ML12339A724
| ML12339A724 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/18/2012 |
| From: | Lahey R State of NY, Office of the Attorney General, Rensselaer Polytechnic Institute |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| References | |
| RAS 22624, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01 | |
| Download: ML12339A724 (35) | |
Text
United States Nuclear Regulatory Commission Official Hearing Exhibit Entergy Nuclear Operations, Inc.
In the Matter of:
(Indian Point Nuclear Generating Units 2 and 3)
ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #: NYS000374-00-BD01 Identified: 10/15/2012 NYS000374 Admitted: 10/15/2012 Withdrawn:
Rejected: Stricken: Submitted: June 19, 2012 Other:
1 UNITED STATES 2 NUCLEAR REGULATORY COMMISSION 3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4 -----------------------------------x 5 In re: Docket Nos. 50-247-LR; 50-286-LR 6 License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 7 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 8 Entergy Nuclear Indian Point 3, LLC, and 9 Entergy Nuclear Operations, Inc. June 18, 2012 10 -----------------------------------x 11 PRE-FILED WRITTEN TESTIMONY OF 12 DR. RICHARD T. LAHEY, JR.
13 REGARDING CONTENTION NYS-38/RK-TC-5 14 On behalf of the State of New York (NYS or the State),
15 the Office of the Attorney General hereby submits the following 16 testimony by RICHARD T. LAHEY, JR. regarding Contention NYS-17 38/RK-TC-5.
18 Q. Please state your full name.
19 A. Richard T. Lahey, Jr.
20 Q. By whom are you employed and what is your position?
21 A. I am the Edward E. Hood Professor Emeritus of 22 Engineering at Rensselaer Polytechnic Institute (RPI), which is 23 located in Troy, New York.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 1
1 Experience 2 Q. Please summarize your educational and professional 3 qualifications.
4 A. I have earned the following academic degrees: a B.S.
5 in Marine Engineering from the United States Merchant Marine 6 Academy, a M.S. in Mechanical Engineering from Rensselaer 7 Polytechnic Institute, a M.E. in Engineering Mechanics from 8 Columbia University, and a Ph.D. in Mechanical Engineering from 9 Stanford University. I have held various technical and 10 administrative positions in the nuclear industry, and I have 11 served as both the Dean of Engineering and the Chairman of the 12 Department of Nuclear Engineering & Science at RPI. Previously, 13 I was responsible for nuclear reactor safety R&D (research &
14 development) for the General Electric Company (GE), and I have 15 extensive experience with both military (i.e., naval) and 16 commercial nuclear reactors. Also, I am a member of a number of 17 professional societies and have served on various expert panels.
18 I was also an Editor of the international Journal of Nuclear 19 Engineering & Design, which focuses on nuclear engineering and 20 nuclear reactor safety technology. I am widely considered to be 21 an expert in matters relating to the design, operations, safety, 22 and aging of nuclear power plants.
23 Q. Which professional societies are you a member of?
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 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.
Contention NYS-38/RK-TC-5 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 14 Previous Submissions 15 Q. I show you what has been marked as Exhibit NYS000299 16 to Exhibit NYS000303. Do you recognize those documents?
17 A. Yes. They are copies of the six declarations that I 18 previously prepared for the State of New York in this 19 proceeding. They include the initial declaration that was 20 submitted in November 2007 in support of the State's petition to 21 intervene and its initial contentions, the April 7, 2008 22 declaration in support of Contention NYS-26A, the September 15, 23 2010 declaration submitted in support of the State's Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 4
1 supplemental bases for Contention 25, the September 9, 2010 2 declaration submitted in support of the amended Contention NYS-3 26B/RK-TC-1B, and the September 30, 2011 and November 1, 2011 4 declarations submitted in support of the present Contention, 5 Contention NYS-38/RK-TC-5.
6 Q. I show you what has been marked as Exhibit NYS000296.
7 Do you recognize that document?
8 A. Yes. It is a copy of the Report that I prepared for 9 the State of New York in this proceeding concerning Contentions 10 NYS-25 and NYS-26B/RK-TC-1B. The Report reflects my analysis 11 and opinions.
12 Q. I show you what has been marked as Exhibit NYS000297.
13 I note that the State has provisionally designated the exhibit 14 as containing confidential information. Do you recognize that 15 document?
16 A. Yes. This is a copy of a Supplemental Report that I 17 previously prepared for the State of New York in this proceeding 18 that addresses aspects of the revised fatigue analysis that 19 Entergy and Westinghouse prepared concerning certain components 20 in the Indian Point reactors and their related systems and 21 boundaries. The Supplemental Report sets out some of my 22 concerns about the use of the WESTEMS computer code by Entergy 23 and Westinghouse to develop a cumulative fatigue analysis of Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 5
1 certain reactor components. The Supplemental Report also 2 reflects my analysis and opinions.
3 Q. I show you a copy of what has been marked as Exhibit 4 NYS000344. Do you recognize that document?
5 A. Yes, this is a copy of the pre-filed testimony that I 6 previously submitted in December 2011 concerning Contention NYS-7 26B/RK-TC1B.
8 9 Overview 10 Q. What is the purpose of your testimony?
11 A. I was retained by the State of New York State to 12 review Entergy's application to the U.S. Nuclear Regulatory 13 Commission (USNRC) and its Staff for two renewed operating 14 licenses for the nuclear power plants known as Indian Point Unit 15 2 and Unit 3. I have reviewed the License Renewal Applications 16 (LRAs) and subsequent filings by Entergy and the USNRC Staff.
17 My declarations and report discuss my concerns and opinions 18 about issuing twenty-year operating licenses for these 19 facilities. My testimony seeks to identify and discuss some 20 age-related safety concerns which have not yet been addressed by 21 Entergy. In my opinion these concerns must be resolved to 22 assure the health and safety of the American public, 23 particularly those in the vicinity of the Indian Point reactors.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 6
1 The purpose of my testimony today is to provide support 2 for, and my views on, certain aspects of New Yorks and 3 Riverkeepers Joint Contention NYS-38/RK-TC-5 (NYS-38/RK-TC-4 5), which was admitted for litigation by the Atomic Safety 5 Licensing Board. Contention NYS-38/RK-TC-5 asserts, among other 6 things, that Entergy has not demonstrated that it has a program 7 that will manage the effects of aging of critical components or 8 systems at the Indian Point nuclear power facilities and that 9 therefore the USNRC does not have a record and a rational basis 10 upon which it can determine whether to grant Entergy a renewed 11 license for the Indian Point facilities.
12 My testimony critiques Entergys proposed approach towards 13 the age related degradation of various components in Indian 14 Point's steam generators during the requested twenty year period 15 of extended operation.
16 My testimony also critiques Entergy's proposed approach 17 that defers important aspects of a program that ostensibly seeks 18 to address the age related degradation caused by metal fatigue.
19 It is also my understanding that the Board has deferred 20 presentation of testimony on another aspect of this contention, 21 namely the ongoing iterative regulatory dialogue between Entergy 22 and USNRC Staff concerning important aspects of Entergy's Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 7
1 approach to address age related degradation caused by 2 embrittlement.
3 Q. Have you reviewed various materials in preparation for 4 your testimony?
5 A. Yes.
6 Q. What is the source of those materials?
7 A. I have reviewed documents prepared by government 8 agencies, Entergy, Westinghouse, the utility industry, or its 9 associations, and various related text books and peer-reviewed 10 articles.
11 Q. I show you Exhibits NYS00146A-C, NYS00147A-D, 12 NYS000150 through NYS000154, NYS000160, NYS000161, NYS000195, 13 NYS000304 through NYS000369 and NYS000375 through NYS000395. Do 14 you recognize these documents?
15 A. Yes. These are true and accurate copies of some of 16 the documents that I referred to, used, or relied upon in 17 preparing my report, declarations, and this testimony. In some 18 cases, where the document was extremely long and only a small 19 portion is relevant to my testimony, an excerpt of the document 20 is provided. If it is only an excerpt, that is noted on the 21 first page of the Exhibit.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 8
1 Q. How do these documents relate to the work that you did 2 as an expert in forming opinions such as those contained in this 3 testimony?
4 A. These documents represent the type of information that 5 persons within my field of expertise reasonably rely upon in 6 forming opinions of the type offered in this testimony.
7 8 Conclusions and Opinions 9 Q. Dr. Lahey I show you what has been marked as Exhibit 10 NYS000160. Do you recognize it?
11 A. Yes, this is a copy of the USNRC Staff's August 2011 12 Supplemental Safety Evaluation Report (SSER) for the requested 13 renewal of the operating licenses for the Indian Point reactors 14 [NUREG-1930, Supp. 1].
15 Q. Did you review the Supplemental Safety Evaluation 16 Report?
17 A. Yes.
18 Q. I also show you Exhibits NYS000151, NYS000152, 19 NYS00153, and NYS00154. Do you recognize them?
20 A. Yes, these are 2011 communications from Entergy to the 21 USNRC Staff in response to Staff questions about the age related 22 degradation of various components at Indian Point Unit 2 and Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 9
1 Indian Point Unit 3. These are Entergy communications NL 2 032, NL-11-074, NL-11-90, and NL-11-096.
3 Q. Did you reach any conclusions based on that review?
4 A. I have reviewed the USNRC Staff's Supplemental Safety 5 Evaluation Report for Indian Point Unit 2 and Unit 3. The SSER 6 makes it clear that a number of important details and questions 7 remain unresolved concerning the aging-induced degradation of 8 various safety-related systems and components and the management 9 of that process. Unfortunately, there are virtually no details 10 given on the future analyses and/or inspections that Entergy 11 will apparently do. The absence of such details makes it 12 difficult, if not impossible, to meaningfully evaluate the 13 approach or program that Entergy proposes. In any event, the 14 dates given for Entergy and the USNRC's anticipated resolution 15 of these issues appear to be beyond the time frame for 16 submission of testimony and the evidentiary hearings in this 17 ASLB proceeding and thus will not allow for a testing of the 18 adequacy of the proposed resolution of these issues in this 19 proceeding. That timeline will also prevent the State of New 20 York from playing any meaningful role in their development or 21 resolution.
22 The details of the inspections for primary water stress 23 corrosion cracking (PWSCC) in the steam generators divider Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 10
1 plates will apparently not be available until after extended 2 operations are expected to begin.
3 Inspections of the steam generators tube-to-tubesheet 4 welds in Indian Point Unit 2 for PWSCC will not be made until 5 sometime between March 2020 and March 2024, which is well after 6 the proposed extended operation period has begun. This is 7 particularly troubling since these welds form part of the 8 primary systems pressure boundary, and if they fail radiation 9 may be released to the secondary side and also to the 10 environment.
11 Inspections of the steam generators tube-to-tubesheet 12 welds in Indian Point Unit 3 for PWSCC will not be made until 13 the first refueling outage after the reactor enters the period 14 of extended operation. Indian Point Unit 3 could enter its 15 period of extended operation in late December 2015. Based on 16 the current refueling schedules, which have Indian Point Unit 3 17 refueling in March of odd numbered years, I anticipate that the 18 first refueling outage for Indian Point Unit 3 after it enters 19 the period of extended operation in December 2015 would be in or 20 around March 2017.
21 In addition, although USNRC Staff has required Entergy and 22 Westinghouse to disclose the parameters surrounding code user 23 interventions in future applications and runs of the WESTEMS Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 11
1 code for Indian Point, Entergy has not disclosed the parameters 2 surrounding user intervention in the previous runs of WESTEMS 3 that provided the basis for the what has been described as the 4 refined metal fatigue analysis that was previously submitted in 5 this proceeding. The absence of this information impedes and 6 prevents a meaningful analysis of the metal fatigue analysis 7 that Entergy has presented here and the aging management program 8 that Entergy has proposed.
9 Furthermore, the State has previously raised concerns about 10 the locations within the reactor coolant pressure boundary that 11 are examined with respect to fatigue. In 2011 Entergy indicated 12 that it will undertake further steps to identify additional 13 limiting locations that are subject to fatigue in the reactor 14 coolant pressure boundary. Based on a May 15, 2012 Entergy 15 letter to the Board, it appears that Entergy and Westinghouse 16 will not complete the initial steps of this analysis until mid 17 September 2012. It is my understanding that Entergy has not yet 18 disclosed the results of any such review. The absence of this 19 information also impedes and prevents an meaningful analysis of 20 the metal fatigue analysis that Entergy has presented here and 21 the aging management program that Entergy has proposed.
22 23 Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 12
1 The Indian Point Reactors 2 Q. Are you familiar with the power reactors that are the 3 subject of this proceeding?
4 A. Yes.
5 Q. Would you briefly describe them?
6 A. Entergy operates two power reactors that are located 7 in northern Westchester County near the Village of Buchanan.
8 The operating reactors are known as the Indian Point Unit 2 and 9 Indian Point Unit 3 power reactors. These Westinghouse-designed 10 plants are 4-loop pressurized water reactors (PWRs), and they 11 are currently rated at power levels of 3,216.4 MWt. Entergy 12 also owns another reactor at the same site. That reactor is 13 known as the Indian Point Unit 1 reactor; however, that reactor 14 has been shut down and no longer produces power.
15 16 Operation of a Pressurized Water Reactor 17 Q. Would you briefly describe the design and operation of 18 a pressurized water reactor?
19 A. Pressurized water nuclear reactors have water (i.e.,
20 the primary coolant) under high pressure flowing through the 21 core in which heat is generated by the fission process. The 22 core is located inside a reactor pressure vessel (RPV). This 23 heat is absorbed by the coolant and then transferred from the Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 13
1 coolant in the primary system to lower pressure water in the 2 secondary system via a large heat exchanger (i.e., a steam 3 generator) which, in turn, produces steam on the secondary side.
4 These steam generator systems, which are part of the plants 5 Nuclear Steam Supply System (NSSS), are located inside a large 6 containment structure. After leaving the containment building, 7 via main steam piping, the steam drives a turbine, which turns a 8 generator to produce electrical power.
9 As I mentioned, Indian Point Unit 2 and Indian Point Unit 3 10 each have a four loop Nuclear Steam Supply System or NSSS. Each 11 loop contains, among other components, a pressurizer, a steam 12 generator, and a reactor coolant pump. Thus, the two operating 13 Indian Point reactors collectively have eight steam generators.
14 The reactor pressure vessel is a large steel container that 15 holds the core (i.e., the nuclear fuel); it also serves as a key 16 part of the primary coolant's pressure boundary.
17 As its name (PWR) suggests, this reactor design uses a 18 pressurizer on the primary side that performs several functions.
19 In particular, it maintains the operating pressure on the 20 primary side of the nuclear reactor and accommodates variations 21 in reactor coolant volume for load changes during reactor 22 operations, as well as reactor heat-up and cool-down. The 23 reactor coolant also moderates the neutrons produced in the core Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 14
1 since a pressurized water nuclear reactor will not function 2 unless the neutrons are moderated (i.e., slowed down due to 3 collisions with the hydrogen molecules in the primary coolant).
4 Q. I show you what has been marked as Exhibit NYS000304.
5 Do you recognize it?
6 A. Yes. It is a schematic diagram from a USNRC document 7 that identifies the relative location of various components in a 8 pressurized water nuclear reactor type of power plant including, 9 from the inside to the outside, the: reactor core, reactor 10 pressure vessel, pressurizer, steam generator, containment 11 structure, turbine, and associated piping. The diagram also 12 identifies various materials that are used or contained in those 13 components.
14 Q. I show you what has been marked as Exhibit NYS000375.
15 Do you recognize it?
16 A. Yes. It is a schematic diagram of a Westinghouse 17 Nuclear Steam Supply System. Among other things, this diagram 18 depicts the reactor coolant pressure boundary. The components 19 on the primary side appear in red or orange, and the components 20 on the secondary side are in blue or green.
21 Q. What is the reactor coolant pressure boundary?
22 A. The USNRC provides a definition of the reactor coolant 23 pressure boundary in its regulations 10 C.F.R. § 50.2. In Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 15
1 essence, the reactor coolant pressure boundary refers to a 2 physical barrier or boundary between the reactor coolant system 3 in the "primary loop" of nuclear steam supply system and the 4 "secondary loop" of the nuclear steam supply system. As I 5 noted, one can see this boundary line in Westinghouse diagram 6 (Exhibit NYS000375) that represents the primary loop in red or 7 yellow and the secondary loop in green or blue. It is critical 8 not to breach the reactor coolant pressure boundary and allow 9 the reactor coolant to escape.
10 11 Steam Generators, Their Components, and 12 Primary Water Stress Corrosion Cracking 13 Q. I show you what has been marked as Exhibit NYS000376.
14 Do you recognize it?
15 A. Yes, it is a diagram of a Westinghouse steam 16 generator.
17 Q. Would you please describe the role of the steam 18 generators in the Indian Point nuclear steam supply systems?
19 A. Each reactor coolant loop contains a vertical shell 20 and U-tube steam generator. Reactor coolant enters the inlet 21 side of the channel head at the bottom of the steam generator 22 through the inlet nozzle, is forced upward through the 23 tubesheet, flows through the U-tubes, returns through the Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 16
1 tubesheet to an outlet plenum and leaves the generator through a 2 bottom nozzle. The inlet and outlet plena in the steam 3 generator are separated by a partition or divider plate. The 4 divider plate is joined to the plenum head and the tubesheet 5 through a stub runner.
6 Q. Did Entergy's License Renewal Application discuss the 7 intended function of the steam generators' components?
8 A. Yes, in the License Renewal Application Tables 2.3.1-9 4-IP2/IP3 of the LRA Entergy states that the channel head, the 10 divider plate, tubes, and the tubesheet each constitutes a 11 pressure boundary for Indian Point Unit 2 and Indian Point Unit 12 3. Entergy acknowledged that the tubes also perform a heat 13 transfer function. Those tables are located in the License 14 Renewal Application at pages 2.3-36, 2.3-39, respectively.
15 Q. Are you familiar with the term "primary water stress 16 corrosion cracking"?
17 A. Yes. Primary water stress corrosion cracking is a 18 recognized aging phenomenon for many alloy/environmental 19 combinations. It presents a challenge since it occurs in 20 otherwise ductile alloys, but only in very specific 21 environments. Occurrence of the phenomenon requires the 22 simultaneous presence of stress, whether residual or applied, 23 and a specific alloy /environment combination.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 17
1 Q. Has primary water stress corrosion cracking occurred 2 in the nuclear energy production area?
3 A. Yes. In operating nuclear reactors, cracking of 4 stressed nickel based alloys, such as Alloy 600, has occurred.
5 It should also be noted that Alloy 600 components are generally 6 welded with Alloys 82 or 182, which are derivatives of Alloy 600 7 that have also been found to be susceptible to primary water 8 stress corrosion cracking.
9 Q. Has Entergy disclosed the material present in the 10 divider plates in the steam generators at Indian Point?
11 A. Yes, in 2011 in response to a request for information 12 by USNRC Staff, Entergy stated that the current Indian Point 13 Unit 2 steam generators use Alloy 600 for the divider plates and 14 that it assumed that the weld material for the divider plate 15 assemblies was Alloy 82/182 weld material. Entergy also stated 16 that the Indian Point Unit 3 steam generators use Alloy 600 for 17 the divider plates and that it assumed that the weld material 18 for the divider plate assemblies was Alloy 82/182 weld material.
19 Q. I show you Exhibit NYS000151; would you describe the 20 document?
21 A. Yes, this is a copy of NL-11-032, which was Entergy's 22 March 28, 2011 initial response to the USNRC Staff's request for 23 additional information. In this document starting on page 20 of Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 18
1 Attachment 1, Entergy discusses, among other things, the 2 material composition of the steam generator divider plates and 3 associated welds.
4 Q. Has Entergy disclosed the composition of the heat 5 transfer tubes in Indian Point's steam generators?
6 A. Yes, in the License Renewal Application at page 2.3-7 21, Entergy stated that the current Indian Point Unit 2 steam 8 generators use Alloy 600 for the heat transfer tubes and that 9 the current Indian Point Unit 3 steam generators use Alloy 690 10 for their tubes.
11 Q. What types of steam generators parts or locations are 12 affected by primary water stress corrosion cracking?
13 A. In addition to the heat transfer tubes, primary water 14 stress corrosion cracking could also affect other components or 15 assemblies that use Alloy 600 or welds that use Alloy 82/182 16 weld material that, as I noted, are derivatives of Alloy 600.
17 In the August 30, 2011 Supplemental Safety Evaluation Report at 18 page 3-21, the USNRC Staff has also expressed concern about the 19 propagation of primary water stress corrosion cracking in 20 tubesheets that have Alloy 600 cladding or related weld even 21 when the heat transfer tubes are made from Alloy 690TT material.
22 According to Staff, "a crack initiated in this region, close to 23 the tube, may propagate into or through the weld, causing a Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 19
1 failure of the weld and of the reactor coolant pressure 2 boundary." These areas of concern would include the channel 3 head-to-tubesheet-to-tube complex, including the divider plate 4 assembly and the tube-to-tubesheet welds.
5 Q. In your opinion, would primary water stress corrosion 6 cracking of the divider plates, weld, or channel head assemblies 7 impact the intended function of the steam generators?
8 A. Yes, in my opinion it would.
9 First, primary water stress corrosion cracking of a divider 10 plate or its weld could compromise the ability of the divider 11 plate to direct fluid through the tubesheet into the heat 12 transfer tubes and hence impede one of the intended functions of 13 the tubes and the steam generator to transfer heat and thus to 14 provide a heat sink for the heat generated in the core. I would 15 consider the loss of that intended function to be a significant 16 safety concern since shock-load-induced failures of the divider 17 plate have apparently not been analyzed (e.g., the 18 thermal/pressure shock loads experience during various 19 postulated LOCA events), but such events may lead to gross 20 failures of cracked divider plates.
21 Second, the USNRC Staff recognized that a primary water 22 stress corrosion crack in the divider plate could propagate into 23 a tubesheet and a tube-to-tubesheet weld. Such a crack in the Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 20
1 lower steam generator assembly area could compromise another 2 intended function of the steam generator, namely the maintenance 3 of the reactor coolant pressure boundary between the primary 4 loop and the secondary loop in the nuclear steam supply system.
5 Q. Do you have any opinion about the sufficiency of the 6 approach that Entergy has proposed here regarding primary water 7 stress corrosion cracking of steam generator components at the 8 Indian Point facilities?
9 A. As I stated last Fall, Entergy's proposal leaves many 10 important details and questions unresolved. As the Supplemental 11 Safety Evaluation Report confirms (at p. 3-19), Entergy proposes 12 that it will perform "an inspection of steam generators for both 13 units to assess the condition of the divider plate assembly."
14 However, this proposal does not describe the inspection 15 methodology nor the number of steam generators to be inspected.
16 Indeed, it is quite vague and does not provide details.
17 Similarly, it does not describe the acceptance criteria for such 18 inspection or the corrective action criteria for divider plates 19 that fail the inspection.
20 Turning to the issue of cracks spreading from tubesheet 21 cladding to tube-to-tubesheet welds, Entergy again proposes an 22 approach that is short on details. Specifically, Entergy 23 proposes to "develop a plan" that will use one of two options:
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 21
1 (1) "perform an analytical evaluation" to establish that 2 tubesheet cladding and welds are not susceptible to primary 3 water stress corrosion cracking, or redefine the reactor coolant 4 pressure boundary; or (2) perform a one time inspection of a 5 representative number of tube-to-tubesheet welds in each steam 6 generator to determine if primary water stress corrosion 7 cracking is present. This plan that Entergy has proposed to 8 develop leaves many questions unanswered, including: the basis 9 for the analytical analysis, how Entergy can change the 10 definition of the reactor coolant pressure boundary after these 11 facilities have relied on that definition for many years, and, 12 the methodology of the alternative one-time inspection.
13 In each regard, Entergy has not presented an aging 14 management program, but rather has presented a vague, conceptual 15 approach.
16 17 WESTEMS and "User Intervention" 18 Q. Turning to the issue of fatigue, could you explain 19 what fatigue is?
20 A. Yes. Fatigue is another important age-related 21 degradation mechanism. It is one of the primary considerations 22 when conducting a time limited aging analysis (TLAA) and an 23 aging management program (AMP) for nuclear power plants.
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 22
1 Fatigue of various structures, components and fittings in a 2 nuclear reactor can result in pipe ruptures, physical failures, 3 and the relocation of loose pieces of metal throughout the 4 reactor system, which, in turn, may result in core blockages and 5 interfere with the safe operation of a nuclear power plant. The 6 main concerns about fatigue are the increased potential for a 7 primary or secondary side LOCA, and the failure of various RPV 8 internals.
9 Q. I would like to direct your attention to NYS000160, 10 NYS000151, NYS000152, NYS00153, and NYS00154. Do you have those 11 documents?
12 A. Yes, I have those documents.
13 Q. Would you identify them for the record?
14 A. Yes. These documents include the August 2011 USNRC 15 Staff Supplemental Safety Evaluation Report (NUREG-1930, Supp.
16 1) and Entergy's 2011 responses to the Staff's request for 17 additional information, including Entergy communications NL 18 032, NL-11-074, NL-11-90, and NL-11-096.
19 Q. Do these documents discuss the use of the WESTEMS 20 computer code?
21 A. Yes, they discuss Entergy's and Westinghouse's use of 22 WESTEMS at Indian Point as part of the license renewal process.
23 Q. Do they also discuss the term "user intervention"?
Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 23
1 A. They do.
2 Q. Would you describe the term user intervention?
3 A. Yes. The term "user intervention" refers to, among 4 other things, the use of assumptions and engineering judgment in 5 the process of calculating the CUFen values using codes such as 6 WESTEMS.
7 Q. Did you review NYS000160, NYS000151, NYS000152, 8 NYS00153, and NYS00154 with respect to the use of WESTEMS and 9 user intervention in the Indian Point license renewal process 10 A. Yes.
11 Q. And did you draw any conclusions from that review?
12 A. Yes.
13 Q. Would you describe those conclusions?
14 A. Certainly.
15 Entergy relies on WESTEMS, a proprietary computer program 16 developed by Westinghouse, as an essential part of Entergys 17 CUFen analysis for the Indian Point facilities. Entergy agreed 18 with the USNRC that the piping system stress model (NB-3600) in 19 WESTEMS will not be used until the USNRC staff resolves some 20 issues concerning its validity. Rather, a finite element method 21 (FEM) design by analysis approach (NB-3200) will be used 22 instead. Unfortunately, this finite element method-based 23 computational approach requires numerous assumptions concerning Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 24
1 the stress-inducing thermal transients and the loads/moments 2 from the piping system; such assumptions must be developed and 3 applied by the WESTEMS code user to the component being 4 analyzed. These assumptions could materially affect the results 5 raising questions concerning their reliability and validity.
6 Thus, it is necessary to have disclosed in advance the 7 assumptions to be used in the analysis and the basis for using 8 those assumptions in order to ascertain whether the approach 9 being proposed will meet the required safety standards for an 10 adequate AMP. Given the role that such user-developed 11 assumptions play in the process, and the fact that Entergy has 12 apparently not done an error analysis of the WESTEMS results, 13 there remain important questions concerning the reliability and 14 validity of these results.
15 In its August 2011 Supplemental Safety Evaluation (at 4-2),
16 USNRC Staff has required Entergy to create records that document 17 and justify any assumptions and engineering judgments developed 18 and used in the CUFen calculations. Such assumptions and 19 engineering judgments affect the WESTEMS results. A systematic 20 and methodical explanation of these assumptions and engineering 21 judgments is essential in evaluating the adequacy of the CUFen 22 calculations done for IP-2 & IP-3 using WESTEMS.
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1 Entergy has agreed that any user intervention in future 2 WESTEMS evaluations will be explained and justified.
3 Unfortunately, nothing was said about the previous WESTEMS 4 evaluations that were done for IP-2 & IP-3 and the affect that 5 user interventions had on those CUFen results (for which no 6 error analysis has been given). Moreover, as the Supplemental 7 Safety Evaluation Report makes clear (at 4-2), the documentation 8 of any new user interventions will not be disclosed or 9 implemented until close to the end of the current licensing 10 terms (i.e., September, 2013 and December, 2015). In addition, 11 Entergy has not disclosed the specific criteria it will use in 12 deciding whether to make a user intervention and what standards 13 will control the extent of these interventions. Thus, the State 14 of New York is being effectively excluded from reviewing this 15 important process.
16 Q. Dr. Lahey, I direct your attention to Exhibits 17 NYS000296 and NYS000297, which are your December 2011 reports.
18 A. Yes, I have those documents.
19 Q. Turning to Exhibit NYS000296 would you describe your 20 observations and conclusions about WESTEMS and user 21 interventions and assumptions?
22 A. As I discussed in this report, for example at 23 paragraphs 35 and 36, it is apparent that the WESTEMS code is Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 based on rather simple models (particularly the thermal-2 hydraulic models) and that the thermal stress results for CUFen 3 are strongly influenced by the code users assumptions, 4 manipulations and interventions. In fact, there is a lot of 5 engineering judgment implicit in the CUFen results, and, since 6 an error analysis has not been done to bound the uncertainty, 7 and many results are disturbingly close to the CUFen = 1.0 limit, 8 I do not believe that one can trust these results to assure the 9 safety of the IP-2 and IP-3 during extended plant operations.
10 Indeed, these results are quite uncertain and this uncertainty 11 should be quantified by doing parametric runs and/or a detailed 12 error analysis. Moreover, because the effect of various shock 13 loads on the failure of these fatigue-weakened components, 14 structures, and fittings has not been considered, it is unclear 15 that the health and safety of the American public is being 16 adequately protected. As previously noted, these results are 17 quite uncertain and this uncertainty needs to be quantified by 18 doing a detailed error analysis.
19 Q. Now, I would like to turn to Exhibit NYS000297, which 20 is your Supplemental Report. I note that this exhibit has been 21 provisionally designated as potentially containing proprietary 22 information. Would you briefly describe your conclusions 23 contained in this report?
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1 A. Yes. The Supplemental Report sets out some of my 2 concerns about the use of the WESTEMS computer code by Entergy 3 and Westinghouse to develop a cumulative fatigue analysis of 4 certain components in the reactors and their reactor coolant 5 pressure boundaries. Among other things, the report describes 6 my concerns about the use of engineering judgment, assumptions, 7 and user interventions as part of the WESTEMS analysis.
8 Q. Do you have any additional conclusions or observations 9 concerning the use of WESTEMS and user intervention in the 10 Indian Point license renewal application process?
11 A. There is a difference between stating that one will 12 develop a program that will comply with the parameters in GALL 13 and actually disclosing the details, judgments, assumptions, and 14 user interventions that underlay the program and the analyses, 15 including computer codes such as WESTEMS, that are critical to 16 the program. Only through the latter can one test an 17 applicant's claim that its proposed program is consistent with 18 GALL. In fact, it is not possible to demonstrate that and aging 19 management program is effective or consistent with GALL unless 20 the details, judgments, assumptions, and user interventions have 21 been disclosed. This is particularly important since neither 22 Westinghouse nor Entergy appear willing to perform an error Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 analysis of WESTEMS results to justify any claims of 2 compliance.
3 4 Identification of Limiting Locations for Fatigue Analysis 5 Q. Earlier in your overview of your testimony you stated 6 that you had concerns about the locations selected for metal 7 fatigue analysis. Could you expand on that?
8 A. Yes. For CUFen fatigue calculations it is important to 9 fully understand the assumptions that will be made and the 10 criteria that will be used in determining which locations will 11 produce the most limiting conditions.
12 It is my understanding that as a result of additional 13 review by USNRC Staff of the Indian Point license renewal 14 application, the USNRC Staff has raised concerns that Entergy 15 may not have chosen the sites of the most limiting fatigue 16 conditions and Entergy has agreed to reanalyze the locations it 17 has previously identified and to determine if more limiting 18 conditions exist at other sites. If so, detailed further 19 analysis will be required.
20 Unfortunately, the exact time for reporting the results of 21 this future review/analysis was not specified, but it will 22 apparently be shortly before extended operations are expected to 23 begin. Postponing the disclosure of the details of that Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 29
1 review/analysis until Indian Point Unit 2 is on the cusp of 2 extended operation will prevent those matters from being tested 3 and resolved in these ASLB hearings and greatly handicaps, if 4 not precludes, the State of New York from any meaningful role in 5 their development and resolution. Moreover, the assumptions to 6 be used and the criteria to be applied for these future reviews, 7 and whether they were properly designed to identify limiting 8 locations and the conditions of such locations, are left for 9 consideration at a later day -- apparently by the USNRC and 10 Entergy, but not the State and other interested parties.
11 Moreover, it also appears that, as before, this review will 12 focus on structures, components and fittings outside the RPV and 13 will thus not include a comprehensive consideration of the 14 fatigue of important RPV internal structures, components and 15 fittings.
16 Q. I show you Exhibit NYS000355. Do you recognize that 17 document?
18 A. Yes. This is a report entitled NUREG/CR-6260 19 "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected 20 Nuclear Power Plant Components." This document identifies 21 generic locations to examine for metal fatigue; however, 22 additional location may be required to be examined.
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1 Q. I show you Exhibit NYS000160, which is the August 2011 2 Supplemental Safety Evaluation Report and direct your attention 3 to page 4-1 and 4-2.
4 A. Yes, I have those pages.
5 Q. Would you describe your understanding of what the 6 USNRC Staff has required?
7 A. As a result of additional review of the Indian Point 8 license renewal application, USNRC Staff raised a question about 9 whether the locations included in the metal fatigue analysis for 10 the Indian Point facilities were the most limiting locations for 11 the facilities.
12 Q. Do you know how Entergy responded?
13 A. Based on the Supplemental Safety Evaluation Report at 14 page 4-1, it is my understanding that Entergy proposed to review 15 the fatigue evaluations for ASME Code Class 1 components to 16 determine whether the NUREG/CR-6260 locations are the limiting 17 locations for Indian Point Unit 2 and Indian Point Unit 3.
18 Further, according to the Supplement safety Evaluation Report, 19 Entergy agreed that if more limiting locations are identified, 20 Entergy will evaluate the most limiting location for the effects 21 of reactor coolant environment on the fatigue usage.
22 However, based on a May 15, 2012 Entergy letter to the 23 Atomic Safety and Licensing Board (ASLB), it appears that Pre-filed Written Testimony of Richard T. Lahey, Jr.
Contention NYS-38/RK-TC-5 31
1 Entergy and Westinghouse will not complete the initial steps of 2 this analysis until mid-September 2012 (Exhibit NYS000395). It 3 is my understanding that Entergy has not yet disclosed the 4 results of any such review.
5 Q. Do you have any conclusions about this process and 6 time lime?
7 A. The schedule for this process and the absence of this 8 information to date impedes a meaningful evaluation of the metal 9 fatigue analysis that Entergy has presented here and the aging 10 management program that Entergy has proposed. Moreover, the 11 apparent unwillingness of Entergy to provide an error analysis 12 or user intervention information for their CUFen results makes it 13 impossible to assess the validity of their conclusions with 14 respect to limiting locations.
15 16 Conclusion 17 Q. Dr. Lahey, have you reviewed the USNRCs Staff Safety 18 Evaluation Report (SER) and Supplemental Safety Evaluation 19 Report (SSER) as it relates to the issue metal fatigue and 20 Entergys proposal for how to address that issue?
21 A. Yes.
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1 Q. Does it provide an adequate basis for acceptance of 2 Entergys claim that it has demonstrated that it has a 3 sufficient AMP to address the metal fatigue issue?
4 A. No. In my opinion it does not.
5 Q. What is the basis for your statement?
6 A. The SER and SSER basically accept, as valid, what 7 Entergy has proposed to address the metal fatigue issue.
8 Therefore, all the criticisms I have identified in this 9 testimony and in my prior testimony related to metal fatigue 10 apply with equal force to the SER and SSER. In addition, 11 because the USNRC Staff has apparently ignored the importance of 12 shock-load-induced failures and the synergistic effects of 13 embrittlement and metal fatigue on the integrity of reactor 14 pressure vessel internals, the credibility and technical 15 validity of the USNRC Staffs position is severely compromised.
16 I have discussed this matter in detail in my previous testimony 17 with respect to these two related degradation phenomena.
18 Q. In your expert opinion, has Entergy done adequate 19 fatigue evaluations to assure the safety of their two nuclear 20 power plants at the Indian Point site during extended 21 operations?
22 A. No, they have not.
23 Q. Does this conclude your testimony?
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1 A. Yes, it does.
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10 11 12 13 14 15 16 17 18 19 20 21 22 23 Pre-filed Written Testimony of Richard T. Lahey, Jr.
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1 UNITED STATES 2 NUCLEAR REGULATORY COMMISSION 3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4 -----------------------------------x 5 In re: Docket Nos. 50-247-LR; 50-286-LR 6 License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 7 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 8 Entergy Nuclear Indian Point 3, LLC, and 9 Entergy Nuclear Operations, Inc. June 18, 2012 10 -----------------------------------x 11 DECLARATION OF RICHARD T. LAHEY, JR.
12 I, Richard T. Lahey, Jr., do hereby declare under penalty 13 of perjury that my statements in the foregoing testimony and my 14 statement of professional qualifications are true and correct to 15 the best of my knowledge and belief.
16 Executed in Accord with 10 C.F.R. § 2.304(d) 17 18 19 ________________________
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.edu Pre-filed Written Testimony of Richard T. Lahey, Jr.
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