ML11356A391

Riverkeeper (Riv) Pre-Filed Evidentiary Hearing Exhibit RIV000034, Dr. Hopenfeld Pre-filed Testimony

ML11356A391
Person / Time
Site:	Indian Point
Issue date:	12/20/2011
From:	Hopenfeld J Riverkeeper
To:	Atomic Safety and Licensing Board Panel
SECY RAS
Shared Package
ML11356A390	List: ML11356A391 ML11356A393 ML11356A395 ML11356A396 ML11356A398 ML11356A400 ML11356A401 ML11356A402 ML11356A404 ML11356A405
References
RAS 21625, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01
Download: ML11356A391 (23)

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RIV000034 Submitted: December 22, 2011 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 ___________________________________________ )

Riverkeeper, Inc. provisionally designates the attached Testimony of Dr. Joram Hopenfeld dated December 20, 2011 as containing Confidential Proprietary Information Subject to Nondisclosure Agreement

REDACTED, PUBLIC VERSION

RIV000034 Submitted: December 22, 2011 1 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 TESTIONY OF DR. JORAM HOPENFELD REGARDING RIVERKEEPER CONTENTION TC

-1B - METAL FATIGUE On behalf of Riverkeeper, Inc. ("Riverkeeper"), Dr. Joram Hopenfeld submits the following testimony regarding Riverkeeper Contention TC

-1B. Q. Please state your name and address.

1 A. My name is Dr. Joram Hopenfeld and my business address is 1724 Yale Place, Rockville 2 Maryland, 20850.

3 4 Q. What is your educational and professional background?

5 A. I have received the following degrees from the University of California in Los Angeles: a 6 B.S. and M.S. in engineering, and a Ph.D. in mechanical engineering. I am an expert in the field 7 relating to nuclear power plant aging management. I have 45 years of professional experience in 8 the fields of nuclear safety regulation and licensing, design basis and severe accidents, thermal-9 hydraulics, material/environment interaction , corrosion, fatigue, radioactivity transport, industrial 10 instrumentation, environmental monitoring, pressurized water reactor steam generator transient 11 testing and accident analysis, design, and project management, including 18 years in the employ 12 of the U.S. Nuclear Regulatory Commission ("NRC"). My education and professional 13 experience are described in my curriculum vita, which is provided as Exhibit RIV00000 4. 14 15 Q. What is the purpose of your testimony?

16 A. The purpose of my testimony is to provide support for, and my views on, Riverkeeper's 17 Contention TC

-1B related to the aging effects of metal fatigue at Indian Point Generating Unit 18 Nos. 2 and 3 during proposed 20

-year extended operating terms. This contention was admitted 19 Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 2 by the Atomic Safety & Licensing Board ("ASLB") on November 4, 2010.

1 Riverkeeper asserts 1 that Entergy Nuclear Operations, Inc. ("Entergy"), the owner of Indian Point, has failed to 2 demonstrate that metal fatigue of reactor components will be adequately managed during the 3 proposed period s of extended operation at the plant as required by 10 C.F.R. § 54.21(c).

4 5 Q. Have you prepared a report in support of your testimony?

6 A. Yes, I prepared an expert report, provided as Exhibit RIV0000 35, which reflects my 7 analysis and opinions.

8 9 Q. What materials have you reviewed in preparation for your expert report and 10 testimony?

11 A. I have reviewed numerous documents in preparation of my expert report and testimony, 12 including the following:

all of the pleadings involving Riverkeeper Contention TC

-1B, the 13 relevant section of Entergy's License Renewal Application ("LRA"), Entergy's Amendment 2 to 14 the LRA, dated January 22, 2008, Entergy's submission to the ASLB on August 10, 2010 15 entitled, "entitled "Notification of Entergy's Submittal Regarding Completion of Commitment 16 33 for Indian Point Units 2 and 3,"

"refined" Environmental Fatigue Evaluations for Indian Point 17 Units 2 and 3 generated by Entergy's vendor Westinghouse in June 2010, relevant requests for 18 additional information from the NRC and responses thereto by Entergy concerning metal fatigue, 19 NRC Staff's Safety Evaluation Report, and Supplement 1 thereto, hundreds of documents 20 identified by Entergy as relevant to Riverkeeper's metal fatigue contentions, numerous relevant 21 NUREG reports, scientific and scholarly reports and articles, industry guidance documents and 22 reports, and other documents generated by NRC, Entergy, industry groups, and scientific 23 organizations. I have used such documents to inform me of the relevant facts and derive my 24 conclusions.

25 1Riverkeeper initially filed Contention TC

-1, followed by Amended Contention TC

-1A, concerning metal fatigue, which were admitted by the ASLB on July 31, 2008. 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-1B, which the ASLB admitted as superseding the previous contentions. 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, Ruling on Motions for Summary Disposition of NYS

-26/26A/RiverkeeperTC

-1/1A (Metal Fatigue of Reactor Components) and Motion for Leave to File New Contention NYS-26B/Riverkeeper TC

-1B) (November 4, 2010) at 29.

Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 3 1 A list of the particular documents that I reference in my expert report is included at the end of the 2 report. Those references have been provided as Exhibits NYS000 146A-146C, NYS00350

-352, 3 NYS000 354-357, NYS000361

-362, and RIV0000 36-0 58, in support of my testimony.

To the 4 best of my knowledge, these are true and accurate copies of each document that I referred to, 5 used and/or relied upon in preparing my report and this testimony.

In some cases where the 6 document was extremely long and only a small portion is relevant to my testimony, an excerpt of 7 the document is provided. If it is only an excerpt, that is noted on the cover of the Exhibit.

8 9 Q. What conclusions have you reached about metal fatigue at Indian Point?

10 A. In my professional judgment, and as I describe in more detail below and in my report, 11 Entergy has failed to demonstrate that the serious aging mechanism of metal fatigue will be 12 adequately managed throughout the proposed extended licensing terms at Indian Point. Though 13 Entergy has proffered "refined" analyses purporting to demonstrate that certain components will 14 remain within allowable acceptance criteria for metal fatigue, these analyses are flawed and 15 inaccurate.

Consideration of all relevant factors reveals that many components may become 16 susceptible to metal fatigue and pose safety risks during the proposed period s of extended 17 operation. In light of this eventuality, Entergy should have, but did not, expand the scope of 18 components to be assessed for metal fatigue. Entergy has otherwise not provided sufficient 19 details concerning an Aging Management Program ("AMP") to ensure that the degradation effect 20 of metal fatigue would be adequately handled during the license renewal periods.

21 22 Q. What is metal fatigue?

23 A. Metal fatigue is an aging phenomenon that refers to when a structure or test specimen is 24 subjected to repeated, "cyclic," loading during plant operation. Under such cyclic loading a 25 crack will be initiated and the structure will fail under stresses that are substantially lower than 26 those that cause failure under static loadings. Material composition, strain rate, temperature and 27 local water chemistry are some of the factors that contribute to fatigue of metal parts. During 28 each loading cycle, a certain fraction of the fatigue life of a component is used up depending on 29 the magnitude of the applied stress. Eventually, after the number of allowable cycles, N, the 30 structure will use all its fatigue life. The number of cycles actually experienced at any given 31 Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 4 stress amplitude, n, divided by the corresponding number of allowable cycles, N, is called the 1 usage fatigue factor ("CUF"). The maximum number of cycles that should be experienced by 2 any structure or component should always result in a CUF that does not exceed 1.0, or unity

. 3 Section III of the American Society of Mechanical Engineers ("ASME") Code provides fatigue 4 curves in air for various materials which specify the allowable number of cycles for a given 5 stress intensity. The Code requires that the CUF at any given location be maintained below one.

6 7 Q. What are the safety implications of metal fatigue?

8 A. Fatigue may result in small leaks, which, if not detected in time, could lead to a pipe 9 ruptures or other equipment failures. Fatigue may also create small cracks that propagate and 10 cause a given component to malfunction or break up and form loose parts which can interfere 11 with the safe operation of the plant. Such failures may have serious consequences to public 12 health and safety. For example, if one of the feed water distribution nozzles (J tubes) were to fail 13 from fatigue, pieces from the broken nozzle could be lodged between steam generator tubes, 14 causing the tubes to rupture and leading to a potential core melt. Components which are 15 susceptible to fatigue must, therefore, have a planned management program to ensure that the 16 plant functions efficiently and safely.

17 18 Q. Please explain how component susceptibility to metal fatigue is predicted. 19 A. Crack growth rate for a given stress intensity can be predicted using an equation that 20 includes empirical constant s that were derived from laboratory tests in air under controlled 21 conditions. However, this equation can predict crack growth reliably only as long as the 22 equation is used under the conditions that were used to calibrate the empirical constants. In 23 order to account for crack propagation in the actual reactor environment , the individual usage 24 factor in air is multiplied by a corresponding environmental correction factor, "F en." F en is the 25 ratio of the fatigue life in air at room temperature to the fatigue life in water at the local 26 temperature. The environmentally corrected CUF is expressed as CUF en. 27 28 Laboratory tests were conducted under controlled conditions at the Argonne National Laboratory 29 ("ANL"), to generate Fen factors.

Because laboratory tests were not prototypic of the reactor 30 environment, ANL provided a detailed discussion of the required adjustments to be made to the 31 Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 5 laboratory data. The ANL equations describe Fen in terms of the temperature (T), dissolved 1 oxygen (DO), sulfur content (S), and strain rate (e): Fen = f(T, DO, S, e). This equation cannot 2 be used without knowing the value of the four variables at the surface of a component at any 3 given time, during both steady state and transient operations

. 4 5 Q. Entergy's LRA contained the results of an analysis of the effects of environmentally 6 assisted fatigue on certain reactor components during the proposed period of extended 7 operation. What was the outcome of this CUF en analysis? 8 A. LRA Tables 4.3

-13 and 4.3

-14 indicated that the CUF en of four risk significant reactor 9 components would exceed unity during the period of extended operation. Due to these results, 10 Entergy committed to performing a refined fatigue analysis in order to lower the predicted CUF en 11 values to less than 1.0.

The results of this "refined" environmentally assisted fatigue ("EAF")

12 analysis, reported in revised LRA Tables 4.3

-13 and 4.3

-14 in August 2010, indicated that the 13 CUF en values for all location s evaluated were below 1.0

. 14 15 Q. Have you reviewed Entergy's refined environmentally assisted fatigue analyses?

16 A. Yes, I have reviewed two reports generated by Westinghouse, both dated June 2010, 17 pertaining to Entergy's refined fatigue analyses for Unit 2 and Unit 3, as well as other documents 18 identified by Entergy as relevant to the refined analyses.

19 20 Q. What is your opinion regarding the validity of the results of Entergy's refined 21 analyses? 22 A. Based on my review of the June 2010 metal fatigue evaluations and related documents, I 23 believe that the methodology employed to calculate Entergy's new CUF en values is highly 24 suspect, and that the validity of the results is questionable. I believe that there is a wide margin 25 of error due to many critical underlying assumptions that Entergy's refined analyses have failed 26 to properly address. Entergy's new calculations have likely grossly under

-predicted the CUF en 27 values for the components evaluated for several reasons in particular: (1) the calculations fail to 28 properly adjust laboratory data to account for the actual reactor environment in the calculation of 29 the Fen factor s to apply, (2) the calculations use incorrect values for dissolved oxygen ("DO")

30 levels in the calculation of the Fen factors to apply, (3), the calculations do not accurate ly 31 Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 6 consider heat transfer coefficients in the calculation of CUF en values, and (4) the calculations use 1 an unjustified number of transients in the calculation of CUF en values. 2 3 Q. Please explain why the use of laboratory data must be adjusted in order to 4 determine appropriate F en factors in the calculation of CUF en. 5 A. The equations for determining F en factors were derived from laboratory tests. Due the 6 significant differences that exist between the laboratory and reactor environment, there are 7 numerous uncertainties in applying the Fen equations to actual reactor components. In, 8 NUREG/CR-6909, Effect of LWR Coolant Environment on Fatigue Life of Reactor Materials , 9 ANL identifies numerous such uncertainties, which include material composition, component 10 size and geometry, surface finish, loading history, strain rate, mean stress, water chemistry, 11 dissolved oxygen levels, temperature, and flow rate.

2 Such uncertainties can have a significant 12 affect upon fatigue life and ignoring them will result in underestimated CUF en calculations. For 13 example, variations of temperature when temperature is below 150°C can reduce fatigue life by a 14 factor of two; increased water conductivity due to the presence of trace anionic impurities in the 15 coolant, which has already been documented to cause stress corrosion cracking at several nuclear 16 plants, may decrease fatigue life of austenitic stainless steels; variation in sulfide morphology at 17 a low strain rate may result in a difference by an order of magnitude in fatigue life; and surface 18 temperature fluctuations and non

-uniform temperature distributions during stratification can 19 increase the potential for crack initiation and growth, thereby reducing fatigue life. So, to 20 appropriately apply the Fen equations to actual reactor components, the user must consider all of 21 the relevant uncertainties, and the results must be adjusted to account for the varying parameters.

22 23 In NUREG/CR-6909, Effect of LWR Coolant Environment on Fatigue Life of Reactor Materials , 24 ANL specifies that appropriate bounding Fen values of 12 for stainless steel and 17 for carbon 25 and low alloy steel to account for the numerous uncertainties in using the Fen equations.

3 These 26 values are based on a review of data from different laboratory tests covering a wide range of 27 parameters. These bounding Fen factors are not necessarily conservative, and it is reasonable to 28 2 See NUREG/CR-6909 at 72.

3 See NUREG/CR-6909 at iii, 3.

Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 7 expect even higher Fen values in the actual reactor environment, especially for those components 1 that experience stratified flows and thermal striping.

2 3 Q. How did you reach the conclusion that Entergy's refined fatigue evaluations fail to 4 properly adjust laboratory data to account for the actual reactor environment?

5 A. Entergy's calculations of Fen factors, used to determine the refined CUF en values, relied 6 on equations that were derived from laboratory tests. Entergy's "refined" EAF analyses have not 7 adequately evaluated the numerous uncertainties associated with determining acceptable Fen 8 values, or, in the alterative, applied bounding Fen values to conservatively ensure that such 9 uncertainties are accounted for. Entergy's calculations did not specify the input parameters that 10 were actually used for each transient. Westinghouse provided no indication at all that the 11 numerous uncertainties associated with applying the Fen equations to actual reactor components 12 were considered in the calculations of the CUF en values. For example, Entergy's calculations do 13 not correct the Fen value to account for the fact that when temperature is below 150°C, fatigue 14 life could be reduced by a factor of two. Nor is there any evidence that Entergy considered the 15 presence of trace impurities on water conductivity, which reduces fatigue life. Instead, Entergy 16 used unrealistically low F en values that are, as yet, not justified in light of the wide range of 17 parameters unaccounted for.

18 19 Q. Please explain how Entergy's failure to properly adjust laboratory data to the actual 20 reactor environment affect s the results of the refined EAF analysis.

21 A. Entergy's calculation of fatigue life for selected components would be significantly 22 affected if all relevant uncertainties were actually considered. Entergy's failure to properly 23 account for the numerous uncertainties inherent in determining the appropriate Fen value from 24 using equations derived from laboratory tests has resulted in calculations that underestimate the 25 CUF en for the analyzed components.

Entergy's assessment that no CUF en for the evaluated 26 components exceeds unity remains unsubstantiated. The use of the bounding F en values 27 recommended by ANL, which represent far more realistic values than most of those calculated 28 and used by Entergy, would increase the CUF en values beyond unity for eight of the components 29 analyzed, as I calculated in my expert report.

30 31 Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 8 Q. Entergy has previously claimed that using the bounding Fen values recommended in 1 NUREG/CR-6909 "would actually yield less conservative CUF en values, because the ASME 2 Code design air curves for carbon steel and low

-alloy steels contained in air [which is what 3 Entergy considered] . . . are more conservative than the newer air curves in NUREG/CR

-4 6909."4 How would you respond to this assertion?

5 A. Entergy's apparent position that its reliance on the ASME code curves automatically 6 results in a more conservative CUF en value than would be reached if Entergy had individually 7 evaluated all the uncertainties identified by ANL, is completely unfounded and simply wrong.

8 While the current ASME code does incorporate a fatigue margin design of 2 on stress and 20 on 9 cycles for carbon and low alloy steels in air, this conservatism was not intended to provide a 10 margin of safety. The ASME code curves are valid in relation to fatigue life in air, and do not 11 reflect the specific effects of the reactor coolant environment. In contrast, NUREG/CR-6909 12 provides guidance on appropriate F en factors to account for uncertainties inherent in CUF en due 13 to the reactor, i.e.

water, environment. The uncertainties discussed in NUREG/CR

-6909 are not 14 reflected anywhere in the current ASME code. The difference between the NUREG/CR

-6909 air 15 curve and the ASME code air curve is small in comparison to the many uncertainties associated 16 with the actual reactor environment that must be considered to calculate an appropriate F en. 17 18 Q. Please explain how DO levels affect the determination of appropriate F en values for 19 calculating CUF en. 20 A. One of the largest uncertainties in determining appropriate F en values is the concentration 21 of dissolved oxygen ("DO") in the water at the surface of each component during the transient.

22 The F en varies exponentially with the DO level, and is therefore sensitive to uncertainties in DO 23 concentrations. The equations for determining F en were experimentally derived under conditions 24 where the temperature and DO at the surface of the specimen were known. In contrast, in a 25 reactor plant, the DO in many cases is unknown. This is in particularly true during startup and 26 shutdown transients. During these transient the oxygen content at the surface of the component 27 varies significantly due to oxygen incursions from external sources and because DO has a 28 negative solubility coefficient in water. The level of DO, therefore, increases significantly 29 4 Declaration of Nelson F. Azevedo in Support of Applicant's Motion for Summary Disposition of Contentions NYS-26/26A and Riverkeeper TC

-1/1A (August 20, 2010), at ¶ 48.

Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 9 during shutdown transients. During startup transients, DO will be at a maximum at the 1 beginning of the transient and then decrease towards its steady state value. Oxygen excursions 2 occur during heatups. Oxygen concentrations can vary with changes in temperature by more 3 than an order of magnitude in comparison to oxygen levels during normal operating conditions.

4 DO levels during transients are not measured at the surface of the reactor components at 5 operating plants. As a result, the actual DO levels, and resulting F en, are subject to uncertainties.

6 For example, an uncertainty of five in DO levels at the surface of a given component could lead 7 to under-predicting the F en by a factor of five at a minimum. This uncertainty can be accounted 8 for by using bounding oxygen values during each transient. These values are different for 9 different materials. The bounding oxygen value for stainless steel is the lowest value of oxygen 10 during the transient, while the opposite is true for carbon and low allo y steels. 11 12 Due to the uncertainties related to DO levels during transients, ANL, as well as the Electric 13 Power Research Institute ("EPRI"), specifically instruct that in determinin g F en, DO can be 14 conservatively taken as the maximum value for the transient.

5 ANL indicates that for carbon and 15 low-alloy steels and austenitic stainless steel, values of 0.4 ppm and 0.05 ppm, respectively, can 16 be used to perform a conservative evaluation.

6 17 18 Q. How did you reach the conclusion that Entergy's refined fatigue evaluations use 19 incorrect values for DO levels in the calculation of the Fen factors applied?

20 A. Entergy's refined EAF calculations did not utilize the recommended bounding DO 21 values, and instead used unrealistically low steady state DO values of less than 0.05 parts per 22 million (ppm) during all transients. Using low oxygen values that exist during normal operating 23 conditions does not reflect those that exist during transients. As such, the F en equations were 24 used incorrectly with respect to the oxygen input. Entergy should have followed the 25 recommendation to apply a DO level of 0.4 parts per million for carbon and low alloy steel 26 components, or increase the Fen factor applied to such components to reflect the uncertainties 27 related to DO.

28 5 NUREG/CR-6583, ANL-97/18, Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels at 78; EPRI's Materials Reliability Program: Guidelines for Addressing Fatigue Environmental Effects in a License Renewal Application, MRP

-47 at 4-19. 6 NUREG/CR-6909 at A-5.

Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 10 1 Q. Has Entergy and/or Westinghouse provided any explanation regarding the 2 treatment of DO levels in the refined EAF analyses?

3 A. Entergy has not provided an adequate rational e for failing to abide by well

-founded 4 recommendations. Entergy's refined calculations do not discuss any tools that are used at Indian 5 Point to determine oxygen levels in the plant at the surfaces of its various components, and do 6 not acknowledge that DO levels represent a major uncertainty in the calculation of Fen.

7 8 In fact, Westinghouse's discussion regarding the uncertainties of DO levels appears limited to 9 one statement, which concludes that during periods of heatup when DO may be higher than 10 0.05ppm, the temperature will be less than 150°C such that T*=0.

7 This does not justify 11 ignoring DO variations during transients because the fact that T*=0 is not an indication that DO 12 is not present or that it ceases to effect fatigue life. As I already mentioned, fatigue life may 13 decrease by a factor of two when the temperature is less than 150°C. In addition, the heating 14 period below 150

°C (300°F) does not represent or bound all transients. During startup and 15 shutdown transient, for example, the temperature varies approximately from 150°F to 600° F. 16 Entergy's use of T*=0, therefore, improperly discounts the presence of oxygen and the effect it 17 will have on fatigue life. The approach espoused by Westinghouse does not reflect the 18 recommendations contained in NUREG/CR

-6909 to use an average temperature experienced 19 during the transients to determine the Fen.

8 20 21 Q. Please explain how Entergy's failure to use appropriate values for DO levels in the 22 calculation of Fen, affects the results of the refined EAF analysis.

23 A. Because the refined EAF analyses do no properly consider DO in the calculation of F en 24 values, Entergy's calculation of the CUF en values for carbon and low alloy steels are grossly 25 underestimated. The flawed methodology employed strongly suggests that at least some of the 26 CUF en values calculated would exceed unity, even assuming DO was the only questionabl e 27 parameter. In fact, the CUF en values are under

-predicted by at least a factor of four due to the 28 7 Environmental Fatigue Evaluation for Indian Point Unit 2, WCAP

-17199-P, Revision 0 (Westinghouse, June 2010), IPECPROP00056486, a p.5

-24; Environmental Fatigue Evaluation for Indian Point Unit 3, WCAP

-17200-P, Revision 0 (Westinghouse, June 2010), IPECPROP00056577, at p.5

-24. 8 NUREG/CR-6909 at 40.

Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 11 use of low oxygen values alone. Increasing the Fen factor to reflect the uncertainties related to 1 DO would result in every component evaluated, except the vessel shell, to either closely reach or 2 exceed unity. Notably, EPRI's Materials Reliability Program: Guidelines for Addressing Fatigue 3 Environmental Effects in a License Renewal Application, MRP-47, Rev. 1, calculates F en values 4 as high as 130 at high DO levels, which is two orders of magnitude higher than those calculated 5 by Entergy.

9 6 7 By failing to appropriately account for DO levels during each transient in the calculation of F en , 8 it is impossible to conclude that the claimed CUF en values, which Westinghouse and Entergy's 9 "refined" analysis purport to predict to a ten

-thousandth of a decimal point, are accurate.

10 11 Q. Can you please explain what a heat transfer coefficient is in relation to determining 12 CUF en. 13 A. A heat transfer coefficient is a parameter used to determine the rate of heat transfer 14 during transients, in order to calculate thermal stress and its impact on fatigue life. Heat transfer 15 is a major factor in the determination of CUF en because it controls the cyclic thermal stresses 16 during transients. Thermal stresses arise when there is a change in the local fluid temperature 17 like during heat

-ups or cool

-downs or due to local mixing of hot and cold fluids. Failures result 18 from either low stress at high cycle or high stress at low cycle. Stresses from thermal fatigue 19 pose a serious risk of damage and have caused cracks in pipes and leakage at several nuclear 20 reactors. This trend is expected to increase with time

. 21 22 To calculate the rate at which heat is transferred to the reactor component surface during a 23 transient (or in other words the temperature distribution of a component during a transient) in 24 order to determine thermal stress, thermal

-hydraulic computer codes are used with plant data to 25 perform a heat transfer analysis. Heat transfer coefficients (h), water temperature, cycling 26 period, and interface motion are all important inputs to this heat transfer analysis, and the 27 consequent determination of the CUF e n values. The CUF en value will vary greatly depending on 28 the inputs used to perform the heat transfer analysis. The heat transfer coefficient h is the most 29 important parameter in this regard. The heat transfer coefficient is commonly expressed in terms 30 9 EPRI, MPR 47 at 4

-22.

Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 12 of geometry (G), fluid properties (P), flow rates (Q), and temperature difference between the 1 = f (G, P, Q, 2 parameter and has been measured and determined for many different geometries, flow rates and 3 rates of temperature change, and is known for well

-defined, controlled conditions. However, the 4 local flow at the surface of many reactor components during transients is not well defined and, 5 therefore, approximations and assumptions are required in calculating the proper h for a given set 6 of conditions. Such approximations lead to uncertainties in the CUF en because uncertainties in h 7 directly impact the errors in the calculated stress. Typical variations in h could increase stress b y 8 a factor of 2. Increase in turbulence due to local discontinuities, and increase in the rate of the 9 local temperature change increases h. Increase in h increases the corresponding stress and 10 reduces fatigue life. For example, h along nozzles and bends varies in intensity because of the 11 large variation in turbulence along their surface. This leads to non

-uniform heat loads and 12 introduces larger uncertainty in the stress distribution in comparison to simpler flow 13 configurations. Stratified flow in the pressurizer surge line, is another example where non

-14 uniform heat loads exist. Such flows occur when a warmer fluid flows on top of a cooler fluid, 15 with a temperature difference between the two fluids as high as 350°F. Instabilities at the 16 interface between the two fluids are known to produce high frequency temperature fluctuations 17 on the surface of the component, with the potential for accelerating crack initiation and growth.

18 Another factor that would lead to non

-uniform stress distributions is preferential wall wear due to 19 flow accelerate d corrosion ("FAC") in low alloy steel components. For example, my review of 20 Entergy ultrasonic examination reports indicates that in components where flow is not fully 21 developed, component wall thickness can vary by more than 400% at Indian Point

. 22 23 Q. How did you reach the conclusion that Entergy's refined fatigue evaluations do not 24 accurately consider heat transfer coefficients in the calculation of the CUF en values? 25 A. For at least certain components, it is apparent that Entergy's calculations employed 26 unrealistically low heat transfer coefficients in the determination of CUF en. A more realistic 27 selection for this key parameter indicates that the CUF en is significantly larger than predicted by 28 Entergy. For other components, Entergy has failed to provide sufficient information to justify 29 that heat transfer was appropriately considered, also casting doubt on the accuracy of Entergy's 30 "refined" CUF en calculations.

31 Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 13 1 Q. Can you please explain how you reached the conclusion that Entergy used 2 unrealistically low heat transfer coefficients for the calculation of CUF en for certain 3 components.

4 A. I reviewed Entergy documentation describing the heat transfer calculations performed in 5 relation to the pressurizer surge line, a component that is highly vulnerable to fatigue failures.

10 6 My review indicates that Entergy applied two different equations for determining the heat 7 transfer coefficient for the surge line, the well

-known Dittus

-Boelter equation for steady state 8 forced convection fully developed turbulent flow in a straight pipe, where h = 0.023 k/D Re .8 9 Pr.3 and an equation for a laminar free convection flow in a horizontal pipe, where h = 0.56 k/L 10 Ra .25. However, these equations were derived for entirely different flow conditions than those 11 in the surge line, and using these equations to calculate stresses would underestimate the rate of 12 heat transfer. This is due to a few reasons. First, because of discontinuities in the nozzle and 13 bends, the local turbulence in these regions is much higher than in the straight sections, 14 increasing the corresponding value of h depending on the intensity of turbulence as determined 15 by the local Reynolds number. Second, the latter equation was obtained for the case where the 16 fluid, in a heated horizontal pipe or a plate, was driven by buoyancy in the laminar flow region.

17 In the surge line, the physical situation is entirely different, as the hot and the cold fluids move 18 with respect to each other at low velocities. The flow at the interface of the two fluids is unstable 19 causing mixing at the interface with intensity depending on the relative velocities of the two 20 fluids and their temperatures. The instabilities at the interface would trigger early transition of 21 the laminar boundary layer to turbulence at the pipe surface. Consequently, the turbulence 22 would increase the heat transfer rate to the surface in comparison to what would been predicted 23 by the equation used. Third, both of the equations used represent average values, and not local 24 conditions. For example, the latter equation does not account for the temperature fluctuations at 25 the surface which result from the instabilities, in stratified flows, at the interface between the hot 26 and the cold fluid streams in the pipe, which makes it imperative that local conditions be take n 27 into account in determining the value of h.

Therefore, Entergy's heat transfer calculations for the 28 pressurizer surge line do not appear to have considered the potential for significant reduction in 29 10 See WCAP-14950, Westinghouse, Mitigation and Evaluation of Pressurizer Insurge/Outsurge Transients (February 1998), IPECPROP00059247.

Docket Nos. 50

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-TC-1B 14 fatigue life due to temperature fluctuation (also known as thermal striping)

. Stratified flow is a 1 major source of error in calculating CUF en. In fact, when the flow is stratified, an uncertainly of 2 at least two in the heat transfer coefficient can be expected. While the accuracy of the heat 3 transfer coefficient for natural convections under controlled conditions is on the order of +/

-30%, 4 for a stratified flow with an unstable mixing at the interface between moving hot and cold fluid, 5 higher uncertainties can be expected. Entergy's calculations do not appear to reflect the 6 extensive amount of data that has been generated in the last decade to resolve fatigue issues in 7 Pressurized Water Reactors ("PWR")

during stratification.

For example, using an advanced 8 mathematical technique, it has been shown that the CUF of some RCS components would 9 exceed unity when thermal stratification stresses were included in the analysis.

11 10 11 Based on these considerations, it is my opinion that the reliance on the two equations used was 12 inappropriate and resulted in non

-conservative assumptions in the selection of the heat transfer 13 coefficient for the pressurizer surge line. Because of the complex nature of the flow in natural 14 convection, the uncertainty in heat transfer coefficients for such flows, including stratified flows, 15 may exceed 30% even for relatively simple geometries. Entergy's application of the equations 16 used could under predict the heat transfer to the surge line by 30% to 200%. This, in turn, would 17 have a significant affect upon the value of CUF en. Entergy's "refined" CUFen values for the 18 surge lines in Indian Point Units 2 and 3 are 0.822 and 0.594 respectively.

More realistic heat 19 transfer calculations alone would have increased the CUF en value by as much as a factor of two.

20 An uncertainty of only 30% in the heat transfer coefficients would cause the corresponding 21 CUF en for Indian Point Unit 2 to exceed unity.

22 23 I also reviewed Entergy documentation related to the heat transfer analysis for the inlet and 24 outlet reactor vessel nozzles.

12 The temperature distributions for these components were based 25 on over 40

-year old Combustion Engineering calculations. These calculations did not use a 26 11 Kwang-Chu Kim et.al., Thermal fatigue estimation due to thermal stratification in the RCS branch line using one

-way FSI scheme, Journal of Mechanical Science and Technology, 22 (2008) 2218

-2227. 12 Combustion Engineering, Inc., Nuclear Components Engineering Department, C.E. Contract No. 17765, "Analytical Report for Indian Point Reactor Vessel Unit No. 2," C.R. Crockrell and J. C. Lowry, CENC

-1110 (April 22, 1968); Combustion Engineering, Inc., Nuclear Components Engineering Department, C.E. Contract No. 3366, "Analytical Report for Indian Point Reactor Vessel Unit No. 3," C.R. Crockrell and J. C. Lowry, CENC

-1122 (June 1969).

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-TC-1B 15 finite element analysis and were based on a simplified 2

-D model where the heat transfer 1 coefficient was assumed constant along the flow and thermal properties were taken as 2 independent of the temperature. The calculations were based on "as installed" nozzle 3 dimensions. Due to nozzle geometry, the heat transfer is not uniform along the nozzle and can 4 vary by 20%

-30% depending on flow velocity location along the nozzle and flow direction.

5 Such variation would result in axial thermal stress, which is not included in the 2

-D analysis.

6 Additionally, it is apparent that this analysis neglected to account for the observed fact that low 7 alloy steels are subjected to wall thinning due to FAC. The reduction in wall thickness after 60 8 years of operation is expected to reduce fatigue life. Based on these considerations, it is apparent 9 that Entergy's calculations used inappropriate heat transfer coefficients in the calculation of 10 CUFen for the inlet and outlet reactor vessel nozzles 11 12 Q. Can you please explain how you reached the conclusion that Entergy failed to justify 13 that heat transfer was appropriately considered in the calculation of CUF en for certain 14 components.

15 A. For the other components evaluated in Entergy's "refined" CUF en analyses, meaning the 16 RCS charging system, RCS injection nozzles, and RHR class 1 piping, Entergy did not provide 17 the actual equations employed to determine the heat transfer coefficients. Instead, Entergy's 18 calculations only indicate that certain therma l-hydraulic phenomena, like stratification, were 19 taken into account, and that a thermal hydraulic model in the WESTEMS program was 20 employed. Though Entergy has purported to provide sufficient information concerning the 21 calculation of heat transfer coefficients for the rest of the components evaluated, Entergy's 22 documentation and analyses do not specify the heat transfer coefficients used, or how h was 23 determined, for these components. To assess the uncertainty of h, it is imperative to know the 24 component geometry, the piping geometry upstream of the component, the flow velocities, and 25 the corresponding expressions for h, none of which are specified by Entergy for the components 26 at issue. Because complex computer code models and empirical equations were used by Entergy 27 to predict thermal fatigue life, the validity of the underlying assumptions must be known.

Given 28 the uncertainties associated with determining the heat transfer coefficient, h, it is imperative to 29 ensure that the methodology and assumptions employed were adequate to account for such 30 uncertainties. Entergy's calculations have failed to do this. Without an understanding of the 31 Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 16 values of h and the assumptions used to arrive at such values, the methodology employed by 1 Entergy to re

-calculate C UF en for the three relevant components remains questionable. Thus, 2 based on the information relied upon by Entergy to support the "refined" evaluation, it is simply 3 impossible to conclude that the new CUF en values for these other components are accurate.

4 5 Q. How did you reach the conclusion that Entergy's refined fatigue evaluations use an 6 unjustified number of transients in the calculation of the CUF en values? 7 A. To evaluate the remaining fatigue life of a given component, it is necessary to consider 8 the past as well as future loading during all transients. In other words the fatigue status of the 9 component must be known from the time it was installed to the time it is removed from service.

10 An accurate number of past and anticipated future transients is essential to the calculation of the 11 CUF en values. However, Entergy did not adequately consider either past or future transients at 12 Indian Point. Because the actual number of transients during the proposed period of extended 13 operation is not known, it was necessary for certain assumptions to be made in obtaining this 14 number. Entergy's documentation concerning the "refined" EAF analyses fails to demonstrate 15 that Entergy employed appropriate assumptions in obtaining the number of transients.

16 17 To assess the severity of past transients, each transient that has occurred must be described to 18 determine its contribution to the CUF en. Yet considering past transients can be difficult because 19 of the fact that degradation of some components was not included in the fatigue analysis when 20 nuclear power plants were originally built. For example, the pressurizer surge line falls into this 21 category. Historical records in such cases are incomplete and insufficient to provide adequate 22 inputs for the number of heat

-up and cool

-down transients, stratification frequency, and system 23 24 from my review of Entergy's documents that plant data for Indian Point Units 2 and 3 prior to 25 1993 was not available and Entergy relied on data from four other plants in considering past 26 transients. However, Entergy did not show or justify that the number of past transients were 27 developed appropriately based on such other data. It would appear that Entergy has relied upon 28 some undisclosed model in using other plant data in developing the number of transients for the 29 new calculations. For example, stratification is a plant

-specific phenomenon that generally 30 requires 3D modeling and Entergy has not demonstrated how the number of transients 31 Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 17 concerning stratification was developed for the pressurized surge line by virtue of using other 1 plant data.

2 3 In relation to future transients, to determine the number of cycles to the end of the 60

-year plant 4 life, it is my understanding from reviewing Entergy's documents that Entergy used the actual 5 number of plant transient cycles to a given date and a straight

-line extrapolation to determine the 6 remaining number of transients.

Generally, the assumption of using a past number of transients 7 as a predictor of future transients must be based on a component and system failure analysis 8 which Entergy has apparently not performed or provided. In fact, Entergy has failed to specify 9 in any respect its underlying assumptions that would justify the use of such an extrapolation.

10 Instead, Entergy merely indicated that the number of transients selected was conservative 11 because it was higher than it was originally anticipated when the plant was designed. Entergy 12 did not quantify how many transients would be considered "conservative." Additionally, 13 Entergy's approach incorrectly assumes that all relevant transients were originally considered, 14 which is not the case. For example, stratification transients were not included when Indian Point 15 was designed. Entergy's explanation simply does not demonstrate that an appropriate number of 16 transients were considered.

17 18 Justification of an appropriate number of future transients is critical in light of the fact that the 19 useful life of most engineering components and structures follow a "bathtub curve" whereby 20 component failure toward the end of operating life will occur at a very high frequency. Because 21 Indian Point will be entering an extended period of operation, Entergy must justify its use of the 22 straight-line extrapolation for the number of transients it assumed in calculating the CUF ens. In 23 addition, even if using a straight

-line extrapolation is justified, Indian Point

-specific data relating 24 to past transients was apparently not available, which would make the anticipated number of 25 transients erroneous

. 26 27 The failure to employ appropriate assumptions in obtaining the number of transients casts doubt 28 on the CUF en methodology used, and, in turn, the accuracy of Entergy's refined EAF 29 calculations.

30 31 Docket Nos. 50

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-TC-1B 18 Q. Please summarize your conclusions about Entergy's refined EAF analyses.

1 A. Informed judgment is required in order to determine whether Entergy has adequately 2 accounted for the significant degree of uncertainty associated with calculating CUF en. In other 3 words, it is critical to understand all of the underlying assumptions employed to arrive at the new 4 EAF calculations. However, Entergy has failed to properly account for all relevant uncertainties 5 in determining the revised CUF en values. While in certain respects, it is clear that Entergy 6 incorrectly applied a given parameter, such as DO levels and heat transfer coefficients for certain 7 components, in other respects, Entergy's analysis fails provide sufficient technical details to 8 understand how a given parameter was accounted for in the resulting CUF en, such as heat 9 transfer coefficients for the majority of components evaluated, and the number of transients 10 chosen. Many assumptions are not even explained, let alone justified, and numerous potential 11 uncertainties are largely not presented and far from adequately addressed.

12 13 Entergy's failure to sufficiently account for all relevant parameters has resulted in predictions 14 that are non

-conservative. Given the large uncertainties in the input parameters and other 15 assumptions used to generate the revised metal fatigue calculations, the methodology employed 16 by Entergy suggests the likelihood of a wide margin of error, and the detrimental effects of the 17 environment on fatigue strength of the components evaluated are likely grossly underestimated.

18 In fact, many of the revised CUF en values remain very close to unity. The fact that Entergy's 19 refined CUF en values are reported to the fifth significance figure (i.e. to a ten

-thousandth of a 20 decimal point), with several just a hair below unity

, clearly shows that Entergy does not 21 appreciate the uncertainties inherent in calculating the CUF en values. With a margin of error to 22 account for varying input data and other undisclosed assumptions, such numbers could be 23 considerably higher than the 1.0 regulatory threshold. In any event, without an error analysis, 24 the claimed high degree of accuracy of the results remains questionable at best. Entergy 25 repeatedly applies the term "bounding" to its analyses and results, implying that such results are 26 conservative, and that no error analysis is necessary. However, there is no explanation to show 27 that the results are in any respect bounding or conservative.

28 29 Docket Nos. 50

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-TC-1B 19 Based on all these considerations, it is my professional opinion that Entergy's refined CUF en 1 calculations cannot be used as the basis for concluding that the aging effects of metal fatigue will 2 be adequately managed at Indian Point during the PEO.

3 4 Q. In your opinion, does Entergy otherwise have an adequate program for managing 5 the effects of metal fatigue during the proposed period of extended operation?

6 A. No. Entergy has chosen to rely upon its refined EAF analyses to demonstrate that metal 7 fatigue will be managed throughout the period of extended operation. This is not adequate for 8 several reasons.

9 10 First, for the foregoing reasons contained in this testimo ny, Entergy's new calculations do not 11 demonstrate that the CUF en for the components evaluated will not exceed unity during the 12 proposed extended licensing terms.

13 14 Second, in order for Entergy to have an effective AMP to monitor for metal fatigue, it must 15 expand the scope of the fatigue analysis beyond simply representative components, to identify 16 other components whose CUF en may be greater than 1.0. This is necessary because Entergy's 17 initial findings presented in in Tables 4.3

-13 and 4.3

-14 of the April 2007 LRA indicated that the 18 CUF en values for various components exceeded the regulatory threshold of 1.0, and under such 19 circumstances, applicable regulatory and industry guidance required Entergy to identify 20 additional reactor locations for potential high susceptibility to metal fatigue. Entergy's refined 21 EAF evaluations did not expand the scope of components analyzed, but rather only assessed 22 those locations identified in NUREG/CR

-6260, Application of NUREG/CR

-5999 Interim Fatigue 23 Curves to Selected Nuclear Power Plant Components (1995). In addition, the most recent 24 version of NUREG

-1801, Generic Aging Lessons Learned (GALL) Report, Revision 2 specifies 25 that the sample set for fatigue calculations should include additional plant

-specific component 26 locations if they may be more limiting than those considered in NUREG/CR

-6260. Entergy's 27 fatigue analyses to date demonstrate that the components analyzed will likely exceed unity, and 28 were, therefore, not necessarily the most limiting locations and bounding for the entire plant. In 29 fact, NRC Staff has now conceded this point, as demonstrated by a request for information 30 issued to Entergy seeking confirmation and justification that the locations selected for EAF 31 Docket Nos. 50

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-TC-1B 20 analyses consisted of the most limiting locations for the plant. However, in a supplement to the 1 Safety Evaluation Report pertaining to the Indian Point license renewal proceeding, NRC Staff 2 accepted only a vague commitment from Entergy to determine at some point in the future 3 whether the locations assessed were the most limiting for Indian Point. This was not appropriate 4 and Entergy's failure to actually confirm and justify bounding and limiting locations for Indian 5 Point leaves Entergy's AMP insufficient, as it does not comply with the directive in the GALL 6 Report or demonstrate that metal fatigue will be appropriately monitored, managed and corrected 7 during the period of extended operation. Entergy must identify the locations that may be more 8 limiting, and which will be the subject of CUF en calculations, now, and not just articulate a plan 9 to determine such locations later. Such a determination is complex and would require 10 consideration of thermal striping during stratification, an assessment of experience at other 11 PWRs, and identifying and ranking of all components susceptible to thermal fatigue in terms of 12 numerous key parameters, including the ratios of the local heat transfer coefficient and the local 13 14 velocities, number of transients, magnitude and cycling frequency of surface temperatures and 15 loads, and surface discontinuities, and flow discontinuities in each component. Entergy has not 16 provided any information about how the analysis to determine the most limiting locations as 17 Indian Point will be performed to allow for meaningfully comment upon the adequacy of the 18 analysis. 19 20 Lastly, Entergy does not have an adequate AMP for metal fatigue because, in the absence of a 21 reliable and complete assessment of CUF en values for susceptible plant components, Entergy has 22 failed to define specific criteria concerning component inspection, monitoring, repair, and 23 replacement. Entergy's plans for correcting metal fatigue related degradation depend initially 24 upon calculating the vulnerability of plant components. Entergy intends to rely upon future 25 CUF en calculations throughout the period of extended operation to manage metal fatigue.

26 Entergy's calculations are meant to signify when components require inspection, monitori ng, 27 repair, or replacement, and, according to Entergy, will trigger when such actions are taken. As 28 such, the validity of Entergy's monitoring program depends upon the accuracy of the 29 calculations of the CUF en. When a fatigue monitoring program is entirely based on a predictive 30 analysis and not on actual measurements, and the analysis is flawed, the monitoring program is 31 Docket Nos. 50

-247-LR & 50-286-LR Pre-filed Testimony of Dr. Joram Hopenfeld In support of RK

-TC-1B 21 invalid. Thus, Entergy's flawed methodology for calculating CUF en, which Entergy ostensibly 1 intends to employ throughout the period of extended operation, as well as Entergy's failure to 2 expand the scope of components to be assessed, renders Entergy's vague commitments to 3 inspect, repair, and replace affected locations insufficient to ensure proper management of metal 4 fatigue during the proposed PEO. Without accurate metal fatigue calculations to properly guide 5 Entergy's aging management efforts, Entergy has failed to define specific criteria to assure that 6 susceptible components are inspected, monitored, repaired, or replaced in a timely manner. 7 Once components with high CUF en values have been properly identified, Entergy must describe 8 a fatigue management plan for each such component that should, at a minimum, rank 9 components with respect to their consequences of failure, establish criteria for repair versus 10 defect monitoring, and establish criteria for the frequency of the inspection (considering, for 11 example defect size changes and uncertainties in the stress analysis and instrumentation), and 12 allow for independent and impartial reviews of scope and frequency of inspection. Entergy has 13 not done this.

14 15 Q. Please summarize your opinions regarding whether or not Entergy has 16 demonstrated that metal fatigue of reactor components will be adequately managed during 17 the proposed periods of extended operation as required by 10 C.F.R. § 54.21(c). 18 A. Based on my review of Entergy's submissions concerning metal fatigue and other 19 relevant documents, in my professional opinion, Entergy has failed to demonstrate that the aging 20 effects of metal fatigue will be adequately managed for the period of extended operation, and 21 has, thus, failed to comply with 10 C.F.R. § 54.21(c) or regulatory guidance, including the GALL 22 Report. Entergy's refined EAF CUF en analyse s did not account for numerous critical factors and 23 as a result have a wide margin of error and have likely underestimated the refined CUF en values. 24 As many of the values derived by Entergy's new evaluation are very close to unity, the 25 regulatory threshold, it is highly likely that if Entergy had accurately considered all relevant 26 factors, many of the CUF en values would, in actuality, exceed 1.0. Entergy also improperly 27 failed confirm that the components evaluated represent the most limiting locations providing a 28 bounding analysis, and failed to expand the scope of components to be subject to a CUF en 29 analysis accordingly. And lastly, in light of the insufficiency of Entergy's CUFen analyses, 30 Entergy has otherwise failed to provide any details concerning the inspection, monitoring, repair, 31 and replacement to ensure that the degradation effects of metal fatigue would be sufficiently 32 handled. 33 34 Q. Does this conclude your initial testimony regarding Riverkeeper Contention TC

-1B? 35 A. Yes. 36

l/:44 In the Matter of NUVt.l'<t-LU UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD Ent:ergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3) ) ) ) ) ) ) Docket Nos. 50-247-LR and 50-286-LR DECLARATION OF DR. JORAM HOPENFELD I, Joram Hopenfeld, do hereby declare under penalty of perjwy 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) December_,2D 2011