ML12335A457

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Official Exhibit - NYS000296-00-BD01 - Report of Dr. Richard T. Lahey, Jr. in Support of Contentions NYS-25 and NYS-26B/RK-TC-1B
ML12335A457
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 12/22/2011
From: Lahey R
State of NY, Office of the Attorney General
To:
Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 21617, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01
Download: ML12335A457 (43)
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United States Nuclear Regulatory Commission Official Hearing Exhibit NYS000296 Entergy Nuclear Operations, Inc. Submitted: December 22, 2011 In the Matter of:

(Indian Point Nuclear Generating Units 2 and 3)

ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #: NYS000296-00-BD01 Identified: 10/15/2012 Admitted: 10/15/2012 Withdrawn:

Rejected: Stricken:

Other:

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. December 20, 2011

x REPORT OF DR. RICHARD T. LAHEY, JR.

IN SUPPORT OF CONTENTIONS NYS-25 AND NYS-26B/RK-TC-1B

Prepared for the State of New York Office of the Attorney General

REPORT OF DR. RICHARD T. LAHEY, JR.

The purpose of this Report is to document my principal technical concerns associated with the relicensing of the two operating nuclear reactors at the Indian Point site in Buchanan, New York (i.e., Indian Point Unit 2 & Indian Point Unit 3 ).

INTRODUCTION

1. I am the Edward E. Hood Professor Emeritus of Engineering at Rensselaer Polytechnic Institute (RPI) in Troy, New York. I have earned the following academic degrees: a B.S. in Marine Engineering from the United States Merchant Marine Academy, a M.S. in Mechanical Engineering from Rensselaer Polytechnic Institute (RPI), a M.E. in Engineering Mechanics from Columbia University, and a Ph.D. in Mechanical Engineering from Stanford University. I have also held various technical and administrative positions in the nuclear industry. In addition, at RPI, I have served as both the Dean of Engineering and the Chairman of the Department of Nuclear Engineering & Science.
2. Previously, I was responsible for nuclear reactor safety R&D (research

& development) for the General Electric Company (GE), and I have extensive experience with both military (i.e., naval) and commercial nuclear reactors.

3. I am a member of various professional societies, including: the American Nuclear Society (ANS), where I was a member of the Board of Directors and the ANSs Executive Committee, and was the founding Chair of the ANS Report of Richard T. Lahey, Jr.

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Thermal-Hydraulics Division; the American Society of Mechanical Engineers (ASME), where I was Chair of the Nucleonics Heat Transfer Committee, K-13; the American Institute of Chemical Engineering (AIChE), where I was the Chair of the Energy Transport Field Committee; and the American Society of Engineering Educators (ASEE), where I was Chair of the Nuclear Engineering Division. I was also an Editor of the international Journal of Nuclear Engineering & Design.

4. In addition, I have served on numerous panels and committees for the United States Nuclear Regulatory Commission (USNRC), Idaho National Engineering Laboratory (INEL), Oak Ridge National Laboratory (ORNL), and the Electric Power Research Institute (EPRI). I am a member of the National Academy of Engineering (NAE), and have been elected Fellow of both the ANS and the ASME.
5. Over the last 50 years, I have published numerous books, monographs, chapters, articles, reports, and journal papers on nuclear engineering and nuclear reactor safety technology, and have received many honors and awards for my career accomplishments, including: the E.O. Lawrence Memorial Award of the Department of Energy (DOE), the Glenn Seaborg Medal of the ANS and the Donald Q. Kern Award of the AIChE. I am widely considered to be an expert in matters relating to the design, operations, safety and aging of nuclear power plants. My Curricula Vitae, which accompanies this report, describes the details of my educational and professional background and qualifications.

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6. I am quite familiar with the type of pressurized water nuclear reactors (PWRs) at the Indian Point site in Buchanan, New York. I have previously submitted a declaration on November 30, 2007 in support of the Notice of Intention to Participate and Petition to Intervene filed by the State of New York in this proceeding, a declaration dated April 7, 2008 in further support of the State's Supplemental Contention 26-A (metal fatigue), a September 8, 2010 declaration in support of the State's New and Amended Contention Concerning Metal Fatigue (NYS-26B/RK-TC-1B), a September 15, 2010 declaration in support of the State's Additional Bases for Previously-Admitted Contention NYS-25 (Embrittlement of Reactor Pressure Vessels and Associated Internals), a September 30, 2011 declaration and November 1, 2011 declaration in support of the State's recent contention following the publication of the recent Supplemental Safety Evaluation Report (SSER) for the Indian Point nuclear reactors.
7. The factual statements and the expression of opinion in this report are based on, among other things, my best professional knowledge, my extensive professional experience in nuclear reactor technology, and my review of various documents including, but not limited to, Entergys 2007 License Renewal Application (LRA), Entergys July 14, 2010 submission of License Renewal Application Amendment No. 9, [NL-10-063] , Entergy's August 9, 2010 letter to NRC concerning License Renewal Application Commitment No. 33 [NL-10-082],

various documents from the Electric Power Research Institute (EPRI) and Report of Richard T. Lahey, Jr.

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Westinghouse (W), as well as many other related technical documents, including those referenced herein.

OVERVIEW

8. Before discussing my specific technical concerns associated with the relicensing of the two operating Indian Point reactors, Indian Point Unit 2 and Indian Point Unit 3 (IP-2 and IP-3), I believe that an overview of the United States Nuclear Regulatory Commissions (USNRCs) relicensing process would help put my concerns in perspective.
9. As commercial nuclear power plants exceed their normal design life of 40 years, key structures and components degrade (i.e., they begin to wear out). To assure safe operation during the proposed 20-year life extension of the two Indian Point reactors, it is imperative not to seriously erode the original design margins, since the probability of reactor system and component failures increases with age.
10. To ensure that aging nuclear reactor systems, structures, components and fittings do not become compromised during the term of an extended operating license, the USNRC has recognized the importance of developing and implementing effective aging management programs (AMPs) for these structures, components and fittings [10 C.F.R. Part 54] to protect the health and safety of the American public. Among other things, it is important to develop and implement an effective AMP to ensure that the nuclear reactor system, structures, components and fittings do not become seriously degraded and/or fail due to neutron-induced embrittlement Report of Richard T. Lahey, Jr.

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and/or fatigue, and other age-related degradation mechanisms, such as various stress corrosion cracking (SCC) mechanisms. Entergy has proposed AMPs to address these technical concerns, but as I will discuss subsequently, their AMPs are grossly deficient.

11. In order to facilitate the licensing renewal process, the USNRC has developed a Standard Review Plan (SRP) for the license renewal applications of nuclear power plants [NUREG-1800, Rev. 1]. This plan is comprised of a highly prescriptive process which does not encourage the discovery of any new aging-related safety concerns during ASLB hearings. In addition, the USNRC staff have prepared a guidance document, entitled the Generic Aging Lessons Learned (GALL) Report [NUREG-1801, Rev. 1], in which the staff give some guidance for Aging Management Programs (AMPs) and Reviews (AMRs) for the extended operations of nuclear power plants. Nevertheless, the USNRCs Standard Review Plan [e.g., NUREG-1800, Rev. 1, at p. 3.1-1; NUREG-1800, Rev. 2, at p.3.1-1]

recognizes the need to also address, AMR results not consistent with or not addressed in the GALL report, and my principle technical concerns are in this category, since these two documents do not address the safety-related aging concerns that I have identified in this proceeding. In December 2010, USNRC Staff revised the GALL Report and the Standard Review Plan [NUREG-1800, Rev. 2; NUREG-1801, Rev. 2], but unfortunately these revisions still do not address my concerns.

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12. Historically, the various safety-related aging issues have always been considered and analyzed in silos. That is, they have been studied separately and with little or no interaction. For example, the analysis of the thermal/mechanical-induced fatigue of the various components, structure and fittings of nuclear reactors has been done without considering the synergistic effect of the degradation due to significant corrosion and/or any irradiation-damage-induced embrittlement of these materials. Moreover, the required new safety analyses (documented in the plant-specific FSARs), such as those done for various loss of coolant accidents (LOCAs),

have been done implicitly assuming that the in-core geometry will remain intact.

Unfortunately, during the period of life extension, many reactor pressure vessel (RPV) internals may become highly fatigued and/or embrittled, and they could fail due to the pressure and/or thermal shock loads associated with various postulated LOCA events. If so, core coolability may be compromised such that core melting and radiation releases from the fuel may occur. Indeed, if a coolable core geometry can not be maintained, due to the failure of various in-core structures, components and fittings, the core cooling may be seriously degraded due to flow blockages, causing it to melt and relocate to the lower RPV head. As has recently been graphically demonstrated during the Fukushima nuclear emergency, this can cause the lower head to fail releasing radioactive materials into the containment and the environment. Clearly this is a very undesirable event. Unfortunately, the AMPs that Entergy have proposed to follow for IP-2 & IP-3 do not address these important aging-related safety concerns.

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13. Another very important aging-related safety concern has to do with how Entergy has proposed to treat the important primary piping systems and fittings of the two Indian Point nuclear reactors. In particular, Entergy has presented the results of fatigue calculations done for them by Westinghouse (W) using the WESTEMS computer code. These results show that a number of the piping systems and fittings that comprise the pressure boundary on the primary side of the nuclear reactors at Indian Point will be very close to their design fatigue limits during the period of extended operation. Nevertheless, they have proposed to take no action to repair or replace these components until just before they reach the design limit on fatigue. Significantly, they have not done an error analysis for these results, even though there were many modeling and input data assumptions and code user interventions made in doing those analyses with WESTEMS, hence the accuracy of the results given is not clear. In addition, this is an excellent example of how doing everything in silos can yield potentially dangerous results. That is, these fatigue evaluations are based on quasi-static cyclic loads which may eventually degrade the structures (i.e., weaken the structures by causing fatigue-induced surface cracking). However, if, for example, part of the primary pressure boundary experiences pressure and /or thermal shock loads due to reactor SCRAMS, some anticipated transients without SCRAMS (ATWS), or a secondary side LOCA, it could fail well before it has reached its so-called high or low cycle design limit, thus inducing a serious primary side LOCA. Clearly, the AMP being followed does not adequately protect the health and safety of the American public, Report of Richard T. Lahey, Jr.

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and is thus unacceptable. In fact, I believe that one must maintain a sufficient design margin to accommodate such occurrences, by, for example, replacing or repairing the fatigued structures well before the fatigue limit is reached.

DISCUSSION

14. Let us now consider in more detail my fundamental concerns with the adequacy of Entergys license renewal application (LRA). In particular, Entergys Indian Point Unit 2 & Unit 3 (IP-2 & IP-3) are currently under consideration for 20-year life extensions beyond their original 40-year design life. If approved, these plants will be licensed for operation levels of about 48 effective full power years (EFPY). These Westinghouse-designed plants are 4-loop PWRs and they are currently 1 rated at power levels of 3,216.4 MWt.
15. These plants are sited on the east bank of the Hudson River in Buchanan, NY, which is about 24 miles north of the New York City (NYC) border. 2



1 The USNRC approved a stretch power increase of 3.26% for IP-2 in 2004 and a 4.85% increase for IP-3 in 2005; IP-2 and IP-3 also received 1.4% power uprates in 2003 and 2002, respectively. Thus, these plants are already being driven harder than their original designs envisioned.

2 By way of additional reference, the Indian Point reactors are approximately 37 miles north of Wall Street in lower Manhattan; 3 miles southwest of Peekskill, NY; 5 miles northeast of Haverstraw, NY; 16 miles southeast of Newburgh, NY; 17 miles northwest of White Plains, NY; 23 miles northwest of Greenwich, Connecticut; 37 miles west of Bridgeport, Connecticut, and 37-39 miles north-northeast of Jersey City and Newark, New Jersey, respectively. In addition, the Indian Point reactors are approximately 5 miles west of the New Croton Reservoir in Westchester County, Report of Richard T. Lahey, Jr.

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Because of their close proximity to a very highly populated area (i.e., the NYC metropolitan area), which is also the worlds leading financial center, it is vital that IP-2 & IP-3 fully and unambiguously meet all reasonable and applicable criteria for safe operation. This is particularly true when considering life extension, since failures due to embrittlement and metal fatigue are much more likely as the plants age.

16. As noted previously, USNRC staff have prepared a guidance document entitled the Generic Aging Lessons Learned (GALL) Report, [NUREG-1801, Rev.

1 (2005)], in which the Staff seeks to describe various acceptable Aging Management Programs (AMPs) for the extended operations of nuclear power plants.

It is important to note that this USNRC document does not specifically describe aging management programs for the embrittlement of internal structures, components and fittings within the reactor pressure vessel (RPV), including, but not limited to, the: control rods and their associated guide tubes, the control rod assemblies, plates and associated seal welds, and many important in-core structures, components and fittings which will be discussed subsequently [See GALL, Chapter XI (Aging Management Programs); see also Entergy NL-10-063, at pg. 84 (Revision 1 of NUREG-1801 includes no aging management program description for PWR reactor vessel internals.)]. Although the USNRC Staff did not originally have an aging management program (AMP) to address reactor pressure



which is part of the reservoir system that provides drinking water to New York City.

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vessel internal components in GALL when the State presented its initial contentions, the USNRC Staff have subsequently revised the GALL Report and have advised applicants to address this important aging-related safety concern to help assure the health and safety of the American public during extended plant operations. [See GALL Report, NUREG-1801, Rev. 2, (December 2010)] Also, the new safety analyses (i.e., in plant-specific Safety Evaluation Reports) associated with the extended operations of the two Indian Point reactors do not currently include an analysis of the synergistic effects of embrittlement, fatigue, and corrosion, 3 including: irradiation-assisted stress corrosion cracking (IASCC), and primarily water assisted stress corrosion cracking (PWSCC) of affected structures, components and fittings both inside and outside the reactor pressure vessels (RPVs). This is a very important oversight which must be corrected. Indeed, Entergy needs to develop an acceptable AMP to manage the degradation of such components during periods of extended reactor operation. In contrast, Entergy has simply proposed to follow industry-developed guidelines [MRP-227] or a revised version of those guidelines [MRP-227A] that will be published later. Unfortunately, it appears that virtually none of my concerns are addressed in these guidelines.



3 To avoid repetition, it should be assumed (when not specifically mentioned), that various stress corrosion cracking (SSC) mechanisms are also a concern when considering the aging- related degradation of nuclear reactor structures, components, and fittings.

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EMBRITTLEMENT (Amended Contention #25)

17. Let us now focus on my specific concerns associated with how radiation-induced embrittlement issues have been treated by Entergy. As previously discussed in my initial November 2007 declaration [¶¶ 6-18], one of the key aging-related phenomena that must be considered in Entergys License Renewal Application (LRA) is the embrittlement of the reactor pressure vessels (RPVs) internal metal structures, components and fittings (i.e., the RPV internals).

This embrittlement occurs due to the extended irradiation (i.e., the neutron fluence, which is the neutron flux times the duration of the irradiation process) that will be experienced by the various stainless steel and nickel-based alloy metal components, particularly those located within the so-called belt line region of the RPV (i.e., the region of the RPV that is closest to the core) where the neutron flux is the highest

[WOG Letter, 3/01]. In addition, RPV internals may experience flow/thermal-transient-induced fatigue degradation, as well as degradation due to various radiation damage mechanisms [Was, 2007], including, but not limited to, the damage due to void swelling which may occur because of transmutation and other effects [Was, 2007; NUREG/CR-6897; Barnes, 1964]. Also, some in-core components may experience irradiation-assisted stress corrosion cracking (IASCC) [WCAP-14577, Rev. 1-A, pgs. 3-6 & 3-8, 2001; Tang, MRP-191, Fig. 3-1, 2006], and/or primary water enhanced stress corrosion cracking (PWSCC) [WCAP-14577, Rev. 1-A, pgs. 3-4 & 3-5, 2001] due to prolonged exposure to the high temperature (i.e., T >

400°F) borated primary coolant. In addition, cast austenitic or martensitic stainless Report of Richard T. Lahey, Jr.

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steel reactor components (e.g., some reactor piping/fittings, pump casings, pressurizer spray heads, etc.), and various cast austenitic stainless steel reactor pressure vessel (RPV) internals (e.g., upper mixing vanes, and upper/lower core assemblies and support columns) that are composed of a duplex stainless steel which contains both austenitic and ferritic phases, are subject to embrittlement due to thermal aging [WCAP-14577, Rev. 1-A, pgs. 3-12 & 3-13, 2001; EPRI Report TR-106092, 1997; NUREG/CR-4513, Rev. 1]. Moreover, the heat affected zone (HAZ) of stainless steel and nickel-alloy weldments may be even more sensitive to various corrosion and radiation-induced embrittlement mechanisms than the base metals being joined [Hawthorne et al., 1986; NUREG/CR-6960; Carey, 2006; Tang, MRP-175, 2005].

18. In any event, embrittlement causes metals to lose ductility and become more susceptible to failures due to cracking and fracture. Also, reactor operations may be restricted by embrittlement since the temperature at which the embrittled metal structures and fittings change from non-ductile to ductile behavior (i.e., the so-called nil ductility temperature, NDT) will increase as the reactor operates over time and ages. Thus, during reactor operations, the temperature at which the embrittled metal structures, components and fittings change from non-ductile to ductile behavior will increase. Significantly, this phenomenon implies that embrittled RPV internals will become progressively more vulnerable to failure due to thermal shocks as reactor operations continue. Obviously, irradiation damage is a serious age-related phenomena, and one that will not be annealed-out (i.e., it will Report of Richard T. Lahey, Jr.

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not be healed) during reactor operations since PWR operating temperatures are too low for this to occur.

19. How the rather complex, interacting metal degradation mechanisms associated with fatigue and irradiation interact is still an area of research (e.g., how fatigue-induced surface cracks propagate in an embrittled, as opposed to ductile, metal structure). Nevertheless, it is well known that, the effects of embrittlement, especially loss of fracture toughness, make existing cracks in the affected materials and components less resistant to growth [USNRC Letter, Grimes to Newton, 2/8/01, pg. 14]; indeed, it is well known that, irradiation embrittlement decreases the resistance to crack propagation, [WOG, WCAP-14577 Rev. 1-A Rpt.,

3/2001, pg. 3-2]. Anyway, the radiation-induced damage to RPV internals can be extensive, since they can experience a neutron fluence of at least 1023 n/cm2 at neutron energy (E) levels of E > 1 MeV (i.e., > 100 dpa) 4 [Was, 2007; EPRI (Dyle),

2008; WCAP-14577 Rev. 1-A Rpt.] by the end of life (EOL) for extended operations.

It should be noted that the EOL Charpy impact Upper Shelf Energy (USE) for some thermally-aged cast stainless steel in-core components could be as low as 28 ft-lbf

[WCAP-14577 Rev. 1-A Rpt., pg. 3-13], which is well below the acceptable ASME code-specified minimum of 50 ft-lbf, and even the 43 ft-lbf variance proposed by Westinghouse [WCAP-13587, Rev. 1, 1993], and subsequently endorsed by the ACRS [ACRS Letter, 9/23/09], as being acceptable for the steel Indian Point reactor pressure vessels (RPVs) at end of life (EOL) for extended operations.



4 Displacements per atom (dpa) is a measure of radiation damage to a material.

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20. Given that a variance on embrittlement from the applicable ASME Code is needed for the RPV itself, it is noteworthy that the RPVs inner wall experiences much less fluence than many of the metal structures, components, and fittings that are inside the RPV, and in much closer proximity to the core and fuel rods. That is, the inner walls of the Indian Point RPVs are expected to experience ~

1.9 x 1019 n/cm2 by the end of extended operations [Entergy Letter, 3/8/10], which, as noted previously, is much less than the ~ 1023 n/cm2 fluences expected to be experienced by some RPV internals. The obvious conclusion is that some RPV internal structures, components and fittings will be significantly embrittled during the period of extended operations of IP-2 and IP-3, and much more so than the RPV inner wall, which historically has been the focus of the USNRC. Some RPV internal components which are particularly vulnerable to radiation-induced embrittlement include the: core baffle, intermediate shells, former plates and bolts (especially the re-entrant corners), and including the baffle-to-baffle bolt locations, the core barrel-to-former bolt locations and baffle-to-former bolt locations, core barrel (and its welds), lower core plate and support structures, clevis bolts, fuel alignment pins, the thermal shield, and the lower support column and mixer. As discussed below, vulnerable in-vessel components also include the control rods and their associated guide tubes, plates, pins and welds (i.e., the so-called internal J-welds).

21. Entergy has acknowledged that, PWR internals aging degradation has been observed in European PWRs, specifically with regard to the cracking of baffle-former bolting [NL-10-063, at pg. 89]. Moreover, EPRI indicates that, A Report of Richard T. Lahey, Jr.

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considerable amount of PWR internals aging degradation has been observed in European PWRs [EPRI MRP-227, at A-4]. Entergy also states that, As with other U.S. commercial PWR plants, cracking of baffle former bolts is recognized as a potential issue for the (Indian Point) units, [NL-10-063, at pg. 890]. Moreover, materials degradation has also been observed in control rod guide tube alignment (split) pins, [EPRI MRP-227, at A-4], and degradation due to irradiation-assisted stress corrosion cracking (IASSC) has led to the failure of control rod blades and a reduction in control rod worth (i.e., the ability to absorb neutrons and thus control the nuclear reaction rate) when they were subjected to thermal shocks [NRC Information Notice 2011-13 (June 29, 2011)].

22. As noted previously, any degradation in ductility will adversely affect the possible pressure-temperature (p-T) operating conditions (i.e., there will be an increase in the nil ductility temperature, NDT). Also, it will adversely affect the ability of embrittled in-core components to withstand thermal shock transients and the decompression shock loads associated with a postulated design basis accident (DBA) loss of coolant accident (LOCA). Significantly, the metal structures, components and fittings within the RPV are subjected to many of the same transients (e.g., a SCRAM - a rapid insertion of the control rods causing a rapid decrease in the cores power level), which are known to cause fatigue-induced degradation of the primary piping, nozzles and structures [see, e.g., September 8, 2010 Declaration of Richard T. Lahey, Jr., submitted to ASLB on metal fatigue].

Hence, fatigue will also degrade the strength and ductility of many of the highly Report of Richard T. Lahey, Jr.

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embrittled metal structures, components and fittings within the RPV.

Unfortunately, virtually no fatigue analyses of RPV internals have been presented by Entergy in their application for the extended operations of IP-2 and IP-3 [see, e.g., September 8, 2010 Declaration of Richard T. Lahey, Jr., submitted to ASLB on metal fatigue]. Moreover, the aging management program (AMP) proposed by Entergy in 2010, via their amendment to the license renewal application [NL 063], does not discuss an analysis of the synergistic impacts of these interacting aging effects (i.e., fatigue and embrittlement). I believe that this is a very serious omission and one that must be corrected prior to the relicensing of Indian Point Unit 2 & Unit 3.

23. In addition, severe pressurized thermal shocks can occur on the primary side of a PWR during postulated accidents which may rapidly depressurize the secondary side of the reactor system and cause a SCRAM. While the pressured thermal shock of the reactor pressure vessel (RPV) itself was discussed in Entergys re-licensing applications, there was no indication of what new accident analysis was done (if any) in which both embrittlement and fatigue were explicitly taken into account when assessing the effect of the accident-induced transient loads on RPV internals. This is quite important since thermal shock may cause highly embrittled and fatigued RPV internal structures, components, and fittings to fail, perhaps leading to an uncoolable core geometry and core melting. Another serious omission from the USNRCs GALL Report and Standard Review Plan is that there was no mention at all of how highly embrittled and fatigued RPV internal structures, Report of Richard T. Lahey, Jr.

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components and fittings will respond to the severe transient decompression and thermal shock loads associated with a DBA LOCA, and the subsequent discharge into the primary side of the reactor of relatively cold emergency core coolant (ECC) from the accumulators. For example, it is well known [e.g., Tong & Weisman, pgs.

147-149, 1970] that a strong decompression shock wave, created during the subcooled blowdown phase of a DBA LOCA, can cause significant transient pressure differentials across various internal RPV structures. Detailed experiments (e.g.,

LOFT) and analyses have shown that, when ductile, these in-vessel metal structures are not likely to deform or fail to the point where a coolable geometry can not be maintained for the core. In contrast, no such experiments and analyses have been presented by Entergy to justify that highly embrittled and fatigued in-vessel structures, components and fittings will not fail, and that a coolable core geometry will be maintained subsequent to a DBA LOCA. This is a very serious and, in my opinion, a totally unacceptable omission since brittle and fatigue-weakened structures are known not to tolerate shock loads well (e.g., they may break loose or fracture) and, if a coolable geometry of the core is not maintained, it can melt, releasing a significant amount of radiation and possibly causing a breach of the lower head of the RPV. It is incumbent on Entergy to prove this will not happen, since Federal regulation, 10 C.F.R. § 54.4(a), clearly states that reactor operators must: provide the capability to shut down the reactor and maintain it in a safe shut-down condition. It is important to stress that while the USNRC Staff noted in their review of the Safety Evaluation Reports (SERs) for IP-2 & 3 that, if certain Report of Richard T. Lahey, Jr.

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reactor vessel internals failed, they could potentially inhibit core coolability during an accident [NUREG-1930, SER, Dockets 50-247 and 50-286, pg. 2-40, (Nov.

2009)], their primary focus was on the reactors sample tubing systems, and did not encompass more critical RPV internal structures, components, and fittings such as those noted previously in paragraphs 16 & 20 (above). Also, the industry programs, which Entergy has proposed to follow under the AMP for RPV internals [MRP-227, MRP-227A (to be published), MPR-228, and the USNRC Staff's Safety Evaluation Reports, NUREG-1930, and NUREG-1930 Supplement], are mute on the serious age-related safety concern of the coolability of PWR cores subsequent to shock load induced failure of highly embrittled and fatigued RPV internals, and thus they do not address my concerns. Indeed, the aging management program (AMP) submitted by Entergy [NL-10-063 (July 14, 2010)] and related reports [USNRC Staff RIS-2011-07 (July 21, 2011)] and a proposed inspection plan proposed by Entergy [NL-11-107 (Sept. 28, 2011)] do not address, or even mention, the synergistic aging effects of embrittlement and fatigue on RPV internals and the impact of accident-induced shock loads on these components. Likewise, the initial version of the industry initiative [MPR-227], is focused on nondestructive evaluation (NDE) and nondestructive testing (NDT) methods and procedures, and thus does not address any of my aging-related safety concerns for RPV internals.

Additionally, it appears that the USNRC Staff's review of MRP-227 [NRC - FSE &

Letter Nelson to Wilmshurst, (June 22, 2011)] also does not address my aging-Report of Richard T. Lahey, Jr.

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related safety concerns for RPV internals. Furthermore, based on the FSAR letter, it does not seem likely that the to-be-issued MRP-227A will address these issues.

24. I believe the USNRC has made a major error in not highlighting these aging-related safety issues in the GALL Report [NUREG-1801, Rev. 1, NUREG-1801, Rev. 2] and the Standard Review Plan [NUREG-1800, Rev. 1, NUREG-1800, Rev. 2]. Perhaps this is because of siloing of the safety evaluations and the various AMP issues, in which each was discussed and analyzed separately, and thus the integrated and synergistic effect of accident-induced shock loads on highly embrittled and fatigued RPV internals was not considered. In fact, the aging phenomena of embrittlement and fatigue acting together, and in concert with one another, has apparently not been considered at all in the development of the AMPs for IP-2 and IP-3. It appears that this might have occurred because of some confusion associated with the USNRCs leak-before-break (LBB) ruling

[NUREG/CR-4572, NUREG-1061, Vol. 3, 10 CFR 50, Appendix-A], in which the USNRCs rules were changed for some of the dynamic ex-vessel LOCA loads (i.e., for the pipe whip and jet loads) associated with the design basis accident (DBA). In particular, the USNRC now allows reactor operators not to use the ex-vessel loads associated with a double-ended pipe break if they can show that the primary side piping would be expected to leak well before it breaks. It is significant to note, however, that the LBB ruling does not apply to the in-vessel DBA LOCA decompression and thermal shock loads. Unfortunately, the implications of this ruling have apparently been misunderstood by many in the nuclear industry, and it Report of Richard T. Lahey, Jr.

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appears to have even led USNRC staff to not be overly concerned about the effect of DBA LOCA decompression-induced shock loads, and emergency core cooling system (ECCS) or secondary side LOCA induced thermal shock loads, on the highly embrittled and fatigued metal components, structures and fittings within the RPV.

Also, as noted previously, it appears that this significant safety issue has been totally overlooked in the Standard Review Plan (SRP) for Licensing Renewal Applications (LRAs).

Integrity of the Control Rods, Their Associated Guide Tubes, Plates, Pins and Welds

25. Any aging management program concerning the embrittlement of reactor pressure vessel (RPV) internals should include the control rods,5 and their associated guide tubes, plates, pins, and welds within the scope of such a program.



5 Although it appears that the USNRC guidance may exclude "reactivity control assemblies [as] . . . not subject to an AMR" (RIS 2011-07 at 3; see also MRP-227 at v), other aspects of the control rod system are included within the scope of Part 54 and reactor vessel internals. GALL Revision 2 includes control rod guide tube (CRGT) assemblies and welds within the scope of reactor pressure vessel internals.

GALL, Rev. 2 at § IV B2, p. IV B2-1; id. at p. IV B2-6. Additionally, the Standard Review Plan Revision 2 states "Cracking due to stress corrosion cracking and fatigue could occur in nickel alloy control rod guide tube assemblies, guide tube support pins exposed to reactor coolant, and neutron flux." SRP, Rev. 2 at § 3.1.2.2.13, p. 3.1-7. SRP also states "The GALL Report recommends further evaluation of cracking due to stress corrosion cracking and fatigue in the nickel alloy control rod guide tube assemblies, guide tube support pins exposed to reactor coolant, and neutron flux." SRP, Rev. 2 at § 3.1.3.2.13; p 3.1-13. Thus, it appears that all the RIS was intending to say, as does MRP-227, is that control rod assemblies are not included within the scope of those documents. As I discuss, this is a serious and indefensible omission. The effect of shock loads on control rods and their related assemblies, and the ability to maintain a coolable core geometry was apparently not considered in this scoping decision.

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The control rods and their associated guide tubes, plates, pins, and welds are very important RPV internals and their integrity is an extremely important safety concern. They are located in the core region of the RPV, and the control rods are inserted into the RPV through the upper head via so-called stub tubes. The function of the control rods is to absorb excess fission neutrons (i.e., those not needed to achieve a chain reaction) so that the power level of a reactor can be controlled.

26. With respect to the control rod assemblies, of particular concern is the significant and reoccurring stress corrosion cracking that has been observed in the J-groove seal welds on the control rod drive (CRD) stub tube penetrations of the upper head of PWR RPVs. By way of example, recently the operator of the Davis-Besse reactor, found evidence of boric acid deposits and indications of primary water stress corrosion cracking in their nozzles and welds[NRC Staff, Division of Components Integrity-NRC Perspectives on PWR Materials Issues, at 7 (June 2010)

ML101520577]. According to the USNRC, the timing and extent of cracking was unexpected [Id.; see also NRC News, III-10-123 (May 26, 2010)]. It is very significant to note that this type of leakage had been found earlier (i.e., in 2002) at Davis Besse, and it nearly resulted in a major LOCA due to a massive corrosion-induced failure of the upper RPV head. In fact, the stress corrosion cracking of these type CRD welds is considered to be one of the single biggest challenges currently facing operating PWRs [NEI 03-08 (Addenda), at D-5 (June 2009)].

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27. In addition, because of geometric considerations, many PWRs (including IP-2 and IP-3) can not meet the USNRCs required minimum coverage for the non-destructive testing (NDT) of these important J-groove welds [Entergy, Walpole, NL-09-130 (Sept. 24,2009)], and thus the integrity of these CRD stub tube welds can not be directly confirmed by inspection. It appears that to help address this chronic problem Entergy has ordered two new RPV heads [Telecon-USNRC/Entergy Report, March 18, 2008], but they have not yet been scheduled for installation at Indian Point [Telecon-USNRC/Entergy, March 18, 2008]. In any event, unlike the rather superficial treatment given this important safety concern by Entergy [NL-10-063], I believe that a tangible, enforceable, and viable aging management program (AMP) should be developed and implemented before re-licensing the Indian Point reactor plants for extended operations, since the integrity of these CRD welds must be assured. If not, due to the leakage of borated primary coolant through cracked welds, there can be aggressive corrosion and wasting of the unclad outer surface of the upper head of the RPVs (such as the serious event that occurred at Davis-Besse and was identified in 2002). Worse yet, there might be an inadvertent control rod ejection (due to a massive failure of the welds in the upper RPV head), which could cause a significant reactivity excursion, leading to core melting and radiation releases.

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Baseline Inspections

28. With respect to Entergys proposal to conduct baseline examinations of the RPV internals, it should be noted that I have previously called on Entergy to conduct such examinations and for USNRC Staff to require the conduct of such examinations before entering the period of extended operations [See November 2007 Declaration of Richard T. Lahey, Jr., at ¶¶ 24, 25; see also State of New York Notice of Intention to Participate and Petition to Intervene, at pgs. 217-220, State of New York Contention-23 (Baseline Inspections)]. 6 Fortunately, both the USNRC and Entergy now seem to have embraced the concept of baseline inspections for RPV internals, but the proposed aging management program (AMP) as set forth in NL-10-063 lacks sufficient details to know when the baseline inspections of the RPV and its internals will begin and end, and the scope of these inspections.
29. In any event, as I have discussed previously, I do not believe that Entergy has developed adequate AMPs to address the embrittlement of RPV internals, including the integrity of the control rods and their associated guide tubes, plates, pins and associated welds, and the baseline inspections for IP-2 and IP-3. This is very significant since I believe that a program to identify, and repair or replace, seriously degraded RPV internals is needed to adequately protect the health and safety of the American public during extended operations.



6 Interestingly, in early 2008 during the initial phase of this proceeding, both Entergy and the USNRC Staff opposed this proposal and NYSs Contention-23, which asserted that Entergy should conduct baseline inspections to determine the present condition of important systems, structures, components and fittings.

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FATIGUE (Amended Contention #26)

30. Let us next focus on New York's amended contention 26, which has been consolidated with Riverkeeper's contention TC-1, and concerns fatigue-induced degradation mechanisms. Fatigue is obviously a very important age-related safety concern, particularly when a significant plant life extension is being considered. In fact, it is one of the primary things that must be considered when doing a time-limited aging analysis (TLAA) or developing a plant-specific aging management program (AMP) for the extended operation of a nuclear reactor. A common figure of merit used in the American Society of Mechanical Engineers (ASME) code (Section-III) to appraise the possibility of fatigue failure is the cumulative usage factor (CUF), which is the ratio of the number of cycles experienced divided by the number of allowable cycles. The maximum number of cycles which can be experienced by a structure or component before significant cracking is expected occurs when CUF =

1.0, and one must have CUF < 1.0 during the period of plant operation. In addition, since the high pressure/temperature primary coolant is known [e.g., NUREG/CR-6909] to degrade the fatigue life of immersed metal structures, components and fittings, the USNRC also requires that environmental corrections be applied to the calculated CUF, and it specifies the formulas/curves to be used for these corrections

[e.g., NUREG/CR-5704; NUREG/CR-6583]. Significantly, CUFen >CUF and the environmentally-adjusted fatigue analyses must satisfy CUFen < 1.0 during Report of Richard T. Lahey, Jr.

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extended plant operations. 7 It is very important to note that these fatigue evaluations do not take into account that the structures, components and fittings being analyzed may have experienced significant corrosion and irradiation-induced embrittlement, and thus can experience early fatigue-induced failures.

31. In the original relicensing submittal for Indian Point Units 2 & 3, Entergy analyzed typical limiting PWR structures and fittings using some of those given in NUREG/CR-6260 [pg. 5-62], and this analysis showed that several important structures and components will significantly exceed the environmentally-adjusted CUFen = 1.0 criterion during the proposed extended operations period. In particular, the pressurizer surge line and nozzle, and the reactor coolant system charging system nozzle (on the primary side), and the steam generator main feed water nozzles and tube/tube-sheet welds (on the secondary side), and the upper joint canopy of the IP-2 control rod drive (CRD) mechanisms, all had unacceptably high CUFen (e.g., CUFen > 9.0 for the IP-2 and IP-3 pressurizer surge lines and CUFen >

15 for the IP-2 RCS charging system nozzle [LRA-Section 4]). Once CUFen violations of this type are found, Entergy was expected [NUREG-1801, Rev.1, Vol.2, pg.XM-2 ; EPRI, MRP-47; Guidelines for Addressing Fatigue Environmental Effects in a Licensing Renewal Application, pg. 3-4 (2005)] to also do fatigue analyses for other potentially limiting reactor structures, components and fittings.



7 The subscript en denotes an environmentally-adjusted cumulative usage factor, CUFen.

Report of Richard T. Lahey, Jr.

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However, a systematic fatigue analysis of other structures, components and fittings was apparently not done.

32. Nevertheless, in order to perform more mechanistic, but less conservative, fatigue evaluations, Entergy contracted with Westinghouse (W) to redo the ex-vessel fatigue analyses for IP-2 and IP-3. These results were reported separately [Entergy NL-10-082 discussing WCAP-17199-P / WCAP-17200-P, Environmental Fatigue Evaluation for Indian Point Unit 2/3 (June 2010)].8 Rather than performing standard ASME code evaluations, as Entergy had done before, these calculations were done using WESTEMS, a proprietary computer code of W. Subsequent to submitting these calculations to the ASLB and some other interested parties, the ASLB directed Entergy to release the WESTEMS code manuals to New York State (NYS) for review. The parts of the manuals transmitted to us contained assumptions and models (particularly for the thermal-hydraulics models) used by W in WESTEMS.
33. Anyway, the new CUFen results filed with the ASLB by Entergy

[Applicants Motion for Summary Disposition of NYS Contentions 26/26A &

Riverkeeper Technical Contentions 1/1A (Metal Fatigue of Reactor Components)

(August 25, 2010)] show that the previously most limiting CUFen were reduced by more than an order of magnitude (e.g., the results for the pressurizer surge line



8 Entergy disclosed the results of Westinghouse's refined WESTEMS calculations in Entergy's NL-10-082 communication and its summary disposition motion. NL 082, Attachment 1, Table 4.4-13, Table 4.3-14; Declaration of Nelson Azevedo, at ¶¶ 39, 50 (Aug. 20, 2010) (included in ML102600058).

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piping and the RCS piping charging system nozzle), which is a very significant reduction, and one that must be very carefully verified since it significantly reduces the design safety margins implicit in the original ASME code evaluations.

Additionally, for the first time, limiting fatigue analysis results were given for the residual heat removal (RHR) system piping and nozzles, and the results for these components were very close to the unity limit. In particular, for the IP-2 RHR line, CUFen = 0.9434, and for the IP-3 RHR line, CUFen = 0.9961. Thus, virtually any error in these results could lead to a violation of the USNRCs CUFen = 1.0 limit.

34. Unfortunately, an error analysis of the WESTEMS results was not made available to the State of New York by either Entergy or Westinghouse, nor were any results provided showing that the computational results exhibited nodal convergence, or how they were bench-marked against representative experimental data and/or analytical solutions. One would normally expect to see a detailed propagation-of-error type of analysis [e.g., Vardeman & Jobe, Basic Engineering Data Collection and Analysis, Duxbury, pp. 310-311 (2001)] to determine the overall uncertainty in the CUFen results given by W. It is well known that all engineering analyses are based on imperfect mathematical models of reality and various code user assumptions which inherently involve some level of error. In addition, as the USNRC Staff confirmed in the SSER [NUREG-1930, Supplement 1 at 4-2], WESTEMS permits the code user to make assumptions and interventions that can affect the outcome. As a consequence, without a well-documented error analysis, the accuracy of Entergys new fatigue results are quite uncertain and thus Report of Richard T. Lahey, Jr.

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cannot be used to establish the integrity of the structures, components or fittings being analyzed. As I discuss in detail in another companion report, what is abundantly clear is that there are many possible sources of error in the results that Entergy (and Westinghouse) have provided to the ASLB and the parties. (This additional companion report includes a discussion that Entergy and Westinghouse may consider as falling within the scope of the Confidentiality Order in this proceeding and thus will be submitted separately in support of Contention NYS-26).

35. Earlier this year, the USNRC Staff required Entergy to disclose and make clear those user interventions that will be used in future WESTEMS analysis of IP-2 and IP-3 [SSER, NUREG-1930, Supplement 1 at 4-2], but, for some reason, this disclosure was not required for the results that Entergy has already submitted to the ASLB for IP-2 and IP-3. Obviously, this information is needed in order to do a proper review of these important results.
36. It is apparent that the WESTEMS code is based on rather simple models (particularly the thermal-hydraulic models) and that the thermal stress results for CUFen are strongly influenced by the code users assumptions, manipulations and interventions. There is a lot of engineering judgment implicit in the CUFen results, and, since an error analysis has not been done to bound the uncertainty, and many results are disturbingly close to the CUFen = 1.0 limit, I do not believe that one can trust these results to assure the safety of the IP-2 and IP-3 during extended plant operations. Indeed, these results are quite uncertain and this uncertainty should be quantified by doing parametric runs and a detailed error Report of Richard T. Lahey, Jr.

28

analysis. Moreover, because the effect of various shock loads on the failure of these fatigue-weakened components, structures, and fittings has not been considered, it is unclear that the health and safety of the American public is being adequately protected. As previously noted, these results are quite uncertain and this uncertainty needs to be quantified by doing a detailed error analysis.

37. It is also significant to note that in-core fatigue failures of irradiated baffle-to-former bolts have been observed in operating PWRs [e.g., WCAP-14577, Rev. 1, License Renewal Evaluation: Aging Renewal Evaluation: Aging Management of Reactor Internals, pg. 2-29 (Oct. 2000); USNRC Staff Report, Final Safety Evaluation by the Office of Nuclear Reactor Regulation Concerning Westinghouse Owners Group Report, WCAP-14575, Revision 1, License Renewal Evaluation: Aging Management for Class 1 Piping and Associate Pressure Boundary Components, Project No. 686, (Nov. 8, 2000)], and B&W designed PWRs have had fatigue-induced failures of various in-core components even when CUF <

1.0 (perhaps due to undetected manufacturing flaws) [Entergy Email: Esquillo to Stuard et al.,

Subject:

Section XI - Cracking (8/30/06)]. Significantly, the possible effect of fatigue on the failure of RPV internals was apparently well known to Entergy [Entergy Email: Batch to Finnin,

Subject:

Need to Evaluate High Cycle Fatigue to IPEC Baffle Bolts? (12/28/06)]. Moreover, unlike postulated nuclear reactor accidents, the fatigue failures of in-core bolts are actual events that have happened and will likely happen again for sufficiently stressed materials. Also, it is not possible to inspect (e.g., using UT techniques) all the bolts within a RPV, and Report of Richard T. Lahey, Jr.

29

thus the nuclear industry has recommended [EPRI Report, MRP-228; Materials Reliability Program: Inspection Standard for PWR Internals, (July 2009)] that an analysis be done to support continued operations if bolt failures are found during in-core non-destructive evaluations (NDE). However, it appears that these analyses do not take into account the possibility of various accident-induced pressure and/or thermal shock loads within the RPV, such as those due to a DBA LOCA. Thus, the number of intact bolts which might be adequate for normal operations may be totally inadequate to accommodate shock loads during accidents. In any event, I believe that not doing a realistic safety analysis is totally unacceptable since shock-load-induced bolting failures may lead to a blocked or distorted core geometry which, in turn, may not allow the ability to adequately cool the core and can lead to core melting.

38. As for all mechanical systems, when nuclear power plants exceed their original design life (i.e., 40 years) they begin to wear out and thus, to assure safe operation during plant life extension, it is important not to erode the original design-basis safety factors in the interest of keeping the plants running. In particular, in addition to the previously discussed in-core bolting fatigue failure concerns, many other highly irradiated in-core structures, components and fittings (e.g., core baffles, formers, etc.) and welds (e.g., on thermal shields [WOG, 3/01]) will be subjected to some of the same (and even more) fatigue-inducing transients as those which effect the structures, components and fittings that are external to the RPV (e.g., those piping systems, components and fittings that were analyzed by W).

Report of Richard T. Lahey, Jr.

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However, no fatigue analysis of these important RPV internals was provided by Entergy, and there was apparently no recognition of the importance of a DBA LOCA, secondary side LOCA and ATWS loads on the integrity of these structures, components and fittings. As for in-core bolting, I believe that not doing a proper fatigue and safety analysis of these embrittled and fatigued RPV internal structures, components and fittings is completely unacceptable since shock-load-induced failures of RPV internals may lead to a distorted or blocked core geometry, which may, in turn, not allow the ability to adequately cool the core and result in core melting. Indeed, seriously degraded RPV internals (including bolting) should be repaired or replaced in order to protect the health and safety of the American public during extended operations.

SUMMATION

39. In summary, there are some very important aging-related safety issues associated with the operation of Indian Point Units 2 & 3 during their proposed 20-year life extension. In my opinion, Entergy has not proposed adequate aging management programs (AMPs) for the embrittlement of RPV internals, the fatigue of important structures, components and fittings, both within and outside the RPV, has not yet specified an acceptable AMP for the control rod structures, components, fittings and associated welds, and the baseline inspection of RPV internals.

Moreover, Entergy has apparently not done any new safety analyses which address core coolability in light of the synergistic effects of the degradation of RPV internals Report of Richard T. Lahey, Jr.

31

due to radiation-induced embrittlement, fatigue and various stress corrosion cracking (SCC) mechanisms. In particular, these interacting degradation mechanisms may lead to the loss of an intact core geometry due to various accident-induced shock loads, resulting in significant core melting and radiation releases.

40. It appears that while many of the individual technical issues that I have raised are known to the USNRC, apparently a silo type mentality has prevented the staff from fully connecting the dots. Nevertheless, the USNRC has recognized the need to accommodate important AMPs and safety issues that were not considered in the development of the Standard Review Plan [NUREG-1800, Rev. 1, NUREG-1800, Rev. 2] and the GALL report [NUREG-1801, Rev.1, NUREG-1801, Rev.2]. In any event, as the owner and operator of the Indian Point Nuclear Reactors, Entergy has the ultimate responsibility to assure that the effects of plant aging are adequately managed [10 C.F.R. §§ 54.21 (a) (3), 54.4] in order to protect the health and safety of the American public. Thus, I believe that the focus of the ASLB hearings on Indian Point should be on this vital objective rather than on legal technicalities concerning whether or not some of the important aging-related safety issues that I have raised are currently on the USNRCs checklist for license extension.

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December 20, 2011 Dr. Richard T. Lahey, Jr.

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Friday; June 16, 2006; 10:25 AM; From: Mark A. Rinckel; To: Ron Finnin; Cc:

Alan Cox, Michael D. Stroud, Virgilio M. Esquillo, and Stan Batch;

Subject:

Section XI - Cracking.

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Allison;

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Need to Evaluate High Cycle Fatigue to IPEC Baffle Bolts? (12/28/06) and email string:

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Don Fronabarger, Ted S. Ivy;

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Need to evaluate high cycle fatigue for IPEC baffle bolts?

Friday; January 12, 2007; 10:14 AM; From: Stan Batch; To: Walter Wittich and Nelson F. Azevedo; Cc: Ron Finnin and Don Fronabarger;

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Need to evaluate high cycle fatigue for IPEC baffle bolts?

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Westinghouse, WCAP-14577, Rev. 1-A, License Renewal Evaluation: Aging Management for Reactor Internals, D.R. Forsyth et al. (March 2001).

Westinghouse (W) - Letter No. LTR-PAFM-03-42, Yang, Procedures for Transfer Function Database Creation and Guidelines for the Associated Finite Element Analysis, (March 2006).

Westinghouse Owners Group (WOG) Report on Reactor Internals, Rev. 1-A (March 2001) [IPEC00006881ff].

Vardeman & Jobe, Basic Engineering Data Collection and Analysis, Duxbury (2001).

Report of Richard T. Lahey, Jr.

39

Entergy Designated Proprietary Documents Westinghouse, WESTEMS computer code manual, version 4.5, Vol. 1, Rev. 0 (March 2007) [excerpts produced by Entergy].

Westinghouse, WCAP14173, Rev. 2, Roarty, et al., "Global to Local Transformations and Stress Transfer Functions for Pressurizer Surge Line, Pressurizer Lower Head and Pressurizer Spray Line," IPECPROP00060808 (August 1995)

Westinghouse, WCAP-12191, Rev. 3,Transient and Fatigue Cycle Monitoring Program Transient History Evaluation Report for Indian Point Unit 2 Addendum 1 (September 2003) IPECPROP00004699.

Westinghouse, WCAP-17199-P, Environmental Fatigue Evaluation for Indian Point 2, (June 2010) IPECPROP00056486.

Westinghouse, WCAP-17200-P, Environmental Fatigue Evaluation for Indian Point Unit 3, (June 2010) IPECPROP00056577.

Westinghouse, WCAP-17199-P, Environmental Fatigue Evaluation for Indian Point 2, (June 2010) IPECPROP00056486.

Westinghouse, WCAP-17200-P, Environmental Fatigue Evaluation for Indian Point Unit 3, (June 2010) IPECPROP00056577.

Westinghouse, WCAP-17149-P, Rev. 1, Evaluation of Pressurizer Insurge/

Outsurge Transients for Indian Point Unit 2, IPECPROP00056663 (July 2010).

Westinghouse, WCAP-17162-P, Rev. 1, Evaluation of Pressurizer Insurge/

Outsurge Transients for Indian Point Unit 3, IPECPROP00056717 (July 2010).

Westinghouse (W) - Calculation Note, CN-PAFM-09-21, IPECPROP0057432, Indian Point Units 2 & 3 Charging Nozzles Environmental Fatigue Evaluation (June 18, 2010).

Westinghouse (W) - Calculation Note, CN-PAFM-09-67, IPECPROP00057917, Pressurizer Surge Nozzle and Lower Head Transfer Functions for Indian Point Units 2 and 3 (June 18, 2010).

Report of Richard T. Lahey, Jr.

40

Westinghouse (W) - Calculation Note. CN-PAFM-09-77, IPECPROP00057881/3, Indian Point Units 2 & 3 Accumulator Nozzle Environmental Fatigue Evaluation (June 18, 2010).

Westinghouse (W) - Calculation Note, CN-PAFM-09-79, IPECPROP00057559, Indian Point Unit 2 Boron Injection Tank Nozzle Environmental Fatigue Evaluations (June 18, 2010)

Westinghouse (W) - Calculation Note, CN-PAFM-09-117, IPECPROP00057276, Indian Point Units 2 and 3 Hot Leg Surge Nozzle Environmental Fatigue Evaluations (June 18, 2010).

Report of Richard T. Lahey, Jr.

41

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