ML12171A537
| ML12171A537 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/14/2012 |
| From: | Duquette D State of NY, Office of the Attorney General |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| Shared Package | |
| ML12171A508 | List: |
| References | |
| RAS 22624, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01 | |
| Download: ML12171A537 (30) | |
Text
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 1
UNITED STATES 1
NUCLEAR REGULATORY COMMISSION 2
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 3
x 4
In re:
Docket Nos. 50-247-LR; 50-286-LR 5
License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 6
Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 7
Entergy Nuclear Indian Point 3, LLC, and 8
Entergy Nuclear Operations, Inc.
June 14, 2012 9
x 10 PRE-FILED WRITTEN 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 testimony by Dr. David J. Duquette regarding Contention NYS-16 38/RK-TC-5.
17 Q.
Please state your name and address.
18 A.
David J. Duquette, Materials Engineering Consulting 19 Services, 4 North Lane, Loudonville, New York 12211.
20 Experience 21 Q.
What is your educational background?
22 A.
My educational and professional experience are 23 NYS000372 Submitted: June 19, 2012
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 2
detailed in the attached curriculum vitae (CV)(Exhibit 1
NYS000166) and report (Exhibit NYS000372); also attached is a 2
list of my publications, awards, and other professional 3
activities. I am a graduate of the United States Coast Guard 4
Academy and the Massachusetts Institute of Technology (MIT). I 5
performed my graduate work at the Corrosion Laboratory at the 6
Massachusetts Institute of Technology, spent two years as a 7
Research Associate at the Advanced Materials Research and 8
Development Laboratory at Pratt and Whitney Aircraft prior to 9
joining the faculty at Rensselaer Polytechnic Institute.
10 Q.
What is your professional experience, particularly as 11 it relates to corrosion prevention?
12 A.
My research is primarily in the area of corrosion 13 science and engineering. I have supervised more than 50 14 graduate research dissertations in corrosion and related 15 sciences. I am the author or co-author of more than 230 16 publications and 20 book chapters. I present invited lectures 17 internationally 20 to 25 times per year. Last year, I completed 18 nine years of service on the United States Nuclear Waste 19 Technical Review Board, having been appointed to the Board by 20 President Bush in 2002. The Nuclear Waste Technical Review 21 Board was created by Congressional legislation to provide 22 scientific oversight and advice on spent nuclear fuel and high 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 3
level nuclear waste and reports to the U.S. Congress and U.S.
1 Secretary of Energy. I also maintain an active consulting 2
practice, primarily in the area of corrosion and mechanical 3
failures. A list of my publications is attached at Exhibit 4
NYS000166.
5 Q.
Can you cite specific examples of recognition by the 6
scientific community?
7 A.
I have been elected a Fellow of three learned 8
societies, ASMI (formerly the American Society of Metals), NACE 9
(formerly known as the National Association of Corrosion 10 Engineers) and ECS (the Electrochemical Society). I have 11 received the Whitney Award from NACE for outstanding corrosion 12 research, an A.V. Humboldt Senior Scientist Award from the 13 German government, as well as other awards from the scientific 14 community.
15 Q.
Do you have experience with respect to nuclear power 16 plants or systems?
17 A.
Yes.
18 Q.
Please describe that experience.
19 A.
I have served on Electric Power Research Institute 20 (EPRI) panels for corrosion control in nuclear power systems, 21 and was funded by EPRI for 5 years and by the Department of 22 Energy for 11 years for corrosion research in nuclear systems.
23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 4
I have supervised Ph.D. students performing research on nuclear 1
systems for U.S. Navy applications at the Knolls Atomic Power 2
Laboratory located in upstate New York. As part of my work, I 3
have also had personal tours of numerous reactors and related 4
waste facilities because of my service on the Nuclear Waste 5
Technical Review Board. These reactors included Dresden, 6
Savannah River, Hanford, several French plants, as well as 7
plants in England, Germany, Spain, and Argentina. In each of 8
those tours high level aspects of technical management of the 9
facilities, including aging and maintenance of the 10 infrastructures were discussed in detail.
11 I have experience with materials degradation and corrosion 12 issues in nuclear plants including consultation for Three Mile 13 Island Unit 1, the closure of Three Mile Island Unit 2, Diablo 14 Canyon Unit 1 and Unit 2 (MIC corrosion of stainless steel 15 piping), Seabrook, and the plants formerly operated by 16 Commonwealth Edison at Braidwood, Byron, Clinton, Dresden, 17 LaSalle, Quad Cities, and Zion.
18 Q.
Do you have experience with respect to steam 19 generators at nuclear power plants?
20 A.
Yes.
21 Q.
Please describe that experience.
22 A.
I have examined the issue of corrosion and materials 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 5
degradation of steam generators at various U.S. nuclear plants 1
including Three Mile Island Unit 1 (stress corrosion cracking of 2
steam generator), Seabrook (corrosion of steam generator), and 3
the Commonwealth Edison plants including Braidwood, Byron, 4
Clinton, Dresden, LaSalle, Quad Cities, and Zion Unit 1 and Unit 5
2 (stress corrosion cracking in steam generators). Those 6
facilities include both pressurized water reactor and boiling 7
water reactor designs and utilize steam generators manufactured 8
by Westinghouse, General Electric, and Babcock & Wilcox.
9 Overview 10 Q.
What is the purpose of your testimony?
11 A.
The purpose of my testimony is to provide support for, 12 and my views on, an aspect of New Yorks Contention 38 (NYS-13 38), which was admitted for litigation by the Atomic Safety 14 Licensing Board (ASLB). Contention NYS-38 asserts, among 15 other things, that Entergy has not demonstrated that it has a 16 program that will manage the effects of aging of critical 17 components or systems at the Indian Point nuclear power 18 facilities and that therefore the NRC does not have a record and 19 a rational basis upon which it can determine whether to grant 20 Entergy a renewed license for the Indian Point facilities. My 21 testimony critiques Entergys proposed approach towards the age 22 related degradation of various components of Indian Point's 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 6
steam generators during the requested twenty year period of 1
extended operation.
2 Q.
I show you what has been marked as Exhibit NYS000372.
3 Do you recognize that document?
4 A.
Yes. It is a copy of the report that I prepared for 5
the State of New York in this proceeding concerning Contention 6
NYS-38/RK-TC-5. The report reflects my review of various 7
documents and my analysis and opinions.
8 Q.
What, in general terms, does this report consist of?
9 A.
This report contains a discussion of my experience, a 10 description of Indian Point's nuclear steam supply systems 11 including the steam generators, a review of stress corrosion 12 cracking and its interaction with certain alloys and welds in 13 nuclear power plants, stress corrosion cracking concerns for 14 nuclear power plant steam generator divider plate assemblies and 15 tube-to-tubesheet welds, Entergy's proposed approach to these 16 concerns, and my opinions and conclusions concerning that 17 approach.
18 Q.
Have you reviewed materials in preparation for your 19 testimony?
20 A.
Yes.
21 Q.
What is the source of those materials?
22 A.
Many are documents prepared by government agencies, 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 7
peer reviewed articles, or documents prepared by Entergy or by 1
the nuclear power industry.
2 Q.
What materials have you reviewed in preparation for 3
your testimony?
4 A.
Among the materials I have reviewed are portions of 5
Entergys License Renewal Application for Indian Point Unit 2 6
and Unit 3 related to the aging management review and aging 7
management programs for steam generators; communications between 8
Entergy and NRC Staff concerning steam generators; the 2011 9
Supplemental Safety Evaluation Report (SSER) for the renewal of 10 the Indian Point operating licenses prepared by NRC Staff; a 11 document known as the Generic Aging Lessons Learned Report 12 (GALL), Final Report (including Revision 1 and Revision 2), a 13 document known as the Standard Review Plan (both Revision 1 and 14 Revision 2); numerous industry documents including Electric 15 Power Research Institute (EPRI), Westinghouse, Nuclear Energy 16 Institute (NEI) documents; scientific and engineering 17 literature, and NRC documents; as well as disclosures in this 18 proceeding related to steam generators.
19 Q.
I show you NYS Exhibits NYS00146A-NYS146C [GALL Rev 1]
20 NYS00147A-NYS00147D [GALL Rev 2], NYS000151 [NL-11-032],
21 NYS000152 [NL-11-074], NYS000153 [NL-11-090] NYS000154 [NL 22 096], NYS000160 [SSER], NYS000161 [SRP Rev 2], NYS000195 [SRP 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 8
Rev 1], NYS000199 [Feb. RAI], and NYS000375 through NYS000394.
1 Do you recognize these documents?
2 A.
Yes. These are true and accurate copies of the 3
documents that I referred to, used and/or relied upon in 4
preparing my report and this testimony. In some cases, where 5
the document was extremely long and only a small portion is 6
relevant to my testimony, an excerpt of the document is 7
provided. If it is only an excerpt, that is noted on the first 8
page of the Exhibit or its description.
9 Q.
How do these documents relate to the work that you do 10 as an expert in forming opinions such as those contained in this 11 testimony?
12 A.
These documents represent the type of information that 13 persons within my field of expertise reasonably rely upon in 14 forming opinions of the type offered in this testimony.
15 Q.
Did you review anything else in preparing your report 16 or this testimony?
17 A.
Yes, I reviewed other documents Entergy produced in 18 this proceeding as of early June and concluded that they were 19 not relevant in preparing my report and this testimony.
20 Conclusions and Opinions 21 Q.
What conclusions, if any, have you reached?
22 A.
In my professional judgment, and as I describe in more 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 9
detail below, and in my report, based on a review of documents 1
provided by Entergy and NRC Staff as of early June as well as 2
industry and engineering literature, a serious concern exists 3
about potential cracking in the divider plate assemblies and 4
tube-to-tubesheet welds of the Westinghouse steam generators at 5
Indian Point Unit 2 and Unit 3. Recent experience in similar 6
steam generators in Europe has discovered primary water stress 7
corrosion cracking (PWSCC) in Alloy 600 divider plates and in 8
the Alloy 82/182 welds connecting the divider plates to the 9
tubesheets. If cracks in the divider plates or in the divider 10 plate welds propagate into the Alloy 600 cladding of the 11 tubesheets it is likely that they will propagate into the tube-12 to-tubesheet welds and accordingly compromise the pressure 13 boundary, resulting in contamination of the secondary water with 14 primary water.
15 At the present time there is no qualified inspection 16 procedure to determine the extent of cracking in the divider 17 plates or associated channel head assemblies or the propagation 18 of cracking from the tubesheet cladding to the tube-to-tubesheet 19 weld. European inspection procedures result in high radiation 20 doses for plant workers/inspectors. According to an August 4, 21 2011 NEI Steam Generator Task Force presentation to NRC, the 22 stresses that may initiate PWSCC or to lead to the propagation 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 10 of either PWSCC or fatigue cracks from the divider plate 1
assemblies into the tube-to-tubesheet welds have not even been 2
determined.
3 EPRI has recently begun a program to determine the 4
susceptibility of divider plates and related structures and 5
assemblies to PWSCC, but the results of that research are not 6
scheduled to be available until 2016, well into the periods of 7
extended operation for Indian Point Unit 2 and Unit 3.
8 Entergy's proposed plan for steam generator divider plate 9
assemblies, tubesheets, and welds contains several unknowns. At 10 present, neither Indian Point nor NRC, EPRI, or the industry 11 have demonstrated that the age related degradation of divider 12 plate assemblies, tubesheets, and welds can be adequately 13 managed.
14 Until the magnitude of the problem is assessed and a 15 qualified inspection program is developed, the Entergy Aging 16 Management Program at Indian Point cannot be considered adequate 17 to assure the safety of the site to workers at the facility and 18 to the general public.
19 Stress Corrosion Cracking 20 Q.
What is stress corrosion cracking?
21 A.
Stress corrosion cracking (or SCC) is a well-22 documented phenomenon for many alloy/environmental combinations.
23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 11 It is a particularly insidious phenomenon since it occurs in 1
otherwise ductile alloys, but only in very specific 2
environments. Occurrence of the phenomenon requires the 3
simultaneous presence of stress, whether residual or applied, 4
and a specific alloy /environment combination. The phenomenon 5
is generally unpredictable for new combinations of alloys and 6
environments and is often only identified through experience.
7 It was originally called pure water stress corrosion cracking 8
and was later relabeled as primary water stress corrosion 9
cracking (or PWSCC).
10 Q.
Can you briefly describe the experience of stress 11 corrosion cracking in the nuclear energy production area?
12 A.
Yes. Cracking of Alloy 600 steam generator tubes was 13 originally observed in the vicinity of the tubesheets and tube 14 support plates in steam generators because of the expansive 15 characteristics of the corrosion products of the carbon steel 16 tubesheets and support plates in the crevices between the 17 support plates and the tubesheets and the rolled-in tubes. The 18 expansion of the corrosion products imparted large stresses on 19 the mill annealed Alloy 600 tubes resulting in plastic 20 deformation of the tubes (denting). Cracking in the deformed 21 tubes in the tubesheet region was brought under some measure of 22 control by judicious water treatment campaigns. However, 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 12 cracking in the U-bends of Alloy 600 tubes has also been 1
observed at nuclear power plants, including a documented rupture 2
of a steam generator tube at Indian Point 2 on February 15, 3
2000.
4 Since the first observations of cracked Alloy 600 5
components in nuclear reactors, and to the present day numerous 6
attempts at quantifying the specific mechanisms of the 7
susceptibility of Alloy 600 to primary water stress corrosion 8
cracking were attempted but only with limited success. It is 9
clear that metallurgical, environmental, and loading variables 10 all contribute to the susceptibility of Alloy 600 to primary 11 water stress corrosion cracking.
12 In 1985, the NRC issued a generic letter to PWR licensees 13 and potential licensees recommending actions for the resolution 14 of unresolved safety issues regarding steam generator tube 15 integrity. Some success has been achieved with specific thermal 16 treatments of the alloy, and the introduction of improved water 17 chemistries. In many cases where steam generator tubes were 18 made of Alloy 600, reactor owners replaced those steam 19 generators with steam generators with tubes that were made from 20 a more PWSCC-resistant alloy designated Alloy 690.
21 However, there are many other components in an operating 22 nuclear plant and steam supply system that still contain Alloy 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 13 600. For example, PWSCC concerns exist for steam generator 1
tubes, steam generator divider plates, heater thermal sleeves 2
and penetrations in the pressurizer, penetrations for the 3
control rod drive mechanisms in reactor pressure vessel heads, 4
and other components of the reactors that are fabricated from 5
Alloy 600. It should also be noted that Alloy 600 components 6
are generally welded with Alloys 82 or 182, derivatives of Alloy 7
600 that have also been found to be susceptible to primary water 8
stress corrosion cracking (PWSCC).
9 Indian Point Power Generation Systems & Steam Generators 10 Q.
Can you briefly describe the design of the Indian 11 Point power generation system?
12 A.
According to Entergy's License Renewal Application 13 Indian Point Unit 2 and Unit 3 each employ a pressurized water 14 reactor (PWR) design and a four loop nuclear steam supply system 15 (NSSS) furnished by Westinghouse Electric Corporation. The 16 reactor coolant system consists of four similar transfer loops 17 connected in parallel to the reactor vessel. Each loop contains 18 a reactor coolant pump and a steam generator. The system also 19 includes a pressurizer, a pressurized relief tank, connecting 20 piping, and instrumentation necessary for operational control.
21 The reactor coolant system transfers the heat generated in the 22 core of the reactor vessel to the steam generators, where steam 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 14 is produced to drive the turbine electric power generators.
1 Q.
I show you Exhibit NYS000375. Do you recognize it?
2 A.
Yes, it is a schematic drawing of a Westinghouse 3
Pressurized Water Reactor Nuclear Steam Supply System that 4
identifies the various components and the reactor coolant 5
pressure boundary.
6 Q.
Would you please describe the role of the steam 7
generators in the Indian Point nuclear steam supply systems?
8 A.
Each reactor coolant loop contains a vertical shell 9
and U-tube steam generator. Reactor coolant enters the inlet 10 side of the channel head at the bottom of the steam generator 11 through the inlet nozzle, is forced upward through the 12 tubesheet, flows through the U-tubes, returns through the 13 tubesheet to an outlet channel and leaves the generator through 14 a bottom nozzle. The inlet and outlet channels in the steam 15 generator are separated by a partition or divider plate. The 16 divider plate is joined to the channel head and the tubesheet 17 through a stub runner.
18 Q.
I show you Exhibit NYS000376. Do you recognize it?
19 A.
Yes, it is a diagram of a Westinghouse steam generator 20 that identifies the various components within a generator.
21 Q.
I show you Exhibits NYS000377 and NYS000378. Would 22 you describe them?
23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 15 A.
These are two NRC documents, one entitled History of 1
"Westinghouse Model 44 Steam Generators," and the second 2
entitled "Steam Generator Tube Operational Experience." They 3
describe the use of the Westinghouse Model 44 Steam Generator, 4
that model's use of Alloy 600 material, incidents of tube 5
ruptures of steam generators using Alloy 600, and the 6
replacement of Model 44 steam generators.
7 Q.
Are you aware of the type of steam generator that was 8
initially used at Indian Point Unit 2 and Unit 3 when they began 9
operation?
10 A.
Yes. According to Entergy and NRC documents, Indian 11 Point Unit 2 and Unit 3 were constructed with Westinghouse Model 12 44 steam generators.
13 Q.
Did there come a time when Indian Point facilities 14 changed the steam generators?
15 A.
Yes. According to Entergy and NRC documents, in 1989, 16 thirteen years after it began operations, Indian Point Unit 3 17 replaced its Westinghouse Model 44 steam generators with 18 Westinghouse Model 44F steam generators that use Alloy 690 for 19 its tubes.
20 Indian Point Unit 2 used Westinghouse Model 44 steam 21 generators from 1973 to 2000. I understand that Indian Point 22 Unit 2 received four additional Model 44 steam generators from 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 16 Westinghouse in the 1980s, but that it did not install them at 1
that time. In February 2000, a tube ruptured on steam generator 2
Number 24, and the plant shut down and remained offline for 11 3
months. During that outage, Indian Point Unit 2 replaced its 4
original Westinghouse steam generators with the ones it received 5
from Westinghouse in 1980s.
6 Q.
Has Entergy disclosed the material used in the current 7
Indian Point Unit 2 steam generators?
8 A.
Yes, Entergy has stated that the current Indian Point 9
Unit 2 steam generators use Alloy 600 for the tubes and for the 10 divider plates. It also stated that it assumed that the weld 11 material for the divider plate assemblies was Alloy 82/182 weld 12 material.
13 Q.
Has Entergy disclosed the material used in the current 14 Indian Point Unit 3 steam generators?
15 A.
Yes, Entergy has stated that the Indian Point Unit 3 16 steam generators use Alloy 690 for the tubes and Alloy 600 for 17 the divider plates. It also stated that it assumed that the 18 weld material for the divider plate assemblies was Alloy 82/182 19 weld material.
20 Q.
What types of steam generators parts or locations are 21 affected by primary water stress corrosion cracking?
22 A.
In addition to the heat transfer tubes, which we have 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 17 already discussed, primary water stress corrosion cracking could 1
also affect other components or assemblies that use Alloy 600 or 2
welds that use Alloy 82/182 weld material that, as I noted, are 3
derivatives of Alloy 600. In the August 2011 Supplemental 4
Safety Evaluation Report at page 3-21, the NRC Staff has also 5
expressed concern about the propagation of primary water stress 6
corrosion cracking in tubesheets that have Alloy 600 cladding or 7
related weld even when the heat transfer tubes are made from 8
Alloy 690TT material. According to Staff, "a crack initiated in 9
this region, close to the tube, may propagate into or through 10 the weld, causing a failure of the weld and of the reactor 11 coolant pressure boundary." These areas of concern would 12 include the channel head to tubesheet to tube complex, including 13 the divider plate assembly and the tube-to-tubesheet welds.
14 Reactor Coolant Pressure Boundary 15 Q.
Are you familiar with the term reactor coolant 16 pressure boundary?
17 A.
Yes, the NRC has a definition of this term in its 18 regulations at 10 C.F.R. § 50.2. That regulation provides:
19 "Reactor coolant pressure boundary means all those 20 pressure-containing components of boiling and 21 pressurized water-cooled nuclear power reactors, such 22
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 18 as pressure vessels, piping, pumps, and valves, which 1
are:
2 (1) Part of the reactor coolant system, 3
or 4
(2) Connected to the reactor coolant system, up to and 5
including any and all of the following:
6 (i) The outermost containment isolation valve in 7
system piping which penetrates primary reactor 8
containment, 9
(ii) The second of two valves normally closed during 10 normal reactor operation in system piping which does 11 not penetrate primary reactor containment, 12 (iii) The reactor coolant system safety and relief 13 valves."
14 Stated differently, the reactor coolant pressure boundary refers 15 to a physical barrier or boundary between the reactor coolant 16 system on the "primary loop" of nuclear steam supply system and 17 the "secondary loop" of the nuclear steam supply system. You 18 can see this boundary line in the Westinghouse NSSS diagram 19 (Exhibit NYS000375) that represents the primary loop in red or 20 yellow and the secondary loop in green or blue. It is critical 21 not to breach the reactor coolant pressure boundary and allow 22 reactor coolant to escape.
23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 19 Q.
Did Entergy's License Renewal Application discuss the 1
function of the steam generators' components?
2 A.
Yes, in the License Renewal Application Tables 2.3.1-3 4-IP2/IP3 of the LRA Entergy states that the channel head, the 4
divider plate, tubes, and the tubesheet each constitutes a 5
pressure boundary for Indian Point Unit 2 and Indian Point Unit 6
- 3. They indicated that the tubes also perform a heat transfer 7
function. Those tables are located in the License Renewal 8
Application at pages 2.3-36, 2.3-39, respectively.
9 Current Concerns About Primary Water Stress Corrosion Cracking 10 Q.
I show you Exhibit NYS000199; do you recognize it?
11 A.
Yes, this is a set of questions prepared by NRC Staff 12 and sent to Entergy in February 2011.
13 Q.
Directing your attention to page 9, would you please 14 read aloud the last full paragraph?
15 A.
Yes.
16 "In some foreign steam generators with a similar 17 design to that of Indian Point Units 2 and 3 steam 18 generators, extensive cracking due to PWSCC has been 19 identified in SG divider plate assemblies made with 20 Alloy 600, even with proper primary water chemistry.
21 Specifically, cracks have been detected in the stub 22 runner, very close to the tubesheet/stub runner weld 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 20 with depths of almost a third of the divider plate 1
thickness. Therefore, the staff noted that the Water 2
Chemistry Control - Primary and Secondary Program may 3
not be effective in managing the aging effect of 4
cracking due to PWSCC in SG divider plate assemblies."
5 Q.
I show you Exhibit NYS000160. Do you recognize 6
it?
7 A.
Yes, it is a copy of the NRC Staff's Supplemental 8
Safety Evaluation Report that they issued at the end of August 9
2011. Among other things, at pages 3-18 to 3-19 and 3-20 to 3-10 23, it discusses Entergy's revised proposal concerning the 11 Westinghouse steam generator divider plates assemblies and the 12 tube-to-tubesheet welds.
13 Q.
Directing your attention to the third paragraph on 14 page 3-18, would you please read that aloud?
15 A.
Yes.
16 "The staff noted that, although these SG divider plate 17 assembly cracks might not have a significant safety 18 impact in and of themselves, these cracks could affect 19 adjacent items that are part of the reactor coolant 20 pressure boundary, such as the tubesheet and the 21 channel head, if they propagate to the boundary with 22 these items. The staff further noted that for the 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 21 tubesheet, PWSCC cracks in the divider plate 1
assemblies fabricated from Alloy 600 and its 2
associated weld metals could propagate to the 3
tubesheet cladding, with possible consequences to the 4
integrity of the tube-to-tubesheet welds.
5 Furthermore, for the channel head, the PWSCC cracks in 6
the divider plate assemblies could propagate to the SG 7
triple point (i.e. the point where the divider plate 8
and tube sheet meet with the shell) and potentially 9
affect the pressure boundary of the SG channel head."
10 Q.
Directing your attention to the second, third, and 11 fourth sentences in the first full paragraph on page 3-21, would 12 you please read that aloud?
13 A.
Yes.
14 "The staff's concern is that, if the tubesheet 15 cladding is Alloy 600 or the associated Alloy 600 weld 16 materials, the region of the autogenous tube-to-17 tubesheet weld may have insufficient chromium content 18 to prevent initiation of PWSCC, even when the SG tubes 19 are made from Alloy 690TT. Consequently, a crack 20 initiated in this region, close to a tube, may 21 propagate into or through the weld, causing a failure 22 of the weld and of the reactor coolant pressure 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 22 boundary (RCPB). This could occur in once-through 1
SGs, as well as in recirculating SGs such as those 2
used at both of the applicant's units."
3 Q.
Are these NRC statements consistent with your 4
understanding of the recent experience with stress corrosion 5
cracking?
6 A.
Yes. As discussed in more detail in my accompanying 7
report, recent EPRI reports and other documents have begun to 8
report incidents of primary water stress corrosion cracking in 9
steam generators with a similar design to the Indian Point Unit 10 2 and Unit 3 steam generators.
11 EPRI and Westinghouse have cited reports of cracking in the 12 divider plate assemblies in French steam generators (Saint 13 Laurent, Gravelines, Chinon) and in a Swedish steam generator 14 (Ringhals) that have similar design and construction details to 15 U.S. reactors. The cracking has been observed in the divider 16 plate itself, in the full penetration welds connecting the stub 17 runner to the tubesheet and connecting the stub runner to the 18 divider plate. In the French steam generators, the cracks are 19 reported to have occurred in the heat affected zone of the stub 20 runner to divider plate weld and have been observed to run 21 nearly the length of the divider plate (~6 feet). Perhaps of 22 more concern, as the cracks approach the triple point of the 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 23 tubesheet-channel head complex, the cracks tend to curve 1
upwards. It has been suggested that this PWSCC could compromise 2
the pressure boundary of the steam generator by propagating 3
through the channel head via corrosion fatigue after the PWSCC 4
crack has initiated. Cracks that form in the divider plate, the 5
stub runner, and/or the associated welds may propagate into the 6
tubesheet, allowing mixing of the primary water with the 7
secondary water and accordingly compromising the integrity of 8
the reactor coolant pressure boundary. Given the crack path, 9
another possibility is propagation of PWSCC into the tubesheet 10 cladding that would then propagate into the tube to tubesheet 11 weld and subsequently into the Alloy 600 tubes. This phenomenon 12 is of particular concern for the IP2 replacement steam 13 generators that were constructed in the 1980s with Alloy 600 14 tubes. Moreover, the steam generators at both IP2 and IP3 have 15 Alloy 600 divider plates and Alloy 82/182 welds.
16 Entergy's Proposed Approach 17 Q.
Following NRC Staff's question in 2011 what, if 18 anything, did Entergy propose to do concerning the reports about 19 primary water stress corrosion cracking?
20 A.
Among other things, in its March 28, 2011 submission 21 (Exhibit NYS000151), Entergy told NRC that:
22 "The industry plans are to study the potential for 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 24 divider plate crack growth and develop a resolution to 1
the concern through the EPRI Steam Generator 2
Management Program Engineering and Regulatory 3
Technical Advisory Group. This industry-lead effort 4
is expected to begin in 2011 and be completed within 5
two years."
6 Acknowledging that the EPRI investigation of the issue is under 7
development and not yet completed, Entergy also stated that it 8
would "inspect all Indian Point steam generators to assess the 9
condition of the divider plate assembly. The examination 10 technique used will be capable of detecting PWSCC in the steam 11 generator divider plate assembly welds."
12 Q.
Did Entergy provide any further information about the 13 inspections it proposed to perform on the steam generator 14 divider plate assemblies?
15 A.
No, it did not.
16 Q.
Did Entergy make any proposals with respect to Staff 17 concern about the propagation of primary water stress corrosion 18 cracking in a steam generator that contains Alloy 690TT heat 19 transfer tubes?
20 A.
Yes. In 2011, Entergy proposed two options as 21 follows:
22 Option 1 (Analysis) 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 25 IPEC will perform an analytical evaluation of the 1
steam generator tube-to-tubesheet welds in order to 2
establish a technical basis for either determining 3
that the tubesheet cladding and welds are not 4
susceptible to PWSCC, or redefining the pressure 5
boundary in which the tube-to-tubesheet weld is no 6
longer included and, therefore, is not required for 7
reactor coolant pressure boundary function. The 8
redefinition of reactor coolant pressure boundary must 9
be approved by the NRC as part of a license amendment 10 request.
11 Option 2 (Inspection) 12 IPEC will perform a one-time inspection of a 13 representative number of tube-to-tubesheet welds in 14 each steam generator to determine if PWSCC cracking is 15 present. If weld cracking is identified:
16
- a.
The condition will be resolved through repair or 17 engineering evaluation to justify continued service, 18 as appropriate, and 19
- b.
An ongoing monitoring program will be established 20 to perform routine tube-to-tubesheet weld inspections 21 for the remaining life of the generators.
22 Q.
Have you seen any information about the status of the 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 26 EPRI investigation into the issue of primary water stress 1
corrosion cracking in divider plate assemblies?
2 A.
Yes, I have read recent EPRI documents that report 3
that the investigation and development of a proposed management 4
program may have not be completed until at least 2016 at which 5
point both Indian Point Unit 2 and Indian Point Unit 3 will be 6
beyond their initial 40 year operating license term.
7 Q.
I show you Exhibits NYS000393 and NYS000394? Do you 8
recognize them?
9 A.
Yes, these documents are two EPRI documents. The 10 first is entitled "Nuclear Sector Roadmaps" (January 2012) and 11 contains sections entitled "In Use: Aging Management of Alloy 12 600 and Alloy 82/182 in the Steam Generator Channel Head 13 Assembly" and "Materials Aging And Degradations, Action Plan 14 Roadmap Summary." The second is entitled "EPRI, 2012 Research 15 Portfolio, Steam Generator Management." These documents discuss 16 the EPRI investigation and its timeline.
17 Q.
Directing your attention to Exhibit NYS000393 would 18 you summarize the section entitled "In Use: Aging Management of 19 Alloy 600 and Alloy 82/182 in the Steam Generator Channel Head 20 Assembly"?
21 A.
In this section EPRI acknowledges that primary water 22 stress corrosion cracking (PWSCC) that initiates in Alloy 600 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 27 and associated weld materials in the steam generator could 1
propagate to pressure boundaries such as the tube-to-tubesheet 2
weld or the carbon steel materials in the bowl. The document 3
further acknowledges that the industry lacks understanding of 4
the impact of cracks that may compromise safe operations as the 5
steam generators age, and proposes a research program to address 6
this issue in AMRs. EPRI also admits that there are no 7
qualified techniques to inspect the steam generator channel 8
head, and that the inspection methods currently used in Europe 9
to inspect the steam generator divider plates result in 10 significant doses to workers.
11 Q.
Do you believe an aging management program is 12 necessary to manage the primary water stress corrosion cracking 13 aging degradation of steam generators at Indian Point?
14 A.
Yes. It is important to develop an Aging Management 15 Program with substantive, meaningful, and enforceable standards.
16 The fact that Indian Point in the past has experienced some 17 primary water stress corrosion issues with Alloy 600 material in 18 its steam generators, detailed in my report, indicates to me 19 that there are already corrosion risks at the facility and that 20 appropriate measures must be taken to prevent steam generator 21 components from failing in the future.
22 Q.
Has Entergy agreed that it needs an Aging Management 23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 28 Program at Indian Point to address primary water stress 1
corrosion cracking of divider plate assemblies and its 2
propagation to pressure boundaries such as the tube-to-tubesheet 3
weld or the carbon steel materials in the bowl?
4 A.
Ostensibly it has, but from a practical and 5
engineering perspective it has not, nor has it addressed safety 6
considerations. Entergy has not proposed a specific inspection 7
procedure for Indian Point except to say that it will be guided 8
by industry standards. Industry standards have not yet been 9
established. Entergy's proposed plan for steam generator 10 divider plate assemblies, tubesheets, and welds contains several 11 unknowns. It provides no detail about the inspection methods or 12 technique (visual, volumetric, or surface inspection),
13 acceptance criteria, monitoring and trending protocols, or 14 corrective action responses that it might employ. The absence 15 of such details prevents meaningful evaluation of the proposed 16 approach. At present, Indian Point (and NRC) have not 17 demonstrated that the age related degradation of divider plate 18 assemblies, tubesheets, and welds resulting from primary water 19 stress corrosion cracking can be adequately managed.
20 Q.
What, generally, is your conclusion about the adequacy 21 of Entergys proposal for steam generators primary water stress 22 corrosion cracking at Indian Point?
23
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 29 A.
While Entergy advanced a proposal it is not an Aging 1
Management Program. There is nothing in the proposal at all to 2
determine what Entergy is committing to do. It is wholly 3
deficient. It seems more like a "wait and see" placeholder 4
proposal while EPRI works on the question.
5 Q.
Are you aware that NRC Staff, in its August 2011 6
Supplemental Safety Evaluation Report, found that Entergy's 7
proposed approach "acceptable" and concluded that Entergy has 8
demonstrated that the effects of primary water stress corrosion 9
cracking in the divider plate assemblies in the steam generators 10 will be adequately managed so that their intended functions 11 would be maintained during the requested period of extended 12 operation?
13 A.
Yes, I am aware that it what NRC Staff stated, and for 14 the reasons stated in my testimony and my report, I do not and 15 cannot agree with that conclusion.
16 Q.
Have you now completed your initial testimony 17 regarding contention NYS-38/RK-TC-5?
18 A.
Yes. However, I reserve the right to express further 19 opinions if new evidence is introduced or disclosed.
20 21
Pre-filed Written Testimony of David J. Duquette Contention NYS-38/RK-TC-5 30 UNITED STATES 1
NUCLEAR REGULATORY COMMISSION 2
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 3
x 4
In re:
Docket Nos. 50-247-LR; 50-286-LR 5
License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 6
Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 7
Entergy Nuclear Indian Point 3, LLC, and 8
Entergy Nuclear Operations, Inc.
June 14, 2012 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 testimony and my 13 statement of professional qualifications are true and correct to 14 the best of my knowledge and belief.
15 Executed in Accord with 10 C.F.R. § 2.304(d)
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 June 14, 2012