However, it is unclear what impact will have 13 on these remaining open commitments. As I noted earlier, the 14 15 16 17 18 Although Entergy disclosed it is not 19 clear what use, if any, Entergy has made, or will make, of the 20 document. In my opinion, Entergy should not -- and cannot --
21 rely on the to avoid inspecting the channel head and 22 divider plate assemblies, including the tube-to-tubesheet welds 27
1 in the eight Indian Point steam generators. To the extent that 2 Entergy remains committed to performing inspections after 3 license renewal, potentially well into the plants periods of 4 extended operation, that is an inadequate assurance for managing 5 aging steam generators at Indian Point. Rather, Entergy should 6 affirmatively and clearly commit to performing inspections as 7 soon as possible for IP2, and certainly before the period of 8 extended operation for IP3. Additionally, Entergy must identify 9 the inspection techniques it intends to use, develop acceptance 10 criteria, and provide a detailed plan for addressing any flaws 11 or indications that it may encounter. I also recommend that 12 Entergy conduct follow-up inspections at least every 10 years, 13 given the primarily Alloy 600TT construction of IP2 steam 14 generator components and assemblies and the age of the IP3 steam 15 generators.
16 In conclusion, from my perspective in 2011 and 2012 there 17 was substantial uncertainty about what pathway Entergy would 18 pursue with respect to steam generators; moreover, essential 19 details were lacking in the various optional pathways Entergy 20 identified. The recent EPRI Report and the operating license 21 amendment have not resolved these uncertainties and unknowns.
22 Q. Does this conclude your supplemental testimony?
28
1 A. Yes. However, I reserve the right to offer further 2 opinions if new information is presented.
3 29
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. June 8, 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 Executed in Accord with 10 C.F.R. § 2.304(d) 17 18 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 19 June 8, 2015 30