ML13081A768

Riverkeeper Post-Hearing Proposed Findings of Fact and Conclusions of Law Regarding Contention RK-TC-2-Flow Accelerated Corrosion

ML13081A768
Person / Time
Site:	Indian Point
Issue date:	03/22/2013
From:	Brancato D, Musegaas P Riverkeeper
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 24281, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01
Download: ML13081A768 (90)

Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD

)

In the Matter of )

)

Entergy Nuclear Operations, Inc. ) Docket Nos.

(Indian Point Nuclear Generating ) 50-247-LR Units 2 and 3) ) and 50-286-LR

___________________________________________ )

RIVERKEEPER POST-HEARING PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW REGARDING CONTENTION RK-TC FLOW ACCELERATED CORROSION March 22, 2013

TABLE OF CONTENTS BACKGROUND .......................................................................................................................1 APPLICABLE LEGAL AND REGULATORY REQUIREMENTS ......................................7 BURDEN OF PROOF............................................................................................................. 11 FINDINGS OF FACT ............................................................................................................. 12 I. WITNESSES ............................................................................................................. 12 A. Riverkeepers Witness .............................................................................................. 12 B. Entergys Witnesses .................................................................................................. 13 C. NRC Staffs Witnesses .............................................................................................. 16 II. THE NATURE AND SIGNIFICANCE OF FLOW ACCELERATED CORROSION ........................................................................................................... 16 A. Definition of FAC...................................................................................................... 16 B. The Local, Non-Linear Nature of FAC .................................................................... 21 C. The Safety Significance of FAC................................................................................ 26 III.

SUMMARY

OF ENTERGYS PROGRAM FOR MANAGING FLOW ACCELERATED CORROSION ............................................................................. 28 IV. THE CHECWORKS COMPUTER MODEL ......................................................... 31 A. The Nature of the CHECWORKS Computer Code ................................................ 31 B. Problematic Assumptions Inherent in the CHECWORKS Model ......................... 33 C. The Role of CHECWORKS at Indian Point ........................................................... 35 D. The Predictive Performance of CHECWORKS at Indian Point ............................ 37 E. The Implication of Poor Predictive Accuracy of CHECWORKS at Indian Point ......................................................................................................................... 43 F. The Track Record of CHECWORKS Performance at Indian Point ...................... 45 G. CHECWORKS and Compliance with the GALL Report ........................................ 48 H. Alternative Computer Modeling Suitable for Managing FAC ............................... 54 V. ENTERGYS OTHER TOOLS FOR DETERMINING THE SCOPE OF FAC INSPECTIONS ................................................................................................ 56 A. Reinspections Based on Actual Wall Thickness Measurements ............................. 56 B. Industry and Plant Operating Experience ............................................................... 59 C. Other Plant Inspection Programs ............................................................................ 59 D. Engineering Judgment.............................................................................................. 60 i

E. Selection of SNM Components ................................................................................. 64 F. Conclusions Regarding Entergys Other Tools for Inspection Selection Scope ........................................................................................................................ 65 VI. THE SAFETY CONSEQUENCES OF IMPROPERLY MANAGED FAC AT INDIAN POINT ........................................................................................................ 66 VII. THE SUFFICIENCY OF ENTERGYS FAC AGING MANAGEMENT PROGRAM ............................................................................................................... 69 A. FAC Program Documentation ................................................................................. 69 B. Insufficiency of Detail of FAC AMP ........................................................................ 73 VIII. FACTUAL CONCLUSIONS TO BE DRAWN FROM THE EVIDENCE ............ 76 CONCLUSIONS OF LAW ..................................................................................................... 80 PROPOSED ORDER .............................................................................................................. 83 ii

In accordance with 10 C.F.R. §§ 2.712, 2.1209, the Atomic Safety and Licensing Boards (ASLB) July 1, 2010 Scheduling Order, 1 and the ASLBs February 28, 2013 Order Granting Parties Joint Motion for Alteration of Filing Schedule, 2 Riverkeeper, Inc. (Riverkeeper),

hereby submits the instant Post-Hearing Proposed Findings of Fact and Conclusions of Law Regarding Contention RK-TC Flow Accelerated Corrosion. The testimony and supporting exhibits in the record in the above-referenced proceeding relating to Contention RK-TC-2 demonstrate that Entergy does not have an adequate program to manage the aging effects of flow accelerated corrosion during the proposed period of extended operation for the Indian Point nuclear power plant.

BACKGROUND On or about April 23, 2007, Entergy Nuclear Operations, Inc. (Entergy) filed a License Renewal Application (LRA) with the U.S. Nuclear Regulatory Commission (NRC) seeking 20-year extended operating licenses for Indian Point nuclear generating Units 2 and 3.3 The LRA purported to include a sufficient and legally acceptable program for managing an aging phenomenon known as flow accelerated corrosion (FAC) throughout the proposed period of extended operation (PEO). Pursuant to 10 C.F.R. § 2.309 and Federal Register notices published by the NRC, on November 30, 2007, Riverkeeper filed a request for hearing and petition to intervene in the Indian Point license renewal proceedings, proffering, inter alia, a contention, Riverkeeper Contention TC-2 (hereinafter referred and cited to as Contention RK-1 In the Matter of Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos.

50-0247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, Scheduling Order (July 1, 2010), at ¶ N.

2 In the Matter of Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos.

50-0247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, Order (Granting Parties Joint Motion for Alteration of Filing Schedule) (Feb. 28, 2013).

3 See Exhibit ENT00015A-B (Indian Point Energy Center LRA (April 2007)).

1

TC-2), challenging the adequacy of Entergys LRA for failure to demonstrate an effective program for managing FAC at the facility. 4 In particular, Contention RK-TC-2 asserted that Entergys program for management of FAC is deficient because it has not demonstrated that components in the Indian Point nuclear power plant that are within the scope of the license renewal rule and are vulnerable to FAC will be adequately inspected and maintained during the license renewal term. 5 Contention RK-TC-2 explained that Entergys program for managing FAC is inadequate due to reliance on the computer code CHECWORKS without sufficient benchmarking or a track record of performance at Indian Point, and because Entergys FAC program does not provide sufficient details to demonstrate that susceptible plant components will be adequately maintained during the PEO. 6 On July 31, 2008, the ASLB admitted Contention RK-TC-2 for an adjudicatory hearing, finding that the contention raises questions regarding the sufficiency of Entergys AMP [aging management program] to demonstrate that a specific class of components subject to FAC will be managed so that their intended functions will be maintained during the period of extended operations.7 Thereafter, in accordance with discovery obligations set forth in 10 C.F.R. § 2.336, the parties began exchanging monthly logs disclosing documents relevant to the contention.

In November 2009, NRC Staff published a final Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3, which contained the results of the NRC Staffs review of Entergys Flow Accelerated Corrosion Program at Indian 4

Riverkeeper, Inc.s Request for Hearing and Petition to Intervene in the License Renewal Proceedings for the Indian Point Nuclear Power Plant (November 30, 2007), ADAMS Accession No. ML073410093, at 15-23 (hereinafter Riverkeeper Petition to Intervene).

5 Riverkeeper Petition to Intervene at 16.

6 Riverkeeper Petition to Intervene at 16, 20-23.

7 See In the Matter of Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos. 50-247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, Memorandum and Order (Ruling on Petitions to Intervene and Requests for Hearing) (July 31, 2008), ADAMS Accession No. ML082130436, at 167-169 (hereinafter ASLB July 31, 2008 Contention Admissibility Order).

2

Point.8 On July 26, 2010, Entergy filed a motion seeking summary disposition of Contention RK-TC-2.9 The ASLB denied this motion because genuine issues of material fact regarding the adequacy of the Applicants plan to manage the effects of flow-accelerated corrosion (FAC) during the proposed period of extended operation must be resolved on the merits after an evidentiary hearing.10 Pursuant to 10 C.F.R. § 2.1207(a)(1) and various ASLB scheduling and procedural orders, in advance of the adjudicatory hearing, on December 22, 2011, Riverkeeper filed an Initial Statement of Position Regarding Contention RK-TC-2 (Exhibit RIV000002), prefiled written direct testimony of Riverkeepers expert, Dr. Joram Hopenfeld (Exhibit RIV000003), as well as numerous exhibits (Exhibits RIV000004 through RIV000033), in support of RK-TC-2.11 Following Riverkeepers extensive initial filings on RK-TC-2, on January 30, 2012, Entergy filed a motion in limine, seeking to exclude certain portions of Riverkeepers testimony, exhibits and initial statement of position. 12 On March 6, 2012, the ASLB denied this motion.13 8

Exhibit NYS00326A-NYS00326F (NUREG-19030, Safety Evaluation Report, at pp.3-21 to 3-285).

9 Applicants Motion for Summary Disposition of Riverkeeper Technical Contention 2 (Flow-Accelerated Corrosion) (July 26, 2010), ADAMS Accession No. ML102140430, (hereinafter Applicants Motion for Summary Disposition of RK-TC-2).

10 In the Matter of Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos.

50-0247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, Memorandum and Order (Ruling on Entergys Motion for Summary Disposition of Riverkeeper TC-2 (Flow-Accelerated Corrosion)) (November 4, 2010), at 1.

11 See Exhibit RIV000002 (Riverkeeper Initial Statement of Position Regarding Contention RK-TC-2 (Flow Accelerated Corrosion) (December 22, 2011)); Exhibit RIV000003 (Prefiled Written Testimony of Dr. Joram Hopenfeld Regarding Riverkeeper Contention TC Flow Accelerated Corrosion (December 21, 2011)); Exhibits RIV000004 through RIV000033.

12 Entergys Motion in Limine to Exclude Portions of Pre-Filed Direct Testimony, Expert Report, Exhibits, and Statements of Position for Contention Riverkeeper TC-2 (Flow-Accelerated Corrosion) (January 30, 2012) (seeking to exclude discussions of Entergys failure to consider the impact of FAC at Indian Point on loss-of-coolant accidents, probabilistic risk assessments, component integrity under seismic loads, component integrity during station blackout loads, and component susceptibility to metal fatigue).

13 In the Matter of Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos.

50-0247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, Order (Granting in Part and Denying in Part Applicants Motions in Limine) at 23 (March 6, 2012) (ASLB Order Denying Entergy Motion in Limine on RK-TC-2) (finding that the issues objected to by Entergy were related and relevant to whether FAC will be adequately managed during the period of extended operationsa question that the Board will weigh on the merits during the evidentiary hearing in this proceeding.).

3

Subsequently, on March 28, 2012 and March 31, 2012, Entergy and NRC Staff submitted statements of position, pre-filed written testimony, and exhibits related to RK-TC-2, respectively. 14 On April 30, 2012, Riverkeeper filed a motion in limine to exclude certain testimony proffered by Entergy on RK-TC-2.15 On June 1, 2012, the ASLB ruled to hold Riverkeepers Motion in Limine in abeyance, pending the development of the factual record on Contention RK-TC-2 at the adjudicatory hearing. 16 Thereafter, on June 29, 2012, Riverkeeper filed a Revised Statement of Position (Exhibit RIV000107), prefiled rebuttal testimony (Exhibit RIV000108), and several additional supporting exhibits related to Contention RK-TC-2 (Exhibits RIV000109 through RIV000113).17 In accordance with 10 C.F.R. § 2.1207(a)(3) and several ASLB orders, on August 29, 2012, Riverkeeper (and ostensibly Entergy and NRC Staff),

14 See Exhibit ENT000028 (Entergy Statement of Position Regarding Contention RK-TC-2 (Flow-Accelerated Corrosion), March 28, 2012); Exhibit ENT000029 (Testimony of Entergy Witnesses Ian D. Mew, Alan B. Cox, Nelson F. Azevedo, Jeffrey S. Horowitz, and Robert M. Aleksick Regarding Contention RK-TC-2 (Flow-Accelerated Corrosion), March 28, 2012); Exhibits ENT00015A-B, ENT000030 to ENT000089; Exhibit NRC000120 (NRC Staffs Statement of Position Regarding RK-TC-2, March 31, 2012); Exhibit NRC000121 (NRC Staff Testimony of Matthew G. Yoder and Allen L. Hiser, Jr. Concerning Riverkeeper Technical Contention RK-TC-2 Flow Accelerated Corrosion, March 31, 2012); Exhibits NRC000122 to NRC000131.

15 Riverkeeper, Inc. Motion in Limine to Exclude Portions of Pre-Filed Testimony and Statement of Position Regarding RK-TC-2 (Flow Accelerated Corrosion) (April 30, 2012) (Riverkeeper Motion in Limine) ((objecting to Entergys witnesses reference to historical data purportedly used to benchmark the CHECWORKS computer code, which is used by Entergy to manage FAC at Indian Point, in light of a discovery ruling from earlier in the proceeding in which the ASLB found that Entergy does not have ready access to the data [] and thus has not, and cannot, rely on it to provide the track record for its AMP [aging management program] or to demonstrate that its use of CHECWORKS is adequately benchmarked. (Quoting In the Matter of Entergy Nuclear Operations, Inc.

(Indian Point Nuclear Generating Units 2 and 3), Docket Nos. 50-0247-LR and 50-286-LR, ASLBP No.07-858 LR-BD01, Order (Ruling on Riverkeepers Motion to Compel) (November 4, 2010), at 5 (emphasis added))).

16 In the Matter of Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos.

50-0247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, Order (Denying New Yorks Motion in Limine and Holding Riverkeepers Motion in Limine in Abeyance) (June 1, 2012) at 12 ((Stating that [b]ecause the relationship of the past use of CHECWORKS to Entergys LRAs AMP for FAC is directly relevant to RK-TC-2, we will probe at the oral stage of the evidentiary hearing how this information was relied on by Entergy in preparing its LRA and that after the factual record is more fully developed, we will be better supplied with information to understand the history of Entergys CHECWORKS benchmarking and to resolve Riverkeepers Motion in Limine.).

17 See Exhibit RIV000107 (Riverkeeper Revised Statement of Position Regarding Contention RK-TC-2 (Flow Accelerated Corrosion) (June 29, 2012)); Exhibit RIV000108 (Prefiled Rebuttal Testimony of Dr. Joram Hopenfeld Regarding Riverkeeper Contention TC Flow Accelerated Corrosion (June 29, 2012)); Exhibits RIV000109 through RIV000113.

4

submitted proposed board examination questions related to Contention RK-TC-2 for the ASLB to consider propounding to witnesses during adjudicatory hearings on the matter. 18 In advance of the adjudicatory hearings, Riverkeeper made certain updates and/or corrections to previously proffered exhibits related to Contention RK-TC-2, and, in addition, filed several supplemental relevant exhibits, as follows. In September 2012, the parties resolved to remove a proprietary designation that Riverkeeper had previously assigned to a Report generated by Riverkeepers expert witness, Dr. Joram Hopenfeld, Exhibit RIV000005; Riverkeeper filed an updated, non-confidential version of the relevant exhibit on September 7, 2012.19 On October 1, 2012, in accordance with the ASLBs directive, Riverkeeper filed a corrected version of Exhibit RIV000008, which previously had an incorrect exhibit designation. 20 Lastly, on October 11, 2012, Riverkeeper filed a motion to file five additional exhibits relevant to Contention RK-TC-2, Exhibits RIV000127 through RIV000131, which the ASLB granted via written order on October 15, 2012.21 Adjudicatory hearings convened in the above-captioned Indian Point license renewal proceeding on October 15, 2012. At the outset, the ASLB admitted the parties theretofore most recent exhibit lists and all the exhibits listed thereon (with certain exceptions, none of which 18 See Riverkeeper Proposed Board Examination Questions (Aug. 29, 2012).

19 See Letter from D. Brancato (Riverkeeper) to ASLB, Re: Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos. 50-247-LR, 50-286-LR (Sept. 7, 2012); RIVR60001 (Riverkeeper Updated Exhibit List). As the only theretofore confidential/proprietary designation, the hearing with respect to Contention RK-TC-2 was a fully open proceeding, for which no portion was closed to the public.

20 See Letter from D. Brancato (Riverkeeper) to ASLB, Re: Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos. 50-247-LR, 50-286-LR (Oct. 1, 2012); RIVR70001 (Riverkeeper Updated Exhibit List).

21 See Riverkeeper, Inc. Motion for Leave to File Additional Exhibits Concerning Contention RK-TC-2 (Flow Accelerated Corrosion) (Oct. 11, 2012); In the Matter of Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos. 50-0247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, Order (Granting Riverkeepers Motion for Leave to File Additional Exhibits) (Oct. 15, 2012); see also Exhibits RIV000127 through RIV000131.

5

affected exhibits related to Contention RK-TC-2).22 Hearings on Contention RK-TC-2 began in the afternoon of October 15, 2012, continued on October 16, 2012, and concluded on the morning of October 17, 2012.23 The hearing consisted of initial presentations by Entergys witness, Dr. Jeffrey S. Horowitz, and Riverkeepers witness, Dr. Joram Hopenfeld, 24 the ASLBs live examination of a panel of witnesses identified by the respective parties as experts, as well as limited re-direct- and cross-examinations of the panel of witnesses by counsel for Riverkeeper, Entergy, and NRC Staff. The panel of witnesses included: Dr. Joram Hopenfeld for Riverkeeper; Ian D. Mew, Alan B. Cox, Nelson F. Azevedo, Dr. Jeffrey S. Horowitz, and Robert M. Aleksick for Entergy; and Dr. Allen Hiser and Matthew Yoder for NRC Staff.

PowerPoint presentations generated by Dr. Horowitz and Dr. Hopenfeld to facilitate their respective initial presentations were marked as board exhibits 1 and 2 for identification, but were not received into evidence as hearing exhibits in the proceeding. 25 During the course of the hearing, Riverkeeper introduced two new exhibits, RIV000132 and RIV000133, that became helpful for clarifying certain of Dr. Hopenfelds concerns relating to Entergys program for managing FAC at Indian Point. The ASLB received these exhibits de bene into evidence at the 22 In the Matter of: Entergy Nuclear Operations, Inc. (Indian Point Generating Units 2 and 3), Docket Nos. 50-0247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, Transcript of Hearing (hereinafter cited as Tr. at [page:line number]), at 1268:21-24, 1270:1-6. Citations to admitted hearing exhibits below in Riverkeepers Findings of Fact and Conclusions of Law will be as follows: Exh. [xxxxxxxxx] ([document description] at __).

23 Tr. at pp.1293-1892.

24 Prior to the commencement of the hearing, the ASLB requested a presentation by Entergys witness, Dr. Jeffrey S.

Horowitz, on the CHECWORKS program. See In the Matter of Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos. 50-0247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, Order (Evidentiary Hearing Administrative Matters) (Sept. 14, 2012), at 2. Riverkeepers expert was afforded the opportunity to provide a responsive presentation. See In the Matter of Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos. 50-0247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, Teleconference Transcript (September 24, 2012), at pp.1226-1229; see also E-mail from S. Lewman (ASLB Law Clerk) to the parties, Re: CHECWORKS Presentations (Oct. 10, 2012).

25 See Tr. at 1315: 13-19 (the PowerPoint presentations themselves are not evidence in this proceeding. They are accepted, one from Dr. Horowitz as Board Exhibit 1; the one from Dr. Hopenfeld as Board Exhibit 2, both for identification. (Whereupon, the documents referred to were marked for identification as Board Exhibits 1 and 2.).

6

hearing.26 In addition, during the hearing, Riverkeepers expert made reference to an admitted hearing exhibit that had been designated as relevant to a different contention, Exhibit RIV000049. On October 17, 2012, per the ASLBs instruction, 27 Riverkeeper filed an updated exhibit list, RIVR10001 (i.e., revision 10) reflecting the addition of Exhibits RIV000132 and RIV000133, and which also designated RIV000049 as additionally relevant to Contention RK-TC-2.28 Several days after hearings in relation to Contention RK-TC-2 concluded, the ASLB orally denied Riverkeepers aforementioned Motion in Limine.29 Overall, the hearings on Contention RK-TC-2 resulted in a stenographic record of approximately 550 pages of transcript 30 and approximately 124 exhibits, including pre-filed written testimony from all the parties, received into evidence. 31 APPLICABLE LEGAL AND REGULATORY REQUIREMENTS NRCs regulations require nuclear power plant license renewal applicants to have programs for effectively managing the aging of in-scope plant systems, structures, and components. 10 C.F.R. §§ 54.21, 54.29. In particular, 10 C.F.R. § 54.21(a)(3) states that for each system, structure and component that is within the scope of NRC license renewal requirements, applicants must demonstrate that the effects of aging will be adequately managed so that the intended function(s) will be maintained consistent with the CLB [current licensing basis] for the period of extended operation. 10 C.F.R. §§ 54.21, 54.29.

26 Tr. at 1791-92; In the Matter of Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3),

Docket Nos. 50-0247-LR and 50-286-LR, ASLBP No. 07-858-03-LR-BD01, Order (Adopting Proposed Transcript Corrections with Minor Edits) (Dec. 27, 2012), at 17 (changing incorrect reference to RIV000049, to RIV000132 and RIV000133).

27 See Tr. at 1792:11-14.

28 See Letter from D. Brancato (Riverkeeper) to ASLB, Re: Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3), Docket Nos. 50-247-LR, 50-286-LR (Oct. 17, 2012); RIVR10001 (Riverkeeper Updated Exhibit List).

29 See Tr. at 2923:3-14.

30 Tr. at 1293-1450, 1477-1778, 1800-1892.

31 See Exhibits RIVR100001; ENTR70001; NRCR30001.

7

According to applicable guidance contained in NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, (hereinafter SRP-LR), an aging management program (AMP) sufficient to meet the regulatory standard should include and implement ten separate elements: (1) the scope of the program; (2) actions for prevention and mitigation of aging degradation; (3) parameters for monitoring and inspecting so as to detect the presence and extent of aging effects; (4) detection of aging effects prior to loss of the structure and component intended function; (5) trending activities to provide predictability of the extent of degradation, in order to effect timely corrective and mitigative actions; (6) acceptance criteria against which the need for corrective actions will be evaluated; (7) timely corrective actions when acceptance criteria are not met; (8) confirmation processes to ensure adequate preventative actions, and complete and effective corrective actions; (9) administrative controls to provide for formal review and approval mechanisms; and (10) consideration of plant-specific and industry operating experience. 32 NRCs NUREG-1801, Generic Aging Lessons Learned (GALL) Report (hereinafter GALL Report), a technical basis document referenced in NUREG-1800, provides license renewal applicants with guidance regarding how an AMP can satisfy the 10 program elements identified in SRP-LR.33 An AMP that is consistent with the GALL Report is acceptable to show compliance with NRCs regulatory standard in 10 C.F.R. § 54.21(a)(3). However, applicants cannot generically claim consistency with this guidance document, and instead must provide a 32 Exhibit NYS000195 (NUREG-1800, Rev. 1, at A.1-3 to A.1-7); Exhibit NYS000161 (NUREG-1800, Rev. 2, at A.1-3 to A.1-7).

33 Exhibits NYS00146A-NYS00146C (GALL Report, Revision 1); Exhibits NYS00147A- NYS00147D (GALL Report, Revision 2).

8

reasonably thorough description of its AMP to show conclusively how th[e] program will ensure that the effects of aging will be managed. 34 In contrast, an applicant merely stating that its AMP meets NUREG-1801 without any specificity falls short of the required demonstration. . . . [W]hether an applicant is successful depends upon whether it is [sic] has shown that the specific plant details of its AMP have adequately addressed this guidance. But a bald reference to NUREG-1801 fails to show how the recommendations of NUREG-1801 are proposed to be implemented for [the facility] . . . and does not demonstrate that the effects of aging are adequately managed for the plant.35 Section XI.M17, Flow-Accelerated Corrosion of the GALL Report, Revision 1, as well as the more recent GALL Report, Revision 2, contain guidance relative to an acceptable FAC AMP.36 The GALL Report indicates that an applicants FAC program can be based on Electric Power Research Institute (EPRI) guidelines in the Nuclear Safety Analysis Center (NSAC)-

202L-R2 (or R3), and, in summary, includes performing (a) an analysis to determine critical locations, (b) limited baseline inspections to determine the extent of thinning at these locations, and (c) follow-up inspections to confirm predictions, or repairing or replacing components as necessary.37 The GALL Report indicates that an acceptable FAC program includes the use of a predictive code, such as CHECWORKS. 38 Such a code is used to predict component degradation in the systems conducive to FAC, as indicated by plant specific data, including 34 Entergy Nuclear Vermont Yankee (Vermont Yankee Nuclear Power Station), LBP-08-25, 68 NRC 763, 870 (Nov.

24, 2008).

35 Id. at 871.

36 Exhibit NYS00146C (GALL Report, Revision 1, at pp. XI M-61 to XI M-63); Exhibit NYS00147D (GALL Report, Revision 2, at pp. XI M17-1 to XI M17-4).

37 See Exhibit NYS00146C (GALL Report, Revision 1, at XI M-61); Exhibit NYS00147D (GALL Report, Revision 2, at XI M17-1).

38 See Exhibit NYS00146C (GALL Report, Revision 1, at XI M-61); Exhibit NYS00147D (GALL Report, Revision 2, at XI M17-1).

9

material, hydrodynamic, and operating conditions.39 Inspections based on the results of such a predictive computer model should provide reasonable assurance that structural integrity will be maintained between inspections and ensure that the extent of wall thinning is adequately determined, that intended function will not be lost, and that corrective actions are adequately identified.40 In relation to the acceptance criteria program element, the GALL Report explains that a predictive code such as CHECWORKS is used to calculate the number of refueling or operating cycles remaining before the component reaches the minimum allowable wall thickness, in order to determine the need for some form of corrective action.41 The most recent revision of the GALL Report, Revision 2, (issued in December 2010 and constituting the most current final and effective iteration of the NRCs position with regard to an acceptable program for managing the aging effects of FAC, as of the date of the adjudicatory hearings in the Indian Point license renewal proceeding as well as this filing), states that CHECWORKS is acceptable because it provides a bounding analysis for FAC. The analysis is bounding because in general the predicted wear rates and component thicknesses are conservative when compared to actual field measurements. It is recognized that CHECWORKS is not always conservative in predicting component thickness; therefore, when measurements show the predictions to be non-conservative, the model must be re-calibrated using the latest field data.42 The GALL Report otherwise provides further explanation regarding how a license renewal applicant can satisfy the various program elements required for a legally sufficient license renewal AMP. While the GALL Report, Revision 2 is the current legally operative version of the guidance document, in July 2012, NRC Staff issued a draft Interim Staff Guidance (ISG) 39 See Exhibit NYS00146C (GALL Report, Revision 1, at XI M-61); Exhibit NYS00147D (GALL Report, Revision 2, at XI M17-1).

40 See Exhibit NYS00146C (GALL Report, Revision 1, at XI M-62); Exhibit NYS00147D (GALL Report, Revision 2, at XI M17-2).

41 See Exhibit NYS00146C (GALL Report, Revision 1, at XI M-62); Exhibit NYS00147D (GALL Report, Revision 2, at XI M17-2).

42 Exhibit NYS00147D (GALL Report, Revision 2, at XI M17-1 to XI M17-2 (emphasis added)).

10

document that proposes certain changes to the AMP for FAC contained in the GALL Report, Revision 2.43 Though Entergy introduced this ISG as a hearing exhibit, at the time of NRC Staffs review of Entergys LRA and at the time of the adjudicatory hearing on Contention RK-TC-2, NRC Staffs draft ISG was still not final or effective.44 As reflected in NRCs guidance documents discussed above, the American Society of Mechanical Engineers (ASME) code requires that licensees maintain minimum design wall thicknesses of nuclear power plant piping during the entire period of plant operations. 45 Additionally, NRCs General Design Criterion 4 requires that plant structures, systems and components be able to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss of coolant accidents and be appropriately protected against dynamic effects . . .

that may result from equipment failures and from events and conditions outside the nuclear power unit.46 BURDEN OF PROOF The ultimate burden of proof on the question of whether the permit or the license should be issued is . . . upon the applicant.47 While an intervenor must go forward with evidence to establish a prima facie case on an admitted contention, the agencys rules of practice . . . place the ultimate burden of proof on . . . the license applicant, with respect to a merits disposition of 43 Exh. ENT000573 (Draft LR-ISG-2012-01).

44 Tr. at 1678:3-5, 17-19 (Hiser); Tr. at 1829:21-24 (ALJ Wardwell); Tr. at 1684:6-10 (Yoder).

45 ASME B31.3; ASME Code Section III, Paragraph NB-3200.

46 10 C.F.R. Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 4Environmental and dynamic effects design bases.

47 See 10 C.F.R. § 2.325; Amergen Energy Co. (Oyster Creek Nuclear Generating Station), CLI-09-7, 69 NRC 235, 269 (2009); Metropolitan Edison Co. (Three Mile Island Nuclear Station, Unit 1), ALAB-697, 16 NRC 1265, 1271 (1982).

11

any substantive matter at issue in th[e] proceeding (i.e., the admitted [] contentions).48 The license renewal applicant must assure that public health, safety, and environmental concerns are protected.49 FINDINGS OF FACT The record compiled on Contention RK-TC-2, through prefiled testimony, exhibits, and witness examination during the adjudicatory hearings, provides a sufficient basis for the following findings of fact.

I. WITNESSES A. Riverkeepers Witness

1. Riverkeepers testimony on Contention RK-TC-2 was presented by Dr. Joram Hopenfeld. Dr. Hopenfeld is an expert in the field of nuclear power plant aging management, with 45 years of professional experience. He holds B.S. and M.S. degrees in engineering, and a Ph.D. in mechanical engineering from the University of California Los Angeles. Exh.

RIV000003 (Hopenfeld Testimony at 1:6-14); Exh. RIV000004 (Hopenfeld CV). Dr. Hopenfeld is recognized expert in mechanical engineering. Tr. at 1691:4-6 (ALJ McDade).

2. Dr. Hopenfeld is an expert with respect to numerous subject matters, including nuclear safety regulation and licensing, design basis and severe accidents, thermal-hydraulics, material/environment interaction, corrosion, fatigue, radioactivity transport, industrial 48 In the Matter of Pacific Gas and Electric Co. (Diablo Canyon Power Plant Independent Spent Fuel Storage Installation) Docket No. 72-26-ISFSI; ASLBP No. 02-801-01-ISFSI, 58 NRC 47, *12; 2003 (2003); see also In the Matter of Carolina Power & Light Company (Shearon Harris Nuclear Power Plant), Docket No. 50-400-LA; ASLBP No. 99-762-02-LA; LBP-00-12, 51 NRC 247 (2000) (the agencys rules of practice . . . place the ultimate burden of proof on CP&L, as the license applicant, with respect to a merits disposition of any substantive matter at issue in this proceeding (i.e., the admitted BCOC contentions).

49 Amergen Energy Co. (Oyster Creek Nuclear Generating Station), CLI-09-7, 69 NRC 235, 263 (2009) (applicant must demonstrate that it satisfies the reasonable assurance standard by a preponderance of the evidence);

Metropolitan Edison Co. (Three Mile Island Nuclear Station, Unit 1), ALAB-697, 16 NRC 1265, 1271 (1982);

Commonwealth Edison Co. (Zion Units 1 and 2), ALAB-616, 12 NRC 419, 421 (1980) (Applicants have to provide reasonable assurance that public health, safety, and environmental concerns were protected, and to demonstrate that assurance by a preponderance of the evidence.).

12

instrumentation, environmental monitoring, pressurized water reactor steam generator transient testing and accident analysis, design, and project management. Exh. RIV000003 (Hopenfeld Testimony at 1:6-14); Exh. RIV000004.

3. Dr. Hopenfeld educational and professional experience makes him well-qualified to provide opinions about the aging phenomenon of FAC. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 4:11-6:14, 7:18-20, 8:14-16). Dr. Hopenfeld has ample educational and professional experience in the fields of electrochemistry, instrumentation, materials, and mass transfer, knowledge of which is required to understand and evaluate corrosion and related aging mechanisms. Id. Dr. Hopenfeld has had vast amounts of direct, hands-on experience with FAC and related issues. Id. Dr. Hopenfeld has published numerous peer-reviewed papers in the area of corrosion, and holds patents related to monitoring the wall thinning of piping components.

Exh. RIV000003 (Hopenfeld Testimony, at 2:7-8); Exh. RIV000004 (Hopenfeld CV at 4-5);

Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 5:27-6:10).

4. Dr. Hopenfeld also has particular experience with the CHECWORKS computer code that is used at U.S. nuclear power plants in relation to FAC. Exh. RIV000003 (Hopenfeld Testimony at 2:8-14); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 5:18-19). As an NRC project manager during the time CHECWORKS was developed, Dr. Hopenfeld has firsthand knowledge about the background and development of the CHECWORKS model. Tr. at 1319:22-25, 1320:1-7 (Hopenfeld); Tr. at 1330:14-16 (Hopenfeld).

B. Entergys Witnesses

5. Entergys testimony on Contention RK-TC-2 was presented by a panel of five witnesses: Ian D. Mew, Alan B. Cox, Nelson F. Azevedo, Dr. Jeffrey S. Horowitz, and Robert M. Aleksick.

13

6. Mr. Ian D. Mew is a Senior Engineer employed by Entergy at Indian Point responsible for FAC issues. Exh. ENT000029 (Entergy Testimony at A2, A4). Mr. Mews curriculum vitae does not demonstrate that he has an in-depth knowledge in all the fields required to understand the capabilities, limitations, acceptability of CHECWORKS, and to properly assess CHECWORKS predictions. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 8:28-30; 9:6-8). In particular, Mr. Mews curriculum vitae does not reflect expertise in mass transfer, nuclear safety analysis, electrochemistry, or materials, which would be demonstrated by technical publications in these fields. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 8:28-30); Exh. ENT000030 (CV of Ian Mew).

7. Mr. Alan B. Cox is employed by Entergy as the Technical Manager of License Renewal, and was responsible developing and reviewing the Indian Point Aging Management Program (AMP) for FAC. Exh. ENT000029 (Entergy Testimony at A6, A8). Mr. Coxs curriculum vitae does not demonstrate that he has an in-depth knowledge in all the fields required to understand the capabilities, limitations, acceptability of CHECWORKS, and to properly assess CHECWORKS predictions. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 8:30-9:2; 9:6-8). In particular, Mr. Coxs curriculum vitae indicates that he does not have expertise in mass transfer, nuclear safety analysis, electrochemistry, or materials, which would be demonstrated by technical publications in these areas. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 8:30-9:2; Exh. ENT000031 (CV of Alan Cox).

8. Mr. Nelson F. Azevedo is employed by Entergy as Supervisor of Code Programs at Indian Point, and has been responsible for FAC-related issues at the plant since 2001. Exh.

ENT000029 (Entergy Testimony at A10, A12). Mr. Azevedos curriculum vitae does not demonstrate that he has an in-depth knowledge in all the fields required to understand the 14

capabilities, limitations, acceptability of CHECWORKS, and to properly assess CHECWORKS predictions. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 9:2-8). In particular, Mr.

Azevedos curriculum vitae does not show that he has expertise in thermal hydraulics, nuclear safety analysis, or electrochemistry, as would be established by technical publications about such topics. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 9:2-8); Exh. ENT000032 (CV of Nelson Azevedo).

9. Dr. Horowitz is an independent consultant hired by Entergy in this proceeding.

Exh. ENT000029 (Entergy Testimony at A14). Dr. Horowitz was a co-developer of the CHECWORKS computer code and predecessor codes, and consultant to the Electric Power Research Institute (EPRI). Id. (Entergy Testimony at A17). As a co-developer of the CHECWORKS computer model and a consultant to EPRI, Dr. Horowitz has a direct financial interest in the promotion and use of CHECWORKS at nuclear power plants. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 9:10-13).

10. Mr. Aleksick is the president and founder of CSI Technologies, Inc., a company specializing in FAC services. Exh. ENT000029 (Entergy Testimony at A21). Since 1992, Mr.

Aleksick has been involved in preparing CHECWORKS models at Indian Point. Id. (Entergy Testimony at A23). Mr. Aleksicks curriculum vitae indicates that he does not have in-depth knowledge or publications in the field of nuclear safety risk assessment. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 9:15-16); Exh. ENT000037 (CV of Robert Aleksick). In addition, as the president and founder of CSI Technologies, Inc., which markets the application of CHECWORKS and is closely affiliated with EPRI and with the development and use of CHECWORKS, Mr. Aleksick has a financial interest in the use of the CHECWORKS computer model to manage FAC at nuclear power plants. Id. (Hopenfeld Rebuttal Testimony at 9:16-20).

15

C. NRC Staffs Witnesses

11. NRC Staffs testimony on Contention RK-TC-2 was presented by Dr. Allen Hiser and Matthew Yoder.

12. Dr. Hiser is a Senior Technical Advisor for License Renewal Aging Management in the Division of License Renewal of the Office of Nuclear Reactor Regulation at NRC. Exh.

NRC000121 (NRC Staff Testimony at A.1(b)). Dr. Hiser provides technical advice to NRCs Division of License Renewal on a variety of nuclear power plant aging management issues. Id.

(NRC Staff Testimony at A.2(b)).

13. Mr. Yoder is a Senior Chemical Engineer in the Steam Generator Tube Integrity and Chemical Engineering Branch in the Division of Engineering of the Office of Nuclear Reactor Regulation at NRC. Exh. NRC000121 (NRC Staff Testimony at A.1(a)). Mr. Yoder is responsible for conducing reviews of flow accelerated corrosion (FAC) programs for applicants for license renewal, the results of which are memorialized in NRC safety evaluations.

Id. (NRC Staff Testimony at A.2(a)).

II. THE NATURE AND SIGNIFICANCE OF FLOW ACCELERATED CORROSION A. Definition of FAC

14. Flow accelerated corrosion, or FAC, is a pipe wall thinning phenomenon in which the thinning rate is accelerated by flow velocity. Exh. RIV000003 (Hopenfeld Testimony at 3:29-4:5); Exh. RIVR00005 (Hopenfeld Expert Report, at 2-3). When metal pipes are exposed to flowing liquid, flow velocity has a significant effect on metal removal. Exh. RIV000003 (Hopenfeld Testimony at 3:30-31); Exh. RIVR00005 (Hopenfeld Expert Report, at 2-3).

15. FAC can affect various nuclear power plant components. This includes steam generator components such as inlet piping, nozzles, blow down piping, and distribution rings, 16

which are highly susceptible to FAC. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 15:27-28, 28:16-26); Exh. RIV000003 (Hopenfeld Testimony at 18:6-9); Tr. at 1516:4-15 (Hopenfeld); Exh. RIVR00008 (NRC IN 1991-019); Exh. RIV000009 (NRC Monitoring Report 5-93-0042).

16. The main causes of FAC include turbulence, intensity, steam quality, material compositions, oxygen content, and coolant pH. Exh. RIV000003 (Hopenfeld Testimony at 4:1-3); Exh. RIVR00005 (Hopenfeld Expert Report, at 2).

17. While Entergys program purporting to manage the aging effects of FAC at Indian Point defines FAC as limited to chemical dissolution mechanisms (see Exh. ENT000038 (EN-DC-315 at 6)), FAC is best understood to encompass not just chemical dissolution processes, but also mechanical processes. In particular, Dr. Hopenfeld explains that two different mechanisms can lead to FAC wall thinning: (1) physical removal of metal by mechanical forces (shear or impact), and (2) chemical or electrochemical dissolution of the metal. Exh. RIVR00005 (Hopenfeld Expert Report, at 2); Tr. at 1321:4-13, 1322:4-7 (Hopenfeld). In addition, in many circumstances, both mechanisms can occur simultaneously and synergistically. Exh.

RIVR00005 (Hopenfeld Expert Report, at 2); Tr. at 1321:14-16, 1331:3-5 (Hopenfeld); Tr. at 1442:6-1444:6 (Hopenfeld), Tr. at 1447:17-19 (Hopenfeld); Tr. at 1517:18-20 (Hopenfeld); Exh.

RIV000127 (Macdonald).

18. Dr. Hopenfeld testified that FAC includes wall thinning by impingement corrosion, electrochemical corrosion, erosion-corrosion, cavitation-erosion, and metal dissolution. Exh. RIV000003 (Hopenfeld Testimony at 3:31-4:1); see also Exh. RIVR00005 (Hopenfeld Expert Report, at 2); Tr. at 1435:22-1437:9 (Hopenfeld); Tr. at 1699:4-12, 1700:24-1701:11 (Hopenfeld). Dr. Hopenfeld further testified that Entergys arbitrary restrictive 17

definition of FAC improperly excludes wall thinning by cavitation, wet steam, galvanic corrosion, and jet impingement/erosion. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 29:3-5). All these phenomenon are properly considered forms of FAC, since all of these mechanisms are affected to one degree or another by flow velocities. Exh. RIVR00005 (Hopenfeld Expert Report, at 2); Tr. at 1701:2-6 (Hopenfeld).

19. Dr. Hopenfeld testified that Entergys narrow understanding that FAC is strictly controlled by chemical dissolution is not universally accepted, and that [o]ther researchers believe that fluid shear forces sufficiently large to remove the protective oxide film and thereby control wall thinning. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 29:10-12). For example, Digby D. Macdonalds theory on erosion/corrosion mechanisms, as discussed in a professional paper entitled The Point Defect Model for the Passive State, makes it clear that FAC is not controlled only by mass transfer processes. Tr. at 1323:15-17 (Hopenfeld); Exh.

RIV000127 (Macdonald). Entergy has only made bare statements, and not provided any data, to support a conclusion that FAC only involves chemical dissolution processes.

20. Numerous Entergy ultrasonic examination reports, which memorialize measurements of pipe wall thicknesses, demonstrate that metal loss observed in piping components at Indian Point is not controlled by chemical dissolution alone. Tr. at 1331:2-1332:10 (Hopenfeld). In particular, Dr. Hopenfeld explained that comparing the corrosion rate of a straight section of pipe to the corrosion rate of a curved section of pipe, such as an elbow, results in a maximum elbow to pipe mass transfer coefficient ratio, or, a metal loss ration; Dr.

Hopenfeld further explains that if simply chemical dissolution, mass transfer processes are present in pipes affected by FAC, this ratio would be less than 1.6, whereas ratios in excess of 1.6 are indicative that more than just chemical dissolution is at work, that erosion is also 18

occurring, and that FAC is not only controlled by mass transfer. Tr. at 1331:2-12, 22-25, 1332:1-10 (Hopenfeld). This explanation is supported by a professional paper by Jianrong Wang and Siamack A. Shirazi, entitled, A CFD Based Correlation for Mass Transfer Coefficient in Elbows. Exh. RIV000131 (Wang & Shirazi). Dr. Hopenfelds review of hundreds of pipe wall thickness measurements of components at Indian Point indicates that the metal loss ratio was far in excess of 1.6, varying from 6 to 52; these ratios for numerous piping components at Indian Point demonstrate that FAC is controlled by both chemical dissolution and erosion mechanisms.

Tr. at 1331:2-12, 22-25, 1332:1-10, 1562:17-19 (Hopenfeld).

21. One ultrasonic examination report discussed by Dr. Hopenfeld to demonstrate the point that FAC involves corrosion and erosion mechanisms related to a component that Entergy determined had lamination. Tr. at 1555:20-1556:19; Exh. RIV000130 (Entergy UT Examination Report). While Entergys witnesses claim that the wide variation of pipe wall thickness recorded for the particular component analyzed was not real, but instead an erroneous reading attributable to lamination, (Tr. at 1558:5-14 (Aleksick)), Dr. Hopenfeld explained that this one particular example in no way negated his analysis of many other Entergy ultrasonic examination reports containing component thickness measurements, which showed metal loss ratios in excess of 1.6, and that his analysis of so many other component measurements confirmed and bolstered his conclusion that more than a straightforward mass transfer phenomenon is occurring in FAC-susceptible components. Tr. at 1562:5-23, 1566:4-10, 1569:21-1570:6; 1574:14-22, 1577:4-1578:2, 1579:5-1580:4, 1580:22-24, 1581:19-1582:3, 1585:22-1586:3, 7-21; 1587:9-20, 1729:14-18, 1730:6-8, 1845:23-1846:10, 1847:10-1848:15, 1857:15-16 (Hopenfeld); Tr. at 1594:25-1595:19 (ALJ McDade, Hopenfeld); Exh. RIV000132 19

(Excerpt of IP FAC Inspection Report); Exh. RIV000133 (Excerpt of IP FAC Inspection Report).

22. For example, Dr. Hopenfeld explained how two Indian Point component examination reports, received into evidence as Exhibits RIV000132 and RIV000133, demonstrated that FAC does not involve only chemical dissolution processes. Tr. at 1845:23-1846:10, 1847:10-1848:15; Exh. RIV000132 (Excerpt of IP FAC Inspection Report); Exh.

RIV000133 (Excerpt of IP FAC Inspection Report). In relation to Exhibit RIV000132, an ultrasonic examination report of an elbow component, Dr. Hopenfeld explained how the ratio of the wall thinning in the straight section to the maximum . . . wear of material lost in the elbow is more than 1.6, indicating that FAC in the pipe is not only a chemical-controlled process. Tr. at 1845:23-1846:10 (Hopenfeld); Exh. RIV000132 (Excerpt of IP FAC Inspection Report at p.2 of 6). Similarly, in relation to Exhibit RIV000133, an ultrasonic examination report of a reducer, Dr. Hopenfeld explained that the changes in thickness measurements indicate that not simply diffusional control was at work. Tr. at 1847:10-22, 1854:21-1857:16 (Hopenfeld); Exh.

RIV000133 (Excerpt of IP FAC Inspection Report).

23. Entergy attempts to attribute variations in component wall thickness to original manufacturing fabrication. Tr. at 1876:22-1877:1, 1878:25-1879:9 (Fagg; Azevedo). However, Dr. Hopenfeld testimony refutes this position, testifying that new components generally have uniform wall thicknesses. Tr. at 1876:22-1878:17 (Hopenfeld). In addition, Dr. Hopenfelds observations about the high magnitude of variations in pipe wall thicknesses at Indian Point further refutes the notion that such differences are somehow attributable to original component manufacturing. See supra ¶¶ 20-22.

20

24. NRC recently acknowledged in a draft guidance document that [e]rosion mechanisms are sometimes perceived as being comparable to wall thinning due to FAC. Exh.

ENT000573 (Draft LR-ISG-2012-01 at 3). In recognition of the significance of mechanical mechanisms, NRCs recent draft regulatory guidance proposes to revise the definition of wall thinning in the GALL Report to include erosion mechanisms such as cavitation, flashing, droplet impingement, and solid particle impingement. Id. (Draft LR-ISG-2012-01 at 4); Tr. at 1825:22-25 (ALJ Wardwell); Tr. at 1700:1-11 (Hopenfeld); Tr. at 1700:24-1701:1-12 (Hopenfeld); see also Tr. at 1729:18-25 (Hopenfeld). In addition, the NRCs draft guidance indicates that AMPs for FAC may also address erosion mechanisms. Id. (Draft LR-ISG-2012-01 at Appendix D); Tr. at 1682:17-22 (Yoder); 1702:11-12 (Hiser).

25. Dr. Hopenfeld testified that while NRCs draft guidance represented an improvement in the agencys recognition of the significance of erosion mechanisms and the importance of managing such mechanism in the context of a FAC aging management program, the draft guidance still did not entirely correct the regulatory definition of FAC. In particular, Dr. Hopenfeld testified that NRCs draft guidance improperly continued to support the notion that FAC is purely a chemical dissolution process, despite academic opinions to the contrary. Tr.

at 1849:7-1850:6 (Hopenfeld); Tr. at 1700:1-11 (Hopenfeld); Tr. at 1700:24-1701:12 (Hopenfeld); see also Tr. at 1729:10-1730:8 (Hopenfeld).

B. The Local, Non-Linear Nature of FAC

26. Wall thinning resulting from FAC is a local phenomenon affected by local geometry, local metal composition, and local turbulences. Exh. RIV000003 (Hopenfeld Testimony at 4:3-4); Exh. RIVR00005 (Hopenfeld Expert Report, at 2).

21

27. FAC is inherently unpredictable. Exh. RIV000003 (Hopenfeld Testimony at 5:25); Exh. RIVR00005 (Hopenfeld Expert Report at 3).

28. Once local corrosion has begun, geometrical changes may further intensify local turbulence, and, as a result, FAC may progress at a non-linear rate. Exh. RIV000003 (Hopenfeld Testimony at 4:4-5); Exh. RIVR00005 (Hopenfeld Expert Report, at 2). As local turbulence and local flow velocity are not directly measured quantities, it is difficult to identify locations where FAC rates are highest. Exh. RIVR00005 (Hopenfeld Expert Report, at 2).

29. Entergys witness, Mr. Aleksick, testified to his opinion that FAC is a line level phenomenon. Tr. at 1553:7-9 (Aleksick). However, Riverkeepers witness, Dr. Hopenfeld explained that this is an improper characterization of FAC. In particular, Dr. Hopenfeld testified that FAC is best characterized as a component phenomenon, explaining that a line approach is a simplification. Tr. at 1841:2-1843:5 (Hopenfeld). Dr. Hopenfelds review of hundreds of Indian Point data points [a]ll conclusively show that it [FAC] is not a line phenomenon or an average phenomenon, but rather, that [i]t is a component phenomenon. Tr. at 1841:19-22, 1842:3-4 (Hopenfeld).

30. Entergys witnesses testified that FAC wear progresses in a linear fashion. Tr. at 1374:18-1375:2 (Aleksick). However, this testimony is not dispositive. As Dr. Hopenfeld testified, his experience and literature support the notion that FAC is a local phenomenon in which component geometry changes with time, and Entergys witnesses have failed to show any data to show that wear at Indian Point has been observed to be linear with time, excluding, possibly, straight pipes. Tr. at 1421:17-1422:7 (Hopenfeld); Tr. at 1542:19-1543:17 (Hopenfeld). In fact, Entergys witnesses conceded that Entergy has never actually analyzed Indian Point data for the purposes of determining the linearity of FAC wear rates. Tr. at 22

1766:12-24 (Aleksick); Tr. at 1767:15-1768:2 (Azevedo). Entergy has only made bare statements, and not provided any data, to support a conclusion that wear at Indian Point is linear and not local in nature.

31. Moreover, Entergys witnesses also acknowledged in the course of their testimony that, FAC might not be linear, such as when operating conditions have changed as a result of wear over the course of many years. Tr. at 1423:6-9 (Aleksick). Entergys witness, Mr. Aleksick, also conceded that FAC wear rates are not consistent within an elbow and that, as a result, [o]ne portion of the elbow may wear more rapidly than another. Tr. at 1423:22-25 (Aleksick); see also Tr. at 1555:12-14 (Aleksick). Mr. Aleksick also plainly acknowledged that of course, its [i.e., FAC] local phenomenon in the sense that the components wear locally and the degree of wear may vary from component to component based on its local geometry. Tr. at 1553:19-22 (Aleksick).

32. In contrast to Entergys witnesses, Riverkeepers witness, Dr. Hopenfeld, presented a solid technical basis to support his position that the rate of FAC can progress in a non-linear manner. In particular, Dr. Hopenfeld reviewed hundreds of grid measurements of Indian Point component wall thicknesses, which showed extreme changes in wall uniformity, confirming Dr. Hopenfelds conclusion about the local, non-linear nature of FAC at Indian Point.

Tr. at 1544:2-1545:8, 1545:14-18 (Hopenfeld).

33. One ultrasonic examination report discussed by Dr. Hopenfeld to demonstrate the point that FAC is a local, non-linear phenomenon related to a component that Entergy determined had lamination. Tr. at 1555:20-1556:19; Exh. RIV000130 (Entergy UT Examination Report). While Entergys witnesses claim that the wide variation of pipe wall thickness recorded for the particular component analyzed was not real, but instead an erroneous 23

reading attributable to lamination, (Tr. at 1558:5-14 (Aleksick)), Dr. Hopenfeld explained that this one particular example in no way negated his analysis of many other Entergy ultrasonic examination reports containing component thickness measurements, which demonstrated the local, non-linear nature of FAC at Indian Point. 1333:15-1334:5, 1441:24-1442:5, 1566:3-10, 1569:21-24, 1577:12-1578:2, 1857:15-16 (Hopenfeld).

34. In fact, Dr. Hopenfeld analyzed and discussed numerous other examples of non-linear wear at Indian Point as demonstrated by component thickness measurements. Tr. at 1547:1-7, 1579:5-1581:6, 1581:19-1582:3, 20-24, 1583:1, 1585:22-1586:2, 1586:9-21, 1587:9-13, 1592:13-25 (Hopenfeld); Exh. RIV000049 (IP Outage Report). For example, Dr. Hopenfeld explained how in one set of grid measurements, admitted in the proceeding as Exhibit RIV000049, it was not possible to conclude that wear on the component was linear. Tr. at 1585:22-1586:2, 1586:9-21, 1587:9-13, 1592:13-25 (Hopenfeld); Exh. RIV000049 (IP Outage Report). In relation to another set of grid measurements, admitted in the proceeding as Exhibit RIV000132, Dr. Hopenfeld explained how the measured wear indicated non-linear FAC rates, undermining Entergys witnesses conclusions about the linearity of FAC. Tr. at 1845:23-1847:6 (Hopenfeld); Exh. RIV000132 (Excerpt of IP FAC Inspection Report at p.2 of 6). Yet another example, admitted into evidence in the proceeding as Exhibit RIV000133, showed significant changes in wall thickness, once again demonstrating a non-linear, local phenomenon at work.

Tr. at 1847:10-1848:15 1854:21-1857:16 (Hopenfeld); Exh. RIV000133 (Excerpt of IP FAC Inspection Report).

35. Entergy attempts to attribute variations in component wall thickness to original manufacturing fabrication. Tr. at 1876:22-1877:1, 1878:25-1879:9 (Fagg; Azevedo). However, Dr. Hopenfeld testimony refutes this position, testifying that new components generally have 24

uniform wall thicknesses. Tr. at 1876:22-1878:17 (Hopenfeld). In addition, Dr. Hopenfelds observations about the high magnitude of variations in pipe wall thicknesses at Indian Point indicative of local, non-linear wear further refutes the notion that such differences are somehow attributable to original component manufacturing. See supra ¶¶ 32, 34.

36. Dr. Hopenfeld explained that numerous instances of leaks and component wall thinning at Indian Point were manifestations of the localized effects of FAC. Tr. at 1545:20-1546:7 (Hopenfeld). Several examples of such leaks and thinning events are memorialized in Entergy documentation that are evidence in this proceeding. Exh. RIV000024 (Entergy Operating Experience Report); Exh. RIV000025 (Entergy Daily Event Report), RIV000026 (Entergy Condition Report List), RIV000027 (Entergy Condition Report List), RIV000028 (Entergy Condition Report); RIV000029 (Entergy Condition Report).

37. U.S. and worldwide nuclear power plant operating experience further supports Dr.

Hopenfelds position that FAC is a non-linear, local phenomenon. For example, Dr. Hopenfeld explained that the FAC-related failure at the Surry nuclear power plant in 1986 involved uneven corrosion that occurred in an elbow component, which Dr. Hopenfeld physically observed. Tr. at 1514:22-1515:5 (Hopenfeld); Exh. RIV000006 (NRC IN 86-106); Exh. RIV000007 (NRC Bulletin 87-01). Dr. Hopenfeld also explained that a FAC-incident at San Onofre nuclear generating station, an extreme reduction in thickness was found in the distribution ring, at an area of very high turbulence. Tr. at 1516:4-15 (Hopenfeld); Exh. RIVR00008 (NRC IN 1991-019); Exh. RIV000009 (NRC Monitoring Report 5-93-0042). Dr. Hopenfeld also discussed how a FAC event that occurred at the Mihama power station in Japan was very, very local and completely unpredicted. Tr. at 1517:10-12; Exh. RIV000011 (NRC IN 2006-008). Dr.

Hopenfeld explained that a close review of the events that occurred at the Mihama station 25

showed a clear indication of how local the phenomenon was; the data in particular showed that there [was] no linearity between time and corrosion in relation to the FAC incident that occurred at Mihama. Tr. at 1530:11-1531:22.

38. Despite the evidence in the record establishing the local, non-linear nature of FAC, Entergys FAC program at Indian Point presumes a linear phenomenon for a fact. Tr. at 1837:24-1838:1 (Azevedo: ALJ McDade). This inhibits the ability of the FAC program at Indian Point from adequately detecting FAC. See Tr. at 1493:5-24 (Hopenfeld).

C. The Safety Significance of FAC

39. FAC poses a significant safety risk at nuclear power plants if left undetected.

Exh. RIV000003 (Hopenfeld Testimony at 4:8-19); Exh. RIVR00005 (Hopenfeld Expert Report, at 2-3). When FAC reduces wall thickness below the minimum design value, the subject component may leak or rupture. Exh. RIV000003 (Hopenfeld Testimony at 4:8-19); Exh.

RIVR00005 (Hopenfeld Expert Report at 2-3); Tr. at 1333:1-14 (Hopenfeld). A FAC-induced rupture of a high pressure component or pipe may have very serious and devastating safety consequences. Exh. RIV000003 (Hopenfeld Testimony at 4:8-19); Exh. RIVR00005 (Hopenfeld Expert Report at 2-3).

40. Numerous instances of undetected FAC at nuclear power plants have resulted in catastrophic events: (1) in 1986, a feed water pipe elbow ruptured at the Surry nuclear power plant resulting in several fatalities, see Exh. RIV000003 (Hopenfeld Testimony at 4:8-19); Exh.

RIVR00005 (Hopenfeld Expert Report, at 3); Exh. RIV000006 (NRC IN 86-106); Exh.

RIV000007 (NRC Bulletin 87-01); (2) at San Onofre, FAC resulted in failures of feed ring and J-tube components, see Exh. RIVR00005 (Hopenfeld Expert Report at 3); Exh. RIVR00008 (NRC IN 1991-019); Exh. RIV000009 (NRC Monitoring Report 5-93-0042); (3) in 1997, extraction 26

steam piping ruptured at the Fort Calhoun nuclear power station, see Exh. RIVR00005 (Hopenfeld Expert Report at 3); Exh. RIV000010 (NRC IN 1997-084); and (4) in 2004, FAC in a secondary loop at the Mihama nuclear power plant in Japan resulted in several worker fatalities. Exh. RIV000003 (Hopenfeld Testimony at 4:8-19); Exh. RIVR00005 (Hopenfeld Expert Report, at 3); Exh. RIV000011 (NRC IN 2006-008).

41. Undetected FAC may result in significant safety issues when plant components are subject to sudden transient loads, including pipe ruptures and damage to nearby and/or connected components. Exh. RIV000003 (Hopenfeld Testimony at 18:28-20:12); Exh.

RIVR00005 (Hopenfeld Expert Report, at 24-25); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 45-46). In particular, undetected FAC poses a risk of loss of coolant accidents (LOCA), since the probability of a pipe failing under a given load is reduced when the pipe walls have been degraded. Exh. RIV000003 (Hopenfeld Testimony at 18:28-20:12); Exh.

RIVR00005 (Hopenfeld Expert Report, at 24-25). For example, earthquake loads (which pose a higher risk than previously thought in the region where Indian Point is located) and station blackouts may result in a LOCA due to FAC-degraded components. Exh. RIV000003 (Hopenfeld Testimony at 18:28-20:12); Exh. RIVR00005 (Hopenfeld Expert Report, at 24-25);

Exh. RIV000031 (Sykes Study); Exh. RIV000032 (NRC GI-199); Exh. RIV000033 (Dedman Article). In addition, synergistic effects of an aging phenomenon known as metal fatigue, on piping components degraded by FAC, may also pose risks for severe accidents with significant safety consequences. Exh. RIV000003 (Hopenfeld Testimony at 18:28-20:12); Exh.

RIVR00005 (Hopenfeld Expert Report, at 24-25); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 45-46).

27

III.

SUMMARY

OF ENTERGYS PROGRAM FOR MANAGING FLOW ACCELERATED CORROSION

42. As described in sections A.2.1.14 and B.1.15 of Entergys LRA, Entergys FAC AMP as [a]n existing program that applies to safety-related and non-safety related carbon and low alloy steel components in systems containing high-energy fluids carrying two-phase or single-phase high energy fluid 2% of plant operating time. Exhibit ENT00015A-B (Indian Point Energy Center LRA (April 2007) at §§ A.2.1.14, p.A-24, B.1.15, p.B-54. The program applies to a subset of total plant systems that Entergy had categorized as FAC-susceptible.

Exh. ENT000029 (Entergy Testimony at 67).

43. Entergys LRA explains that Entergys program is based on EPRI guidelines in Nuclear Safety Analysis Center (NSAC)-202L, Recommendations for an Effective Flow-Accelerated Corrosion Program, (Exh. RIV000012 (NSAC-202L)) which purports to outline how licensees should predict, detect, and monitor FAC in piping and other pressure retaining components. Exhibit ENT00015A-B (Indian Point Energy Center LRA (April 2007) at §§ A.2.1.14, p.A-24, B.1.15, p.B-54.

44. The LRA explains that the FAC program includes (a) an evaluation to determine critical locations, (b) initial operational inspections to determine the extent of thinning at these locations, and (c) follow-up inspections to confirm predictions, or repair or replace components as necessary. Exhibit ENT00015A-B (Indian Point Energy Center LRA (April 2007) at §§ A.2.1.14, p.A-24, B.1.15, p.B-54.

45. Entergy claims that the FAC program at Indian Point includes the ten program elements identified in SRP-LR and the GALL Report. Exh. RIV000014 (AMP Evaluation Report, IP-RPT-06-LRD07, Revision 5).

28

46. Entergy purports to implement its FAC program, and the guidance and recommendations contained in the GALL Report and NSAC-202L pertaining thereto, via a fleet-wide procedure, EN-DC-315, Revision 6, Flow Accelerated Corrosion Program. Exh.

ENT000029 (Entergy Testimony at A60); Exh. ENT000038 (EN-DC-315, Rev. 6). This procedure requires piping and piping component inspections to be conducted, and ultrasonic thickness measurements to be performed to determine pipe wall thickness. Exh. ENT000038 (EN-DC-315, Rev. 6).

47. Entergys FAC program involves modelable/modeled systems and/or lines, and susceptible non-modeled (SNM) systems and/or lines. Exh. ENT000029 (Entergy Testimony at A68). Modeled systems and lines include major piping systems that typically pose the highest consequence of failure. Exh. ENT000029 (Entergy Testimony at A68). These lines are modeled in the computer code CHECWORKS. Id. SNM systems and/lines at Indian Point consist of piping susceptible to FAC but not modeled in the computer code CHECWORKS. Id.

48. During refueling outages, Entergy undertakes FAC-related inspections. Entergys witness, Mr. Aleksick explained that during [e]ach outage we inspect perhaps 100 components, perhaps 50 of those have been inspected before. Perhaps 50 of them are new. Tr. at 1507:4-8 (Aleksick); see also Tr. at 1603:7-11, 1869:15-22 (Aleksick). The goal of the FAC inspection program at Indian Point is to ensure that components are inspected prior to reaching a critical wall thickness. Tr. at 1674:1-5 (Yoder); Exhibit NYS00146C (GALL Report, Revision 1, at pp.

XI M-61 to XI M-62); Exhibit NYS00147D (GALL Report, Revision 2, at pp. XI M17-1, XI M17-2).

29

49. For modeled components, Entergys method of selecting components for wall measurements and scheduling inspections is based in part on predictions generated from the computer code, CHECWORKS. Exh. ENT000029 (Entergy Testimony at A72, A73). The guidance contained in EPRIs guidance document, NSAC-202L, and in Entergys implementing procedure, EN-DC-315, makes it clear that Entergys FAC management program relies on the use of the computer program CHECWORKS to predict timing and locations of wall thinning.

Entergy asserts that the criteria for selecting components for FAC inspections of modeled components are consistent with the criteria in NSAC-202L, and that the selection is based on (1) the trending of pipe wall thickness measurements from past outages; (2) predictive evaluations performed using the CHECWORKS code; (3) industry and IPEC-specific operating experience related to FAC; (4) results from other plant inspection programs . . . and (5) engineering judgment. Id. (Entergy Testimony at A72); Exh. RIV000012 (NSAC-202L).

50. At Indian Point, there are approximately 8,000 components across both units that are modeled in CHECWORKS, which are contained in 40 analysis lines per unit; approximately 3,700 of those [8,000 components] have been inspected since 1992, i.e., over the plants lifetime. Tr. at 1507:23-1508:2, 5-11, 1748:9-10, 1752:4-7, 1862:21-1863:15, 1619:22-24 (Aleksick). Entergys witnesses testify that the approximate 8,000 components modeled in CHECWORKS represents 22% of susceptible lines at Indian Point Unit 2, and 20% of susceptible lines at Indian Point Unit 3. Exh. ENT000029 (Entergy Testimony at A76).

51. For SNM components, Entergys method of selecting components for wall measurements and scheduling inspections is based on a separate analysis that does not use CHECWORKS. Exh. ENT000029 (Entergy Testimony at A72); Tr. at 1509:9-12 (Aleksick).

30

52. Entergys witness, Mr. Aleksick, estimated that there are approximately 700 SNM lines per unit, but indicated that Entergy does not count every component in those lines and that it was too difficult to estimate the number of SNM components covered in the FAC program. Tr. at 1865:12-18 (Aleksick); Tr. at 1866:19-1867:2 (Aleksick). Entergys witnesses also indicated that they did not know precisely how many SNM components at Indian Point had been inspected over the life of the plant, but did indicate that roughly 50% of such components had been inspected to date. Tr. at 1865:12-18 (Aleksick); Tr. at 1866:19-1867:2 (Aleksick); Tr.

1867:11-1867:16, 1867:24-1868:9 (Mew).

53. Entergy purports to follow the guidance contained in NSAC-202L with respect to inspection methodology, and reviewing inspection results and implementing corrective actions such as follow-up inspections, component repair, or component replacement. Exh. ENT000029 (Entergy Testimony at A78).

54. Entergys FAC program does not consider the effect of FAC on risk-significant, FAC-susceptible components in the steam generators. Exh. ENT000029 (Entergy Testimony at A64); Tr. at 1522:6-15 (Azevedo); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 28:16-17).

IV. THE CHECWORKS COMPUTER MODEL A. The Nature of the CHECWORKS Computer Code

55. The CHECWORKS computer code is software that was developed as a predictive tool to assist utilities in planning inspections and evaluating the inspection data to prevent piping failures caused by FAC. Exh. RIV000012 (NSAC-202L); Exh. RIV000003 (Hopenfeld Testimony at 4:22-25).

31

56. Because FAC is an unpredictable phenomenon, CHECWORKS is based on statistics, meaning that it is based on a collection of selective data which represents only a fraction of the total flow area. Exh. RIV000003 (Hopenfeld Testimony at 4:25-27); Exh.

RIVR00005 (Hopenfeld Expert Report, at 3). As a result, CHECWORKS will not produce reliable predictive results unless it is adequately calibrated, or benchmarked. Exh. RIV000003 (Hopenfeld Testimony at 4:27-5:3); Exh. RIVR00005 (Hopenfeld Expert Report, at 3-4). This notion of calibration and benchmarking is different from Entergys use of the terms calibrated and non-calibrated in relation to CHECWORKS analysis lines. Tr. at 180713-22 (Hopenfeld, Wardwell).

57. The NRC has recognized that benchmarking analytic codes is necessary, stating that analytical methods and codes are assessed and benchmarked against measurement data. . . .

The validation and benchmarking process provides the means to establish the associated biases and uncertainties. Exh. RIV000013 (VY Safety Eval. at § 2.8.7.1).

58. When plant parameters change, re-calibration of the CHECWORKS code to update the model becomes necessary. Exh. RIV000003 (Hopenfeld Testimony at 4:27-5:3);

Exh. RIVR00005 (Hopenfeld Expert Report, at 3-4). This is because changes in power output affect various plant parameters, including velocities, temperatures, coolant chemistry, and steam moisture. Exh. RIV000003 (Hopenfeld Testimony at 4:27-5:3); Exh. RIVR00005 (Hopenfeld Expert Report, at 4). EPRI has acknowledged that even small power uprates can have a significant affect on FAC rates. Exh. RIV000012 (NSAC-202L).

32

B. Problematic Assumptions Inherent in the CHECWORKS Model

59. Several assumptions underlie the CHECWORKS computer code, which greatly limit the usefulness of the model for accurately predicting component locations susceptible to FAC.

60. First, the CHECWORKS code assumes that wall thinning by FAC is controlled solely by chemical dissolution. Tr. at 1320:10-1322:7, 1322:22-1323:3, 1323:9-18, 1331:4-5, 1334:12-19, 1432:6-20, 1437:10-15 (Hopenfeld); Tr. at 1437:23-25 (Aleksick); Tr. at 1703:24-1704:3 (Hiser). As articulated in the findings discussed above, existing theoretical and experimental data, including data from Indian Point, indicate that FAC is controlled by both dissolution and erosion. See supra ¶¶ 17-25. Thus, the assumption underlying the CHECWORKS model about the definition of FAC is a fundamental flaw, since the model does address all relevant forms of component wall thinning. See id.; see also Tr. at 1320:10-1322:7, 1322:22-1323:3, 1323:9-18, 1331:4-5, 1736:14-1736:19, 1737:4-1738:6, 1738:8-17 (Hopenfeld).

This underlying flaw makes CHECWORKS deficient for purposes of detecting FAC at Indian Point. Tr. at 1735:5-1736:19 (Hopenfeld).

61. Second, the CHECWORKS code does not acknowledge that FAC is a local, non-linear phenomenon. As articulated in findings discussed above, Entergys position that FAC is local and non-linear is not dispositive, and Indian Point-specific data confirms that FAC at the plant is in fact properly characterized as a local, non-linear phenomenon. See supra ¶¶ 26-38.

Unpredictable component to component local thinning variations are possible since geometry changes with time. See id. Instead of recognizing the local nature of FAC, CHECWORKS was designed to predict average thinning rates by corrosion, instead of local rates by either corrosion or erosion. See Tr. at 1332:18-25, 1334:12-19, 1329:7-13, 1433:8-10 (Hopenfeld). Entergys 33

witness, Mr. Aleksick explained that CHECWORKS provides Entergy with a single wear rate per component. Tr. at 1428:5 (Aleksick). For example, Mr. Aleksick testified that CHECWORKS provides a single prediction of wear rates and wear for the entire elbow. Tr. at 1427:8-10 (Aleksick). Riverkeepers witness, Dr. Hopenfeld, further testified that the CHECWORKS equation cannot report maximum values, and that as a result, predictions are based on averages. Tr. at 1433:2-10, 1446:20-21 (Hopenfeld). Dr. Hopenfeld refuted Entergys witnesses representations that CHECWORKS is not based on averages. Tr. at 1654:24-1655:3, 9-16 (Horowitz). While Entergys witnesses state that CHECWORKS accounts for geometrical effects on FAC, in actuality, the model only accounts for the differences in the average values between components; for example, the wear in an elbow may be on the average two times larger than the average wear in the straight section, but at the turning point on the elbow, the local value may be ten times larger. Tr. at 1654:24-1655:3, 9-16 (Horowitz); Tr. at 1513:12-13, 23-24, 1514:2-17 (Hopenfeld). Thus, CHECWORKS cannot accurately predict wear in components with complex geometries. Tr. at 1513:12-13, 23-24, 1514:2-17 (Hopenfeld).

62. Third, CHECWORKS assumes that component chromium content is known, when in fact, chromium content represents an uncertainty in the CHECWORKS model. See Exh. ENT000029 (Entergy Testimony at A110); Tr. at 1323:23-1324:19-1326:5, 1729:2-9 (Hopenfeld); Tr. at 1645:12-18 (Horowitz). Entergys witness Mr. Azevedo indicated that while Entergy has the ability to measure chromium content, such measurements are only performed on a limited basis, and not conducted enough to affect CHECWORKS results. Tr. at 1748:19-1749:4, 8-14 (Azevedo). Dr. Hopenfeld explained that FAC is very sensitive to the chrome content and that an uncertainty in chromium alone could affect F AC wear by an order of magnitude. Tr. at 1323:23-1324:19 (Hopenfeld); Exh. RIV000108 (Hopenfeld Rebuttal 34

Testimony at 31:3); see also Tr. at 1325:10-11 (Hopenfeld). Dr. Hopenfeld testified that a factor of six in chromium uncertainty leads to a factor of 10 in FAC uncertainty. See Tr. at 1327:5-9 (Hopenfeld).

C. The Role of CHECWORKS at Indian Point

63. Entergys witnesses testify that CHECWORKS is a tool used at Indian Point to select and schedule piping components for inspection. Exh. ENT000029 (Entergy Testimony at A73). As Entergys witness, Mr. Aleksick explained, CHECWORKS is a tool used to help select inspection locations, to increase that population of inspected components. Tr. at 1299:8-11 (Aleksick).

64. As noted above, 22% of FAC-susceptible components are modeled at Indian Point Unit 2 and 20% of FAC-susceptible components are modeled at Indian Point Unit 3, totaling approximately 8,000 modeled components. See supra ¶ 50.

65. Also noted above, since the inception of the FAC program at Indian Point in 1992, less than half of these modeled components, 3,700, have actually been inspected. See supra ¶ 50.

66. In a given inspection scope for inspections conducted during plant refueling outages, upwards of 38% of inspections are based on CHECWORKS predictions. Exh.

ENT000029 (Entergy Testimony at Figure 1 - IP2 Inspection Scope Distribution and Figure 2 -

IP3 Inspection Scope Distribution). During five of the past ten inspection outages conducted at Indian Point Units 2 and 3, CHECWORKS was attributed to approximately one-third or more of inspection locations. Id.; see also Tr. at 1603:16-18 (Aleksick).

67. In relation to new components selected for inspections, Entergys witness, Mr.

Aleksick explained that roughly half of those come from CHECWORKS, sometimes more, 35

sometimes less, but thats the general order of magnitude. Tr. at 1486:7-18 (Aleksick); see also 1507:1-8 (Aleksick). Mr. Aleksick indicated that for those uninspected components, its extremely important to have the CHECWORKS predictions. Tr. at 1299:5-7 (Aleksick). Mr.

Aleksick testified that the role CHECWORKS plays at Indian Point is to predict whats going on in uninspected components. Tr. at 1641:24-1642:1 (Aleksick).

68. In relation to components modeled in CHECWORKS, Entergys witness, Mr.

Aleksick testified that its fair to say that CHECWORKS is the primary tool used in identifying and selecting them for inspection. Tr. at 1502:22-1503:6 (Aleksick). Mr. Aleksick clarified that while about half of new inspections are informed by CHECWORKS, of the new inspections of modeled components, virtually all of them are informed by CHECWORKS. Tr.

at 1541:10-14 (Aleksick (emphasis added)).

69. In relation to reinspected, i.e., non-new, components, [r]oughly half were initially identified for inspection through CHECWORKS. Tr. at 1609:9-25 (Aleksick; ALJ McDade). In addition, earlier in the Indian Point plant inspection history, that percentage would be larger. Tr. at 1610:2-5 (Aleksick).

70. Despite Entergys witnesses representations that CHECWORKS only plays a minor role in the FAC program at Indian Point (see, e.g., Tr. at 1479:18-20, 1480:12-18 (Mew);

Exh. ENT000029 (Entergy Testimony at A73, A93); Tr. at 1739:6-11, 1739:24-1740:1 (ALJ McDade)), the evidence in the record indicates that CHECWORKS plays an integral and primary role in determinations at Indian Point regarding what new modeled components to inspect for FAC-related degradation. In addition, Entergys witness, Mr. Cox testified that determinations about remaining service life and resulting correct actions are based on the output of the CHECWORKS program. Tr. at 1478:12-25 (Cox).

36

71. There are certain categories of components for which CHECWORKS is not used.

This includes a range of SNM components, including small bore piping. Exh. ENT000029 (Entergy Testimony at A68); see supra ¶ 47. In addition, safety-significant and FAC susceptible components with the steam generators are not monitored by CHECWORKS. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 28:16-18); Tr. at 1335:6-16 (Hopenfeld).

D. The Predictive Performance of CHECWORKS at Indian Point

72. The predictive capability of the CHECWORKS computer code can be evaluated by comparing wall thickness predictions generated by CHECWORKS with actual thickness measurements. Exh. RIVR00005 (Hopenfeld Expert Report, at 5); Tr. at 1304:22-1305:1 (Aleksick). Entergys witness, Mr. Aleksick, testified that CHECWORKS results are a good way to assess the validity of the models predictions. Tr. at 1299:25-1300:2 (Aleksick). In addition, Entergys witness, Dr. Horowitz, agreed that such comparisons provide a measure of the accuracy of the predictive capabilities of the CHECWORKS model. Tr. at 1647:15-24 (Horowitz, ALJ Wardwell).

73. Several decades-worth of graphical CHECWORKS data pertaining to code predictions versus actual component wall thickness was made available by Entergy in this proceeding. Exh. RIV000003 (Hopenfeld Testimony at 5-7); Exh. RIVR00005 (Hopenfeld Expert Report, at 5); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 34:14-16); Exh.

RIV00016A (CHECWORKS Graphs) and RIV00016B (CHECWORKS Graphs); Exh.

RIV000112 (CHECWORKS Graphs). As Entergys witness, Mr. Aleksick explains, CHECWORKS creates charts, scatter plots showing predicted values versus measured values, so that the analyst can assess, for each analysis line, the degree to which the program accurately reflects field conditions. Tr. at 1304:22-1305:1 (Aleksick).

37

74. The data produced in this proceeding was generated both before and after the changes in plant operating conditions at Indian Point due to power uprates that occurred at Unit 2 in 2004, and at Unit 3 and 2005. Exh. RIV000003 (Hopenfeld Testimony at 5:28-6:1); Exh.

RIVR00005 (Hopenfeld Expert Report, at 5); Exh. RIV00016A (CHECWORKS Graphs) and RIV00016B (CHECWORKS Graphs); Exh. RIV000112 (CHECWORKS Graphs).

75. Riverkeepers witness, Dr. Hopenfeld, testified that he reviewed this data, constituting several thousand data points and several hundred graphs. Exh. RIV000003 (Hopenfeld Testimony at 5-7); Exh. RIVR00005 (Hopenfeld Expert Report, at 5); Exh.

RIVR00005 (Hopenfeld Expert Report, at 5); Exh. RIV00016A (CHECWORKS Graphs) and RIV00016B (CHECWORKS Graphs); Exh. RIV000112 (CHECWORKS Graphs). Based on his review and analysis of the CHECWORKS data that was provided, Dr. Hopenfeld concluded that the computer model as employed at Indian Point is highly inaccurate and produces results that demonstrate a complete lack of correlation between component wear predictions and actual wall thickness measurements. Exh. RIV000003 (Hopenfeld Testimony at 5-7); Exh. RIVR00005 (Hopenfeld Expert Report, at 5-6); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 34:14-28). Dr. Hopenfeld explained that the CHECWORKS model is highly inconsistent, with widely varying results and significant outliers. Tr. at 1537:10-1538:18 (ALJ McDade, Hopenfeld). Dr.

Hopenfeld characterized his findings about CHECWORKS as follows: All their predictions are very bad. Tr. at 1813:24-25 (Hopenfeld).

76. Dr. Hopenfeld explains that while a perfect correlation would manifest as data falling on the 45° line that appears in each graph, instead, most of the data reviewed exhibited a wide scatter. Exh. RIV000003 (Hopenfeld Testimony at 6); Exh. RIVR00005 (Hopenfeld Expert Report, at 6); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 34:14-28). Entergys 38

witness also testified that the 45° line, better termed a middle line since it does not actually represent 45°, represents the ideal. Tr. at 1625:25-1626:14 (Aleksick). While Entergys witness, Mr. Aleksick conceded that some of data exhibited poor agreement between predictions and field observations, he also testified that some of the predictions do agree very well with the field observations. Tr. at 1305:2-5 (Aleksick). This testimony is not dispositive, and is in fact undermined by the testimony and analysis of Dr. Hopenfeld, which demonstrated that most of data plots exhibited disagreement. Exh. RIV000003 (Hopenfeld Testimony at 6);

Exh. RIVR00005 (Hopenfeld Expert Report, at 6); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 34:14-28); Tr. at 1813:24-25 (Hopenfeld).

77. As the x-axes of the CHECWORKS graphs represent a prediction of zero wear, data points that fall between e 45° line and x-axis represent non-conservative predictions. Exh.

RIV000003 (Hopenfeld Testimony at 6:12-13); Exh. RIVR00005 (Hopenfeld Expert Report, at 6). Dr. Hopenfelds review of all of Entergys plotted data points revealed that CHECWORKS yielded non-conservative predictions about 40-60% of the time. Exh. RIV000003 (Hopenfeld Testimony at 6:13-14); Exh. RIVR00005 (Hopenfeld Expert Report, at 6, Table 1 Column A);

Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 34, 37, 48); Tr. at 1664:19-20, 1727:6-8, Tr.

at 1813:24-25 (Hopenfeld). Dr. Hopenfelds rearrangement of a CHECWORKS plot illustrates the magnitude of the models non-conservatism. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 48); Exh. RIV000111 (Re-plotted CHECWORKS data); Tr. at 1661:4-1664:6 (Hopenfeld).

78. Dr. Hopenfeld also observed that many CHECWORKS predictions yield widely different measured points, even though with an ideal correlation, each predicted point would 39

have a single measured value. Exh. RIV000003 (Hopenfeld Testimony at 6:16-18); Exh.

RIVR00005 (Hopenfeld Expert Report at 6).

79. Dr. Hopenfeld testified that the degree of inaccuracy of Entergys actual wear measurements to CHECWORKS predictions has been very high. In particular, Dr. Hopenfeld observed that many CHECWORKS predictions varied by at least a factor of 2, and upwards of by a factor of 10, or higher. Exh. RIV000003 (Hopenfeld Testimony at 6:18-29); Exh.

RIVR00005 (Hopenfeld Expert Report at 6-8, Table 1); Tr. at 1317:22-25, 1328:5-11, 1446:10-12, 1650:22-24, 1658:25-1660:21, 16639-15 (Hopenfeld). The high degree of over- and under-predicting wear on a significant number of components indicates that CHECWORKS cannot predict FAC at Indian Point with any degree of precision, but rather only predict a range of corrosion that is far too wide for practical applications. Exh. RIV000003 (Hopenfeld Testimony at 6-7); Exh. RIVR00005 (Hopenfeld Expert Report at 7-8).

80. Two lines designated by a +50% and -50% are present on every graph reviewed by Dr. Hopenfeld. Dr. Hopenfeld explained how these lines are misleading, and that data falling within this range actually represents a wide margin of error. RIV000003 (Hopenfeld Testimony at 6:18-29); Exh. RIVR00005 (Hopenfeld Expert Report at 6-8); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 24:13-26:2). In particular, the data within these two lines is not bounded within 50%, but rather, indicates a wear ratio varying by a factor of .7 and 2 for the +50% and -

50%, respectively. RIV000003 (Hopenfeld Testimony at 6:18-29); Exh. RIVR00005 (Hopenfeld Expert Report at 6-8). Entergys witnesses confirmed that these arbitrary lines are confusing and improperly labeled. Tr. at 1631:1-4, 9-15 (Aleksick); Tr. at 1631:16-22 (Cox).

81. Dr. Hopenfeld explained that many data points he reviewed fell outside of the arbitrary and misleading +/-50% lines, indicating inaccurate predictions by upwards of a factor 40

of 10, or more. RIV000003 (Hopenfeld Testimony at 6:25-28); Exh. RIVR00005 (Hopenfeld Expert Report at 6-8, Table 1 Column B).

82. Each CHECWORKS data plot includes a line correction factor or LCF.

RIV000003 (Hopenfeld Testimony at 7:9-23); Exh. RIVR00005 (Hopenfeld Expert Report at 8);

Tr. at 1304:17-21, 1309:11-21, 1310:11-13 (Aleksick). The LCF indicates the degree to which CHECWORKS over or under-predicts wear. Exh. RIV000020 (CHECWORKS SFA Model Report Excerpt at 26). While an LCF of 1 would indicate an exact agreement between CHECWORKS predictions and actual wall thickness measurements (Id.), and LCF greater than 1, measured wear is higher than predicted wear, and vice versa. See id.; Tr. at 1311:5-18 (Horowitz).

83. Entergy has stated that reasonable LCF should be between 0.5 and 2.5. Exh.

RIV000020 (CHECWORKS SFA Model Report Excerpt at 26). However, Dr. Hopenfeld observed that Entergy has failed to justify the conclusion that this range is acceptable, or how a data plot with an LCF within this range would indicate that CHECWORKS can be used to accurately predict inspection locations. Exh. RIV000003 (Hopenfeld Testimony at 7:17-20);

Exh. RIVR00005 (Hopenfeld Expert Report at 8); Tr. at 1754:4-20 (Horowitz). It is not clear at all that this range is based on safety considerations, as it should be. Tr. at 1756:15-1757:12 (Hopenfeld).

84. Dr. Hopenfeld has observed that none of the CHECWORKS graphs he reviewed had an LCF of 1, except for those figures with no data in them. Exh. RIVR00005 (Hopenfeld Expert Report at 8). He further observed that the LCF was reported to be outside Entergys arbitrary acceptable range numerous times, a clear demonstration that CHECWORKS is 41

unreasonably failing to predict wear rates. Exh. RIV000003 (Hopenfeld Testimony at 7:20-23);

Exh. RIVR00005 (Hopenfeld Expert Report at 8, Table 1 Column C).

85. Dr. Hopenfelds assessment of the predictive accuracy of CHECWORKS using 11 years worth of data pre- and post- dating power uprates that occurred at Indian Point revealed that CHECWORKS predictions are not improving at all with time, indicating an ongoing lack of adequate benchmarking. Exh. RIV000003 (Hopenfeld Testimony at 10); Exh. RIVR00005 (Hopenfeld Expert Report at 13); Tr. at 1755:10-25 (Hopenfeld). The data further indicates that CHECWORKS cannot be successfully calibrated or benchmarked in the future prior to the start of Entergys proposed PEO, or otherwise. Exh. RIV000003 (Hopenfeld Testimony at 9:28-10:7); Exh. RIVR00005 (Hopenfeld Expert Report at 13, 19); Ex. RIV000108 (Hopenfeld Rebuttal Testimony at 30:8-31:11); Tr. at 1727:21-1728:10, 1732:16-24 (Hopenfeld). Dr.

Hopenfeld also indicated that benchmarking the CHECWORKS model at Indian Point is impossible in light of the faulty assumptions about the definition of FAC, and chromium content levels in modeled components. Tr. at 1728:22-1730:15, 1737:4-1738:17 (Hopenfeld).

86. The lack of benchmarking in the CHECWORKS model at Indian Point indicates that the model has failed to fully account for changes in plant parameters that have occurred at Indian Point. Exh. RIV000003 (Hopenfeld Testimony at 5-8, 9:28-10:7); Exh. RIVR00005 (Hopenfeld Expert Report at 13). This refutes Entergys position that the CHECWORKS model adequately accounts for changes in plant operating conditions. Tr. at 1298:14-24, 1747:6-13 (Aleksick).

87. Based on Dr. Hopenfelds explanation about the highly erratic behavior of CHECWORKS, and the lack of correlation exhibited between CHECWORKS predictions and actual measurements, CHECWORKS is ineffective for objective quantitative assessments, and 42

not a useful or reliable tool for determining inspection locations for modeled components to timely detect and mitigate FAC during the proposed PEO. Exh. RIV000003 (Hopenfeld Testimony at 5-10); Exh. RIVR00005 (Hopenfeld Expert Report at 5-13); Tr. at 1532: 25-1533:20 (Hopenfeld); Tr. at 1650:25-1651:20 (Hopenfeld).

E. The Implication of Poor Predictive Accuracy of CHECWORKS at Indian Point

88. CHECWORKS predictions of wall thinning by FAC at Indian Point plant are too inaccurate to prevent pipe wall thicknesses from being reduced below minimum design values.

Exh. RIV000003 (Hopenfeld Testimony at 10:9-11:4); Exh. RIVR00005 (Hopenfeld Expert Report at 13-15). In particular, more than half of modeled, FAC-susceptible components have never been inspected, and CHECWORKS predictions, which are used to screen and choose locations for inspection, will not accurately predict where thinning below minimum acceptable values will and will not occur in these components, and/or adequately inform where representative, i.e., a limited and relatively small number of, inspections should and will be undertaken. Exh. RIV000003 (Hopenfeld Testimony at 10:9-11:4); Exh. RIVR00005 (Hopenfeld Expert Report at 13-15); Tr. at 1449:4-14, 1587:21-25, 1665:24-1666:8, 1669:5-1670:4, 1740:12-22, 1741:1-9 (Hopenfeld). The failure to inspect components that should be inspected as a result of relying on CHECWORKS predictions makes it impossible to pursue correct or adequate decisions about which modeled components require corrective action. Exh.

RIV000003 (Hopenfeld Testimony at 10:9-11:4); Exh. RIVR00005 (Hopenfeld Expert Report at 13-15); Tr. at 1449:4-14, 1665:24-1666:8, 1740:12-22, 1741:1-9 (Hopenfeld). Entergys witness, Mr. Azevedo conceded that modeled components with wall thinning below allowable limits would not be identified for inspection in the event a CHECWORKS prediction indicates wear is low, when in fact, wear on the component is high, and that such components would never 43

be evaluated under Entergys criteria for undertaking corrective action. Tr. at 1671:18-24, 1672:1-20 (Azevedo). Mr. Azevedo further acknowledged that Entergys use of CHECWORKS to date has only provided a sense of where future inspections should occur, [f]or the locations that we [Entergy] have inspected. Tr. at 1360:15-22 (Azevedo).

89. The increase in operating life from 40 to 60 years represents a significant potential for pipe wall thicknesses to fall below designated minimum critical design levels during extended operations, and it can be expected that an increasing number of components will become prone to failure after 40 years of service. Exh. RIVR00005 (Hopenfeld Expert Report at 13-15).

90. Even small changes in the corrosion rate can result in unacceptable levels of FAC.

Exh. RIVR00005 (Hopenfeld Expert Report at 14).

91. The use of the inaccurate, non-conservative CHECWORKS model may result in unsafe plant operations. Exh. RIV000003 (Hopenfeld Testimony at 10:9-11:4); Exh.

RIVR00005 (Hopenfeld Expert Report at 13-15). CHECWORKS predictions affect plant safety because they fail to accurately indicate when modeled, previously uninspected components are reaching a critical wall thickness. Exh. RIV000003 (Hopenfeld Testimony at 10:9-11:4); Exh.

RIVR00005 (Hopenfeld Expert Report at 13-15). This results in incorrect component ranking or prioritization, and ultimately untimely component inspection corrective action as a result. Exh.

RIV000003 (Hopenfeld Testimony at 10:9-11:4); Exh. RIVR00005 (Hopenfeld Expert Report at 13-15).

92. Conservative predictions can affect plant safety as well. Entergy attributes findings that components have low remaining life and should be replaced to an often overpredicted wear value by CHECWORKS. Exh. RIV000021 (CHECWORKS SFA Model 44

Report at Appendix K). This assumption is incorrect, since CHECWORKS produces non-conservative results upwards of 60% of the time. Exh. RIV000003 (Hopenfeld Testimony at 10:26-11:4); Exh. RIVR00005 (Hopenfeld Expert Report at 15). If predictions are commonly perceived to be based on conservative estimates, component replacement could be incorrectly postponed, which could result in an excessive, undetected wall thinning, thus posing the potential for component malfunction and safety consequences. Exh. RIV000003 (Hopenfeld Testimony at 10:26-11:4); Exh. RIVR00005 (Hopenfeld Expert Report at 15).

F. The Track Record of CHECWORKS Performance at Indian Point

93. The predictive capabilities of the CHECWORKS computer code can also be assessed in terms of the ability of the model to prevent wall thinning incidents. Exh. RIV000003 (Hopenfeld Testimony at 11:6-12:24); Exh. RIVR00005 (Hopenfeld Expert Report at 15-18).

94. Because CHECWORKS produces inaccurate results, and in light of the long industry-wide history in which CHECWORKS has been unsuccessful in predicting wall thinning, a demonstrated record of performance is necessary to show that CHECWORKS is able to manage FAC during the PEO at Indian Point. Exh. RIV000003 (Hopenfeld Testimony at 11:6-12:24); Exh. RIVR00005 (Hopenfeld Expert Report at 15-18).

95. CHECWORKS has had questionable effectiveness at nuclear power plants since the program was introduced. Exh. RIV000003 (Hopenfeld Testimony at 11:6-12:24); Exh.

RIVR00005 (Hopenfeld Expert Report at 15-18); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 40-41). NRCs Advisory Committee on Reactor Safeguards has stated that because of the poor correlation between CHECWORKS predictions and operating data, stating that [i]f you look at the data base, you dont really have too much confidence in CHECWORKS. Exh.

RIV000022 (2005 ACRS Transcript at 198). There is also evidence suggesting that the rate of 45

failures attributable to FAC actually went up in the period of time after CHECWORKS was put into use, demonstrating CHECWORKS was not effective in reducing the number of pipe failures. Exh. RIV000023 (NUREGCR-6936). Despite the use of CHECWORKS, FAC documented below minimum acceptable limits has been detected at numerous nuclear power plants across the United States, including Duane Arnold, Hope Creek, Clinton, Braidwood, LaSalle, Peach Bottom, Palo Verde, Palisades, Catawba, Calvert Cliffs, Kewaunee, Browns Ferry, ANO, and Salem. Exh. RIV000023 (NUREGCR-6936); Exh. RIVR00005 (Hopenfeld Expert Report at 16-17); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 40-41).

96. The NRC has acknowledged the seriousness and persistence of FAC throughout the nuclear industry. Exh. RIV000006 (NRC IN 86-106); Exh. RIV000007 (NRC Bulletin 87-01); Exh. RIVR00008 (NRC IN 1991-019); Exh. RIV000009 (NRC Monitoring Report 5 0042); Exh. RIV000011 (NRC IN 2006-008).

97. At Indian Point, the inability of the CHECWORKS model to produce accurate results establishes that the model has no track record of performance. Exh. RIV000003 (Hopenfeld Testimony at 12:1-24); Exh. RIVR00005 (Hopenfeld Expert Report at 17-18).

98. In addition, at Indian Point, a history of FAC related incidents confirms that, to date, the code has not been successful in preventing FAC. Exh. RIV000003 (Hopenfeld Testimony at 12:1-24); Exh. RIVR00005 (Hopenfeld Expert Report at 17-18).

99. Several Entergy documents report numerous leaks and/or the discovery of wall thinning below the minimum acceptable values in mechanical systems at Indian Point. Exh.

RIV000003 (Hopenfeld Testimony at 12:1-24); Exh. RIVR00005 (Hopenfeld Expert Report at 17-18); Exh. RIV000024 (Entergy Operating Experience Report); Exh. RIV000025 (Entergy Daily Event Report), RIV000026 (Entergy Condition Report List), RIV000027 (Entergy 46

Condition Report List), RIV000028 (Entergy Condition Report); RIV000029 (Entergy Condition Report); Exh. RIV000030 (ACRS Meeting Transcript September 2009); Tr. at 1694:7-21, 1696:4-10 (Hopenfeld).

100. Entergys documentation demonstrates a history of unacceptable FAC-related thinning events and/or leaks that have occurred at Indian Point notwithstanding the use of CHECWORKS at the plant. Exh. RIV000003 (Hopenfeld Testimony at 12:1-24); Exh.

RIVR00005 (Hopenfeld Expert Report at 17-18); Exh. RIV000024 (Entergy Operating Experience Report); Exh. RIV000025 (Entergy Daily Event Report), RIV000026 (Entergy Condition Report List), RIV000027 (Entergy Condition Report List), RIV000028 (Entergy Condition Report); RIV000029 (Entergy Condition Report); Exh. RIV000030 (ACRS Mtg Transcript Sept 2009).

101. Leaks pose tangible safety related concerns, and the leak-before-break concept is not an excuse for operating with excessively worn-out components. Exh. RIV000003 (Hopenfeld Testimony at 11:25-26); Exh. RIVR00005 (Hopenfeld Expert Report at 17); Tr. at 1333:1-14 (Hopenfeld). Yet, Entergys witness, Mr. Azevedo, testified that locations susceptible to wall thinning are identified when leaks occur. Tr. at 1439:23-1440:10 (Azevedo). Dr.

Hopenfelds testimony that component leaks are indicative of an ineffective aging management program is persuasive, since such instances should be prevented by an adequate program, and not viewed as acceptable. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 39-40); Tr. at 1333:5-7, 1693:15-1694:21, 1695:13-1696:10, 1699:1-24 (Hopenfeld).

102. Entergys history of unacceptable FAC-related thinning events and component leaks demonstrates that Entergy currently has no track record of performance of the code at the plant in preventing excessive FAC and/or leaks, and no track record of performance under post-47

power uprate conditions. Exh. RIV000003 (Hopenfeld Testimony at 12:1-24); Exh. RIVR00005 (Hopenfeld Expert Report at 17-18).

103. Because of the lack of a track record of performance of CHECWORKS in the nuclear industry or at Indian Point in particular, the CHECWORKS model cannot be considered an appropriate or useful tool for managing FAC at Indian Point during the PEO.

G. CHECWORKS and Compliance with the GALL Report 104. The Indian Point LRA was evaluated for compliance with the GALL Report, Revision 1. Tr. at 1891:16-19 (Hiser). A new revision, Revision 2, of the GALL Report was issued in December 2010 while the Indian Point license renewal proceeding was ongoing, and NRC Staff considered this new report in its assessment of the FAC program at Indian Point. Tr.

at 1892:1-8 (Hiser).

105. Entergy and NRC Staff both maintain that that the FAC program at Indian Point is consistent with Revisions 1 and 2 of the GALL Report. Tr. at 1348:17 (Cox); Tr. at 1892:12-17 (Hiser); Exh. ENT000029 (Entergys Testimony at A141).

106. While Revision 2 is the most recent finalized version of the GALL Report, in July 2012, NRC Staff issued a draft Interim Staff Guidance (ISG) document that suggests certain changes to Revision 2 of the GALL Report, that is, the current operative version of the GALL Report. Tr. at 1678:3-5, 17-19 (Hiser); Exh. ENT000573 (Draft LR-ISG-2012-01). At the time of NRC Staffs review of Entergys LRA and the adjudicatory hearing, NRC Staffs draft ISG was still a draft and therefore not final or official, and did not implicate any mandates for compliance. Tr. at 1829:21-24 (ALJ Wardwell); Tr. at 1684:6-10 (Yoder).

107. Entergys use of CHECWORKS at Indian Point does not comply with the GALL Report, Revision 1. The GALL Report, Revision 1 states that in order to have acceptable 48

Monitoring and Trending, CHECWORKS or a similar predictive code is used to predict component degradation in the systems conducive to FAC. Exhibit NYS00146C (GALL Report, Revision 1, at pp. XI M-61). The GALL Report focuses on the use of a quantitative predictive code for the purposes of monitoring and detecting FAC. See id.; Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 43:11-16); Tr. at 1598:2-5, 1600:8-12 (Hopenfeld). As discussed above, Entergy only relies upon a quantitative predictive code for a portion of the FAC program at Indian Point. See supra ¶¶ 47, 49, 50, 64. Moreover, CHECWORKS does not ensure that all forms of FAC will be adequately managed. See supra ¶ 60; Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 18:20-24). Entergys use of CHECWORKS is not consistent with the language in the GALL Report, which focuses on the use of computer modeling in order to adequately manage FAC.

108. The GALL Report, Revision 1 also states that [t]he inspection schedule developed by the licensee on the basis of the results of such a predictive code provides reasonable assurance that structural integrity will be maintained between inspections, and that inspections should ensure that the extent of wall thinning is adequately determined, that intended function will not be lost, and that corrective actions are adequately identified. Exh.

NYS00146C (GALL Report, Revision 1, at pp. XI M-61 to XI M-62). As discussed above, Entergys use of CHECWORKS at Indian Point to determine inspection priorities does not result in accurate inspection priorities, and thus, does not provide the requisite reasonable assurance or ensure that wall thinning will be adequately identified. See supra ¶¶ 72-103; Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 18:14-18). Whether or not intended functions of susceptible components have been lost to date at Indian Point is not relevant; rather, Entergy must, but has 49

not, assured that CHECWORKS will prevent the loss of intended functions of relevant components during the PEO. Tr. at 1593:2-1594:1 109. The GALL Report, Revision 1 guidance pertaining to acceptance criteria states that [i]nspection results are to inputs for a predictive computer code, such as CHECWORKS, to calculate the number of refueling or operating cycles remaining before the component reaches the minimum allowable wall thickness. Exhibit NYS00146C (GALL Report, Revision 1, at p.

XI M-62). As NRC Staffs witness, Mr. Yoder, testified, the entire program contained in the GALL Report must rely on your ability to inspect prior to reaching a critical thickness for a component. Tr. at 1674:3-7 (Yoder). However, as discussed above, at Indian Point, CHECWORKS does not provide accurate results and so does not result in the selection of appropriate inspection locations. See supra ¶¶ 72-92. CHECWORKS, thus, does not assure that components will be inspected prior to wall thicknesses dipping below allowable limits. Dr.

Hopenfeld specifically testified that the ultimate problem with CHECWORKS is that it does not adequately identify at-risk components which can fail before Entergy inspects them. 1741:1-9 (Hopenfeld; ALJ McDade); see also Tr. at 1449:4-14, 1665:24-1666:8, 1740:12-22 (Hopenfeld).

110. Entergys use of CHECWORKS at Indian Point also does not comply with the GALL Report, Revision 2. Like Revision 1, Revision 2 also emphasizes the use of computer modeling in order to adequately manage FAC, states that inspections conducted based on predictive code results provide assurance of component integrity, and states that inspections are used to ensure component wall thicknesses do not reach minimum allowable limits. Exhibit NYS00147D (GALL Report, Revision 2, at pp. XI M17-1, XI M17-2). For the same reasons just articulated, Entergys FAC program fails to be consistent with this guidance. See supra ¶¶107, 108, 109.

50

111. In addition, the GALL Report, Revision 2 contains revised language that ostensibly clarified the previous version of the report in relation to the appropriateness of relying on CHECWORKS. In particular, while the GALL Report, Revision 1 states that CHECWORKS is acceptable because in general it provides a bounding analysis for FAC, (Exhibit NYS00146C (GALL Report, Revision 1, at pp. XI M-61 to XI M-62), the GALL Report, Revision 2 appeared to define what this means, explaining that [t]he analysis is bounding because in general the predicted wear rates and component thicknesses are conservative when compared to actual field measurements. Exhibit NYS00147D (GALL Report, Revision 2, at pp.

XI M17-1). Revision 2 further indicated that [i]t is recognized that CHECWORKS is not always conservative in predicting component thickness and that therefore, when measurements show the predictions to be non-conservative, the model must be re-calibrated using the latest field data. Id. (GALL Report, Revision 2, at pp. XI M17-1 to XI M17-2). This language indicates that the use of CHECWORKS is acceptable if it provides conservative results, and if not, if it can be re-calibrated to do so. Exh. RIV000003 (Hopenfeld Testimony at 16:27-18:12);

Exh. RIVR00005 (Hopenfeld Expert Report at 18-19); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 41-43); Tr. at 1330:17-23, 1597:23-1598:1, 1661:20-1662:2, 11-15, 1668:18-1669:4, 1689:18-1690:16 (Hopenfeld). This is a valid and reasonable interpretation of the current language in the GALL Report, Revision 2. Tr. at 1330:17-23, 1689:18-1690:16 (Hopenfeld); Exh. RIV000003 (Hopenfeld Testimony at 16:27-18:12); Exh. RIVR00005 (Hopenfeld Expert Report at 18-19); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 41-43).

112. Entergys use of CHECWORKS is not consistent with the expanded guidance contained in the GALL Report, Revision 2. Entergy accepts the use of CHECWORKS as a best-estimate model, which produces highly non-conservative data upwards of 60% of the 51

time. Tr. at 1641:11-13 (Aleksick); Tr. at 1300:2-6 (Aleksick); see supra ¶¶ 75-81. The CHECWORKS model at Indian Point does not produce conservative results, and cannot be re-calibrated to provide conservative results. See supra ¶¶ 76, 77, 79, 85; Exh. RIV000003 (Hopenfeld Testimony at 17:21-18:12); Exh. RIVR00005 (Hopenfeld Expert Report at 19).

These circumstances indicate that Entergys use of CHECWORKS does not comply with the GALL Report, Revision 2. Exh. RIV000003 (Hopenfeld Testimony at 16:27-18:12); Exh.

RIVR00005 (Hopenfeld Expert Report at 18-19).

113. As the current legally operative version of the GALL Report, it is appropriate and necessary for Entergys FAC program to comply with Revision 2. See Exh. RIV000002 (Riverkeeper Initial SOP).

114. In relation to the management of FAC, Entergy has effectively committed to complying with the guidance contained in the GALL Report, Revision 2. See Exh. ENT000029 (Entergys Testimony at A141); Tr. at 1688:13-20 (Hopenfeld).

115. While Revision 2 is the most recent and effective version of the GALL Report and NRC Staffs ISG is not binding in any way (see Tr. at 1678:3-5, 17-19 (Hiser); Tr. at 1829:21-24 (ALJ Wardwell), Tr. at 1684:6-10 (Yoder)), Entergys FAC program at Indian Point also fails to be consistent with the NRC Staffs proposed changes.

116. First, the changes proposed in the ISG would not materially change the language in the GALL Report, Revision 2 that emphasizes the use of CHECWORKS in order to detect component degradation, states that inspections conducted based on predictive code results provide assurance of component integrity, or that states that inspections are used to ensure component wall thicknesses do not reach minimum allowable limits. Exh. ENT000573 (Draft LR-ISG-2012-01 at D-7, D-8). As a result, for the same reasons explained above, Entergys 52

FAC program would still fail to be consistent with the GALL Report if the changes proposed in the ISG were effective. See supra ¶¶107, 108, 109.

117. Second, while the ISG proposes to delete language in the GALL Report, Revision 2 stating that CHECWORKS is bounding and acceptable because in general the predicted wear rate and component thicknesses are conservative when compared to actual field measurements (Exh. ENT000573 (Draft LR-ISG-2012-01 at D-7, ¶ 5), the proposed changes would leave undisturbed the following language: It is recognized that CHECWORKS is not always conservative in predicting component thickness; therefore, when measurements show the predictions to be non-conservative, the model must be recalibrated using the latest field data.

Id. The proposed language continues to indicate that the use of CHECWORKS is acceptable only if non-conservative results can be corrected by re-calibrating the model. Exh. RIV000003 (Hopenfeld Testimony at 16:27-18:12); Exh. RIVR00005 (Hopenfeld Expert Report at 18-19).

As a result, Entergys FAC program would still fail to be consistent with the GALL Report if the changes proposed in the ISG were effective.

118. As NRC Staffs witness, Mr. Yoder, testified, it was not the intent of the proposed changes to disturb the notion that entire program contained in the GALL Report must rely on your ability to inspect prior to reaching a critical thickness for a component. Tr. at 1674:3-9, 1675:4-9 (Yoder). NRC Staffs other witness, Dr. Hiser, also explained that the proposed changes to the GALL Report was intended to clarify the goal of the existing wording, which is to ensure adequate predictions of component performance and to ensure that inspections corrective actions would be taken as necessary, before the acceptance criteria would be exceeded for those components. Tr. at 1676:21-1677:1 (Hiser). For the reasons already discussed above, Entergys use of CHECWORKS at Indian Point does not ensure the ability to predict which components 53

will reach critical wall thickness. See supra ¶¶ 88-92. Thus, based on the representation of NRC Staff, Entergys FAC program would still fail to be consistent with the GALL Report if the changes proposed in the ISG were effective.

H. Alternative Computer Modeling Suitable for Managing FAC 119. Riverkeepers witness, Dr. Hopenfeld, testified that the use of CHECWORKS at Indian Point is inappropriate. Tr. at 1532:15-1533:15, 1734:22-1735:3 (Hopenfeld). Dr.

Hopenfeld explained that the underlying fault assumptions of linear wear rate, a narrow definition of FAC, and chromium uncertainties, coupled with the indisputable evidence about the poor accuracy of CHECWORKS at Indian Point, renders it of no use for managing FAC at Indian Point. Tr. at 1532:15-1533:15, 1734:22-1735:3 (Hopenfeld).

120. Dr. Hopenfeld testified that a computer model should play a role in managing FAC. Tr. at 1539:17-1540:24 (Hopenfeld). This is consistent with the guidance contained in the GALL Report. See supra ¶ 107. In order to be adequate any computer model used to manage FAC at Indian Point must not suffer from the same deficiencies and underlying assumptions identified by Dr. Hopenfeld. See supra ¶¶ 59-62.

121. Dr. Hopenfeld explained that a conservative code is the right engineering approach. Tr. at 1728:8-14 (Hopenfeld). Dr. Hopenfeld explained that alternative computer models and codes are available and preferable in relation to managing FAC at the plant. Tr. at 1434-1435, 1446-1447 (Hopenfeld).

122. Dr. Hopenfeld explained that data from a different computer model, BRT-CICEROTM, has demonstrated very narrow uncertainty and resulted in much less data scatter in wear predictions. Tr. at 1534-1537, 1879-1881, 1741:9-1742:2 (Hopenfeld); Exh. RIV000110 (Trevin, Moutrille Study at Figure 7). BRT-CICEROTM is a bounding code, and other such 54

codes exist. Tr. at 1649:6-8 (Horowitz). Dr. Hopenfeld indicated that the BRT-CICEROTM model employs a better approach than CHECWORKS. Tr. at 1534-1537, 1548:13-19 (Hopenfeld).

123. Dr. Hopenfeld testified that the BRT-CICEROTM model showed an order of magnitude better accuracy than the CHECWORKS model. Tr. at 1810:3-10 (Hopenfeld; ALJ Wardwell); Tr. at 1810:25-1811:20, 1814:24-25, 1815:12 (Hopenfeld). As opposed to 900%

inaccuracy in CHECWORKS data, data from the BRT-CICEROTM model exhibited only 40%

inaccuracy, and a much more bunched scatter of data, and on one side of the middle ideal line.

Tr. at 1741:9-1742:2, 1811:8-20, 1814:24-25, 1815:12 (Hopenfeld).

124. Dr. Hopenfeld admitted that the scales on the data plot for the BRT-CICEROTM model are different, and that changing the scales could alter the visual impact of the scatter, but that doing so would not impact the value of the numbers, or Dr. Hopenfelds opinions about the model. Tr. at 1813:17-23, 1814:2-15, 1814:24-25, 1815:12 (Hopenfeld).

125. Dr. Hopenfeld refuted Entergys witnesses testimony that the BRT-CICEROTM model assumes linear corrosion rates (Tr. at 1882:8-9 (Horowitz)), by explaining that wear in the model is based on averages, and, as a result, it cannot be concluded from that whether its linear or non-linear. Tr. at 1881:19-1882:1 (Hopenfeld); Tr. at 1548:13-19 (Hopenfeld).

126. Dr. Hopenfeld also testified that the equation underlying the BRT-CICEROTM model is better because it measures chromium, and because it may consider all mechanisms that are properly encompassed by the term, FAC, i.e., more than just chemical dissolution processes.

Tr. at 1536:1-1537:9, 1549:1-21, 1734:15-21, 1742:1-7, 1815:19-25, 1816:7-8, 1817:19-23, 1818:12-15, 1819-1822 (Hopenfeld).

55

V. ENTERGYS OTHER TOOLS FOR DETERMINING THE SCOPE OF FAC INSPECTIONS 127. Entergy maintains that, in addition to predictive evaluations from CHECWORKS, the FAC program at Indian Point uses other tools in order to determine the scope of FAC inspections. For modeled components, Entergy purports to rely on trending of actual pipe wall thickness measurements from past outages, industry and Indian Point operating experience, results of other plant inspection programs, and engineering judgment, while, for non-modeled components, Entergy relies on a separate ranking and selection analysis. Exh. ENT000029 (Entergy Testimony at A72, A73); Tr. at 1509:9-12 (Aleksick); 1480:2325, 1481:1-3 (Mew).

128. The goal of using these tools is to ensure that components are inspected prior to reaching a critical wall thickness. Tr. at 1674:1-5 (Yoder); Exhibit NYS00146C (GALL Report, Revision 1, at pp. XI M-61 to XI M-62); Exhibit NYS00147D (GALL Report, Revision 2, at pp.

XI M17-1, XI M17-2).

A. Reinspections Based on Actual Wall Thickness Measurements 129. Entergy attributes approximately half, but upwards of two-thirds, of its inspection scope, to reinspections based upon previous inspection results. Exh. ENT000029 (Entergy Testimony at A77, Figure 1, Figure 2).

130. Entergys witnesses emphasize that determinations about reinspection locations are based on inspection results, and not on CHECWORKS or other factors. Tr. at 1606:8-16 (Aleksick); Tr. at1633:7-14 (Aleksick); Tr. at 1835:10-16 (Azevedo); Tr. at 1605:23-1606:7 (Aleksick); Tr. at 1836:2-17 (Mew, Azevedo); Tr. at 1610:10-13 (Aleksick).

131. However, Dr. Hopenfeld explains that actual pipe wall thickness measurements from past outages are only useful when used in combination with a predictive tool which would prevent the wall thickness of a given component from being reduced to below the minimum 56

design thickness while in service. Exh. RIV000003 (Hopenfeld Testimony at 13); Exh.

RIVR00005 (Hopenfeld Expert Report at 21); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 11-14).

132. Dr. Hopenfeld further demonstrated that reinspections based upon trending of actual measurements are not entirely independent of CHECWORKS. Exh. RIV000003 (Hopenfeld Testimony at 13); Exh. RIVR00005 (Hopenfeld Expert Report at 21); Exh.

RIV000108 (Hopenfeld Rebuttal Testimony at 11-14). For components initially and newly selected for inspection by CHECWORKS, any decisions regarding future inspection scope based on actual pipe wall thickness measurements and wear rate trending of the actual inspection results, necessarily depends upon use of the CHECWORKS computer model. Exh. RIV000003 (Hopenfeld Testimony at 13); Exh. RIVR00005 (Hopenfeld Expert Report at 21); Exh.

RIV000108 (Hopenfeld Rebuttal Testimony at 11-14).

133. In fact, Entergys witness, Mr. Aleksick explained that [r]oughly half of reinspected components were initially identified for inspection through CHECWORKS, and that in the Indian Point plant inspection history, that percentage would be larger. Tr. at 1609:9-25 (Aleksick; ALJ McDade); Tr. at 1610:2-5 (Aleksick). As a result, a large number of components are only eligible to be selected for reinspection because they were originally selected for an initial inspection by CHECWORKS. It is, thus, fair to conclude that reinspections stem in part from, and are informed by initial inspections initiated because of CHECWORKS predictions. See Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 11-12).

134. The fact that CHECWORKS is not a reliable predictor of FAC (see supra ¶¶ 72-

87) reduces the efficacy of reinspections, since CHECWORKS prevents components with critical wall thicknesses from being selected in the first instance. Tr. at 1449:4-14, 1665:24-57

1666:8, 1669:5-1670:4, 1740:12-22, 1741:1-9 (Hopenfeld). This is highlighted by Mr.

Azevedos acknowledgement that Entergy only has a sense of where future inspections should occur, [f]or the locations that we [Entergy] have inspected. Tr. at 1360:15-22 (Azevedo).

135. Entergys position that reinspection determinations do not involve CHECWORKS in any way also runs afoul of the acceptance criteria contained in the GALL Report, which indicates that inspection results should be inputs to a predictive code in order to calculate the number of refueling or operating cycles remaining before the component reaches the minimum allowable wall thickness, in order to determine the need for some form of corrective action. See Exhibit NYS00146C (GALL Report, Revision 1, at XI M-62); Exhibit NYS00147D (GALL Report, Revision 2, at XI M17-2); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 12).

136. In addition, reinspections assume that FAC wear progresses at a linear rate. Tr. at 1611:4-14 (Aleksick); See Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 13:25-29.

However, the evidence in the record indicates that FAC is a non-linear, local phenomenon. See supra ¶¶ 26-37. Thus, Entergys inspection selection scope on the basis of reinspections is flawed and does not assure that inspections will occur, or that corrective action will be taken, before components reach critical wall thicknesses. See Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 13).

137. Dr. Hopenfeld further testifies that, for reinspections to be considered a valid independent tool, i.e. one used independent of a quantitative predictive model, reinspections and data trending would require a very large number of inspection points, and the considerations of numerous factors. See Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 13-14). Entergy has failed to show that its use of reinspections meets the rigorous standards necessary to assure the 58

adequate detection of wall thinning at Indian Point. See Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 13-14).

B. Industry and Plant Operating Experience 138. Entergy chooses upwards of a third of its FAC inspections based upon operating experience. Exh. ENT000029 (Entergy Testimony at Figure 1 and Figure 2).

139. Dr. Hopenfeld explained that industry and plant experience with pipe wall thinning are types of information that feed into the CHECWORKS model, and, are thus, not entirely independent tools for identifying inspection scope during outages. Exh. RIV000003 (Hopenfeld Testimony at 13); Exh. RIVR00005 (Hopenfeld Expert Report at 21); Exh.

RIV000108 (Hopenfeld Rebuttal Testimony at 14-15). The usefulness of such information for determining future inspections rests, at least in part, on how the CHECWORKS model processes the inputs and how such information affects the model over time. Exh. RIV000003 (Hopenfeld Testimony at 13); Exh. RIVR00005 (Hopenfeld Expert Report at 21); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 14-15).

140. Dr. Hopenfeld further explained that to the extent Entergy relies on industry and plant experience as an independent tool, Entergy has failed to provide an adequate description of its meaning or to provide sufficient details to assure that this tool establishes inspection locations so as to prevent undetected FAC. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 14-15).

C. Other Plant Inspection Programs 141. Though Entergy indicates that a tool for selecting FAC inspection locations is other plant inspection programs (Exh. ENT000029 (Entergy Testimony at A72)), the record 59

does not establish what percentage of inspections are attributable to this tool. See id. (Entergy Testimony at A77, Figure 1, Figure 2).

142. Entergy has not provided enough information in order to assess and evaluate the effectiveness of Entergys reliance other plant inspection programs. Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 15:5-17). Thus, the record does not establish that Entergys use of other plant inspection programs is used to establish inspection locations in a manner that prevents undetected FAC. Id.

D. Engineering Judgment 143. Entergy purports to rely on engineering judgment for a small fraction of determining inspection locations. Exh. ENT000029 (Entergy Testimony at Figure 1 and Figure 2).

144. However, as Dr. Hopenfeld testifies, to the extent Entergys other tools for selecting inspection locations (that is, actual pipe wall thickness, plant and industry experience, and other plant inspection programs) do not rely upon CHECWORKS in order to meaningfully contribute to inspection scope selection, they can only be properly categorized as inputs which assist in the formulation of an engineering judgment. Exh. RIV000003 (Hopenfeld Testimony at 14); Exh. RIVR00005 (Hopenfeld Expert Report at 21-22). As the only tool that can properly be categorized as independent for purposes of selecting inspection locations, engineering judgment plays a larger role than Entergys witnesses portray. Exh. RIV000003 (Hopenfeld Testimony at 14); Exh. RIVR00005 (Hopenfeld Expert Report at 21-22).

145. EPRI has explained, engineering judgment cannot substitute for other factors.

Exh. RIV000012 (NSAC-202L at 2-4). Dr. Hopenfeld explains how it is commonly recognized in all major industrial plants that engineering judgment alone is not sufficiently reliable to 60

manage FAC. Exh. RIV000003 (Hopenfeld Testimony at 14); Exh. RIVR00005 (Hopenfeld Expert Report at 22); Tr. at 1318:4-12 (Hopenfeld). The development of the CHECWORKS computer model itself arose out of the realization by the nuclear industry that engineering judgment was not enough to be able to detect unacceptable and unsafe component wall thinning.

Exh. RIV000003 (Hopenfeld Testimony at 14); Exh. RIVR00005 (Hopenfeld Expert Report at 22); Tr. at 1318:4-12 (Hopenfeld).

146. Dr. Hopenfeld observes, and Entergys witnesses do not dispute, that engineering judgment is intrinsically subjective. Exh. RIV000003 (Hopenfeld Testimony at 14); Exh.

RIVR00005 (Hopenfeld Expert Report at 22); Exh. ENT000029 (Entergy Testimony at A75).

147. Dr. Hopenfeld explains that when engineering judgment is identified as a predictive tool, a very high degree of knowledge is required by those who conduct the assessment and specify the required steps for the prevention of component failures. Exh.

RIV000003 (Hopenfeld Testimony at 14); Exh. RIVR00005 (Hopenfeld Expert Report at 22).

Even with the same input data, different assumptions could lead to different results because each assessment would depend heavily on the individual skill and experience of the responsible engineer. Exh. RIV000003 (Hopenfeld Testimony at 14); Exh. RIVR00005 (Hopenfeld Expert Report at 22).

148. Dr. Hopenfeld testifies that in order to assess the validity of the use of engineering judgment, it is imperative to fully understand how it is used and all relevant underlying assumptions informing any judgment related determinations. Exh. RIV000003 (Hopenfeld Testimony at 14); Exh. RIVR00005 (Hopenfeld Expert Report at 22).

149. Dr. Hopenfeld testified that numerous Entergy documents pertaining to Entergys FAC program did not reveal what engineering judgment even means in relation to FAC 61

inspections at Indian Point, and what role it actually plays in inspection scope selection. Exh.

RIV000003 (Hopenfeld Testimony at 14); Exh. RIVR00005 (Hopenfeld Expert Report at 22).

Dr. Hopenfeld has observed that Entergy has not identified any kind of systematic methodology which demonstrates that engineering judgment is a separate predictive tool that would adequately meet applicable regulatory guidelines, including those contained in the GALL Report, and which would manage FAC related component degradation throughout the proposed PEO. Exh.

RIV000003 (Hopenfeld Testimony at 14); Exh. RIVR00005 (Hopenfeld Expert Report at 22);

Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 15-16); Tr. at 1318:13-24 (Hopenfeld).

150. Entergy relies upon generic guidance contained in an EPRI report, NSAC-202L, to guide the use of engineering judgment for purposes of selecting components for FAC inspections. Tr. at 1503:24-25, 1504:1-2 (Cox); Tr. at 1850:9-1851:3 (Azevedo).

151. Entergy also uses a fleet-wide FAC program implementation document, EN-DC-315, for guidance on employing engineering judgment for selecting components for FAC inspections. Tr. at 1504:3-7 (Cox).

152. Though Entergys witnesses take the position that the guidance contained in NSAC-202L and EN-DC-315 is pretty specific, these documents contain no discernible criteria relating to how to properly exercise engineering judgment. Tr. at 1850:9-1851:3 (Azevedo);

Exh. RIV000012 (NSAC-202L at § 2.5); Exh. ENT000038 (EN-DC-315). The guidance relating to engineering judgment in NSAC-202L, consisting of about half a page of text, pertains largely to what training a FAC engineer should attain, while there is no section in EN-DC-315 relating to engineering judgment. Exh. RIV000012 (NSAC-202L at § 2.5); Exh. ENT000038 (EN-DC-315). Entergy could not point to any additional documentation containing guidance on how to employ engineering judgment at Indian Point, making it difficult to study. Tr. at 1503:24-25, 62

1504:1-7 (Cox); Tr. at 1505:1-19 (Mew); Tr. at 1850:9-1851:3 (Azevedo); Tr. at 1853:22-23 (Azevedo). Entergys witness, Mr. Azevedo testified that the same generic criteria contained in the NSAC-202L and EN-DC-315 are used any time engineering judgment is employed at Indian Point in the FAC program. Tr. at 1853:22-23 (Azevedo).

153. Dr. Hopenfeld explains that there are several key elements necessary to form a sound engineering judgment as it relates to FAC at Indian Point, (a) good documentation of historical FAC assessments; (b) good communication between the organization that conducts analytical assessments and plant operators; (c) knowledge of FAC assessment methods; and (d) knowledge of risks and consequences. Exh. RIV000003 (Hopenfeld Testimony at 14-15); Exh.

RIVR00005 (Hopenfeld Expert Report at 22-23).

154. Dr. Hopenfeld provided a detailed description and assessment of these four elements. Exh. RIV000003 (Hopenfeld Testimony at 14-15); Exh. RIVR00005 (Hopenfeld Expert Report at 22-23). Based on this review, Dr. Hopenfeld properly concluded that Entergy does not appear to espouse any of the key elements that are necessary to form a sound engineering judgment. Exh. RIV000003 (Hopenfeld Testimony at 14-15); Exh. RIVR00005 (Hopenfeld Expert Report at 22-23); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 15-16).

155. The evidence in the record does not establish that engineering judgment is employed in such a way so as to ensure the inspection of components prior to wall thinning dipping below minimum levels. Exh. RIV000003 (Hopenfeld Testimony at 14-15); Exh.

RIVR00005 (Hopenfeld Expert Report at 22-23); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 15-16).

63

E. Selection of SNM Components 156. Entergy attributes upwards of 27% of its inspection scope to inspections of susceptible non-modeled, or SNM, components. Exh. ENT000029 (Entergy Testimony at Figure 1 and Figure 2); see supra ¶¶ 47, 51, 52.

157. Selecting SNM components for inspections does not involve a quantitative, predictive computer model, but a manual ranking and assessment process. Exh. ENT000029 (Entergy Testimony at A72); Tr. at 1509:9-12 (Aleksick).

158. SNM components represent 78% of all susceptible lines at Indian Point Unit 2 and 80% of all susceptible lines at Indian Point Unit 3. See Exh. ENT000029 (Entergy Testimony at A76). About half of these components, which encompass well over the majority of FAC-susceptible components at Indian Point, have never been inspected. See Tr. at 1865:12-18 (Aleksick); Tr. at 1866:19-1867:2 (Aleksick); Tr. 1867:11-1867:16, 1867:24-1868:9 (Mew).

159. Entergys process for selecting SNM components for inspections involves the same tools used for selecting modeled components, except for the use of CHECWORKS. That is, Entergys selection process considers operating experience, engineering judgment, and trending of past measurements. Exh. ENT000029 (Entergys Testimony at A69). For the reasons already discussed above, Entergy has not established that relying on these tools is adequate to ensure that wall thicknesses will be detected before falling below acceptable limits.

See supra ¶¶ 131, 135-137, 140, 144-155.

160. Entergys inspection selection process for SNM components, like that for modeled components, involves selecting a representative, i.e. limited, set of components for inspection based upon manual rankings. Tr. at 1589:10-17 (Aleksick). Because Entergys 64

selection methodology is flawed, it is not clear that components requiring corrective action will be chosen to be inspected. See supra ¶ 157, 159.

161. A record of excessive wall thinning and component leaks at Indian Point indicates that Entergy has no track record of performance in managing FAC in SNM components. See supra ¶¶99-102. Entergys witness, Mr. Azevedo, acknowledged that Entergy has experienced leaks in SNM components. Tr. at 1590:12-14, 24, 1591:3 (Azevedo). Such leaks are significant, even if they did not involve a loss of intended function, and indicate that Entergys methodology for selecting SNM components is not adequate. Tr. at 1593:2-1594:1 (Hopenfeld).

F. Conclusions Regarding Entergys Other Tools for Inspection Selection Scope 162. The other tools identified by Entergy for selecting components for FAC inspections are not adequate to ensure that all necessary components will be inspected so as to effectively manage FAC at Indian Point during the proposed PEO. These additional criteria, taken together with the use of CHECWORKS, are not sufficient to establish an accurate FAC inspection scope so as to assure Entergys ability to inspect susceptible components prior to reaching a critical thickness. Exh. RIV000003 (Hopenfeld Testimony at 13-15); Exh.

RIVR00005 (Hopenfeld Expert Report at 21-23); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 15-16).

163. This is for the same reasons the deficiencies in CHECWORKS implicate undetected wall thinning, namely, that the necessary components will not be selected for inspection. See generally supra ¶¶ 88-92.

164. A history of excessive wall thinning and component leaks indicates that Entergy has no track record of performance in managing FAC, which indicates that Entergys other 65

tools aside from CHECWORKS are not sufficient for managing FAC at Indian Point. See supra

¶¶99-102.

VI. THE SAFETY CONSEQUENCES OF IMPROPERLY MANAGED FAC AT INDIAN POINT 165. Entergys inability to adequately determine which components require inspections to avoid reaching critical wall thicknesses poses significant safety concerns. Exh.

RIV000003 (Hopenfeld Testimony at 18-20); Exh. RIVR00005 (Hopenfeld Expert Report at 23-25); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 43-46). As Dr. Hopenfeld testified, the concern revolves around the fact that Entergy runs Indian Point with unknowable condition[s].

Tr. 1335:18-23, 1694:22-1695:7 (Hopenfeld). Dr. Hopenfeld explained that a critical aspect of an adequate AMP for FAC is an assessment of the safety implications of not detecting wall thinning below minimum acceptable levels required by the ASME. Tr. at 1498:2-12, 22-25, 1499:1-16 (Hopenfeld); Exh. RIV000003 (Hopenfeld Testimony at 4:13-16). As Dr. Hopenfeld testified, waiting until thickness dips below allowable levels to determine whether it can handle local transient loads is an unsafe way to run a nuclear plant. Tr. at 1335:18-23 (Hopenfeld).

166. Dr. Hopenfeld explained his concern that undetected FAC degradation may result in significant safety consequences if Indian Point is subject to sudden transient loads, including operational transients, design basis accident (DBA) transients, earthquake loads, station blackouts (SBOs), and transients without scrams (ATWS). Exh. RIV000003 (Hopenfeld Testimony at 18-20); Exh. RIVR00005 (Hopenfeld Expert Report at 23-24); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 28:20-23, 43-46).

167. This is particularly of concern in relation to steam generators, which Entergy does not model in CHECWORKS, or otherwise include as part of its FAC program, despite the fact that steam generator components are highly susceptible to FAC. See supra ¶¶ 15, 54, 71. In 66

particular, with severely degraded walls, the feed water distribution piping ring inside the steam generators, which is subjected to high local velocities and turbulence, may rupture under transient loads causing damage to other structures within the steam generators. Exh. RIV000003 (Hopenfeld Testimony at 19); Exh. RIVR00005 (Hopenfeld Expert Report at 24).

168. Dr. Hopenfeld explains that undetected FAC during the proposed PEO also poses a risk of loss of coolant accidents (LOCAs). Exh. RIV000003 (Hopenfeld Testimony at 18-20); Exh. RIVR00005 (Hopenfeld Expert Report at 23-24); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 43-46). As Dr. Hopenfeld explains, when the original Indian Point probabilistic risk assessments were developed, it was assumed that pipes were in pristine conditions, since the effects of aging were not included. Exh. RIV000003 (Hopenfeld Testimony at 19); Exh. RIVR00005 (Hopenfeld Expert Report at 24). As a result, when the walls have been degraded, the probability of a pipe failing under a given load will be affected. Exh. RIV000003 (Hopenfeld Testimony at 19); Exh. RIVR00005 (Hopenfeld Expert Report at 24).

169. Dr. Hopenfeld introduced and discussed persuasive evidence that Indian Point is vulnerable to accidents from earthquake loads. Exh. RIV000003 (Hopenfeld Testimony at 19-20); Exh. RIVR00005 (Hopenfeld Expert Report at 25); Exh. RIV000031 (Sykes Study); Exh.

RIV000032 (NRC GI-199); Exh. RIV000033 (Dedman Article). The evidence in the record indicates that Indian Point is susceptible to an earthquake of up to 7.0 magnitude. Exh.

RIV000031 (Sykes Study). An NRC report from August 2010 (in conjunction with supplemental data regarding power plants not reviewed in the report) indicates that Indian Point Unit 3 has the highest risk of seismic related core damage than any other nuclear power plant in the country. Exh. RIV000032 (NRC GI-199); Exh. RIV000033 (Dedman Article). The evidence in the record indicates that Entergy has failed to consider how the uncertainty related to 67

pipe wall thickness at Indian Point will affect the integrity of components under earthquake loads in light of the recent information about such risks. Exh. RIV000003 (Hopenfeld Testimony at 18-20); Exh. RIVR00005 (Hopenfeld Expert Report at 23-25); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 43-46).

170. Dr. Hopenfeld explains that a lack of adequate knowledge about component wall thickness also poses a risk of a station blackout accident scenario. Exh. RIV000003 (Hopenfeld Testimony at 18-20); Exh. RIVR00005 (Hopenfeld Expert Report at 23-25); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 43-46). The evidence in the record indicates that Entergy has failed to consider how the uncertainty related to pipe wall thickness at Indian Point will affect the integrity of components under station blackout transients. Exh. RIV000003 (Hopenfeld Testimony at 18-20); Exh. RIVR00005 (Hopenfeld Expert Report at 23-25); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 43-46).

171. An aging phenomenon known as metal fatigue may also affect components degraded by FAC. See supra ¶ 41; Exh. RIV000003 (Hopenfeld Testimony at 20); Exh.

RIVR00005 (Hopenfeld Expert Report at 25); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 45-46); Tr. at 1523:25-1526:1. This combined, synergistic effect poses risks for severe accidents with significant safety consequences. Exh. RIV000003 (Hopenfeld Testimony at 20);

Exh. RIVR00005 (Hopenfeld Expert Report at 25); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 45-46). The evidence in the record indicates that Entergy has not considered how the operation of Indian Point with unknown pipe wall thicknesses will affect the likelihood of components succumbing to the effects of metal fatigue. Exh. RIV000003 (Hopenfeld Testimony at 20); Exh. RIVR00005 (Hopenfeld Expert Report at 25); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 45-46).

68

172. Without an adequate knowledge of the degree to which the strength of various components have been degraded due to FAC-related wear, there is uncertainty as to whether Indian Point would continue to operate without a severe accident occurring. Exh. RIV000003 (Hopenfeld Testimony at 18-20); Exh. RIVR00005 (Hopenfeld Expert Report at 23-25); Exh.

RIV000108 (Hopenfeld Rebuttal Testimony at 43-46). The record contains inadequate justification that Indian Point can operate safely in spite of the very large uncertainties in CHECWORKS predictions, and the lack of any other adequate tools to detect FAC during the proposed PEO. Exh. RIV000003 (Hopenfeld Testimony at 18-20); Exh. RIVR00005 (Hopenfeld Expert Report at 23-25); Exh. RIV000108 (Hopenfeld Rebuttal Testimony at 43-46). As Dr.

Hopenfeld explains, it is not a sound practice to wait until component wall thinning results occur before assessing and determining whether the component can handle local transient loads. Tr. at 1695:13-1696:10 (Hopenfeld).

VII. THE SUFFICIENCY OF ENTERGYS FAC AGING MANAGEMENT PROGRAM A. FAC Program Documentation 173. Entergys AMP for FAC is described in Appendices A and B of Entergys LRA.

Exhibit ENT00015A-B (Indian Point Energy Center LRA (April 2007) at §§ A.2.1.14, p.A-24, B.1.15, p.B-54; Tr. at 1342:1-9 (Cox). The LRA description of the FAC AMP is a short, page and a half narrative that parrots what the GALL Report says as a FAC program description. Tr.

at 1342:19, 1383:3-12, 1405:25-1406:1-3 (Cox); Tr. at 1415:16-19 (Hiser); Tr. at 1721:20-25 (Hiser).

174. Entergys witnesses claim that additional details of the program are incorporated into the LRA by reference to the GALL Report. Tr. at 1342:19-21, 1343:18-23, 1344:7-10 (Cox).

69

175. The FAC program contained in the GALL Report is not a site-specific program and so does not contain any criteria specifically relating to implementation of an effective FAC program at Indian Point. Tr. at 1363:25-1364:5 (Cox); Tr. at 1370-1371 (ALJ McDade).

176. Entergy considers the FAC program contained in the GALL Report an actual program, as opposed to a program description. Tr. at 1368:2-1369:5 (Cox). Entergy believes that it is sufficient to refer to the program contained in the GALL Report to demonstrate an effective FAC program. Tr. at 1369:24-1370:3 (Cox). Entergys witness, Mr. Cox, testified that a reference to, or copy of, the FAC AMP contained in the GALL Report constitutes the level of information needed to establish an effective and adequate AMP for FAC in the LRA. Tr. at 1381:22-1382:22, 1404:14-18 (Cox).

177. NRC Staffs witness, Dr. Hiser, took the position that the AMP contained in the GALL Report is not an implementable program but rather an Aging Management Program description, which other documents implement. Tr. at 1392:9-13 (Hiser).

178. NRC requires a demonstration of consistency and not just a statement of consistency with a GALL Report AMP. Tr. at 1388:6-13 (Hiser).

179. Entergys witnesses assert that the FAC program at Indian Point is consistent with the guidance contained in the GALL Report. Tr. at 1344:15-20 (Cox).

180. Entergys witnesses claim of consistency is based on the fact that Entergy relies on the industry guidance document, NSAC-202L, which is referenced and incorporated into the GALL Report. Tr. at 1342:1-9, 1346:11-15 (Cox). Tr. at 1342:1-9, 1346:11-1347:4, 1347:18-22, 1355:16-19, 1355:20-1356:22 (Cox). Entergys witness indicated that NSAC-202L defines enough of the essence of the program for purposes of determining the adequacy of a FAC program and whether it is consistent with regulatory guidance. Tr. at 1372:17-24 (Cox); Tr. at 70

1356:17-20 (Cox); Tr. at 1357:9-11. NSAC-202L is not a site-specific document and so does not contain any criteria specifically relating to implementation of an effective FAC program at Indian Point. Tr. at 1363:18-23 (Cox). Entergys witness, Mr. Cox, explained that Entergys reference to NSAC-202L in the LRA establishes that Entergy has a sufficiently detailed FAC AMP. Tr. at 1380:6-24, 1383:12-16 (Cox).

181. Entergys witnesses indicate that the industry guidance document, NSAC-202L, referenced in the GALL Report, is implemented via a variety of fleet-wide implementing procedures including, EN-DC-315. Tr. at 1349:13-17 (Cox); Tr. at 1352:10-12 (Azevedo); Tr. at 1353:14-16 (Cox); Tr. at 1358:2-13 (Azevedo); Tr. at 1372:8-16 (Cox); 1385:2-5 (Mew). These implementing procedures contain just a description of how you are going to implement it, not an actual implementation of it yet . . . . It is not necessarily the results and the actual applications procedure, but its the procedure itself that tells you how to implement the program. Tr. at 1376:24-1377:11 (Cox; ALJ Wardwell).

182. Entergys witness, Mr. Cox indicated that he would not be as comfortable claiming consistency with the GALL Report in the absence of EN-DC-315. Tr. at 1376:20-22 (Cox). While applicable to Indian Point, EN-DC-315 is a fleet-wide procedure, and thus, does not contain procedures specific to implementation at Indian Point, versus at other Entergy nuclear plants. Tr. at 1377:15-1378:5 (Azevedo); 1364:6-9 (ALJ Wardwell).

183. Implementing procedure details, operating experience results, and any site-specific information about the FAC program at Indian Point were not included as part of Entergys license renewal application. Tr. at 1366:6-13 (Cox). Instead, Entergy considers a commitment to follow generic guidance sufficient to establish an effective FAC AMP. Tr. at 1365:18-1366:13 (Cox). Entergys witness, Mr. Cox, explained that actually finding the Entergy 71

FAC AMP is not something that can be accomplished without coming onsite and auditing implementing procedures. Tr. at 1405:18-23 (Cox).

184. Entergys FAC Program documentation is generic in nature, containing no specificity as to applicability at Indian Point, and is virtually identical to FAC programs at other nuclear plants. Tr. at 1407:18-23 (Cox); Tr. at 1720:14-1721:25 (Hiser).

185. Entergys witness, Mr. Cox, indicated that it is feasible for Entergy to incorporate additional details regarding Entergys FAC AMP into the LRA. Tr. at 1403:21-1404:8 (Cox).

NRC Staffs witness likewise testified that it clearly would be possible to provide more information in the [license renewal] application. Tr. at 1722:24-1724:13 (Hiser).

186. Changes to Entergys implementing procedures can be accomplished via the procedures outlined in 10 C.F.R. § 50.59. Tr. at 1349:23-1350:4 (Cox); 1364:6-1365:14 (ALJ Wardwell); Tr. at 1489:16-1490:17 (Cox); Tr. at 1491:1-4 (Hiser). This process is not specific to the FAC program at Indian Point, and there are no specific criteria relating to FAC that would inform determinations regarding program changes. Tr. at 1854:2-13; 1857:11-1859:10 (Hiser, Azevedo). The 50.59 amendment process is a prescriptive process with a lot of leeway, and which implicates a lot of judgment. Tr. at 1492:21-4 (Hopenfeld).

187. There are no license commitments or conditions associated with the use of Entergys FAC AMP during the PEO. Tr. at 1823:23-1824:14 (Hiser) 188. NRC Staffs witness, Dr. Hiser, indicated that one way to determine if Entergys FAC program is working as intended is to see if the plant suddenly finds that it is doing inspections and finding that it needs to do replacements, because they are no longer meeting their inspection criteria on wall thickness; according to NRC Staff, that would be an indication that 72

maybe the program performance is not what it should be or that a problem [is] occurring. Tr.

at 1391:5-13 (Hiser).

B. Insufficiency of Detail of FAC AMP 189. Entergys reliance on CHECWORKS and other tools for determining locations to inspect for FAC is flawed, and results in inspection scopes that will not ensure the inspection of all components suffering from unacceptable wall thinning. See supra §§ IV, V.

190. In addition, fundamentally incorrect assumptions about the linearity and definition of FAC, and the categorical exclusion of steam generators from the FAC management program also prevent the FAC AMP from effectively selecting inspection locations, and ultimately detecting and managing FAC at Indian Point. See supra § II.

191. Thus, as a threshold matter, Entergys AMP for FAC does not assure that appropriate inspections will be undertaken, which renders Entergys other program criteria, such as those governing inspection methodology, frequency, and corrective actions, inadequate as well, for assuring component integrity and the timely detection of unacceptable FAC. See supra

§§ IV, V.

192. Entergys generic program documents, that is, the GALL Report and NSAC-202L, provide general guidance on what constitutes an adequate FAC AMP. See supra ¶¶ 173-188.

These documents establish that a critical aspect of an effective FAC AMP is ensuring that representative inspections are performed on the most critical components. Exh. NYS00146C (GALL Report, Revision 1, at § XI.M17); Exh. NYS00147D (GALL Report, Revision 2, at § XI.M17); Exh. RIV000012 (NSAC-202L). The GALL Report, Revisions 1 and 2, state that

[t]he extent and schedule of the inspections assure detection of wall thinning before the loss of intended function and that the inspection schedule developed by the licensee . . . provides 73

reasonable assurance that structural integrity will be maintained between inspections. Exh.

NYS00146C (GALL Report, Revision 1, at XI M-61 to XI M-62, ¶¶ 4, 5); Exh. NYS00147D (GALL Report, Revision 2, at XI M17-1 to XI M17-2, ¶¶ 4, 5). NSAC-202L states that [a]

primary objective of FAC analysis is to identify the most susceptible components, and that sample locations should be chosen to select the components with the greatest susceptibility to FAC. Exh. RIV000012 (NSAC-202L at § 2.2). In addition, to the extent CHECWORKS is used in a FAC program pursuant to the GALL Report and NSAC-202L, it must be based on the proper use of the model, that is, one that can be re-calibrated, and which can provide accurate results. See supra § IV.G.

193. Riverkeepers witness, Dr. Hopenfeld, provided testimony which confirms the importance of selecting appropriate components in order to have an effective FAC AMP. In particular, Dr. Hopenfeld testified that the first step of an adequate FAC AMP is to make sure you inspect the component before you reach critical thickness, a critical part of which is identify[ing] the right components . . . to inspect out of the thousands of components to choose from. Tr. at 1449:4-14,1493:5-24; 1497:10-1498:1, 1500:7-10 (Hopenfeld). While the GALL Report has a procedures for doing so, Entergy does not currently meet such guidance, and, accordingly, must provide adequate details of how it will meet such guidance. Tr. at 1497:10-1498:1, 1500:7-15 (Hopenfeld); see supra § IV, V.

194. Because Entergys current FAC program does not comply with this threshold tenet governing what constitutes an adequate AMP, Entergy must provide sufficient details regarding inspection scope, frequency, component replacement and repair criteria, etc., to demonstrate that FAC will be appropriately managed at Indian Point during the PEO. Exh.

RIVR00005 (Hopenfeld Expert Report at 25). As Dr. Hopenfeld explains, because Entergys 74

FAC program does not assure an adequate inspection scope, and as a result, the adequate management of FAC, Entergy must address all the elements identified in the SRP-LR, and the GALL Report, including the method for determining component inspections, frequency of such inspections, and attendant criteria for component repair and replacement. Exh. NYS000195 (NUREG-1800, Rev. 1, at A.1-3 to A.1-7); Exh. NYS000161 (NUREG-1800, Rev. 2, at A.1-3 to A.1-7); Exh. NYS00146A-NYS00146C (GALL Report, Revision 1); Exh. NYS00147A-NYS00147D (GALL Report, Revision 2). Exh. RIVR00005 (Hopenfeld Expert Report at 25).

Entergy must describe an AMP that ensures that the aging effects of FAC will be adequately managed throughout the PEO. Exh. RIVR00005 (Hopenfeld Expert Report at 25).

195. An articulation of sufficient program details is necessary because the controlling guidance documents, the GALL Report and NSAC-202L, lack specificity and, standing alone, do not guarantee that there is a reasonable assurance that FAC will be adequately detected and maintained during the proposed PEO. Tr. at 1496:5-25 (Hopenfeld); see also supra ¶¶ 173-188.

196. Currently, Entergy does not articulate details in its AMP with respect to the frequency of FAC inspections. See Tr. at 1362:5-10, 16-25 (Aleksick); Tr. at 1400:1-5 (Cox).

Entergys operative AMP documents also provide general guidance in relation to conducting inspections and implementing corrective actions, such as component repairs, or replacements.

See Exh. RIV000012 (NSAC-202L at § 4.8). For example, the guidance Entergy employs to determine grid inspection sizes is subjective and affected by cost considerations. Tr. at 1433:25-1435:1 (Hopenfeld). It is also unclear whether the procedures Entergy follows to conduct ultrasonic testing result in accurate measurements. Tr. at 1760- 1762 (Hopenfeld); Tr. at 1758:23-1759:4 (Azevedo). In addition, criteria governing repair and replacement decisions are fundamentally flawed in light of the inadequate scope of inspections in the first instance. See Tr.

75

at 1448:22-1450:4 (Hopenfeld). Additional details are necessary in relation to FAC inspection scope, inspection methodology, corrective actions, and other GALL Report program elements, which address the various deficiencies in Entergys current FAC AMP discussed above. See supra ¶¶ 189, 190.

VIII. FACTUAL CONCLUSIONS TO BE DRAWN FROM THE EVIDENCE The following factual conclusions can be drawn from the above-discussed findings of fact:

197. FAC is a pipe wall thinning phenomenon that encompasses chemical dissolution processes in addition to mechanical, erosive, processes. See supra ¶¶ 14, 16-25.

198. FAC affects a range of nuclear power plants components, including those located in the steam generator. See supra ¶ 15. However, Entergy does not consider the effect of FAC on risk-significant, FAC-susceptible components in the steam generators. See supra ¶ 54.

199. Entergys AMP for FAC improperly assumes that FAC is limited solely to chemical dissolution processes, and ignores the fact that additional mechanical processes are at work as well, limiting the effectiveness of the program for adequately managing FAC. See supra

¶ 17, 60.

200. FAC is a local phenomenon that progresses in a non-linear manner. See supra ¶¶ 26-37.

201. Entergys AMP for FAC improperly assumes that FAC is linear with time and a non-local phenomenon, which severely limits the effectiveness of the program for adequately managing FAC. See supra ¶ 38.

202. FAC poses a significant safety risk at nuclear power plants if left undetected. See supra ¶¶ 39-41.

76

203. The CHECWORKS computer code models a subset of susceptible components at Indian Point. Less than half of the 8,000 components modeled in CHECWORKS have been inspected to date over the life of Indian Point. See supra ¶¶ 47, 50.

204. Susceptible non-modeled (SNM) components comprise the large majority of FAC susceptible components at Indian Point, and are not modeled in CHECWORKS; less than half the components in the 700 SNM analysis lines have been inspected to date over the life of Indian Point. See supra ¶¶ 47, 51-52.

205. The CHECWORKS code is a statistical predictive model used to plan inspections of nuclear power plant components, in order to prevent FAC-related failures. See supra ¶¶ 55-

56. The code requires benchmarking in order to account for changes in plant parameters. See supra ¶¶ 57-58.

206. The CHECWORKS model incorrectly assumes that FAC is controlled solely by chemical dissolution, is a linear, non-local phenomenon, and that chromium content is a known variable. See supra ¶¶ 59-62.

207. At Indian Point, Entergy uses CHECWORKS to select and schedule modeled piping components for inspection. See supra ¶¶ 63-65.

208. CHECWORKS plays an integral and primary role in determinations at Indian Point regarding what new modeled components to inspect for FAC-related degradation. See supra ¶¶ 66-70.

209. For over a decade, CHECWORKS has been producing highly non-conservative, inaccurate results, indicating a lack of adequate benchmarking; the evidence in the record demonstrates that the CHECWORKS model at Indian Point cannot be successfully calibrated or benchmarked in the future prior to the start of Entergys proposed PEO. See supra ¶¶ 72-86.

77

Based on the highly erratic behavior of CHECWORKS and the lack of correlation exhibited between CHECWORKS predictions and actual measurements, CHECWORKS is ineffective for objective quantitative assessments, and not a useful or reliable tool for determining inspection locations for modeled components to timely detect and mitigate FAC during the proposed PEO.

See supra ¶¶ 87.

210. CHECWORKS poor predictive accuracy results in inspection scopes that will not determine where wall thinning below minimum ASME safe values will occur, which poses safety consequences. See supra ¶¶ 88-92.

211. CHECWORKS does not have a track record of performance at Indian Point, as exhibited by a history of wall thinning and leakage events. See supra ¶¶93-102. This also renders the CHECWORKS model an inappropriate tool for informing inspection locations at Indian Point. See supra ¶ 103.

212. The use of CHECWORKS at Indian Point runs afoul of the guidance contained in revisions 1 and 2 of the GALL Report, as well as with proposed draft guidance contained in interim NRC Staff guidance, since the non-conservative model does not assure the adequate detection and management of FAC. See supra ¶¶ 104-118.

213. Alternative computer modeling systems exist that produce more accurate results than the CHECWORKS model. See supra ¶¶ 119-126.

214. Entergys FAC program employs several other tools in addition to CHECWORKS in order to determine inspection locations. See supra ¶ 127. These tools have failed to ensure the management of FAC at Indian Point, as exhibited by a history of wall thinning and leakage events at the plant, and are not adequate to ensure that components will be inspected prior to reaching critical wall thicknesses. See supra ¶¶ 128-164.

78

215. Entergys inability to adequately determine what components require inspections in order to avoid reaching critical wall thicknesses may result in significant safety consequences if Indian Point is subject to operational transients, design basis accident transients, earthquake loads, station blackouts, transients without scrams, as well as metal fatigue. See supra ¶¶ 165-172. Entergy has failed to consider how the uncertainty related to pipe wall thickness at Indian Point will affect the integrity of components under such conditions. Id.

216. Entergys FAC AMP is comprised of a series of generic guidance documents and implementing procedures, which contain general platitudes and no specificity as to applicability of the program at Indian Point, and no details about how to actually accomplish the program in an effective manner. See supra ¶¶ 173-188. Entergy asserts these documents are sufficient to demonstrate an adequate FAC AMP. Id. The FAC program at Indian Point relies heavily on trusting that Entergy can and will implement an adequate AMP, with minimal accountability. Id.

217. Because Entergys FAC program cannot properly determine appropriate inspection locations, the program does not currently meet the general AMP guidance for managing FAC. See supra ¶¶ 189-194. As a result, Entergy must provide, but has not, an AMP containing sufficient details to assure that components will be selected for inspection so as to prevent wall thinning below acceptable levels, and otherwise ensure that the aging effects of FAC will be adequately managed throughout the PEO. See supra ¶¶ 194-196. Entergys AMP must address and rectify the numerous ways in which Entergys existing FAC program is inadequate. See supra ¶¶ 189, 190, 196.

79

CONCLUSIONS OF LAW The ASLB has considered all of the evidence presented by the parties on Contention RK-TC-2. Based upon a review of the entire record on Contention RK-TC-2 in this proceeding, and based on the foregoing proposed findings of fact, the ASLB reaches the following conclusions.

218. Contention RK-TC-2 challenges the adequacy of Entergys Aging Management Program (AMP) for flow accelerated corrosion (FAC). Controlling regulations require license renewal applicants to demonstrate that the aging effects will be adequately managed so that the intended function(s) will be maintained consistent with the CLB [current licensing basis]

for the period of extended operation. 10 C.F.R. § 54.21(a)(3); 10 C.F.R. § 54.29. Entergy maintains the burden to make this demonstration. See 10 C.F.R. § 2.325; Amergen Energy Co.

(Oyster Creek Nuclear Generating Station), CLI-09-7, 69 NRC 235, 269 (2009). Thus, the legal issue with respect to this contention is whether Entergy has demonstrated that it has an AMP that will adequately manage the aging effects of fact throughout Entergys proposed PEO.

219. The ALSB is persuaded by the evidence presented by intervenor, Riverkeeper, that the aging effects of FAC will not be adequately managed at Indian Point during the PEO.

220. Entergys FAC program at Indian Point does not comply with the guidance contained in the GALL Report. Findings of Fact ¶¶ 104-118; 173-196. Entergys program relies on the use of a computer code CHECWORKS, as well as other tools, in order to determine the scope of inspections at Indian Point. Findings of Fact ¶¶ 63-70, 127. The evidence in the record shows that CHECWORKS does not produce accurate results so as to meaningfully inform inspection priorities at Indian Point for relevant components. Findings of Fact ¶¶ 59-62, 72-87,93-103. The evidence in the record further shows that Entergy has failed to demonstrate that its other tools accurately predict inspection locations. Findings of Fact ¶¶ 127-164. Lastly, the 80

evidence shows that Entergys FAC program fails to recognize fundamental principles that FAC is a non-linear, local phenomenon, which encompasses chemical dissolution as well as mechanical processes. Findings of Fact ¶¶ 15, 17, 26-38, 54, 60. The fact that Entergys FAC program at Indian Point is premised upon incorrect assumptions that FAC is a limited, linear, and non-local phenomenon severely limits the ability of the program to accurately choose inspection locations. Findings of Fact ¶¶ 17, 38, 60. All told, the evidence in the record overwhelmingly demonstrates that Entergy is not capable of selecting all appropriate components for inspections so as to ensure the adequate detection of wall thinning before thicknesses are reduced below allowable levels. Findings of Fact ¶¶ 88-92, 128-164. An adequate inspection scope is a threshold part of an adequate FAC AMP, and Entergys inability to demonstrate that minimum wall thicknesses of FAC-susceptible components will not be reduced by FAC to below acceptable levels during the PEO is inconsistent with various aspects of the GALL Report, Revisions 1 and 2, including the fundamental premise that a FAC program must ensure that the extent of wall thinning is adequately determined, that intended function will not be lost, and that corrective actions are adequately identified. Exhibit NYS00146C (GALL Report, Revision 1, at pp. XI M-61, XI M-62); Exhibit NYS00147D (GALL Report, Revision 2, at pp. XI M17-1, XI M17-2); Findings of Fact ¶¶ 104-118.

221. In light of Entergys failure to demonstrate the effective use of CHECWORKS, or other independent tools for managing FAC at Indian Point, Entergy must, but has failed, to articulate a thorough description of an AMP which shows conclusively how this program will ensure that the effects of aging will be managed in light of the flaws in Entergys existing FAC program. Entergy Nuclear Vermont Yankee, 68 NRC 763, 870; see id. at 871; Findings of Fact

¶¶ 189-194. Entergy relies upon generic guidance documents to demonstrate an adequate AMP 81

for FAC. Findings of Fact ¶¶ 173-188. However, Entergy cannot generically claim consistency with NRCs guidance documents or promise to follow the same generic program recommendations provided to all plants does not clear the bar required by the regulations.

Entergy Nuclear Vermont Yankee, 68 NRC 763, 870. Moreover, Entergy has failed to refute the evidence that the FAC program at Indian Point is not currently complying with such generic guidance. Findings of Fact ¶¶ 104-118, 189-196.

222. In addition, the ASLB finds persuasive evidence in the record that undetected FAC during the proposed PEO poses a risk of loss of coolant accidents in violation of NRCs General Design Criterion (GDC) 4. Findings of Fact ¶¶ 165-172. This criterion requires that plant structures, systems and components be able to accommodate the effects of . . . loss of coolant accidents and be appropriately protected against dynamic effects . . . that may result from equipment failures and from events and conditions outside the nuclear power unit. 10 C.F.R. Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 4 Environmental and dynamic effects design bases. Entergy has failed to consider how uncertainties in pipe wall thicknesses will affect the integrity of components under varying transient loads, and, as a result, failed to demonstrate that the intended functions of FAC-susceptible components will be maintained. Findings of Fact ¶¶ 165-172. Entergys failure to assess the safety consequences of transient loads on FAC-degraded components fails to demonstrate that such components will be able to accommodate LOCAs and be protected against dynamic effects . . . from events and conditions outside the nuclear plants. 10 C.F.R. Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 4Environmental and dynamic effects design bases; Findings of Fact ¶¶ 165-172.

82

223. Based on the foregoing, and the Findings of Fact stated above, it is the ASLBs determination that Entergy has failed to meet its burden of demonstrating compliance with the guidance in the GALL Report, and that, pursuant to 10 C.F.R. § 54.21(a)(3) and § 54.29, the aging effects of FAC will be adequately managed during the proposed periods of extended operation.

224. Therefore, Contention RK-TC-2 is resolved in favor of intervenor Riverkeeper.

PROPOSED ORDER WHEREFORE, IT IS ORDERED, that Contention RK-TC-2 (Flow Accelerated Corrosion) in the Indian Point license renewal proceeding is resolved in favor of the intervenor, Riverkeeper. Accordingly, the Director of NRCs Office of Nuclear Reactor Regulation is authorized to deny Entergy Nuclear Operations, Inc.s application for a renewed operating license for the Indian Point nuclear power plant.

IT IS FURTHER ORDERED, in accordance with 10 C.F.R. § 2.341(b)(1), that any party to this proceeding may file a petition for review of this Initial Decision with the Commission within twenty-five (25) days after service of this partial initial decision. Filing of a petition for review is mandatory for a party to exhaust its administrative remedies before seeking judicial review. 10 C.F.R. § 2.341(b)(1).

IT IS FURTHER ORDERED, in accordance with 10 C.F.R. § 2.340(g) and § 2.1210, that this Initial Decision shall constitute the final decision of the Commission forty (40) days after its issuance, unless there is a petition for Commission review filed in accordance with 10 C.F.R. §§ 2.1212 and 2.341(b), or the Commission decides to review this Initial Decision under 10 C.F.R.

§2.1210(a)(2) or (3).

It is so ORDERED.

83

Respectfully submitted, Signed (electronically) by Deborah Brancato, Esq.

Phillip Musegaas, Esq.

Riverkeeper, Inc.

20 Secor Road Ossining, NY 10562 (914) 478-4501 dbrancato@riverkeeper.org phillip@riverkeeper.org Dated: March 22, 2013 84

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD

)

In the Matter of ) Docket Nos.

) 50-247-LR Entergy Nuclear Operations, Inc. ) and 50-286-LR (Indian Point Nuclear Generating )

Units 2 and 3) ) March 22, 2013

___________________________________________ )

CERTIFICATE OF SERVICE I certify that on March 22, 2013, copies of Riverkeeper Post-Hearing Proposed Findings of Fact and Conclusions of Law Regarding Contention RK-TC Flow Accelerated Corrosion, were served on the following by NRCs Electronic Information Exchange:

Lawrence G. McDade, Chair Shelbie Lewman Richard E. Wardwell, Administrative Judge Law Clerk Michael F. Kennedy, Administrative Judge Anne Siarnacki Atomic Safety and Licensing Board Panel Law Clerk U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board Panel Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Lawrence.McDade@nrc.gov Washington, D.C. 20555 Richard.Wardwell@nrc.gov shelbie.lewman@nrc.gov Michael.Kennedy@nrc.gov anne.siarnacki@nrc.gov John J. Sipos, Esq. Bobby R. Burchfield, Esq.

Assistant Attorney General Matthew M. Leland, Esq.

Office of the New York Attorney General Clint A. Carpenter, Esq.

for the State of New York McDermott Will & Emery LLC The Capitol 600 13th Street, NW Albany, NY 12224 Washington, DC 20005-3096 E-mail: John.Sipos@oag.state.ny.us bburchfield@mwe.com mleland@mwe.com ccarpenter@mwe.com Richard A. Meserve, Esq. Martin J. ONeill, Esq.

Covington & Burling LLP Morgan, Lewis & Bockius, LLP 1201 Pennsylvania Avenue, NW 1000 Louisiana Street, Suite 4000 Washington, DC 20004-2401 Houston, TX 77002 Phone: (202) 662-6000 E-mail: martin.oneill@morganlewis.com Fax: (202) 662-6291 E-mail: rmeserve@cov.com 1

Janice A. Dean, Esq. William C. Dennis, Esq.

Assistant Attorney General Entergy Nuclear Operations, Inc.

Office of the Attorney General 440 Hamilton Avenue 120 Broadway, 26th Floor White Plains, NY 10601 New York, NY 10271 E-mail: wdennis@entergy.com E-mail: Janice.dean@oag.state.ny.us Office of the Secretary Office of Commission Appellate Adjudication Rulemakings and Adjudications Staff U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 E-mail: OCAAMAIL@nrc.gov E-mail: HEARINGDOCKET@nrc.gov Manna Jo Greene Thomas F. Wood, Esq.

Karla Raimundi Daniel Riesel, Esq.

Stephen C. Filler, Board Member Victoria S. Treanor Hudson River Sloop Clearwater, Inc. Sive, Paget and Riesel, P.C.

724 Wolcott Ave 460 Park Avenue Beacon, New York 12508 New York, NY 10022 E-mail: Mannajo@clearwater.org E-mail: driesel@sprlaw.com karla@clearwater.org vtreanor@sprlaw.com stephenfiller@gmail.com Melissa-Jean Rotini, of counsel Sean Murray, Mayor Assistant County Attorney Village of Buchanan Office of Robert F. Meehan, Westchester Municipal Building County Attorney 236 Tate Avenue 148 Martine Avenue, 6th Floor Buchanan, NY 10511-1298 White Plains, NY 10601 E-mail: vob@bestweb.net, E-mail: MJR1@westchestergov.com SMurray@villageofbuchanan.com, Administrator@villageofbuchanan.com Elise N. Zoli, Esq. Richard Webster, Esq.

Goodwin Procter, LLP Public Justice, P.C.

53 State Street Suite 200 Boston, MA 02109 1825 K Street, NW E-mail: ezoli@goodwinprocter.com Washington, DC 20006 rwebster@publicjustice.net Robert D. Snook, Esq. Michael J. Delaney Assistant Attorney General Department of Environmental Protection 55 Elm Street, P.O. Box 120 59-17 Junction Boulevard Hartford, CT 06141-0120 Flushing NY 11373 E-mail: Robert.Snook@po.state.ct.us E-mail: mdelaney@dep.nyc.gov 2

Sherwin E. Turk Kathryn M. Sutton, Esq.

Beth N. Mizuno Paul M. Bessette, Esq.

Brian G. Harris Jonathan M. Rund, Esq.

David E. Roth Raphael Kuyler, Esq.

Joseph A. Lindell Morgan, Lewis & Bockius, LLP Anita Ghosh, Esq. 1111 Pennsylvania Ave. N.W.

Office of General Counsel Washington, D.C. 20004 Mail Stop: 0-15D21 E-mail:

U.S. Nuclear Regulatory Commission pbessette@morganlewis.com Washington, D.C. 20555-0001 ksutton@morganlewis.com E-mail: jrund@morganlewis.com Sherwin.Turk@nrc.gov rkuyler@morganlewis.com Beth.Mizuno@nrc.gov brian.harris@nrc.gov David.Roth@nrc.gov Joseph.Lindell@nrc.gov anita.ghosh@nrc.gov Edward F. McTiernan Deputy Counsel Office of General Counsel New York State Department of Environmental Conservation 625 Broadway, 14th floor Albany, NY 12233-150024 (518) 402-9185 phone (518) 402-9018 fax efmctier@gw.dec.state.ny.us Signed (electronically) by Deborah Brancato Deborah Brancato March 22, 2013 3