ML15161A309
| ML15161A309 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/09/2015 |
| From: | State of NY, Office of the Attorney General |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| References | |
| RAS 27923, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR | |
| Download: ML15161A309 (80) | |
Text
NYS000530 Submitted: June 9, 2015 Public, Redacted 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 9, 2015 10 -----------------------------------x 11 REVISED PRE-FILED WRITTEN TESTIMONY OF 12 Dr. RICHARD T. LAHEY, JR.
13 REGARDING CONSOLIDATED CONTENTION NYS-26B/RK-TC-1B 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 testimony by RICHARD T. LAHEY, JR., PhD. regarding Consolidated 17 Contention NYS-26B/RK-TC-1B.
18 Q. Please state your full name.
19 A. Richard T. Lahey, Jr.
20 Q. By whom are you employed and what is your position?
21 A. I am retired and am currently the Edward E. Hood 22 Professor Emeritus of Engineering at Rensselaer Polytechnic 23 Institute (RPI), which is located in Troy, New York.
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 1
1 Q. Please summarize your educational and professional 2 qualifications.
3 A. I have earned the following academic degrees: a B.S.
4 in Marine Engineering from the United States Merchant Marine 5 Academy, a M.S. in Mechanical Engineering from Rensselaer 6 Polytechnic Institute, a M.E. in Engineering Mechanics from 7 Columbia University, and a Ph.D. in Mechanical Engineering from 8 Stanford University. I have held various technical and 9 administrative positions in the nuclear industry, and I have 10 served as both the Dean of Engineering and the Chairman of the 11 Department of Nuclear Engineering & Science at RPI. Previously, 12 I was responsible for nuclear reactor safety R&D (research &
13 development) for the General Electric Company (GE), and I have 14 extensive experience with both military (i.e., naval) and 15 commercial pressurized water and boiling water nuclear reactors 16 (PWR and BWR). Also, I am a member of a number of professional 17 societies and have served on numerous expert panels. I was also 18 an Editor of the international Journal of Nuclear Engineering &
19 Design, which focuses on nuclear engineering and nuclear reactor 20 safety technology. I am widely considered to be an expert in 21 matters relating to the design, operations, safety, and aging of 22 nuclear power plants.
23 Q. Which professional societies are you a member of?
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 A. I am a member of a number of professional societies, 2 including: the American Nuclear Society (ANS), where I was a 3 member of the Board of Directors and the ANSs Executive 4 Committee, and was the founding Chair of the ANSs Thermal-5 Hydraulics Division; the American Society of Mechanical 6 Engineers (ASME), where I was Chair of the Nucleonics Heat 7 Transfer Committee, K-13; the American Institute of Chemical 8 Engineering (AIChE), where I was the Chair of the Energy 9 Transport Field Committee; and the American Society of 10 Engineering Educators (ASEE), where I was Chair of the Nuclear 11 Engineering Division.
12 Q. What expert panels have you served on?
13 A. I have served on numerous panels and committees for 14 the: United States Nuclear Regulatory Commission (USNRC), Idaho 15 National Engineering Laboratory (INEL), Oak Ridge National 16 Laboratory (ORNL), National Aeronautics and Space Administration 17 (NASA), National Research Council(NRC) and the Electric Power 18 Research Institute (EPRI). I am a member of the National 19 Academy of Engineering (NAE), have been elected Fellow of both 20 the ANS and the ASME, and have been a Fulbright-Hays, Alexander 21 von Humboldt and Japanese Society for the Promotion of Science 22 (JSPS) Scholar.
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 3
1 A. Have you published any papers in the field of nuclear 2 engineering and nuclear reactor safety technology?
3 Q. Yes. Over the last 50 years, I have published 4 numerous books, monographs, chapters, articles, reports, and 5 journal papers on nuclear engineering and nuclear reactor safety 6 technology. Those articles are listed in my Curricula Vitae.
7 Q. Have you received any professional awards?
8 A. Yes, I have received many honors and awards for my 9 career accomplishments in the area of nuclear reactor thermal-10 hydraulics and safety technology, including: the E.O. Lawrence 11 Memorial Award of the Department of Energy (DOE), the Glenn 12 Seaborg Medal of the ANS and the Donald Q. Kern Award of the 13 AIChE.
14 Q. I show you what has been marked as Exhibit NYS000295.
15 Do you recognize that document?
16 A. Yes. It is a copy of my Curricula Vitae, which 17 summarizes, among other things, my experience, publications, and 18 honors & awards.
19 Q. I show you what has been marked as Exhibit NYS000299 20 to Exhibit NYS000303, and Exhibit NYS000483. Do you recognize 21 those documents?
22 A. Yes. They are copies of the seven declarations that I 23 previously prepared to date for the State of New York in this Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 4
1 proceeding. They include my initial declaration that was 2 submitted in November 2007 in support of the States petition to 3 intervene and its initial contentions, the April 7, 2008 4 declaration in support of Contention NYS-26A, the September 15, 5 2010 declaration submitted in support of the States 6 supplemental bases for Contention 25, the September 9, 2010 7 declaration submitted in support of the amended Contention 8 NYS26B/RK-TC-1B, the September 30, 2011 and November 1, 2011 9 declarations submitted in support of Joint Contention NYS-38/RK-10 TC-5, and the February 12, 2015 declaration submitted in support 11 of additional bases for Contention NYS-25 and Joint Contention 12 NYS-38/RK-TC-5.
13 Q. I show you what has been marked as Exhibit NYS000296.
14 Do you recognize that document?
15 A. Yes. It is a copy of the Report that I prepared for 16 the State of New York in this proceeding. This Report documents 17 my analysis and opinions.
18 Q. I show you what has been marked as Exhibit NYS000297.
19 Do you recognize that document?
20 A. Yes. This is a copy of a Supplemental Report that I 21 prepared for the State of New York in this proceeding that 22 addresses aspects of the revised fatigue analysis that Entergy 23 and Westinghouse prepared for certain components in the Indian Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 Point reactors. The supplemental report also documents my 2 assessment and opinions of this work.
3 Q. I show you what has been marked as Exhibit NYS000294, 4 NYS000344, and NYS000440. Do you recognize those documents?
5 A. Yes, those documents contain my previous pre-filed 6 testimony filed in December 2011 and June 2012 in support of 7 Contentions NYS-25 and NYS-26B.
8 Q. What is the purpose of your testimony?
9 A. I have been retained by the State of New York State to 10 review Entergys application to the U.S. Nuclear Regulatory 11 Commission (USNRC) and its Staff for two renewed operating 12 licenses for the nuclear power plants known as Indian Point Unit 13 2 and Unit 3. I have reviewed the License Renewal Applications 14 (LRAs) and subsequent filings by Entergy and the USNRC Staff.
15 My declarations and report discuss my concerns and opinions 16 about issuing twenty-year extended operating licenses for these 17 facilities. My testimony seeks to identify and discuss some 18 age-related safety concerns which have not yet been addressed by 19 Entergy. In my opinion these concerns must be resolved to 20 assure the health and safety of the American public, 21 particularly those in the vicinity of the Indian Point reactors.
22 Q. Have you reviewed various materials in preparation for 23 your testimony?
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 6
1 A. Yes.
2 Q. What is the source of those materials?
3 A. I have reviewed documents prepared by government 4 agencies, Entergy, Westinghouse, the utility industry, or its 5 associations (e.g., EPRI), and various related text books and 6 peer-reviewed articles.
7 Q. I show you Exhibits NYS00146A-C, NYS00147A-D, 8 NYS000160, NYS000161, NYS000195, NYS000304 through NYS000369, 9 and NYS000484 through NYS000525. Do you recognize these 10 documents?
11 A. Yes. These are true and accurate copies of some of 12 the documents that I referred to, used, or relied upon in 13 preparing my report, declarations, previous testimony, and this 14 testimony. In some cases, where the document was extremely long 15 and only a small portion is relevant to my testimony, an excerpt 16 of the document is provided. If it is only an excerpt, that is 17 noted on the first page of the Exhibit.
18 Q. I direct your attention to latter part of your Report 19 (Exh. NYS000296) entitled Reference Documents, which contains 20 a list of documents. Would you describe that list?
21 A. Yes that section of the Report lists various salient 22 documents that I referred to, used or relied on, in preparing my 23 Report and the Supplemental Report.
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 Q. I direct your attention to the latter part of your 2 February 12, 2015 Declaration (Exh. NYS000483) entitled 3 Reference Documents, which contains a list of documents.
4 Would you describe that list?
5 A. Yes, that section of the Declaration lists various 6 additional salient documents that I referred to, used or relied 7 on, in preparing my February 12, 2015 Declaration.
8 Q. How do these documents relate to the work that you do 9 as an expert in forming opinions such as those contained in this 10 testimony?
11 A. These documents represent the type of information that 12 persons within my field of expertise reasonably rely upon in 13 forming opinions of the type offered in this testimony.
14 The Indian Point Reactors 15 Q. Are you familiar with the power reactors that are the 16 subject of this proceeding?
17 A. Yes.
18 Q. Would you briefly describe them?
19 A. Entergy operates two nuclear power reactors that are 20 located in northern Westchester County near the Village of 21 Buchanan. The operating nuclear reactors are known as the 22 Indian Point Unit 2 and Indian Point Unit 3 reactors. These 23 Westinghouse-designed plants are 4-loop pressurized water Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 reactors (PWRs), and they are currently rated at power levels of 2 3,216.4 MWt. Entergy also owns another reactor at the same site.
3 That reactor is known as the Indian Point Unit 1 reactor; 4 however, that reactor has been shut down and no longer produces 5 power.
6 Operation of a Pressurized Water Reactor 7 Q. Would you briefly describe the design and operation of 8 a pressurized water reactor?
9 A. Pressurized water nuclear reactors have water (i.e.,
10 the primary coolant) under high pressure flowing through the 11 core in which heat is generated by the fission process. The 12 core is located inside a reactor pressure vessel (RPV). This 13 heat is absorbed by the coolant and then transferred from the 14 coolant in the primary system to lower pressure water in the 15 secondary system via a large heat exchanger (i.e., a steam 16 generator) which, in turn, produces steam on the secondary side.
17 These steam generator systems, which are part of the plants 18 Nuclear Steam Supply System (NSSS), are located inside a large 19 containment structure. After leaving the containment building, 20 via main steam piping, the steam drives a turbine, which turns a 21 generator to produce electrical power.
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 The reactor pressure vessel is a large steel container that 2 holds the core (i.e., the nuclear fuel); it also serves as a key 3 part of the primary coolants pressure boundary.
4 As the name Pressurized Water nuclear Reactor (PWR) 5 suggests, this reactor design uses a pressurizer on the primary 6 side that performs several functions. In particular, it 7 maintains the operating pressure on the primary side of the 8 nuclear reactor and accommodates variations in reactor coolant 9 volume for load changes during reactor operations, and during 10 reactor heat-up and cool-down. The reactor coolant also 11 moderates the neutrons produced in the core since a pressurized 12 water nuclear reactor will not function unless the neutrons are 13 moderated (i.e., slowed down due to collisions with the hydrogen 14 molecules in the primary coolant).
15 Q. I show you what has been marked as Exhibit NYS000304.
16 Do you recognize it?
17 A. Yes. It is a schematic diagram from a USNRC document 18 that identifies the relative location of various components in a 19 pressurized water nuclear reactor type of power plant including, 20 from the inside to the outside, the reactor core, reactor 21 pressure vessel, pressurizer, steam generator, containment 22 structure, turbine, and associated piping. The diagram also Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 identifies the various materials that are used or contained in 2 those components.
3 Reactor Pressure Vessel Internals 4 Q. I show you what has been marked as Exhibit NYS000306.
5 Do you recognize it?
6 A. Yes. It is a series of schematic diagrams or figures, 7 including Figure 3-5, from an Electric Power Research Institute 8 (EPRI) document known as MRP-227 that identifies various 9 components within pressurized water nuclear reactor designed by 10 the Westinghouse Company. The title of Figure 3-5 is, Overview 11 of typical Westinghouse internals.
12 Q. Please describe what is encompassed by the term 13 reactor pressure vessel (RPV) internals?
14 A. The term reactor pressure vessel internals (i.e.,
15 RVIs) includes various structures, components, and fittings 16 inside the reactor pressure vessel including the: core barrel 17 (and its welds), core baffle, intermediate shells, former 18 plates, lower core plate and support structures, clevis bolts, 19 fuel alignment pins, thermal shield, the lower support column 20 and mixer, upper mixing vanes, and the upper/lower core 21 assemblies and support column, and the control rods and their 22 associated guide tubes, plates, and welds. Reactor pressure 23 vessel internals (RVIs) also include the bolts that hold various Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 components together or to other components including: the 2 baffle-to-baffle bolts, the core barrel-to-former bolts, and 3 baffle-to-former bolts as well as the welds or weldments that 4 hold sections of these components together.
5 Q. Was the aging management of RVIs initially considered 6 as part of the LRA for the Indian Point facilities?
7 A. No, it was not. Fortunately, during the course of 8 these ASLB hearings on Indian Point the USNRC has now recognized 9 and highlighted the importance of RVIs [see, e.g., USNRC Report, 10 Final Interim Guidance LR-ISG-2011-04 Updated Aging Management 11 Criteria for Reactor Vessel Internal Components for Pressurized 12 Water Reactors, NRC-ISG-2011-04 (May 28, 2013) (NYS000524)].
13 Q. Are there any reactor components that you believe 14 should be considered as reactor vessel internals, but that 15 Entergy has claimed are not reactor vessel internals?
16 A. Yes. Entergy has argued that the control rods are not 17 reactor vessel internals. However, the control rods and their 18 associated guide tubes, plates, pins and welds are located in 19 the core region of the RPV, and the control rods are inserted 20 into the RPV through the upper head through so-called stub 21 tubes. The function of the control rods is to absorb excess 22 fission neutrons (i.e., those not needed to achieve a chain 23 reaction) so that the power level of a reactor can be Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 controlled. Accordingly, the control rods and associated 2 components are very important RPV internals and their integrity 3 is an extremely important safety concern. While the control 4 rods are moving parts and can be replaced as required, many of 5 the other associated components are not moving parts and are not 6 normally replaced. In any event, if a shock load occurs (e.g.,
7 during a LOCA or severe earthquake) any of these seriously 8 embrittled structures may fail and lead to degraded core 9 cooling. Thus, in my opinion, omitting the control rod 10 assemblies and associated fittings from an RPV internals (RVIs) 11 aging management program is a serious and indefensible omission.
12 Q. Coming back to Exhibit NYS000306, would you describe 13 the other diagrams?
14 A. Yes. They are a collection of additional schematic 15 figures from the Electric Power Research Institutes Report MRP-16 227 that provide additional detail concerning various reactor 17 pressure vessel internals and their location within the reactor 18 pressure vessel. The reactor pressure vessel internals shown 19 include the control rod guide tube assembly, the control rod 20 guide cards, guide tube support pins, the control rods, baffles, 21 formers, baffle-former assemblies, baffle-to-former bolts, 22 corner edge bracket baffle to former bolts, core barrel to 23 former bolts, baffle plate edge bolts, core support structures, Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 and various weldments, including welds within the reactor 2 pressure vessel for the core barrel plates.
3 Overview 4 Q. In your expert opinion what is the most important age-5 related safety issue associated with the relicensing of the two 6 Indian Point reactors?
7 A. My over-arching concern relates to Entergys silo 8 type approach to evaluating the impact of various aging 9 mechanisms such as embrittlement and fatigue, and the companys 10 failure to consider, as part of plant safety analyses, the 11 potential consequence of unanticipated shock loads (e.g., those 12 due to design basis accidents) on severely fatigued and 13 embrittled components. Entergy implicitly assumes that there is 14 no interplay between the various material aging degradation 15 phenomena and that degraded components will have no impact on 16 the plants ability to safely operate, particularly during 17 unanticipated shock loads. For example, Entergys fatigue 18 evaluations, performed by Westinghouse using the WESTEMS 19 computer code, used the metric CUFen to appraise environmentally 20 assisted fatigue in various reactor components. However, these 21 evaluations were quasi-static low and high cycle fatigue 22 evaluations that considered neither the effect of neutron-23 induced embrittlement nor the combined effects of fatigue damage Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 and other degradation mechanisms such as radiation enhanced 2 corrosion-induced cracking. In both the fatigue analyses and 3 the plant safety analyses, it was implicitly assumed that 4 fatigue weakened and embrittled structures, components and 5 fittings would respond to shock loads in the same way as if they 6 were ductile, which is simply not true. Also, no error analyses 7 were presented to quantify the WESTEMS predictions for the 8 various internals, piping systems and fittings even though some 9 of them were extremely close to the CUFen = 1.0 failure limit. In 10 any event, under these circumstances, various operational and 11 accident-induced shock loads could cause failures well before 12 the fatigue limit is reached (i.e., when CUFen < 1.0), and 13 therefore reliance on inspection-based fatigue monitoring does 14 not provide adequate assurance that the degraded components will 15 not fail.
16 Once again, the most serious short-coming of this siloing 17 approach is that synergistic interactions between radiation-18 induced embrittlement, corrosion-induced cracking, and fatigue-19 induced degradation mechanisms have not been considered. For 20 example, neither Entergys license renewal application nor its 21 proposed aging management plan consider the potential for, or 22 the consequences of, fatigue-induced failure of seriously 23 embrittled reactor pressure vessel internals (RVIs). Also, when Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 the plants safety analyses were done by Entergy it was 2 implicitly assumed that the in-core geometry would remain intact 3 during postulated accidents. Unfortunately, unlike ductile 4 metals, seriously embrittled and fatigued RPV internals may not 5 be able to survive the shock loads associated with significant 6 seismic events or the pressure and/or thermal shock loads 7 induced by various accidents and severe operational transients.
8 If not, they can fail and relocate, possibly causing core 9 blockages that degrade core cooling and may lead to core melting 10 and massive radiation releases.
11 Entergy has an obligation to show that its plants can be 12 safely operated beyond their 40 year design lives. I believe 13 that this will require much more study and analysis than has 14 been presented to date to identify any limiting RPV internals 15 that require repair or replacement. Nevertheless, this must be 16 done to verify that the two Indian Point reactors can be safely 17 operated for another 20 years beyond the design life of these 18 plants.
19 Q. What do you mean by synergistic interactions between 20 aging-related degradation mechanisms?
21 A. I mean that the concurrent exposure of reactor 22 components - especially RVI components - to multiple aging 23 mechanisms that occur in a reactor core (including fatigue, Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 irradiation embrittlement, and corrosion) may result in 2 cumulative material degradation that exceeds the predicted 3 combined degradation for each aging mechanism acting alone.
4 Q. Are there any studies or reports that support your 5 concern regarding synergistic aging effects?
6 A. Yes. However, the rather complex and interacting 7 metal degradation mechanisms associated with fatigue, 8 irradiation and corrosion interact is still an area of active 9 research (e.g., how fatigue-induced cracks propagate in an 10 embrittled, as opposed to ductile, metal structure). In fact, 11 the Department of Energy (DOE) and USNRC, in conjunction with 12 various national laboratories, have recently embarked on an 13 ambitious R&D program to understand and resolve issues related 14 to these interacting and synergistic effects [NUREG/CR-7153, 15 Vol. 2, Expanded Materials Degradation Assessment (EMDA), Aging 16 of Core Internals and Piping Systems (October 2014), at 1-5 17 (Exh. NYS00484A-B)]. In addition, the federal government has 18 also embarked on a fairly large research program, known as the 19 Light Water Reactor Sustainability Program, which includes 20 research into whether the different materials and LWR components 21 can continue to perform their intended function during the 22 extended operation of a nuclear reactor. [DOE, Light Water Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 Sustainability Program, Material Aging and Degradation Technical 2 Program Plan (August 2014) (Exh. NYS000485)].
3 Nevertheless, it is well known that, the effects of 4 embrittlement, especially loss of fracture toughness, make 5 existing cracks in the affected materials and components less 6 resistant to growth [USNRC Letter, Grimes to Newton, at 16 7 (Feb. 10, 2001) (Exh. NYS000324); see Stevens, Gary L.,
8 Presentation to the ACRS on Technical Brief on Regulatory 9 Guidance for Evaluating the Effects of Light Water Reactor 10 Coolant Environments in Fatigue Analyses of Metal Components 11 (December 2, 2014), at 56-58 (Exh. NYS000486); Chopra, O.K.,
12 Degradation of LWR Core Internal Materials due to Neutron 13 irradiation, NUREG/CR-7027 (Dec. 2010) (Exh. NYS000487)], and, 14 irradiation embrittlement decreases the resistance to crack 15 propagation [Westinghouse Owners Group WCAP-14577 Rev. 1-A 16 Report, at 3-2 (March 2001) (Exh. NYS00307A-D)]. Moreover, a 17 recent report, prepared by Argonne National Laboratory for the 18 USNRC, acknowledges, with respect to cast austenitic stainless 19 steels (CASS), that a combined effect of thermal aging and 20 irradiation embrittlement could reduce the fracture resistance 21 even further to a level neither of these degradation mechanisms 22 can impart alone [Chen, et al., Crack Growth Rate and Fracture 23 Toughness Tests on Irradiated Cast Stainless Steels, NUREG/CR-Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 7184 (Revised December 2014), at xv (Exh. NYS00488A-B)].
2 Indeed, nuclear industry groups have now recognized the 3 potential for synergistic aging effects in CASS RVI components 4 [EPRI, Slides, Industry-NRC Meeting on CASS Screening Criteria 5 for Thermal and Irradiation Embrittlement for BWR and PWR 6 Internals (July 15, 2014) (Exh. NYS000489)].
7 Q. Are synergistic aging effects limited to CASS 8 components?
9 A. No. All components within the RPV are subject to 10 multiple aging degradation mechanisms. Different materials may 11 undergo aging in different ways, but all materials are 12 susceptible to synergistic effects.
13 Q. Are these synergistic aging effects fully understood?
14 A. Not at all. Multiple recent reports and studies from 15 USNRC, DOE, and associated contractors recognize the lack of 16 understanding of the interrelationship between embrittlement, 17 high or low cycle fatigue, and shock loads for highly fatigued 18 and/or embrittled components made of CASS, non-cast stainless 19 steels, or other alloys. In addition, the consequences of the 20 interaction of embrittlement, fatigue, and the corrosion-induced 21 degradation of various reactor pressure vessel internals (RVI),
22 and safety-related components/systems during shock loads, 23 remains unknown [see, e.g., NUREG/CR-6909 Rev. 1 (March 2014 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 (draft) (Exh. NYS000490)), at 11 (it is not possible to 2 quantify the impact of irradiation on the prediction of fatigue 3 lives in PWR primary water environments compared to those in 4 air.); NUREG/CR-7153, Vol. 2, Expanded Materials Degradation 5 Assessment (EMDA), Aging of Core Internals and Piping Systems 6 (October 2014), at 3 (Exh. NYS00484A-B)]. The Argonne National 7 Laboratory report described above states that, no data are 8 available at present with regard to the combined effect of 9 thermal aging and irradiation embrittlement on CASS [Chen, et 10 al., NUREG/CR-7184, at xv (Exh. NYS00488A-B); see also Chopra, 11 O.K., Degradation of LWR Core Internal Materials due to Neutron 12 irradiation, NUREG/CR-7027 (Dec. 2010) (Exh. NYS000487)]. As 13 noted before, the same is also true for the interaction of 14 irradiation-induced embrittlement, corrosion, and fatigue of 15 non-cast stainless steel RVIs.
16 A recent paper presented at an MPA Seminar in Stuttgart, 17 Germany confirms that, at present, the USNRC staff does not have 18 a clear solution to the challenges posed by synergistic age-19 related degradation mechanisms [Stevens, Gary L., et al.,
20 Observations and Recommendations for Further Research Regarding 21 Environmentally Assisted Fatigue Evaluation Methods, 40th MPA-22 Seminar, Materials Testing Institute, University of Stuttgart, 23 Stuttgart, Germany (October 6-7, 2014) (Exh. NYS000491)]. A Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 recent draft report on the Effect of LWR Coolant Environments 2 on the Fatigue Life of Reactor Materials, prepared by Argonne 3 National Laboratory (ANL) and USNRC Staff, recognizes the 4 inconclusive nature of existing data on the synergistic 5 effects of irradiation and fatigue, and other aging mechanisms 6 in LWR environments, and concludes that, additional fatigue 7 data on reactor structural materials irradiated under LWR 8 operating conditions are needed. [NUREG/CR-6909, Rev. 1 (March 9 2014 [draft]), at 11 (Exh. NYS000490)]. Furthermore, during a 10 Briefing on Subsequent License Renewal to the USNRC, the 11 USNRCs Chief of the Corrosion and Metallurgy Branch, Dr. Mirela 12 Gravila, testified that the Piping and Core Internals Panel had 13 recognized significant gaps in our technical knowledge with 14 respect to the effects of irradiation-induced degradation of the 15 RVI components [Trans. of Briefing on Subsequent License 16 Renewal, at 77 (May 2014) (Exh. NYS000492)].
17 Q. With respect to the aging management of nuclear 18 facilities, how has the USNRC responded to these embrittlement 19 concerns with respect to its synergistic effects on fatigue?
20 A. Notwithstanding the significant concerns and 21 considerable uncertainty regarding synergistic aging effects, 22 the USNRC has so far declined to require that plant operators 23 repair or replace degraded systems, structures, and fittings, Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 opting instead to manage aging through periodic inspections, and 2 the use of an empirical environmental factor method (Fen) for 3 fatigue life when evaluating the in situ degradation of 4 structures and components [Stevens, et al., (October 2014), at 5 10 (Exh. NYS000491)], a method which is not necessarily 6 conservative and one that certainly does not address all the 7 synergistic effects (e.g., embrittlement) that New York State is 8 concerned about.
9 Q. Would you please explain in more detail the various 10 degradation mechanisms that you are concerned with?
11 A. Yes, let me begin with embrittlement.
12 Embrittlement 13 Q. Would you explain what embrittlement is?
14 A: Embrittlement refers to the change in the mechanical 15 properties (and structure) of materials, such as metals, that 16 can occur over time under the bombardment of neutrons. The 17 degree of exposure to neutrons is normally expressed in terms of 18 a fluence (i.e., the neutron flux times the duration of the 19 irradiation process). The extended exposure to neutrons causes 20 damage to metals and makes them more brittle so that they become 21 more susceptible to failures due to cracking or fracture. In 22 particular, this radiation-induced damage results in a decrease 23 in fracture toughness and ductility.
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 Embrittlement is an age-related degradation mechanism 2 whereby a component experiences a decrease in ductility, a loss 3 of facture toughness, and an increase in yield strength. While 4 the initial aging effect is loss of ductility and toughness, 5 unstable crack propagation is the eventual adverse aging effect 6 if a crack is present and the local applied stress intensity is 7 sufficient. Moreover, when subjected to a sufficient load, a 8 component which has been highly embrittled by neutron 9 irradiation may experience sudden, brittle fracture well before 10 a surface crack is detected. This is a particular problem for 11 the large pressure and/or thermal shock loads associated with 12 postulated accidents. For this reason, USNRC regulations set 13 forth at 10 C.F.R. § 50.61 impose fracture toughness 14 requirements and/or operating parameters to prevent brittle 15 facture of reactor pressure vessels. Indeed, NUREG-1800, Rev. 2 16 (Table 4.1-3) (Exh. NYS000161) identifies reduced fracture 17 toughness of reactor vessel internals as a candidate for a time 18 limited aging analysis. Because loss of ductility due to 19 radiation embrittlement was not considered in the design of the 20 stainless steel reactor vessels internal components(RVIs), it 21 is all the more important to evaluate the degree of 22 embrittlement of RVIs during license renewal review. [Chopra, 23 O., Public Comment on NRC-2010-0180-0001, Availability of Draft Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 NUREG-1800, Revision 2 and Draft NUREG-1801, Revision 2 (June 9, 2 2010) (Exh. NYS000493)].
3 The Consequences of Embrittlement 4 Q. Is embrittlement a concern for pressurized water 5 nuclear reactors?
6 A. Yes. For a pressurized water nuclear reactor to 7 operate safely, the metals involved need to be sufficiently 8 ductile, which means that they must be able to deform without 9 experiencing failures. When metals, such as steel, experience a 10 significant neutron fluence, which happens to the materials in 11 close proximity to the reactor core (e.g., the steel reactor 12 pressure vessels interior wall and the associated RVIs), the 13 temperature required for them to maintain sufficient ductility 14 is increased as the metal is continually bombarded by a neutron 15 flux. The temperature at which there is a marked change from 16 ductile to non-ductile behavior is often called the nil 17 ductility temperature (NDT). However, even for temperatures 18 well above the NDT, the irradiated metals continue to be damaged 19 and further embrittled due to the neutron bombardment. Indeed, 20 the neutron damage will not be annealed out (i.e., be 21 neutralized) unless the damaged metals are taken to temperatures 22 that are well above PWR operating temperatures.
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 24
1 Q. Could embrittlement impact a nuclear reactors ability 2 to respond to a transient, shock load, or an accident scenario?
3 A. Yes. Reduced ductility (or embrittlement) will 4 adversely affect a PWRs ability to withstand severe seismic 5 events and pressure and/or thermal shock loads, and thus there 6 is a threat to the integrity of highly embrittled internal 7 structures in the reactor pressure vessel. For example, during 8 a recent meeting regarding Indian Point, a member of the 9 Advisory Committee on Reactor Safeguards Plant License Renewal 10 Subcommittee expressed concern that embrittled RVI components 11 could fail during a seismic event. [Trans. of Advisory Committee 12 on Reactor Safeguards, Plan License Renewal Subcommittee, at 13 209-210 (April 23, 2015) (Exh. NYS000526)].
14 Various accidents and abnormal transients can expose a 15 reactor pressure vessel and its internal structures, components 16 and fittings (i.e., RVIs) to significant pressure and/or thermal 17 shock loads. If the reactor pressure vessels internal 18 structures (RVIs) are sufficiently degraded due to corrosion-19 induced cracking, fatigue and/or radiation-induced 20 embrittlement, these shock loads can have significant 21 consequences. Indeed, the resultant stresses from such 22 accidents may cause the RVIs to fail structurally and relocate Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 25
1 within the RPV. If so, the ability to effectively cool the 2 decay heat in the core may be lost due to core blockage.
3 One well known safety concern associated with embrittlement 4 is the ability of metals to withstand a thermal shock event. A 5 thermal shock can occur in various ways, for example: (1) during 6 loss of coolant accidents (e.g., postulated primary or secondary 7 side LOCAs), or, (2) during a reactor SCRAM (i.e., a rapid 8 insertion of the control rods which terminates the nuclear chain 9 reaction). A particularly bad LOCA event is one in which there 10 is a rapid depressurization of the secondary side (e.g., a steam 11 line break) which causes a reactor SCRAM and thus a rapid 12 cooling of the primary coolant via the steam generators. This 13 type of accident can lead to severe thermal shock of the reactor 14 pressure vessel and the associated RPV internals (RVIs).
15 Severe thermal shocks can also occur during a design basis 16 accident (DBA) LOCA event (i.e., a complete breach of main 17 coolant piping on the primary side), which rapidly depressurizes 18 the primary side and leads to the injection of relatively cool 19 emergency core coolant into the reactor pressure vessel (e.g.,
20 from the accumulators). As noted previously, this may lead to 21 the sudden fracture and relocation of highly embrittled RVI 22 structures, components and fittings, and thus impede their 23 ability to perform their intended functions, and adversely Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 26
1 impact their core-cooling functions. In the past, most of the 2 USNRCs attention has been focused on the integrity of the 3 reactor pressure vessel. However, the RVIs are much less 4 massive and are much closer to the core, and thus they suffer a 5 lot more radiation damage and embrittlement. Notably, the 6 USNRCs fluence threshold for irradiation embrittlement of the 7 reactor pressure vessel beltline is 1 x 1017 n/cm2 [10 C.F.R. Part 8 50, Appendix G; USNRC Regulatory Issue Summary 2014-11 (Exh.
9 NYS000494)]. In contrast, Westinghouse RVIs can experience 10 fluence in the range 1 x 1021 to 5 x 1022 n/cm2, or higher. Thus, 11 RVI are subject to neutron irradiation which is several orders 12 of magnitude higher than levels known to cause reduced fracture 13 toughness in reactor pressure vessel materials. [MRP 191 (Nov.
14 2006), Table 4-6 (Exh. NYS000321)].
15 Q. Are there other effects of embrittlement that can 16 compromise the ability to maintain a coolable core geometry in 17 the event of thermal or decompression shock loads following a 18 DBA LOCA?
19 A. Yes. As described previously, the synergistic 20 interactions between the metal degradation mechanisms associated 21 with fatigue, irradiation and corrosion are not well understood.
22 However, it is well known that irradiation embrittlement reduces 23 fracture toughness and decreases the resistance to crack Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 27
1 propagation in the metal. [USNRC Letter, Grimes to Newton, at 2 16 (Feb. 10, 2001) (Exh. NYS000324); Westinghouse Owners Group, 3 WCAP-14577, Rev. 1-A Report (March 2001), at 3-2 (Exh.
4 NYS000341)].
5 The radiation-induced damage to some RPV internals can be 6 extensive, since they can experience a neutron fluence of at 7 least 1023 n/cm2 at neutron energy (E) levels of E > 1 MeV (i.e.,
8 > 100 dpa) [Was (2007) (Exh. NYS000339); EPRI, Dyle (2008) 9 (Exh. NYS000322); WOG WCAP-14577 Rev. 1-A Report (March 2001) 10 (Exh. NYS000341)] by the end of life (EOL) for extended 11 operations. According to one study, the crack growth rate for 12 materials irradiated to only 3x1020 n/cm2 fluence can be up to 40 13 times higher than that for unirradiated materials [Chopra, O.K.,
14 Degradation of LWR Core Internal Materials due to Neutron 15 irradiation, NUREG/CR-7027 (Dec. 2010) (Exh. NYS000487)]. It 16 should be stressed that the fluence experienced by some RPV 17 internals is about four orders of magnitude (i.e., ~ 10,000 18 times) larger than will be experienced by the inner wall of the 19 reactor pressure vessel by the end of life (EOL) for extended 20 operations [Rao, A.S. (USNRC), Irradiation Assisted Degradation 21 of LWR Core Internal Materials; Brief Review, (Apr. 14, 2015) 22 (Exh. NYS000495)]. Thus, the RPV internals will be much more 23 embrittled than the RPV walls, which have historically been the Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 28
1 focus of USNRC embrittlement concerns. A highly embrittled RPV 2 internal component subjected to a severe earthquake or 3 thermal/decompression shock, could thus fail and relocate within 4 the RPV, which, in turn, could result in the loss of a coolable 5 core geometry.
6 GALL, Revision 1 7 Q. I show you a document marked as Exhibit NYS00146A-C 8 and entitled NUREG-1801, Revision 1, the Generic Aging Lessons 9 Learned Report, GALL. Are you familiar with this document?
10 A. Yes.
11 Q. When did the USNRC Staff release that document?
12 A. In September of 2005.
13 Q. Does NUREG-1801, Revision 1 include an aging 14 management program (AMP) for reactor pressure vessel internals 15 in a pressurized water nuclear reactor?
16 A. No. Revision 1 of NUREG-1801 includes no aging 17 management program description for PWR reactor pressure vessel 18 internals (RVIs). NUREG-1801, Revision 1,Section XI.M16, 19 entitled PWR Vessel Internals, instead defers to the guidance 20 provided in Chapter IV line items as appropriate. The Chapter 21 IV line item guidance simply recommends actions to:
22 ..(1) participate in the industry programs for 23 investigating and managing aging effects on reactor Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 29
1 internals; (2) evaluate and implement the results of the 2 industry programs as applicable to the reactor internals; 3 and, (3) upon completion of these programs, but not less 4 than 24 months before entering the period of extended 5 operation, submit an inspection plan for reactor internals 6 to the NRC for review and approval.
7 That statement appears a number of times in GALL, Revision 8 1, Chapter IV. For example, that statement appears on pages IV 9 B2-4, IV B2-5, IV B2-8, IV B2-14, IV B2-16, and IV B2-17 with 10 respect to the embrittlement of reactor pressure vessel 11 internals.
12 Q. I show you what has been marked as Exhibit NYS000313, 13 which is a July 15, 2010 submission from Entergy that forwarded 14 a document to the Atomic Safety and Licensing Board (ASLB). Do 15 you recognize the attachment to that submission?
16 A. Yes, it contains a copy of a July 14, 2010 17 communication, NL-10-063, from Entergy to the USNRCs document 18 control desk that concerns embrittlement of reactor pressure 19 vessel internals. In addition, NL-10-063 contains an 20 Attachment 1.
21 Q. Directing your attention to NL-10-063, Attachment 1, 22 page 84 of 90, what does Entergy say there about GALL, NUREG-23 1801, Revision 1 and reactor pressure vessel internals?
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 30
1 A. Entergy states that Revision 1 of NUREG-1801 includes 2 no aging management program description for PWR reactor vessel 3 internals.
4 Standard Review Plan, Revision 1 5 Q. I show you a document marked as Exhibit NYS000195 that 6 is entitled NUREG-1800, Revision 1, USNRC Staffs Standard 7 Review Plan (SRP). Are you familiar with this document?
8 A. Yes.
9 Q. When did the USNRC Staff release that document?
10 A. In September of 2005.
11 Q. Does the Standard Review Plan, Revision 1 recognize 12 that the reactor pressure vessel internals could experience 13 embrittlement?
14 A. Yes, the Standard Review Plan, Revision 1 at § 15 3.1.2.2.6 recognized that reactor pressure vessel internals 16 could experience embrittlement.
17 Q. Would you elaborate?
18 A. In § 3.1.2.2.6 on page 3.1-5, the Standard Review 19 Plan, Revision 1 states, Loss of fracture toughness due to 20 neutron irradiation embrittlement and void swelling could occur 21 in stainless steel and nickel alloy reactor vessel internals 22 components exposed to reactor coolant and neutron flux.
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 31
1 Q. Did the Standard Review Plan, Revision 1 make 2 provision for an aging management program (AMP) for reactor 3 pressure vessel internals in a pressurized water reactor?
4 A. No, it did not. At § 3.1.3.2.6, the Standard Review 5 Plan, Revision 1 stated that The GALL Report recommends no 6 further evaluation of programs to manage loss of fracture 7 toughness due to neutron irradiation embrittlement . . . That 8 statement is on page 3.1-12. This is also confirmed by § 9 3.1.2.2.6 and Table 3.1-1 which made clear that GALL and the 10 Standard Review Plan did not propose a specific aging management 11 plan and repeated the language from GALL about staying up to 12 date with industry discussions about embrittlement and 13 submitting a plan in the future for consideration by USNRC 14 Staff.
15 Entergys Opposition to NYS Contention 25 16 Q. In November 2007 you submitted a declaration in 17 support of the State of New Yorks Contention 25 concerning 18 embrittlement. Do you know if Entergy submitted a response?
19 A. Yes, Entergy did.
20 Q. What did Entergy say in its response?
21 A. Entergy opposed the admission of Contention 25 and 22 presented various arguments. One of Entergys principal 23 arguments was that stainless steel components are not Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 32
1 susceptible to a decrease in fracture toughness as a result of 2 neutron embrittlement. Entergy stated: The core barrel, 3 thermal shield, baffle plates and baffle former plates 4 (including bolts) are, however, made of stainless steel and are 5 not susceptible to a decrease in fracture toughness as a result 6 of neutron embrittlement. [Entergy January 22, 2008 Answer at 7 137]. This is a surprisingly uninformed statement from the 8 operators of a nuclear power plant. Anyway, while this may have 9 been a popular belief many years ago, it is incorrect.
10 GALL, Revision 2 11 Q. I show you a document marked as Exhibit NYS00147A-D 12 that is entitled Revision 2 of the Generic Aging Lessons Learned 13 Report or GALL. Are you familiar with this document?
14 A. Yes, I have reviewed it.
15 Q. When did the USNRC Staff release that document?
16 A. December of 2010.
17 Q. What does GALL, Revision 2 say about embrittlement?
18 A. GALL, Revision 2 includes the following statement:
19 Neutron irradiation embrittlement - Irradiation by neutrons 20 results in embrittlement of carbon and low-alloy steels. It may 21 produce changes in mechanical properties by increasing the 22 tensile and yield strengths with a corresponding decrease in 23 fracture toughness and ductility. The extent of embrittlement Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 33
1 depends on the neutron fluence, temperature, and trace material 2 chemistry. [GALL, Revision 2 at page IX-34 (Exh. NYS000147)].
3 I note that the phrase low-alloy steels includes stainless 4 steel.
5 Q. Does GALL, Revision 2 discuss the aging degradation of 6 PWR reactor pressure vessel internals?
7 A. Yes. Chapter IV and Chapter XI now discuss the aging 8 degradation of PWR reactor pressure vessel internals through 9 various aging mechanisms including embrittlement.
10 Q. What does GALL, Revision 2, Chapter IV state about 11 embrittlement of PWR reactor pressure vessel internals?
12 A. Chapter IV summarizes which reactor vessel internals 13 are subject to embrittlement (and other aging mechanisms) and is 14 organized by nuclear steam supply system vendors. There is a 15 section (B2) concerning components in nuclear steam supply 16 systems designed by Westinghouse, the company that designed 17 those systems at Indian Point Unit 2 and Unit 3. That section 18 recognizes that reactor pressure vessel internals in 19 Westinghouse-designed PWRs are subject to degradation due to 20 embrittlement. It further recognizes that for Westinghouse 21 PWRs, reactor pressure vessel internal components made of 22 stainless steel and nickel alloy experience a loss of fracture 23 toughness due to neutron irradiation embrittlement. These Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 34
1 statements appear on GALL, Revision 2 at pages IV B2-2 to IV B2-2 14.
3 Q. Directing your attention to GALL, Revision 2, pages IV 4 B2-12 and IV B2-13, do you see the items numbered IV.B2.RP-268 5 and IV.B2.RP-269?
6 A. Yes, those items concern reactor vessel internal 7 components in inaccessible locations.
8 Q. What is the aging effect or mechanism of concern?
9 A. There are a number including loss of fracture 10 toughness due to neutron irradiation embrittlement, void 11 swelling, and corrosion-induced cracking.
12 Q. And these are inaccessible RPV internals in 13 Westinghouse PWRs?
14 A. Yes.
15 Q. Does GALL Revision 2 make any suggestions about the 16 reactor pressure vessel components that are located in 17 inaccessible locations?
18 A. Yes, it recommends an evaluation of the internals 19 located in inaccessible locations if other similar components 20 indicate aging effects that need management.
21 Q. You mentioned that GALL, Revision 2, Chapter XI also 22 discussed reactor pressure vessel internals. Where is that 23 discussion?
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 35
1 A. Chapter XI contains a section numbered XI.M16A 2 entitled PWR Vessel Internals, which starts at page XI M16A-1.
3 Q. Would you summarize that section?
4 A. Yes. Like Chapter IV, it recognizes that PWR reactor 5 pressure vessel internals experience a loss of fracture 6 toughness due to either thermal aging or neutron irradiation 7 embrittlement, as well as other age-related degradation 8 mechanisms, such as various corrosion-induced cracking 9 mechanisms. It provides a template for license renewal 10 applicants to include in their license renewal applications that 11 discusses embrittlement and other aging mechanisms that degrade 12 reactor pressure vessel internals. It recommends that 13 applicants propose an inspection plan that is then submitted to 14 the USNRC Staff for review and approval. The template is 15 derived from a document prepared as a result of an effort 16 coordinated by the Electric Power Research Institute (EPRI) to 17 develop guidelines concerning the inspection of reactor pressure 18 vessel internals.
19 Q. Directing your attention to GALL, Revision 2, page XI 20 M16A-3, do you see item 3, titled Parameters Monitored/
21 Inspected?
22 A. Yes.
23 Q. Would you summarize that section?
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 36
1 A. Yes, this section provides recommendations for an 2 inspection plan for reactor pressure vessel internals, and 3 specifically what I would describe as the scope or focus of the 4 plan. This section is titled Parameters Monitored/Inspected 5 and states that the recommended inspection program does not 6 directly monitor for loss of fracture toughness that is induced 7 by thermal aging or neutron embrittlement. Instead, it states 8 that the embrittlement of reactor pressure vessel internal 9 components is indirectly monitored through visual or volumetric 10 inspection techniques that look for cracking (i.e., the 11 detection of failures after they have occurred). It is 12 important to note that the focus of this document is on non-13 destructive testing (NDT) and non-destructive evaluation (NDE) 14 techniques. In particular it does not consider the implications 15 on core coolability subsequent of any shock load induced 16 failures of highly degraded RPV internals.
17 MRP-227, Revision 0 18 Q. I show you a document marked as Exhibit NYS00307A-D.
19 Do you recognize it?
20 A. Yes, I have reviewed it. It is a copy of the document 21 prepared as a result of the nuclear industrys efforts 22 coordinated by the Electric Power Research Institute (EPRI).
23 Q. What is the title of that document?
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 37
1 A. The documents title is, Material Reliability 2 Program: Pressurized Water Reactor Internals Inspection and 3 Evaluation Guidelines (MRP-227-Rev. 0), 1016596, Final Report, 4 December 2008. Unfortunately, as I discussed previously, it is 5 focused on NDT and NDE inspection techniques rather than my 6 aging-related safety concerns.
7 MRP-227-A 8 Q. I show you a document marked as Exhibit NRC00114A-F 9 [MRP 227-A]. Do you recognize it?
10 A. Yes, I have reviewed it. It is the version of the 11 MRP-227, Revision 0 [Exh. NYS00307A-D] that was reviewed and 12 approved by the USNRC Staff, and includes various edits and 13 additional materials in response to USNRC Staff comments and 14 questions. It was submitted to the USNRC in January 2012.
15 Unfortunately, as I have noted previously, it is focused on NDT 16 and NDE inspection techniques rather than my aging-related 17 safety concerns.
18 Q. Does MRP-227-A say anything about embrittlement?
19 A. Yes. The industry has recognized that, there are no 20 recommendations for inspection to determine embrittlement level 21 because these mechanisms cannot be directly observed [MRP-227-22 A, Footnote 1 for Table 3-3 (December 2011) [Exh. NRC00114A-F].
23 That is, the level of degradation due to embrittlement of RPV Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 38
1 internal components, fittings and structures, and their ability 2 to withstand fatigue and shock loads cannot be determined using 3 the inspection techniques proposed in MRP-227-A.
4 Q. Do you have specific concerns with the approach to 5 aging management for reactor vessel internals set forth in MRP-6 227-A?
7 A. Yes. MRP-227-A is an inspection-based aging 8 management plan, which I believe is inadequate. To begin with, 9 depending on the type of component, inspection may not be 10 possible for the entire component, or for the entire set of such 11 components, given the location of the components and their 12 possible inaccessibility. For example, a visual or ultrasonic 13 inspection of the external head of a bolt does not necessarily 14 provide insight into the integrity of the remainder of the bolt 15 which is not visible. Moreover, an inspection focused on one 16 type of age-related degradation mechanism does not necessarily 17 work for another ongoing degradation process that is affecting 18 the same component, and the effect of shock loads on the 19 integrity of various RVIs and primary pressure boundary systems 20 is certainly not addressed by inspections. An inspection-based 21 approach to aging management, such as the one developed by the 22 nuclear industry in MRP-227 and condoned by USNRC in MRP-227-A, 23 is useful but it fails to account for the possibility that Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 39
1 highly embrittled and fatigued RVI components may not have signs 2 of degradation that can be detected by an inspection, but such 3 weakened components could nonetheless fail as a result of a 4 severe seismic event or thermal or pressure shock load. In 5 short, many of my concerns about the cumulative and ongoing 6 synergistic aging effects are not adequately addressed by MRP-7 227-A.
8 Entergys License Renewal Application 9 Q. Directing your attention to Entergys 2007 License 10 Renewal Application (LRA), did you find any indication in the 11 LRA that Entergy recognized that embrittlement could affect the 12 reactor pressure vessel?
13 A. Yes.
14 Q. Where was that?
15 A. The License Renewal Application at § 3.1.2.1.1 16 recognized that reactor pressure vessels are constructed of the 17 following materials:
18
- carbon steel; 19
- carbon steel with stainless steel or nickel alloy; 20
- cladding; 21
- nickel alloys; and, 22
- stainless steel.
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 40
1 The same LRA section further recognized that reactor pressure 2 vessels experience the following aging effects that require 3 management:
4
- cracking; 5
- loss of material; and, 6
- reduction of fracture toughness, a term which 7 encompasses embrittlement.
8 Q. Did you find any indication in the LRA that Entergy 9 has now recognized that embrittlement could affect reactor 10 pressure vessel internals?
11 A. Yes.
12 Q. Where was that?
13 A. The License Renewal Application at § 3.1.2.1.2 14 recognized that reactor pressure vessel internals are constructed 15 of the following materials:
16
- cast austenitic stainless steel (CASS);
17
- nickel alloy; and, 18
- stainless steel.
19 The same LRA section further recognized that the reactor 20 pressure vessel internals experience the following aging effects 21 that require management:
22
- change in dimensions; 23
- cracking; Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 41
1
- loss of material; 2
- loss of preload; and, 3
- reduction of fracture toughness, a term which, as 4 noted previously, encompasses embrittlement.
5 The 2007 LRA and the IP3 Reactor Pressure Vessel 6 Q. I direct your attention to License Renewal Application 7 Appendix A, § A.3.2.1.4. Do you have that?
8 A. Yes.
9 Q. What is that section of the License Renewal 10 Application concerned with?
11 A. That section concerns the IP3 reactor pressure vessel 12 itself.
13 Q. And what did Entergy say there?
14 A. Entergy stated that a part of the IP3 pressure vessel, 15 specifically plate B2803-3, exceeded the screening criteria for 16 pressurized thermal shock (PTS).
17 Q. Did Entergy acknowledge any specific concern about the 18 reactor pressure vessels at Indian Point?
19 A. Yes, Entergy acknowledged that with respect to IP3 20 that the reactor pressure vessel plate B2803-3 exceeds the 21 screening criterion by 9.9°F. [Entergy January 22, 2008 Answer 22 at 139; citing LRA § A.3.2.1.4].
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 42
1 Q. What if anything did Entergy propose to do about the 2 IP3 pressure vessel?
3 A. Entergy proposed to submit to USNRC Staff a safety 4 analysis for plate B2803-3 three years before the plate reached 5 the reference temperature for pressurized thermal shock (RTPTS) 6 criterion.
7 The 2007 LRA and RPV Internals 8 Q. In your review of the April 2007 Indian Point License 9 Renewal Application, did you see an aging management program 10 (AMP) for reactor pressure vessel internals?
11 A. No, I did not. The 2007 License Renewal Application 12 did not contain an aging management program that specifically 13 focused on reactor pressure vessel internals. Rather, Appendix 14 A stated that sometime in the future Entergy would develop an 15 aging management program for the reactor pressure vessel 16 internals of their plants [LRA Appendix A, § A.2.1.41 with 17 respect to IP2, and § A.3.1.41 with respect to IP3]. This 18 deferred approach concerning IP2 and IP3 reactor pressure vessel 19 internals is also repeated at LRA, § 3.1.2.2.6.
20 Q. Do reactor pressure vessels and their associated 21 internal structures, components and fittings experience 22 embrittlement?
23 A. Yes.
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 43
1 Q. Are there any reactor pressure vessel internal 2 structures that are neglected in Entergys discussion of future 3 programs it will develop to address such structures?
4 A. Yes. It should be noted that the control rods and 5 their associated guide tubes, plates, pins, and welds are not 6 highlighted, but they are also very important RPV internals and 7 their integrity is an extremely important safety concern. As I 8 have previously noted, they are located in the core region of 9 the RPV, and are inserted into the RPV through the upper head 10 via so-called stub tubes. Their function is to absorb excess 11 fission neutrons (i.e., those not needed to achieve a chain 12 reaction) so that the power level of a reactor can be 13 controlled. The control rods themselves are currently 14 considered by the USNRC to be moving components (which can be 15 replaced) and are thus not required to have an aging management 16 plan (AMP). Nevertheless, the other associated CRD structures, 17 components and fittings need an AMP since if these highly 18 embrittled structures, components and fittings are subjected to 19 significant shock loads they may fail, leading to possible core 20 cooling issues.
21 Q. Do you believe there are any special problems 22 associated with providing an adequate aging management program Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 44
1 for control rods and their associated guide tubes, plates and 2 welds?
3 A. Yes. For example, because of geometric 4 considerations, many PWRs (including IP2 and IP3) cannot meet 5 the USNRCs required minimum coverage for the non-destructive 6 testing (NDT) of the so-called J-groove welds [Entergy, 7 Walpole, NL-09-130 (Sept. 24, 2009) (Exh. NYS000311)], and thus 8 the integrity of these important CRD stub tube welds cannot be 9 directly confirmed by inspection. It appears that to help 10 address this chronic problem Entergy has ordered two new RPV 11 heads [Telecom-USNRC/Entergy Report (March 18, 2008) (Exh.
12 NYS000317)], but they have not yet been scheduled for 13 installation at Indian Point [Telecom-USNRC/Entergy (March 18, 14 2008) (Exh. NYS000317)]. In any event, unlike the rather 15 superficial treatment given this important safety concern by 16 Entergy [NL-10-063 (Exh. NYS000313)], I believe that a tangible, 17 enforceable, and viable aging management program (AMP) should be 18 developed and implemented before re-licensing the Indian Point 19 reactor plants for extended operations, since the integrity of 20 these CRD welds must be assured. If not, due to the leakage of 21 borated primary coolant through cracked welds, there can be 22 aggressive corrosion and wasting of the unclad outer surface of 23 the upper head of the RPVs (such as the serious event that Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 45
1 occurred at Davis-Besse and was identified in 2002). Worse yet, 2 there might be an inadvertent control rod ejection (due to a 3 massive failure of the welds in the upper RPV head), which could 4 cause a significant reactivity excursion, leading to core 5 melting and radiation releases.
6 Q. Are there places within the reactor pressure vessel 7 that you believe warrant particular aging management attention?
8 A. Yes. For the relicensing of the two reactors at 9 Indian Point, corrosion-induced cracking (e.g., SCC) and 10 radiation-induced embrittlement of the RPVs and their associated 11 internals is an important age-related safety concern, 12 particularly in the so-called belt line region of the RPV, 13 which is the region that is the closest to the reactor core. In 14 addition, as noted previously, the integrity of the so-called J-15 welds, which are part of the control rod drive seal in the upper 16 head of reactor pressure vessels, is important to avoid 17 corrosion-induced failures of the upper head and the possibility 18 of control rod ejection (and thus an uncontrolled reactivity 19 excursion).
20 Entergys NL-10-063 Communication 21 Q. I direct your attention to Exhibit NYS000313. Do you 22 recognize it?
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 46
1 A. Yes, I have reviewed this document. As noted above, 2 it contains a copy of a July 14, 2010 communication, NL-10-063, 3 from Entergy to the USNRC document control desk that concerns 4 embrittlement of reactor pressure vessel internals. In turn, 5 NL-10-063 contains an Attachment 1.
6 Q. Does Entergy make any statements here about 7 embrittlement of reactor pressure vessel internals?
8 A. Yes. Entergy acknowledges that, PWR internals aging 9 degradation has been observed in European PWRs, specifically 10 with regard to cracking of baffle-former bolting. [NL-10-063, 11 at 89 (Exh. NYS000313)]. Entergy also states: As with other 12 U.S. commercial PWR plants, cracking of baffle-former bolts is 13 recognized as a potential issue for the [Indian Point] units.
14 [NL-10-063, at 89 (Exh. NYS000313)]. Moreover, EPRI has stated 15 that, a considerable amount of PWR internals aging degradation 16 has been observed in European PWRs. [EPRI MRP-227, at A-4 17 (Exh. NYS00307A-D)]. Material degradation has also been 18 observed in control rod guide tube alignment (split) pins [EPRI 19 MRP-227, at A-4 (December 2008) (Exh. NYS00307A-D)]. It is 20 important to note that MRP-227 has also recommended that 21 analysis be done to show when it is acceptable to continue to 22 operate PWRs in which there have been bolt failures (e.g., due 23 to embrittlement and/or fatigue). While this type of temporary, Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 47
1 short-term fix might be adequate for normal operations, it may 2 lead to structural and component failures due to the shock loads 3 associated with various postulated accidents. If so, the failed 4 internal structures and components may relocate, cause core 5 blockages, or otherwise result in uncoolable core geometry, and 6 thus lead to seriously degraded core cooling, core melting and 7 massive radiation releases.
8 Q. Do you have any additional problems with the 9 inspection program for RVIs as proposed in the MRP-227 and 10 adopted by Entergy?
11 A. Yes. With respect to Entergys proposal to conduct 12 baseline examinations of the RPV internals (RVIs), it should be 13 noted that I have previously called on Entergy to conduct such 14 examinations and for USNRC Staff to require the conduct of such 15 examinations before entering the period of extended operations 16 [See November 2007 Declaration of Richard T. Lahey, Jr., at ¶¶ 17 24, 25 (Exh. NYS000298); see also State of New York Notice of 18 Intention to Participate and Petition to Intervene, at 217-220, 19 State of New York Contention-23 (Baseline Inspections)].
20 Fortunately, both the USNRC and Entergy now seem to have 21 embraced the concept of baseline inspections for RPV internals, 22 but the proposed aging management program (AMP) as set forth in 23 NL-10-063 lacks sufficient details to know when the baseline Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 48
1 inspections of the RPV and its internals will begin and end, and 2 the scope of these inspections. Thus, it is not possible to 3 know whether the proposed baseline inspections will be 4 comprehensive and adequate.
5 Q. Are there other problems that you believe need to be 6 addressed if Entergy is to have an adequate aging management 7 program for RPV internals?
8 A. Yes. My Report provides more details on my concerns 9 with Entergys failure to conduct an evaluation of the 10 synergistic impacts of embrittlement, corrosion-induced 11 cracking, and metal fatigue on the degradation of RPV internals, 12 and its failure to consider how those interacting degradation 13 mechanisms will impact the ability of the RPV internals to 14 withstand the effect of thermal and decompression shock loads as 15 a result of a DBA LOCA. I am also concerned that the design of 16 the inspection programs -- including their frequency, the type 17 of inspections to be conducted, the acceptance criteria and the 18 criteria for actions to be taken in the event of a failure of a 19 component -- does not consider these synergistic degradation 20 mechanisms. Finally, Entergys AMP for RPV internals does not 21 include specific programs with objective criteria for either 22 preventative measures or for corrective actions to be taken when Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 49
1 inspections show that certain components are not able to safely 2 undergo extended plant operations.
3 Entergys NL-11-107 Communication 4 Q. I show you what has been marked as Exhibit NYS000314.
5 Do you recognize that document?
6 A. Yes, this is a copy of Entergys September 28, 2011 7 communication, NL-11-107, with the USNRC's document control 8 desk.
9 Q. Would you please turn to Table 5-2 at page 36 of the 10 Attachment to NL-11-107.
11 A. Yes, I have that.
12 Q. What does the document say there?
13 A. In discussing the baffle-former assemblies and their 14 related baffle-edge bolts, it recognizes that irradiated-15 assisted stress corrosion cracking and fatigue can cause 16 cracking which, in turn, leads to failed or missing bolts 17 connecting a baffle to a former.
18 Q. What else does communication NL-11-107 state?
19 A. In it, Entergy tells the USNRC that it has completed 20 commitment number 30 wherein Entergy stated that it would submit 21 an inspection plan to the USNRC for reactor pressure vessel 22 internals (RVIs) no later than two years before the plant 23 entered the period of extended operations. However, none of my Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 50
1 safety concerns associated with the synergistic effects of 2 embrittlement, fatigue and corrosion on the integrity of RPV 3 internals, and post-accident core coolability (i.e., due to 4 shock load induced failures), were addressed. In my opinion an 5 adequate inspection plan for RPV internals is a necessary, but 6 not sufficient, means of assuring safe extended plant 7 operations. Indeed, a systematic safety evaluation of the 8 degraded RPV internals is also needed to identify the limiting 9 structures, components and fittings that need to be repaired or 10 replaced before the onset of extended operations.
11 Entergys Amended and Revised RVI Plan, and 12 USNRC Staffs November 2014 SSER2 13 14 Q. I direct your attention to Exhibits NYS000496 through 15 NYS000506. Do you recognize these exhibits?
16 A. Yes, I have reviewed these documents. In 17 communication NL-12-037, dated February 17, 2012 [Exh.
18 NYS000496], the applicant submitted an amendment to its license 19 renewal application entitled Revised Reactor Vessel Internals 20 Program and Inspection Plan. Thereafter, the applicant 21 explained and modified this proposed plan in response to various 22 requests for information (RAIs) from the USNRC. [Exhs.
23 NYS000497 through NYS000506]. Collectively, I will refer to Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 51
1 this collection of communications as the applicants Amended 2 and Revised RVI Plan.
3 Q. I direct your attention to Exhibit NYS000507. Do you 4 recognize this exhibit?
5 A. Yes, I have reviewed this exhibit. It is the Second 6 Supplemental Safety Evaluation Report, or SSER2, prepared by 7 USNRC Staff and released in November 2014. In the SSER2, the 8 USNRC Staff evaluated and approved the applicants Amended and 9 Revised RVI Plan.
10 Q. Does the SSER2 discuss the potential synergism between 11 various aging mechanisms?
12 A. Yes, to some degree. The USNRC recognized the 13 potential synergy between thermal and irradiation embrittlement 14 for cast austenitic stainless steel components (CASS). [SSER2 15 at 3-42 (Exh. NYS000507)]. In particular, in its Safety 16 Evaluation Report (SER) for MRP-227, the USNRC Staff 17 acknowledged the potential for synergistic interaction between 18 embrittlement and other aging mechanisms. For example, the 19 USNRC noted that the synergistic effects of SCC, fatigue, and 20 thermal embrittlement . . . could potentially cause greater 21 degradation in the welds [of Combustion Engineering lower 22 support columns] than just the consideration of IASCC 23 (irradiation assisted stress corrosion cracking) and irradiation Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 52
1 embrittlement alone. Degradation in these welds could then be 2 equivalent to or greater than other components susceptible only 3 to IASCC and irradiation embrittlement due to the synergistic 4 effects. [SE at 15 (Exh. NYS000309)]. The USNRC staff could 5 have - indeed, should have - made the same observation about 6 potential synergistic aging effects for Westinghouse RVI 7 components, fittings, and structures at IP2 and IP3.
8 Q. Does the applicants Amended and Revised RVI Plan say 9 anything about preventative actions to manage aging effects?
10 A. Yes. In Attachment 1 to NL-12-037, Entergy has 11 indicated that the Amended and Revised RVI Plan is a condition 12 monitoring program that does not include preventative actions.
13 [Attachment 1 to NL-12-037, at 5 (Exh. NYS000496)]. Generally, 14 the applicant continues to approach the problem of synergistic 15 aging effects on RVI components through condition monitoring 16 (i.e., periodic inspections per MRP-227-A) rather than a 17 comprehensive approach which includes detailed analyses and/or 18 preventative actions (i.e., repair and replacement) [Revised 19 Reactor Vessel Internals Program and Inspection Plan, 20 Attachment 1 to NL-12-037, at 5 (Exh. NYS000496)]. This 21 approach implies that aging effects and degradation will not be 22 addressed until cracks or other degradation mechanisms (e.g.,
23 wear) have been directly observed [Revised Reactor Vessel Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 53
1 Internals Program and Inspection Plan, Attachment 1 to NL 2 037, at 5 (Exh. NYS000496)].
3 In short, component degradation will be addressed only 4 after it occurs. The applicant incorrectly concludes that 5 preventative actions, such as component replacement, are not 6 required for most RVI components because cracking or other flaws 7 can be detected before the failure of a component affects the 8 safe operation of the reactor. This is apparently based on the 9 erroneous assumption that IP2 and IP3 will continuously operate 10 during the 20-year period of extended operation within normal 11 steady-state parameters. Entergy ignores the possibility that 12 significantly fatigued, embrittled and corrosion-weakened, or 13 otherwise degraded, RVI components, structures, or fittings may 14 be exposed to various shock loads which can cause them to deform 15 or relocate and thereby impair core cooling. In fact, the 16 applicants reactor safety analyses implicitly assume that the 17 reactor core will maintain a coolable geometry during emergency 18 core cooling system (ECCS) operation subsequent to a DBA LOCA, 19 notwithstanding the degradation and possible deformation or 20 relocation of various RVI components and potential flow 21 blockages and degraded core cooling which may result.
22 Q. Does Entergy make any statements about the degradation 23 of RVI components in the Amended and Revised RVI Plan?
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 54
1 A. Yes, similar to NL-10-063 (Exh. NYS000313), the 2 applicant acknowledges, in NL-12-037, that other PWRs have 3 experienced material degradation and failure of multiple RVI 4 components, including cracking of baffle-former bolting, 5 cracking in other important bolting, wear in thimble tubes, and 6 potential wear in control rod guide tube guide plates 7 [Attachment 1 to NL-12-037, at 8 (Exh. NYS000496)]. Also, the 8 applicant has committed to replace one affected IP2 component -
9 the degraded guide tube support pins (split pins) - by 2016 10 [Revised Reactor Vessel Internals Program and Inspection Plan, 11 Attachment 1 to NL-12-037, at 8 (Exh. NYS000496); Commitment 50, 12 Attachment 1 to NL-13-122, at 7 (Exh. NYS000502)].
13 Interestingly, the applicant has agreed to replace the IP2 split 14 pins, even though they were already replaced once in 1995, and 15 even though the applicant claims that the failure of a split pin 16 would not compromise reactor vessel functions [Response to RAI 17 16, Attachment 2 to NL-12-166, at 1 (Exh. NYS000500)]. However, 18 for many other affected RVI components, the applicant proposes a 19 wait-and-see approach.
20 Q. Could you provide an example?
21 A. Yes. The applicant acknowledges that cracking of 22 baffle former bolts is recognized as a potential issue for the 23 Indian Point units [Revised Reactor Vessel Internals Program, Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 55
1 Attachment 1 to NL-12-037, at 8 (Exh. NYS000496)], but the 2 applicant does not propose to replace the degraded bolts, only 3 to continue monitoring them [Revised Reactor Vessel Internals 4 Inspection Plan, Attachment 2 to NL-12-037, at 40, tbl. 5-2 5 (Exh. NYS000496)]. In fact, the applicant has not yet developed 6 inspection acceptance criteria for baffle former bolts in either 7 IP2 or IP3 [SSER, at 3-20 (Exh. NYS000507)]. Instead, the 8 applicant has agreed to develop a technical justification 9 including acceptance criteria for baffle former bolts sometime 10 prior to the first round of inspections, which might not occur 11 until 2019 for IP2 and 2021 for IP3 [SSER2, at 3-20 (Exh.
12 NYS000507); Response to RAI 5, Attachment 1 to NL-12-089, at 11 13 (Exh. NYS000497)].
14 Another example of the applicants wait-and-see approach 15 for the RVIs is the applicants proposal for managing aging 16 effects on the clevis insert bolts. [SSER2, at 3-23 to 3-26 17 (Exh. NYS000507)]. Like the split pins that the applicant is 18 replacing in IP2 for the second time, clevis insert bolts are 19 susceptible to primary water stress corrosion cracking (PWSCC) 20 [MRP-227-A, Appendix A, at A-2 (Exh. NRC00114A-F)]. Failures of 21 clevis insert bolts, apparently caused by PWSCC, were detected 22 at a Westinghouse-designed reactor in 2010. Out of 48 clevis 23 bolts in this reactor, 29 were partially or completely fractured Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 56
1 but only 7 of those damaged bolts were visually detected as 2 having failed [SSER2, at 3-25 (Exh. NYS000507)]. Despite this 3 high rate of failure (about 60% of the total bolts were damaged) 4 and low rate of visual detection (only about 24% of the damaged 5 bolts were detected), the applicant proposes to manage the aging 6 degradation of clevis insert bolts with visual (VT-3) 7 inspections rather than pre-emptive replacement [Revised 8 Reactor Vessel Internals Inspection Plan, Attachment 2 to NL-9 12-037, tbl. 5-4, at 51 (Exh. NYS000496)].
10 The applicant apparently acknowledges that visual 11 inspections will not detect the majority of clevis bolt cracks 12 prior to failure, but justifies this approach on the grounds 13 that crack detection prior to bolt failure is not required due 14 to design redundancy [Response to RAI 17, Attachment 1 to NL-15 13-122, at 8 (Exh. NYS000502)]. In fact, the applicant appears 16 to suggest that the failure of multiple clevis insert bolts will 17 not seriously affect the operation of the reactor. The 18 applicant then analyzes the effect of clevis bolt failures on 19 various other components.
20 The applicants analysis of the effects of clevis bolt 21 failures assumes that all other components will be functioning 22 according to their design specifications, and does not consider 23 the fact that the other components may also be undergoing Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 57
1 degradation from various interacting aging mechanisms.
2 Moreover, the applicant fails to consider the possibility that a 3 shock load (e.g., due to a LOCA) may cause the sudden failure of 4 the remaining intact clevis bolts, which, in turn, may lead to 5 an uncoolable core geometry. In short, rather than taking 6 proactive steps to replace the degraded clevis bolts prior to 7 failure, the applicant proposes to wait for clevis bolt failures 8 to occur before taking steps to address the problem, an approach 9 which is totally unacceptable in my opinion.
10 The baffle former bolts and clevis insert bolts are just 11 two examples of Entergys overarching approach to RVI aging 12 management, which foregoes preventative component repair or 13 replacement in favor of running the reactor until detectable 14 damage or component failure occurs.
15 Q. Do you have any other concerns regarding specific 16 components discussed in the Amended and Revised RVI Plan?
17 A. Yes. The applicants approach for analyzing the lower 18 support structures functionality and fracture toughness is also 19 flawed [Response to RAI-11-A, Attachment 1 to NL-13-052, at 1-4 20 (Exh. NYS000501)]. The applicant suggested that irradiation 21 embrittlement effects would only be significant in the presence 22 of pre-existing flaws or service induced defects, together with 23 a stress level capable of crack propagation. In its analysis, Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 58
1 the applicant, based on the lack of documented fractures of core 2 support columns, assumed that only a limited number of columns 3 could actually contain flaws of significant size. The 4 applicant further assumed that the columns would be subject to 5 nominal normal operating stresses [SSER2, at 3-43 (Exh.
6 NYS000507)]. When the USNRC Staff inquired about the most 7 recent visual inspections of the core support structures, the 8 applicant acknowledged that the CASS support column caps were 9 inaccessible to inspection and that VT-3 visual inspection 10 offered no meaningful information regarding the structural 11 integrity of the columns. [Id. at 3044.] Under these 12 circumstances, the applicants conclusion that irradiation-13 induced cracking of core support columns is unlikely 14 represents wishful thinking and is contrary to recent studies 15 [e.g., NUREG/CR-7184, at xv (Revised December 2014) (Exh.
16 NYS00488A-B)], which show the extreme sensitivity of crack 17 growth rate and fracture toughness to irradiation. Moreover, it 18 ignores the fact that these and other non-CASS RVI structures 19 and components undergo a range of aging degradation mechanisms 20 simultaneously under steady-state and transient conditions, and 21 that their embrittlement or susceptibility to fracture simply 22 cannot always be adequately detected using currently available 23 inspection techniques.
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 59
1 Also, not all of the core support structures are accessible 2 for inspection, so surrogate structures have been chosen by 3 Entergy to assess age-related degradation mechanisms. For 4 example, the girth weld of the core barrel has been proposed by 5 the applicant as a leading indicator for irradiation-induced 6 embrittlement (IE) and irradiation-assisted stress corrosion 7 cracking (IASCC) of the core support column caps, even though 8 these components are very different, and they may be exposed to 9 different degradation mechanisms and shock loads. In fact, as 10 pointed out recently by a member of the ACRS Plant License 11 Renewal Subcommittee, [t]he relationship between a lower core 12 barrel weld and the tops of these columns is a bit of a stretch 13 . . . [t]heyre totally different type of components, totally 14 different loadings. Moreover, to have a failure due to a 15 seismic event you dont even need to have a crack if these 16 columns are really brittle . . . . [ACRS Plant License Renewal 17 Subcommittee Transcript, at 209-211 (April 23, 2015) (Exh.
18 NYS000526)].
19 Q. Does the applicants Amended and Revised RVI Plan 20 adequately account for the potential cumulative effect of 21 synergistic aging mechanisms on RVIs?
22 A. No. By merely relying on MRP 227-A for its aging 23 management plan, the applicant has ignored the large Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 60
1 uncertainties that exist with respect to the effects of 2 irradiation-induced aging phenomena. [Chen, et al., at xv (Exh.
3 NYS000488A); NUREG/CR-7153, Vol. 2: Aging of Core Internals and 4 Piping Systems, at 181, 187, 210-211 (Exh. NYS00484A-B);
5 Stevens, et al. (October 2014), at 9-10 (Exh. NYS000491)].
6 While the applicants Thermal Aging and Neutron Irradiation of 7 Cast Austenitic and Stainless Steel (CASS) program generally 8 recognizes the potential adverse synergistic effects of elevated 9 coolant temperature and irradiation on the fracture toughness of 10 CASS materials, a broader recognition of this principle is 11 needed by the applicant, since RVI components made from non-cast 12 stainless steel will also experience the combined effects of 13 irradiation-induced embrittlement, corrosion, and other aging 14 mechanisms. The applicant has failed to evaluate the 15 synergistic mechanisms that occur for many other important and 16 vulnerable RVI components, such as the core baffles, baffle 17 bolts, and formers. Compared to the baffles, baffle bolts, and 18 formers, the core support columns (which are obviously very 19 important incore structures) are located in an area of the 20 reactor pressure vessel which is subject to less radiation 21 fluence (and thus are less susceptible to embrittlement).
22 Q. Do you have any other concerns with the applicants 23 Amended and Revised RVI Plan?
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 61
1 A. Yes. The applicant proposes to rely on visual (VT-3) 2 inspection techniques for many RVI components. However, there 3 are significant shortcomings of this technique to detect 4 material cracking, degradation, or wear prior to failure, as has 5 been noted by USNRC staff [Tregoning, at 2-3 (Exh. NYS000508);
6 Case, at 1 (Exh. NYS000509)], and illustrated by the visual 7 detection of only 7 out of 29 fractured clevis insert bolts at a 8 Westinghouse PWR in 2010 [SSER2, at 3-25 (Exh. NYS000507)]. In 9 an RAI to Entergy, the USNRC staff observed that VT-3 visual 10 examination may not be adequate for all components for detecting 11 fatigue cracking prior to the occurrence of structurally 12 significant cracking. [Attachment 1 to NL-13-052, at 5 (May 7, 13 2013) (Exh. NYS000501)]. Moreover, as I have noted previously, 14 the level of embrittlement can not be detected at all using 15 visual inspection techniques.
16 Fatigue 17 Q. Turning to fatigue. Could you explain what fatigue 18 is?
19 A. Yes. Fatigue is another important age-related 20 degradation mechanism. It is one of the primary considerations 21 when conducting a time limited aging analysis (TLAA) and an 22 aging management program (AMP) for nuclear power plants.
23 Fatigue of various structures, components and fittings in a Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 62
1 nuclear reactor can result in piping and pressure boundary 2 component and fitting ruptures, physical failures, and the 3 relocation of loose pieces of RVI metal throughout the reactor 4 system, which, in turn, may result in core blockages and 5 interfere with the effective core cooling of a nuclear power 6 plant. My main concerns about fatigue are the increased 7 potential for a primary or secondary side LOCA, and the failure 8 of various RPV internals (RVIs). It should be noted that the 9 fatigue life of a PWR component, fitting or structure is 10 normally evaluated in terms of a cumulative usage factor (CUF) 11 which is corrected for the degradation in fatigue life due to 12 the reactor coolant environment (i.e., CUFen). The cumulative 13 usage factor is defined as, CUF = N/Na-AIR, where N is the number 14 of the various fatigue cycles that have occurred (or are 15 expected by the end of plant life, EOL), and Na-AIR is the number 16 of allowable fatigue cycles obtained from data (taken in air) at 17 which failure (i.e., significant surface cracking) is expected.
18 The observed degradation in fatigue that occurs due to hot 19 reactor coolant is quantified by an environmental fatigue 20 correction factor, Fen = Na-AIR/Na-RC, where Na-AIR is the allowable 21 number of fatigue cycles measured in air, and Na-RC is the 22 allowable fatigue cycles measured in a simulated reactor coolant 23 (RC) enviroment; thus, CUFen = CUF x Fen = N/Na-RC. The criterion Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 63
1 for acceptance by the USNRC is that CUFen < 1.0 by the end of 2 life (EOL) for the component, fitting or structure in question.
3 Anyway, the allowable cycles to failure (Na-RC) are determined 4 from small scale experiments using metal test samples which are 5 exposed to simulated reactor coolant environments.
6 Unfortunately, to date, there have not been any systematic 7 fatigue experiments done in simulated reactor coolant 8 environments using highly embrittled metal test samples, which 9 have less fatigue life than ductile materials. That is, the 10 synergistic degradation effect of embrittlement has not been 11 included in CUFen (= N/Na-RC) evaluations, thus the results are 12 expected to be non-conservative since the denominator (Na-RC) will 13 be too large, and thus CUFen will be too small.
14 Q. I show you what has been marked as Exhibit NYS000527.
15 Do you recognize this document?
16 A. Yes. This is Entergys Fatigue Monitoring Plan for 17 IP2 and IP3.
18 Q. Is Entergy required to conduct fatigue evaluations of 19 internal and external components?
20 A. Yes. In this proceeding, the applicant agreed, in 21 Commitments 33, 43 and 49, to calculate the CUFen for 22 external(i.e., primary pressure boundary) and internal (RVI) 23 components in certain locations [Dacimo, Fred, Entergy, letter Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 64
1 to Document Control Desk, USNRC, Reply to Request for 2 Additional Information Regarding the License Renewal 3 Application, NL-13-122 (September 27, 2013), at 20 4 (NYS000502)]. Additionally, the USNRC has recently proposed to 5 require all applicants for license renewal to evaluate the 6 fatigue life of limiting components beyond those originally 7 specified in NUREG/CR-6260, and to evaluate the effect of 8 reactor coolant environment on the fatigue life of both 9 external and internal (i.e., RVIs) structures and systems [79 10 Fed. Reg. 69,884 (November 24, 2014) (NYS000522); USNRC, Draft 11 Regulatory Guide DG-1309 (Proposed Revision 1 of Regulatory 12 Guide 1.207, dated March 2007), Guidelines for Evaluating the 13 Effects of Light-Water Reactor Coolant Environments in Fatigue 14 Analyses of Metal Components (November 2014) (NYS000523)].
15 Q. In your expert opinion, has Entergy done adequate 16 fatigue evaluations to assure the safety of their two nuclear 17 power plants at the Indian Point site during extended 18 operations?
19 A. No.
20 21 22 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 65
1 2
3 4
5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 66
1 2
3 4
5 6
7 Unfortunately, Westinghouse did 8 not perform error analyses to quantify the modeling 9 uncertainties and the effect of code user interactions in its 10 fatigue calculations using WESTEMS. It would appear that 11 virtually any error would put some of the calculated values of 12 CUFen over the CUFen = 1.0 fatigue failure limit.
13 Q. I show you a document marked as Exhibit NYS000513. Do 14 you recognize it?
15 A. Yes, it is a paper presented by Westinghouse at a 16 recent Pressure Vessels & Piping Conference of the American 17 Society of Mechanical Engineers held in Anaheim, California in 18 July 2014; it is entitled License Renewal and Environmental 19 Fatigue Screening Application and its authors were Mark Gray 20 and Christopher Kupper.
21 Q. Are you familiar with its contents?
22 A. Yes, I have reviewed this article and it clearly shows 23 the iterative process used by Westinghouse in which safety Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 67
1 margin is removed in its environmentally assisted fatigue (EAF) 2 calculations in an effort to reduce the output or result below 3 CUFen = 1.0.
4 Q. Returning to our discussion about Westinghouses 5 fatigue evaluations of reactor coolant pressure boundary 6 components, has Entergy addressed the issue of fatigue in the 7 context of shock loads?
8 A. No, 9
10 11 12 13 14 15 16 17 18 19 20 Even 21 assuming this CUFen calculation is accurate, it does not account 22 for the possibility that a highly fatigued component, which does 23 not yet have signs of significant surface cracking, may be Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 68
1 exposed to an unexpected seismic event or shock load that could 2 cause it to fail. This is a good example of the type of silo 3 thinking (i.e., the fatigue and safety analyses are treated 4 entirely separately) that NYS is concerned about.
5 Q. I show you your Supplemental Report, which has been 6 marked as Exhibit NYS000297. I note that the State has 7 provisionally designated it as containing confidential 8 information. Would you provide a brief summary of the Report?
9 A. I prepared this Supplemental Report to set out some of 10 my concerns about the use of the WESTEMS computer code to 11 develop a cumulative fatigue analysis of certain components in 12 the Indian Point reactors and their reactor coolant pressure 13 boundaries.
14 Q. Would you briefly summarize your concerns?
15 A. Yes. First, I am concerned that without an error 16 analysis it is difficult to be in a position to meaningfully 17 analyze the results of the 2010 and subsequently refined CUFen 18 analyses presented by Entergy and Westinghouse.
19 Q. Why is an error analysis important?
20 A. It is well known that all engineering analyses are 21 based on imperfect mathematical models of reality and various 22 code user assumptions which inherently involve some level of 23 error. Error analyses help readers and decision makers Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 69
1 understand what level of confidence to attach to the calculated 2 results and the proposed conclusions.
3 Q. Is the preparation of an error analysis an accepted 4 practice in the field of engineering?
5 A. Yes. Engineers frequently prepare error analyses. In 6 my submissions in this proceeding I noted that one would 7 normally expect to see at least a hybrid propagation-of-error 8 type of analysis [Kline & McClintock (1953) (Exh. NYS000514)] to 9 determine the overall uncertainty in the CUFen results given by 10 Westinghouse. I also referenced a standard enineering text 11 book, Basic Engineering Data Collection and Analysis, pp. 310-12 311, by Vardeman & Jobe [2001], to demonstrate the various types 13 of error analyses which are regularly done by engineers [Exh.
14 NYS000347].
15 Q. I show you what has been marked as Exhibit NYS000515.
16 Are you familiar with it?
17 A. Yes, it is a recent USNRC inspection report with 18 notices of non-conformance for Westinghouses Quality Assurance 19 Program. In that report, the USNRC determined that Westinghouse 20 failed to adequately implement its QA program in the areas of 21 corrective actions, oversight of suppliers, and audits. Since 22 Entergy relies on Westinghouse services to, among other things, 23 provide appropriate guidance on corrective action and other Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 70
1 activities affecting safety-related functions, the USNRCs 2 findings of non-conformance are all the more unsettling. Under 3 these circumstances, the USNRC should insist that an error 4 analysis be performed to ensure the validity of Westinghouses 5 fatigue evaluations for IP2 and IP3 components.
6 Q. Are you aware of an instance where an error analysis 7 was prepared for a project at Indian Point?
8 A. Yes, for example, in 1980, the Consolidated Edison 9 Company of New York prepared an error analysis in support of a 10 proposal to add more spent fuel into the spent fuel pool at 11 Indian Point Unit 2.
12 Q. I show you what has been marked as Exhibit NYS000348; 13 do you recognize it?
14 A. Yes. That is a copy of the 1980 Con Edison error 15 analysis for the re-racking of spent fuel in the Unit 2 spent 16 fuel pool.
17 Q. Do you have other concerns about the refined CUFen 18 reanalysis?
19 A. Yes, as discussed in my Supplemental Report, I am 20 concerned that engineering judgment or user intervention could 21 have affected the results. I note that when USNRC Staff issued 22 the Supplemental Safety Evaluation Report, Staff instructed 23 Entergy and Westinghouse, on a going forward basis, to document Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 and disclose the use of engineering judgment and user 2 intervention when conducting future fatigue analysis using the 3 WESTEMS code. This is noted in Exhibit NYS000160 at page 4-2.
4 To my knowledge, Westinghouse has provided such information for 5 some, but not all of the fatigue evaluations performed to date.
6 Also, USNRC Staff instructed Entergy not to use WESTEMS when 7 conducting analyses under the ASME Standard know as NB-3600 [at 8 4-2, 4-3 (Exh. NYS000160)]. Furthermore, I am concerned about 9 the analytical framework employed by the WESTEMS code. As 10 detailed, in my Supplemental Report, I believe that the codes 11 thermal-hydraulic models and framework are too simplified to 12 predict accurate results.
13 14 15 16 17 18 19 Q. Do you know how Entergy is proposing to address 20 fatigue as part of its overall aging management plan for IP2 and 21 IP3?
22 A. Yes. Entergy has an existing Fatigue Monitoring Plan 23 for addressing metal fatigue. The program is designed to Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 monitor operational cycles and transients so that the various 2 CUFen remain below unity. The company has also proposed to 3 include RVIs in its Fatigue Monitoring Program; however, for the 4 reasons already stated, WESTEMS may be non-conservative, so any 5 program that relies on values, such as the fatigue cycles to 6 failure, derived from the WESTEMS methodology is inherently 7 unreliable for ensuring that aging RVIs avoid failure. This is 8 particularly true when these results include no accompanying 9 error analysis.
10 Conclusion 11 Q. Could you summarize your general concerns with the 12 applicants license renewal application?
13 A. Yes, I am very concerned that Entergy has continually 14 eroded the safety margins and conservatisms built into the 15 current licensing basis for the Indian Point reactors. For 16 example, Entergy has relied on CUFen calculations that remove 17 various conservatisms but are still very close to unity (the 18 fatigue failure limit). Entergy also relies on the detection of 19 degradation, wear or cracking prior to component failure, rather 20 than repairing or replacing the aging parts - particularly the 21 RVIs - preemptively. Moreover, Entergy implicitly assumes that 22 the plant will operate in a steady-state, and has not taken into Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 account unanticipated severe seismic events or thermal/pressure 2 shock loads which can cause failures to occur.
3 As reactors and their constituent components age, it 4 becomes very important to preserve - rather than erode -
5 operational safety margins. Uncertainties exist in all systems, 6 and calculation or modeling mistakes are always possible. For 7 example, the USNRC recently became aware that certain 8 methodologies prescribed in its NUREG-0800 Branch Technical 9 Position (BTP) 5-3 for estimating the initial fracture toughness 10 of reactor vessel materials may be non-conservative. [See, 11 e.g., Troyer, et al., An Assessment of Branch Technical 12 Position 5-3 to Determine Unirradiated RTNDT for SA-508 Cl.2 13 Forgings, Paper No. PCP2014-28897, Proceedings of the ASME 2014 14 Pressure Vessels and Piping Conference, Anaheim, California 15 (July 20-24, 2014) (Exh. NYS00516); Letter from Pedro Salas, 16 Regulatory Affairs Director, AREVA, to USNRC regarding Potential 17 Non-conservatism in NRC Branch Technical Position 5-3 (January 18 30, 2014)(Exh. NYS000517); USNRC, Slides, Assessment of BTP 5-3 19 Protocols to Estimate RTNDT(u) and USE (June 4, 2014) (Exh.
20 NYS000518); NUREG-0800, Rev. 2 (Exh. NYS000521)].
21 22 23 Anyway, Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 since unexpected errors of this type do occur, maintaining 2 safety margins helps to guard against potentially adverse 3 impacts due to precisely this type of unexpected finding of non-4 conservatism in safety evaluations. Lastly, I would like to 5 note that at a recent American Society of Mechanical Engineers 6 (ASME) Pressure Vessels & Piping Conference, USNRC staff also 7 highlighted newly-identified non-conservatisms in sections of 8 the ASME Code regarding fracture toughness applicable to nuclear 9 reactor operations. [Kirk, M. et al., Assessment of Fracture 10 Toughness Models for Ferritc Steels Used in Section XI of the 11 ASME Code Relative to Current Data-Based Model, PVP 2014-28540 12 (Exh. NYS000520)]. This is yet another reason to preserve, 13 rather than erode safety margins in the aging management of 14 light water nuclear reactors (e.g., PWRs).
15 Q. You have reviewed NUREG-1801, GALL Report Revision 1 16 (Exh. NYS00146A-C); NUREG-1800, Standard Review Plan Revision 1 17 (Exh. NYS000195); NUREG-1801, GALL Report Revision 2 (Exh.
18 NYS00147A-D); NUREG-1800, Standard Review Plan Revision 2 (Exh.
19 NYS000161); EPRIs MRP-227 Revision 0 (Exh. NYS00307A-D); EPRIs 20 MRP-227-A (Exh. NYS000507); Entergys July 2010 NL-10-063 21 communication, Entergys February 2012 NL-12-037 communication 22 (Exh. NYS000313) and subsequent communications constituting its 23 Amended and Revised RVI Plan (Exhs. NYS000496-506); USNRC Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 Staffs June 22, 2011 Safety Evaluation of MRP-227 Revision 0 2 (Exh. NYS000309); NUREG-1930, USNRC Staffs August 30, 2011 3 Supplemental Safety Evaluation (Exh. NYS000160); and NUREG-1930, 4 USNRC Staffs November 2014 Second Supplement Safety Evaluation 5 Report for the Indian Point License Renewal Application (Exh.
6 NYS000507); and Entergys NL-11-107 communication (Exh.
7 NYS000314), correct?
8 A. Yes.
9 Q. Do you have any opinion about those documents with 10 respect to the degradation of reactor pressure vessel internals 11 (RVIs)?
12 A. Yes.
13 Q. Please summarize your testimony.
14 A. As I stated in my initial November 2007 declaration in 15 support of the State of New Yorks Contentions 25 and 26, my 16 April 2008 declaration in support of Contention NYS-26A, my 17 September 2010 declarations in support of the States 18 supplemental filings on Contentions NYS-25 and NYS-26B/RK-TC-1B, 19 and my previously filed testimony on Contentions NYS-25 and NYS-20 26B/RK-TC-1B, and my February 2015 declaration in support of the 21 States further supplemental filings on Contention NYS-25, in my 22 professional judgment Entergy has failed to demonstrate that it 23 has adequately accounted for the aging phenomena of Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 embrittlement and fatigue for structures, components and 2 fittings inside the reactor pressure vessels (i.e., RVIs) at 3 Indian Point Unit 2 and Indian Point Unit 3. My professional 4 judgment has not fundamentally changed based upon Entergys July 5 14, 2010 submission of License Renewal Application, Amendment 6 No. 9 [NL-10-063 (Exh. NYS000313)], Entergys September 28, 2011 7 submission of NL-11-107 [Exh. NYS000314], or Entergys Amended 8 and Revised RVI Plan, consisting of the February 17, 2012 9 submission of NL-12-037 [Exh. NYS000496] as amended by its 10 subsequent communications [Exhs. NYS000497-506] and approved by 11 the USNRC Staff in the SSER2 [Exh. NYS000507]. I do not believe 12 that Entergys July 15, 2010 communication to the Board [NL 13 063 (Exh. NYS000313)] concerning a new AMP for RPV internals, or 14 its September 28, 2011 communication [NL-11-107 (Exh.
15 NYS000314)], are adequate to address the safety concerns and 16 technical issues that I have raised herein. They do not address 17 my age-related safety concerns, nor do they recognize the 18 importance of the various synergistic degradation mechanisms 19 that I am concerned with. The Amended and Revised RVI Plan, 20 which the USNRC Staff evaluated and approved in the November 21 2014 SSER2, also does not resolve my concerns over the 22 simultaneous and synergistic age-related degradation mechanisms 23 that may affect various RVI components and structures.
Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 While some age-related safety issues might eventually be 2 resolved analytically or experimentally, in many cases it 3 appears that the easiest and most cost-effective way to resolve 4 them is to simply repair or replace the most seriously degraded 5 structures, components and fittings, and this approach is what 6 NYS has been proposing for some time (particularly for the 7 degraded RVIs).
8 Q. Does this conclude your testimony?
9 A. Not quite. I want to stress that during the course of 10 my involvement in these relicensing proceedings I have 11 discovered what I believe to be some important new age-related 12 safety concerns which, to the best of my knowledge, have not 13 been previously considered in relicensing proceedings. These 14 concerns include: the synergistic effect on the degradation and 15 integrity of RPV internals (RVIs) of radiation-induced 16 embrittlement, corrosion and fatigue, and the potential for the 17 unanticipated failure of RPV internals (RVIs) due to a severe 18 seismic event or accident-induced thermal and/or pressure shock 19 loads, and the implications of the failure of RPV internal 20 structures, components and fittings (i.e., RVIs) on post-21 accident core coolability. While in the past many of these 22 issues and concerns have been noted separately, the implications 23 of their synergistic interaction has apparently been overlooked Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 and not evaluated (i.e., they have been evaluated in silos).
2 Since I first raised these technical issues in 2007, the USNRC, 3 DOE and various nuclear industry groups have slowly begun to 4 recognize their significance. In fact, the evaluation and study 5 of these important issues is underway, but major uncertainties 6 still exist. As a consequence, I believe that these important 7 age-related safety concerns must be resolved in order to have 8 assurance that the Indian Point reactors can operated safely 9 beyond their design life of 40 years. Indeed, I believe that 10 the most vulnerable RPV internals (RVIs) need to be carefully 11 identified and repaired or replaced prior to extended operations 12 since it is beyond the current state-of-the-art to perform 13 realistic and accurate calculations on the relocation of failed 14 RPV internals (RVIs) and the resultant potential for core 15 blockages and degraded core cooling.
16 Q. Does this complete your testimony.
17 A. Yes, it does. I do, however, reserve the right to 18 supplement my testimony if new information is disclosed or 19 introduced.
20 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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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 9, 2015 10 -----------------------------------x 11 DECLARATION OF RICHARD T. LAHEY, JR.
12 I, Richard T. Lahey, Jr., do hereby declare under penalty 13 of perjury that my statements in the foregoing testimony and my 14 statement of professional qualifications are true and correct to 15 the best of my knowledge and belief.
16 Executed in Accord with 10 C.F.R. § 2.304(d) 17 18 ________________________
19 Dr. Richard T. Lahey, Jr.
20 The Edward E. Hood Professor Emeritus of Engineering 21 Rensselaer Polytechnic Institute, Troy, NY 12180 22 (518) 495-3884, laheyr@rpi.edu Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
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