ML15161A301
| ML15161A301 | |
| 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 27921, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR | |
| Download: ML15161A301 (80) | |
Text
UNITED STATES 1 NUCLEAR REGULATORY COMMISSION 2 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 3 -----------------------------------x 4 In re: Docket Nos. 50
-247-LR; 50-286-LR 5 License Renewal Application Submitted by ASLBP No. 07
-858-03-LR-BD01 6 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 7 Entergy Nuclear Indian Point 3, LLC, and 8 Entergy Nuclear Operations, Inc.
June 9 , 2015 9 -----------------------------------x 10 REVISED PRE-FILED WRITTEN TESTIMONY OF 11 Dr. RICHARD T. LAHEY, JR.
12 REGARDING CONTENTION NYS-25 13 On behalf of the State of New York ("NYS" or "the State"), 14 the Office of the Attorney General hereby submits the following 15 testimony by RICHARD T. LAHEY, JR., PhD. regarding Contention 16 NYS-25. 17 Q. Please state your full name.
18 A. Richard T. Lahey, Jr.
19 Q. By whom are you employed and what is your position?
20 A. I am retired and am currently the Edward E. Hood 21 Professor Emeritus of Engineering at Rensselaer Polytechnic 22 Institute (RPI), which is located in Troy, New York.
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 1
Q. Please summarize your educational and professional 1 qualifications.
2 A. I have earned the following academic degrees: a B.S.
3 in Marine Engineering from the United States Merchant Marine 4 Academy, a M.S. in Mechanical Engineering from Rensselaer 5 Polytechnic Institute, a M.E. in Engineering Mechanics from 6 Columbia University, and a Ph.D. in Mechanical Engineering from 7 Stanford University. I have held various technical and 8 administrative positions in the nuclear industry, and I have 9 served as both the Dean of Engineering and the Chairman of the 10 Department of Nuclear Engineering & Science at RPI. Previously, 11 I was responsible for nuclear reactor safety R&D (research &
12 development) for the General Electric Company (GE), and I have 13 extensive experience with both military (i.e., naval) and 14 commercial pressurized water and boiling water nuclear reactors 15 (PWR and BWR). Also, I am a member of a number of professional 16 societies and have served on numerous expert panels. I was also 17 an Editor of the international Journal of Nuclear Engineering &
18 Design, which focuses on nuclear engineering and nuclear reactor 19 safety technology. I am widely considered to be an expert in 20 matters relating to the design, operations, safety, and aging of 21 nuclear power plants.
22 Q. Which professional societies are you a member of?
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 2
A. I am a member of a number of professional societies, 1 including: the American Nuclear Society (ANS), where I was a 2 member of the Board of Directors and the ANS's Executive 3 Committee, and was the founding Chair of the ANS's Thermal-4 Hydraulics Division; the American Society of Mechanical 5 Engineers (ASME), where I was Chair of the Nucleonics Heat 6 Transfer Committee, K-13; the American Institute of Chemical 7 Engineering (AIChE), where I was the Chair of the Energy 8 Transport Field Committee; and the American Society of 9 Engineering Educators (ASEE), where I was Chair of the Nuclear 10 Engineering Division.
11 Q. What expert panels have you served on?
12 A. I have served on numerous panels and committees for 13 the: United States Nuclear Regulatory Commission (USNRC), Idaho 14 National Engineering Laboratory (INEL), Oak Ridge National 15 Laboratory (ORNL), National Aeronautics and Space Administration 16 (NASA), National Research Council(NRC) and the Electric Power 17 Research Institute (EPRI). I am a member of the National 18 Academy of Engineering (NAE), have been elected Fellow of both 19 the ANS and the ASME, and have been a Fulbright-Hays, Alexander 20 von Humboldt and Japanese Society for the Promotion of Science 21 (JSPS) Scholar.
22 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 3
A. Have you published any papers in the field of nuclear 1 engineering and nuclear reactor safety technology?
2 Q. Yes. Over the last 50 years, I have published 3 numerous books, monographs, chapters, articles, reports, and 4 journal papers on nuclear engineering and nuclear reactor safety 5 technology. Those articles are listed in my Curricula Vitae.
6 Q. Have you received any professional awards?
7 A. Yes, I have received many honors and awards for my 8 career accomplishments in the area of nuclear reactor thermal-9 hydraulics and safety technology, including: the E.O. Lawrence 10 Memorial Award of the Department of Energy (DOE), the Glenn 11 Seaborg Medal of the ANS and the Donald Q. Kern Award of the 12 AIChE. 13 Q. I show you what has been marked as Exhibit NYS000295.
14 Do you recognize that document?
15 A. Yes. It is a copy of my Curricula Vitae, which 16 summarizes, among other things, my experience, publications, and 17 honors & awards.
18 Q. I show you what has been marked as Exhibit NYS000299 19 to Exhibit NYS000303, and Exhibit NYS000483. Do you recognize 20 those documents?
21 A. Yes. They are copies of the seven declarations that I 22 previously prepared to date for the State of New York in this 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 4
proceeding. They include my initial declaration that was 1 submitted in November 2007 in support of the State's petition to 2 intervene and its initial contentions, the April 7, 2008 3 declaration in support of Contention NYS-26A, the September 15, 4 2010 declaration submitted in support of the State's 5 supplemental bases for Contention 25, the September 9, 2010 6 declaration submitted in support of the amended Contention 7 NYS26B/RK-TC-1B, the September 30, 2011 and November 1, 2011 8 declarations submitted in support of Joint Contention NYS-38/RK-9 TC-5, and the February 12, 2015 declaration submitted in support 10 of additional bases for Contention NYS-25 and Joint Contention 11 NYS-38/RK-TC-5.
12 Q. I show you what has been marked as Exhibit NYS000296.
13 Do you recognize that document?
14 A. Yes. It is a copy of the Report that I prepared for 15 the State of New York in this proceeding. This Report documents 16 my analysis and opinions.
17 Q. I show you what has been marked as Exhibit NYS000297.
18 Do you recognize that document?
19 A. Yes. This is a copy of a Supplemental Report that I 20 prepared for the State of New York in this proceeding that 21 addresses aspects of the revised fatigue analysis that Entergy 22 and Westinghouse prepared for certain components in the Indian 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 5
Point reactors. The supplemental report also documents my 1 assessment and opinions of this work.
2 Q. I show you what has been marked as Exhibit NYS000294, 3 NYS000344, and NYS000440. Do you recognize those documents?
4 A. Yes, those documents contain my previous pre-filed 5 testimony filed in December 2011 and June 2012 in support of 6 Contentions NYS-25 and NYS-26B.
7 Q. What is the purpose of your testimony?
8 A. I have been retained by the State of New York State to 9 review Entergy's application to the U.S. Nuclear Regulatory 10 Commission (USNRC) and its Staff for two renewed operating 11 licenses for the nuclear power plants known as Indian Point Unit 12 2 and Unit 3. I have reviewed the License Renewal Applications 13 (LRAs) and subsequent filings by Entergy and the USNRC Staff.
14 My declarations and report discuss my concerns and opinions 15 about issuing twenty-year extended operating licenses for these 16 facilities. My testimony seeks to identify and discuss some 17 age-related safety concerns which have not yet been addressed by 18 Entergy. In my opinion these concerns must be resolved to 19 assure the health and safety of the American public, 20 particularly those in the vicinity of the Indian Point reactors.
21 Q. Have you reviewed various materials in preparation for 22 your testimony?
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 6
A. Yes. 1 Q. What is the source of those materials?
2 A. I have reviewed documents prepared by government 3 agencies, Entergy, Westinghouse, the utility industry, or its 4 associations (e.g., EPRI), and various related text books and 5 peer-reviewed articles.
6 Q. I show you Exhibits NYS00146A-C, NYS00147A-D, 7 NYS000160, NYS000161, NYS000195, NYS000304 through NYS000369, 8 and NYS000484 through NYS000525. Do you recognize these 9 documents?
10 A. Yes. These are true and accurate copies of some of 11 the documents that I referred to, used, or relied upon in 12 preparing my report, declarations, previous testimony, and this 13 testimony. In some cases, where the document was extremely long 14 and only a small portion is relevant to my testimony, an excerpt 15 of the document is provided. If it is only an excerpt, that is 16 noted on the first page of the Exhibit.
17 Q. I direct your attention to latter part of your Report 18 (Exh. NYS000296) entitled "Reference Documents," which contains 19 a list of documents. Would you describe that list?
20 A. Yes that section of the Report lists various salient 21 documents that I referred to, used or relied on, in preparing my 22 Report and the Supplemental Report.
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 7
Q. I direct your attention to the latter part of your 1 February 12, 2015 Declaration (Exh. NYS000483) entitled 2 "Reference Documents," which contains a list of documents.
3 Would you describe that list?
4 A. Yes, that section of the Declaration lists various 5 additional salient documents that I referred to, used or relied 6 on, in preparing my February 12, 2015 Declaration.
7 Q. How do these documents relate to the work that you do 8 as an expert in forming opinions such as those contained in this 9 testimony?
10 A. These documents represent the type of information that 11 persons within my field of expertise reasonably rely upon in 12 forming opinions of the type offered in this testimony.
13 The Indian Point Reactors 14 Q. Are you familiar with the power reactors that are the 15 subject of this proceeding?
16 A. Yes. 17 Q. Would you briefly describe them?
18 A. Entergy operates two nuclear power reactors that are 19 located in northern Westchester County near the Village of 20 Buchanan. The operating nuclear reactors are known as the 21 Indian Point Unit 2 and Indian Point Unit 3 reactors. These 22 Westinghouse-designed plants are 4-loop pressurized water 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 8
reactors (PWRs), and they are currently rated at power levels of 1 3,216.4 MW
- t. Entergy also owns another reactor at the same site.
2 That reactor is known as the Indian Point Unit 1 reactor; 3 however, that reactor has been shut down and no longer produces 4 power. 5 Operation of a Pressurized Water Reactor 6 Q. Would you briefly describe the design and operation of 7 a pressurized water reactor?
8 A. Pressurized water nuclear reactors have water (i.e., 9 the primary coolant) under high pressure flowing through the 10 core in which heat is generated by the fission process. The 11 core is located inside a reactor pressure vessel (RPV). This 12 heat is absorbed by the coolant and then transferred from the 13 coolant in the primary system to lower pressure water in the 14 secondary system via a large heat exchanger (i.e., a steam 15 generator) which, in turn, produces steam on the secondary side.
16 These steam generator systems, which are part of the plant's 17 Nuclear Steam Supply System (NSSS), are located inside a large 18 containment structure. After leaving the containment building, 19 via main steam piping, the steam drives a turbine, which turns a 20 generator to produce electrical power.
21 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 9
The reactor pressure vessel is a large steel container that 1 holds the core (i.e., the nuclear fuel); it also serves as a key 2 part of the primary coolant's pressure boundary.
3 As the name Pressurized Water nuclear Reactor (PWR) 4 suggests, this reactor design uses a pressurizer on the primary 5 side that performs several functions. In particular, it 6 maintains the operating pressure on the primary side of the 7 nuclear reactor and accommodates variations in reactor coolant 8 volume for load changes during reactor operations, and during 9 reactor heat-up and cool-down. The reactor coolant also 10 moderates the neutrons produced in the core since a pressurized 11 water nuclear reactor will not function unless the neutrons are 12 moderated (i.e., slowed down due to collisions with the hydrogen 13 molecules in the primary coolant).
14 Q. I show you what has been marked as Exhibit NYS000304.
15 Do you recognize it?
16 A. Yes. It is a schematic diagram from a USNRC document 17 that identifies the relative location of various components in a 18 pressurized water nuclear reactor type of power plant including, 19 from the inside to the outside, the reactor core, reactor 20 pressure vessel, pressurizer, steam generator, containment 21 structure, turbine, and associated piping. The diagram also 22 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 10 identifies the various materials that are used or contained in 1 those components.
2 Reactor Pressure Vessel Internals 3 Q. I show you what has been marked as Exhibit NYS000306.
4 Do you recognize it?
5 A. Yes. It is a series of schematic diagrams or figures, 6 including Figure 3-5, from an Electric Power Research Institute 7 (EPRI) document known as MRP-227 that identifies various 8 components within pressurized water nuclear reactor designed by 9 the Westinghouse Company. The title of Figure 3-5 is, "Overview 10 of typical Westinghouse internals."
11 Q. Please describe what is encompassed by the term 12 "reactor pressure vessel (RPV) internals"?
13 A. The term "reactor pressure vessel internals" (i.e., 14 RVIs) includes various structures, components, and fittings 15 inside the reactor pressure vessel including the: core barrel 16 (and its welds), core baffle, intermediate shells, former 17 plates, lower core plate and support structures, clevis bolts, 18 fuel alignment pins, thermal shield, the lower support column 19 and mixer, upper mixing vanes, and the upper/lower core 20 assemblies and support column, and the control rods and their 21 associated guide tubes, plates, and welds. Reactor pressure 22 vessel internals (RVIs) also include the bolts that hold various 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 11 components together or to other components including: the 1 baffle-to-baffle bolts, the core barrel-to-former bolts, and 2 baffle-to-former bolts as well as the welds or weldments that 3 hold sections of these components together.
4 Q. Was the aging management of RVIs initially considered 5 as part of the LRA for the Indian Point facilities?
6 A. No, it was not. Fortunately, during the course of 7 these ASLB hearings on Indian Point the USNRC has now recognized 8 and highlighted the importance of RVIs [see, e.g., USNRC Report, 9 "Final Interim Guidance LR-ISG-2011-04 Updated Aging Management 10 Criteria for Reactor Vessel Internal Components for Pressurized 11 Water Reactors," NRC-ISG-2011-04 (May 28, 2013) (NYS000524)].
12 Q. Are there any reactor components that you believe 13 should be considered as reactor vessel internals, but that 14 Entergy has claimed are not reactor vessel internals?
15 A. Yes. Entergy has argued that the control rods are not 16 reactor vessel internals. However, the control rods and their 17 associated guide tubes, plates, pins and welds are located in 18 the core region of the RPV, and the control rods are inserted 19 into the RPV through the upper head through so-called stub 20 tubes. The function of the control rods is to absorb excess 21 fission neutrons (i.e., those not needed to achieve a chain 22 reaction) so that the power level of a reactor can be 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 12 controlled. Accordingly, the control rods and associated 1 components are very important RPV internals and their integrity 2 is an extremely important safety concern. While the control 3 rods are moving parts and can be replaced as required, many of 4 the other associated components are not moving parts and are not 5 normally replaced. In any event, if a shock load occurs (e.g., 6 during a LOCA or severe earthquake) any of these seriously 7 embrittled structures may fail and lead to degraded core 8 cooling. Thus, in my opinion, omitting the control rod 9 assemblies and associated fittings from an RPV internals (RVIs) 10 aging management program is a serious and indefensible omission.
11 Q. Coming back to Exhibit NYS000306, would you describe 12 the other diagrams?
13 A. Yes. They are a collection of additional schematic 14 figures from the Electric Power Research Institute's Report MRP-15 227 that provide additional detail concerning various reactor 16 pressure vessel internals and their location within the reactor 17 pressure vessel. The reactor pressure vessel internals shown 18 include the control rod guide tube assembly, the control rod 19 guide cards, guide tube support pins, the control rods, baffles, 20 formers, baffle-former assemblies, baffle-to-former bolts, 21 corner edge bracket baffle to former bolts, core barrel to 22 former bolts, baffle plate edge bolts, core support structures, 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 13 and various weldments, including welds within the reactor 1 pressure vessel for the core barrel plates.
2 Overview 3 Q. In your expert opinion what is the most important age-4 related safety issue associated with the relicensing of the two 5 Indian Point reactors?
6 A. My over-arching concern relates to Entergy's "silo" 7 type approach to evaluating the impact of various aging 8 mechanisms such as embrittlement and fatigue, and the company's 9 failure to consider, as part of plant safety analyses, the 10 potential consequence of unanticipated shock loads (e.g., those 11 due to design basis accidents) on severely fatigued and 12 embrittled components. Entergy implicitly assumes that there is 13 no interplay between the various material aging degradation 14 phenomena and that degraded components will have no impact on 15 the plants' ability to safely operate, particularly during 16 unanticipated shock loads. For example, Entergy's fatigue 17 evaluations, performed by Westinghouse using the WESTEMS 18 computer code, used the metric CUF en to appraise environmentally 19 assisted fatigue in various reactor components. However, these 20 evaluations were quasi-static low and high cycle fatigue 21 evaluations that considered neither the effect of neutron-22 induced embrittlement nor the combined effects of fatigue damage 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 14 and other degradation mechanisms such as radiation enhanced 1 corrosion-induced cracking. In both the fatigue analyses and 2 the plant safety analyses, it was implicitly assumed that 3 fatigue weakened and embrittled structures, components and 4 fittings would respond to shock loads in the same way as if they 5 were ductile, which is simply not true. Also, no error analyses 6 were presented to quantify the WESTEMS predictions for the 7 various internals, piping systems and fittings even though some 8 of them were extremely close to the CUF en = 1.0 failure limit. In 9 any event, under these circumstances, various operational and 10 accident-induced shock loads could cause failures well before 11 the fatigue limit is reached (i.e., when CUF en < 1.0), and 12 therefore reliance on inspection-based fatigue monitoring does 13 not provide adequate assurance that the degraded components will 14 not fail.
15 Once again, the most serious short-coming of this "siloing" 16 approach is that synergistic interactions between radiation-17 induced embrittlement, corrosion-induced cracking, and fatigue-18 induced degradation mechanisms have not been considered. For 19 example, neither Entergy's license renewal application nor its 20 proposed aging management plan consider the potential for, or 21 the consequences of, fatigue-induced failure of seriously 22 embrittled reactor pressure vessel internals (RVIs). Also, when 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 15 the plant's safety analyses were done by Entergy it was 1 implicitly assumed that the in-core geometry would remain intact 2 during postulated accidents. Unfortunately, unlike ductile 3 metals, seriously embrittled and fatigued RPV internals may not 4 be able to survive the shock loads associated with significant 5 seismic events or the pressure and/or thermal shock loads 6 induced by various accidents and severe operational transients.
7 If not, they can fail and relocate, possibly causing core 8 blockages that degrade core cooling and may lead to core melting 9 and massive radiation releases.
10 Entergy has an obligation to show that its plants can be 11 safely operated beyond their 40 year design lives. I believe 12 that this will require much more study and analysis than has 13 been presented to date to identify any limiting RPV internals 14 that require repair or replacement. Nevertheless, this must be 15 done to verify that the two Indian Point reactors can be safely 16 operated for another 20 years beyond the design life of these 17 plants. 18 Q. What do you mean by synergistic interactions between 19 aging-related degradation mechanisms?
20 A. I mean that the concurrent exposure of reactor 21 components - especially RVI components - to multiple aging 22 mechanisms that occur in a reactor core (including fatigue, 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 16 irradiation embrittlement, and corrosion) may result in 1 cumulative material degradation that exceeds the predicted 2 combined degradation for each aging mechanism acting alone.
3 Q. Are there any studies or reports that support your 4 concern regarding synergistic aging effects?
5 A. Yes. However, the rather complex and interacting 6 metal degradation mechanisms associated with fatigue, 7 irradiation and corrosion interact is still an area of active 8 research (e.g., how fatigue-induced cracks propagate in an 9 embrittled, as opposed to ductile, metal structure). In fact, 10 the Department of Energy (DOE) and USNRC, in conjunction with 11 various national laboratories, have recently embarked on an 12 ambitious R&D program to understand and resolve issues related 13 to these interacting and synergistic effects [NUREG/CR-7153, 14 Vol. 2, "Expanded Materials Degradation Assessment (EMDA), Aging 15 of Core Internals and Piping Systems" (October 2014), at 1-5 16 (Exh. NYS00484A-B)]. In addition, the federal government has 17 also embarked on a fairly large research program, known as the 18 Light Water Reactor Sustainability Program, which includes 19 research into whether the different materials and LWR components 20 can continue to perform their intended function during the 21 extended operation of a nuclear reactor.
[DOE, Light Water 22 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 17 Sustainability Program, Material Aging and Degradation Technical 1 Program Plan (August 2014) (Exh. NYS000485)].
2 Nevertheless, it is well known that, "the effects of 3 embrittlement, especially loss of fracture toughness, make 4 existing cracks in the affected materials and components less 5 resistant to growth" [USNRC Letter, Grimes to Newton, at 16 6 (Feb. 10, 2001) (Exh. NYS000324); see Stevens, Gary L., 7 Presentation to the ACRS on "Technical Brief on Regulatory 8 Guidance for Evaluating the Effects of Light Water Reactor 9 Coolant Environments in Fatigue Analyses of Metal Components" 10 (December 2, 2014), at 56-58 (Exh. NYS000486); Chopra, O.K., 11 "Degradation of LWR Core Internal Materials due to Neutron 12 irradiation," NUREG/CR-7027 (Dec. 2010) (Exh. NYS000487)], and, 13 "irradiation embrittlement decreases the resistance to crack 14 propagation" [Westinghouse Owners Group WCAP-14577 Rev. 1-A 15 Report, at 3-2 (March 2001) (Exh. NYS00307A-D)]. Moreover, a 16 recent report, prepared by Argonne National Laboratory for the 17 USNRC, acknowledges, with respect to cast austenitic stainless 18 steels (CASS), that "a combined effect of thermal aging and 19 irradiation embrittlement could reduce the fracture resistance 20 even further to a level neither of these degradation mechanisms 21 can impart alone" [Chen, et al., "Crack Growth Rate and Fracture 22 Toughness Tests on Irradiated Cast Stainless Steels," NUREG/CR-23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 18 7184 (Revised December 2014), at xv (Exh. NYS00488A-B)].
1 Indeed, nuclear industry groups have now recognized the 2 potential for synergistic aging effects in CASS RVI components 3 [EPRI, Slides, Industry-NRC Meeting on CASS Screening Criteria 4 for Thermal and Irradiation Embrittlement for BWR and PWR 5 Internals" (July 15, 2014) (Exh. NYS000489)].
6 Q. Are synergistic aging effects limited to CASS 7 components?
8 A. No. All components within the RPV are subject to 9 multiple aging degradation mechanisms. Different materials may 10 undergo aging in different ways, but all materials are 11 susceptible to synergistic effects.
12 Q. Are these synergistic aging effects fully understood?
13 A. Not at all. Multiple recent reports and studies from 14 USNRC, DOE, and associated contractors recognize the lack of 15 understanding of the interrelationship between embrittlement, 16 high or low cycle fatigue, and shock loads for highly fatigued 17 and/or embrittled components made of CASS, non-cast stainless 18 steels, or other alloys. In addition, the consequences of the 19 interaction of embrittlement, fatigue, and the corrosion-induced 20 degradation of various reactor pressure vessel internals (RVI), 21 and safety-related components/systems during shock loads, 22 remains unknown [see, e.g., NUREG/CR-6909 Rev. 1 (March 2014 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 19 (draft) (Exh. NYS000490)), at 11 ("it is not possible to 1 quantify the impact of irradiation on the prediction of fatigue 2 lives in PWR primary water environments compared to those in 3 air."); NUREG/CR-7153, Vol. 2, "Expanded Materials Degradation 4 Assessment (EMDA), Aging of Core Internals and Piping Systems" 5 (October 2014), at 3 (Exh. NYS00484A-B)]. The Argonne National 6 Laboratory report described above states that, "no data are 7 available at present with regard to the combined effect of 8 thermal aging and irradiation embrittlement" on CASS [Chen, et 9 al., NUREG/CR-7184, at xv (Exh. NYS00488A-B); see also Chopra, 10 O.K., "Degradation of LWR Core Internal Materials due to Neutron 11 irradiation," NUREG/CR-7027 (Dec. 2010) (Exh. NYS000487)]. As 12 noted before, the same is also true for the interaction of 13 irradiation-induced embrittlement, corrosion, and fatigue of 14 non-cast stainless steel RVIs.
15 A recent paper presented at an MPA Seminar in Stuttgart, 16 Germany confirms that, at present, the USNRC staff does not have 17 a clear solution to the challenges posed by synergistic age-18 related degradation mechanisms [Stevens, Gary L., et al., 19 "Observations and Recommendations for Further Research Regarding 20 Environmentally Assisted Fatigue Evaluation Methods," 40th MPA-21 Seminar, Materials Testing Institute, University of Stuttgart, 22 Stuttgart, Germany (October 6-7, 2014) (Exh. NYS000491)]. A 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 20 recent draft report on the "Effect of LWR Coolant Environments 1 on the Fatigue Life of Reactor Materials," prepared by Argonne 2 National Laboratory (ANL) and USNRC Staff, recognizes the 3 "inconclusive" nature of existing data on the synergistic 4 effects of irradiation and fatigue, and other aging mechanisms 5 in LWR environments, and concludes that, "additional fatigue 6 data on reactor structural materials irradiated under LWR 7 operating conditions are needed." [NUREG/CR-6909, Rev. 1 (March 8 2014 [draft]), at 11 (Exh. NYS000490)]. Furthermore, during a 9 "Briefing on Subsequent License Renewal" to the USNRC, the 10 USNRC's Chief of the Corrosion and Metallurgy Branch, Dr. Mirela 11 Gravila, testified that the Piping and Core Internals Panel had 12 recognized "significant gaps" in our technical knowledge with 13 respect to the effects of irradiation-induced degradation of the 14 RVI components [Trans. of Briefing on Subsequent License 15 Renewal, at 77 (May 2014) (Exh. NYS000492)].
16 Q. With respect to the aging management of nuclear 17 facilities, how has the USNRC responded to these embrittlement 18 concerns with respect to its synergistic effects on fatigue?
19 A. Notwithstanding the significant concerns and 20 considerable uncertainty regarding synergistic aging effects, 21 the USNRC has so far declined to require that plant operators 22 repair or replace degraded systems, structures, and fittings, 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 21 opting instead to manage aging through periodic inspections, and 1 the use of an empirical environmental factor method (F en) for 2 fatigue life when evaluating the in situ degradation of 3 structures and components [Stevens, et al., (October 2014), at 4 10 (Exh. NYS000491)], a method which is not necessarily 5 conservative and one that certainly does not address all the 6 synergistic effects (e.g., embrittlement) that New York State is 7 concerned about.
8 Q. Would you please explain in more detail the various 9 degradation mechanisms that you are concerned with?
10 A. Yes, let me begin with embrittlement.
11 Embrittlement 12 Q. Would you explain what embrittlement is?
13 A: Embrittlement refers to the change in the mechanical 14 properties (and structure) of materials, such as metals, that 15 can occur over time under the bombardment of neutrons. The 16 degree of exposure to neutrons is normally expressed in terms of 17 a "fluence" (i.e., the neutron flux times the duration of the 18 irradiation process). The extended exposure to neutrons causes 19 damage to metals and makes them more brittle so that they become 20 more susceptible to failures due to cracking or fracture. In 21 particular, this radiation-induced damage results in a decrease 22 in fracture toughness and ductility.
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 22 Embrittlement is an age-related degradation mechanism 1 whereby a component experiences a decrease in ductility, a loss 2 of facture toughness, and an increase in yield strength. While 3 the initial aging effect is loss of ductility and toughness, 4 unstable crack propagation is the eventual adverse aging effect 5 if a crack is present and the local applied stress intensity is 6 sufficient. Moreover, when subjected to a sufficient load, a 7 component which has been highly embrittled by neutron 8 irradiation may experience sudden, brittle fracture well before 9 a surface crack is detected. This is a particular problem for 10 the large pressure and/or thermal shock loads associated with 11 postulated accidents. For this reason, USNRC regulations set 12 forth at 10 C.F.R. § 50.61 impose fracture toughness 13 requirements and/or operating parameters to prevent brittle 14 facture of reactor pressure vessels. Indeed, NUREG-1800, Rev. 2 15 (Table 4.1-3) (Exh. NYS000161) identifies reduced fracture 16 toughness of reactor vessel internals as a candidate for a time 17 limited aging analysis. Because loss of ductility due to 18 radiation embrittlement was not considered in the design of the 19 stainless steel reactor vessel's internal components(RVIs), it 20 is all the more important to evaluate the degree of 21 embrittlement of RVIs during license renewal review. [Chopra, 22 O., Public Comment on NRC-2010-0180-0001, Availability of Draft 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 23 NUREG-1800, Revision 2 and Draft NUREG-1801, Revision 2 (June 9, 1 2010) (Exh. NYS000493)].
2 The Consequences of Embrittlement 3 Q. Is embrittlement a concern for pressurized water 4 nuclear reactors?
5 A. Yes. For a pressurized water nuclear reactor to 6 operate safely, the metals involved need to be sufficiently 7 ductile, which means that they must be able to deform without 8 experiencing failures. When metals, such as steel, experience a 9 significant neutron fluence, which happens to the materials in 10 close proximity to the reactor core (e.g., the steel reactor 11 pressure vessel's interior wall and the associated RVIs), the 12 temperature required for them to maintain sufficient ductility 13 is increased as the metal is continually bombarded by a neutron 14 flux. The temperature at which there is a marked change from 15 ductile to non-ductile behavior is often called the "nil 16 ductility temperature" (NDT). However, even for temperatures 17 well above the NDT, the irradiated metals continue to be damaged 18 and further embrittled due to the neutron bombardment. Indeed, 19 the neutron damage will not be annealed out (i.e., be 20 neutralized) unless the damaged metals are taken to temperatures 21 that are well above PWR operating temperatures.
22 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 24 Q. Could embrittlement impact a nuclear reactor's ability 1 to respond to a transient, shock load, or an accident scenario?
2 A. Yes. Reduced ductility (or embrittlement) will 3 adversely affect a PWR's ability to withstand severe seismic 4 events and pressure and/or thermal shock loads, and thus there 5 is a threat to the integrity of highly embrittled internal 6 structures in the reactor pressure vessel. For example, during 7 a recent meeting regarding Indian Point, a member of the 8 Advisory Committee on Reactor Safeguards Plant License Renewal 9 Subcommittee expressed concern that embrittled RVI components 10 could fail during a seismic event. [Trans. of Advisory Committee 11 on Reactor Safeguards, Plan License Renewal Subcommittee, at 12 209-210 (April 23, 2015) (Exh. NYS000526)].
13 Various accidents and abnormal transients can expose a 14 reactor pressure vessel and its internal structures, components 15 and fittings (i.e., RVIs) to significant pressure and/or thermal 16 shock loads. If the reactor pressure vessel's internal 17 structures (RVIs) are sufficiently degraded due to corrosion-18 induced cracking, fatigue and/or radiation-induced 19 embrittlement, these shock loads can have significant 20 consequences. Indeed, the resultant stresses from such 21 accidents may cause the RVIs to fail structurally and relocate 22 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 25 within the RPV. If so, the ability to effectively cool the 1 decay heat in the core may be lost due to core blockage.
2 One well known safety concern associated with embrittlement 3 is the ability of metals to withstand a thermal shock event. A 4 thermal shock can occur in various ways, for example: (1) during 5 loss of coolant accidents (e.g., postulated primary or secondary 6 side LOCAs), or, (2) during a reactor SCRAM (i.e., a rapid 7 insertion of the control rods which terminates the nuclear chain 8 reaction). A particularly bad LOCA event is one in which there 9 is a rapid depressurization of the secondary side (e.g., a steam 10 line break) which causes a reactor SCRAM and thus a rapid 11 cooling of the primary coolant via the steam generators. This 12 type of accident can lead to severe thermal shock of the reactor 13 pressure vessel and the associated RPV internals (RVIs).
14 Severe thermal shocks can also occur during a design basis 15 accident (DBA) LOCA event (i.e., a complete breach of main 16 coolant piping on the primary side), which rapidly depressurizes 17 the primary side and leads to the injection of relatively cool 18 emergency core coolant into the reactor pressure vessel (e.g., 19 from the accumulators). As noted previously, this may lead to 20 the sudden fracture and relocation of highly embrittled RVI 21 structures, components and fittings, and thus impede their 22 ability to perform their intended functions, and adversely 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 26 impact their core-cooling functions. In the past, most of the 1 USNRC's attention has been focused on the integrity of the 2 reactor pressure vessel. However, the RVIs are much less 3 massive and are much closer to the core, and thus they suffer a 4 lot more radiation damage and embrittlement. Notably, the 5 USNRC's fluence threshold for irradiation embrittlement of the 6 reactor pressure vessel beltline is 1 x 10 17 n/cm 2 [10 C.F.R. Part 7 50, Appendix G; USNRC Regulatory Issue Summary 2014-11 (Exh.
8 NYS000494)]. In contrast, Westinghouse RVIs can experience 9 fluence in the range 1 x 10 21 to 5 x 10 22 n/cm 2 , or higher. Thus, 10 RVI are subject to neutron irradiation which is several orders 11 of magnitude higher than levels known to cause reduced fracture 12 toughness in reactor pressure vessel materials. [MRP 191 (Nov.
13 2006), Table 4-6 (Exh. NYS000321)].
14 Q. Are there other effects of embrittlement that can 15 compromise the ability to maintain a coolable core geometry in 16 the event of thermal or decompression shock loads following a 17 DBA LOCA?
18 A. Yes. As described previously, the synergistic 19 interactions between the metal degradation mechanisms associated 20 with fatigue, irradiation and corrosion are not well understood.
21 However, it is well known that irradiation embrittlement reduces 22 fracture toughness and decreases the resistance to crack 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 27 propagation in the metal. [USNRC Letter, Grimes to Newton, at 1 16 (Feb. 10, 2001) (Exh. NYS000324); Westinghouse Owners Group, 2 WCAP-14577, Rev. 1-A Report (March 2001), at 3-2 (Exh.
3 NYS000341)].
4 The radiation-induced damage to some RPV internals can be 5 extensive, since they can experience a neutron fluence of at 6 least 10 23 n/cm 2 at neutron energy (E) levels of E > 1 MeV (i.e., 7 > 100 dpa) [Was (2007) (Exh. NYS000339); EPRI, Dyle (2008) 8 (Exh. NYS000322); WOG WCAP-14577 Rev. 1-A Report (March 2001) 9 (Exh. NYS000341)] by the end of life (EOL) for extended 10 operations. According to one study, the crack growth rate for 11 materials irradiated to only 3x10 20 n/cm 2 fluence can be up to 40 12 times higher than that for unirradiated materials [Chopra, O.K., 13 "Degradation of LWR Core Internal Materials due to Neutron 14 irradiation," NUREG/CR-7027 (Dec. 2010) (Exh. NYS000487)]. It 15 should be stressed that the fluence experienced by some RPV 16 internals is about four orders of magnitude (i.e., ~ 10,000 17 times) larger than will be experienced by the inner wall of the 18 reactor pressure vessel by the end of life (EOL) for extended 19 operations [Rao, A.S. (USNRC), "Irradiation Assisted Degradation 20 of LWR Core Internal Materials; Brief Review," (Apr. 14, 2015) 21 (Exh. NYS000495)]. Thus, the RPV internals will be much more 22 embrittled than the RPV walls, which have historically been the 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 28 focus of USNRC embrittlement concerns. A highly embrittled RPV 1 internal component subjected to a severe earthquake or 2 thermal/decompression shock, could thus fail and relocate within 3 the RPV, which, in turn, could result in the loss of a coolable 4 core geometry.
5 GALL, Revision 1 6 Q. I show you a document marked as Exhibit NYS00146A-C 7 and entitled NUREG-1801, Revision 1, the Generic Aging Lessons 8 Learned Report, GALL. Are you familiar with this document?
9 A. Yes. 10 Q. When did the USNRC Staff release that document?
11 A. In September of 2005.
12 Q. Does NUREG-1801, Revision 1 include an aging 13 management program (AMP) for reactor pressure vessel internals 14 in a pressurized water nuclear reactor?
15 A. No. Revision 1 of NUREG-1801 includes no aging 16 management program description for PWR reactor pressure vessel 17 internals (RVIs). NUREG-1801, Revision 1,Section XI.M16, 18 entitled "PWR Vessel Internals," instead defers to the guidance 19 provided in Chapter IV line items as appropriate. The Chapter 20 IV line item guidance simply recommends actions to:
21 "..(1) participate in the industry programs for 22 investigating and managing aging effects on reactor 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 29 internals; (2) evaluate and implement the results of the 1 industry programs as applicable to the reactor internals; 2 and, (3) upon completion of these programs, but not less 3 than 24 months before entering the period of extended 4 operation, submit an inspection plan for reactor internals 5 to the NRC for review and approval."
6 That statement appears a number of times in GALL, Revision 7 1, Chapter IV. For example, that statement appears on pages IV 8 B2-4, IV B2-5, IV B2-8, IV B2-14, IV B2-16, and IV B2-17 with 9 respect to the embrittlement of reactor pressure vessel 10 internals.
11 Q. I show you what has been marked as Exhibit NYS000313, 12 which is a July 15, 2010 submission from Entergy that forwarded 13 a document to the Atomic Safety and Licensing Board (ASLB). Do 14 you recognize the attachment to that submission?
15 A. Yes, it contains a copy of a July 14, 2010 16 communication, NL-10-063, from Entergy to the USNRC's document 17 control desk that concerns embrittlement of reactor pressure 18 vessel internals. In addition, NL-10-063 contains an 19 "Attachment 1."
20 Q. Directing your attention to NL-10-063, Attachment 1, 21 page 84 of 90, what does Entergy say there about GALL, NUREG-22 1801, Revision 1 and reactor pressure vessel internals?
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 30 A. Entergy states that "Revision 1 of NUREG-1801 includes 1 no aging management program description for PWR reactor vessel 2 internals."
3 Standard Review Plan, Revision 1 4 Q. I show you a document marked as Exhibit NYS000195 that 5 is entitled NUREG-1800, Revision 1, USNRC Staff's Standard 6 Review Plan (SRP). Are you familiar with this document?
7 A. Yes. 8 Q. When did the USNRC Staff release that document?
9 A. In September of 2005.
10 Q. Does the Standard Review Plan, Revision 1 recognize 11 that the reactor pressure vessel internals could experience 12 embrittlement?
13 A. Yes, the Standard Review Plan, Revision 1 at § 14 3.1.2.2.6 recognized that reactor pressure vessel internals 15 could experience embrittlement.
16 Q. Would you elaborate?
17 A. In § 3.1.2.2.6 on page 3.1-5, the Standard Review 18 Plan, Revision 1 states, "Loss of fracture toughness due to 19 neutron irradiation embrittlement and void swelling could occur 20 in stainless steel and nickel alloy reactor vessel internals 21 components exposed to reactor coolant and neutron flux."
22 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 31 Q. Did the Standard Review Plan, Revision 1 make 1 provision for an aging management program (AMP) for reactor 2 pressure vessel internals in a pressurized water reactor?
3 A. No, it did not. At § 3.1.3.2.6, the Standard Review 4 Plan, Revision 1 stated that "The GALL Report recommends no 5 further evaluation of programs to manage loss of fracture 6 toughness due to neutron irradiation embrittlement . . ." That 7 statement is on page 3.1-12. This is also confirmed by § 8 3.1.2.2.6 and Table 3.1-1 which made clear that GALL and the 9 Standard Review Plan did not propose a specific aging management 10 plan and repeated the language from GALL about staying up to 11 date with industry discussions about embrittlement and 12 submitting a plan in the future for consideration by USNRC 13 Staff. 14 Entergy's Opposition to NYS Contention 25 15 Q. In November 2007 you submitted a declaration in 16 support of the State of New York's Contention 25 concerning 17 embrittlement. Do you know if Entergy submitted a response?
18 A. Yes, Entergy did.
19 Q. What did Entergy say in its response?
20 A. Entergy opposed the admission of Contention 25 and 21 presented various arguments. One of Entergy's principal 22 arguments was that stainless steel components are not 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 32 susceptible to a decrease in fracture toughness as a result of 1 neutron embrittlement. Entergy stated: "The core barrel, 2 thermal shield, baffle plates and baffle former plates 3 (including bolts) are, however, made of stainless steel and are 4 not susceptible to a decrease in fracture toughness as a result 5 of neutron embrittlement." [Entergy January 22, 2008 Answer at 6 137]. This is a surprisingly uninformed statement from the 7 operators of a nuclear power plant. Anyway, while this may have 8 been a popular belief many years ago, it is incorrect.
9 GALL, Revision 2 10 Q. I show you a document marked as Exhibit NYS00147A-D 11 that is entitled Revision 2 of the Generic Aging Lessons Learned 12 Report or GALL. Are you familiar with this document?
13 A. Yes, I have reviewed it.
14 Q. When did the USNRC Staff release that document?
15 A. December of 2010.
16 Q. What does GALL, Revision 2 say about embrittlement?
17 A. GALL, Revision 2 includes the following statement:
18 "Neutron irradiation embrittlement - Irradiation by neutrons 19 results in embrittlement of carbon and low-alloy steels. It may 20 produce changes in mechanical properties by increasing the 21 tensile and yield strengths with a corresponding decrease in 22 fracture toughness and ductility. The extent of embrittlement 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 33 depends on the neutron fluence, temperature, and trace material 1 chemistry." [GALL, Revision 2 at page IX-34 (Exh. NYS000147)].
2 I note that the phrase "low-alloy steels" includes stainless 3 steel. 4 Q. Does GALL, Revision 2 discuss the aging degradation of 5 PWR reactor pressure vessel internals?
6 A. Yes. Chapter IV and Chapter XI now discuss the aging 7 degradation of PWR reactor pressure vessel internals through 8 various aging mechanisms including embrittlement.
9 Q. What does GALL, Revision 2, Chapter IV state about 10 embrittlement of PWR reactor pressure vessel internals?
11 A. Chapter IV summarizes which reactor vessel internals 12 are subject to embrittlement (and other aging mechanisms) and is 13 organized by nuclear steam supply system vendors. There is a 14 section ("B2") concerning components in nuclear steam supply 15 systems designed by Westinghouse, the company that designed 16 those systems at Indian Point Unit 2 and Unit 3. That section 17 recognizes that reactor pressure vessel internals in 18 Westinghouse-designed PWRs are subject to degradation due to 19 embrittlement. It further recognizes that for Westinghouse 20 PWRs, reactor pressure vessel internal components made of 21 stainless steel and nickel alloy experience a "loss of fracture 22 toughness due to neutron irradiation embrittlement." These 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 34 statements appear on GALL, Revision 2 at pages IV B2-2 to IV B2-1 14. 2 Q. Directing your attention to GALL, Revision 2, pages IV 3 B2-12 and IV B2-13, do you see the items numbered IV.B2.RP-268 4 and IV.B2.RP-269?
5 A. Yes, those items concern reactor vessel internal 6 components in "inaccessible locations."
7 Q. What is the aging effect or mechanism of concern?
8 A. There are a number including loss of fracture 9 toughness due to neutron irradiation embrittlement, void 10 swelling, and corrosion-induced cracking.
11 Q. And these are inaccessible RPV internals in 12 Westinghouse PWRs?
13 A. Yes. 14 Q. Does GALL Revision 2 make any suggestions about the 15 reactor pressure vessel components that are located in 16 inaccessible locations?
17 A. Yes, it recommends an "evaluation" of the internals 18 located in inaccessible locations if other similar components 19 "indicate aging effects that need management."
20 Q. You mentioned that GALL, Revision 2, Chapter XI also 21 discussed reactor pressure vessel internals. Where is that 22 discussion?
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 35 A. Chapter XI contains a section numbered XI.M16A 1 entitled "PWR Vessel Internals," which starts at page XI M16A-1.
2 Q. Would you summarize that section?
3 A. Yes. Like Chapter IV, it recognizes that PWR reactor 4 pressure vessel internals experience a "loss of fracture 5 toughness due to either thermal aging or neutron irradiation 6 embrittlement," as well as other age-related degradation 7 mechanisms, such as various corrosion-induced cracking 8 mechanisms. It provides a template for license renewal 9 applicants to include in their license renewal applications that 10 discusses embrittlement and other aging mechanisms that degrade 11 reactor pressure vessel internals. It recommends that 12 applicants propose an inspection plan that is then submitted to 13 the USNRC Staff for review and approval. The template is 14 derived from a document prepared as a result of an effort 15 coordinated by the Electric Power Research Institute (EPRI) to 16 develop guidelines concerning the inspection of reactor pressure 17 vessel internals.
18 Q. Directing your attention to GALL, Revision 2, page XI 19 M16A-3, do you see item 3, titled "Parameters Monitored/
20 Inspected"?
21 A. Yes. 22 Q. Would you summarize that section?
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 36 A. Yes, this section provides recommendations for an 1 inspection plan for reactor pressure vessel internals, and 2 specifically what I would describe as the scope or focus of the 3 plan. This section is titled "Parameters Monitored/Inspected" 4 and states that the recommended inspection "program does not 5 directly monitor for loss of fracture toughness that is induced 6 by thermal aging or neutron embrittlement." Instead, it states 7 that the embrittlement of reactor pressure vessel internal 8 components is indirectly monitored through visual or volumetric 9 inspection techniques that look for cracking (i.e., the 10 detection of failures after they have occurred). It is 11 important to note that the focus of this document is on non-12 destructive testing (NDT) and non-destructive evaluation (NDE) 13 techniques. In particular it does not consider the implications 14 on core coolability subsequent of any shock load induced 15 failures of highly degraded RPV internals.
16 MRP-227, Revision 0 17 Q. I show you a document marked as Exhibit NYS00307A-D.
18 Do you recognize it?
19 A. Yes, I have reviewed it. It is a copy of the document 20 prepared as a result of the nuclear industry's efforts 21 coordinated by the Electric Power Research Institute (EPRI).
22 Q. What is the title of that document?
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 37 A. The document's title is, "Material Reliability 1 Program: Pressurized Water Reactor Internals Inspection and 2 Evaluation Guidelines (MRP-227-Rev. 0), 1016596, Final Report, 3 December 2008." Unfortunately, as I discussed previously, it is 4 focused on NDT and NDE inspection techniques rather than my 5 aging-related safety concerns.
6 MRP-227-A 7 Q. I show you a document marked as Exhibit NRC00114A-F 8 [MRP 227-A]. Do you recognize it?
9 A. Yes, I have reviewed it. It is the version of the 10 MRP-227, Revision 0 [Exh. NYS00307A-D] that was reviewed and 11 approved by the USNRC Staff, and includes various edits and 12 additional materials in response to USNRC Staff comments and 13 questions. It was submitted to the USNRC in January 2012.
14 Unfortunately, as I have noted previously, it is focused on NDT 15 and NDE inspection techniques rather than my aging-related 16 safety concerns.
17 Q. Does MRP-227-A say anything about embrittlement?
18 A. Yes. The industry has recognized that, "there are no 19 recommendations for inspection to determine embrittlement level 20 because these mechanisms cannot be directly observed" [MRP-227-21 A, Footnote 1 for Table 3-3 (December 2011) [Exh. NRC00114A-F].
22 That is, the level of degradation due to embrittlement of RPV 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 38 internal components, fittings and structures, and their ability 1 to withstand fatigue and shock loads cannot be determined using 2 the inspection techniques proposed in MRP-227-A.
3 Q. Do you have specific concerns with the approach to 4 aging management for reactor vessel internals set forth in MRP-5 227-A? 6 A. Yes. MRP-227-A is an inspection-based aging 7 management plan, which I believe is inadequate. To begin with, 8 depending on the type of component, inspection may not be 9 possible for the entire component, or for the entire set of such 10 components, given the location of the components and their 11 possible inaccessibility. For example, a visual or ultrasonic 12 inspection of the external head of a bolt does not necessarily 13 provide insight into the integrity of the remainder of the bolt 14 which is not visible. Moreover, an inspection focused on one 15 type of age-related degradation mechanism does not necessarily 16 work for another ongoing degradation process that is affecting 17 the same component, and the effect of shock loads on the 18 integrity of various RVIs and primary pressure boundary systems 19 is certainly not addressed by inspections. An inspection-based 20 approach to aging management, such as the one developed by the 21 nuclear industry in MRP-227 and condoned by USNRC in MRP-227-A, 22 is useful but it fails to account for the possibility that 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 39 highly embrittled and fatigued RVI components may not have signs 1 of degradation that can be detected by an inspection, but such 2 weakened components could nonetheless fail as a result of a 3 severe seismic event or thermal or pressure shock load. In 4 short, many of my concerns about the cumulative and ongoing 5 synergistic aging effects are not adequately addressed by MRP-6 227-A. 7 Entergy's License Renewal Application 8 Q. Directing your attention to Entergy's 2007 License 9 Renewal Application (LRA), did you find any indication in the 10 LRA that Entergy recognized that embrittlement could affect the 11 reactor pressure vessel?
12 A. Yes. 13 Q. Where was that?
14 A. The License Renewal Application at § 3.1.2.1.1 15 recognized that reactor pressure vessels are constructed of the 16 following materials:
17
- carbon steel; 18
- carbon steel with stainless steel or nickel alloy; 19
- cladding; 20
- nickel alloys; and, 21
- stainless steel.
22 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 40 The same LRA section further recognized that reactor pressure 1 vessels experience the following aging effects that require 2 management:
3
- cracking; 4
- loss of material; and, 5
- reduction of fracture toughness, a term which 6 encompasses embrittlement.
7 Q. Did you find any indication in the LRA that Entergy 8 has now recognized that embrittlement could affect reactor 9 pressure vessel internals?
10 A. Yes. 11 Q. Where was that?
12 A. The License Renewal Application at § 3.1.2.1.2 13 recognized that reactor pressure vessel internals are constructed 14 of the following materials:
15
- cast austenitic stainless steel (CASS);
16
- nickel alloy; and, 17
- stainless steel.
18 The same LRA section further recognized that the reactor 19 pressure vessel internals experience the following aging effects 20 that require management:
21
- change in dimensions; 22
- cracking; 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 41
- loss of material; 1
- loss of preload; and, 2
- reduction of fracture toughness, a term which, as 3 noted previously, encompasses embrittlement.
4 The 2007 LRA and the IP3 Reactor Pressure Vessel 5 Q. I direct your attention to License Renewal Application 6 Appendix A, § A.3.2.1.4. Do you have that?
7 A. Yes.
8 Q. What is that section of the License Renewal 9 Application concerned with?
10 A. That section concerns the IP3 reactor pressure vessel 11 itself. 12 Q. And what did Entergy say there?
13 A. Entergy stated that a part of the IP3 pressure vessel, 14 specifically plate B2803-3, exceeded the screening criteria for 15 pressurized thermal shock (PTS).
16 Q. Did Entergy acknowledge any specific concern about the 17 reactor pressure vessels at Indian Point?
18 A. Yes, Entergy acknowledged that with respect to IP3 19 that the reactor pressure vessel plate B2803-3 "exceeds the 20 screening criterion by 9.9°F." [Entergy January 22, 2008 Answer 21 at 139; citing LRA § A.3.2.1.4].
22 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 42 Q. What if anything did Entergy propose to do about the 1 IP3 pressure vessel?
2 A. Entergy proposed to submit to USNRC Staff a safety 3 analysis for plate B2803-3 three years before the plate reached 4 the reference temperature for pressurized thermal shock (RT PTS) 5 criterion.
6 The 2007 LRA and RPV Internals 7 Q. In your review of the April 2007 Indian Point License 8 Renewal Application, did you see an aging management program 9 (AMP) for reactor pressure vessel internals?
10 A. No, I did not. The 2007 License Renewal Application 11 did not contain an aging management program that specifically 12 focused on reactor pressure vessel internals. Rather, Appendix 13 A stated that sometime in the future Entergy would develop an 14 aging management program for the reactor pressure vessel 15 internals of their plants [LRA Appendix A, § A.2.1.41 with 16 respect to IP2, and § A.3.1.41 with respect to IP3]. This 17 deferred approach concerning IP2 and IP3 reactor pressure vessel 18 internals is also repeated at LRA, § 3.1.2.2.6.
19 Q. Do reactor pressure vessels and their associated 20 internal structures, components and fittings experience 21 embrittlement?
22 A. Yes. 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 43 Q. Are there any reactor pressure vessel internal 1 structures that are neglected in Entergy's discussion of future 2 programs it will develop to address such structures?
3 A. Yes. It should be noted that the control rods and 4 their associated guide tubes, plates, pins, and welds are not 5 highlighted, but they are also very important RPV internals and 6 their integrity is an extremely important safety concern. As I 7 have previously noted, they are located in the core region of 8 the RPV, and are inserted into the RPV through the upper head 9 via so-called stub tubes. Their function is to absorb excess 10 fission neutrons (i.e., those not needed to achieve a chain 11 reaction) so that the power level of a reactor can be 12 controlled. The control rods themselves are currently 13 considered by the USNRC to be moving components (which can be 14 replaced) and are thus not required to have an aging management 15 plan (AMP). Nevertheless, the other associated CRD structures, 16 components and fittings need an AMP since if these highly 17 embrittled structures, components and fittings are subjected to 18 significant shock loads they may fail, leading to possible core 19 cooling issues.
20 Q. Do you believe there are any special problems 21 associated with providing an adequate aging management program 22 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 44 for control rods and their associated guide tubes, plates and 1 welds? 2 A. Yes. For example, because of geometric 3 considerations, many PWRs (including IP2 and IP3) cannot meet 4 the USNRC's required minimum coverage for the non-destructive 5 testing (NDT) of the so-called "J-groove" welds [Entergy, 6 Walpole, NL-09-130 (Sept. 24, 2009) (Exh. NYS000311)], and thus 7 the integrity of these important CRD stub tube welds cannot be 8 directly confirmed by inspection. It appears that to help 9 address this chronic problem Entergy has ordered two new RPV 10 heads [Telecom-USNRC/Entergy Report (March 18, 2008) (Exh.
11 NYS000317)], but they have not yet been scheduled for 12 installation at Indian Point [Telecom-USNRC/Entergy (March 18, 13 2008) (Exh. NYS000317)]. In any event, unlike the rather 14 superficial treatment given this important safety concern by 15 Entergy [NL-10-063 (Exh. NYS000313)], I believe that a tangible, 16 enforceable, and viable aging management program (AMP) should be 17 developed and implemented before re-licensing the Indian Point 18 reactor plants for extended operations, since the integrity of 19 these CRD welds must be assured. If not, due to the leakage of 20 borated primary coolant through cracked welds, there can be 21 aggressive corrosion and wasting of the unclad outer surface of 22 the upper head of the RPVs (such as the serious event that 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 45 occurred at Davis-Besse and was identified in 2002). Worse yet, 1 there might be an inadvertent control rod ejection (due to a 2 massive failure of the welds in the upper RPV head), which could 3 cause a significant reactivity excursion, leading to core 4 melting and radiation releases.
5 Q. Are there places within the reactor pressure vessel 6 that you believe warrant particular aging management attention?
7 A. Yes. For the relicensing of the two reactors at 8 Indian Point, corrosion-induced cracking (e.g., SCC) and 9 radiation-induced embrittlement of the RPVs and their associated 10 internals is an important age-related safety concern, 11 particularly in the so-called "belt line" region of the RPV, 12 which is the region that is the closest to the reactor core. In 13 addition, as noted previously, the integrity of the so-called J-14 welds, which are part of the control rod drive seal in the upper 15 head of reactor pressure vessels, is important to avoid 16 corrosion-induced failures of the upper head and the possibility 17 of control rod ejection (and thus an uncontrolled reactivity 18 excursion).
19 Entergy's NL-10-063 Communication 20 Q. I direct your attention to Exhibit NYS000313. Do you 21 recognize it?
22 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 46 A. Yes, I have reviewed this document. As noted above, 1 it contains a copy of a July 14, 2010 communication, NL-10-063, 2 from Entergy to the USNRC document control desk that concerns 3 embrittlement of reactor pressure vessel internals. In turn, 4 NL-10-063 contains an "Attachment 1."
5 Q. Does Entergy make any statements here about 6 embrittlement of reactor pressure vessel internals?
7 A. Yes. Entergy acknowledges that, "PWR internals aging 8 degradation has been observed in European PWRs, specifically 9 with regard to cracking of baffle-former bolting." [NL-10-063, 10 at 89 (Exh. NYS000313)]. Entergy also states: "As with other 11 U.S. commercial PWR plants, cracking of baffle-former bolts is 12 recognized as a potential issue for the [Indian Point] units."
13 [NL-10-063, at 89 (Exh. NYS000313)]. Moreover, EPRI has stated 14 that, a "considerable amount of PWR internals aging degradation 15 has been observed in European PWRs." [EPRI MRP-227, at A-4 16 (Exh. NYS00307A-D)]. Material degradation has also been 17 observed in control rod guide tube alignment (split) pins [EPRI 18 MRP-227, at A-4 (December 2008) (Exh. NYS00307A-D)]. It is 19 important to note that MRP-227 has also recommended that 20 analysis be done to show when it is acceptable to continue to 21 operate PWRs in which there have been bolt failures (e.g., due 22 to embrittlement and/or fatigue). While this type of temporary, 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 47 short-term "fix" might be adequate for normal operations, it may 1 lead to structural and component failures due to the shock loads 2 associated with various postulated accidents. If so, the failed 3 internal structures and components may relocate, cause core 4 blockages, or otherwise result in uncoolable core geometry, and 5 thus lead to seriously degraded core cooling, core melting and 6 massive radiation releases.
7 Q. Do you have any additional problems with the 8 inspection program for RVIs as proposed in the MRP-227 and 9 adopted by Entergy?
10 A. Yes. With respect to Entergy's proposal to conduct 11 baseline examinations of the RPV internals (RVIs), it should be 12 noted that I have previously called on Entergy to conduct such 13 examinations and for USNRC Staff to require the conduct of such 14 examinations before entering the period of extended operations 15 [See November 2007 Declaration of Richard T. Lahey, Jr., at ¶¶ 16 24, 25 (Exh. NYS000298); see also State of New York Notice of 17 Intention to Participate and Petition to Intervene, at 217-220, 18 State of New York Contention-23 (Baseline Inspections)].
19 Fortunately, both the USNRC and Entergy now seem to have 20 embraced the concept of baseline inspections for RPV internals, 21 but the proposed aging management program (AMP) as set forth in 22 NL-10-063 lacks sufficient details to know when the baseline 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 48 inspections of the RPV and its internals will begin and end, and 1 the scope of these inspections. Thus, it is not possible to 2 know whether the proposed baseline inspections will be 3 comprehensive and adequate.
4 Q. Are there other problems that you believe need to be 5 addressed if Entergy is to have an adequate aging management 6 program for RPV internals?
7 A. Yes. My Report provides more details on my concerns 8 with Entergy's failure to conduct an evaluation of the 9 synergistic impacts of embrittlement, corrosion-induced 10 cracking, and metal fatigue on the degradation of RPV internals, 11 and its failure to consider how those interacting degradation 12 mechanisms will impact the ability of the RPV internals to 13 withstand the effect of thermal and decompression shock loads as 14 a result of a DBA LOCA. I am also concerned that the design of 15 the inspection programs -- including their frequency, the type 16 of inspections to be conducted, the acceptance criteria and the 17 criteria for actions to be taken in the event of a failure of a 18 component -- does not consider these synergistic degradation 19 mechanisms. Finally, Entergy's AMP for RPV internals does not 20 include specific programs with objective criteria for either 21 preventative measures or for corrective actions to be taken when 22 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 49 inspections show that certain components are not able to safely 1 undergo extended plant operations.
2 Entergy's NL-11-107 Communication 3 Q. I show you what has been marked as Exhibit NYS000314.
4 Do you recognize that document?
5 A. Yes, this is a copy of Entergy's September 28, 2011 6 communication, NL-11-107, with the USNRC's document control 7 desk. 8 Q. Would you please turn to Table 5-2 at page 36 of the 9 Attachment to NL-11-107.
10 A. Yes, I have that.
11 Q. What does the document say there?
12 A. In discussing the baffle-former assemblies and their 13 related baffle-edge bolts, it recognizes that irradiated-14 assisted stress corrosion cracking and fatigue can cause 15 cracking which, in turn, leads to failed or missing bolts 16 connecting a baffle to a former.
17 Q. What else does communication NL-11-107 state?
18 A. In it, Entergy tells the USNRC that it has completed 19 commitment number 30 wherein Entergy stated that it would submit 20 an inspection plan to the USNRC for reactor pressure vessel 21 internals (RVIs) no later than two years before the plant 22 entered the period of extended operations. However, none of my 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 50 safety concerns associated with the synergistic effects of 1 embrittlement, fatigue and corrosion on the integrity of RPV 2 internals, and post-accident core coolability (i.e., due to 3 shock load induced failures), were addressed. In my opinion an 4 adequate inspection plan for RPV internals is a necessary, but 5 not sufficient, means of assuring safe extended plant 6 operations. Indeed, a systematic safety evaluation of the 7 degraded RPV internals is also needed to identify the limiting 8 structures, components and fittings that need to be repaired or 9 replaced before the onset of extended operations.
10 Entergy's Amended and Revised RVI Plan, and 11 USNRC Staff's November 2014 SSER2 12 13 Q. I direct your attention to Exhibits NYS000496 through 14 NYS000506. Do you recognize these exhibits?
15 A. Yes, I have reviewed these documents. In 16 communication NL-12-037, dated February 17, 2012 [Exh.
17 NYS000496], the applicant submitted an amendment to its license 18 renewal application entitled "Revised Reactor Vessel Internals 19 Program and Inspection Plan." Thereafter, the applicant 20 explained and modified this proposed plan in response to various 21 requests for information (RAIs) from the USNRC. [Exhs.
22 NYS000497 through NYS000506]. Collectively, I will refer to 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 51 this collection of communications as the applicant's "Amended 1 and Revised RVI Plan."
2 Q. I direct your attention to Exhibit NYS000507. Do you 3 recognize this exhibit?
4 A. Yes, I have reviewed this exhibit. It is the Second 5 Supplemental Safety Evaluation Report, or SSER2, prepared by 6 USNRC Staff and released in November 2014. In the SSER2, the 7 USNRC Staff evaluated and approved the applicant's Amended and 8 Revised RVI Plan.
9 Q. Does the SSER2 discuss the potential synergism between 10 various aging mechanisms?
11 A. Yes, to some degree. The USNRC recognized the 12 potential synergy between thermal and irradiation embrittlement 13 for cast austenitic stainless steel components (CASS). [SSER2 14 at 3-42 (Exh. NYS000507)]. In particular, in its Safety 15 Evaluation Report (SER) for MRP-227, the USNRC Staff 16 acknowledged the potential for synergistic interaction between 17 embrittlement and other aging mechanisms. For example, the 18 USNRC noted that "the synergistic effects of SCC, fatigue, and 19 thermal embrittlement . . . could potentially cause greater 20 degradation in the welds [of Combustion Engineering lower 21 support columns] than just the consideration of IASCC 22 (irradiation assisted stress corrosion cracking) and irradiation 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 52 embrittlement alone. Degradation in these welds could then be 1 equivalent to or greater than other components susceptible only 2 to IASCC and irradiation embrittlement due to the synergistic 3 effects." [SE at 15 (Exh. NYS000309)]. The USNRC staff could 4 have - indeed, should have - made the same observation about 5 potential synergistic aging effects for Westinghouse RVI 6 components, fittings, and structures at IP2 and IP3.
7 Q. Does the applicant's Amended and Revised RVI Plan say 8 anything about preventative actions to manage aging effects?
9 A. Yes. In Attachment 1 to NL-12-037, Entergy has 10 indicated that the Amended and Revised RVI Plan "is a condition 11 monitoring program that does not include preventative actions."
12 [Attachment 1 to NL-12-037, at 5 (Exh. NYS000496)]. Generally, 13 the applicant continues to approach the problem of synergistic 14 aging effects on RVI components through "condition monitoring" 15 (i.e., periodic inspections per MRP-227-A) rather than a 16 comprehensive approach which includes detailed analyses and/or 17 preventative actions (i.e., repair and replacement) ["Revised 18 Reactor Vessel Internals Program and Inspection Plan,"
19 Attachment 1 to NL-12-037, at 5 (Exh. NYS000496)]. This 20 approach implies that aging effects and degradation will not be 21 addressed until cracks or other degradation mechanisms (e.g., 22 wear) have been directly observed ["Revised Reactor Vessel 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 53 Internals Program and Inspection Plan," Attachment 1 to NL 1 037, at 5 (Exh. NYS000496)].
2 In short, component degradation will be addressed only 3 after it occurs. The applicant incorrectly concludes that 4 preventative actions, such as component replacement, are not 5 required for most RVI components because cracking or other flaws 6 can be detected before the failure of a component affects the 7 safe operation of the reactor. This is apparently based on the 8 erroneous assumption that IP2 and IP3 will continuously operate 9 during the 20-year period of extended operation within normal 10 "steady-state" parameters. Entergy ignores the possibility that 11 significantly fatigued, embrittled and corrosion-weakened, or 12 otherwise degraded, RVI components, structures, or fittings may 13 be exposed to various shock loads which can cause them to deform 14 or relocate and thereby impair core cooling. In fact, the 15 applicant's reactor safety analyses implicitly assume that the 16 reactor core will maintain a coolable geometry during emergency 17 core cooling system (ECCS) operation subsequent to a DBA LOCA, 18 notwithstanding the degradation and possible deformation or 19 relocation of various RVI components and potential flow 20 blockages and degraded core cooling which may result.
21 Q. Does Entergy make any statements about the degradation 22 of RVI components in the "Amended and Revised RVI Plan"?
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 54 A. Yes, similar to NL-10-063 (Exh. NYS000313), the 1 applicant acknowledges, in NL-12-037, that other PWRs have 2 experienced material degradation and failure of multiple RVI 3 components, including cracking of baffle-former bolting, 4 cracking in other important bolting, wear in thimble tubes, and 5 potential wear in control rod guide tube guide plates 6 [Attachment 1 to NL-12-037, at 8 (Exh. NYS000496)]. Also, the 7 applicant has committed to replace one affected IP2 component -
8 the degraded guide tube support pins (split pins) - by 2016 9 ["Revised Reactor Vessel Internals Program and Inspection Plan,"
10 Attachment 1 to NL-12-037, at 8 (Exh. NYS000496); Commitment 50, 11 Attachment 1 to NL-13-122, at 7 (Exh. NYS000502)].
12 Interestingly, the applicant has agreed to replace the IP2 split 13 pins, even though they were already replaced once in 1995, and 14 even though the applicant claims that the failure of a split pin 15 would not compromise reactor vessel functions [Response to RAI 16 16, Attachment 2 to NL-12-166, at 1 (Exh. NYS000500)]. However, 17 for many other affected RVI components, the applicant proposes a 18 "wait-and-see" approach.
19 Q. Could you provide an example?
20 A. Yes. The applicant acknowledges that "cracking of 21 baffle former bolts is recognized as a potential issue for the 22 Indian Point units" ["Revised Reactor Vessel Internals Program,"
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 55 Attachment 1 to NL-12-037, at 8 (Exh. NYS000496)], but the 1 applicant does not propose to replace the degraded bolts, only 2 to continue monitoring them ["Revised Reactor Vessel Internals 3 Inspection Plan," Attachment 2 to NL-12-037, at 40, tbl. 5-2 4 (Exh. NYS000496)]. In fact, the applicant has not yet developed 5 inspection acceptance criteria for baffle former bolts in either 6 IP2 or IP3 [SSER, at 3-20 (Exh. NYS000507)]. Instead, the 7 applicant has agreed to develop a technical justification 8 including acceptance criteria for baffle former bolts sometime 9 prior to the first round of inspections, which might not occur 10 until 2019 for IP2 and 2021 for IP3 [SSER2, at 3-20 (Exh.
11 NYS000507); Response to RAI 5, Attachment 1 to NL-12-089, at 11 12 (Exh. NYS000497)].
13 Another example of the applicant's "wait-and-see" approach 14 for the RVIs is the applicant's proposal for managing aging 15 effects on the clevis insert bolts. [SSER2, at 3-23 to 3-26 16 (Exh. NYS000507)]. Like the split pins that the applicant is 17 replacing in IP2 for the second time, clevis insert bolts are 18 susceptible to primary water stress corrosion cracking (PWSCC) 19 [MRP-227-A, Appendix A, at A-2 (Exh. NRC00114A-F)]. Failures of 20 clevis insert bolts, apparently caused by PWSCC, were detected 21 at a Westinghouse-designed reactor in 2010. Out of 48 clevis 22 bolts in this reactor, 29 were partially or completely fractured 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 56 but only 7 of those damaged bolts were visually detected as 1 having failed [SSER2, at 3-25 (Exh. NYS000507)]. Despite this 2 high rate of failure (about 60% of the total bolts were damaged) 3 and low rate of visual detection (only about 24% of the damaged 4 bolts were detected), the applicant proposes to manage the aging 5 degradation of clevis insert bolts with visual (VT-3) 6 inspections rather than pre-emptive replacement ["Revised 7 Reactor Vessel Internals Inspection Plan," Attachment 2 to NL-8 12-037, tbl. 5-4, at 51 (Exh. NYS000496)].
9 The applicant apparently acknowledges that visual 10 inspections will not detect the majority of clevis bolt cracks 11 prior to failure, but justifies this approach on the grounds 12 that "crack detection prior to bolt failure is not required due 13 to design redundancy" [Response to RAI 17, Attachment 1 to NL-14 13-122, at 8 (Exh. NYS000502)]. In fact, the applicant appears 15 to suggest that the failure of multiple clevis insert bolts will 16 not seriously affect the operation of the reactor. The 17 applicant then analyzes the effect of clevis bolt failures on 18 various other components.
19 The applicant's analysis of the effects of clevis bolt 20 failures assumes that all other components will be functioning 21 according to their design specifications, and does not consider 22 the fact that the other components may also be undergoing 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 57 degradation from various interacting aging mechanisms.
1 Moreover, the applicant fails to consider the possibility that a 2 shock load (e.g., due to a LOCA) may cause the sudden failure of 3 the remaining intact clevis bolts, which, in turn, may lead to 4 an uncoolable core geometry. In short, rather than taking 5 proactive steps to replace the degraded clevis bolts prior to 6 failure, the applicant proposes to wait for clevis bolt failures 7 to occur before taking steps to address the problem, an approach 8 which is totally unacceptable in my opinion.
9 The baffle former bolts and clevis insert bolts are just 10 two examples of Entergy's overarching approach to RVI aging 11 management, which foregoes preventative component repair or 12 replacement in favor of running the reactor until detectable 13 damage or component failure occurs.
14 Q. Do you have any other concerns regarding specific 15 components discussed in the Amended and Revised RVI Plan?
16 A. Yes. The applicant's approach for analyzing the lower 17 support structures' functionality and fracture toughness is also 18 flawed [Response to RAI-11-A, Attachment 1 to NL-13-052, at 1-4 19 (Exh. NYS000501)]. The applicant suggested that irradiation 20 embrittlement effects would only be significant in the presence 21 of pre-existing flaws or service induced defects, together with 22 a stress level capable of crack propagation. In its analysis, 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 58 the applicant, based on the lack of documented fractures of core 1 support columns, "assumed that only a limited number of columns 2 could actually contain flaws of significant size." The 3 applicant further assumed that the columns would be subject to 4 "nominal normal operating stresses" [SSER2, at 3-43 (Exh.
5 NYS000507)]. When the USNRC Staff inquired about the most 6 recent visual inspections of the core support structures, the 7 applicant acknowledged that the CASS support column caps were 8 inaccessible to inspection and that VT-3 visual inspection 9 offered "no meaningful information regarding the structural 10 integrity of the columns." [Id. at 3044.] Under these 11 circumstances, the applicant's conclusion that irradiation-12 induced cracking of core support columns is "unlikely" 13 represents wishful thinking and is contrary to recent studies 14 [e.g., NUREG/CR-7184, at xv (Revised December 2014) (Exh.
15 NYS00488A-B)], which show the extreme sensitivity of crack 16 growth rate and fracture toughness to irradiation. Moreover, it 17 ignores the fact that these and other non-CASS RVI structures 18 and components undergo a range of aging degradation mechanisms 19 simultaneously under steady-state and transient conditions, and 20 that their embrittlement or susceptibility to fracture simply 21 cannot always be adequately detected using currently available 22 inspection techniques.
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 59 Also, not all of the core support structures are accessible 1 for inspection, so surrogate structures have been chosen by 2 Entergy to assess age-related degradation mechanisms. For 3 example, the girth weld of the core barrel has been proposed by 4 the applicant as a leading indicator for irradiation-induced 5 embrittlement (IE) and irradiation-assisted stress corrosion 6 cracking (IASCC) of the core support column caps, even though 7 these components are very different, and they may be exposed to 8 different degradation mechanisms and shock loads. In fact, as 9 pointed out recently by a member of the ACRS Plant License 10 Renewal Subcommittee, "[t]he relationship between a lower core 11 barrel weld and the tops of these columns is a bit of a stretch 12 . . . [t]hey're totally different type of components, totally 13 different loadings." Moreover, to have a failure due to a 14 seismic event "you don't even need to have a crack if these 15 columns are really brittle . . . ." [ACRS Plant License Renewal 16 Subcommittee Transcript, at 209-211 (April 23, 2015) (Exh.
17 NYS000526)].
18 Q. Does the applicant's Amended and Revised RVI Plan 19 adequately account for the potential cumulative effect of 20 synergistic aging mechanisms on RVIs?
21 A. No. By merely relying on MRP 227-A for its aging 22 management plan, the applicant has ignored the large 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 60 uncertainties that exist with respect to the effects of 1 irradiation-induced aging phenomena. [Chen, et al., at xv (Exh.
2 NYS000488A); NUREG/CR-7153, Vol. 2: Aging of Core Internals and 3 Piping Systems, at 181, 187, 210-211 (Exh. NYS00484A-B);
4 Stevens, et al. (October 2014), at 9-10 (Exh. NYS000491)].
5 While the applicant's Thermal Aging and Neutron Irradiation of 6 Cast Austenitic and Stainless Steel (CASS) program generally 7 recognizes the potential adverse synergistic effects of elevated 8 coolant temperature and irradiation on the fracture toughness of 9 CASS materials, a broader recognition of this principle is 10 needed by the applicant, since RVI components made from non-cast 11 stainless steel will also experience the combined effects of 12 irradiation-induced embrittlement, corrosion, and other aging 13 mechanisms. The applicant has failed to evaluate the 14 synergistic mechanisms that occur for many other important and 15 vulnerable RVI components, such as the core baffles, baffle 16 bolts, and formers. Compared to the baffles, baffle bolts, and 17 formers, the core support columns (which are obviously very 18 important incore structures) are located in an area of the 19 reactor pressure vessel which is subject to less radiation 20 fluence (and thus are less susceptible to embrittlement).
21 Q. Do you have any other concerns with the applicant's 22 Amended and Revised RVI Plan?
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 61 A. Yes. The applicant proposes to rely on visual (VT-3) 1 inspection techniques for many RVI components. However, there 2 are significant shortcomings of this technique to detect 3 material cracking, degradation, or wear prior to failure, as has 4 been noted by USNRC staff [Tregoning, at 2-3 (Exh. NYS000508);
5 Case, at 1 (Exh. NYS000509)], and illustrated by the visual 6 detection of only 7 out of 29 fractured clevis insert bolts at a 7 Westinghouse PWR in 2010 [SSER2, at 3-25 (Exh. NYS000507)]. In 8 an RAI to Entergy, the USNRC staff observed that "VT-3 visual 9 examination may not be adequate for all components for detecting 10 fatigue cracking prior to the occurrence of structurally 11 significant cracking." [Attachment 1 to NL-13-052, at 5 (May 7, 12 2013) (Exh. NYS000501)]. Moreover, as I have noted previously, 13 the level of embrittlement can not be detected at all using 14 visual inspection techniques.
15 Fatigue 16 Q. Turning to fatigue. Could you explain what fatigue 17 is? 18 A. Yes. Fatigue is another important age-related 19 degradation mechanism. It is one of the primary considerations 20 when conducting a time limited aging analysis (TLAA) and an 21 aging management program (AMP) for nuclear power plants.
22 Fatigue of various structures, components and fittings in a 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 62 nuclear reactor can result in piping and pressure boundary 1 component and fitting ruptures, physical failures, and the 2 relocation of loose pieces of RVI metal throughout the reactor 3 system, which, in turn, may result in core blockages and 4 interfere with the effective core cooling of a nuclear power 5 plant. My main concerns about fatigue are the increased 6 potential for a primary or secondary side LOCA, and the failure 7 of various RPV internals (RVIs). It should be noted that the 8 fatigue life of a PWR component, fitting or structure is 9 normally evaluated in terms of a cumulative usage factor (CUF) 10 which is corrected for the degradation in fatigue life due to 11 the reactor coolant environment (i.e., CUF en). The cumulative 12 usage factor is defined as, CUF
= N/Na-AIR , where N is the number 13 of the various fatigue cycles that have occurred (or are 14 expected by the end of plant life, EOL), and Na-AIR is the number 15 of allowable fatigue cycles obtained from data (taken in air) at 16 which failure (i.e., significant surface cracking) is expected.
17 The observed degradation in fatigue that occurs due to hot 18 reactor coolant is quantified by an environmental fatigue 19 correction factor, F en = Na-AIR/Na-RC , where Na-AIR is the allowable 20 number of fatigue cycles measured in air, and Na-RC is the 21 allowable fatigue cycles measured in a simulated reactor coolant 22 (RC) enviroment; thus, CUF en = CUF x F en = N/Na-RC. The criterion 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 63 for acceptance by the USNRC is that CUF en < 1.0 by the end of 1 life (EOL) for the component, fitting or structure in question.
2 Anyway, the allowable cycles to failure (Na-RC) are determined 3 from small scale experiments using metal test samples which are 4 exposed to simulated reactor coolant environments.
5 Unfortunately, to date, there have not been any systematic 6 fatigue experiments done in simulated reactor coolant 7 environments using highly embrittled metal test samples, which 8 have less fatigue life than ductile materials. That is, the 9 synergistic degradation effect of embrittlement has not been 10 included in CUF en (= N/Na-RC) evaluations, thus the results are 11 expected to be non-conservative since the denominator (Na-RC) will 12 be too large, and thus CUF en will be too small.
13 Q. I show you what has been marked as Exhibit NYS000527.
14 Do you recognize this document?
15 A. Yes. This is Entergy's Fatigue Monitoring Plan for 16 IP2 and IP3.
17 Q. Is Entergy required to conduct fatigue evaluations of 18 internal and external components?
19 A. Yes. In this proceeding, the applicant agreed, in 20 Commitments 33, 43 and 49, to calculate the CUF en for 21 external(i.e., primary pressure boundary) and internal (RVI) 22 components in certain locations [Dacimo, Fred, Entergy, letter 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 64 to Document Control Desk, USNRC, "Reply to Request for 1 Additional Information Regarding the License Renewal 2 Application," NL-13-122 (September 27, 2013), at 20 3 (NYS000502)]. Additionally, the USNRC has recently proposed to 4 require all applicants for license renewal to evaluate the 5 fatigue life of limiting components beyond those originally 6 specified in NUREG/CR-6260, and to evaluate the effect of 7 reactor coolant environment on the fatigue life of both 8 external and internal (i.e., RVIs) structures and systems [79 9 Fed. Reg. 69,884 (November 24, 2014) (NYS000522); USNRC, Draft 10 Regulatory Guide DG-1309 (Proposed Revision 1 of Regulatory 11 Guide 1.207, dated March 2007), "Guidelines for Evaluating the 12 Effects of Light-Water Reactor Coolant Environments in Fatigue 13 Analyses of Metal Components" (November 2014) (NYS000523)].
14 Q. In your expert opinion, has Entergy done adequate 15 fatigue evaluations to assure the safety of their two nuclear 16 power plants at the Indian Point site during extended 17 operations?
18 A. No. 19 20 21 22 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 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.
Contention NYS-25 66 1 2 3 4 5 6 Unfortunately, Westinghouse did 7 not perform error analyses to quantify the modeling 8 uncertainties and the effect of code user interactions in its 9 fatigue calculations using WESTEMS. It would appear that 10 virtually any error would put some of the calculated values of 11 CUF en over the CUF en = 1.0 fatigue failure limit.
12 Q. I show you a document marked as Exhibit NYS000513. Do 13 you recognize it?
14 A. Yes, it is a paper presented by Westinghouse at a 15 recent Pressure Vessels & Piping Conference of the American 16 Society of Mechanical Engineers held in Anaheim, California in 17 July 2014; it is entitled "License Renewal and Environmental 18 Fatigue Screening Application" and its authors were Mark Gray 19 and Christopher Kupper.
20 Q. Are you familiar with its contents?
21 A. Yes, I have reviewed this article and it clearly shows 22 the iterative process used by Westinghouse in which safety 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 67 margin is removed in its environmentally assisted fatigue (EAF) 1 calculations in an effort to reduce the output or result below 2 CUF en = 1.0. 3 Q. Returning to our discussion about Westinghouse's 4 fatigue evaluations of reactor coolant pressure boundary 5 components, has Entergy addressed the issue of fatigue in the 6 context of shock loads?
7 A. No, 8 9 10 11 12 13 14 15 16 17 18 19 Even 20 assuming this CUF en calculation is accurate, it does not account 21 for the possibility that a highly fatigued component, which does 22 not yet have signs of significant surface cracking, may be 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 68 exposed to an unexpected seismic event or shock load that could 1 cause it to fail. This is a good example of the type of "silo 2 thinking" (i.e., the fatigue and safety analyses are treated 3 entirely separately) that NYS is concerned about.
4 Q. I show you your Supplemental Report, which has been 5 marked as Exhibit NYS000297. I note that the State has 6 provisionally designated it as containing confidential 7 information. Would you provide a brief summary of the Report?
8 A. I prepared this Supplemental Report to set out some of 9 my concerns about the use of the WESTEMS computer code to 10 develop a cumulative fatigue analysis of certain components in 11 the Indian Point reactors and their reactor coolant pressure 12 boundaries.
13 Q. Would you briefly summarize your concerns?
14 A. Yes. First, I am concerned that without an error 15 analysis it is difficult to be in a position to meaningfully 16 analyze the results of the 2010 and subsequently refined CUF en 17 analyses presented by Entergy and Westinghouse.
18 Q. Why is an error analysis important?
19 A. It is well known that all engineering analyses are 20 based on imperfect mathematical models of reality and various 21 code user assumptions which inherently involve some level of 22 error. Error analyses help readers and decision makers 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 69 understand what level of confidence to attach to the calculated 1 results and the proposed conclusions.
2 Q. Is the preparation of an error analysis an accepted 3 practice in the field of engineering?
4 A. Yes. Engineers frequently prepare error analyses. In 5 my submissions in this proceeding I noted that one would 6 normally expect to see at least a hybrid 'propagation-of-error' 7 type of analysis [Kline & McClintock (1953) (Exh. NYS000514)] to 8 determine the overall uncertainty in the CUF en results given by 9 Westinghouse. I also referenced a standard enineering text 10 book, "Basic Engineering Data Collection and Analysis," pp. 310-11 311, by Vardeman & Jobe [2001], to demonstrate the various types 12 of error analyses which are regularly done by engineers [Exh.
13 NYS000347].
14 Q. I show you what has been marked as Exhibit NYS000515.
15 Are you familiar with it?
16 A. Yes, it is a recent USNRC inspection report with 17 notices of non-conformance for Westinghouse's Quality Assurance 18 Program. In that report, the USNRC determined that Westinghouse 19 failed to adequately implement its QA program in the areas of 20 corrective actions, oversight of suppliers, and audits. Since 21 Entergy relies on Westinghouse services to, among other things, 22 provide appropriate guidance on corrective action and other 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 70 activities affecting safety-related functions, the USNRC's 1 findings of non-conformance are all the more unsettling. Under 2 these circumstances, the USNRC should insist that an error 3 analysis be performed to ensure the validity of Westinghouse's 4 fatigue evaluations for IP2 and IP3 components.
5 Q. Are you aware of an instance where an error analysis 6 was prepared for a project at Indian Point?
7 A. Yes, for example, in 1980, the Consolidated Edison 8 Company of New York prepared an error analysis in support of a 9 proposal to add more spent fuel into the spent fuel pool at 10 Indian Point Unit 2.
11 Q. I show you what has been marked as Exhibit NYS000348; 12 do you recognize it?
13 A. Yes. That is a copy of the 1980 Con Edison error 14 analysis for the re-racking of spent fuel in the Unit 2 spent 15 fuel pool.
16 Q. Do you have other concerns about the refined CUF en 17 reanalysis?
18 A. Yes, as discussed in my Supplemental Report, I am 19 concerned that engineering judgment or user intervention could 20 have affected the results. I note that when USNRC Staff issued 21 the Supplemental Safety Evaluation Report, Staff instructed 22 Entergy and Westinghouse, on a going forward basis, to document 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 71 and disclose the use of engineering judgment and user 1 intervention when conducting future fatigue analysis using the 2 WESTEMS code. This is noted in Exhibit NYS000160 at page 4-2.
3 To my knowledge, Westinghouse has provided such information for 4 some, but not all of the fatigue evaluations performed to date.
5 Also, USNRC Staff instructed Entergy not to use WESTEMS when 6 conducting analyses under the ASME Standard know as NB-3600 [at 7 4-2, 4-3 (Exh. NYS000160)]. Furthermore, I am concerned about 8 the analytical framework employed by the WESTEMS code. As 9 detailed, in my Supplemental Report, I believe that the code's 10 thermal-hydraulic models and framework are too simplified to 11 predict accurate results.
12 13 14 15 16 17 18 Q. Do you know how Entergy is proposing to address 19 fatigue as part of its overall aging management plan for IP2 and 20 IP3? 21 A. Yes. Entergy has an existing Fatigue Monitoring Plan 22 for addressing metal fatigue. The program is designed to 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 72 monitor operational cycles and transients so that the various 1 CUF en remain below unity. The company has also proposed to 2 include RVIs in its Fatigue Monitoring Program; however, for the 3 reasons already stated, WESTEMS may be non-conservative, so any 4 program that relies on values, such as the fatigue cycles to 5 failure, derived from the WESTEMS methodology is inherently 6 unreliable for ensuring that aging RVIs avoid failure. This is 7 particularly true when these results include no accompanying 8 error analysis.
9 Conclusion 10 Q. Could you summarize your general concerns with the 11 applicant's license renewal application?
12 A. Yes, I am very concerned that Entergy has continually 13 eroded the safety margins and conservatisms built into the 14 current licensing basis for the Indian Point reactors. For 15 example, Entergy has relied on CUF en calculations that remove 16 various conservatisms but are still very close to unity (the 17 fatigue failure limit). Entergy also relies on the detection of 18 degradation, wear or cracking prior to component failure, rather 19 than repairing or replacing the aging parts - particularly the 20 RVIs - preemptively. Moreover, Entergy implicitly assumes that 21 the plant will operate in a steady-state, and has not taken into 22 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 73 account unanticipated severe seismic events or thermal/pressure 1 shock loads which can cause failures to occur.
2 As reactors and their constituent components age, it 3 becomes very important to preserve - rather than erode -
4 operational safety margins. Uncertainties exist in all systems, 5 and calculation or modeling mistakes are always possible. For 6 example, the USNRC recently became aware that certain 7 methodologies prescribed in its NUREG-0800 Branch Technical 8 Position (BTP) 5-3 for estimating the initial fracture toughness 9 of reactor vessel materials may be non-conservative. [See, 10 e.g., Troyer, et al., "An Assessment of Branch Technical 11 Position 5-3 to Determine Unirradiated RTNDT for SA-508 Cl.2 12 Forgings," Paper No. PCP2014-28897, Proceedings of the ASME 2014 13 Pressure Vessels and Piping Conference, Anaheim, California 14 (July 20-24, 2014) (Exh. NYS00516); Letter from Pedro Salas, 15 Regulatory Affairs Director, AREVA, to USNRC regarding Potential 16 Non-conservatism in NRC Branch Technical Position 5-3 (January 17 30, 2014)(Exh. NYS000517); USNRC, Slides, "Assessment of BTP 5-3 18 Protocols to Estimate RTNDT(u) and USE (June 4, 2014) (Exh.
19 NYS000518); NUREG-0800, Rev. 2 (Exh. NYS000521)].
20 21 22 Anyway, 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 74 since unexpected errors of this type do occur, maintaining 1 safety margins helps to guard against potentially adverse 2 impacts due to precisely this type of unexpected finding of non-3 conservatism in safety evaluations. Lastly, I would like to 4 note that at a recent American Society of Mechanical Engineers 5 (ASME) Pressure Vessels & Piping Conference, USNRC staff also 6 highlighted newly-identified non-conservatisms in sections of 7 the ASME Code regarding fracture toughness applicable to nuclear 8 reactor operations. [Kirk, M. et al., "Assessment of Fracture 9 Toughness Models for Ferritc Steels Used in Section XI of the 10 ASME Code Relative to Current Data-Based Model," PVP 2014-28540 11 (Exh. NYS000520)]. This is yet another reason to preserve, 12 rather than erode safety margins in the aging management of 13 light water nuclear reactors (e.g., PWRs).
14 Q. You have reviewed NUREG-1801, GALL Report Revision 1 15 (Exh. NYS00146A-C); NUREG-1800, Standard Review Plan Revision 1 16 (Exh. NYS000195); NUREG-1801, GALL Report Revision 2 (Exh.
17 NYS00147A-D); NUREG-1800, Standard Review Plan Revision 2 (Exh.
18 NYS000161); EPRI's MRP-227 Revision 0 (Exh. NYS00307A-D); EPRI's 19 MRP-227-A (Exh. NYS000507); Entergy's July 2010 NL-10-063 20 communication, Entergy's February 2012 NL-12-037 communication 21 (Exh. NYS000313) and subsequent communications constituting its 22 Amended and Revised RVI Plan (Exhs. NYS000496-506); USNRC 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 75 Staff's June 22, 2011 Safety Evaluation of MRP-227 Revision 0 1 (Exh. NYS000309); NUREG-1930, USNRC Staff's August 30, 2011 2 Supplemental Safety Evaluation (Exh. NYS000160); and NUREG-1930, 3 USNRC Staff's November 2014 Second Supplement Safety Evaluation 4 Report for the Indian Point License Renewal Application (Exh.
5 NYS000507); and Entergy's NL-11-107 communication (Exh.
6 NYS000314), correct?
7 A. Yes. 8 Q. Do you have any opinion about those documents with 9 respect to the degradation of reactor pressure vessel internals 10 (RVIs)? 11 A. Yes. 12 Q. Please summarize your testimony.
13 A. As I stated in my initial November 2007 declaration in 14 support of the State of New York's Contentions 25 and 26, my 15 April 2008 declaration in support of Contention NYS-26A, my 16 September 2010 declarations in support of the State's 17 supplemental filings on Contentions NYS-25 and NYS-26B/RK-TC-1B, 18 and my previously filed testimony on Contentions NYS-25 and NYS-19 26B/RK-TC-1B, and my February 2015 declaration in support of the 20 State's further supplemental filings on Contention NYS-25, in my 21 professional judgment Entergy has failed to demonstrate that it 22 has adequately accounted for the aging phenomena of 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 76 embrittlement and fatigue for structures, components and 1 fittings inside the reactor pressure vessels (i.e., RVIs) at 2 Indian Point Unit 2 and Indian Point Unit 3. My professional 3 judgment has not fundamentally changed based upon Entergy's July 4 14, 2010 submission of License Renewal Application, Amendment 5 No. 9 [NL-10-063 (Exh. NYS000313)], Entergy's September 28, 2011 6 submission of NL-11-107 [Exh. NYS000314], or Entergy's Amended 7 and Revised RVI Plan, consisting of the February 17, 2012 8 submission of NL-12-037 [Exh. NYS000496] as amended by its 9 subsequent communications [Exhs. NYS000497-506] and approved by 10 the USNRC Staff in the SSER2 [Exh. NYS000507]. I do not believe 11 that Entergy's July 15, 2010 communication to the Board [NL 12 063 (Exh. NYS000313)] concerning a new AMP for RPV internals, or 13 its September 28, 2011 communication [NL-11-107 (Exh.
14 NYS000314)], are adequate to address the safety concerns and 15 technical issues that I have raised herein. They do not address 16 my age-related safety concerns, nor do they recognize the 17 importance of the various synergistic degradation mechanisms 18 that I am concerned with. The Amended and Revised RVI Plan, 19 which the USNRC Staff evaluated and approved in the November 20 2014 SSER2, also does not resolve my concerns over the 21 simultaneous and synergistic age-related degradation mechanisms 22 that may affect various RVI components and structures.
23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 77 While some age-related safety issues might eventually be 1 resolved analytically or experimentally, in many cases it 2 appears that the easiest and most cost-effective way to resolve 3 them is to simply repair or replace the most seriously degraded 4 structures, components and fittings, and this approach is what 5 NYS has been proposing for some time (particularly for the 6 degraded RVIs).
7 Q. Does this conclude your testimony?
8 A. Not quite. I want to stress that during the course of 9 my involvement in these relicensing proceedings I have 10 discovered what I believe to be some important new age-related 11 safety concerns which, to the best of my knowledge, have not 12 been previously considered in relicensing proceedings. These 13 concerns include: the synergistic effect on the degradation and 14 integrity of RPV internals (RVIs) of radiation-induced 15 embrittlement, corrosion and fatigue, and the potential for the 16 unanticipated failure of RPV internals (RVIs) due to a severe 17 seismic event or accident-induced thermal and/or pressure shock 18 loads, and the implications of the failure of RPV internal 19 structures, components and fittings (i.e., RVIs) on post-20 accident core coolability. While in the past many of these 21 issues and concerns have been noted separately, the implications 22 of their synergistic interaction has apparently been overlooked 23 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 78 and not evaluated (i.e., they have been evaluated in "silos").
1 Since I first raised these technical issues in 2007, the USNRC, 2 DOE and various nuclear industry groups have slowly begun to 3 recognize their significance. In fact, the evaluation and study 4 of these important issues is underway, but major uncertainties 5 still exist. As a consequence, I believe that these important 6 age-related safety concerns must be resolved in order to have 7 assurance that the Indian Point reactors can operated safely 8 beyond their design life of 40 years. Indeed, I believe that 9 the most vulnerable RPV internals (RVIs) need to be carefully 10 identified and repaired or replaced prior to extended operations 11 since it is beyond the current state-of-the-art to perform 12 realistic and accurate calculations on the relocation of failed 13 RPV internals (RVIs) and the resultant potential for core 14 blockages and degraded core cooling.
15 Q. Does this complete your testimony.
16 A. Yes, it does. I do, however, reserve the right to 17 supplement my testimony if new information is disclosed or 18 introduced.
19 20 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 79 UNITED STATES 1 NUCLEAR REGULATORY COMMISSION 2 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 3 -----------------------------------x 4 In re: Docket Nos. 50
-247-LR; 50-286-LR 5 License Renewal Application Submitted by ASLBP No. 07
-858-03-LR-BD01 6 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 7 Entergy Nuclear Indian Point 3, LLC, and 8 Entergy Nuclear Operations, Inc.
June 9 , 2015 9 -----------------------------------x 10 DECLARATION OF RICHARD T. LAHEY, JR.
11 I, Richard T. Lahey, Jr., do hereby declare under penalty 12 of perjury that my statements in the foregoing testimony and my 13 statement of professional qualifications are true and correct to 14 the best of my knowledge and belief.
15 Executed in Accord with 10 C.F.R. § 2.304(d) 16 17 ________________________
18 Dr. Richard T. Lahey, Jr.
19 The Edward E. Hood Professor Emeritus of Engineering 20 Rensselaer Polytechnic Institute, Troy, NY 12180 21 (518) 495-3884, laheyr@rpi.edu 22 Revised Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-25 80