Text

Official Exhibit - RIV000144-PUB-00-BD01 - Supplemental Report of Dr. Joram Hopenfeld in Support of Contention NYS-26B/RK-TC-1B and Amended Contention NYS-38/RK-TC-5 (Public/Redacted)
	ML15331A266
Person / Time
Site:	Indian Point
Issue date:	06/09/2015
From:	; Riverkeeper
To:	; Atomic Safety and Licensing Board Panel
	SECY RAS
References
	RAS 27919, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
	Download: ML15331A266 (39)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD

x In re:

Docket Nos. 50-247-LR; 50-286-LR License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 Entergy Nuclear Indian Point 3, LLC, and Entergy Nuclear Operations, Inc.

June 9, 2015

x Riverkeeper, Inc. provisionally designates the attached Report of Dr. Joram Hopenfeld dated June 8, 2015 as containing Entergy/Westinghouse Designated Confidential Proprietary Information Subject to Nondisclosure Agreement REDACTED, PUBLIC VERSION RIV000144 Date Submitted: June 9, 2015 United States Nuclear Regulatory Commission Official Hearing Exhibit In the Matter of:

Entergy Nuclear Operations, Inc.

(Indian Point Nuclear Generating Units 2 and 3)

ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #:

Identified:

Admitted:

Withdrawn:

Rejected:

Stricken:

Other:

RIV000144-PUB-00-BD01 11/5/2015 11/5/2015

UNITED STATES NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD

x In re:

Docket Nos. 50-247-LR; 50-286-LR License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 Entergy Nuclear Indian Point 3, LLC, and Entergy Nuclear Operations, Inc.

June 8, 2015

x SUPPLEMENTAL REPORT OF DR. JORAM HOPENFELD IN SUPPORT OF CONTENTION NYS-26/RK-TC-1B AND AMENDED CONTENTION NYS-38/RK-TC-5

Contains Entergy/Westinghouse Designated Proprietary Information Docket Nos. 50-247-LR & 50-286-LR Supplemental Report of Dr. Joram Hopenfeld (June 2015) i TABLE OF CONTENTS

SUMMARY

.................................................................................................................................... 2 INTRODUCTION.......................................................................................................................... 3 DISCUSSION................................................................................................................................. 6

1. The CUFen Equation............................................................................................................ 6 1.1 Argonne National Laboratories Methodology of Accounting for Environmental Effects....................................................................................................................... 6 1.2 ANL Tests and Data Extrapolation........................................................................... 6 1.3 The Fen Equation....................................................................................................... 7

2. Inadequate Consideration of Dissolved Oxygen.................................................................. 8 2.3 Entergys Consideration of Dissolved Oxygen at IP2 and IP3................................. 9 2.4 Discussion of Entergys Incorrect Dissolved Oxygen Theory............................ 10

3. Inadequate Consideration of Radiation and Stress Corrosion Effects............................... 14 3.1 Radiation and Stress Corrosion Effects.................................................................. 14 3.2 Entergys Theory Why Radiation and the Synergistic Effects of Stress Corrosion cracking (SCC) on Metal Fatigue can be Neglected............................................... 16 3.3. Discussion of Entergys Theory on Radiation and Synergistic Effects......................... 17

4. Errors From Assuming CUF = CUF of Record................................................................. 19 4.1 Geometry Changes.................................................................................................. 20 4.2 Surface Finish......................................................................................................... 20 4.3 Heat Transfer.......................................................................................................... 21 4.4 Strain Rate............................................................................................................... 23 4.5 Radiation Effects..................................................................................................... 23

5. Inadequate Determination of the Most Limiting Locations............................................... 25 5.1 Pressurizer Surge Line, Mixing Tees, Unisolable Branches Connected to RCS Piping...................................................................................................................... 26 5.2 Steam Generator Tubes and Steam Generator Secondary Side.............................. 27 5.3 Reactor Head Penetrations, Outlet Inlet Nozzle Safe Ends................................... 28

6. Summary Assessment and Safety Implications of Entergys Results............................... 28 CONCLUSIONS........................................................................................................................... 31 REFERENCES............................................................................................................................. 34

Contains Entergy/Westinghouse Designated Proprietary Information Docket Nos. 50-247-LR & 50-286-LR Supplemental Report of Dr. Joram Hopenfeld (June 2015) 4 amended its consolidated contention to challenge Entergys refined fatigue evaluations and the ongoing inadequacy of Entergys aging management plan for metal fatigue at Indian Point.9 Subsequently, in 2011, the NRC acknowledged and conceded that there may be more limiting locations at Indian Point than those identified in NUREG/CR-6260 that were analyzed by Entergy, and requested that Entergy confirm and justify bounding locations for IP2 and IP3.10 However, as memorialized in Supplement 1 to the Indian Point Safety Evaluation Report dated August 2011, NRC Staff accepted Entergys vague Commitment 43 to address this issue. In response, Riverkeeper and the State of New York filed an additional contention, Contention NYS-38/RK-TC-5, which, among other bases, contested Entergys program for managing metal fatigue due to Entergys failure to expand the scope of its fatigue analysis and conduct a bounding metal fatigue assessment.11 In connection with this contention, I explained, among other things, that Entergy must expand its analysis to include balance-of-plant and reactor vessel internal (RVI) components. Notably, Entergy justified its failure to conduct fatigue analysis for balance-of-plant components by claiming that the fatigue life of such components had been conservatively analyzed. However, industry guidelines do not specify that balance-of-plant components can be excluded from CUFen analysis, as I have raised in submissions related to Riverkeepers admitted contentions.

Subsequently, the NRC Staff undertook a supplemental safety review, which culminated in the issuance of Supplement 2 to the Indian Point Safety Evaluation Report in November 2014. In this report, NRC Staff memorialized Entergys Commitment 49 to manage the effects of fatigue on RVI components at Indian Point during the proposed periods of extended operations by relying on its Fatigue Monitoring Program and recalculating CUF values for RVI components to include reactor coolant environment effects.12 In response, Riverkeeper and the State of New York successfully raised amended bases to contention NYS-38/RK-TC-5, with support of my expert declaration which criticized Entergys commitment and flawed methodology for determining CUFen values for RVI components.13 9 See State of New Yorks and Riverkeepers Motion for Leave to File a New and Amended Contention Concerning the August 9, 2010 Entergy Reanalysis of Metal Fatigue (Sept. 9, 2010); Declaration of Dr. Joram Hopenfeld in Support of Petitioners State of New York and Riverkeeper, Inc.s New and Revised Contention Concerning Metal Fatigue (Sept. 9, 2010).

10 See NRC Letter, Request for Additional Information for the Review of the Indian Point Nuclear Generating Unit Numbers 2 and 3, License Renewal Application (February 10, 2011) (Exhibit NYS000199); NUREG-1930, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3, Supplement 1 (August 2011) (Exhibit NYS000160).

11 State of New York and Riverkeepers New Joint Contention NYS-38/RK-TC-5 (September 30, 2011), ADAMS Accession No. ML11273A196.

12 See NUREG-1930, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3, Supplement 2 (November 2014), at 3-51 to 3-52 (Exhibit NYS000507).

13 State of New York and Riverkeepers Joint Motion for Leave to Supplement Previously-Admitted Joint Contention NYS-38/RK-TC-5 (February 13, 2015), ADAMS Accession No. ML15044A498; Declaration of Dr.

Joram Hopenfeld (February 12, 2015) (Exhibit RIV000148).

Contains Entergy/Westinghouse Designated Proprietary Information Docket Nos. 50-247-LR & 50-286-LR Supplemental Report of Dr. Joram Hopenfeld (June 2015) 5 In accordance with Entergys regulatory Commitment 43 to determine the limiting locations for IP2 and IP3, and regulatory Commitment 49 to calculate CUFen values for RVI components, and after years of delay, Entergy vendor, Westinghouse, conducted and issued refined fatigue analyses for Indian Point.14 The purpose of this report is to explain how these most recent fatigue evaluations are fundamentally flawed in various respects, and how Entergy continues to lack an adequate aging management program for metal fatigue at Indian Point. Westinghouse and Entergy either ignored important parameters or selected inputs that would minimize the effect of the environment on fatigue life, and was thereby able to obtain CUFen values that were <1. In particular, Entergys calculations are deficient in the following ways, as will be described in further detail below:

1.) Westinghouse/Entergy failed to properly account for the effects of dissolved oxygen on component fatigue; 2.) Westinghouse/Entergy failed to account for radiation and stress corrosion effects on metal fatigue; 3.) Westinghouse/Entergy continued the flawed approach of assuming a CUF of record in the fatigue analyses; and 4.) Westinghouse/Entergy failed to properly expand the scope of analysis to bound the most limiting locations.

In light of these various deficiencies, and given an uncertainty analysis should have been conducted, but was not.

Moreover, Westinghouse/Entergy failed to conduct a safety assessment to show that IP2 and IP3 can operate safely during normal operations and DBAs, despite the fact that many of the refined CUFen values are very close to 1 without any uncertainty allowance.

Based on a review of Entergys latest fatigue evaluations, the conclusion remains that Entergy has failed to demonstrate that the CUFs of components at Indian Point will not exceed unity and/or succumb to metal fatigue during the proposed periods of extended operation, or that it otherwise has an adequate program for managing the effects of metal fatigue at Indian Point during the proposed periods of extended operations for IP2 and IP3.

14 Westinghouse, Indian Point Unit 2 and Unit 2 EAF Screening Evaluations, Calculation Note Number CN-PAFM-12-35 (November 2012), IPECPROP00072778 (Exhibit NYS000510); Westinghouse, Indian Point Unit 2 (IP2) and Unit 3 (IP3) Refined EAF Analyses and EAF Screening Evaluations, Calculation Note Number CN, PAFM-13-32 (August 2013), IPECPROP00078338 (Exhibit NYS000511).

Contains Entergy/Westinghouse Designated Proprietary Information Docket Nos. 50-247-LR & 50-286-LR Supplemental Report of Dr. Joram Hopenfeld (June 2015) 6 DISCUSSION

1. The CUFen Equation Before the degree of inaccuracy and lack of conservatism in Entergys final predicted CUFen values can be appreciated, it is useful to review key concepts underlying the calculations of the environmentally corrected cumulative usage factors, CUFen, and then to describe how Energy improperly selected the input data that misleadingly caused all the most recent final CUFen values to be less than one.

1.1 Argonne National Laboratories Methodology of Accounting for Environmental Effects Since the American Society of Mechanical Engineers (ASME) code does not specify a methodology for how to determine the effects of the light water reactor (LWR) environment on metal fatigue, Argonne National Laboratories (ANL) undertook this task in the mid-1990s, with work still in progress. The ANL methodology is based on the following equation (hereinafter referred to as Equation 1):

CUFen (water) = Fen (lab) x CUF (air)

CUFen is the environmentally corrected cumulative usage factor (CUF), Fen is an environmental correction factor to relate laboratory fatigue data in water to laboratory fatigue data in air, and CUF (air) represents a CUF of a given plant component based on data that was obtained in air, commonly specified by the ASME code. Since the Fen is an experimental factor, the underlying principle of using the above equation is that the user would not extrapolate the Fen to conditions other than those that existed in its derivation.

1.2 ANL Tests and Data Extrapolation In the ANL tests, small polished specimens were exposed to cyclic loads in a loop where the temperature and water chemistry at the surface of the specimens were well known and were kept steady during the tests. The tests did not represent prototypic conditions such as would be experienced by actual components during thermal transients. In 2007, ANL published a detailed report, NUREG/CR-6909, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials,15 which summarized the test results and proposed models and methodologies to explain and account for how water chemistry and material composition affect metal fatigue. The report proposed a model for the transition from crack initiation to crack propagation, and described the role of oxygen and metal composition on crack propagation. The report presented experimental correlation on the effects of oxygen, temperature, and strain rate on the reduction in metal fatigue life when the specimen was exposed to water instead air.

15 NUREG/CR-6909, ANL-06-08, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials (2007) (Exhibit NYS000357).

Contains Entergy/Westinghouse Designated Proprietary Information Docket Nos. 50-247-LR & 50-286-LR Supplemental Report of Dr. Joram Hopenfeld (June 2015) 7 Recognizing the lack of prototypical results, NUREG/CR-6909 provided a lengthy list of the differences between the laboratory setting and the actual nuclear plant setting, emphasizing the limitation of the experimental data. In addition, Dr. O.K. Chopra, the principal investigator of the ANL research, discussed the results with the NRCs Advisory Committee on Reactor Safeguards (ACRS) and emphasized that it is the responsibility of the operator to account for the differences between the lab and plant environments when applying the results to draw conclusions regarding fatigue life.16 Dr. Chopra emphasized that the ANL results may not be conservative.17 Recognizing that the results of the ANL tests were not prototypical, the Electric Power Research Institute (EPRI) issued a list of guidelinesMRP-47regarding how the ANL data should be corrected to account for differences between laboratory and plant environments.18 As discussed below, throughout the course of the Indian Point license renewal proceeding, Entergy has taken the erroneous position that the ANL data can be applied to IP2 and IP3 directly for accurate metal fatigue predictions without accounting for the known differences between the laboratory and plant environments. Entergy has also incorrectly maintained the position that that CUF in Equation 1 above can be substituted by the CUF of records, without major reanalysis of the CUF of records, which is again discussed in further detail below.

1.3 The Fen Equation The environmental correction factor, Fen, in Equation 1 above was obtained as a best statistical fit to experimental data. When compared to the experimental data, the ANL best-fit data model, as set forth in NUREG/CR-6909, predicts fatigue life within a factor of 3 for low carbon and low alloy steel and austenitic steels.19 Equation 1 is an approximate, not an exact equation, and some data points will fall outside of predictions derived from using the equation.

The Fen equation provided in NUREG/CR-6909 is as follows:

Fen = exp (K - K1( S*

T* O*))

In this equation the different variable represent the following:

K, K1 =

constants depending on material type (carbon steel, low alloy steel stainless steel)

S*

=

Transformed Sulphur 16 Official Transcript of Proceedings, Nuclear Regulatory Commission, Advisory Committee on Reactor Safeguards Subcommittee on Materials, Metallurgy and Reactor Fuels (December 6, 2006), ADAMS Accession No. ML063550058, at 22 (Exhibit RIV000037).

17 See generally id.

18 EPRI, MRP-47, Materials Reliability Program: Guidelines for Addressing Fatigue Environmental Effects in a License Renewal Application, Final Report, Revision 1 (September 2005) (Exhibit NYS000350).

19 See NUREG/CR-6909, ANL-06-08, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials (2007) at 26, 62 (Exhibit NYS000357).

Contains Entergy/Westinghouse Designated Proprietary Information Docket Nos. 50-247-LR & 50-286-LR Supplemental Report of Dr. Joram Hopenfeld (June 2015) 8 O*

=

Transformed dissolved oxygen near metal surface T*

=

Transformed temperature tem at metal surface: T-150

=

Transformed strain rate O or ( DO)

=

dissolved oxygen concentration at the metal surface O*

=

0 when O <0.04 parts per million (ppm)

O*

=

ln(O/.04) for O> 0.04 ppm

2. Inadequate Consideration of Dissolved Oxygen Dissolved oxygen, DO, is the most important of 12 major variables listed by ANL that must be considered when applying Equation 1 to the LWR environment.

The formation of oxide films on the surface of stainless steel is believed to play a major role in metal fatigue, however that effect is still not completely understood. Until this mechanism is better understood, the application of the Fen equation to plant components must employ DO values that resemble those that were used in the development of that equation. Oxygen concentrations in the laboratory tests were uniform in liquid and were conducted at steady state with controlled water chemistry and temperature. On the other hand, in the plant, local oxygen concentrations are not well known during transients because measurements in the plant are made by bulk sampling periodically during steady state operations,20 usually far removed from the component of interest during thermal transients.

Reactor coolant contains high concentrations of corrosive products that vary in composition around the flow path. Oxygen enters the reactor system usually during heat-up and cool-down operations and is also generated by electrolysis in the core. Hydrazine is used to maintain low DO levels during steady state operating temperatures. Hydrazine is no more than a catalyst which facilitates the formation of metal oxides and hydroxides, (Fe2)O3, Fe(OH)3, respectively.

Hydrazine does not remove oxygen from the reactor system; it only changes its form, which varies with the temperature. Thus, the oxygen could be bound to metal surfaces or be floating crud in the system. It is not possible to arbitrarily assume that the sampled steady state DO concentration in a typical PWR would be the same as the DO at the surfaces of all components at IP2 and IP3 during all transients where the temperature undergoes abrupt changes. It is the concentration of DO during temperature transients that must be entered into and accounted for in the Fen equation. Such concentrations must be measured, as they can neither be assumed nor calculated. In recognition of the difficulties of determining the DO in real life environments, ANL and EPRI developed guidelines for how to enter the oxygen term in the Fen equation for plant application.

20 See EPRI, MRP-47, Materials Reliability Program: Guidelines for Addressing Fatigue Environmental Effects in a License Renewal Application, Final Report, Revision 1 (September 2005) (Exhibit NYS000350).

Contains Entergy/Westinghouse Designated Proprietary Information Docket Nos. 50-247-LR & 50-286-LR Supplemental Report of Dr. Joram Hopenfeld (June 2015) 15 cycles below 5000 per second.41 Tests in the U.S. and Japan show that irradiation can increase crack growth rates by more than a factor of 5 in low oxygen boiling water reactor (BWR) environments.42 Similarly to affecting fatigue, radiation also effects stress corrosion cracking (SCC). The main difference between SCC and metal fatigue is that the former occurs under cyclic loads while the latter occurs at static loads. In a 2012 ANL study, S. Mohanty, et al discussed several deterministic models with empirical crack growth rates to predict SCC and fatigue life.43 Crack growth rates for alloy 600 were presented in terms of an experimental stress intensity factor and cyclic frequency, stress ratio, and surface temperature. The study concluded that many empirical models are available for reactor component base metals but very few for dissimilar metal welds.44 The study further concluded that metal fatigue, flow accelerated corrosion (FAC), and SCC can act in combination with each other to magnify their individual effects.45 Thus, any analysis of the effects of the LWR environment on fatigue must consider the synergistic effects of radiation SCC and thermal embrittlement. A first step towards this end would be to incorporate the effects of radiation into the Fen equation.

Based on extensive literature review of the effects of radiation on fatigue in the LWR environment, a draft revised NUREG/CR-6909 concluded that the limited available data was inconclusive with regard to the impact of irradiation on fatigue in the LWR environment.46 However, in this revised and still draft report, ANL and NRC inappropriately recommend that the effects of radiation on the Fen be ignored because the existing data cannot be used to quantize these effects and that to do so, data on the effects of radiation on the Fen would have to be obtained.47 The draft report gives no indication that such data will ever be obtained, so instead supports ignoring the effects of radiation on fatigue life.

41 G. E. Korth & M. D. Harper, Effects of Neutron Radiation on the Fatigue and Creep/Fatigue Behavior of Type 308 Stainless Steel Weld Metal at Elevated Temperatures (Seventh ASTM International Symposium on Effects of Radiation on Structural Materials, June 1974), available at, http://www.osti.gov/scitech/servlets/purl/4294682 (Exhibit RIV000152).

42 See Argonne National Laboratory, Corrosion and Mechanics of Materials, Light Water Reactors, Irradiation-Induced Stress Corrosion Cracking of Austenitic Stainless Steels, http://www.ne.anl.gov/capabilities/cmm/highlights/ssc austenic ss.html (last visited June 2, 2015) (Exhibit RIV000153).

43 S. Mohanty, S. Majumdar, & K. Natesan, Argonne National Laboratory, A Review of Stress Corrosion Cracking/Fatigue Modeling for Light Water Reactor Cooling System Components (June 2012) (Exhibit RIV000154).

44 See id.

45 See id.

46 See NUREG/CR-6909, Revision 1, ANL-12/60, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, Draft Report for Comment (March 2014), available at, http://pbadupws.nrc.gov/docs/ML1408/ML14087A068.pdf (Exhibit NYS000490).

47 Id.

Contains Entergy/Westinghouse Designated Proprietary Information Docket Nos. 50-247-LR & 50-286-LR Supplemental Report of Dr. Joram Hopenfeld (June 2015) 17 3.3. Discussion of Entergys Theory on Radiation and Synergistic Effects Entergys apparent theory as described above disregards scientific facts and the prevailing views of many researchers on the effects of radiation/neutron fluence, and SCC on metal fatigue.

Entergys reliance on the notion that components with CUF values of less than 1 will not be impacted by radiation effects or subject to fatigue initiation or propagation is based on an incorrect perception of the meaning of the CUF. In the absence of the appropriate consideration of radiation effects and other critical parameters, a CUFen value that is less than 1.0 does not necessarily indicate that fatigue issues will not arise during the proposed periods of extended operation. Importantly, a CUFen of less than one does not necessarily demonstrate that fatigue initiation is not expected during the life of the component.

The position that a component will not be subject to fatigue cracks because the CUFen for that component is less than one stands in contradiction of the conventional understanding of metal fatigue. The CUF does not represent the absence of cracks or flaws in the material when the CUF or CUFen is less than 1. The CUF is used to assess the possibility of fatigue failure in a given environment and is based on a criterion that the CUF should be kept below one. Large numbers of fatigue tests have shown that a statistically significant number of test specimens would fail under cyclic loading when the CUF exceeded one. These results were based only on the observations of specimen failures, not on crack size history during the tests. The CUF is strictly an empirical criterion that has proven to work well over half a century for full size components in diverse applications.

The common understanding of crack initiation and propagation under cyclic loads stands in contradiction to Entergys theory. NUREG/CR-6909 schematically depicts crack formation during the fatigue life of specimens under cyclic loads, and clearly indicates that cracks (flaws) are present from the beginning of the test, throughout the fatigue life of a specimen under cyclic loads; the report explains that fatigue life may be considered to constitute propagation of cracks from 10 to 3000 [micro meters] long.52 Schematically, the transition from microscopic cracks to macroscopic cracks occurs about the time it reaches its half-life. The initiation stage of fatigue involves the growth of microscopic cracks and this stage is not characterized by any specific value of CUFen. Importantly, NUREG/CR-6909 as well as other studies give no indication that fatigue initiation does not exist when CUFen <1.0, or that there is a correlation between crack size and the CUFen. A 2012 ANL study also shows that the time when crack initiation occurs depends on empirical constants and the strain rate, with no mention that cracks do not propagate when CUFen <1.0.53 In addition, NRC studies at the Oak Ridge National Laboratory (ORNL) show that there is no correlation between a propagating crack size and the CUF as it approaches unity. For example, one study has indicated that when a component 52 NUREG/CR-6909, ANL-06-08, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials (2007) at 7 (Exhibit NYS000357).

53 See S. Mohanty, S. Majumdar, & K. Natesan, Argonne National Laboratory, A Review of Stress Corrosion Cracking/Fatigue Modeling for Light Water Reactor Cooling System Components (June 2012) (Exhibit RIV000154).

Contains Entergy/Westinghouse Designated Proprietary Information Docket Nos. 50-247-LR & 50-286-LR Supplemental Report of Dr. Joram Hopenfeld (June 2015) 33 inspected components and the frequency of inspections. Without accurate CUFen evaluations, the number and frequency of inspections will not minimize the risk of failures by metal fatigue.

Non-conservative CUFens increase the possibility that fatigue susceptible components would remain in service due to inadequately chosen inspection intervals.

Furthermore, because Entergys screening evaluation was flawed in numerous respects, Entergy has yet to conduct a bounding analysis and has continued to fail to adequately expand the scope of its fatigue analysis as is necessary.

In light of the absence of comprehensive, accurate metal fatigue calculations to properly guide Entergys aging management efforts, Entergy has failed to define specific criteria to assure that susceptible components are inspected, monitored, repaired, or replaced in a timely manner.

Once components with high CUFs have been properly identified, Entergy must describe a fatigue management plan for each such component that should, at a minimum, rank components with respect to their consequences of failure, establish criteria for repair versus defect monitoring, and establish criteria for the frequency of the inspection, and corrective actions.

In light of the foregoing, Entergy has failed to demonstrate that it has an adequate program to monitor, manage, and correct metal fatigue related degradation sufficient to comply with 10 C.F.R. § 54.21(c), or the regulatory guidance of NUREG-1801, Generic Aging Lessons Learned (GALL) Report.

Contains Entergy/Westinghouse Designated Proprietary Information Docket Nos. 50-247-LR & 50-286-LR Supplemental Report of Dr. Joram Hopenfeld (June 2015) 34 REFERENCES ANL, Report on Assessment of Environmentally-Assisted Fatigue for LWR Extended Service Conditions, ANL-LWRS-47 (September 2011) (Exhibit RIV000150)

Argonne National Laboratory, Corrosion and Mechanics of Materials, Light Water Reactors, Irradiation-Induced Stress Corrosion Cracking of Austenitic Stainless Steels, http://www.ne.anl.gov/capabilities/cmm/highlights/ssc austenic ss.html (last visited June 2, 2015) (Exhibit RIV000153)

Combustion Engineering, Inc., CENC-1110 (Exhibit RIV000052A-D)

Declaration of Dr. Joram Hopenfeld (February 12, 2015) (Exhibit RIV000148)

Declaration of Dr. Joram Hopenfeld in Support of Petitioners State of New York and Riverkeeper, Inc.s New and Revised Contention Concerning Metal Fatigue (Sept. 9, 2010)

Entergy Ultrasonic Examination Report, IPEC00020853 (Exhibit RIV000130)

EPRI, MRP-47, Materials Reliability Program: Guidelines for Addressing Fatigue Environmental Effects in a License Renewal Application, Final Report, Revision 1 (September 2005) (Exhibit NYS000350)

G. E. Korth & M. D. Harper, Effects of Neutron Radiation on the Fatigue and Creep/Fatigue Behavior of Type 308 Stainless Steel Weld Metal at Elevated Temperatures (Seventh ASTM International Symposium on Effects of Radiation on Structural Materials, June 1974), available at, http://www.osti.gov/scitech/servlets/purl/4294682 (Exhibit RIV000152)

G.T. Yahr, et al, Case Study of the Propagation of a Small Flaw Under PWR Loading Conditions and Comparison with the ASME Code Design Life (Exhibit RIV000118)

Herve Bodineau & Thierry Sollier, Tube support plate clogging up of French PWR steam generators, Eurosafe, IRSN - Reactor Safety Division, BP17 (2009) (Exhibit RIV000158).

Indian Point Energy Center License Renewal Application (Exhibit ENT00015A-B)

J.S. Kim & J.S. Seo, Development of Engineering Formulae for Stress Concentration Factors of Local Wall Thinning in CANDU Feeder Pipe Under Pressure, Proceedings of the ASME 2011 Pressure Vessels & Piping Division Conference, PVP2011, July 17-21, 2011, Baltimore, Maryland, USA, PVP2011-57930 (Exhibit RIV000138)

K. L. Murty, Materials Ageing and Degradation in Light Water Reactors (Woodhead Publishing 2013) (Exhibit RIV000145)

Contains Entergy/Westinghouse Designated Proprietary Information Docket Nos. 50-247-LR & 50-286-LR Supplemental Report of Dr. Joram Hopenfeld (June 2015) 35 Material Aging Institute International Conference on Plants Materials Degradations, Chemical conditioning of Light Water reactors Systems (2008), IPEC00265853 (Exhibit RIV000149)

NASA, Nuclear and Space Radiation Effects on Materials, NASA SP-8053 (June 1970),

available at, http://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19710015558.pdf (Exhibit RIV000151)

NL-10-082, License Renewal Application - Completion of Commitment #33 Regarding the Fatigue Monitoring Program (August 9, 2010) (Exhibit NYS000352),

http://pbadupws.nrc.gov/docs/ML1023/ML102300504.pdf NRC Bulletin No. 88-11: Pressurizer Surge Line Thermal Stratification (December 20, 1988)

(Exhibit RIV000115).

NRC Letter, Request for Additional Information for the Review of the Indian Point Nuclear Generating Unit Numbers 2 and 3, License Renewal Application (February 10, 2011) (Exhibit NYS000199)

NUREG-75/014 (WASH-1400), Reactor Safety Study: An Assessment of Accident Risks in U.S.

Commercial Nuclear Power Plants (1975) (Exhibit RIV000147)

NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants (1990), available at http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1150/ (Exhibit RIV00146A-RIV00146B).

NUREG-1801, Generic Aging Lessons Learned (GALL) Report (Exhibit NYS00146A-146C)

NUREG-1930, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3, Supplement 1 (August 2011) (Exhibit NYS000160).

NUREG-1930, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3, Supplement 2 (November 2014) (Exhibit NYS000507)

NUREG/BR-0058, Rev.4, Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission (2004), available at, http://www.nrc.gov/reading-rm/doc-collections/nuregs/brochures/br0058/br0058r4.pdf (Exhibit RIV000159)

NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components (February 1995) (Exhibit NYS000335)

NUREG/CR-6909, ANL-06-08, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials (2007) (Exhibit NYS000357)

Contains Entergy/Westinghouse Designated Proprietary Information Docket Nos. 50-247-LR & 50-286-LR Supplemental Report of Dr. Joram Hopenfeld (June 2015) 36 NUREG/CR-6909, Revision 1, ANL-12/60, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, Draft Report for Comment (March 2014), available at, http://pbadupws.nrc.gov/docs/ML1408/ML14087A068.pdf (Exhibit NYS000490)

Official Transcript of Proceedings, Nuclear Regulatory Commission, Advisory Committee on Reactor Safeguards Subcommittee on Materials, Metallurgy and Reactor Fuels (December 6, 2006), ADAMS Accession No. ML063550058 (Exhibit RIV000037)

Pilkey & Pilkey, Petersons Stress Concentration Factors, 3rd Edition, ISBN: 978-0-470-04824-5 (2008) (Exhibit RIV000157)

Prefiled Written Testimony of Dr. Joram Hopenfeld Regarding Contention NYS-38/RK-TC-5 (June 19, 2012) (Exhibit RIV000102)

Prefiled Rebuttal Testimony of Dr. Joram Hopenfeld Regarding Contention NYS-38/RK-TC-5 (November 9, 2012) (Exhibit RIV000134)

Prefiled Written Testimony of Dr. Joram Hopenfeld Regarding Riverkeeper Contention TC Flow Accelerated Corrosion (December 22, 2011) (Exhibit RIV000003).

Report of Dr. Joram Hopenfeld in Support of Contention NYS-26-B/RK-TC-1B - Metal Fatigue (December 22, 2011) (Exhibit RIV000035)

Riverkeeper, Inc.s Request for Hearing and Petition to Intervene in the License Renewal Proceeding for the Indian Point Nuclear Power Plant (November 30, 2007), ADAMS Accession No. ML073410093 S. McKelvey & A. Fatemi, Effect of Forging Surface on Fatigue Behavior of Steels: A Literature Review (University of Toledo) (Exhibit RIV000155)

S. Mohanty, S. Majumdar, & K. Natesan, Argonne National Laboratory, A Review of Stress Corrosion Cracking/Fatigue Modeling for Light Water Reactor Cooling System Components (June 2012) (Exhibit RIV000154)

State of New Yorks and Riverkeepers Motion for Leave to File a New and Amended Contention Concerning the August 9, 2010 Entergy Reanalysis of Metal Fatigue (Sept. 9, 2010)

State of New York and Riverkeepers Joint Motion for Leave to Supplement Previously-Admitted Joint Contention NYS-38/RK-TC-5 (February 13, 2015), ADAMS Accession No. ML15044A498 State of New York and Riverkeepers New Joint Contention NYS-38/RK-TC-5 (September 30, 2011), ADAMS Accession No. ML11273A196.

Contains Entergy/Westinghouse Designated Proprietary Information Docket Nos. 50-247-LR & 50-286-LR Supplemental Report of Dr. Joram Hopenfeld (June 2015) 37 Testimony of Entergy Witnesses Nelson F. Azevedo, Alan B. Cox, Jack R. Strosnider, Robert E.

Nickell, and Mark A. Gray Regarding Contention NYS-26B/RK-TC-1B (Metal Fatigue) (March 29, 2012) (Exhibit ENT0000183)

U.S. NRC, Commissioner Apostolakis, Application of Risk Assessment and Management to Nuclear Safety (DOE Nuclear Safety Workshop, September 2012), available at, http://energy.gov/sites/prod/files/2013/12/f5/Apostolakis.pdf (Exhibit RIV000160).

Westinghouse, Environmental Fatigue Evaluation for Indian Point Unit 2, WCAP-17199-P, Revision 0 (June 2010), IPECPROP00056486 (Exhibit NYS000361)

Westinghouse, Environmental Fatigue Evaluation for Indian Point Unit 3, WCAP-17200-P, Revision 0 (June 2010), IPECPROP00056577 (Exhibit NYS000362)

Westinghouse, Indian Point Unit 2 and Unit 2 EAF Screening Evaluations, Calculation Note Number CN-PAFM-12-35 (November 2012), IPECPROP00072778 (Exhibit NYS000510)

Westinghouse, Indian Point Unit 2 (IP2) and Unit 3 (IP3) Refined EAF Analyses and EAF Screening Evaluations, Calculation Note Number CN-PAFM-13-32 (August 2013),

IPECPROP00078338 (Exhibit NYS000511)

Westinghouse, Indian Point Unit 2 Reactor Vessel Core Support Pad Fatigue Reevaluation, CN-MRCDA-13-11, IPECPROP00079179 (Exhibit RIV000156)