ML11361A395
| ML11361A395 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 12/27/2011 |
| From: | Lahey R Rensselaer Polytechnic Institute |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| Shared Package | |
| ML11361A394 | List: |
| References | |
| RAS 21650, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01 | |
| Download: ML11361A395 (46) | |
Text
Exhibit NYSR00344 December 27, 2011 1 UNITED STATES 2 NUCLEAR REGULATORY COMMISSION 3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4 -----------------------------------x 5 In re: Docket Nos. 50-247-LR; 50-286-LR 6 License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 7 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 8 Entergy Nuclear Indian Point 3, LLC, and 9 Entergy Nuclear Operations, Inc. December 22, 2011 10 -----------------------------------x 11 PRE-FILED WRITTEN TESTIMONY OF 12 RICHARD T. LAHEY, JR.
13 REGARDING CONSOLIDATED CONTENTION NYS-26B / RK-TC-1B 14 On behalf of the State of New York (NYS or the State),
15 the Office of the Attorney General hereby submits the following 16 testimony by RICHARD T. LAHEY, JR. regarding Contention 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 the Edward E. Hood Professor Emeritus of 21 Engineering at Rensselaer Polytechnic Institute (RPI), which is 22 located in Troy, New York.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 1
1 Q. Please summarize your educational and professional 2 qualifications.
3 A. I have earned the following academic degrees: a B.S.
4 in Marine Engineering from the United States Merchant Marine 5 Academy, a M.S. in Mechanical Engineering from Rensselaer 6 Polytechnic Institute, a M.E. in Engineering Mechanics from 7 Columbia University, and a Ph.D. in Mechanical Engineering from 8 Stanford University. I have held various technical and 9 administrative positions in the nuclear industry, and I have 10 served as both the Dean of Engineering and the Chairman of the 11 Department of Nuclear Engineering & Science at RPI. Previously, 12 I was responsible for nuclear reactor safety R&D (research &
13 development) for the General Electric Company (GE), and I have 14 extensive experience with both military (i.e., naval) and 15 commercial nuclear reactors. Also, I am a member of a number of 16 professional societies and have served on various expert panels.
17 I was also an Editor of the international Journal of Nuclear 18 Engineering & Design, which focuses on nuclear engineering and 19 nuclear reactor safety technology. I am widely considered to be 20 an expert in matters relating to the design, operations, safety, 21 and aging of nuclear power plants.
22 Q. Which professional societies are you a member of?
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 2
1 A. I am a member of a number of professional societies, 2 including: the American Nuclear Society (ANS), where I was a 3 member of the Board of Directors and the ANSs Executive 4 Committee, and was the founding Chair of the ANSs Thermal-5 Hydraulics Division; the American Society of Mechanical 6 Engineers (ASME), where I was Chair of the Nucleonics Heat 7 Transfer Committee, K-13; the American Institute of Chemical 8 Engineering (AIChE), where I was the Chair of the Energy 9 Transport Field Committee; and the American Society of 10 Engineering Educators (ASEE), where I was Chair of the Nuclear 11 Engineering Division.
12 Q. What expert panels have you served on?
13 A. I have served on numerous panels and committees for 14 the United States Nuclear Regulatory Commission (USNRC), Idaho 15 National Engineering Laboratory (INEL), Oak Ridge National 16 Laboratory (ORNL), and the Electric Power Research Institute 17 (EPRI). I am a member of the National Academy of Engineering 18 (NAE), and have been elected Fellow of both the ANS and the 19 ASME.
20 A. Have you published any papers in the field of nuclear 21 engineering and nuclear reactor safety technology?
22 Q. Yes. Over the last 50 years, I have published 23 numerous books, monographs, chapters, articles, reports, and Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 3
1 journal papers on nuclear engineering and nuclear reactor safety 2 technology. Those articles are listed in my Curricula Vitae.
3 Q. Have you received any professional awards?
4 A. Yes, I have received many honors and awards for my 5 career accomplishments, including: the E.O. Lawrence Memorial 6 Award of the Department of Energy (DOE), the Glenn Seaborg Medal 7 of the ANS and the Donald Q. Kern Award of the AIChE.
8 Q. I show you what has been marked as Exhibit NYS000295.
9 Do you recognize that document?
10 A. Yes. It is a copy of my Curricula Vitae, which 11 summarizes, among other things, my experience, publications, and 12 awards.
13 Q. I show you what has been marked as Exhibit NYS000299 14 to Exhibit NYS000303. Do you recognize those documents?
15 A. Yes. They are copies of the six declarations that I 16 previously prepared to date for the State of New York in this 17 proceeding. They include the initial declaration that was 18 submitted in November 2007 in support of the State's petition to 19 intervene and its initial contentions, the April 7, 2008 20 declaration in support of Contention NYS-26A, the September 15, 21 2010 declaration submitted in support of the State's 22 supplemental bases for Contention 25, the September 9, 2010 23 declaration submitted in support of the amended Contention Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 4
1 NYS26B/RK-TC-1B, and the September 30, 2011 and November 1, 2011 2 declarations submitted in support of Contention NYS-38/RK-TC-5.
3 Q. I show you what has been marked as Exhibit NYS000296.
4 Do you recognize that document?
5 A. Yes. It is a copy of the Report that I prepared for 6 the State of New York in this proceeding. The Report reflects 7 my analysis and opinions.
8 Q. I show you what has been marked as Exhibit NYS000297.
9 Do you recognize that document?
10 A. Yes. This is a copy of a Supplemental Report that I 11 prepared for the State of New York in this proceeding that 12 addresses aspects of the revised fatigue analysis that Entergy 13 and Westinghouse prepared concerning certain components the 14 Indian Point reactors. The supplemental report also reflects my 15 analysis and opinions.
16 Q. What is the purpose of your testimony?
17 A. I was retained by the State of New York State to 18 review Entergy's application to the U.S. Nuclear Regulatory 19 Commission (USNRC) and its Staff for two renewed operating 20 licenses for the nuclear power plants known as Indian Point Unit 21 2 and Unit 3. I have reviewed the License Renewal Applications 22 (LRAs) and subsequent filings by Entergy and the USNRC Staff.
23 My declarations and report discuss my concerns and opinions Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 5
1 about issuing twenty-year operating licenses for these 2 facilities. My testimony seeks to identify and discuss some 3 age-related safety concerns which have not yet been addressed by 4 Entergy. In my opinion these concerns must be resolved to 5 assure the health and safety of the American public, 6 particularly those in the vicinity of the Indian Point reactors.
7 Q. Have you reviewed various materials in preparation for 8 your testimony?
9 A. Yes.
10 Q. What is the source of those materials?
11 A. I have reviewed documents prepared by government 12 agencies, Entergy, Westinghouse, the utility industry, or its 13 associations, and various related text books and peer-reviewed 14 articles.
15 Q. I show you Exhibits NYS00146A-C, NYS00147A-D, 16 NYS000160, NYS000161, NYS000195, and NYS000304 through 17 NYS000369. Do you recognize these documents?
18 A. Yes. These are true and accurate copies of some of 19 the documents that I referred to, used, or relied upon in 20 preparing my report, declarations, and this testimony. In some 21 cases, where the document was extremely long and only a small 22 portion is relevant to my testimony, an excerpt of the document Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 6
1 is provided. If it is only an excerpt, that is noted on the 2 first page of the Exhibit.
3 Q. I direct your attention to latter part of your Report 4 (Exh. NYS000296) entitled "Reference Documents," which contains 5 a list of documents. Would you describe that list?
6 A. Yes that section of the Report list various salient 7 documents that I referred to, used or relied on, in preparing my 8 Report and the Supplemental Report 9 Q. How do these documents relate to the work that you do 10 as an expert in forming opinions such as those contained in this 11 testimony?
12 A. These documents represent the type of information that 13 persons within my field of expertise reasonably rely upon in 14 forming opinions of the type offered in this testimony.
15 16 The Indian Point Reactors 17 Q. Are you familiar with the power reactors that are the 18 subject of this proceeding?
19 A. Yes.
20 Q. Would you briefly describe them?
21 A. Entergy operates two power reactors that are located 22 in northern Westchester County near the Village of Buchanan.
23 The operating reactors are known as the Indian Point Unit 2 and Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 7
1 Indian Point Unit 3 power reactors. These Westinghouse-designed 2 plants are 4-loop pressurized water reactors (PWRs), and they 3 are currently rated at power levels of 3,216.4 MWt. Entergy 4 also owns another reactor at the same site. That reactor is 5 known as the Indian Point Unit 1 reactor; however, that reactor 6 has been shut down and no longer produces power.
7 8 Operation of a Pressurized Water Reactor 9 Q. Would you briefly describe the design and operation of 10 a pressurized water reactor?
11 A. Pressurized water nuclear reactors have water (i.e.,
12 the primary coolant) under high pressure flowing through the 13 core in which heat is generated by the fission process. The 14 core is located inside a reactor pressure vessel (RPV). This 15 heat is absorbed by the coolant and then transferred from the 16 coolant in the primary system to lower pressure water in the 17 secondary system via a large heat exchanger (i.e., a steam 18 generator) which, in turn, produces steam on the secondary side.
19 These steam generator systems, which are part of the plants 20 Nuclear Steam Supply System (NSSS), are located inside a large 21 containment structure. After leaving the containment building, 22 via main steam piping, the steam drives a turbine, which turns a 23 generator to produce electrical power.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 The reactor pressure vessel is a large steel container that 2 holds the core (i.e., the nuclear fuel); it also serves as a key 3 part of the primary coolant's pressure boundary.
4 As its name suggests, this reactor design uses a 5 pressurizer on the primary side that performs several functions.
6 In particular, it maintains the operating pressure on the 7 primary side of the nuclear reactor and accommodates variations 8 in reactor coolant volume for load changes during reactor 9 operations, as well as reactor heat-up and cool-down. The 10 reactor coolant also moderates the neutrons produced in the core 11 since a pressurized water nuclear reactor will not function 12 unless the neutrons are moderated (i.e., slowed down due to 13 collisions with the hydrogen 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 identifies various materials that are used or contained in those 23 components.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 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" includes 14 various structures, components and fittings inside the reactor 15 pressure vessel including the: core barrel (and its welds),
16 core baffle, intermediate shells, former plates, lower core 17 plate and support structures, clevis bolts, fuel alignment pins, 18 thermal shield, the lower support column and mixer, upper mixing 19 vanes, and the upper/lower core assemblies and support column, 20 and the control rods and their associated guide tubes, plates, 21 and welds. Reactor pressure vessel internals also include the 22 bolts that hold various components together or to other 23 components including: the baffle-to-baffle bolts, the core Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 10
1 barrel-to-former bolts, and baffle-to-former bolts as well as 2 the welds or weldments that hold sections of these components 3 together.
4 Q. Coming back to Exhibit NYS000306, would you describe 5 the other diagrams?
6 A. Yes. They are a collection of additional schematic 7 figures from the Electric Power Research Institutes Report MRP-8 227 that provide additional detail concerning various reactor 9 pressure vessel internals and their location within the reactor 10 pressure vessel. The reactor pressure vessel internals shown 11 include the control rod guide tube assembly, the control rod 12 guide cards, guide tube support pins, the control rods, baffles, 13 formers, baffle-former assemblies, baffle to former bolts, 14 corner edge bracket baffle to former bolts, core barrel to 15 former bolts, baffle plate edge bolts, core support, and various 16 weldments, including welds in the reactor pressure vessel for 17 the core barrel plates.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 Overview 2 Q. In your expert opinion what is the most important age-3 related safety issue associated with the relicensing of the two 4 Indian Point reactors?
5 A. The over-arching age-related safety issues that I am 6 concerned with have to do with the fact that the various 7 degradation mechanisms and the safety analyses have been done 8 separately (i.e., in silos), and thus it has been implicitly 9 assumed that there is no interaction between them. In addition, 10 the fatigue evaluations which were presented by Entergy (i.e.,
11 those done for them by Westinghouse using the WESTEMS computer 12 code) focused on the fatigue of various piping systems and 13 piping fittings, but not the fatigue of reactor pressure vessel 14 internals. Moreover, the metric used to appraise their fatigue-15 induced damage (CUFen) only considered quasi-static low and high 16 cycle fatigue evaluations, rather than accident-induced shock 17 loads which may cause failures well before the standard fatigue 18 limit is reached (i.e., CUFen = 1.0 ). Also, no error analyses 19 were presented to quantify the WESTEMS predictions for various 20 piping systems and fittings even though some of them were 21 extremely close to the CUFen = 1.0 limit.
22 Anyway, in my opinion, probably the most serious short 23 coming of this siloing approach is that synergistic Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 interactions between radiation-induced embrittlement, stress 2 corrosion cracking, and fatigue-induced degradation mechanisms 3 have not been considered; for example, the fatigue-induced 4 failure of seriously embrittled reactor pressure vessel (RPV) 5 internals. Also, when the plants safety analyses were done by 6 Entergy it was implicitly assumed that the incore geometry would 7 remain intact during postulated accidents. Unfortunately, 8 unlike ductile metals, seriously embrittled and fatigued RPV 9 internals may not be able to survive the pressure and/or thermal 10 shock loads induced by various accidents and severe transients.
11 If not, they can fail and relocate, possibly causing core 12 blockages that degrade core cooling and may lead to core melting 13 and massive radiation releases.
14 Entergy has an obligation to show that their plants can be 15 safely operated beyond the 40 year design life of these plants.
16 I believe that this will require much more study and analysis 17 than has been presented to date to identify any limiting RPV 18 internals that require repair or replacement. Nevertheless, 19 this must be done to verify that the two Indian Point reactors 20 can be safely operated for another 20 years beyond the design 21 life of these plants.
22 Q. Would you please explain in more detail the various 23 degradation mechanisms that you are concerned with?
Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 A. Yes, let me begin with Embrittlement.
2 3 Embrittlement 4 Q. Would you explain what embrittlement is?
5 A: Embrittlement refers to the change in the mechanical 6 properties (or structure) of materials, such as metals, that can 7 occur over time under the bombardment of neutrons. The degree 8 of exposure to neutrons is normally expressed in terms of a 9 "fluence" (i.e., the neutron flux times the duration of the 10 irradiation process). The extended exposure to neutrons causes 11 damage to metals and makes them more brittle so that they become 12 more susceptible to failures due to cracking or fracture. In 13 particular, this radiation-induced damage results in a decrease 14 in fracture toughness and ductility.
15 Embrittlement is an age-related degradation mechanism 16 whereby a component experiences a decrease in ductility, a loss 17 of facture toughness, and an increase in strength. While the 18 initial aging effect is loss of ductility and toughness, 19 unstable crack propagation is the eventual aging effect if a 20 crack is present and the local applied stress intensity exceeds 21 the reduced fracture toughness. This is particularly true for 22 the large pressure and/or thermal shock loads associated with 23 postulated accidents.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 The USNRC's guidance document GALL, Revision 2 includes the 2 following statement: "Neutron irradiation embrittlement - -
3 Irradiation by neutrons results in embrittlement of carbon and 4 low-alloy steels. It may produce changes in mechanical 5 properties by increasing the tensile and yield strengths with a 6 corresponding decrease in fracture toughness and ductility. The 7 extent of embrittlement depends on the neutron fluence, 8 temperature, and trace material chemistry." That is from GALL, 9 Revision 2 at page IX-34 (Exhibit NYS000147). I note that the 10 phrase "low-alloy steels" includes stainless steel.
11 12 The Consequences of Embrittlement 13 Q. Is embrittlement a concern for pressurized water 14 nuclear reactors.
15 A. Yes. For a pressurized water nuclear reactor to 16 operate safely, the metals involved need to be sufficiently 17 ductile, which means that they must be able to deform without 18 experiencing failures. When metals, such as steel, experience a 19 significant fluence, which happens to the materials in close 20 proximity to the reactor core (e.g., the steel reactor pressure 21 vessels interior wall and the associated RPV internals), the 22 temperature required for them to maintain some ductility is 23 increased as the metal is continually bombarded by a neutron Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 15
1 flux. The temperature at which there is a marked change from 2 ductile to non-ductile behavior is often called the "nil 3 ductility temperature" (NDT). However, even for temperatures 4 well above the NDT, the irradiated metals continue to be damaged 5 and further embrittled due to the neutron bombardment. Indeed, 6 the neutron damage will not be annealed out (i.e., be 7 neutralized) unless the damaged metals are taken to temperatures 8 that are well above PWR operating temperatures.
9 Q. Could embrittlement impact a nuclear reactor's ability 10 to respond to a transient, shock load, or an accident scenario?
11 A. Yes. A degradation in ductility (or embrittlement) 12 will adversely affect a PWR's ability to withstand pressure 13 and/or thermal shock loads, and thus there is a threat to the 14 integrity of highly embrittled internal structures in the 15 reactor pressure vessel.
16 Various accidents can expose a reactor pressure vessel and 17 its internal structures, components and fittings to significant 18 pressure and/or thermal shock loads. If the reactor pressure 19 vessels internal structures are sufficiently degraded due to 20 stress corrosion cracking (SCC), fatigue and/or radiation-21 induced embrittlement, such accidents can have significant 22 consequences. Indeed, the resultant stresses from such 23 accidents may cause the reactor pressure vessel internals to Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 16
1 fail structurally. If so, the ability to effectively cool the 2 decay heat in the core may be lost due to core blockage.
3 One well known safety concern associated with embrittlement 4 is the ability of metals to withstand a thermal shock event. A 5 thermal shock can occur in various ways, for example: (1) during 6 a loss of coolant accidents (LOCAs), or, (2) during a reactor 7 SCRAM (i.e., a rapid insertion of the control rods which 8 terminates the nuclear chain reaction). A particularly bad LOCA 9 event is one in which there is a rapid depressurization of the 10 secondary side which causes a reactor SCRAM and thus a rapid 11 cooling of the primary coolant via the steam generators. This 12 accident leads to a severe pressurized thermal shock of the 13 reactor pressure vessel and the associated RPV internals.
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 failure of a highly embrittled reactor pressure vessel 21 internal structures, components and fittings and thus the 22 inability to subsequently cool the core. It should be noted 23 that in the past most of the attention has been focused on the Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 17
1 integrity of the reactor pressure vessel. However, the RPV 2 internals are much less massive and are much closer to the core, 3 and thus suffer a lot more radiation damage.
4 Q. Are there other effects of embrittlement that can 5 compromise the ability to retain a coolable core geometry in the 6 event of thermal or decompression shock loads following a DBA 7 LOCA?
8 A. Yes. How the rather complex, interacting metal 9 degradation mechanisms associated with fatigue, irradiation and 10 corrosion interact is still an area of research (e.g., how 11 fatigue-induced surface cracks propagate in an embrittled, as 12 opposed to ductile, metal structure). Nevertheless, it is well 13 known that, the effects of embrittlement, especially loss of 14 fracture toughness, make existing cracks in the affected 15 materials and components less resistant to growth [USNRC 16 Letter, Grimes to Newton, pg. 16 (Feb. 10, 2001) (Exh.
17 NYS000324)]. Indeed, it is also well known that, irradiation 18 embrittlement decreases the resistance to crack propagation, 19 [Westinghouse Owners Group WCAP-14577 Rev. 1-A Report, pg. 3-2 20 (March 2001) (Exh. NYS00307A-D)]. Anyway, the radiation-induced 21 damage to RPV internals can be extensive, since they can 22 experience a neutron fluence of at least 1023 n/cm2 at neutron 23 energy (E) levels of E > 1 MeV (i.e., > 100 dpa) [Was, 2007 Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 18
1 (NYS000339); EPRI (Dyle), 2008 (NYS000322); WOG WCAP-14577 Rev.
2 1-A Report, March 2001 (NYS000341)] by the end of life (EOL) for 3 extended operations. It should be noted that the fluence 4 experienced by some RPV internals is about four orders of 5 magnitude ( i.e., ~ 10,000 times ) larger than will be 6 experienced by the inner wall of the reactor pressure vessel by 7 the end of life (EOL) for extended operations. Thus, the RPV 8 internals will be much more embrittled.
9 10 GALL, Revision 1 11 Q. I show you a document marked as Exhibit NYS00146A-C 12 and entitled NUREG-1801, Revision 1, the Generic Aging Lessons 13 Learned Report, GALL. Are you familiar with this document?
14 A. Yes.
15 Q. When did the USNRC Staff release that document?
16 A. In September 2005.
17 Q. Does NUREG-1801, Revision 1 include an aging 18 management program (AMP) for reactor pressure vessel internals 19 in a pressurized water nuclear reactor?
20 A. No. Revision 1 of NUREG-1801 includes no aging 21 management program description for PWR reactor pressure vessel 22 internals. NUREG-1801, Revision 1,Section XI.M16, entitled 23 "PWR Vessel Internals," instead defers to the guidance provided Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 in Chapter IV line items as appropriate. The Chapter IV line 2 item guidance simply recommends actions to:
3 "..(1) participate in the industry programs for 4 investigating and managing aging effects on reactor 5 internals; (2) evaluate and implement the results of the 6 industry programs as applicable to the reactor internals; 7 and, (3) upon completion of these programs, but not less 8 than 24 months before entering the period of extended 9 operation, submit an inspection plan for reactor internals 10 to the NRC for review and approval."
11 That statement appears a number of times in GALL, Revision 12 1, Chapter IV. For example, that statement appears on pages IV 13 B2-4, IV B2-5, IV B2-8, IV B2-14, IV B2-16, and IV B2-17 with 14 respect to the embrittlement of reactor pressure vessel 15 internals.
16 Q. I show you what has been marked as Exhibit NYS000313, 17 which is a July 15, 2010 submission from Entergy that forwarded 18 a document to the Atomic Safety and Licensing Board (ASLB). Do 19 you recognize the attachment to that submission?
20 A. Yes, it contains a copy of a July 14, 2010 21 communication, NL-10-063, from Entergy to USNRC document control 22 desk that concerns embrittlement of reactor pressure vessel 23 internals. In addition, NL-10-063 contains an "Attachment 1."
Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 Q. Directing your attention to NL-10-063, Attachment 1, 2 page 84 of 90, what does Entergy say there about GALL, NUREG-3 1801, Revision 1 and reactor pressure vessel internals?
4 A. Entergy states that "Revision 1 of NUREG-1801 includes 5 no aging management program description for PWR reactor vessel 6 internals."
7 8 Standard Review Plan, Revision 1 9 Q. I show you a document marked as Exhibit NYS000195 that 10 is entitled NUREG-1800, Revision 1, the USNRC Staff's Standard 11 Review Plan (SRP). Are you familiar with this document?
12 A. Yes.
13 Q. When did the USNRC Staff release that document?
14 A. In September 2005.
15 Q. Does the Standard Review Plan, Revision 1 recognize 16 that the reactor pressure vessel internals could experience 17 embrittlement?
18 A. Yes, the Standard Review Plan, Revision 1 at § 19 3.1.2.2.6 recognized that reactor pressure vessel internals 20 could experience embrittlement.
21 Q. Would you elaborate?
22 A. In § 3.1.2.2.6 on page 3.1-5, the Standard Review 23 Plan, Revision 1 states, "Loss of fracture toughness due to Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 21
1 neutron irradiation embrittlement and void swelling could occur 2 in stainless steel and nickel alloy reactor vessel internals 3 components exposed to reactor coolant and neutron flux."
4 Q. Did the Standard Review Plan, Revision 1 make 5 provision for an aging management program (AMP) for reactor 6 pressure vessel internals in a pressurized water reactor?
7 A. No, it did not. At § 3.1.3.2.6, the Standard Review 8 Plan, Revision 1 stated that "The GALL Report recommends no 9 further evaluation of programs to manage loss of fracture 10 toughness due to neutron irradiation embrittlement . . ." That 11 statement is on page 3.1-12. This is also confirmed by § 12 3.1.2.2.6 and Table 3.1-1 which made clear that GALL and the 13 Standard Review Plan did not propose a specific aging management 14 plan and repeated the language from GALL about staying up to 15 date with industry discussions about embrittlement and 16 submitting a plan in the future for consideration by USNRC 17 Staff.
18 19 Entergy's Opposition to NYS Contention 25 20 Q. In November 2007 you submitted a declaration in 21 support of the State of New York's Contention 25 concerning 22 embrittlement. Do you know if Entergy submitted a response?
23 A. Yes, it did.
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1 Q. What did Entergy say in its response?
2 A. Entergy opposed the admission of Contention 25 and 3 presented various arguments. One of Entergy's principal 4 arguments was that stainless steel components are not 5 susceptible to a decrease in fracture toughness as a result of 6 neutron embrittlement. Entergy stated: "The core barrel, 7 thermal shield, baffle plates and baffle former plates 8 (including bolts) are, however, made of stainless steel and are 9 not susceptible to a decrease in fracture toughness as a result 10 of neutron embrittlement." Entergy January 22, 2008 Answer at 11 137. While this was a popular belief some years ago, 12 unfortunately it is incorrect.
13 14 GALL, Revision 2 15 Q. I show you a document marked as Exhibit NYS00147A-D 16 that is entitled Revision 2 of the Generic Aging Lessons Learned 17 Report or GALL. Are you familiar with this document?
18 A. Yes, I have reviewed it.
19 Q. When did the USNRC Staff release that document?
20 A. December 2010.
21 Q. Does GALL, Revision 2 discuss the aging degradation 22 of PWR reactor pressure vessel internals?
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1 A. Yes. Chapter IV and Chapter XI discuss the aging 2 degradation of PWR reactor pressure vessel internals through 3 various aging mechanisms including embrittlement.
4 Q. What does GALL, Revision 2, Chapter IV state about 5 embrittlement of PWR reactor pressure vessel internals?
6 A. Chapter IV summarizes which reactor vessel internals 7 are subject to embrittlement (and other aging mechanisms) and is 8 organized by nuclear steam supply system vendors. There is a 9 section ("B2") concerning components in nuclear steam supply 10 systems designed by Westinghouse, the company that designed 11 those systems at Indian Point Unit 2 and Unit 3. That section 12 recognizes that reactor pressure vessel internals in 13 Westinghouse-designed PWRs, are subject to degradation due to 14 embrittlement. It further recognizes that for Westinghouse 15 PWRs, reactor pressure vessel internal components made of 16 stainless steel and nickel alloy experience a "loss of fracture 17 toughness due to neutron irradiation embrittlement." These 18 statements appear on GALL, Revision 2 at pages IV B2-2 to IV B2-19 14.
20 Q. Directing your attention to GALL, Revision 2, pages IV 21 B2-12 and IV B2-13, do you see the items numbered IV.B2.RP-268 22 and IV.B2.RP-269?
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1 A. Yes, those items concern reactor vessel internal 2 components in "inaccessible locations."
3 Q. What is the aging effect or mechanism of concern?
4 A. There are a number including loss of fracture 5 toughness due to neutron irradiation embrittlement, void 6 swelling, and stress corrosion cracking.
7 Q. And these are inaccessible RPV internals in 8 Westinghouse PWRs?
9 A. Yes.
10 Q. Does GALL Revision 2 make any suggestions about the 11 reactor pressure vessel components that are located in 12 inaccessible locations?
13 A. Yes, it recommends an "evaluation" of the internals 14 located in inaccessible locations if other components "indicate 15 aging effects that need management."
16 Q. You mentioned that GALL, Revision 2, Chapter XI also 17 discussed reactor pressure vessel internals. Where is that 18 discussion?
19 A. Chapter XI contains a section numbered XI.M16A 20 entitled "PWR Vessel Internals," which starts at page XI M16A-1.
21 Q. Would you summarize that section?
22 A. Yes. Like Chapter IV, it recognizes that PWR reactor 23 pressure vessel internals experience a "loss of fracture Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 25
1 toughness due to either thermal aging or neutron irradiation 2 embrittlement," as well as other age-related degradation 3 mechanisms, such as various stress corrosion cracking 4 mechanisms. It provides a template for license renewal 5 applicants to include in their license renewal applications that 6 discusses embrittlement and other aging mechanisms that degrade 7 reactor pressure vessel internals. It recommends that 8 applicants propose an inspection plan that is then submitted to 9 the USNRC Staff for review and approval. The template is 10 derived from a document prepared as a result of an effort 11 coordinated by the Electric Power Research Institute (EPRI) to 12 develop guidelines concerning reactor pressure vessel internals.
13 Q. Directing your attention to GALL, Revision 2, page XI 14 M16A-3, do you see item 3, titled "Parameters Monitored/
15 Inspected"?
16 A. Yes.
17 Q. Would you summarize that section?
18 A. Yes, this section provides recommendations for an 19 inspection plan for reactor pressure vessel internals, and 20 specifically what I would describe as the scope or focus of the 21 plan. This section is titled "Parameters Monitored/Inspected" 22 and states that the recommended inspection "program does not 23 directly monitor for loss of fracture toughness that is induced Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 26
1 by thermal aging or neutron embrittlement." Instead, it states 2 that the embrittlement of reactor pressure vessel internal 3 components is indirectly monitored through visual or volumetric 4 inspection techniques that look for cracking. It is important to 5 note that the focus of this document is on non-destructive 6 testing (NDT) and non-destructive evaluation (NDE) rather than 7 on the implications on core coolability subsequent to any 8 failures of highly degraded RPV internals.
9 10 MRP-227, Revision 0 11 Q. I show you a document marked as Exhibit NYS00307A-D.
12 Do you recognize it?
13 A. Yes, I have reviewed it. It is a copy of the document 14 prepared as a result of the nuclear industrys efforts 15 coordinated by the Electric Power Research Institute (EPRI).
16 Q. What is the title of that document?
17 A. The document's title is, "Material Reliability 18 Program: Pressurized Water Reactor Internals Inspection and 19 Evaluation Guidelines (MRP-227-Rev. 0), 1016596, Final Report, 20 December 2008." Unfortunately, as I discussed previously, it is 21 focused on NDT and NDE techniques rather than my aging-related 22 safety concerns.
23 Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 27
1 Entergy's License Renewal Application 2 Q. Directing you attention to Entergy's 2007 License 3 Renewal Application (LRA), did you find any indication in the 4 LRA that Entergy recognized that embrittlement could affect the 5 reactor pressure vessel?
6 A. Yes.
7 Q. Where was that?
8 A. The License Renewal Application at § 3.1.2.1.1 9 recognized that reactor pressure vessels are constructed of the 10 following materials.
11
- carbon steel 12
- Cladding , and, 14
- nickel alloy 15
- stainless steel 16 The same LRA section further recognized that reactor pressure 17 vessels experience the following aging effects that require 18 management:
19
- cracking 20
- loss of material 21
- reduction of fracture toughness, a term which 22 encompasses embrittlement.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 28
1 Q. Did you find any indication in the LRA that Entergy 2 recognized that embrittlement could affect reactor pressure 3 vessel internals?
4 A. Yes.
5 Q. Where was that?
6 A. The License Renewal Application at § 3.1.2.1.2 7 recognized that reactor pressure vessel internals are constructed 8 of the following materials:
9
- cast austenitic stainless steel (or CASS) 10
- nickel alloy 11
- stainless steel 12 The same LRA section further recognized that the reactor 13 pressure vessel internals experience the following aging effects 14 that require management:
15
- change in dimensions 16
- cracking 17
- loss of material 18
- loss of preload 19
- reduction of fracture toughness, a term which, as 20 noted previously, encompasses embrittlement.
21 22 23 Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 29
1 The 2007 LRA and the IP3 Reactor Pressure Vessel 2 Q. I direct your attention to License Renewal Application 3 Appendix A, § A.3.2.1.4. Do you have that?
4 A. Yes.
5 Q. What is that section of the License Renewal 6 Application concerned with?
7 A. That section concerns the IP-3 reactor pressure vessel 8 itself.
9 Q. And what did Entergy say there?
10 A. Entergy stated that a part of the IP-3 pressure 11 vessel, specifically plate B2803-3, exceeded the screening 12 criteria for pressurized thermal shock.
13 Q. Did Entergy acknowledge any specific concern about the 14 reactor pressure vessels at Indian Point?
15 A. Yes, Entergy acknowledged that with respect to IP-3 16 that the reactor pressure vessel plate B2803-3 "exceeds the 17 screening criterion by 9.9°F." That statement is in Entergy 18 January 22, 2008 Answer at 139; citing LRA § A.3.2.1.4.
19 Q. What if anything did Entergy propose to do about the 20 IP-3 pressure vessel?
21 A. Entergy proposed to submit to USNRC Staff a safety 22 analysis for plate B2803-3 three years before the plate reached Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 30
1 the reference temperature for pressurized thermal shock (RTPTS) 2 criterion.
3 4 The 2007 LRA and RPV Internals 5 Q. In your review of the April 2007 Indian Point License 6 Renewal Application, did you see an aging management program 7 (AMP) for reactor pressure vessel internals?
8 A. No, I did not. The 2007 License Renewal Application 9 did not contain an aging management program that specifically 10 focused on reactor pressure vessel internals. Rather, 11 Appendix A stated that in the future Entergy would develop an 12 aging management program for the reactor pressure vessel 13 internals of their plants. That is stated at LRA Appendix A, 14 § A.2.1.41 with respect to IP-2 and § A.3.1.41 with respect to 15 IP3. This deferred approach concerning IP-2 and IP-3 reactor 16 pressure vessel internals is also repeated at LRA, § 3.1.2.2.6.
17 Q. Do reactor pressure vessels and their associated 18 internal structures, components and fittings experience 19 embrittlement?
20 A. Yes.
21 Q. Are there any reactor pressure vessel internal 22 structures that are neglected in Entergys discussion of future 23 programs it will develop to address such structures.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 31
1 A. Yes. It should be noted that the control rods and 2 their associated guide tubes, plates, pins, and welds are not 3 highlighted, but they are also very important RPV internals and 4 their integrity is an extremely important safety concern. They 5 are located in the core region of the RPV, and are inserted into 6 the RPV through the upper head via so-called stub tubes. Their 7 function is to absorb excess fission neutrons (i.e., those not 8 need to achieve a chain reaction) so that the power level of a 9 reactor can be controlled.
10 Q. Do you believe there are any special problems 11 associated with providing an adequate aging management program 12 for control rods and their associated guide tubes, plates and 13 welds?
14 A. Yes. Because of geometric considerations, many PWRs 15 (including IP-2 and IP-3) cannot meet the USNRCs required 16 minimum coverage for the non-destructive testing (NDT) of the 17 so-called "J-groove" welds [Entergy, Walpole, NL-09-130 (Sept.
18 24, 2009)], and thus the integrity of these important CRD stub 19 tube welds cannot be directly confirmed by inspection. It 20 appears that to help address this chronic problem Entergy has 21 ordered two new RPV heads [Telecon-USNRC/Entergy Report, March 22 18, 2008 (Exh. NYS000317)], but they have not yet been scheduled 23 for installation at Indian Point [Telecon-USNRC/Entergy, March Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 32
1 18, 2008]. In any event, unlike the rather superficial 2 treatment given this important safety concern by Entergy [NL 3 063 (NYS000313)], I believe that a tangible, enforceable, and 4 viable aging management program (AMP) should be developed and 5 implemented before re-licensing the Indian Point reactor plants 6 for extended operations, since the integrity of these CRD welds 7 must be assured. If not, due to the leakage of borated primary 8 coolant through cracked welds, there can be aggressive corrosion 9 and wasting of the unclad outer surface of the upper head of the 10 RPVs (such as the serious event that occurred at Davis-Besse and 11 was identified in 2002). Worse yet, there might be an 12 inadvertent control rod ejection (due to a massive failure of 13 the welds in the upper RPV head), which could cause a 14 significant reactivity excursion, leading to core melting and 15 radiation releases.
16 Q. Are the places within the reactor pressure vessel that 17 you believe warrant particular aging management attention?
18 A. Yes. For the relicensing of the two reactors at 19 Indian Point, stress corrosion cracking (SCC) and radiation-20 induced embrittlement of the RPVs and their associated internals 21 is an important age-related safety concern, particularly in the 22 so-called "belt line" region of the RPV, which is the region 23 that is the closest to the reactor core. In addition, as noted Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 33
1 previously, the integrity of the so-called J-welds, which are 2 part of the control rod drive seal in the upper head of reactor 3 pressure vessels, is important to avoid corrosion-induced 4 failures of the upper head and the possibility of control rod 5 ejection (and thus uncontrolled reactivity excursion).
6 7 Entergy's NL-10-063 Communication 8 Q. I direct your attention to Exhibit NYS000313. Do you 9 recognize it?
10 A. Yes, I have reviewed this document. As noted above, 11 it contains a copy of a July 14, 2010 communication, NL-10-063, 12 from Entergy to the USNRC document control desk that concerns 13 embrittlement of reactor pressure vessel internals. In turn, 14 NL-10-063 contains an "Attachment 1."
15 Q. Does Entergy make any statements here about 16 embrittlement of reactor pressure vessel internals?
17 A. Yes. Entergy acknowledges that, PWR internals aging 18 degradation has been observed in European PWRs, specifically 19 with regard to cracking of baffle-former bolting. NL-10-063, 20 at pg. 89. Moreover, EPRI has stated that, considerable amount 21 of PWR internals aging degradation has been observed in European 22 PWRs. EPRI MRP-227, at A-4 (NYS00307A-D). Entergy also states:
23 As with other U.S. commercial PWR plants, cracking of baffle Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 34
1 former bolts is recognized as a potential issue for the [Indian 2 Point] units. NL-10-063, at pg. 89. Moreover, material 3 degradation has also been observed in control rod guide tube 4 alignment (split) pins. EPRI MRP-227, at A-4.
5 It is important to note that MRP-227 has also recommended 6 that analysis be done to show when it is acceptable to continue 7 to operate PWRs in which there have been bolt failures (e.g.,
8 due to embrittlement and/or fatigue ). While this type of 9 temporary, short-term fix might be adequate for normal 10 operations, it may lead to structural and component failures due 11 to the shock loads associated with various postulated accidents.
12 If so, the failed internal structures and components may 13 relocate, cause core blockages, or an otherwise uncoolable core 14 geometry, and thus lead to seriously degraded core cooling, core 15 melting and massive radiation releases.
16 Q. Do you have any additional problems with the 17 inspection program for RPVIs as proposed in the MRP-227 and 18 adopted by Entergy?
19 A. Yes. With respect to Entergys proposal to conduct 20 baseline examinations of the RPV internals, it should be noted 21 that I have previously called on Entergy to conduct such 22 examinations and for USNRC Staff to require the conduct of such 23 examinations before entering the period of extended operations Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 35
1 [See November 2007 Declaration of Richard T. Lahey, Jr., at ¶¶ 2 24, 25; see also State of New York Notice of Intention to 3 Participate and Petition to Intervene, at pgs. 217-220, State of 4 New York Contention-23 (Baseline Inspections)]. Fortunately, 5 both the USNRC and Entergy now seem to have embraced the concept 6 of baseline inspections for RPV internals, but the proposed 7 aging management program (AMP) as set forth in NL-10-063 lacks 8 sufficient details to know when the baseline inspections of the 9 RPV and its internals will begin and end, and the scope of these 10 inspections. Thus, it is not possible to know whether the 11 proposed baseline inspections will be comprehensive and 12 adequate.
13 Q. Are there other problems that you believe need to be 14 addressed if Entergy is to have an adequate aging management 15 program for RPV internals?
16 A. Yes. My Report provides more details on my concerns 17 with Entergys failure to conduct an evaluation of the 18 synergistic impacts of embrittlement, stress corrosion cracking, 19 and metal fatigue on the degradation of RPV internals and its 20 failure to consider how those interacting degradation mechanisms 21 will impact the ability of the RPV internals to withstand the 22 cumulative effect of thermal and decompression shock loads as a 23 result of a DBA LOCA. I am also concerned that the design of Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 36
1 the inspection programs, including their frequency, the type of 2 inspections to be conducted, the acceptance criteria and the 3 criteria for actions to be taken in the event of a failure of 4 acceptance of a component, does not consider these synergistic 5 degradation mechanisms. Finally, Entergys AMP for RPV 6 internals does not include specific programs with objective 7 criteria for either preventative measures or for corrective 8 actions to be taken when inspections show that certain 9 components are not acceptable.
10 11 Fatigue 12 Q. Let us now consider fatigue. Could you explain what 13 fatigue is?
14 A. Yes. Fatigue is another important age-related 15 degradation mechanism. It is one of the primary considerations 16 when conducting a time limited aging analysis (TLAA) and an 17 aging management program for nuclear power plants. Fatigue of 18 various structures, components and fittings in a nuclear reactor 19 can result in pipe ruptures, physical failures, and the 20 relocation of loose pieces of metal throughout the reactor 21 system, which, in turn, may result in core blockages and 22 interfere with the safe operation of a nuclear power plant. The 23 main concerns about fatigue are the increased potential for a Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 37
1 primary or secondary side LOCA, and the failure of various RPV 2 internals.
3 Q. In your expert opinion, has Entergy done adequate 4 fatigue evaluations to assure the safety of their two nuclear 5 power plants at the Indian Point site during extended 6 operations?
7 A. No. As I have testified to earlier in my Overview 8 remarks, they have focused on the fatigue of some selected 9 piping systems and components and, even then, have calculated 10 some near limiting values (e.g., CUFen = 0.9961 for the RHR line 11 of IP-3 ). No error analyses were presented to quantify the 12 modeling uncertainties and the effect of code user interactions 13 in WESTEMS, however, it appears that virtually any error would 14 put the calculated value of CUFen over the CUFen=1.0 fatigue 15 limit. In any event, Entergy has not done a systematic fatigue 16 evaluation of limiting RPV internals, considering the 17 synergistic effect of radiation-induced embrittlement and stress 18 corrosion cracking on fatigue-induced failures, and the possible 19 failures due to accident-induced pressure and/or thermal shock 20 loads. These latter safety evaluations are particularly 21 important in the evaluation of core coolability subsequent to 22 postulated accidents (i.e., the potential for core blockages due 23 to the relocation of failed RPV internals ).
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 38
1 Q. I show you your Supplemental Report, which has been 2 marked as Exhibit NYS000297. I note that the State has 3 provisionally designated as containing confidential information.
4 Would you provide a brief summary of the Report?
5 A. I prepared this Supplemental Report to set out some of 6 my concerns about the use of the WESTEMS computer code to 7 develop a cumulative fatigue analysis of certain components in 8 the Indian Point reactors and their reactor coolant pressure 9 boundaries.
10 Q. Would you briefly summarize your concerns?
11 A. Yes. First, I am concerned that without an error 12 analysis it is difficult to be in a position to meaningfully 13 analyze the results of the 2010 refined CUFen analysis presented 14 by Entergy and Westinghouse.
15 Q. Why is an error analysis important?
16 A. It is well known that all engineering analyses are 17 based on imperfect mathematical models of reality and various 18 code user assumptions which inherently involve some level of 19 error. Error analyses help readers and decision makers 20 understand what level of confidence to attach to the calculated 21 results and the proposed conclusions.
22 Q. Is the preparation of an error analysis an accepted 23 practice in the field of engineering?
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 39
1 A. Yes. Engineers frequently prepare error analyses.
2 In my submissions in this proceeding I noted that one would 3 normally expect to see a detailed propagation-of-error type of 4 analysis to determine the overall uncertainty in the CUFen 5 results given by Westinghouse. I referenced a text, Basic 6 Engineering Data Collection and Analysis, pp. 310-311, by 7 Vardeman & Jobe, to support this position, which is Exhibit 8 NYS000347.
9 Q. Are you aware of an instance where an error analysis 10 was prepared for a project at Indian Point.
11 A. Yes, for example in 1980, the Consolidated Edison 12 Company of New York prepared an error analysis in support of a 13 proposal to add more spent fuel into the spent fuel pool at 14 Indian Point Unit 2.
15 Q. I show you what has been marked as Exhibit NYS000348; 16 do you recognize it?
17 A. Yes. That is a copy of the 1980 Con Edison error 18 analysis for the re-racking of spent fuel in the Unit 2 spent 19 fuel pool.
20 Q. Do you have other concerns about the refined CUFen 21 reanalysis?
22 A. Yes, as discussed in my Supplemental Report, I am 23 concerned that engineering judgment or user intervention could Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 40
1 have affected the results. I note that when USNRC Staff issued 2 the Supplemental Safety Evaluation Report, Staff instructed 3 Entergy and Westinghouse, on a going forward basis, to document 4 and disclose the use of engineering judgment and user 5 intervention when conducting future fatigue analysis using the 6 WESTEMS code. That is set our in Exhibit NYS000160 at page 4-2.
7 Also, USNRC Staff instructed Entergy not to use WESTEMS when 8 conducting analyses under the ASME Standard know as NB-3600 (at 9 4-2, 4-3). Furthermore, I am concerned about the analytical 10 framework employed by the WESTEMS code. As detailed, in my 11 Supplemental Report, I believe that the code's thermal-hydraulic 12 models and framework are too simplified to predict accurate 13 results.
14 15 Entergy's NL-11-107 Communication 16 Q. I show you what has been marked as Exhibit NYS000314.
17 Do you recognize that?
18 A. Yes, this is a copy of Entergy's September 28, 2011 19 communication, NL-11-107, with the USNRC's document control 20 desk.
21 Q. Would you please turn to Table 5-2 at page 36 of the 22 Attachment to NL-11-107.
23 A. Yes, I have that.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 41
1 Q. What does the document say there?
2 A. In discussing the baffle-former assemblies and their 3 related baffle-edge bolts, it recognizes that irradiated-4 assisted stress corrosion cracking and fatigue can cause 5 cracking which, in turn, leads to failed or missing bolts 6 connecting a baffle to a former.
7 Q. What does communication NL-11-107 state?
8 A. In it, Entergy tells the USNRC that it has completed 9 commitment number 30 wherein Entergy stated that it would submit 10 an inspection plan to the USNRC for reactor pressure vessel 11 internals no later than two years before the plant entered the 12 period of extended operations. However, none of my safety 13 concerns associated with the synergistic effects of 14 embrittlement, fatigue and corrosion on the integrity of RPV 15 internals, and post-accident core coolability, were addressed.
16 In my opinion an adequate inspection plan for RPV internals is a 17 necessary , but not sufficient, means of assuring safe extended 18 plant operations. In addition, a systematic evaluation of the 19 degraded RPV internals is needed to identify the limiting 20 structures, components and fittings that need to be repaired or 21 replaced before the onset of extended operations.
22 23 Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 42
1 Conclusion 2 Q. You have reviewed the GALL Report Revision 1, the 3 Standard Review Plan Revision 1, the GALL Report Revision 2, the 4 Standard Review Plan Revision 2, EPRI's MRP-227 Revision 0, 5 Entergy's July 2010 NL-10-063 communication, NRC Staff's June 6 22, 2011 Safety Evaluation of MRP-227 Revision 0, NRC Staff's 7 August 30, 2011 Supplemental Safety Evaluation for the Indian 8 Point License Renewal Application, and Entergy's NL-11-107 9 communication, correct?
10 A. Yes.
11 Q. Do you have any opinion about those documents with 12 respect to the degradation of reactor pressure vessel internals?
13 A. Yes.
14 Q. Please summarize your testimony.
15 A. As I stated in my initial November 2007 declaration in 16 support of the State of New Yorks Contention 25 and in my 17 September 2010 declarations in support of the State's 18 supplemental filings, in my professional judgment Entergy has 19 failed to demonstrate that it had adequately accounted for the 20 aging phenomena of embrittlement for structures, components and 21 fittings inside the reactor pressure vessels at Indian Point 22 Unit 2 and Indian Point Unit 3. My professional judgment has 23 not fundamentally changed based upon Entergys July 14, 2010 Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 43
1 submission of License Renewal Application, Amendment No. 9 [NL-2 10-063] and Entergy's September 28, 2011 submission of NL 3 107. I do not believe the Entergys July 15, 2010 communication 4 to the Board [NL-10-063] concerning a new AMP for RPV internals 5 is adequate to address the safety concerns and technical issues 6 that I have raised herein. NL-11-107 likewise does not address 7 my age-related safety concerns, nor does it recognize the 8 importance of the various synergistic degradation mechanisms 9 that I am concerned with.
10 Q. Does this conclude your testimony?
11 A. Not quite. I want to stress that during the course of 12 my involvement in these relicensing proceedings I have 13 discovered what I believe to be some important new age-related 14 safety concerns which, to the best of my knowledge, have not 15 been previously considered in relicensing proceedings. In 16 particular, the synergistic effect on the degradation and 17 integrity of RPV internals of radiation-induced embrittlement, 18 corrosion and fatigue, and the potential for the failure of RPV 19 internals due to accident-induced pressure and/or shock loads, 20 and the implications of the failure of RPV internal structures, 21 components and fittings on post-accident core coolability.
22 While in the past many of these issues and concerns have been 23 noted separately, the implications of their interaction has Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 44
1 apparently been overlooked and not evaluated (i.e., they have 2 been evaluated in silos). I believe that these important age-3 related safety concerns must be resolved in order to have 4 assurance that the Indian Point reactors can operated safely 5 beyond their design life of 40 years. Indeed, I believe that 6 the most vulnerable RPV internals need to be identified and 7 repaired or replaced prior to extended operations since it is 8 beyond the current state-of-the-art to perform realistic and 9 accurate calculations on the relocation of failed RPV internals 10 and the resultant potential for core blockages and degraded core 11 cooling.
12 Q. Does this complete your testimony.
13 A. Yes, it does.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 45
1 UNITED STATES 2 NUCLEAR REGULATORY COMMISSION 3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4 -----------------------------------x 5 In re: Docket Nos. 50-247-LR; 50-286-LR 6 License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 7 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 8 Entergy Nuclear Indian Point 3, LLC, and 9 Entergy Nuclear Operations, Inc. December 22, 2011 10 -----------------------------------x 11 DECLARATION OF RICHARD T. LAHEY, JR.
12 I, Richard T. Lahey, Jr., do hereby declare under penalty of 13 perjury that my statements in the foregoing testimony and my 14 statement of professional qualifications are true and correct to 15 the best of my knowledge and belief.
16 Executed in Accord with 10 C.F.R. § 2.304(d) 17 18 19 ________________________
20 Dr. Richard T. Lahey, Jr.
21 The Edward E. Hood Professor Emeritus of Engineering 22 Rensselaer Polytechnic Institute, Troy, NY 23 (518) 495-3884, laheyr@rpi.ed Pre-filed Written Testimony of Richard T. Lahey, Jr.
Consolidated Contention NYS-26B/RK-TC-1B 46