ML12340A558
| ML12340A558 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/19/2012 |
| From: | Hopenfeld J Riverkeeper |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| References | |
| RAS 22625, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01 | |
| Download: ML12340A558 (17) | |
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l 05000286 RIV000102 Exhibit #: RIV000102-00-BD01 Identified: 10/15/2012 Submitted: June 19, 2012 Admitted: 10/15/2012 Withdrawn:
Rejected: Stricken:
Other:
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD
)
In the Matter of ) Docket Nos.
) 50-247-LR Entergy Nuclear Operations, Inc. ) and 50-286-LR (Indian Point Nuclear Generating )
Units 2 and 3) ) June 19, 2012
___________________________________________ )
PREFILED WRITTEN TESTIMONY OF DR. JORAM HOPENFELD REGARDING CONTENTION NYS-38/RK-TC-5 On behalf of Riverkeeper, Inc. (Riverkeeper), Dr. Joram Hopenfeld submits the following testimony regarding the State of New York and Riverkeepers Joint Contention NYS-38/RK-TC-5.
1 Q. Please state your name and address.
2 A. My name is Dr. Joram Hopenfeld and my business address is 1724 Yale Place, Rockville, 3 Maryland 20850.
4 5 Q. Please describe your educational and professional background?
6 A. I have received a B.S. and M.S. in engineering, and a Ph.D. in mechanical engineering 7 from the University of California in Los Angeles. I am an expert in the field relating to nuclear 8 power plant aging management. I have 45 years of professional experience in the fields of 9 nuclear safety regulation and licensing, design basis and severe accidents, thermal-hydraulics, 10 material/environment interaction, corrosion, erosion, fatigue, cavitation (i.e. fatigue induced 11 metal degradation), fouling, radioactivity transport, industrial instrumentation, environmental 12 monitoring, pressurized water reactor (PWR) steam generator (SG) transient testing and 13 accident analysis, design, and project management. My curriculum vitae, which has previously 14 been provided in this proceeding as Exhibit RIV000004, fully describes my education, 15 professional experience, and publications.
16 17 18 1
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1 Q. What is the purpose of your testimony?
2 A. I was retained by Riverkeeper as an expert witness in the proceedings concerning the 3 application by Entergy Nuclear Operations, Inc. (Entergy) for the renewal of two separate 4 operating licenses for the nuclear power generating facilities located at Indian Point on the east 5 bank of the Hudson River in the Village of Buchanan, Westchester County, New York, for 6 twenty years beyond their current expiration dates. The purpose of this testimony is to provide 7 support for, and my views on Contention NYS-38/RK-5, jointly filed by Riverkeeper and the 8 State of New York, concerning Entergys failure to demonstrate that it has programs to 9 effectively manage the aging of several critical components or systems during the proposed 20-10 year extended operating terms. Contention NYS-38/RK-5, which was admitted by the Atomic 11 Safety & Licensing Board (ASLB) on November 20, 2011, identified three aging management 12 programs (AMP) at Indian Point that Entergy has failed to demonstrate meet NRC regulations 13 because they rely upon commitments to take future action: Entergys programs for managing (1) 14 the fatigue of metal components, (2) primary water stress corrosion cracking for the steam 15 generator divider plates, and (3) reactor vessel internals. 1 My testimony specifically addresses 16 the first of these programs, that is, Entergys failure to demonstrate that metal fatigue of reactor 17 components will be adequately managed during the proposed periods of extended operation at 18 the plant as required by 10 C.F.R. § 54.21(c).
19 20 Q. Please describe your professional experience specifically as it relates to metal 21 fatigue.
22 A. My education, experience, extensive knowledge, and public recognition make me well 23 qualified to provide opinions and testimony related to the material degradation phenomenon 24 known as metal fatigue, that is, the fatigue or cyclic stress of metal parts due to repeated 25 stresses during plant operation. Understanding fatigue analyses requires knowledge of 26 temperature distributions and oxygen concentrations during plant transient and demands intimate 27 knowledge of heat and mass transfer. As outlined in my curriculum vitae, my education, 28 experience, and publications have afforded me with relevant expertise.
1 In the Matter of Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos.
50-0247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, State of New York and Riverkeepers New Joint Contention NYS-38/RK-TC-5 (Sept. 30, 2011), ADAMS Accession No. ML11273A196.
2
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1
2 The major fields of study I pursued to obtain my Doctorate degree were heat transfer and mass 3 transfer, fluid dynamics, and electrochemistry. This educational background qualifies me to 4 opine on issues related to metal fatigue because a major element in fatigue analysis is water 5 chemistry and mass transfer since such issues are related to the calculations of environmental 6 correction factors, or Fen. The numerous peer reviewed articles I have published, as listed in 7 my curriculum vitae, include a professional article related to such issues. 2 8
9 While employed at the Atomic Energy Commission (AEC) and Energy Research and 10 Development Administration (ERDA) from 1971 to 1977, I managed a program involving 11 various matters, including the following: experimental and numerical modeling of flow mixing 12 and heat transfer in fuel assemblies and heat exchangers; stratified flow/thermal striping and 13 fatigue analysis; cavitation (a fatigue induced metal degradation) and corrosion in water and in 14 sodium; natural circulation; jet mixing; and effects of the leak environment on fatigue crack 15 growth in sodium. On this last topic, I published a report. 3 Notably, the effects of stratified flow 16 and thermal striping on fatigue was recognized at the AEC years before Westinghouse 17 recognized that it could also exist in PWRs, and now stratification is a major element in fatigue 18 analysis.
19 20 For 18 years in the employ of the U.S. Nuclear Regulatory Commission (NRC), I worked on 21 assessing and resolving issues relating to PWR SGs including conducting extensive studies on 22 tube degradation (via fatigue, stress corrosion cracking, wall thinning, and denting) and their 23 consequences in PWR SGs. Knowledge of crack formation and detection is directly related to 24 fatigue analysis and fatigue management. In addition, SG components under steady or transient 25 conditions are also related to fatigue analysis. Notably, in the mid 1990s, I formulated and raised 26 new concerns, designated a Differing Professional Opinion (DPO), about tube cracking, crack 2
See Curriculum Vitae of Joram Hopenfeld, at p.4 (citing Experience and Modeling of Radioactivity Transport Following Steam Generator Tube Rupture, Nuclear Safety, 26, 286, 1985).
3 See Hopenfeld, et al., Small Sodium to Gas Leak Behavior in relation to LMFBR Leak Detection (International Conference on Liquid Metal Technology, May 3-6, 1976), http://www.osti.gov/bridge/servlets/purl/7252195-KSuoLF/7252195.pdf (Exhibit RIV000103). I also was the U.S. representative in attendance at the International Conference on Cavitation in Fast Breeder Reactors, where I presented on experimental investigation of cavitation inception in a flowing sodium environment.
3
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1 detection, and safety consequences following certain transients (such as steam line breaks, tube 2 ruptures, and station blackouts), and, following hearings in 2000, the Advisory Committee on 3 Reactor Safeguards largely agreed with the issues raised in the DPO and the NRC initiated a 4 decade long program to address such concerns. 4 5
6 While at NRC, I also managed an international multimillion dollar cooperative program 7 involving NRC, Westinghouse, the United Kingdom and EPRI, called the MB2, for the testing of 8 a full size section of Westinghouse Model F steam generator under steady state and operational 9 transients, the data from which was extensively used to validate computer codes and analytical 10 models. Understanding the temperature response of components under transients is crucial to 11 understanding fatigue analyses.
12 13 Over the past several decades, I have been a sought after source of expertise about tube 14 degradation. For example, in the early 1990s, I testified in Congress in connection with material 15 degradation at the Trojan Nuclear Power Plant; a major law firm representing Indian Point 16 retained me as a consultant in connection with the tube accident that occurred at Indian Point; 17 and most recently, I was extensively sought after for an expert opinion by the media in relation to 18 the discovery of severe tube degradation at the San Onofre Nuclear Generating Station in 19 January 2012. 5 20 21 Q. What materials have you reviewed in preparation for your testimony?
22 A. I have reviewed numerous documents in preparation of my testimony, including the 23 following: Entergys License Renewal Application (LRA) concerning the Indian Point nuclear 24 power plant, submitted to the NRC on or about April 30, 2007; Entergys Amendment 2 to the 25 LRA, dated January 22, 2008, which contained information regarding Entergys aging 4
Memorandum from S. Collins (NRR) to W. Travers (EDO), Re: Steam Generator Action Plan Revision to Address Differing Professional Opinion on Steam Generator Tube Integrity (WITS ITEM 200100026), May 11, 2011, http://pbadupws.nrc.gov/docs/ML0113/ML011300073.pdf , ADAMS Accession No. ML011300073 (Exhibit RIV000104); see also NUREG-1740, Voltage-Based Alternative Repair Criteria, A Report to the Advisory Committee on Reactor Safeguards by the Ad Hoc Subcommittee on a Differing Professional Opinion (March/Feb.
2001), at page 5http://pbadupws.nrc.gov/docs/ML0107/ML010750315.pdf, ADAMS Accession No. ML010750315 (Exhibit RIV000105).
5 See, e.g., Associated Press, Nuke inspectors focus on 'unusual' wear on tubes, Fox News.com, February 3, 2012, http://www.foxnews.com/us/2012/02/03/nuke-inspectors-focus-on-unusual-wear-on-tubes/ (Exhibit RIV000106).
4
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1 management program for addressing metal fatigue; all of the pleadings involving Riverkeeper 2 Contentions TC-1, TC-1A, TC-1B, and TC-5; Entergys submission to the ASLB on August 10, 3 2010 entitled, Notification of Entergys Submittal Regarding Completion of Commitment 33 4 for Indian Point Units 2 and 3, NL-10-082; Entergys refined Environmental Fatigue 5 Evaluations for Indian Point Units 2 and 3 generated by Entergys vendor Westinghouse in June 6 2010; NRC Staffs Request for Additional Information for the Review of the Indian Point 7 Nuclear Generating Unit Numbers 2 and 3, License Renewal Application, dated February 10, 8 2011; Entergys Response to Request for Additional Information (RAI), Aging Management 9 Programs, Indian Point Nuclear Generating Unit Nos. 2 & 3, Docket Nos. 50-247 and 50-286, 10 License Nos. DPR-26 and DPR-64, NL-11-032, dated March 28, 2011; NRC Staffs Safety 11 Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos.
12 2 and 3, Supplement 1 dated August 2011 (SER Supplement 1); and Entergys Letter to the 13 ASLB dated May 15, 2012 pertaining to the timing additional metal fatigue evaluations to be 14 performed by Entergy. In addition, I have reviewed hundreds of documents identified by 15 Entergy as relevant to Riverkeepers technical safety contentions, numerous relevant NUREG 16 reports, scientific and scholarly reports and articles, industry guidance documents and reports, 17 and other documents generated by NRC, Entergy, industry groups, and scientific organizations.
18 I have used the above-referenced documents to inform me of the relevant facts and derive my 19 conclusions.
20 21 Numerous documents I have relied upon in forming the opinions contained in this testimony 22 have been previously submitted in this proceeding in support of Contention NYS-26B/RK-TC-23 1B, as follows: Exhibits RIV000036-058, NYS00146A-146C, NYS00147A-147D, NYS000160, 24 NYS000161, NYS000195, NYS000325, NYS00326A-326F, NYS000346, NYS000349-352, 25 NYS000354-358B, NYS000361-369B. In addition, I have relied upon certain additional 26 documents in forming the opinions contained in this testimony; these documents are provided in 27 support of Contention NYS38/RK-5 as Exhibits RIV000103 to RIV000106 and NYS000395. To 28 the best of my knowledge, these are all true and accurate copies of each document that I used 29 and/or relied upon in preparing this testimony.
30 31 5
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1 Q. What is metal fatigue?
2 A. As I explained at length in written testimony and a report submitted in support of 3 Riverkeeper Contention RK-TC-1B earlier in this proceeding, metal fatigue is an aging 4 phenomenon that refers to when a structure or test specimen is subjected to repeated, cyclic, 5 loading during plant operation, under which a crack will initiate and the structure will fail under 6 stresses that are substantially lower than those that cause failure under static loadings. 6 Material 7 composition, strain rate, temperature and local water chemistry are some of the factors that 8 contribute to fatigue of metal parts. 7 During each loading cycle, a certain fraction of the fatigue 9 life of a component is used up depending on the magnitude of the applied stress, and eventually, 10 after the number of allowable cycles, N, the structure will use all its fatigue life. 8 The number of 11 cycles actually experienced at any given stress amplitude, n, divided by the corresponding 12 number of allowable cycles, N, is called the usage fatigue factor (CUF). 9 The maximum 13 number of cycles that should be experienced by any structure or component should always result 14 in a CUF that does not exceed 1.0, or unity.Section III of the American Society of Mechanical 15 Engineers (ASME) Code provides fatigue curves in air for various materials which specify the 16 allowable number of cycles for a given stress intensity. 10 The ASME Code requires that the 17 CUF at any given location be maintained below one. 11 18 19 Q. Does metal fatigue have safety implications?
20 A. Yes. As I discussed in previous submittals support of Riverkeeper Contention RK-TC-21 1B, metal fatigue may result in small leaks or cracks that could lead to pipe ruptures and/or other 22 equipment malfunctions. 12 Such failures can interfere with the safe operation of the plant and 23 have serious consequences to public health and safety. 13 24 6
RIV000034 at 4:23-21, 5:1-6; RIV000035 at pp.1-3.
7 RIV000034 at 4:23-21, 5:1-6; RIV000035 at pp.1-3.
8 RIV000034 at 4:23-21, 5:1-6; RIV000035 at pp.1-3.
9 RIV000034 at 4:23-21, 5:1-6; RIV000035 at pp.1-3.
10 RIV000034 at 4:23-21, 5:1-6; RIV000035 at pp.1-3.
11 RIV000034 at 4:23-21, 5:1-6; RIV000035 at pp.1-3.
12 RIV000034 at 5:8-17; RIV000035 at p.3.
13 RIV000034 at 5:8-17; RIV000035 at p.3.
6
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1 Q. Please explain how component susceptibility to metal fatigue is predicted.
2 A. Once again, as I explained in previous submittals in support of Riverkeeper Contention 3 RK-TC-1B, crack growth rate for a given stress intensity can be predicted by multiplying an 4 individual usage factor by a corresponding environmental correction factor, or Fen, to account 5 for the actual reactor environment. 14 Fen is the ratio of the fatigue life in air at room temperature 6 to the fatigue life in water at the local temperature, and the environmentally corrected CUF is 7 expressed as CUFen. 15 Argonne National Laboratory (ANL) has developed equations for 8 determining Fen factors in terms of temperature (T), dissolved oxygen (DO), sulfur content (S),
9 and strain rate (e): Fen = f(T, DO, S, e). 16 ANL conducted laboratory tests under controlled 10 conditions to generate Fen factors, and described required adjustments to be made to the 11 laboratory data to account for the actual reactor environment. 17 12 13 Q. Entergys LRA contained the results of an analysis of the effects of environmentally 14 assisted fatigue on certain reactor components during the proposed period of extended 15 operation. What was the outcome of this CUFen analysis?
16 A. Entergys LRA included the results of assessment of the effect of the reactor water 17 environment on fatigue life for a sample of six components prescribed in NUREG/CR-6260, 18 Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant 19 Components (1995). 18 LRA Tables 4.3-13 and 4.3-14 indicated that the CUFen of four of these 20 risk significant reactor components would exceed unity during the period of extended operation.
21 22 Q. Did Entergy undertake any steps in response to these findings?
23 A. To purportedly demonstrate that metal fatigue will be managed throughout the period of 24 extended operation in light of these findings, Entergy committed to performing a refined fatigue 25 analysis, in relation to the same components, in order to lower the predicted CUFen values to less 26 than 1.0. The results of this refined environmentally assisted fatigue (EAF) analysis, 14 RIV000034 at 5:19-31, 6:1-4; RIV000035 at pp.1-3.
15 RIV000034 at 5:19-31, 6:1-4; RIV000035 at pp.1-3.
16 RIV000034 at 5:19-31, 6:1-4; RIV000035 at pp.1-3.
17 RIV000034 at 5:19-31, 6:1-4; RIV000035 at pp.1-3.
18 NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components (1995), at 4-1 (Exhibit NYS000355).
7
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1 reported in revised LRA Tables 4.3-13 and 4.3-14 in August 2010, indicated that the CUFen 2 values for the locations evaluated, i.e. the same six sample locations prescribed in NUREG/CR-3 6260, were all below 1.0. 19 As explained at length in written testimony submitted in support of 4 Riverkeeper Contention RK-TC-1B earlier in this proceeding, Entergys refined analyses do 5 not demonstrate that the CUFen for the components evaluated will not exceed unity (1.0) during 6 the proposed extended licensing terms, because Entergy employed a flawed methodology that 7 failed to account for all relevant plant parameters, and which resulted in underestimated fatigue 8 predictions. 20 9
10 Q. In your opinion, were Entergys refined fatigue analyses an adequate response to 11 Entergys initial finding that the CUFen of four reactor components would exceed unity 12 during the period of extended operation?
13 A. No, because Entergy did not expand the scope of the fatigue analysis beyond simply 14 representative components, to identify other components whose CUFen may be greater than 1.0.
15 Since the CUFen of several NUREG/CR-6260 components exceeded unity, and the non-16 environmentally corrected CUFs of other components were very close to unity, Entergy should 17 have expanded their fatigue analysis. However, Entergys refined EAF evaluations did not 18 expand the scope of components analyzed, but rather only assessed those locations identified in 19 NUREG/CR-6260.
20 21 According to regulatory and industry guidance, since the CUFen for various components were 22 initially found to exceed the regulatory threshold of 1.0, as presented in original LRA Tables 4.3-23 13 and 4.3-14, Entergy is required to identify and investigate additional reactor locations for 24 potential high susceptibility to metal fatigue. In particular, according to industry guidance 25 document, MRP-47, Revision 1, Electric Power Research Institute, Materials Reliability 26 Program: Guidelines for Addressing Fatigue Environmental Effects in a License Renewal 27 Application (2005),
19 NL-10-082,Completion of Commitment #33 Regarding the Fatigue Monitoring Program, Entergy Nuclear Operations, Inc., Indian Point Nuclear Generating Unit Nos. 2 & 3, Docket Nos. 50-247 and 50-286, License Nos.
DPR-26 and DPR-64 (August 9, 2010) (Exhibit NYS000352).
20 RIV000034 at pp.6-20; RIV000035 at pp.4-21.
8
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1 plant-unique evaluations may show that some of the NUREG/CR-2 6260 [2] locations do not remain within allowable limits for 60 3 years of plant operation when environmental effects are 4 considered. In this situation, plant specific evaluations should 5 expand the sampling of locations accordingly to include other 6 locations where high usage factors might be a concern. 21 7
8 In addition, NUREG-1801, Generic Aging Lessons Learned (GALL) Report, (GALL Report) 9 Revision 1, specifies that [f]or programs that monitor a sample of high fatigue usage locations, 10 corrective actions include a review of additional affected reactor coolant pressure boundary 11 locations, and that sample locations identified in NUREG/CR-6260 are simply the minimum 12 set of components to analyze. 22 Furthermore, Revision 2 of the GALL Report (the most recent 13 version of the report) specifies that the sample set for fatigue calculations that consider the 14 effects of the reactor water environment should include the locations identified in NUREG/CR-15 6260 and additional plant-specific component locations in the reactor coolant pressure boundary 16 if they may be more limiting than those considered in NUREG/CR-6260. 23 Entergys fatigue 17 analyses to date demonstrate that the components analyzed will likely exceed unity, and were, 18 therefore, not necessarily the most limiting locations and bounding for the entire plant.
19 20 Q. Are you aware of whether or not Entergy ever plans to expand the scope of its 21 fatigue analysis to identify other components at Indian Point whose CUFen may exceed 22 unity during the proposed periods of extended operation?
23 A. My review of an NRC Staff RAI dated February 10, 2011, Entergys response thereto 24 dated March 28, 2011, and NRC Staffs SER Supplement 1, dated August 2011, indicates that 25 Entergy may expand the scope of its fatigue analysis at some point in the future before entering 26 the period of extended operation. In particular, NRC Staffs RAI requested that Entergy 27 [c]onfirm and justify that the locations selected for environmentally assisted fatigue analyses in 28 LRA Tables 4.3-13 and 4.3-14 consist of the most limiting locations for the plant (beyond the 29 generic components identified in the NUREG/CR-6260 guidance) and clarify which locations 21 Exhibit NYS000350 at 3-4 (emphasis added).
22 NUREG-1801, GALL Report, Rev. 1 § X.M1, Metal Fatigue of Reactor Coolant Pressure Boundary, ¶¶ 5, 7 (emphasis added)) (Exhibit NYS00146A-146C).
23 NUREG-1801, Gall Report, Rev. 2 § X.M1, Fatigue Monitoring, ¶ 1 (emphasis added) (Exhibit NYS00147A-147D).
9
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1 require an environmentally-assisted fatigue analysis and the actions that will be taken for these 2 additional locations. 24 Thus, NRC Staff has now conceded that there may be more limiting 3 components, and that the CUFen values in LRA Tables 4.3-13 and 4.3-14 may not be bounding.
4 5 In response, Entergy provided a vague commitment (Commitment 43), as follows:
6 Entergy will review design basis ASME Code Class 1 fatigue 7 evaluations to determine whether the NUREG/CR-6260 locations 8 that have been evaluated for the effects of the reactor coolant 9 environment on fatigue usage are the limiting locations for the 10 Indian Point 2 and 3 plant configurations. If more limiting 11 locations are identified, the most limiting location will be 12 evaluated for the effects of the reactor coolant environment on 13 fatigue usage. 25 14 15 NRC Staffs SER Supplement 1 memorializes NRC Staffs acceptance of this commitment. 26 16 17 My review of recent correspondence from Entergy to the ASLB dated May 15, 2012, indicates 18 that [t]he due date for completion of this Commitment [43] is prior to September 28, 2013 for 19 Indian Point Energy Center (IPEC) Unit 2 and prior to December 12, 2015 for IPEC Unit 3 20 and that Entergy has determined that the initial screening review of design basis ASME Code 21 Class 1 fatigue evaluations, as described in Commitment 43, to determine whether the 22 NUREG/CR-6260 locations are the limiting locations for IPEC, is expected to be completed 23 within approximately the next four months and that [i]f more limiting locations are identified, 24 then Entergy will take further actions as necessary in accordance with Commitment 43. 27 25 24 U.S. NRC Request for Additional Information for the Review of the Indian Point Nuclear Generating Unit Numbers 2 and 3, License Renewal Application (February 10, 2011), at 13 (Exhibit RIV000057).
25 Entergy Response to Request for Additional Information (RAI), Aging Management Programs Indian Point Nuclear Generating Unit Nos. 2 & 3, Docket Nos. 50-247 and 50-286, License Nos. DPR-26 and DPR-64 (March 28, 2011), at p.26 of 27 (Exhibit RIV000058).
26 Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3, Docket Nos. 50-247 and 50-286, NUREG-1930, Supplement 1 (August 2011), at 4-2 (Exhibit NYS000160).
27 Correspondence from K. Sutton, P. Bessette (Counsel for Entergy) to (L. McDade, R. Wardwell, M. Kennedy (ASLB) (May 15, 2012) (Exhibit NYS000395).
10
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1 Q. In your opinion, is Entergys Commitment 43 adequate to demonstrate that Entergy 2 has a program to effectively manage metal fatigue during the proposed periods of extended 3 operation at Indian Point?
4 A. No, because Entergy has failed to identify the locations that may be more limiting, and 5 which will be the subject of CUFen calculations, now (that is, during the license renewal 6 proceeding), and, has instead, only articulated a plan to determine all critical component 7 locations later, at some point before entering the proposed extended periods of operation. This 8 approach disallows meaningful review by NRC and the public of a critical safety issue.
9 10 Entergys initial EAF analyses, as memorialized in Entergys original LRA Tables 4.3-13 and 11 4.3-14, as well as Entergys refined EAF analyses, demonstrate that the components analyzed 12 will likely exceed unity, and that, therefore, those components were not necessarily the most 13 limiting locations and bounding for the entire plant. Likewise, NRC Staff has now 14 acknowledged that there may be more limiting components and that the CUFen values in LRA 15 Tables 4.3-13 and 4.3-14 may not be bounding. NRC Staff has rightly questioned Entergys 16 claim that LRA Tables 4.3-13 and 4.3-14 represent limiting conditions for the entire Indian Point 17 plant. The most recent version of the GALL Report clearly specifies that fatigue calculations 18 should include . . . additional plant-specific component locations in the reactor coolant pressure 19 boundary if they may be more limiting than those considered in NUREG/CR-6260. 28 Therefore, 20 it was not appropriate for NRC Staff to accept Entergys vague commitment to determine at 21 some point in the future what additional locations must be analyzed. An actual analysis to 22 determine the most limiting locations must be performed before a determination is made about 23 license renewal.
24 25 Entergys commitment to review design basis ASME Code Class 1 fatigue evaluations is not 26 confirmation or justification that the locations selected for environmentally-assisted fatigue 27 analyses in LRA Tables 4.3-13 and 4.3-14 consist of the most limiting locations for the plant.
28 Entergy has, to date, not provided any analysis that would support a conclusion that the CUFen 29 values in LRA Tables 4.3-13 and 4.3-14 bound all other components at the plant. Entergy has 28 NUREG-1801, Gall Report, Rev. 2 § X.M1, Fatigue Monitoring, ¶ 1 (emphasis added) (Exhibit NYS00147A-147D).
11
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1 not conducted the analysis that would be required to confirm and justify that the components in 2 LRA Tables 4.3-13 and 4.3-14 are bounding. Instead, Entergy has indicated that the process to 3 be used to determine the most limiting locations for which CUFen calculations, and the selection 4 of the locations to be analyzed, will be disclosed in the future, apart from the license renewal 5 review process. Entergy has failed to identify the limiting locations, or describe the 6 methodology to be used to select such locations. This does not affirmatively demonstrate that 7 the effects of aging on the intended function(s) will be adequately managed for the period of 8 extended operation as required by 10 C.F.R. § 54.21(c)(1)(iii). Entergy has simply failed to 9 provide sufficient information in order to assess whether Entergys AMP for metal fatigue is 10 adequate.
11 12 Q. Please describe what an analysis to determine the most limiting locations at Indian 13 Point will require.
14 A. Determining the most limiting locations at Indian Point is not a clearly defined analysis.
15 To the contrary, there are numerous considerations and factors.
16 17 The first step for determining the most limiting locations is component identification: selecting 18 and listing all components that are susceptible to fatigue, including but not limited to nozzles, 19 reducers, mixing tees and bends in feed water lines, surge lines, spray lines, and volume control 20 system lines. The second step is component screening: the selected components must be 21 screened and ranked with respect to their most vulnerable locations, considering parameters that 22 are known to effect fatigue life. These include the ratios of the local heat transfer coefficient, the 23 local material conductivity, wall thickness, fluid temperature, 'T, dissolved oxygen levels, flow 24 velocities, number of transients, magnitude and cycling frequency of surface temperatures, 25 (thermal striping in stratified flows) and loads, and surface discontinuities and flow 26 discontinuities in each component. Moreover, a determination of the most limiting locations 27 should also include an assessment of actual experience at Indian Point as well as at other PWR 28 plants. In addition, thermal striping during stratification should be generally considered as these 29 effect fatigue life, and since the GALL Report requires that environmental effects be included in 30 the calculations and does not exclude thermal striping from such requirements. Only after all 12
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1 such considerations can a detailed numerical fatigue analysis be conducted in relation to the 2 identified areas.
3 4 Additionally, an adequate assessment must also consider the synergistic aging effects of primary 5 water stress corrosion cracking (PWSCC) and thermal fatigue. For example, as discussed in 6 Riverkeeper and the State of New Yorks joint contention, Entergy has acknowledged a problem 7 with PWSCC for the nickel alloy or nickel-alloy clad SG divider plates exposed to reactor 8 coolant. 29 As stated in Entergys LRA, the CUF of record for the SG divider plate is already 9 considerably high: 0.683 for Indian Point Unit 2 and 0.789 for Indian Point Unit 3. 30 These 10 CUFs may exceed unity when they are corrected for the effects of PWSCC and the environment.
11 Importantly, the effect of opening the divider plate on the natural convection flow through the 12 steam generator following station blackouts (SBO) and anticipated transients without scram 13 (ATWS) is an important consideration in this regard.
14 15 Despite the complexity involved in determining the most limiting locations at Indian Point, 16 Entergy has failed to provide any information about how this analysis will be performed to allow 17 for meaningfully comment upon the adequacy of the analysis. This leaves Entergys AMP 18 insufficient, as it does not comply with the directive in the GALL Report or demonstrate that 19 metal fatigue will be appropriately monitored, managed and corrected during the period of 20 extended operation.
21 22 Q. Do you have an opinion regarding what components Entergy should evaluate to 23 determine whether they may be more limiting?
24 A. Yes. Since Entergy has not provided any indication to date about how expansive their 25 search for more limiting locations will be, I have prepared the following table consisting of a list 26 of components, as a sample of what Entergys effort must consider at a minimum to determine 27 whether Indian Point can operate safely during the proposed life extension periods:
29 In the Matter of Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos.
50-0247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, State of New York and Riverkeepers New Joint Contention NYS-38/RK-TC-5 (Sept. 30, 2011), ADAMS Accession No. ML11273A196; Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3, Docket Nos. 50-247 and 50-286, NUREG-1930, Supplement 1 (August 2011), at 3-18 to 3-19 (Exhibit NYS000160).
30 LRA Tables 4.3-9, 4.3-10.
13
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1
Component Location CUF on Record 31 IP2 Reactor Vessel Inlet Nozzles at weldements 0.050 IP3 Reactor Vessel Inlet Nozzles at weldements 0.049 IP2 Reactor Vessel Outlet Nozzles at 0.281 weldements IP3 Reactor Vessel Outlet Nozzles at 0.259 weldements IP2 Reactor Vessel Lower Core Support Plate 0.521 Internals IP3 Reactor Vessel Lower Core Plate 0.237 Internals IP2 Pressurizer Spray Nozzle 0.996 IP3 Pressurizer Spray Nozzle 0.974 IP2/IP3 Pressurizer Lower Head at all n/a weldements and at heater penetration IP2 Steam Generator, Divider Plate 0.683 Primary Side IP3 Steam Generator, Divider Plate 0.789 Primary Side IP2 Steam Generator, Tube to tubesheet welds 0.809 Primary Side IP2 Steam Generator, Main feedwater nozzle 0.898 Secondary Side IP3 Steam Generator, Main feedwater nozzle 1.00 Secondary Side IP2 Steam Generator, Steam nozzle 0.212 Secondary Side IP3 Steam Generator, Steam nozzle 0.023 Secondary Side IP2 Steam Generator, Steam nozzle support ring 0.220 Secondary Side 31 See LRA Tables 4.3-3, 4.3-4, 4.3-5, 4.3-6, 4.3-7, 4.3-8, 4.3-9, 4.3-10.
14
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 IP3 Steam Generator, Steam nozzle support ring 0.894 Secondary Side IP2 Steam Generator, Tubes 0.484 Primary Side IP3 Steam Generator, Tubes 0.161 Primary Side IP2/IP3 Reactor Outlet Nozzle n/a Pump IP2/IP3 RHR SI Nozzle n/a IP2/IP3 Mixing Tees RHR system n/a IP2/IP3 Piping Pressurizer Spray line n/a IP2/IP3 Piping Unisolable branches n/a connected to RCS piping 1
2 I derived this list based on the LRA, reactor experience, and extensive literature review.
3 4 Q. Are you familiar with Entergys use of the computer model WESTEMS' in 5 relation to metal fatigue at Indian Point?
6 A. I am aware that Entergy has indicated that it relies upon the WESTEMS' computer 7 model in performing CUFen calculations.
8 9 Q. Do you have an opinion regarding NRC Staffs approval of Entergys commitment 10 to provide explanations and justifications of any user intervention in future calculations 11 using the WESTEMS' computer model at some point in the future, prior to the expiration 12 of the current Indian Point reactor operating licenses, but not prior to a decision on license 13 renewal, as memorialized in NRC Staffs SER Supplement 1? 32 14 A. Entergy must specify the criteria and assumptions upon which it will rely to modify the 15 WESTEMS' computer model for the calculation of CUFen prior to a decision on license 16 renewal. Varying criteria and assumptions to prospectively be employed may affect the validity 17 and robustness of the analysis. Thus, without specifying the modifications to be made to the 18 model, or the process for deciding when and how to have user intervention in the use of the 32 Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3, Docket Nos. 50-247 and 50-286, NUREG-1930, Supplement 1 (August 2011), at 4-2 to 4-3 (Exhibit NYS000160).
15
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of NYS-38/RK-TC-5 1 model, Entergy has not demonstrated that the aging effects of metal fatigue will be adequately 2 managed.
3 4 Q. Please summarize your opinions regarding whether or not Entergy has 5 demonstrated that metal fatigue of reactor components will be adequately managed during 6 the proposed periods of extended operation as required by 10 C.F.R. § 54.21(c).
7 A. In light of NRC Staffs acceptance of vague commitments to perform necessary metal 8 fatigue investigations, analyses, and justifications, in the future, Entergy has failed to 9 demonstrate that the aging effects of metal fatigue will be adequately managed for the proposed 10 periods of extended operation, and has, thus, failed to comply with 10 C.F.R. § 54.21(c) or 11 regulatory guidance, including the GALL Report. Entergy has failed to make the affirmative 12 demonstration that it has a program to sufficiently monitor, manage, and correct metal fatigue-13 related degradation at Indian Point.
14 15 Q. Does this conclude your initial testimony regarding Contention NYS-38/RK-TC-5?
16 A. Yes.
16
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD
)
In the Matter of )
)
Entergy Nuclear Operations, Inc. ) Docket Nos.
(Indian Point Nuclear Generating ) 50-247-LR Units 2 and 3) ) and 50-286-LR-
~----------------------)
DECLARATION OF DR. JORAM HOPENFELD I, Joram Hopenfeld, do hereby decJare under penalty of perjury that my statements in the foregoing testimony and my statement ofprofessional qualifications are true and correct to the best of my knowledge and belief.
Executed in Accord with 10 C.F.R. § 2.304(d)
~oram openfeld, Ph.D.
, /
'.. /
, 724 Yale Place Rockville, MD 20850 June~2012 17