ML11356A427

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New York State (NYS) Pre-Filed Evidentiary Hearing Exhibit NYS000293, State of New York Statement of Position, NYS-25A (NYS-25A SOP)
ML11356A427
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 12/22/2011
From: Sipos J J
State of NY, Office of the Attorney General
To:
Atomic Safety and Licensing Board Panel
SECY RAS
Shared Package
ML11356A427 List:
References
RAS 21627, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01
Download: ML11356A427 (24)
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Text

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 22, 2011

x

STATE OF NEW YORK INITIAL STATEMENT OF POSITION CONTENTION NYS-25

Office of the Attorney General for the State of New York

The Capitol

State Street

Albany, New York 12224 State of New York Initial Statement of Position In Support of Contention NYS-25

1PRELIMINARY STATEMENT In accordance with 10 C.F.R. § 2.1207(a)(1) and the Atomic Safety and Licensing Board's ("Board") July 1, 2010 Memorandum and Orde r, the State of New York (the "State") hereby submits its Initial Statement of Position on the State's admitted Contention 25 ("NYS-

25") concerning the integrity of Indian Point's embrittled reactor pressure vessels and their internal components. Embrittlement of reactor pressure vessels and their internal components is one of the most important age-related phenomena that NRC must consider in its review of Entergy's application to extend the operating life of the Indian Point reactors for an additional twenty years. Failure to consider the effects of embrittlement and the synergistic effects of embrittlement and other age related degradation mechanisms could prevent the maintenance of a coolable core geometry which in turn, could result in a meltdown of the core, a release of radiation, and profound safety consequen ces for the State and its citizens. In this proceeding, the State of New York has satisfied the standards contained in 10 C.F.R. § 2.309 governing contention admissibil ity - standards that NRC and Entergy have described as "strict by design." The State now submits the accompanying December 20, 2011 Report of Dr. Richard T. Lahey, Jr., Edward S. Hood Professor Emeritus of Engineering, Rensselaer Polytechnic Inst itute (Exh. NYS000296), and Supplemental Report by Dr. Lahey concerning Entergy's use of the WESTEMS computer code as a tool to analyze the cumulative fatigue condition of important reactor components (Exh. NYS000297), Dr. Lahey's prefiled written testimony (Exh. NYS 000294), and related evidence to show that Entergy's license State of New York Initial Statement of Position In Support of Contention NYS-25 2renewal application ("LRA") should be denied because the application does not demonstrate that the reactor pressure vessel internals will remain functional during various design basis events as required by 10 C.F.R. §§ 50.54, 50.46, 54.3, 54.4, 54.21, 54.29, 54.33(a) , and because the

application does not provide an adequate program to monitor and manage the effects of aging degradation of the reac tor pressure vesse l internals as re quired by 10 C.F.R. §§ 54.21(a)(3), 54.21(c)(iii) and 54.29(a)(1).

1 PROCEDURAL HISTORY The State of New York's Petition to Intervene and Contention 25 On November 30, 2007, the State submitted a Petition to Intervene, which Petition included proposed contentions regarding critical deficiencies in Entergy's Indian Point relicensing application with respect to public safety, health, and the environment.

State of New York Notice of Intention to Part icipate and Petition to Intervene , (Nov. 30, 2007) ("NYS Petition") ML073400187. Among those proposed c ontentions was Contention NYS-25 that challenged Entergy's approach to the embrittlement of the reactor pressure vessel and their internal components. NYS Petition at 223-27. NYS-25, asserted: Entergy's License Renewal Application Does Not Include An Adequate Plan To Monitor And Manage The

Effects Of Aging Due To Em brittlement Of The Reactor Pressure Vessels ("RPVs") And The Associated Internals.

1 Any attempt by Entergy to remedy this deficiency in its license renewal application by amending it previous submissions should be filed w ith the Board and all pa rties should be given a reasonable opportunity to file, with the Board, new conten tions based on Entergy's revised submittals.

State of New York Initial Statement of Position In Support of Contention NYS-25

3NYS Petition at 223. NYS-25 contended that the LRA fails to include an adequate plan to monitor and manage the effects of aging due to embrittlement of the reactor pressure vessels

("RPV") and the associated internals as required by 10 C.F.R. § 54.21(a), and does not comply with 10 C.F.R. § 54.21(c).

Id. The State further contended that the LRA does not document "that Entergy performed any age-related accident analyses, or that it took embrittlement into

account when it assessed the effect of transient lo ads or the loss of coolan t accident ("LOCA")." NYS Petition at 224.

Dr. Richard Lahey, a Professor of Engineering at Rensselaer Polytechnic Institute with extensive experience in the de sign, operations, safety, and ag ing of nuclear power plants, submitted a declaration in support of various contentions presented by the State including NYS-

25. Declaration of Dr. Richard T.

Lahey, Jr., at ¶¶ 6-18 (Nov. 2008) (included in ML073400193) (Exh. NYS000298). Dr. Lahey stated that embrittlement of the RPVs and their associated internal components is one of the most important age-related phenomena, and that failure to carefully consider the effects of embrittlement could result in a meltdown of the core.

Id. at ¶¶ 6, 9. Furthermore, Entergy failed to document in its LRA any experiments or analysis to justify that the embrittled RPV internal structures will not fail and that a coolable core geometry will be maintained subsequent to a Design Basis Accident Loss of Coolant Accident.

Id. at ¶¶ 14-16. According to Dr. Lahey "[t]his is a serious and unacceptable omission by Entergy because embrittled structures are known not to tolerate shock loads well."

Id. at ¶ 16.

State of New York Initial Statement of Position In Support of Contention NYS-25 4 Entergy and NRC Staff opposed the admission of the State's contention.

2 Among other things, Entergy argued that the Board should not admit NYS-25 because "The core barrel, thermal shield, baffle plates and baffle former pl ates (including bolts) are, however, made of stainless steel and are not susceptible to a decrease in fracture toughness as a result of neutron embrittlement." Entergy Answer at 137. On February 22, 2008, the State submitted a reply in further support of Contention 25.

3 On March 11, 2008, the Board heard oral argument concerning the admission of Contention 25.

4 On July 31, 2008, the Board issued a memorandum and order rejecting Entergy's and NRC Staff's arguments and admitting contention NYS-25.

Entergy Nuclear Operations, Inc.

(Indian Point, Units 2 and 3), LBP-08-13, 68 N.R.C. 43 (July 31, 2008) ML082130436. The

decision reviewed the State's c ontention and Dr. Lahey's supporti ng declaration. 68 N.R.C. at 129-131. The Board recognized Dr. Lahey's opinion that components in the Indian Point reactors have serious embrittlement issues that are not adequately addressed in Entergy's LRA.

2 Answer of Entergy Nuclear Operations Inc. Opposing New York State Notice of Intention to Participate and Petition to Intervene (Jan. 22, 2008) (ML080300149) ("Entergy Answer"); NRC Staff's Response to Petitions for Leave to Intervene (Jan. 22, 2008) (ML080230543).

3 State of New York Reply in Support of Petition to Intervene (Feb. 22, 2008) (ML080600444).

4 Transcript of Proceedings, at 401-410 (Mar. 11, 2008) (ML080740257). The parties submitted an errata sheet, which the Board subsequently accepted, for the transcript of the three day oral argument.

State of New York Initial Statement of Position In Support of Contention NYS-25

5 Id. at 131. The Board noted:

Dr. Lahey states that Entergy fails to document in its LRA "any experiments or analysis to justify that the embrittled RPV internal structures will not fail and that a coolable core geometry will be maintain subsequent to a [Design Basis Accident] LOCA." According to Dr. Lahe y "[t]his is a serious and unacceptable omission by Entergy because embrittled st ructures are known not to tolerate shock loads well."

Id. (citations omitted).

Entergy's Revised LRA and the State's Additional Bases Regarding Reactor Vessel Internals On July 15, 2010, Entergy submitted to the Board an amendment to the LRA.

See Entergy Letter to ASLB enclosing Entergy Communication NL-10-063 (LRA Amendment No.

9) (Exh. NYS000313) (ML102030120). NL-10-063 forwarded to NRC Staff a document entitled "Reactor Vessel Internals Program" or RVI Program. Entergy stated that "[t]he RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, MRP-227." NL-10-063, Attachment 1, p. 8. On September 15, 2010, the State submitted a motion for leave to file additional bases to NYS Contention 25.

State of New York's Motion for Leave to File Additional Ba ses for Previously-Admitted Contention NYS-25, etc., (Sept. 15, 2010) (ML103050402). Dr. Lahey submitted a declaration in support of the additional bases, and identifying concerns with Entergy's NL-10-063 RVI Program. Declaration of Richard T. Lahey, Jr. (Sept. 15.

2010) (Exh. NYS000301)(included in ML103050402). Among other things, that declaration discussed the synergistic effects of embrittlement and fatigue. Lahey Sept. 15, 2010 Decl. at ¶¶ 13, 14, 15. In the September 15, 2010 declaration, Dr.

State of New York Initial Statement of Position In Support of Contention NYS-25 6Lahey referenced another declaration that he had then recently submitted in connection with NYS-26B.

Id. at ¶ 13 (referencing Declaration of Richard T. Lahey, Jr. (Sept. 8, 2010) (Redacted, Public Version) (Exh. NYS000300) (i ncluded in ML102670665) There, Dr. Lahey noted note that in-core fatigue failures of irradiated baffle-to-former bolts have been observed in operating PWRs.; Lahey Se pt. 8, 2010 Decl. at ¶ 12. Entergy and NRC Staff opposed the admission of the State's proposed additional bases.

5 On October 22, 2010, the State submitted a reply in further support of the additional bases for Contention NYS-25.

6 On April 29, 2011, the State submitted to the Board additional information that reflected concerns by NRC Staff about defici encies in the visual and remote examination techniques that Entergy and industry had proposed to employ as part of the aging management program for the embrittlement of reactor internals. Letter from AAG J. Sipos to ASLB (Apr. 29, 2011) (enclosing two memoranda documenting non-concurrences by NRC Staff members about a safety evaluation of EPRI MRP-227) (ML11133A288). In those documents, two NRC Staffers recognized that two types of examination methodol ogies (EVT-1 and UT) were superior to an alternative examination methodology (VT-3) recommended by EPRI in MRP-227-Rev. 0. R.L. Tregoning, & M. J. Case Memoranda (Exh. NYS000370).

5 NRC Staff's Answer to State of New York's Motion for Leave to File Additional Bases for Previously-Admit ted Contention NYS-25 (Oct. 12, 2010) (ML102850764); Applicant's Answer to Amended Contention NYS 25 Concerning Aging Management of Reactor Pressure Vessel Internal s (Oct. 12, 2010)(ML103010104).

6 State of New York's Joint Reply to Entergy and NRC Staff's Separate Answers to the State's Additional Bases for Previ ously Admitted Contention NYS-25 (Oct. 22, 2010)

State of New York Initial Statement of Position In Support of Contention NYS-25

7 On July 6, 2011, the Atomic Safety and Licensing Board issued a memorandum and order that, among other things, admitted the State's amended bases for Contention NYS-25 and rejected Entergy's and NRC Staff's arguments opposing the State's submission.

Entergy Nuclear Operations, Inc. (Indian Point, Units 2 and 3), Memorandum and Order Ruling on Pending Motions for Leave to File New and Amended Contentions (July 6, 2011) (ML111870344).

NRC Staff's Supplemental Safety Evaluation Report On July 27, 2011, NRC Staff notified the Board that the Staff had completed a Safety Evaluation Report on EPRI's MRP-227 document. Letter from NRC Counsel S. Turk to ASLB, (July 27, 2011) ML11208C309 (enclosing NRC RIS 2011-07, License Renewal Submittal Information For Pressurized Water Reactor Internals Aging Management ML111990086) (Exhs.

NYS000310). . RIS 2011-07 stated in part, "LRAs for pressurized water reactor (PWR) plants have identified that an aging management program (AMP) is needed to manage the effects of aging for reactor vessel internal (RVI) components that are within the scop e of license renewal, in accordance with 10 CFR Part 54, 'Requirements for Renewal of Operating Licenses for Nuclear Power Plants.'"

7 On August 31, 2011, NRC Staff informed the Board and the parties that Staff had issued a supplement to the Safety Evaluation Report c oncerning the application to renew the operating

(ML103000060).

7 In June 2011, Staff prepared a Safety Evaluation of EPRI's MRP-227-Rev 0. Exh.

NYS000309. EPRI had previously withdrawn a confidentiality designation for MRP-227. EPRI State of New York Initial Statement of Position In Support of Contention NYS-25 8licenses for Indian Point Unit 2 a nd Indian Point Unit 3. Letter from NRC Staff Counsel S. Turk to ASLB (ML11243A109) (enclosing NUREG-1930, Supplement 1 (SSER)) (Exh. NYS 000130). The State and Riverkeeper's Joint Contention NYS-38/RK-TC-5 On September 30, 2011, the State and Riverkeep er jointly filed a contention challenging the failure of various proposed Aging Management Programs to provide sufficient details about their contents and methods.

State of New York and Riverkeep er Joint Contention NYS-38/RK-TC-5 (Sept. 30, 2011) (ML11273A196). Entergy and NRC Staff opposed the admission of the proposed contention. On November 1, 2011, the State and Riverkeeper submitted a reply in further support of Contention 38. On November 10, 2011, the Atomic Safety and Licensing Board issued a decision and order admitting Contention 38 and rejecting Entergy's and NRC Staff's arguments opposing the contention.

Entergy Nuclear Operations, Inc. (Indian Point, Units 2 and 3), Memorandum and Order Ruling (admitting New Conten tion NYS-38/RK-TC-5) (ML11314A211).

Over the last three years, the State and Entergy have disclosed various documents concerning the aging of reactor vessel internals. NRC Staff has provided limited disclosures concerning documents within the possession of NRC concerning aging of reactor internals. Further, the State has identified documents reflecting recently acknowledged concerns on the part of NRC and industry about age-related degr adation of reactor pre ssure vessel internals caused by embrittlement, fatigue, and corrosion. For example, at a recent conference in Indian,

MRP 2010-016 letter (Mar. 2, 2010) (Exh NYS 000308).

State of New York Initial Statement of Position In Support of Contention NYS-25

9NRC Staff highlighted concerns over aging management of reactor pressure vessel internals and other components. Slides, NRC Aging Management Program Including Long Term Operation (LTO), Workshop on Challenges on the Long Te rm Operation, vg5, 10, 21, New Delhi, India (ML111801154) (Nov. 8-11, 2011) (Exh. NYS000305).

SUMMARY

OF ARGUMENT Contention 25 is based on (i) 10 C.F.R. § 54.4(a)(1), which requires that the applicant have an aging management plan ("AMP") that will ensure the following functions: (i) The integrity of the react or coolant pressure boundary; (ii) The capability to shut down the reactor a nd maintain it in a safe shutdown condition;

[and]

(iii) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparab le to those referred to in § 50.34(a)(1), §50.67(b)(2), or § 100.11 of this chapter, as applicable;

(ii) 10 C.F.R. § 54.21(a)(3), which requires that the applicant "demonstrat e that the effects of aging will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis ("CLB") fo r the period of extended operation;" and (iii) 10 C.F.R. § 54.21(c) (iii), which specifically re quires that "the applicant [Entergy] shall demonstrate that . . . (iii) [t]he effects of aging on the intended function(s) will be adequately managed for the period of extended operation." Entergy has failed to make the required demonstrations and, thus, the Board cannot find, pursuant to 10 C.F.R. § 54.29(a)(1), that:

State of New York Initial Statement of Position In Support of Contention NYS-25 10 there is reasonable assurance that the ac tivities authorized by the renewed license will continue to be conducted in accordance with the CLB . . . [including] (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under §54.21(a)(1)

The focus of Contention 25 is reactor pressure vessel internals ("RPVI") and Entergy's deficient AMP for RPVIs. The principal suppor t for Contention 25 is the report and prefiled testimony of Dr. Richard T. Lahey, Jr., the Edward E. Hood Professor Emeritus of Engineering at Rensselaer Polytechnic Institute (RPI) in Troy, New York, and numerous documents, all of which demonstrate that: (i) Entergy's AMP for RP VIs is not based on an an alysis that addresses the critical issue of the synergistic degradation of RPVIs caused by the combination of embrittlement, metal fatigue, irradiation-assisted stress corrosion cracking ("IASCC"), and primarily water assisted stress corrosion cracking ("PWSCC"); (ii) Entergy's analysis fails to adequately consider the full range of transient shock loads (thermal and decompression) to which RPVIs will be subjected in the event of various postulated accidents, such as a design basis accident ("DBA"), and thus fails to develop a plan which considers those shock loads, and their resultant impact on core coolabil ity, in setting either inspection, acceptance or corrective action criteria; (iii) the AMP lacks sufficient detail regarding the timing and nature of the inspections to be conducted, or proposes to conduct such inspec tions too late, both with regard to baseline inspections, and inspections during proposed extended operation, to allow Entergy to demonstrate that it has an effective inspection program; and () the AMP does not include a commitment to take preventative actions or to implement corrective actions, nor does it provide State of New York Initial Statement of Position In Support of Contention NYS-25

11specific, enforceable acceptance criteria for many components and contains no criteria to determine when to repair or replace embrittled components. Report of Dr. Richard T. Lahey, Jr. (Exh. NYS000296). The State also reli es on Dr. Lahey's previous de clarations that have been filed in this proceeding, as well as NRC, Entergy, and industry documents (Exhs. NYS000300, 301). .

Dr. Lahey demonstrates the importance of RPVI and explains how the failure of RPVIs during a DBA may result in the loss of a coolable geometry and thus core meltdown and a catastrophic release of radiation to the environment. Many of the RPVIs hold critical reactor components in place and their failure, during an accident, can cause those components, such as baffles, former plates, core support plates, thermal shields, control rods, gui de tubes, plates and welds to either deform and/or create core blockages within the RPV pressure boundary.

See NYS000296 at ¶¶ 12-13, 16-28, 38-40. Entergy's response to New York State's C ontention 25 has evolved from an initial LRA that contained no AMP for RPVIs, to the current proposal that mostly includes adoption of the Electric Power Research Institutes ("EPRI

") MRP-227 (Exh. NYS00307A-D)inspection plan with a representation that, in the future, it w ould endeavor to develop some modifications to MRP-227 that are required to comply w ith concerns raised by the NRC.

See LRA at 3.1.2.1.2. (which does not list an AMP for RPVIs); NL 063, Attachment 1 at 7 (which includes an AMP for RPVIs) (Exh. NYS000313); NL-11-107 Attachment 1 (whi ch purports to include the inspection plan for RPVIs)(Exh. NYS000314); a nd NL-11-101, Attachment 1 at 4 ("This State of New York Initial Statement of Position In Support of Contention NYS-25 12 inspection plan will include the inspections specified in MRP-227, as modified by the conditions and limitations and applicant/licensee action items in the NRC SER on MRP-227, Revision 0. . . . Following issuance of MRP-227-A, Entergy will review the inspection plan to determine the need for revision, and will modify the inspection plan to include the necessary revisions, if any"). Entergy's evolving position fails to definitively commit to a precise program to address the problem of embrittlement of RPVIs and lacks analyses of some of the most important aspects of RPVI degradation, including the absence of any assessment of the s ynergistic affects of embrittlement, metal fatigue and stress corrosion cracking and the absence of specific criteria for acceptance, prevention and corrective actions. Rather, Entergy frequently asserts that its programs will be "consistent" with various more detailed programs, and avoids making a specific commitment to follow a specific requirement of any of those programs.

See e.g. NL-11-107, Attachment 1, Indian Point Energy Center Reactor Vessel Internals Inspection Plan at 21("Equipment, techniques, procedures and personnel used to perform examinations required under this program will be consistent with the requirements of MRP-228" (emphasis added)). Given the enormous importance of retaining the stru ctural integrity of RPVIs in the event of an accident, such as a DBA LOCA, this lack of definitive plans and definitive commitments to specific programs is unacceptable and should be cause for concern. As the Commission has held, when it comes to reactor safety requirements, "[w]e do not simply take the applicant at its word." Entergy Nuclear Vt. Yankee, LLC (Vt. Yankee Nuclear Power Station), CLI-10-17, 72 N.R.C. __, slip op. at 45 (July 8, 2010).

State of New York Initial Statement of Position In Support of Contention NYS-25

13 RELEVANT FACTS The RPVI AMP that is currently before the Board for review consists of the general plan described in NL-10-063 and the in spection plan portion of that AMP described in NL-11-107, as supplemented by NL-11-101. The latter reference relates to Entergy's commitment to include any modifications contained in MR P-227A that it believes are releva nt to Indian Point, in the inspection program for Indian Point. Thus, the adequacy of the AMP for RPVIs must stand or fall on the adequacy of these documents.

With regard to the issues that are relevant here the following fact s are established by Dr. Lahey's Report, (Exh. NYS 000296), his Prefil ed Direct Testim ony, (Exh. NYS000294) and numerous exhibits: 1. Neutron fluence is a substantial degradation phenomena, known as embrittlement, that will increasingly affect Indian Point if it operates beyond its original 40 year life (Lahey Report at ¶ 17) (Exh. NYS 000296); 2. Entergy has provided no analysis in support of its AMPs that addresses the synergistic affects of embrittlement, metal fatigue, stress corrosion, and unique RPVI material on the degradation of RPVIs duri ng the period of extended operation (Lahey Report at ¶ 16);

3. In NL-10-063 Entergy states that its RPVI AMP consists of its agreement to follow MRP-227 and, when available and if applicable, the modifi cations to MRP-227 identified in MRP-227A (Exh. NYS000313);

State of New York Initial Statement of Position In Support of Contention NYS-25 14 4. MRP-227 purports to be an internal s inspection and evaluation guide for PWR RPVIs that requires individual plant development to create a true plant-specific AMP (MRP-227 at v);

5. Entergy, in NL-11-107 provides its plant-specific AMP for RPVIs; 6. The plant specific AMP for Indian Point's RPVIs contains numerous instances in which no actual acceptance criteria are provided or the criteria are wholly subjective.

See, e.g., NL-11-107 (Exh. NYS000314), Physical Measurements Examination Acceptance Criteria (NL-11-107, Attachment 1 at 25("Specific acceptance criteria will be developed as required, and thus are not provided generically in this plan"); Control Rod Guide Tube Assembly, id. at 47 ("The specific rele vant condition is wear that could lead to loss of control rod alignment and impede control assembly insertion"); Baffle Former Assembly (Baffle Former Bolts), id. at 49 ("The examination acceptance criteria for the UT of the baffle-former bol ts shall be established as part of the examination technical justification"), 8 Thermal Shield Assembly, id. at 51 (The specific relevant conditions for thermal shield flexures are excessive wear, fracture, or complete separation"). MRP-227 is similarly vague when it comes to enforceable and objective standards for corrective actions and acceptance criteria. MRP-227 is replete with phrases such as "should," "could," "may," and "assumed." These terms render MRP-227, which

8 NYS-RK Contention 38 challenges the practice of postponing development of AMP plan components until after the hearing. Here , Contention 25 challenges the absence of any acceptance criteria and thus the inability of Entergy to demonstrate compliance with AMP requirements.

State of New York Initial Statement of Position In Support of Contention NYS-25

15 is itself only a guideline, unenforceable. See, e.g., MRP-227 at 6-9; 6-1 ("various options - are available for the disposition of conditions detected during examinations - that are unable to satisfy the examination acceptance criteria").

See also Lahey Report Exh. NYS000296 at ¶ 16.

7. Entergy's AMP does not specify what actions will take place when inspections reveal that accep tance criteria has not been achieved. For example, under Examination Acceptance Criteria for Visual (VT-3) Examination, Entergy merely states, "The disposition can include a supplementary examination to further characterize the relevant condition, an engineering evaluation to show that the component is capable of continued operation with a known relevant condition, or repair/replacement to remediate

the relevant conditi on." NL-11-107 at 23. 8. Entergy's AMP does not include any preventative action measures. NL-10-063, Attachment 1 at 86 ("The Reactor Vessel Internals Program is a condition monitoring program that does not include preventive actions"). 9. Although Entergy has agreed to do some baseline inspections for the baffle former assembly, the timing of thos e inspections is potentially well-beyond the date on which extended operation would commence. NL-11-107, Attachment 1 at 36-38 indicating that baseline inspections will occur in a range of 20-40 EFPY. 10. RPVIs are more susceptible to embrittlement than the RPV, for which Entergy has already had to seek an exemption from ASME End of Life Charpy impact State of New York Initial Statement of Position In Support of Contention NYS-25 16Upper Shelf Energy limits (Lahey Report at ¶¶ 19-20); 11. As the RPVIs age, embrittlement causes the nil ductility temperature (NDT) to rise thus expanding the temperature range at which RPVIs lose ductility (Lahey Report at ¶ 18);

12. There are factors other than embrittlement degrading the structural integrity of RPVIs including metal fatigue a nd stress corrosion cracking (Lahey Report at

¶¶ 17 and 19);

13. Embrittled components are more prone to crack propagation (Lahey Report at ¶ 19);
14. The control rod drives and their asso ciated guide tubes, plates, pins, and welds have been identified as particularly susceptible to stress corrosion cracking which cracking is made more severe by embrittlement (Lahey Report at ¶¶ 25-27); 15. Entergy's AMP for RPVI is inadequate with respect to the embrittlement of the control rod drives and their associated guide tubes, plates, pins, and welds (Lahey Report at ¶¶ 25-27, n.

5; MRP-227 at v); 16. In the event of a DBA LOCA, RPVI s can be subjected to substantial external forces from thermal shock transients and decompression shock loads (Lahey

Report at ¶ 23);

17. Entergy's AMP for RPVIs is not base d upon an analysis that considers the combined degradation affects of embrittlement, metal fatigue and stress corrosion cracking in determining the fr equency of inspections, the criteria for acceptance of State of New York Initial Statement of Position In Support of Contention NYS-25

17components after inspections, the criteria for when to expand inspections, the criteria for preventative actions and the criteria for wh en and how to repair or replace RPVI components (Lahey Report at ¶ 16);

18. Entergy intends to rely extensively on Visual (VT-3) Examination although there is substantial ev idence that this inspection methodology is often subjective and ineffectual (NL-10-063 at 87; MRP-227 at 4-4, 4-5 and 4-14 to 4-16); 19. Without timely and properly conducted baseline inspections of RPVIs, it is not possible to accurately measure the progression of degradation of components over the period of extended operation and to take timely action to prevent failure of the component (Lahey Report at ¶¶ 28-29).

ARGUMENT ENTERGY HAS NOT DEMONSTRATED THAT THE EFFECTS OF AGING ON CERTAIN INTENDED FUNCTIONS WILL BE ADEQUATELY MANAGED FOR THE PERIOD OF EXTENDED OPERATION.

Pursuant to 10 C.F.R. § 54.33(a), all license re newals require the holde r of the license to be subject to 10 C.F.R. § 50.54 which includes, inter alia, the obligation to comply with "all rules, regulations, and orders of the Commission".

Id. § 50.54(h). Among those rules and regulations is 10 C.F.R. § 50.46(b)(4), which requires that following a LOCA "[c]alculated changes in core geometry shall be such that the core remains amenable to cooling." The obligation to maintain a coolable core geomet ry following a DBA LOCA has been described as State of New York Initial Statement of Position In Support of Contention NYS-25 18one of the requirements that is "fundamental to providing reasonable assurance that [a proposed action] will not endanger the health and safety of the public."

Duke Energy Corporation (Catawba Nuclear Station, Units 1 an d 2) 60 N.R.C. 713, 724, LBP-04-32 (2004). The purpose of an AMP is to assure that the "effects of aging on [certain] intended function(s) will be adequately managed for the period of extended operation." 10 C.F.R. § 54.21(c)(iii). Maintaining a coolable core geometry is one of the functions that must be maintained throughout the period of extended oper ation. 10 C.F.R. §§ 54.33(a), 54.4(a)(i)(ii)(iii) and 54.4(b). An applicant must demonstrate that: there is reasonable assurance that the ac tivities authorized by the renewed license will continue to be conducted in accordance with the CLB . . . [including] (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under §54.21(a)(1)

10 C.F.R. § 54.29(a)(1). Here, Entergy must prove that its AMP for components whose continued integrity following a DBA LOCA is essential to maintaining a coolable core geometry will achieve the regulation's objective. Entergy has failed to meet that obligation.

First, maintaining the integrity of RPVIs following postulated accidents, such as a DBA LOCA, is essential to maintaining a coolable core geometry. Safety Evaluation Report Related to the License Renewal of Indian Point Nu clear Generating Unit No

s. 2 and 3, NUREG-1930 ("SER"), Vol. 1 at 2-41 (November 2009)("if cert ain reactor vessel intern als failed, they could potentially inhibit core coolability during an accident"); Lahey Report at ¶23 (Exh.

NYS000296)..

Second, integrity of the RPVI s is directly dependent upon how well the RPVI withstands State of New York Initial Statement of Position In Support of Contention NYS-25

19the hostile environment inside the reactor over an extended period of time including the synergistic effects of embrittlement, metal fatigue and stress corrosion cracking. However, in conducting its analysis of vari ous accident conditions, and in developing its AMP, Entergy does not provide evidence of the abilit y of the RPVIs to withstand acc ident-induced shock loads given all these degradation mechanisms to which they have been and will be subjected. Thus, Entergy has not demonstrated that its AMP for RPVIs will detect and correct problems with RPVIs that could cause them to fail in the event of accident-induced decompression and/or thermal shock loads.

Third, Entergy does not have an adequate AMP to manage and monitor the aging degradation of control rods and their associat ed guide tubes, plates, pins, and welds, even though these components are particularly vu lnerable to stress corrosion cracking and embrittlement makes such stress corrosion more severe. Without any explanation in either MRP-227 or in NL-11-107, these RPVIs are given limite d consideration in the inspection program being offered by Entergy. MRP-227 at v., NR C Regulatory Issue Summary 2011-07 License Renewal Submittal Information For Pressurized Water Reactor Internals Aging Management ("RIS-2011-07"), an internal NRC guidance document, asserts: The scope of the program does not include consumable items such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation because these components are not subject to an AMR, as defined in 10 CFR 54.21(a)(1).

Id. at 3. However, this statement is demonstrably inaccurate. First, the control rods and their associated guide tubes, plates, pins, and welds are not consumables, at least not in the normal State of New York Initial Statement of Position In Support of Contention NYS-25 20 sense of that word. In fact, Dr. Lahey has urged that Entergy be required to make them consumable -

i.e., to be replaced frequently to account for the embrittlement, stress corrosion and metal fatigue degradation to which they ar e being subjected. Sec ond, 10 C.F.R. §54.21(a)(1) excludes "control rod drives," not control rods and their associated guide tubes, plates, pins, and welds. The control rod drives are more akin to pumps because they have moving parts and thus are not passive. But, like pumps, the non-moving parts of the system remain subject to AMR and AMP, where appropriate. Thus, for example, while the Commission makes clear that internal parts of components that move are exempt from aging management, the housing of the component is not included if it is part of the pressure boundary of the containment: the Commission determined that structures and components that perform active functions are not subject to an aging management review (e.g., pumps (except casing), valves (except body), motors, diesel generators, air compressors, snubbers, the control rod drive, ventilation dampers, pressure transmitters, pressure indicators, water level indicators, switchgears, coolin g fans, transistors, batteries, breakers, relays, switches, pow er inverters, circuit boards, battery chargers, and power supplies). Howeve r, pressure-retain ing boundaries (e.g., pump casings, valve bodies, fluid system piping) and struct ural supports (e.g., diesel generator structural supports) th at are necessary for the structure or component to perform its intended function meet the description of passive, and will be subject to an aging management review.

Statement of Considerations Regarding Li cense Renewal Regulations, 60 Fed. Reg. 22461, 22464, 22,477 (May 8, 1995). If the control rods or their associated guide tubes, plates, pins, and welds fail due to embrittlement, stress corrosion cracking and metal fatigue, in the event of an accident, the control rod drives will not be able to perform their function of controlling core reactivity. See also Lahey Report at ¶¶ 25-27NYS 000296 . Fourth, even had Entergy provided a basic analysis to justify the limited AMP inspection State of New York Initial Statement of Position In Support of Contention NYS-25

21program it has offered for RPVIs, the program is deficient in several respects. Entergy's limited RPVI inspection program does not fully provide criteria for determining when the status of the RPVIs is acceptable or when the inspection s hould be expanded, nor does does it provide any criteria for determining what actions shoul d be taken when a RPVI is found to be "unacceptable." Similarly, it does not provide any criteria for a program that defines repair and replacement actions, lacks any preventative action measures for RPVI embrittlement and, while in certain instances provides for baseline inspec tions, the schedule for those baseline inspections allows postponement of the inspections until well into the period of extended operation..

Fifth, the inspection program offered relies substantially on Visual (VT-3) Examination (NL-10-063 at 87; EPRI MRP-227 at 4-4 to 4-5, 4-14 to 4-16), although there is serious doubt that this method will accurately or reliably detect RPVI degradation in all critical situations. The problem is that the MRP-227-approved VT-3 me thodology is incapable of detecting substantial cracks in RPVI that do not exhibit as gross deformation of the component, and that use of other methodologies such as UT, EV-1 and VT-1, also recommended by MRP-227, are more reliable and better able to detect stress corrosion cr acking. Reasons for N on-concurrence on "Draft Safety Evaluation for the Electric Power Research Institute's Topical Report (TR) Materials Reliability Program (MRP) Report 1016596 (MRP-227), Revision 0, 'Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines.'" R.L. Tregoning RES/DE, ML110770169; Comments for the Document Spons or to Consider Pertaining to Non-Concurrence on "Draft Safety Evaluation for the Electric Power Research Institute's Topical State of New York Initial Statement of Position In Support of Contention NYS-25 22 Report (TR) Materials Reliability Program (MRP) Report 1016596 (MRP-227), Revision 0, 'Pressurized Water Reactor (PWR) Internals Insp ection and Evaluation Guidelines.'" Michael J. Case RES/DE, ML110810787. Reliance on the assumpti on that there will be a grossly visible defect before the component reaches critical degr adation is a hypothesis that has not been tested and as to which Entergy offers no supporting evidence. Notably, MRP-227 also contains text that raises question about how Entergy will inspect for and manage RPVI embrittlement. While MRP-227 and NL-10-063 seem to imply that a licensee will examine RPVIs for signs of embrittlement, footnote 1 to MRP-227 Table 3-3 states as follows: The significance of thermal and irradiation embrittlement is directly related to the probability of a flaw existing in the component. There are no recommendations for inspection to determine embrittlement level because these mechanisms cannot be directly observed. However, potential embrittlement must be considered in fl aw tolerance evaluations.

MRP-227, p. 3 3-24, Table 3-3, "Final Disposition of Category B and C Westinghouse internals," note 1 (emphasis added). The recognition that embrittlement mechanisms "cannot be directly observed" further supports the concerns about the effectiveness of visual exams. Even MRP-227 recognizes that EV-1 and VT

-1 have detection capabilities that are superior to VT-3. MRP-227 at 4-4 ("Unlike the detection of general degradation conditions by visual (VT-3) examination, visu al (VT-1) and enhanced visual (EVT-1) examinations are conducted to detect discontinuities and imperfections on the su rface of components, including such conditions as cracks, wear, corrosion, or erosion. Specifically, VT-1 is used for the detection of surface discontinuities such as gaps, while EVT-1 is used for the detection of surface State of New York Initial Statement of Position In Support of Contention NYS-25

23 breaking flaws"). However, ne ither MRP-227 nor NL-11-107 given any rational justification for the widespread approval of the use of VT-3 for such critical components as baffle-former bolt assemblies.

In summary, Entergy recogni zes that it is obligated to provide an AMP for RPVIs. However, the AMP offered is deficient in several fundamental ways which make it impossible for Entergy to demonstrate the adequacy of the RPVI AMP to assure the coolability of the core geometry following accidents, such as a DBA LOCA. Thus, license renewal must be denied because the Board does not have a basis upon which it can find that there is reasonable assurance that the core will retain a coolable geometry in the event of a DBA LOCA. 10 C.F.R. § 54.29(a)(1). CONCLUSION For the above reasons Entergy's applicati on to renew the operating licenses for Indian Point Unit 2 and Indian Point Unit 3 should be denied.

Respectfully submitted, Signed (electronically) by John J. Sipos Assistant Attorney General Office of the Attorney General for the State of New York The Capitol Albany, New York 12227 (518) 402-2251 Dated: December 22, 2011

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