ML12335A400
| ML12335A400 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 12/21/2011 |
| From: | Hopenfeld J Riverkeeper |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| References | |
| RAS 21600, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01 | |
| Download: ML12335A400 (21) | |
Text
United States Nuclear Regulatory Commission Official Hearing Exhibit RIV000003 Entergy Nuclear Operations, Inc. Submitted: December 22, 2011 In the Matter of:
(Indian Point Nuclear Generating Units 2 and 3) c,.",tJ"oP< REGlJ<..q" ASLBP #: 07-858-03-LR-BD01
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Docket #: 05000247 l 05000286 "0 Exhibit #: RIV000003-00-BD01 Admitted: 10/15/2012 i Rejected:
Identified: 10/15/2012 Withdrawn:
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Stricken:
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
)
PREFILED WRITTEN TESTIMOY OF DR. JORAM HOPENFELD REGARDING RIVERKEEPER CONTENTION TC FLOW ACCELERATED CORROSION On behalf of Riverkeeper, Inc. ("Riverkeeper"), Dr. Joram Hopenfe1d submits the following testimony regarding Riverkeeper Contention TC-2.
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. What is your educational and professional background?
6 A. I have received the following degrees from the University of California in Los Angeles: a 7 B.S. and M.S. in engineering, and a Ph.D. in mechanical engineering. I am an expert in the field 8 relating to nuclear power plant aging management. I have 45 years of professional experience in 9 the fields of nuclear safety regulation and licensing, design basis and severe accidents, thermal-10 hydraulics, material/environment interaction, corrosion, fatigue, radioactivity transport, industrial 11 instrumentation, environmental monitoring, pressurized water reactor steam generator transient 12 testing and accident analysis, design, and project management, including 18 years in the employ 13 of the U.S. Nuclear Regulatory Commission ("NRC"). My education and professional 14 experience are described in my curriculum vitae, which is provided as Exhibit RIV000004.
15 16 Q. What is the purpose of your testimony?
17 A. The purpose of my testimony is to provide support for, and my views on, Riverkeeper's 18 Contention TC-2 related to the aging effects of flow-accelerated corrosion ("FAC") at Indian 19 Point Generating Unit Nos. 2 and 3 during proposed 20-year extended operating terms. This
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 contention was admitted by the Atomic Safety & Licensing Board ("ASLB") on July 31 , 2008. 1 2 Riverkeeper asserts that Entergy Nuclear Operations, Inc. ("Entergy"), the owner of Indian Point, 3 has failed to demonstrate that FAC will be adequately managed during the proposed periods of 4 extended operation at the plant as required by 10 C.F.R. § 54.21(c).
5 6 Q. Please describe your professional experience specifically as it relates to FAC.
7 A. I have published numerous peer-reviewed papers in the area of corrosion, and hold 8 patents related to monitoring of wall thinning of piping components. I have knowledge and 9 expertise regarding the use ofthe CHECWORKS computer code, a program that was developed 10 in an attempt to manage FAC at nuclear power plants. My familiarity with the CHECWORKS 11 code dates back to 1988, when it was known as CHEC. Most recently, I was a technical 12 consultant and expert witness for the New England Coalition in the Vermont Yankee license 13 renewal proceeding, where I testified at an adjudicatory hearing concerning FAC and the use of 14 CHECWORKS.
15 16 Q. Have you prepared a report in support of your testimony?
17 A. Yes, I prepared an expert report, provided as Exhibit RIV000005, which reflects my 18 analysis and opinions.
19 20 Q. What materials have you reviewed in preparation for your expert report and 21 testimony?
22 A. I have reviewed numerous documents in preparation of my expert report and testimony, 23 including the following: the relevant section of Entergy's License Renewal Application 24 ("LRA"), all ofthe pleadings involving Riverkeeper Contention TC-2, including Entergy's 25 Motion for Summary Disposition of Riverkeeper's Contention TC-2 and supporting attachments 26 thereto,2 relevant portions of NRC Staffs Safety Evaluation Report pertaining to the Indian Point 1 See In the Matter of Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos. 50-247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, Memorandum and Order (Ruling on Petitions to Intervene and Requests for Hearing) (July 31, 2008), at 161-62. In response to new metal fatigue evaluations performed by Entergy, Riverkeeper and NYS jointly filed an amended contention, NYS-26B/RK-TC" lB, which the ASLB admitted as superseding the previous contentions.
2 Applicant's Motion for Summary Disposition of Riverkeeper Technical Contention 2 (Flow-Accelerated Corrosion) (July 26, 2010), ADAMS Accession No. MLI02140430.
2
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 license renewal proceeding, hundreds of documents identified by Entergy as relevant to 2 Riverkeeper's FAC contention, numerous relevant NUREG reports, scientific and scholarly 3 reports and articles, industry guidance documents and reports, and other documents generated by 4 NRC, Entergy, industry groups, and scientific organizations. I have used such documents to 5 inform me of the relevant facts and derive my conclusions.
6 7 A list of the particular documents that I reference in my expert report, and which I rely upon in 8 this testimony, is included at the end of the report. Those references have been provided as 9 RIV000006 through RIV000033, (or have been previously been provided as exhibits by another 10 party in the proceeding), in support of my testimony. To the best of my knowledge, these are 11 true and accurate copies of each document that I referred to, used and/or relied upon in preparing 12 my report and this testimony. In some cases where the document was extremely long and only a 13 small portion is relevant to my testimony, an excerpt ofthe document is provided. Ifit is only an 14 excerpt, that is noted on the cover of the Exhibit.
15 16 Q. What conclusions have you reached about Entergy's program for managing FAC at 17 Indian Point during the proposed period of extended operation?
18 A. In my professional judgment, and as I describe in more detail below and in my report, 19 Entergy has failed to demonstrate that the serious aging mechanism of F AC will be adequately 20 managed throughout the proposed extended licensing terms at Indian Point. Entergy plans to 21 rely far too heavily on the CHECWORKS computer code to manage FAC during the PEO. This 22 leaves the F AC Aging Management Program ("AMP") at Indian Point fundamentally deficient, 23 since CHECWORKS is not an effective tool for predicting and managing FAC at the plant.
24 Entergy's AMP must contain sufficient details about how F AC will otherwise be monitored and 25 addressed, but it does not. As a result, Entergy has failed to demonstrate that it has a program for 26 handling FAC that meets all relevant criteria and standards.
27 28 Q. What is Flow-Accelerated Corrosion?
29 A. F AC is a pipe wall thinning phenomenon in which the thinning rate is accelerated by 30 flow velocity. When the metal is exposed to flowing liquid, flow velocity has a significant effect 31 on metal removal. FAC includes wall thinning by impingement corrosion, electrochemical 3
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfe\d In support ofRK-TC-2 (FAC) 1 corrosion, erosion-corrosion, cavitation-erosion, and chemical dissolution. F AC is affected 2 mainly by turbulence intensity, steam quality, material compositions, oxygen content, and 3 coolant pH. The rate ofFAC depends upon the local geometry, local metal composition, and 4 local turbulences. FAC can manifest in the formation of round holes and grooving. Once local 5 corrosion has begun, geometrical changes may cause F AC to increase in a non-linear rate.
6 7 Q. What are the safety implications ofFAC?
8 A. Undetected FAC can pose a significant safety risk at nuclear power plants. In order to 9 ensure that the pipes susceptible to F AC will operate safely throughout their lifetime, they are 10 designed with a corrosion allowance that is added to the minimum design wall thickness, 11 commonly known as Tcr. When FAC reduces wall thickness of a component below the 12 minimum design value, the potential exists for the component to rupture. In many cases, this 13 will be preceded by a leak, which can be detected before catastrophic rupture. A FAC-induced 14 rupture of a high pressure component or pipe may have very serious safety consequences. For 15 this reason, the ASME code, specifically requires that components and pipes do not operate 16 below design limit wall thicknesses. 3 Numerous instances of undetected FAC have previously 17 resulted in catastrophic events, including severil fatalities at the Surry nuclear power plant in 18 1986 due to a feed water pipe elbow rupture, and several fatalities at the Mihama nuclear power 19 plant as a result of FAC in the secondary loop.
20 21 Q. What is the CHECWORKS computer code?
22 A. In the wake of several industry F AC events, the CHECWORKS computer code was 23 developed to assist utilities in preventing FAC-related failures. The goal of the CHECWORKS 24 software is to predict what locations may succumb to F AC-induced wear before wall thinning 25 reaches unacceptable levels. Due to the inherent unpredictability of F AC, CHECWORKS is 26 based on statistics, meaning, a collection of selected data which represents only a fraction of the 27 total flow area. Accordingly, CHECWORKS is not a reliable predictive tool unless it is 28 adequately benchmarked for each component and for relevant plant parameters. Changes in 29 plant parameters necessitate appropriate re-calibration of the CHECWORKS code. Updating the 30 model is especially important in the event of a power increase, since power changes affects 3 ASME B31.3; ASME Code Section III, Paragraph NB-3200.
4
Docket Nos. SO-247-LR & SO-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 various plant parameters, including velocities, temperatures, coolant chemistry, and steam 2 moisture. The Electric Power Research Institute ("EPRI") has recognized that power uprates, 3 even small ones, can significantly affect the rate of F AC. 4 4
5 Q. Please describe Entergy's program for managing FAC during the PEO, as you 6 understand it.
7 A. Entergy's LRA sections § A.2.1.14 and B.1.15 indicate that its FAC program is based on 8 an EPRI guidance document, EPRI, Recommendations for an Effective Flow-Accelerated 9 Corrosion Program, and involves determining critical locations, performing inspections, and 10 undertaking corrective action if necessary. The LRA further indicates that Entergy believes its 11 F AC AMP is consistent with NUREG-180 1, Generic Aging Lessons Learned (GALL) Report 12 ("GALL Report"). Based on my review of EPRI's guidance document, as well as Entergy's 13 program implementation documents,S it is apparent that Entergy's FAC management program is 14 largely based on the use CHECWORKS to predict timing and locations of wall thinning.
15 16 Q. What is your opinion regarding Entergy's use ofCHECWORKS to manage FAC 17 during the PEO?
18 A. Entergy's dependence on CHECWORKS to select components for wall measurements 19 and to determine the time between successive thickness measurements is misplaced, because 20 CHECWORKS will fail to assure that timely detection ofFAC-related wall thinning will occur 21 during the proposed PE~. This is primarily because CHECWORKS is not properly 22 benchmarked and fails to provide reliable predictive results at Indian Point.
23 24 Q. How did you reach the conclusion that CHECWORKS is not properly 25 benchmarked at Indian Point?
26 A. Entergy provided and I reviewed more than 6,500 data points, contained in Indian Point-27 specific CHECWORKS modeling reports. These reports memorialize CHECWORKS-related 28 results from FAC-related inspections which occur during plant outages. The data I reviewed was 29 collected in relation to more than 10 outages at Indian Point, from both before and after power 4 EPRI, Recommendations for an Effective Flow-Accelerated Corrosion Program, NSAC-202L-R3, at p.4-S.
5 EN-DC-31S, Revision 3, Flow Accelerated Corrosion Program (March 1,2010).
5
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK-TC-2 (FAC) 1 uprates that occurred at Unit 2 in 2004 and Unit 3 in 2005. The data plots CHECWORKS wear 2 predictions of component wall thickness relative to actual measurements. The data I reviewed, 3 excerpted from Entergy's CHECWORKS modeling reports, has been included in support of my 4 testimony as RIV00016Aand RIVOOOI6A. Based on my extensive analysis of Entergy's data, it 5 is my professional opinion that the CHECWORKS computer code produces highly unreliable 6 and non-conservative component wear predictions. Specifically, the data shows that 7 CHECWORKS has consistently been inaccurate at Indian Point. There is a complete lack of 8 correlation between component wear predictions and actual wall thickness measurements. Allow 9 me to explain. Each graph shows three lines: a central 45° line, and lines designated +50% and 10 -50%. With a perfect correlation, all the data would fall on the 45° line. However, Entergy's 11 data shows that very few points fall on the 45° line. Generally, the graphs exhibit a wide scatter.
12 The data points that fall between the 45° line and the abscissa, i.e., the zero wear prediction x-13 axis, represent non-conservative predictions. My review of all of Entergy' s plotted data points 14 reveals that CHECWORKS has yielded non-conservative predictions about 40-60% ofthe time.
15 I have documented this finding in relation to the different data sets I reviewed in Table 1, column 16 (A) contained in my report. Additionally, with an ideal correlation each predicted point would 17 have a single measured value. Instead, Entergy's data shows that a given prediction yields many 18 widely different measured points. Furthermore, the data shows that the degree of inaccuracy of 19 CHECWORKS' predictions has been quite large. While the +/-50% lines of each graph imply 20 that CHECWORKS bounds the data within +/-50%, this is highly misleading and incorrect. The 21 data falling within this range actually represents a much larger degree of imprecision. The +50%
22 line represents conservative over-predictions in wear that vary by a factor of .7, while the -50%
23 line represents non-conservative under-predictions of wear that vary by a factor of 2. Simply 24 because this data appears within the arbitrary lines placed on these graphs does not indicate that 25 the CHECWORKS has an appropriate degree of accuracy. In any event, many data points fall 26 outside the arbitrary +/-50% lines, which indicates that CHECWORKS cannot even bound the 27 data conservatively within a factor of 2. I have documented the degree to which data fell outside 28 the +/-50% lines in Table 1, column (B) in my expert report. My review ofthe data reveals that 29 CHECWORKS can over or under predict actual measured FAC by more than a factor of 10. The 30 over-prediction or under-prediction of the data by a factorof 10 exhibited by a significant 31 number of components clearly demonstrates that the CHECWORKS model employed at Indian 6
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 Point cannot predict FAC to any degree of accuracy or precision. Instead, CHECWORKS can 2 only predict an overall range of corrosion rate that is far too wide for practical applications, 3 especially when the consequences of component failure are safety related. The consistent 4 inaccuracy of CHECWORKS, from before the power uprates at Indian Point, to well after, 5 demonstrates that the CHECWORKS model has never been properly benchmarked, that the 6 model is certainly not currently benchmarked to account for changes in plant operating 7 parameters as a result of the power increases.
8 9 Q. The graphs plotting the CHECWORKS data you reviewed all contain an "LCF."
10 Can you explain what this means?
11 A. Based on my review of Entergy's CHECWORKS reports, it is my understanding that 12 LCF stands for "line correction factor." According to Entergy, this indicates the degree of 13 CHECWORKS' under- or over- predictions. 6 Entergy uses the LCF to "adjust" the predictions 14 to match the inspection data. 7 While an LCF of 1 would represent an exact agreement between 15 CHECWORKS predictions and actual wall thickness measurements, Entergy considers the LCF 16 to be acceptable if it is between 0.5 and 2.5. 8 Interestingly, the only graphs showing an LCF of 17 1, are those figures with no data in them. Based on my review of Entergy's documentation, there 18 is no justification to support the conclusion that an LCF within the range of 0.5 to 2.5 is 19 acceptable or that an LCF within this range would be an indication that CHECWORKS can be 20 used to accurately predict inspection locations. In any event, the CHECWORKS data I reviewed 21 reveals many instances where the LCF was outside the range that Entergy claims is acceptable. I 22 have documented this in Table 1, column (c) in my expert report. Based on Entergy's own 23 criteria, it is apparent that CHECWORKS is unreasonably failing to predict wear rates.
24 25 6 CSI, Technologies, Inc., Indian Point Unit 3 CHECWORKS SF A Model, Calculation No. 0705.100-01, Revision 2, August 2,2011, at p.26.
7 Applicant's Motion for Summary Disposition of Riverkeeper Techllical Contention 2 (Flow-Accelerated Corrosion) (July 26, 2010), ADAMS Accession No. MLl 02140430, at Attachment 2, ~ 48.
8 CSI, Technologies, Inc., Indian Point Unit 3 CHECWORKS SFA Model, Calculation No. 0705.100-01, Revision 2, August 2,2011, at p.26.
7
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 Q. Are there any limitations to your review of the CHECWORKS data that Entergy 2 provided?
3 A. Yes. Despite the fact that CHECWORKS has been in use since the early 1990s when 4 EPRI sold the program to nuclear power plant operators, Entergy has no CHECWORKS related 5 documentation related to Indian Point Unit 2 generated prior to the year 2000. 9 Further, Entergy 6 did not provide any CHECWORKS related documentation related to Indian Point Unit 3 7 generated prior to 2001 , since "locating such documentation, to the extent it exists, would be 8 extremely burdensome.,,1o Thus, the data analyzed represents only a fraction of the total plant 9 data that was allegedly used to benchmark CHECWORKS at Indian Point. The available data, 10 however, is sufficient to show that the model is not adequately benchmarked for use at Indian 11 Point.
12 13 Q. You indicated that you participated in the Vermont Yankee license renewal 14 proceeding in relation to FAC and the use of CHECWORKS. Are you aware of the ASLB 15 determination in that proceeding relating to benchmarking of the CHECWORKS model at 16 Vermont Yankee?
17 A. Yes, it is my understanding that the ASLB in the Vermont Yankee license renewal 18 proceeding determined that, at Vermont Yankee, a prolonged period of benchmarking of 19 CHECWORKS at VY was not necessary.
20 21 Q. Is there any reason to believe that the ASLB determination in the Vermont Yankee 22 license renewal proceeding about the benchmarking of CHECWORKS at Vermont Yankee 23 has any relevance to the use of CHECWORKS at Indian Point?
24 Q. No, not at all. In fact, there several key differences between the use of CHECWORKS at 25 Vermont Yankee versus Indian Point. First, at Vermont Yankee, there was no post-power uprate 26 data to assess when the F AC program was under review. Instead the ASLB assumed that future 27 updates to the CHECWORKS model before Vermont Yankee entered a proposed PEO would 9 See 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, Order (Ruling on Riverkeeper's Motion to Compel) (November 4,2010), at 3.
10 See In the Matter ofEntergy 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-BDOl, Order (Ruling on Riverkeeper's Motion to Compel) (November 4,2010), at 4.
8
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 benchmark the model. My review of three post-power uprate data sets for Indian Point Unit 2 2 and three post-power uprate data sets for Indian Point Unit 2 already shows that CHECWORKS 3 has not been adequately benchmarked despite ample post-power increase model updates. In 4 addition, at Vermont Yankee, prolonged benchmarking was found not to be necessary because 5 the plant had the benefit of data dating back to 1989. There is no such historical data at Indian 6 Point. Third, in Vermont Yankee, CHECWORKS was determined to be only a small part of the 7 FAC management program. At Indian Point, Entergy's FAC program relies integrally upon 8 CHECWORKS for determining inspection locations.
9 10 Additionally, a plant specific evaluation is prudent and necessary in order to account for unique 11 differences in flow velocities, temperatures, geometry, material, and coolant chemistry. An 12 evaluation of a nuclear power plant operator's use of CHECWORKS necessarily depends upon 13 plant specific data. Accessibility for inspections, past history with respect to the number of 14 components and frequency of wall measurements that were used in the calibration of 15 CHECWORKS, the quality of the correlation of predictions with measurements, and the number 16 of component failures from wall thinning, will necessarily vary depending on the facility. Indian 17 Point is a much larger facility in comparison to Vermont Yankee, and a different and more 18 susceptible type of reactor. As such, it is not appropriate to simply assume that the power uprate 19 that occurred at Vermont Yankee bounds the one that occurred at Indian Point.
20 21 Based on these numerous differences in the circumstances surrounding the use of 22 CHECWORKS at Vermont Yankee versus Indian Point, the findings made in the Vermont 23 Yankee proceeding are not relevant to whether CHECWORKS is adequately benchmarked for 24 use at Indian Point.
25 26 Q. Please summarize your conclusions relating to whether the CHECWORKS model is 27 adequately benchmarked in relation to the operation of Indian Point.
28 A. The available universe of CHECWORKS comparison data demonstrates very poor 29 predictive accuracy of the CHECWORKS model at Indian Point. The data further reveals no 30 signs that predictions are improving with time. The highly erratic predictive behavior of the 31 CHECWORKS model renders the CHECWORKS code useless for objective quantitative 9
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 assessments. The failure of the CHECWORKS code to produce accurate or useful results at 2 Indian Point despite decades of use unmistakably shows that the model continues to lack 3 adequate benchmarking, and will not be properly calibrated before Indian Point enters the rapidly 4 approaching proposed PE~. The lack of adequate benchmarking renders CHECWORKS an 5 ineffective tool for selecting and prioritizing piping and piping component locations at Indian 6 Point for inspections and wall thickness measurements during outages to timely detect and 7 mitigate F AC during the proposed PE~.
8 9 Q. What are the implications of CHECWORKS' poor predictive accuracy due to 10 inadequate benchmarking?
11 A. CHECWORKS predictions of wall thinning by F AC at the plant are by far too inaccurate 12 to prevent pipe wall thickness from being reduced below minimum design values. Non-13 conservative predictions affect plant safety because they fail to indicate when a component is 14 reaching a critical wall thickness and thereby result in untimely component inspection and 15 replacement. I have discussed some examples to illustrate this point in my expert report. Wear 16 rates, initial pipe wall thickness, and minimum design thickness vary widely, and even small 17 changes in the corrosion rate can result in unacceptable levels of F AC and unsafe plant 18 operations. As such, the inaccuracy of CHECWORKS is likely to allow many components to 19 operate outside allowable critical design limits during the PE~. The increase in operating life 20 from 40 to 60 years represents a significant potential for pipe wall thicknesses to fall below 21 designated minimum critical design levels during extended operations, and one would expect 22 that more and more components would become prone to failures after 40 years of service. In my 23 professional opinion, the non-conservative nature ofthe CHECWORKS code will result in 24 unacceptable F AC occurring during the PE~, which could pose serious safety issues.
25 26 In addition, conservative predictions may also affect plant safety. Entergy's documentation 27 indicates that Entergy often attributes findings of "negative time to Tcr" (i.e., a finding that a 28 component has low remaining life and should be replaced), to an "often overpredicted," meaning 29 conservative, wear value by CHECWORKS. i I However, CHECWORKS generally produces 11 See CSI, Technologies, Inc., Indian Point Unit 3 CHECWORKS SFA Model, Calculation No. 0705.100-01, Revision 2, August 2, 2011, IPEC 00238096, at Appendix K - Components with Negative Time to Tcrit.
10
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 non-conservative wear predictions. Additionally, Entergy's apparent assumption is highly 2 problematic from a safety perspective: if predictions are commonly perceived to be based on 3 conservative estimates, component replacement could be erroneously postponed potentially 4 resulting in an excessive wall thinning.
5 6 Q. Aside from assessing the degree to which CHECWORKS is benchmarked, is there 7 any other way to evaluate the effectiveness of the model at Indian Point?
8 A. Yes. To employ CHECWORKS in the face of evidence of inadequate benchmarking, it 9 would be necessary to prove that the model has an adequate track record of performance at 10 actually predicting and preventing FAC. Thus, an assessment of the number and severity of 11 actual component and pipe failures, including leaks and ruptures, that have occurred at Indian 12 Point over the years since CHECWORKS was introduced will also be indicative of whether or 13 not the program is useful and effective.
14 15 On an industry-wide basis, CHECWORKS has not been successful at predicting FAC-induced 16 wall thinning. This was recognized by the Advisory Committee on Reactor Safeguard ("ACRS")
17 Subcommittee on Thermal Hydraulics in 2005. 12 In addition, NVREG/CR-6936, PNNL 18 16186,Probabilities of Failure and Uncertainty Estimate Informationfor Passive Components-19 a Literature Review (May 2007) indicates that rate of FAC-related failures at pressurized water 20 reactors did not decrease, and actually went up, during the period oftime after CHECWORKS 21 was introduced. In fact, in the years since the nuclear power plant industry began using 22 CHECWORKS, incidents of undetected FAC, including leaks and pipe ruptures, have continued 23 to be reported at numerous nuclear power plants across the United States. While there have been 24 no recent catastrophic events resulting from FAC, this is not permission for the plant to operate 25 with pipes of unknown and unacceptable wall thickness. The "leak-before-break" concept is not 26 an excuse for operating with excessively worn-out components. There is simply no evidence to 27 suggest that CHECWORKS has been a reliable tool in the industry for predicting and preventing 28 FAC.
29 12 Statement by Dr. F. Peter Ford, transcript of January 26, 2005 meeting of the ACRS Subcommittee on Thermal Hydraulics (January 26, 2005), at 198, ADAMS Accession No. ML05040At 0613.
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Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 At Indian Point, a history of incidents demonstrate that, to date, CHECWORKS has not been 2 successful in mitigating the effects of FAC. My review of documents provided by Entergy as 3 relevant to Riverkeeper Contention TC-2 indicates that numerous leaks and reports of excessive 4 wall thinning in mechanical systems at Indian Point have been reported. 13 Entergy has 5 documented many instances where FAC has caused component wall thickness to decrease below 6 acceptable code limits. There have also been numerous FAC-induced leaks. The long and 7 consistent history of such occurrences at Indian Point clearly indicates that CHECWORKS has 8 not been successful at predicting where unacceptable FAC is likely to manifest.
9 10 While it is not possible to assess whether the number of failures has increased since the owners 11 ofIndian Point began using CHECWORKS in the 1980s due to the fact that data for years prior 12 to approximately 2000 is "unavailable," it is clear that FAC-related failures continue to occur 13 despite the use ofCHECWORKS at the plant. As Indian Point continues to age past 40 years, it 14 is reasonably foreseeable that more and more components will be prone to thinning and failure.
15 16 Entergy's own documentation irrefutably shows that there is currently no track record of 17 performance of the CHECWORKS code at Indian Point. The model has not been able to 18 preventatively detect FAC before component wall thickness dips below minimum design 19 requirements. Numerous such instances demonstrate the Entergy's use ofCHECWORKS 20 violates the ASME code, and poses tangible safety related concerns, as manifested by the various 21 leaks that have occurred at Indian Point due to undetected FAC. Entergy's failure to demonstrate 22 that the computer model has a demonstrated record of performance is further evidence that 23 CHECWORKS cannot be considered an appropriate or useful tool for managing FAC at Indian 24 Point during the PEO.
25 26 13 See Entergy Engineering Report, Operating Experience Review Report, IP-RPT-06-LRD05, Rev. 3 (2008),
IPEC00186046; Daily DER Report, DER-01-01522, April 25, 2001, IPEC00020501; Entergy Operations, Inc.,
Condition Report List, IPECOOI85743; Entergy Operations, Inc., Condition Report List, IPEC00092552; Entergy Condition Report CR-IP2-2001-10525, IPEC00092616; Entergy Condition Report CR-IP3-2006-02270, IPEC00025699.
12
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 Q. What is your understanding regarding whether Entergy has any other methods, 2 aside from CHECWORKS, for managing the aging effects ofFAC during the PEO?
3 A. I understand that Entergy has stated its position that it does not solely rely upon the use of 4 CHECWORKS to manage FAC. In particular, Entergy has stated that the F AC program at 5 Indian Point will be effective because "CHECWORKS is only one of several bases used by 6 Entergy to select and schedule in-scope components for inspection.,,14 Entergy believes that the 7 FAC program at Indian Point would be effective even without CHECWORKS because 8 inspection scope is also based on (1) actual pipe wall thickness measurements from past outages, 9 (2) industry experience related to F AC, (3) results from other plant inspection programs, and (4) 10 engineering judgment. 15 11 12Q. Please evaluate the effectiveness of these other "tools" for managing FAC during the 13 PEO.
14 A. These other tools that Entergy apparently relies upon do not demonstrate the 15 effectiveness of Entergy's FAC AMP. This is because these additional criteria cannot be viewed 16 as tools that would independently establish an accurate F AC inspection scope. In fact, they 17 largely depend upon CHECWORKS. For example, actual pipe wall thickness measurements 18 from past outages are only useful when used in combination with a predictive tool which would 19 prevent the wall thickness of a given component from being reduced to below the minimum 20 design thickness while in service. In other words, this is a required input for the use of 21 CHECWORKS. It certainly cannot be classified as a stand-alone tool for component selection.
22 In addition, if a component is initially selected to be inspected because of a CHECWORKS 23 prediction, than, necessarily, future decisions about inspection scope based on actual wall 24 thickness measurements, and wear rate trending of the actual inspection results, depend upon use 25 of the CHECWORKS computer model.
26 27 Industry and plant experience includes information about wall thinning events at Indian Point as 28 well as other plants, or changed plant parameters. These are also types of information that feed 14 Applicant's Motion for Summary Disposition of Riverkeeper Technical Contention 2 (Flow-Accelerated
_ Corrosion) (July 26,2010), ADAMS Accession No. MLl02140430, at 17.
15 Applicant's Motion for Summary Disposition of Riverkeeper Technical Contention 2 (Flow-Accelerated Corrosion) (July 26, 2010), ADAMS Accession No. MLl02140430, at 17 and Attach. 2, ~~ 39.
13
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 directly into the CHECWORKS model. These are not necessarily independent tools for 2 identifying and specifying the inspection scope during outages. Again, the usefulness of such 3 information for determining future inspections largely rests on how the CHECWORKS model 4 processes the inputs and how such information affects the model over time.
5 6 The only tool that can be considered an independent method for managing FAC is engineering 7 judgment. If actual pipe wall thickness, plant, and industry experience are not relying on 8 CHECWORKS, they can only otherwise be characterized as inputs which assist the formulation 9 of an engineering judgment. However, alone, this is not a sufficiently reliable tool for managing 10 FAC at Indian Point. The development of the CHECWORKS computer model itself stemmed 11 from the realization by the nuclear industry that engineering judgment alone was no longer 12 enough to be able to detect unacceptable and unsafe wall thinning occurrences. Engineering 13 judgment is problematic because it is intrinsically subjective. When engineering judgment is 14 identified as an independent predictive tool, a very high degree of knowledge is required by 15 those who conduct the assessment and specify the required steps for the prevention of component 16 failures. However, even with the same input data, different assumptions could lead to different 17 results because each assessment would depend heavily on the individual skill and experience of 18 the responsible engineer. In order to assess the validity of the use of engineering judgment, it is 19 imperative to fully understand how it is used and all relevant underlying assumptions informing 20 any judgment related determinations.
21 22 Entergy's FAC program fails to clearly describe what exactly engineering judgment even means 23 in relation to FAC inspections at Indian Point, and what role it actually plays in inspection scope 24 selection. Entergy has not identified any kind of systematic methodology which demonstrates 25 that engineering judgment is a separate predictive tool that would adequately manage FAC 26 related component degradation during the PE~ . In fact, based on my review of Entergy's 27 numerous documents pertaining to its FAC program at Indian Point, it is apparent that Entergy 28 fails to espouse several elements that are key to forming a sound engineering judgment about 29 FAC. First, good documentation of historical FAC assessments is critical in order to ensure that 30 engineering judgment will be based on sound knowledge of plant history. All aspects of the 31 FAC experience at the plant must be maintained, including the accuracy of past predictions, 14
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 repairs, and changes in plant operating conditions like water chemistry. At Indian Point, Entergy 2 has lost more than half of the overall amount of CHECWORKS-related data and documentation.
3 The lack of an institutional knowledge relating to FAC hinders sound engineering judgment.
4 Second, good communication between the organization that conducts analytical assessments and 5 the plant operators is essential to ensure that problems are identified early and appropriate 6 actions are taken to resolve them. At Indian Point, despite anomalies in CHECWORKS results, 7 the record of documents I reviewed did not show adequate discussions between Entergy and 8 Entergy's vendor, CSI, Technologies, Inc. about the significance of such anomalies. It is not 9 apparent that Entergy and its outside vendors have the level of communication necessary to yield 10 an adequate engineering judgment. Third, to render a sound engineering judgment, the operator 11 must have knowledge of FAC assessment methods. In other words, it is critical to understand 12 the model employed, which at Indian Point, is predominantly, ifnot solely, CHECWORKS.
13 However, this is difficult to accomplish. For example, one important way to understand the 14 validity of CHECWORKS is to observe how the model responds to changes in plant parameters.
15 While the opportunity to observes the model's response to input variables arose in 2004 and 16 2005 at Indian Point, a comparison to past performance is very limited in light of the fact that 17 Entergy has lost most ofthe documentation and data. Additionally, Entergy has indicated its 18 belief that historical documentation related to CHECWORKS is irrelevant for purposes of 19 assessing the validity of the model. This is an inappropriate attitude for purposes of 20 demonstrating the ability to exercise well-founded engineering judgment. Fourth and lastly, 21 engineering judgment requires knowledge or risks and consequences. In other words, it is 22 necessary to understand and take into account the varying safety risks posed by FAC. For 23 example, a pipe rupture of a small pipe in the service water system does not pose a risk that 24 could lead to a severe reactor accident, while a FAC-induced rupture of a main feedwater or 25 steam line pipe may lead to an uncontrolled severe accident. The documents and information 26 provided by Entergy do not reflect adequate consideration of inspection priorities of FAC-27 susceptible components relative to the safety risks posed due to a FAC-related failure.
28 29 Based on this assessment, Entergy has completely failed to demonstrate that engineering 30 judgment alone will safely manage FAC at Indian Point.
31 15
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 In turn, it is apparent that Entergy does not employ any meaningful tools that, separate and apart 2 from CHECWORKS, would sufficiently manage the aging effects of F AC at Indian Point.
3 Rather, Entergy's program for managing FAC relies heavily on the unreliable CHECWORKS 4 code.
5 6 Q. Aside from CHECWORKS and the four other tools Entergy claims inform the 7 scope of F AC inspections, has Entergy identified any other details or program elements 8 that would adequately manage F AC during the PEO?
9 A. No. Entergy's FAC program implementation documents rely heavily on the appropriate 10 use ofCHECWORKS. However, Entergy's use ofCHECWORKS has not been successful and 11 the model cannot be used to predict wall thinning during the PEO. Entergy has further failed to 12 identify any other tools that operate independent from CHECWORKS that would adequately 13 manage F AC. Accordingly, Entergy has failed to demonstrate that it has an adequate AMP to 14 manage F AC during the PEO. In order to comply with the GALL Report and NUREG-1800, 15 Standard Review Plan/or License Renewal Applications, Entergy must provide sufficient details 16 to address all relevant program elements, including the method for determining component 17 inspections, frequency of such inspections, and attendant criteria for component repair and 18 replacement. 16 Entergy cannot simply rely on procedural documents which depend upon the 19 proper use of CHECWORKS. Instead, Entergy must provide sufficient details regarding 20 inspection scope, frequency, component replacement and repair criteria, etc., to demonstrate that 21 FAC will be appropriately managed. Entergy has not done this. Its F AC AMP lacks sufficient 22 details to demonstrate that, aside from the use ofCHECWORKS, the aging effects ofFAC will 23 be adequately managed throughout the proposed PE~.
24 25 Q. You indicated earlier that it is your understanding that Entergy believes that its 26 FAC AMP complies with NRC's GALL Report. How would you assess that position?
27 A. Entergy's position that the FAC AMP at Indian Point complies with the guidance 28 contained in the GALL Report is unfounded and wrong. The GALL Report clearly indicates that 29 when a licensee uses CHECWORKS to predict wall thinning, the code must be properly 30 benchmarked before it can be used as a management tool to control F AC. In fact, though 16 See SRP-LR at § Al.2.3.
16
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 Entergy ostensibly continues to rely on the GALL Report, Revision 1, the newest version of the 2 GALL Report, Revision 2, makes this explicit. In particular, the GALL Report, Revision 2, 3 indicates that CHECWORKS is acceptable if it provides a "bounding analysis," which the report 4 defines as one that provides conservative results. 17 If the results are not conservative, 5 CHECWORKS must be re-calibrate it accordingly. 18 6
7 Notably, Entergy assigns very different criteria to the acceptability of CHECWORKS.
8 According to Entergy, as long as CHECWORKS predicts wear within its +/-50% range, which is 9 actually a much bigger margin than implied, or if the LCF is between a 0.5 and 2.5, 10 CHECWORKS is acceptable. Under Entergy's criteria, much of the results could be non-11 conservative, but the use of the code still appropriate. This is not what is contemplated by the 12 GALL Report, as clarified in the most recent version. This is also not what has been 13 contemplated in a more general way by the NRC in using analytical tools to predict plant 14 behavior. In particular, the NRC has stated that, "[i]n general, the analytical methods and codes 15 are assessed and benchmarked against measurement data, comparisons to actual nuclear plant 16 test data and research reactor measurement data. The validation and benchmarking process 17 provides the means to establish the associated biases and uncertainties. The uncertainties 18 associated with the predicted parameters and the correlations modeling the physical phenomena 19 are accounted for in the analyses.,,19 20 21 By the applicable, and appropriate standard set forth in the GALL Report, Entergy's use of 22 CHECWORKS at Indian Point is not acceptable, due to the predominately non-conservative 23 wear results achieved at Indian Point. The CHECWORKS model at Indian Point has produced 24 non-conservative results about 50% ofthe time, and many times data falls outside the broad 25 range of what Entergy considered "bounding," i.e., the +/-50% lines. The model is, therefore, 26 not properly calibrated and, according to GALL, Revision 2, it must be re-calibrated. However, 27 years of apparent attempts to recalibrate CHECWORKS at Indian Point have not resulted in any 28 improvement in the predictive capability of the code. CHECWORKS would have to be 17 GALL Report, Rev. 2 at XI MI7-1 .
18 GALL Report, Rev. 2 at XI MI7-2.
19 See Safety Evaluation by the Office of Nuclear Reactor Regulations Related to Amendment No. 229 to Facility Operating License No. DPR-28 Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc.
(Vermont Power Station, Docket No. 50-571), at § 2.8.7.1, p. 190, ADAMS Accession No. ML060050028.
17
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK-TC-2 (F AC) 1 recalibrated continuously in order to meet the standard in the GALL Report. Under such 2 circumstances, CHECWORKS ceases to be a predictive tool. Calibration of CHECWORKS on 3 a continuous basis would prevent plant operators from being able to determine whether the 4 critical thickness of any given component will be reached before the next cycle, and, in turn, 5 when component repair or replacement is necessary. Because CHECWORKS continues to 6 produce non-conservative results after decades of use, there is no way to ensure appropriate 7 calibration of the model prior to, or even during, the proposed period of extended operations at 8 Indian Point. Thus, according to the GALL Report, Revision 2, which only condones the use of 9 CHECWORKS if it produces conservative results or if can be re-calibrated to the extent results 10 are not conservative, the use of CHECWORKS at Indian Point is not acceptable. In other words, 11 Entergy's reliance upon CHECWORKS, does not demonstrate an AMP for FAC that is 12 consistent and compliant with the GALL Report.
13 14 In addition, CHECWORKS cannot be used to demonstrate that Entergy's FAC program will 15 meet applicable acceptance criteria as discussed in the GALL Report, since CHECWORKS is not 16 capable of accurately calculating the number of operating cycles remaining before a component 17 will reach the minimum allowable wall thickness.2o Notably, the inability to ensure the 18 maintenance of minimum design wall thicknesses also violate the ASME code.
19 20 Furthermore, the use ofCHECWORKS also fails to meet the guidance of the GALL Report 21 because it does not ensure that all forms ofFAC will be adequately managed. In particular, 22 while the GALL Report does not limit the obligation of licensees to manage wall thinning by 23 FAC, CHECWORKS is limited to predicting FAC caused only by electrochemical reaction. As 24 explained above, there are various other forms of flow-induced wall thinning.
25 26 Q. What, if any, safety issues does the continued operation of Indian Point for an 27 additional 20 years pose, given the inadequacy of Entergy's FAC AMP?
28 A. The delay of necessary pipe inspections, repairs, replacements, and/or other corrective 29 action due to over-dependence on a demonstrably ineffective predictive tool could result in 30 serious safety issues at Indian Point during the proposed period of extended operation. The 20 See GALL Report, Rev. 1 at XI M-62; GALL Report, Rev. 2 at XI M17-2.
18
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 operation of the plant without an adequate knowledge of the degree to which the strength of 2 various components have been degraded due to FAC-related wear poses significant safety 3 concerns.
4 5 This is particularly important when Indian Point is subject to sudden transient loads where it may 6 be too late to detect a leak and prevent a component failure. For example, the feed water 7 distribution piping ring inside the steam generators is subjected to high local velocities, (greater 8 than 20 feet per second), and turbulence. With severely degraded walls, this pipe may rupture 9 under transient loads causing damage to other structures within the steam generators, like tubes.
10 Notably, Entergy has not provided data on CHECWORKS predictions for components inside the 11 steam generators.
12 13 In addition, undetected FAC during the extended operating terms at Indian Point also poses a risk 14 of loss of coolant accidents ("LOCA") in violation of NRC's General Design Criterion ("GDC")
15 4, which requires plant structures, systems and components be able to handle such accidents, 16 including equipment failures due to circumstances outside the plant. 21 Notably, when the 17 original Indian Point probabilistic risk assessments ("PRAs") were developed, the effects of 18 aging were not included, and it was assumed that pipes were in pristine conditions. In actuality, 19 the probability of a pipe failing under a given load will be reduced when the walls have been 20 degraded.
21 22 Adequate protection is particularly important at Indian Point because recent risk assessments 23 show that Indian Point is vulnerable to core melts from earthquake loads. In fact, while the area 24 around Indian Point is susceptible to earthquakes of up to 7.0 magnitude, an NRC report from 25 August 2010 reveals that Indian Point Unit 3 has the highest risk of seismic related core damage 26 than any other nuclear power plant in the country. Another important class of accidents that 27 depends on reliable knowledge of wall thickness of various components are station blackouts, 28 SBOs. The fact that Entergy has not demonstrated that it has any reliable method of predicting 29 component wall thinning casts a doubt about Entergy's risk predictions from such accidents.
21 10 C.F.R. Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 4-Environmental and dynamic effects design bases.
19
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support ofRK-TC-2 (FAC) 1 2 Entergy should, but has failed to consider how the uncertainty related to pipe wall thickness at 3 Indian Point will affect the integrity of components under transient loads other than plant 4 transients, such as earthquakes and station blackouts. In addition, Entergy has not considered 5 how the operation of Indian Point with such large uncertainties about pipe wall thicknesses will 6 affect the likelihood of components succumbing to the effects of metal fatigue.
7 8 Pipes at Indian Point have already been reduced in strength due to almost 40 years of operation.
9 Entering an extended period of operation with no valid tool to predict wall thinning limits 10 Entergy's ability to determine the degree of pipe degradation and reduction in strength. Entergy 11 has failed to show that despite such uncertainty, Indian Point would continue to operate in 12 compliance with GDC 4, and without a severe accident occurring.
13 14 Q. Please summarize your opinions regarding whether or not Entergy has 15 demonstrated that FAC will be adequately managed during the proposed period of 16 extended operation as required by 10 C.F.R. § S4.21(c).
17 A. Based on my review of Entergy's documentation concerning FAC and other relevant 18 documents, in my professional opinion, Entergy's has failed to demonstrate that the aging effect 19 ofFAC will be adequately managed during the PEO. Entergy intends to rely on CHECWORKS, 20 which is not adequately benchmarked so as to be an effective tool for predicting FAC at Indian 21 Point during the PEO, and which has no proven track record of performance. Entergy has 22 revealed no other tools that are meaningfully independent of CHECWORKS that will assure that 23 the aging effects of F AC will be sufficiently managed during the PEO. As a result, Entergy's 24 AMP must contain, but does not, sufficient details about methods and frequency of component 25 inspections, and criteria for component repair and replacement, to demonstrate that FAC will be 26 appropriately managed throughout the entire PEO. Failure to do so poses serious safety concerns 27 due to potential FAC if Entergy continues to operate Indian Point for an additional 20 years.
28 29 Q. Does this conclude your initial testimony regarding Riverkeeper Contention TC-2?
30 A. Yes.
20
Docket Nos. 50-247-LR & 50-286-LR Pre-filed Testimony of Dr. loram HopenfeJd In support ofRJ<..-TC-2 (FAC)
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 declare under penalty of perjury that my statements in the foregoing testimony and my statement of professional qualifications are true and correct to the best of my knowledge and belief.
Executed in Accord with 10 C.F.R. § 2.304(d o Ph.D.
724 Yale Place Rockville, MD 20850 J)ecembe~, 2011 21